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Sample records for accident sequence precursor

  1. Introduction to accident sequence precursor methodology

    The report consists of the following sections: (1) Introduction to the issue (Use of accident sequence precursor (ASP) programs; Definition of ASP, indicators, information; Status of ASP evaluation methodology); (2) Selection of potential precursors; (3) Detailed analysis of selected potential precursors (Methodology; ASP analysis software; Indicators and presentation of results); Potential for ASP evaluation at Czech nuclear power plants (Procedure for Conditional Core Damage Probability and Event Importance assessment by the NRC method; Experience in ASP in the Czech Republic; Application of the ASP evaluation method to the Dukovany NPP); Plan of Activities; and Conclusions. (P.A.)

  2. Accident sequence precursor events with age-related contributors

    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 (∼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

  3. Advanced accident sequence precursor analysis level 2 models

    Galyean, W.J.; Brownson, D.A.; Rempe, J.L. [and others

    1996-03-01

    The U.S. Nuclear Regulatory Commission Accident Sequence Precursor program pursues the ultimate objective of performing risk significant evaluations on operational events (precursors) occurring in commercial nuclear power plants. To achieve this objective, the Office of Nuclear Regulatory Research is supporting the development of simple probabilistic risk assessment models for all commercial nuclear power plants (NPP) in the U.S. Presently, only simple Level 1 plant models have been developed which estimate core damage frequencies. In order to provide a true risk perspective, the consequences associated with postulated core damage accidents also need to be considered. With the objective of performing risk evaluations in an integrated and consistent manner, a linked event tree approach which propagates the front end results to back end was developed. This approach utilizes simple plant models that analyze the response of the NPP containment structure in the context of a core damage accident, estimate the magnitude and timing of a radioactive release to the environment, and calculate the consequences for a given release. Detailed models and results from previous studies, such as the NUREG-1150 study, are used to quantify these simple models. These simple models are then linked to the existing Level 1 models, and are evaluated using the SAPHIRE code. To demonstrate the approach, prototypic models have been developed for a boiling water reactor, Peach Bottom, and a pressurized water reactor, Zion.

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

    Lui, C.H. [Nuclear Regulatory Commission, Washington, DC (United States); Galyean, W.J.; Brownson, D.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States)] [and others

    1997-02-01

    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.

  5. Research activities on accident sequence precursor study at JAERI

    At JAERI, as part of the research activity on operating experience analysis, we started the Accident Sequence Precursor (ASP) study in 1996. Since then, we have been carrying out the ASP analyses of actual events that have the potential of risk significance, the review of the ASP documents published by the USNRC and development of the event tree models for providing a more realistic ASP analyses. The objectives of the study are to obtain the risk significant trends, to characterize risk insights useful for identifying plant vulnerabilities, to feed the lessons learned from the study back to plant operations, and to establish risk indicators for event assessment. Although the ASP analysis could provide useful information on not only the safety significance of individual events but also management of plant operation and risk-informed regulation, its usefulness has not been fully recognized yet in Japan. In order to demonstrate the usefulness and effectiveness of the ASP analysis, we have been carrying out the following activities. 1- In-depth analyses of specific precursor events. The review of the operating experience, including the ASP documents, revealed that some specific events such as steam generator tube rupture (SGTR) might have risk significance and as well, some specific anomalies observed during the events might have contributed to the plant risk. In order to identify the risk significant anomalies observed during the events and to obtain generic insights useful for enhancing the safety of nuclear power plants, we have been carrying out in-depth analyses of specific precursor events. To date, the ASP analyses were performed for ten actual and one potential SGTR events with use of the consistent ASP models. 2- Sophistication of event tree models for ASP analysis. A comparison of the event trees used in the USNRC's ASP Program and those constructed in the PSA studies identified differences in the accident sequences defined. As well, some anomalies observed

  6. Accident sequence precursor analysis of Daya Bay and Ling'ao nuclear power plants

    The accident sequence precursor analysis method was adopted to evaluate the events of Daya Bay and Ling'ao Nuclear Power Plants, and some risk significant events which are called the accident sequence precursor were identified. By the statistics, classification and trend analysis, some useful insight can be obtained to support the nuclear safety management for the nuclear power plant. (authors)

  7. Advanced accident sequence precursor analysis level 1 models

    Sattison, M.B.; Thatcher, T.A.; Knudsen, J.K.; Schroeder, J.A.; Siu, N.O. [Idaho National Engineering Lab., Idaho National Lab., Idaho Falls, ID (United States)

    1996-03-01

    INEL has been involved in the development of plant-specific Accident Sequence Precursor (ASP) models for the past two years. These models were developed for use with the SAPHIRE suite of PRA computer codes. They contained event tree/linked fault tree Level 1 risk models for the following initiating events: general transient, loss-of-offsite-power, steam generator tube rupture, small loss-of-coolant-accident, and anticipated transient without scram. Early in 1995 the ASP models were revised based on review comments from the NRC and an independent peer review. These models were released as Revision 1. The Office of Nuclear Regulatory Research has sponsored several projects at the INEL this fiscal year to further enhance the capabilities of the ASP models. Revision 2 models incorporates more detailed plant information into the models concerning plant response to station blackout conditions, information on battery life, and other unique features gleaned from an Office of Nuclear Reactor Regulation quick review of the Individual Plant Examination submittals. These models are currently being delivered to the NRC as they are completed. A related project is a feasibility study and model development of low power/shutdown (LP/SD) and external event extensions to the ASP models. This project will establish criteria for selection of LP/SD and external initiator operational events for analysis within the ASP program. Prototype models for each pertinent initiating event (loss of shutdown cooling, loss of inventory control, fire, flood, seismic, etc.) will be developed. A third project concerns development of enhancements to SAPHIRE. In relation to the ASP program, a new SAPHIRE module, GEM, was developed as a specific user interface for performing ASP evaluations. This module greatly simplifies the analysis process for determining the conditional core damage probability for a given combination of initiating events and equipment failures or degradations.

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

    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

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

    Martz, H.F.

    1998-12-01

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

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

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

    2014-01-01

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

  11. Low-power and shutdown models for the accident sequence precursor (ASP) program

    Sattison, M.B.; Thatcher, T.A.; Knudsen, J.K. [Idaho National Engineering Lab., Idaho Falls, ID (United States)] [and others

    1997-02-01

    The US Nuclear Regulatory Commission (NRC) has been using full-power. Level 1, limited-scope risk models for the Accident Sequence Precursor (ASP) program for over fifteen years. These models have evolved and matured over the years, as have probabilistic risk assessment (PRA) and computer technologies. Significant upgrading activities have been undertaken over the past three years, with involvement from the Offices of Nuclear Reactor Regulation (NRR), Analysis and Evaluation of Operational Data (AEOD), and Nuclear Regulatory Research (RES), and several national laboratories. Part of these activities was an RES-sponsored feasibility study investigating the ability to extend the ASP models to include contributors to core damage from events initiated with the reactor at low power or shutdown (LP/SD), both internal events and external events. This paper presents only the LP/SD internal event modeling efforts.

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

    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

  13. NASA's Accident Precursor Analysis Process and the International Space Station

    Groen, Frank; Lutomski, Michael

    2010-01-01

    This viewgraph presentation reviews the implementation of Accident Precursor Analysis (APA), as well as the evaluation of In-Flight Investigations (IFI) and Problem Reporting and Corrective Action (PRACA) data for the identification of unrecognized accident potentials on the International Space Station.

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

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

  15. Accident Precursor Analysis and Management: Reducing Technological Risk Through Diligence

    Phimister, James R. (Editor); Bier, Vicki M. (Editor); Kunreuther, Howard C. (Editor)

    2004-01-01

    Almost every year there is at least one technological disaster that highlights the challenge of managing technological risk. On February 1, 2003, the space shuttle Columbia and her crew were lost during reentry into the atmosphere. In the summer of 2003, there was a blackout that left millions of people in the northeast United States without electricity. Forensic analyses, congressional hearings, investigations by scientific boards and panels, and journalistic and academic research have yielded a wealth of information about the events that led up to each disaster, and questions have arisen. Why were the events that led to the accident not recognized as harbingers? Why were risk-reducing steps not taken? This line of questioning is based on the assumption that signals before an accident can and should be recognized. To examine the validity of this assumption, the National Academy of Engineering (NAE) undertook the Accident Precursors Project in February 2003. The project was overseen by a committee of experts from the safety and risk-sciences communities. Rather than examining a single accident or incident, the committee decided to investigate how different organizations anticipate and assess the likelihood of accidents from accident precursors. The project culminated in a workshop held in Washington, D.C., in July 2003. This report includes the papers presented at the workshop, as well as findings and recommendations based on the workshop results and committee discussions. The papers describe precursor strategies in aviation, the chemical industry, health care, nuclear power and security operations. In addition to current practices, they also address some areas for future research.

  16. Probabilistic accident sequence recovery analysis

    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)

  17. NASA Accident Precursor Analysis Handbook, Version 1.0

    Groen, Frank; Everett, Chris; Hall, Anthony; Insley, Scott

    2011-01-01

    accident precursors by evaluating anomaly occurrences for their system safety implications and, through both analytical and deliberative methods used to project to other circumstances, identifying those that portend more serious consequences to come if effective corrective action is not taken. APA builds upon existing safety analysis processes currently in practice within NASA, leveraging their results to provide an improved understanding of overall system risk. As such, APA represents an important dimension of safety evaluation; as operational experience is acquired, precursor information is generated such that it can be fed back into system safety analyses to risk-inform safety improvements. Importantly, APA utilizes anomaly data to predict risk whereas standard reliability and PRA approaches utilize failure data which often is limited and rare.

  18. An Accident Precursor Analysis Process Tailored for NASA Space Systems

    Groen, Frank; Stamatelatos, Michael; Dezfuli, Homayoon; Maggio, Gaspare

    2010-01-01

    Accident Precursor Analysis (APA) serves as the bridge between existing risk modeling activities, which are often based on historical or generic failure statistics, and system anomalies, which provide crucial information about the failure mechanisms that are actually operative in the system and which may differ in frequency or type from those in the various models. These discrepancies between the models (perceived risk) and the system (actual risk) provide the leading indication of an underappreciated risk. This paper presents an APA process developed specifically for NASA Earth-to-Orbit space systems. The purpose of the process is to identify and characterize potential sources of system risk as evidenced by anomalous events which, although not necessarily presenting an immediate safety impact, may indicate that an unknown or insufficiently understood risk-significant condition exists in the system. Such anomalous events are considered accident precursors because they signal the potential for severe consequences that may occur in the future, due to causes that are discernible from their occurrence today. Their early identification allows them to be integrated into the overall system risk model used to intbrm decisions relating to safety.

  19. Precursors to potential severe core damage accidents: 1992, a status report; Volume 18: Appendices B, C, D, E, F, and G

    NONE

    1993-12-01

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

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

    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

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

    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

  2. Precursors to potential severe core damage accidents: 1997 - A status report. Volume 26

    This report describes the five operational events in 1997 that affected five commercial light-water reactors (LWRs) and that are considered to be precursors to potential severe core damage accidents. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10-6. These events were identified by first computer-screening the 1997 licensee event reports from commercial LWRs to identify those events that could be precursors. Candidate precursors were selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters to ensure that the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969--1996 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for the events

  3. Precursors to potential severe core damage accidents: 1996. A status report. Volume 25

    This report describes the 14 operational events in 1996 that affected 13 commercial light-water reactors and that are considered to be precursors to potential severe core damage accidents. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10-6. These events were identified by first computer-screening the 1996 licensee event reports from commercial light-water reactors to identify those events that could potentially be precursors. Candidate precursors were selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters and regional offices to ensure the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969--1995 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for the events

  4. Precursors to potential severe core damage accidents: 1997 -- A status report. Volume 26

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.; Muhlheim, M.D. [Oak Ridge National Lab., TN (United States); Dolan, B.W.; Minarick, J.W. [Science Applications International Corp., Oak Ridge, TN (United States)

    1998-11-01

    This report describes the five operational events in 1997 that affected five commercial light-water reactors (LWRs) and that are considered to be precursors to potential severe core damage accidents. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 {times} 10{sup {minus}6}. These events were identified by first computer-screening the 1997 licensee event reports from commercial LWRs to identify those events that could be precursors. Candidate precursors were selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters to ensure that the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969--1996 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for the events.

  5. Accident sequence quantification of phased mission systems using Markov approach

    A Markov approach to incorporate a phased mission analysis into an accident sequence quantification is developed in this study. The Markov approach describes more accurately the dynamic characteristics of phased mission systems, dependency among various phases, and repair of failed components than a cut set approach. For comparison, an accurate cut set approach to quantify the accident sequences is also presented in this study. The results show that it is desirable to use the Markov approach in phased missions systems with large failure rates of components or long mission time intervals. Furthermore, the Markov approach is not constrained by the magnitudes of failure rates or mission time intervals. (author)

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

    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

  7. PSA modeling of long-term accident sequences

    In the traditional Level 1 PSA, the long term of the accident sequences is usually taken into account in a simplified manner. For example, some of the mitigations which are needed at long term are taken into account in the PSA, but the analysis and the associated failures probabilities quantification are estimated based on generic assessments. In the context of the extension of PSA scope to include the external hazards, in France, both operator (EDF) and IRSN work for the improvement of methods to better take into account in the PSA the long term of accident sequences induced by initiators which affect the whole site containing several nuclear installations (reactors, fuel pools, ...). This is an essential prerequisite for the development of external hazards PSA. It has to be noted that in the French PSA, even before Fukushima, this type of accident sequences was already taken into account, many insight being used, as complementary information, to enhance the safety level of the plants. The recent French and international operating experience is an opportunity for tuning the actual PSA methods for long term accident sequences modeling. The paper presents the main results of the ongoing efforts in this area. (author)

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

    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

  9. Accident Sequence Evaluation Program: Human reliability analysis procedure

    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.

  10. Accident Sequence Evaluation Program: Human reliability analysis procedure

    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

  11. Defining accident sequences for conducting probabilistic risk assessments

    A major portion of any PRA activity is devoted to the identification of potential accident sequences which could lead to radiological releases. A systematic application of acceptable methods and techniques is desirable. Event tree and fault tree analysis form the basis for plant/system modeling. Consensus approaches for acquisition of information, model development, assumptions, constraints, inclusion of common cause events, interrelationships of event trees and fault trees, accuracy, completeness and manner of documentation are presented. Alternative methods which may be used or directed toward a specialized purpose are also of interest. Particular emphasis is placed on defining in practical terms the specific tasks of accident sequence definition and what is required, why it is necessary and how each task relates to the overall risk assessment process

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

    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

  13. Human reliability analysis for accident sequences in NPP

    The purpose of this paper is to perform a human performance analysis in accident conditions for the operating NPP. This analysis is realized using Human Reliability Analysis (HRA) methods. HRA methods have necessary tools to analyze the human actions, to estimate the human error probabilities and to identify the major factors which could have a negative influence on the mitigating of the consequences of the abnormal events in NPP. The analyzed events are from CANDU 600 NPP. In order to achieve the analysis of these events the THEP and SPAR-H methods were used. After analyzing the results the actuated equipment, the negative influence factors on the human performance and the dependence levels between the human actions and between the human actions and diagnosis were established. In addition, some recommendations were formulated which could influence positive the human performance on the mitigating of the consequences of the accident sequences in NPP. (authors)

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

    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)

  15. BNL severe accident sequence experiments and analysis program

    Analyses of LWR degraded core accidents require mathematical characterization of two major sources of pressure and temperature loading on the reactor containment buildings: (1) steam generation from core debris-water thermal interactions and (2) molten core-concrete interactions. Experiments are in progress at BNL in support of analytical model development related to aspects of the above containment loading mechanisms. The work supports development and evaluation of the CORCON, MARCH, CONTAIN and MEDICI computer under development at other NRC-contractor laboratories. The thermal-hydraulic behavior of hot debris located within the reactor core region upon sudden introduction of cooling water is being investigated in a joint experimental and analytical program. This work supports development and evaluation of the SCDAP computer code being developed at EG and G to characterize in-vessel severe core damage accident sequences. Progress is described in the two areas of: 1) core debris thermal-hydraulic phenomenology and 2) heat transfer in core-concrete interactions

  16. Accident sequence modeling: human actions, system response, intelligent decision support

    In Probabilistic Safety Assessment (PSA) of large technological systems, accident sequence modeling represents the synthesis of expert judgement, system modeling, and operational evidence. This book contains the papers that were presented at a two-day Seminar that was held in Munich in August 1987. The aim of this Seminar was to provide a forum for in-depth discussion in a workshop atmosphere of the key elements in the modeling process, such as operator actions and system response, and to assess the possibilities of using such models to design operator decision support systems in the form of expert systems or interactive man computer structures. While this evaluation of the state of the art was done in the context of nuclear power reactor safety, most of the models and ideas advanced by the participants have wide applicability and can be used in safety assessments and reliability enhancement programs for other fields, for example the chemical process and aerospace industries. (author)

  17. Development of NASA's Accident Precursor Analysis Process Through Application on the Space Shuttle Orbiter

    Maggio, Gaspare; Groen, Frank; Hamlin, Teri; Youngblood, Robert

    2010-01-01

    Accident Precursor Analysis (APA) serves as the bridge between existing risk modeling activities, which are often based on historical or generic failure statistics, and system anomalies, which provide crucial information about the failure mechanisms that are actually operative in the system. APA docs more than simply track experience: it systematically evaluates experience, looking for under-appreciated risks that may warrant changes to design or operational practice. This paper presents the pilot application of the NASA APA process to Space Shuttle Orbiter systems. In this effort, the working sessions conducted at Johnson Space Center (JSC) piloted the APA process developed by Information Systems Laboratories (ISL) over the last two years under the auspices of NASA's Office of Safety & Mission Assurance, with the assistance of the Safety & Mission Assurance (S&MA) Shuttle & Exploration Analysis Branch. This process is built around facilitated working sessions involving diverse system experts. One important aspect of this particular APA process is its focus on understanding the physical mechanism responsible for an operational anomaly, followed by evaluation of the risk significance of the observed anomaly as well as consideration of generalizations of the underlying mechanism to other contexts. Model completeness will probably always be an issue, but this process tries to leverage operating experience to the extent possible in order to address completeness issues before a catastrophe occurs.

  18. An analysis of station blackout sequences for the severe accident analysis database (II)

    Park, Soo Yong; Kim, Dong Ha

    2006-08-15

    This report contains analysis methodologies and calculation results of station blackout sequences for the severe accident analysis database system. The Korean standard nuclear power plant has been selected as a reference plant. Based on the probabilistic safety analysis of the corresponding plant. Eight accident scenarios, which was predicted to have more than 10{sup -10}/ry occurrence frequency have been analyzed as base cases for the station blackout sequence database. Furthermore, the sensitivity studies for operational plant systems and for phenomenological models of the analysis computer code have been performed. The functions of the severe accident analysis database system will be to make a diagnosis of the accident by some input information from the plant symptoms, to search a corresponding scenario, and finally to provide the user phenomenological information based on the pre-analyzed results. The MAAP 4.06 calculation results of station blackout sequence in this report will be utilized as input data of the severe accident analysis database system.

  19. An analysis of station blackout sequences for the severe accident analysis database (II)

    This report contains analysis methodologies and calculation results of station blackout sequences for the severe accident analysis database system. The Korean standard nuclear power plant has been selected as a reference plant. Based on the probabilistic safety analysis of the corresponding plant. Eight accident scenarios, which was predicted to have more than 10-10/ry occurrence frequency have been analyzed as base cases for the station blackout sequence database. Furthermore, the sensitivity studies for operational plant systems and for phenomenological models of the analysis computer code have been performed. The functions of the severe accident analysis database system will be to make a diagnosis of the accident by some input information from the plant symptoms, to search a corresponding scenario, and finally to provide the user phenomenological information based on the pre-analyzed results. The MAAP 4.06 calculation results of station blackout sequence in this report will be utilized as input data of the severe accident analysis database system

  20. Survey of the use of rapid sequence induction in the accident and emergency department

    Walker, A.; Brenchley, J

    2000-01-01

    Objectives—To determine the current position regarding the use of rapid sequence induction (RSI) by accident and emergency (A&E) medical staff and the attitudes of consultants in A&E and anaesthetics towards this.

  1. Precursors to potential severe core damage accidents: 1994, a status report. Volume 22: Appendix I

    Nine operational events that affected eleven commercial light-water reactors (LWRs) during 1994 and that are considered to be precursors to potential severe core damage are described. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10-6. These events were identified by computer-screening the 1994 licensee event reports from commercial LWRs to identify those that could be potential precursors. Candidate precursors were then selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters and regional offices to ensure that the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969--1981 and 1984--1993 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for events. This document is bound in two volumes: Vol. 21 contains the main report and Appendices A--H; Vol. 22 contains Appendix 1

  2. Precursors to potential severe core damage accidents: 1994, a status report. Volume 22: Appendix I

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.; Vanden Heuvel, L.N. [Oak Ridge National Lab., TN (United States); Dolan, B.W.; Minarick, J.W. [Oak Ridge National Lab., TN (United States)]|[Science Applications International Corp., Oak Ridge, TN (United States)

    1995-12-01

    Nine operational events that affected eleven commercial light-water reactors (LWRs) during 1994 and that are considered to be precursors to potential severe core damage are described. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 {times} 10{sup {minus}6}. These events were identified by computer-screening the 1994 licensee event reports from commercial LWRs to identify those that could be potential precursors. Candidate precursors were then selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters and regional offices to ensure that the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969--1981 and 1984--1993 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for events. This document is bound in two volumes: Vol. 21 contains the main report and Appendices A--H; Vol. 22 contains Appendix 1.

  3. Precursors to potential severe core damage accidents: 1995 A status report

    Ten operational events that affected 10 commercial light-water reactors during 1995 and that are considered to be precursors to potential severe core damage are described. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10-6. These events were identified by first computer-screening the 1995 licensee event reports from commercial light-water reactors to identify those events that could potentially be precursors. Candidate precursors were selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters and regional offices to ensure the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969-1981 and 1984-1994 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for the events

  4. Study on event sequence of loss-of-coolant accident of primary systems for MNPP based on ESD

    According to the characteristics of accident analysis of nuclear power plant, the combination method of event sequence diagram (ESD) and operational safety analysis is used to build the ESD model for loss-of-coolant accident of marine reactor primary system from the operational safety view. The accident evolution and results are studied and the entire event sequences are required. (authors)

  5. Precursors to potential severe core damage accidents: 1992, A status report

    Twenty-seven operational events with conditional probabilities of subsequent severe core damage of 1.0 x 10E-06 or higher occurring at commercial light-water reactors during 1992 are considered to be precursors to potential core damage. These are described along with associated significance estimates, categorization, and subsequent analyses. The report discusses (1) the general rationale for this study, (2) the selection and documentation of events as precursors, (3) the estimation and use of conditional probabilities of subsequent severe core damage to rank precursor events, and (4) the plant models used in the analysis process

  6. Time-dependent accident sequences including human actions

    During an accident, transitions between plant states can occur due to operator intervention and the failure of systems while running. The latter cause of transition is much less likely than the first, which includes errors of commission and omission as well as recovery of lost functions. A methodology has been developed to model these transitions in the time domain. As an example, it is applied to the analysis of Three-Mile-Island-type accidents. Statistical evidence is collected and used in assessing the frequency of stuck-open power-operated relief valves at Babcock and Wilcox plants as well as the frequency of misdiagnosis. Statistical data are also used in modeling the timing of operator actions during the accident, i.e., turning off and on the high-pressure injection system and closing the block valves

  7. Precursors to potential severe core damage accidents. A status report, 1982--1983

    Forester, J.A.; Mitchell, D.B.; Whitehead, D.W. [and others

    1997-04-01

    This study is a continuation of earlier work that evaluated 1969-1981 and 1984-1994 events affecting commercial light-water reactors. One-hundred nine operational events that affected 51 reactors during 1982 and 1983 and that are considered to be precursors to potential severe core damage are described. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10{sup {minus}6}. These events were identified by first computer screening the 1982-83 licensee event reports from commercial light-water reactors to select events that could be precursors to core damage. Candidates underwent engineering evaluation that identified, analyzed, and documented the precursors. This report discusses the general rationale for the study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for the events.

  8. Precursors to potential severe core damage accidents. A status report, 1982--1983

    This study is a continuation of earlier work that evaluated 1969-1981 and 1984-1994 events affecting commercial light-water reactors. One-hundred nine operational events that affected 51 reactors during 1982 and 1983 and that are considered to be precursors to potential severe core damage are described. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10-6. These events were identified by first computer screening the 1982-83 licensee event reports from commercial light-water reactors to select events that could be precursors to core damage. Candidates underwent engineering evaluation that identified, analyzed, and documented the precursors. This report discusses the general rationale for the study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for the events

  9. Cloning and sequence analysis of cDNA for the canine neurotensin/neuromedin N precursor.

    Dobner, P R; Barber, D L; Villa-Komaroff, L; McKiernan, C

    1987-01-01

    Cloned cDNAs encoding neurotensin were isolated from a cDNA library derived from primary cultures of canine enteric mucosa cells. Nucleotide sequence analysis has revealed the primary structure of a 170-amino acid precursor protein that encodes both neurotensin and the neurotensin-like peptide neuromedin N. The peptide-coding domains are located in tandem near the carboxyl terminus of the precursor and are bounded and separated by the paired, basic amino acid residues Lys-Arg. An additional c...

  10. Carboxyl-terminal sequences influence the import of mitochondrial protein precursors in vivo

    The large subunit of carbamoyl phosphate synthase A from Neurospora crassa is encoded by a nuclear gene but is localized in the mitochondrial matrix. The authors have utilized N. crassa strains that produce both normal and carboxyl-terminal-truncated forms of carbamoyl phosphate synthase A to ask whether the carboxyl terminus affects import of the carbamoyl phosphate synthase A precursor. They found that carboxyl-terminal-truncated precursors were directed to mitochondria but that they were imported less efficiently than full-length proteins that were synthesized in the same cytoplasm. The results suggest that effective import of proteins into mitochondria requires appropriate combinations of targeting sequences and three-dimensional structure

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

    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)

  12. Japan: Accident Sequence Study for Seismic Event at the Multi-Unit Site

    One or more units of a multi-unit nuclear power plant (NPP) could fail simultaneously at seismic event depending on the seismic ground motion and its influence to units of the site. The approach proposed here is to analyze the multi-unit accident sequences with core damage frequency (CDF) at seismic event explicitly by best applying and improving the existing technology and providing interfacing capability of the Level 2 and 3 parts of PSA. The identified accident sequence at each unit, the end state of which could be failure (damaged core) as well as success (intact core), is linked mutually and a set of these sequences are conditionally quantified. If all potential accident sequences would be conditioned mutually, quite a number of accident sequences have to be analyzed. To circumvent such an unnecessary and quite resource-intensive burden, a screening process is effective and important for this approach. Therefore two-stage screening method was developed for the approach proposed here. The first and second stage screenings were applied to initiating events and accident sequences respectively. In these screening processes the correlation analysis of seismic-induced component failures was necessary and important, and was performed in use of representative response and capacity correlation factors, in application on which the floor response spectra were used for structures and components. The correlation analysis results were applied to quantify the concurrent seismic-induced failure probability. The above-mentioned approach was applied to an example of twin-unit BWR5 site for verification purpose. The preliminary results revealed that the evaluation of correlation factors or concurrent seismic-induced failure probabilities affected significantly on the dominant accident sequences and therefore it is important to improve the development of the correlation factors as long as detailed consequence evaluation at a multi-unit site would be required in future application

  13. Sequence of an intestinal cDNA encoding human gastric inhibitory polypeptide precursor

    Gastric inhibitory polypeptide (GIP) is a 42-amino acid hormone that stimulates insulin secretion in the presence of glucose. Complementary DNA clones encoding human GIP were isolated from a library prepared with RNA from duodenum. The predicted amino acid sequence indicates that GIP is derived by proteolytic processing of a 153-residue precursor, preproGIP. The GIP moiety is flanked by polypeptide segments of 51 and 60 amino acids at its NH2 and COOH termini, respectively. The former includes a signal peptide of about 21 residues and an NH2-terminal propeptide of 30 amino acids. GIP is released from the precursor by processing at single arginine residues. There is a region of nine amino acids in the COOH-terminal propeptide of the GIP precursor that has partial homology with a portion of chromogranin A as well as pancreastatin

  14. Vasotocin and isotocin precursors from the white sucker, Catostomus commersoni: cloning and sequence analysis of the cDNAs.

    Heierhorst, J; Morley, S D; Figueroa, J.; Krentler, C; Lederis, K; D. Richter

    1989-01-01

    The nucleotide sequences of cloned cDNAs encoding the precursors for vasotocin and isotocin have been elucidated by analyzing a lambda gt11 library constructed from poly(A)+ RNA from the hypothalamic region of the teleost fish Catostomus commersoni. Screening of the library was carried out with synthetic oligonucleotide probes deduced from the amino acid sequences of the nonapeptides vasotocin and isotocin. The cDNA nucleotide sequences predict isotocin and vasotocin prohormone precursors eac...

  15. Fire as a primary event of accident domino sequences: The case of BLEVE

    The domino sequences found in major accidents have been analyzed for a large set of cases (330); the first event triggering the domino effect was an explosion or a fire, both with approximately the same contribution; the same proportion has been found when all domino effect steps were considered. Although fire effects usually reach a distance much shorter than those of an explosion, as fire is the most frequent major accident it is often found as the first step of domino sequences. This is especially true in the case of BLEVEs. Both in fixed plants and in the transportation of hazardous materials, in the event of a fire, if flames affect a vessel and the fireproofing layer has been damaged, a BLEVE can occur at any moment. 127 BLEVE accidents involving domino effect have been analyzed. It has been found that fire is significantly more frequent than explosion, both in triggering the domino effect sequence and in intermediate steps. The time to failure can range from 1 min or even less up to several hours, an aspect that should be very important for the management of the emergency. A set of conclusions are inferred from this survey. - Highlights: • Fire is the prevailing first event in BLEVE domino accidents. • With flames engulfment or impingement on a vessel, a BLEVE can occur at any moment. • In BLEVE accidents, TTF can vary from less than 1 min to more than 1 h. • Fireproofing and PSV can increase TTF but do not guarantee avoiding the BLEVE

  16. BNL severe accident sequence experiments and analysis program

    A major source of containment pressurization during severe accidents is the transfer of stored energy from the hot core material to available cooling water. One mode of thermal interaction involves the quench of superheated beds of debris which could be present in the reactor cavity following melt-through or failure of the reactor vessel. This work supports development of models of superheated bed quench phenomena which are to be incorporated into containment analysis computer codes such as MARCH, CONTAIN, and MEDICI. A program directed towards characterization of the behavior of superheated debris beds has been completed. This work addressed the quench of superheated debris which is postulated to exist in the reactor cavity of a PWR following melt ejection from the primary system. The debris is assumed to be cooled by a pool of water overlying the bed of hot debris. This work has led to the development of models to predict rate of steam generation during the quench process and, in addition, the ability to assess the coolability of the debris during the transient quench process. A final report on this work has been completed. This report presents a brief description of some relevant results and conclusions. 15 refs

  17. CFD analysis of air ingress distribution during mid-loop accident sequences

    The accident management approach affects nuclear technology and safety with a new formulation of basic hypotheses for the evaluation of the Source Term and radiological impact on the population due to Fission Product release following Severe Accidents. Considering also the wide spectrum of hypothetical and low probability accident scenarios having these kind of consequences, the sequences having potential for air ingress into the reactor coolant system or involving the interaction between fuel and air, which can flow into the reactor coolant system from the containment, have recently gained more and more interest. The research activities summarised in this paper have been carried out at the Department of Mechanical, Nuclear and Production Engineering of Pisa University, in the frame of an international Project of the IV European Community Framework Programme. The activity included a review of the spectrum of accident sequences to be considered for the investigation of the air ingress probability, the behaviour and the effects of air ingress into the reactor core. Two classes of scenarios were identified for a more in-depth analysis: (a) mid-loop sequences, and (b) scenarios including vessel melt-through. In this frame, mid-loop sequences, having more probabilistic interest than vessel melt-through scenarios, have been investigated by using 3D analytical tools (i.e. Fluent V5.0 fluid-dynamic code). (author)

  18. Reconstructing SALMFamide neuropeptide precursor evolution in the phylum Echinodermata: ophiuroid and crinoid sequence data provide new insights

    Elphick, Maurice R; Semmens, Dean C.; Blowes, Liisa M.; Judith eLevine; Lowe, Christopher J.; Maria Ina Arnone; Clark, Melody S.

    2015-01-01

    The SALMFamides are a family of neuropeptides that act as muscle relaxants in echinoderms. Analysis of genome/transcriptome sequence data from the sea urchin Strongylocentrotus purpuratus (Echinoidea), the sea cucumber Apostichopus japonicus (Holothuroidea) and the starfish Patiria miniata (Asteroidea) reveals that in each species there are two types of SALMFamide precursor: an L-type precursor comprising peptides with a C-terminal LxFamide-type motif and an F-type precursor solely or largely...

  19. Reconstructing SALMFamide Neuropeptide Precursor Evolution in the Phylum Echinodermata: Ophiuroid and Crinoid Sequence Data Provide New Insights

    Elphick, Maurice R; Semmens, Dean C.; Blowes, Liisa M.; Levine, Judith; Lowe, Christopher J.; Arnone, Maria I.; Clark, Melody S.

    2015-01-01

    The SALMFamides are a family of neuropeptides that act as muscle relaxants in echinoderms. Analysis of genome/transcriptome sequence data from the sea urchin Strongylocentrotus purpuratus (Echinoidea), the sea cucumber Apostichopus japonicus (Holothuroidea), and the starfish Patiria miniata (Asteroidea) reveals that in each species there are two types of SALMFamide precursor: an L-type precursor comprising peptides with a C-terminal LxFamide-type motif and an F-type precursor solely or largel...

  20. Analysis of causes and sequences of the accident on Fukushima NPP as a factor of sever accidents prevention in the vessel reactor

    In this monograph, the provisional analysis of the causes and sequences of the sever accidents on the Fukushima NPP is presented. The analysis of the possibility of the origin of extreme events connected with the flooding of Zaporizhzhia NPP industrial site, emergency of the steam-gas explosions on NPPs with WWER and other phenomena occurred under sever accidents was carried out. It was presented the authors original working-out on symptom-oriented approaches of sever accident initiating event list identification, on criteria substantiation of explosion safety and optimization of processes management at sever accidents, as well as on the methodological support of the accident beyond the design basis management at the WWER for prevention of their transition in the stage of sever accidents.

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

    The Level 1 results of Laguna Verde Nuclear Power Plant PRA are presented in the Internal 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)

  2. Cloning and sequence analysis of cDNA for the canine neurotensin/neuromedin N precursor

    Dobner, P.R.; Barber, D.L.; Villa-Komaroff, L.; McKiernan, C.

    1987-05-01

    Cloned cDNAs encoding neurotensin were isolated from a cDNA library derived from primary cultures of canine enteric mucosa cells. Nucleotide sequence analysis using /sup 32/P-labeled nucleotides, has revealed the primary structure of a 170-amino acid precursor protein that encodes both neurotensin and the neurotensin-like peptide neuromedin N. The peptide-coding domains are located in tandem near the carboxyl terminus of the precursor and are bounded and separated by the paired, basic amino acid residues Lys-Arg. An additional coding domain, resembling neuromedin N, occurs immediately after an Arg-Arg basic amino acid pair located in the central region of the precursor. Additional amino acid homologies suggest that tandem duplications have contributed to the structure of the gene. RNA blot analysis, using the cloned cDNA probe, has revealed several mRNA species ranging in size from 500 to 980 nucleotides in the canine enteric mucosa. In contrast a single RNA species of 1500 nucleotides was detected in bovine hypothalamus poly-(A)/sup +/ RNA. The ability of the canine probe to cross-hybridize with bovine mRNA suggest that this probe can be used to isolate neurotensin/neuromedin N genes from other mammalian species.

  3. Cloning and sequence analysis of cDNA for the canine neurotensin/neuromedin N precursor

    Cloned cDNAs encoding neurotensin were isolated from a cDNA library derived from primary cultures of canine enteric mucosa cells. Nucleotide sequence analysis using 32P-labeled nucleotides, has revealed the primary structure of a 170-amino acid precursor protein that encodes both neurotensin and the neurotensin-like peptide neuromedin N. The peptide-coding domains are located in tandem near the carboxyl terminus of the precursor and are bounded and separated by the paired, basic amino acid residues Lys-Arg. An additional coding domain, resembling neuromedin N, occurs immediately after an Arg-Arg basic amino acid pair located in the central region of the precursor. Additional amino acid homologies suggest that tandem duplications have contributed to the structure of the gene. RNA blot analysis, using the cloned cDNA probe, has revealed several mRNA species ranging in size from 500 to 980 nucleotides in the canine enteric mucosa. In contrast a single RNA species of 1500 nucleotides was detected in bovine hypothalamus poly-(A)+ RNA. The ability of the canine probe to cross-hybridize with bovine mRNA suggest that this probe can be used to isolate neurotensin/neuromedin N genes from other mammalian species

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

    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

  5. An approach to modelling operator behaviour in integrated dynamic accident sequence analysis

    The paper describes an integrated dynamic methodology for simulating nuclear power plant accidents, with special focus on the operator behaviour model. The overall model consists of accident sequence pre-processor, operator response model, safety and support system model, plant dependence model, thermal hydraulics model, and accident sequence scheduler. The operator model consists of the knowledge base (KB) and the decision making module (DM). KB consists of rules of behaviour. Behaviour is guided by emergency operating procedures (EOPs), thermal hydraulics parameters of the plant, system status, and other factors including stress, training, experience, etc. Possible error mechanisms in following symptom based EOPs are mentioned, and factors which cause some of these errors are identified. Plant parameters are classified as ''diagnostic'' and ''control''. Comparison of operator expectations and plant inputs guides the behaviour. System states affect only control action and not diagnosis. The decision maker simulates the operator behaviour in the way it accesses the KB, assuming that the KB contains all the knowledge that is necessary for managing the accident. This is modelled through a ''filter'' concept where the factors that affect behaviour are filters that affect the access to KB. Actions are categorized in verifying the response of reactor protection systems, and in controlling inventory and heat removal. System modelling is done at system rather than component level since operator actions affect the plant at system level. The methodology is being implemented in PC environment. Possible applications include analysis of causes and consequences of operator actions, particularly errors of commission, EOP validation, analysis of dynamic effects of accident sequences, and performing probabilistic risk assessments. 15 refs, 2 figs, 1 tab

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

    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)

  7. Probabilistic methods for identification of significant accident sequences in loop-type LMFBRs

    The aim of the Probabilistic Accident Progression Analysis (PAPA) described herein is to establish a framework for better use of the probability measure; first, as a basis for deterministic calculations, and second, as a part of a comprehensive method for risk assessment in its own right. The achievement of this goal rests on: (1) improvements in the existing approaches for acquisition and analysis of accident sequences; (2) defining a new measure of probabilistic importance that aids in the ranking of sequences of highly uncertain events; and (3) implementation of new techniques for quantification of dependent failures of similar components. The existing techniques related to the above three topics are discussed and the state of the art is reviewed. The PAPA approach is described. The techniques of PAPA are applied to a class of Protected Transients (transients in which the reactor is successfully shutdown) in the Clinch River Breeder Reactor Plant (CRBRP). The results of the application of these techniques are described

  8. Probabilistic methods for identification of significant accident sequences in loop-type LMFBRs

    Jamali, K M.A.

    1983-06-01

    The aim of the Probabilistic Accident Progression Analysis (PAPA) described herein is to establish a framework for better use of the probability measure; first, as a basis for deterministic calculations, and second, as a part of a comprehensive method for risk assessment in its own right. The achievement of this goal rests on: (1) improvements in the existing approaches for acquisition and analysis of accident sequences; (2) defining a new measure of probabilistic importance that aids in the ranking of sequences of highly uncertain events; and (3) implementation of new techniques for quantification of dependent failures of similar components. The existing techniques related to the above three topics are discussed and the state of the art is reviewed. The PAPA approach is described. The techniques of PAPA are applied to a class of Protected Transients (transients in which the reactor is successfully shutdown) in the Clinch River Breeder Reactor Plant (CRBRP). The results of the application of these techniques are described.

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

    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

  10. Co-ordinated research programme on reference studies on probabilistic modelling of accident sequences

    The co-ordinated research programme (CRP) on probabilistic modelling of accident sequences was established in order to ensure that International Atomic Energy Agency (IAEA) Member States not previously involved in international benchmark exercises obtain adequate practice in applying the available PSA techniques and benefit from the extensive international experience. A supportive peer review group was formed to provide guidance and transfer the insights derived from similar European projects. Seventeen countries participate in this programme which will be completed during 1991. Three working groups have been organized around different reactor types, namely WWER-440 PWRs (with a subgroup analysing AST-500, a district heating plant), Framatome PWRs and CANDU. Each participant in a group studied the same initiating event for a reference plant. For detailed analysis one particular accident sequence has been selected by each team. The logic models (event trees and fault trees) were developed and accident sequences were quantified. Sensitivity analyses are presently in progress. The paper presents some preliminary results and insights. The experiences gained from this CRP are considered as extremely useful for the national PSA programmes in several IAEA Member States. (author). 8 refs, 2 figs, 1 tab

  11. Severe accident source term characteristics for selected Peach Bottom sequences predicted by the MELCOR Code

    Carbajo, J.J. [Oak Ridge National Lab., TN (United States)

    1993-09-01

    The purpose of this report is to compare in-containment source terms developed for NUREG-1159, which used the Source Term Code Package (STCP), with those generated by MELCOR to identify significant differences. For this comparison, two short-term depressurized station blackout sequences (with a dry cavity and with a flooded cavity) and a Loss-of-Coolant Accident (LOCA) concurrent with complete loss of the Emergency Core Cooling System (ECCS) were analyzed for the Peach Bottom Atomic Power Station (a BWR-4 with a Mark I containment). The results indicate that for the sequences analyzed, the two codes predict similar total in-containment release fractions for each of the element groups. However, the MELCOR/CORBH Package predicts significantly longer times for vessel failure and reduced energy of the released material for the station blackout sequences (when compared to the STCP results). MELCOR also calculated smaller releases into the environment than STCP for the station blackout sequences.

  12. Severe accident source term characteristics for selected Peach Bottom sequences predicted by the MELCOR Code

    The purpose of this report is to compare in-containment source terms developed for NUREG-1159, which used the Source Term Code Package (STCP), with those generated by MELCOR to identify significant differences. For this comparison, two short-term depressurized station blackout sequences (with a dry cavity and with a flooded cavity) and a Loss-of-Coolant Accident (LOCA) concurrent with complete loss of the Emergency Core Cooling System (ECCS) were analyzed for the Peach Bottom Atomic Power Station (a BWR-4 with a Mark I containment). The results indicate that for the sequences analyzed, the two codes predict similar total in-containment release fractions for each of the element groups. However, the MELCOR/CORBH Package predicts significantly longer times for vessel failure and reduced energy of the released material for the station blackout sequences (when compared to the STCP results). MELCOR also calculated smaller releases into the environment than STCP for the station blackout sequences

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

    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.

  14. Development of a component Monte Carlo program for accident sequence analysis to apply for reprocessing facility

    In consideration of application for reprocessing facility, where a variety of causal events such as equipment failure and human error might occur, and the event progression would take place with relatively substantial time delay before getting to the accident stage, a component Monte Carlo program for accident sequence analysis has been developed to pursue chronologically the probabilistic behavior of each component failure and repair in an exact manner. In comparison with analytical formulation and its calculated results, this Monte Carlo technique is shown to predict a reasonable result. Then, taking an example for a sample problem from a German reprocessing facility model, an accident sequence of red-oil explosion in a plutonium evaporator is analyzed to give a comprehensive interpretation about statistic variation range and computer time elapsed for random walk history calculations. Furthermore, to discuss about its applicability for the practical case of plant system with complex component constitution, a possibility of drastic speed-up of computation is shown by parallelization of the computer program. (author)

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

    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

  16. Reconstructing SALMFamide neuropeptide precursor evolution in the phylum Echinodermata: ophiuroid and crinoid sequence data provide new insights

    Maurice R Elphick

    2015-02-01

    Full Text Available The SALMFamides are a family of neuropeptides that act as muscle relaxants in echinoderms. Analysis of genome/transcriptome sequence data from the sea urchin Strongylocentrotus purpuratus (Echinoidea, the sea cucumber Apostichopus japonicus (Holothuroidea and the starfish Patiria miniata (Asteroidea reveals that in each species there are two types of SALMFamide precursor: an L-type precursor comprising peptides with a C-terminal LxFamide-type motif and an F-type precursor solely or largely comprising peptides with a C-terminal FxFamide-type motif. Here we have identified transcripts encoding SALMFamide precursors in the brittle star Ophionotus victoriae (Ophiuroidea and the feather star Antedon mediterranea (Crinoidea. We have also identified SALMFamide precursors in other species belonging to each of the five echinoderm classes. As in S. purpuratus, A. japonicus and P. miniata, in O. victoriae there is one L-type precursor and one F-type precursor. However, in A. mediterranea only a single SALMFamide precursor was found, comprising two peptides with a LxFamide-type motif, one with a FxFamide-type motif, five with a FxLamide-type motif and four with a LxLamide-type motif. As crinoids are basal to the Echinozoa (Holothuroidea + Echinoidea and Asterozoa (Asteroidea + Ophiuroidea in echinoderm phylogeny, one model of SALMFamide precursor evolution would be that ancestrally there was a single SALMFamide gene encoding a variety of SALMFamides (as in crinoids, which duplicated in a common ancestor of the Echinozoa and Asterozoa and then specialised to encode L-type SALMFamides or F-type SALMFamides. Alternatively, a second SALMFamide precursor may remain to be discovered or may have been lost in crinoids. Further insights will be obtained if SALMFamide receptors are identified, which would provide a molecular basis for experimental analysis of the functional significance of the cocktails of SALMFamides that exist in echinoderms.

  17. Reconstructing SALMFamide Neuropeptide Precursor Evolution in the Phylum Echinodermata: Ophiuroid and Crinoid Sequence Data Provide New Insights.

    Elphick, Maurice R; Semmens, Dean C; Blowes, Liisa M; Levine, Judith; Lowe, Christopher J; Arnone, Maria I; Clark, Melody S

    2015-01-01

    The SALMFamides are a family of neuropeptides that act as muscle relaxants in echinoderms. Analysis of genome/transcriptome sequence data from the sea urchin Strongylocentrotus purpuratus (Echinoidea), the sea cucumber Apostichopus japonicus (Holothuroidea), and the starfish Patiria miniata (Asteroidea) reveals that in each species there are two types of SALMFamide precursor: an L-type precursor comprising peptides with a C-terminal LxFamide-type motif and an F-type precursor solely or largely comprising peptides with a C-terminal FxFamide-type motif. Here, we have identified transcripts encoding SALMFamide precursors in the brittle star Ophionotus victoriae (Ophiuroidea) and the feather star Antedon mediterranea (Crinoidea). We have also identified SALMFamide precursors in other species belonging to each of the five echinoderm classes. As in S. purpuratus, A. japonicus, and P. miniata, in O. victoriae there is one L-type precursor and one F-type precursor. However, in A. mediterranea only a single SALMFamide precursor was found, comprising two peptides with a LxFamide-type motif, one with a FxFamide-type motif, five with a FxLamide-type motif, and four with a LxLamide-type motif. As crinoids are basal to the Echinozoa (Holothuroidea + Echinoidea) and Asterozoa (Asteroidea + Ophiuroidea) in echinoderm phylogeny, one model of SALMFamide precursor evolution would be that ancestrally there was a single SALMFamide gene encoding a variety of SALMFamides (as in crinoids), which duplicated in a common ancestor of the Echinozoa and Asterozoa and then specialized to encode L-type SALMFamides or F-type SALMFamides. Alternatively, a second SALMFamide precursor may remain to be discovered or may have been lost in crinoids. Further insights will be obtained if SALMFamide receptors are identified, which would provide a molecular basis for experimental analysis of the functional significance of the "cocktails" of SALMFamides that exist in echinoderms. PMID:25699014

  18. Reconstructing SALMFamide Neuropeptide Precursor Evolution in the Phylum Echinodermata: Ophiuroid and Crinoid Sequence Data Provide New Insights

    Elphick, Maurice R.; Semmens, Dean C.; Blowes, Liisa M.; Levine, Judith; Lowe, Christopher J.; Arnone, Maria I.; Clark, Melody S.

    2015-01-01

    The SALMFamides are a family of neuropeptides that act as muscle relaxants in echinoderms. Analysis of genome/transcriptome sequence data from the sea urchin Strongylocentrotus purpuratus (Echinoidea), the sea cucumber Apostichopus japonicus (Holothuroidea), and the starfish Patiria miniata (Asteroidea) reveals that in each species there are two types of SALMFamide precursor: an L-type precursor comprising peptides with a C-terminal LxFamide-type motif and an F-type precursor solely or largely comprising peptides with a C-terminal FxFamide-type motif. Here, we have identified transcripts encoding SALMFamide precursors in the brittle star Ophionotus victoriae (Ophiuroidea) and the feather star Antedon mediterranea (Crinoidea). We have also identified SALMFamide precursors in other species belonging to each of the five echinoderm classes. As in S. purpuratus, A. japonicus, and P. miniata, in O. victoriae there is one L-type precursor and one F-type precursor. However, in A. mediterranea only a single SALMFamide precursor was found, comprising two peptides with a LxFamide-type motif, one with a FxFamide-type motif, five with a FxLamide-type motif, and four with a LxLamide-type motif. As crinoids are basal to the Echinozoa (Holothuroidea + Echinoidea) and Asterozoa (Asteroidea + Ophiuroidea) in echinoderm phylogeny, one model of SALMFamide precursor evolution would be that ancestrally there was a single SALMFamide gene encoding a variety of SALMFamides (as in crinoids), which duplicated in a common ancestor of the Echinozoa and Asterozoa and then specialized to encode L-type SALMFamides or F-type SALMFamides. Alternatively, a second SALMFamide precursor may remain to be discovered or may have been lost in crinoids. Further insights will be obtained if SALMFamide receptors are identified, which would provide a molecular basis for experimental analysis of the functional significance of the “cocktails” of SALMFamides that exist in echinoderms. PMID:25699014

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

    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.

  20. Precursors to potential severe core damage accidents: 1994, a status report. Volume 21: Main report and appendices A--H

    Nine operational events that affected eleven commercial light-water reactors (LWRs) during 1994 and that are considered to be precursors to potential severe core damage are described. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10-6. These events were identified by computer-screening the 1994 licensee event reports from commercial LWRs to identify those that could be potential precursors. Candidate precursors were then selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters and regional offices to ensure that the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969--1981 and 1984--1993 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for events. This document is bound in two volumes: Vol. 21 contains the main report and Appendices A--H; Vol. 22 contains Appendix 1

  1. Precursor incident program at EDF

    The precursor program was started by EDF in 1994, after an investigation of the US NRC's Accident Sequence Precursor Program. Since then, reported operational events identified as Safety Outstanding Events have been analyzed whenever possible using probabilistic methods based on PSAs. Analysis provides an estimate of the remaining protection against core damage at the time the incident occurred. Measuring the incidents' severity enables to detect incidents important regarding safety. Moreover, the most efficient feedback actions can be derived from the main accident sequences identified through the analysis. Therefore, incident probabilistic analysis provides a way to assess priorities in terms of treatment and resource allocation, and so, to implement countermeasures preventing further occurrence and development of the most significant incidents. As some incidents cannot be analyzed using this method, probabilistic analysis can only be one among the methods used to assess the nuclear power plants' safety level. Nevertheless, it provides an interesting complement to classical methods of deterministic studies. (author)

  2. Trending analysis of precursor events

    The Accident Sequence Precursor (ASP) Program of United States Nuclear Regulatory Commission (U.S.NRC) identifies and categorizes operational events at nuclear power plants in terms of the potential for core damage. The ASP analysis has been performed on yearly basis and the results have been published in the annual reports. This paper describes the trends in initiating events and dominant sequences for 459 precursors identified in the ASP Program during the 1969-94 period and also discusses a comparison with dominant sequences predicted in the past Probabilistic Risk Assessment (PRA) studies. These trends were examined for three time periods, 1969-81, 1984-87 and 1988-94. Although the different models had been used in the ASP analyses for these three periods, the distribution of precursors by dominant sequences show similar trends to each other. For example, the sequences involving loss of both main and auxiliary feedwater were identified in many PWR events and those involving loss of both high and low coolant injection were found in many BWR events. Also, it was found that these dominant sequences were comparable to those determined to be dominant in the predictions by the past PRAs. As well, a list of the 459 precursors identified are provided in Appendix, indicating initiating event types, unavailable systems, dominant sequences, conditional core damage probabilities, and so on. (author)

  3. Evaluation of operating experience: The precursor study (GPS) performed in the Federal Republic of Germany. Working material

    Probabilistic Safety Assessments (PSAs) are systematic and quantitative predictions of possible accident scenarios at technical installations on the basis of data gained from the past experience on similar technical installations. For supporting PSAs by operational experience as far as possible Precursor studies are performed. An Accident Sequence Precursor is defined as an observed event which could results, in coincidence with additional postulated events, in a potential severe core damage accident. In the presented case study the procedure of such Precursor studies is explained. Particularly, the methodology and the results of the plant-specific Precursor (GPS) performed in the Federal Republic of Germany are shown in detail. 26 refs, 13 figs, 8 tabs

  4. Ribonuclease "XlaI," an activity from Xenopus laevis oocytes that excises intervening sequences from yeast transfer ribonucleic acid precursors.

    Otsuka, A; de Paolis, A; Tocchini-Valentini, G P

    1981-01-01

    A ribonuclease (RNase) activity, RNase "XlaI," responsible for the excision of intervening sequences from two yeast transfer ribonucleic acid (tRNA) precursors, pre-tRNA(Tyr) and pre-tRNA(3Leu), has been purified 54-fold from nuclear extracts of Xenopus laevis oocytes. The RNase preparation is essentially free of contaminating RNase. A quantitative assay for RNase XlaI was developed, and the reaction products were characterized. RNase XlaI cleavage sites in the yeast tRNA precursors were identical to those made by yeast extracts (including 3'-phosphate and 5'-hydroxyl termini). Cleavage of pre-tRNA(3Leu) by RNase XlaI and subsequent ligation of the half-tRNA molecules do not require removal of the 5' leader or 3' trailer sequences. Images PMID:6765601

  5. Nucleotide sequence of the cDNA encoding the precursor of the beta subunit of rat lutropin.

    Chin, W W; Godine, J E; Klein, D. R.; Chang, A S; Tan, L K; Habener, J F

    1983-01-01

    We have determined the nucleotide sequences of cDNAs encoding the precursor of the beta subunit of rat lutropin, a polypeptide hormone that regulates gonadal function, including the development of gametes and the production of steroid sex hormones. The cDNAs were prepared from poly(A)+ RNA derived from the pituitary glands of rats 4 weeks after ovariectomy and were cloned in bacterial plasmids. Bacterial colonies containing transfected plasmids were screened by hybridization with a 32P-labele...

  6. Severe-accident-sequence assessment of hypothetical complete-station blackout at the Browns Ferry Nuclear Plant

    Yue, D.D.; Condon, W.A.

    1981-01-01

    An investigation has been made of various accident sequence which may occur following a complete loss of offsite and onsite ac power at a Boiling Water Reactor nuclear power plant. The investigation was performed for the Browns Ferry Nuclear Power Plant, and all accident sequences resulted in a hypothetical core meltdown. Detailed calculations were performed with the MARCH computer meltdown. Detailed calcuations were performed with the MARCH computer code containing a decay power calculation which was modified to include the actinides. This change has resulted in shortening the time before core uncovery by approx. 18%, and reducing the time before the start of core melting by approx. 26%. Following the hypothetical core meltdown accident, the drywell electric penetration assembly seals have been identified as the most likely leak pathway outside the containment. This potential mode of containment failure occurs at a pressure approx. 30% lower than that analyzed in the Reactor Safety Study.

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

    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

  8. Characterization and gene sequence of the precursor of elafin, an elastase-specific inhibitor in bronchial secretions.

    Sallenave, J M; Silva, A

    1993-04-01

    Human bronchial mucous secretions have been shown to contain inhibitors of serine proteinases secreted by neutrophils. The role of these inhibitors is probably to control the enzymes secreted in the airways and in the lung interstitium. Three of these inhibitors have been identified and characterized: alpha 1-proteinase inhibitor, mucus proteinase inhibitor, and elafin. The elafin molecule, a 6.0 kD inhibitor of serine proteinases shows homology with mucus proteinase inhibitor. We recently isolated both molecules in bronchial secretions. In this report, we present evidence for the existence of a precursor of the elafin molecule. We have cloned and sequenced the gene for this precursor and show that it is composed of three exons. The coding information for a 117 amino acid precursor protein of elafin (inclusive of the signal peptide) is contained in the first two exons. This was confirmed at the mRNA and protein levels. By Northern Blot analysis we detected a 800 bp long product, and by immunoaffinity we detected in sputum and in cultured epithelial cell supernatant (NCI-H322 cell line) a 12 kD protein species cross-reacting with anti-elafin IgG. The finding of possible cross-linking function for the precursor in addition to its antiproteinase activity indicates a possible role for this molecule as a cross-linker agent in the extracellular matrix. PMID:8476637

  9. Presumption of accident progression of Fukushima Dai-ichi and comparison of event sequences among NPSs affected by Tsunami

    In order to contribute to an improvement of accident managements and station blackout countermeasures for NPSs in Japan after the accident of Fukushima Dai-ichi Nuclear Power Stations (NPSs), Japan Nuclear Energy Safety Organization (JNES) has analyzed the progress of the accident scenarios and situation of reactor of the unit 1, 2 and 3 of Fukushima Dai-ichi NPS based on the information such as fragmentary plant chronologies, parameters and information disclosed by TEPCO and tried to figure out the rational explanation of behavior of each plant. In addition, JNES has compared and categorized the event sequences by using event tree method based on the event progression of Fukushima Dai-ichi, Fukushima Dai-ni, Tokai Dai-ni and Onagawa NPSs in order to extract important events and key factors from the viewpoint of prevention of core damage events and impact to event progression. (author)

  10. Environment source terms for ex-vessel FW/SB LOCA accident sequences in ITER EDA

    The paper presents the environmental source term EST evaluation results for some ITER accident sequences, with reference to the activated materials contribution. The assessment is based on the end-of-1994 ITER baseline design and it refers to ex-vessel LOCAs in the first wall FW and shielding blanket SB heat transport system HTS. The main ITER characteristics are: fusion power 1.5 GW, average neutron power load on the outboard first wall 1 MW/m2, fluence 3 MW-y/m2, average pulse length 1,000 s, dwell time 1,200 s. The structural material for FW and SB is AISI 316L. (Mn 1.8 wt.%, Co 0.17 wt.%) stainless steel. Four independent loops for FW and SB heat transport system are considered. The European multi-code approach is briefly described jointly with the computer tools used for the assessment. A sensitivity analysis has been performed to verify the impact of the water chemistry on the environmental release composition and mass

  11. Environment source terms for ex-vessel FW/SB LOCA accident sequences in ITER EDA

    Cambi, G. [Bologna Univ. (Italy). Physics Dept.; Cepraga, D.G. [ENEA, Bologna (Italy). Innovation Dept.; Di Pace, L. [ENEA, Frascati (Italy). Fusion Sector CR di Frascati; Porfiri, M.T. [ENEA, Frascati (Italy). Fusion Dept.

    1995-12-31

    The paper presents the environmental source term EST evaluation results for some ITER accident sequences, with reference to the activated materials contribution. The assessment is based on the end-of-1994 ITER baseline design and it refers to ex-vessel LOCAs in the first wall FW and shielding blanket SB heat transport system HTS. The main ITER characteristics are: fusion power 1.5 GW, average neutron power load on the outboard first wall 1 MW/m{sup 2}, fluence 3 MW-y/m{sup 2}, average pulse length 1,000 s, dwell time 1,200 s. The structural material for FW and SB is AISI 316L. (Mn 1.8 wt.%, Co 0.17 wt.%) stainless steel. Four independent loops for FW and SB heat transport system are considered. The European multi-code approach is briefly described jointly with the computer tools used for the assessment. A sensitivity analysis has been performed to verify the impact of the water chemistry on the environmental release composition and mass.

  12. Rbfox proteins regulate microRNA biogenesis by sequence-specific binding to their precursors and target downstream Dicer

    Chen, Yu; Zubovic, Lorena; Yang, Fan; Godin, Katherine; Pavelitz, Tom; Castellanos, Javier; Macchi, Paolo; Varani, Gabriele

    2016-01-01

    Rbfox proteins regulate tissue-specific splicing by targeting a conserved GCAUG sequence within pre-mRNAs. We report here that sequence-specific binding of the conserved Rbfox RRM to miRNA precursors containing the same sequence motif in their terminal loops, including miR-20b and miR-107, suppresses their nuclear processing. The structure of the complex between precursor miR-20b and Rbfox RRM shows the molecular basis for recognition, and reveals changes in the stem-loop upon protein binding. In mammalian cells, Rbfox2 downregulates mature miR-20b and miR-107 levels and increases the expression of their downstream targets PTEN and Dicer, respectively, suggesting that Rbfox2 indirectly regulates many more cellular miRNAs. Thus, some of the widespread cellular functions of Rbfox2 protein are attributable to regulation of miRNA biogenesis, and might include the mis-regulation of miR-20b and miR-107 in cancer and neurodegeneration. PMID:27001519

  13. Rbfox proteins regulate microRNA biogenesis by sequence-specific binding to their precursors and target downstream Dicer.

    Chen, Yu; Zubovic, Lorena; Yang, Fan; Godin, Katherine; Pavelitz, Tom; Castellanos, Javier; Macchi, Paolo; Varani, Gabriele

    2016-05-19

    Rbfox proteins regulate tissue-specific splicing by targeting a conserved GCAUG sequence within pre-mRNAs. We report here that sequence-specific binding of the conserved Rbfox RRM to miRNA precursors containing the same sequence motif in their terminal loops, including miR-20b and miR-107, suppresses their nuclear processing. The structure of the complex between precursor miR-20b and Rbfox RRM shows the molecular basis for recognition, and reveals changes in the stem-loop upon protein binding. In mammalian cells, Rbfox2 downregulates mature miR-20b and miR-107 levels and increases the expression of their downstream targets PTEN and Dicer, respectively, suggesting that Rbfox2 indirectly regulates many more cellular miRNAs. Thus, some of the widespread cellular functions of Rbfox2 protein are attributable to regulation of miRNA biogenesis, and might include the mis-regulation of miR-20b and miR-107 in cancer and neurodegeneration. PMID:27001519

  14. A dynamic event tree informed approach to probabilistic accident sequence modeling: Dynamics and variabilities in medium LOCA

    In Probability Safety Assessments, accident scenario dynamics are addressed in the accident sequence analysis task. In an analyst-driven, iterative process, assumptions are made about equipment responses and operator actions and simulations of the scenario evolution are performed. To calculate how scenario dynamics and stochastic variabilities may affect the results of this process in terms of estimated risk, this work applies Dynamic Event Trees (DETs) to more comprehensively examine the accident scenario space. Alternative event tree models are developed and the core damage frequency is quantified to reveal the effects of different delineations of the sequences and of the bounding assumptions underlying success criteria. The results from a case study on Medium-break Loss of Coolant Accident scenarios in a Pressurized Water Reactor are presented, considering the break size, available injection trains, and the timing of rapid cooldown and the switchover to recirculation. The results show not only that estimated risk can be very sensitive to the numerous assumptions made in current accident sequence analysis but also that bounding assumptions do not always result in conservative risk estimates, thereby confirming the benefits that DETs provide in terms of characterizing scenario dynamics. - Highlights: • The overall most challenging MLOCA break is at neither extreme of the size range. • Selecting the limiting break size influenced estimated risk strongly (6″ vs 7″). • Success criteria can be defined more realistically by splitting the MLOCA range. • A more demanding success criterion for one top event can reduce overall risk. • Non-limiting success branches may lead to more demanding subsequent success criteria

  15. Determination of the Specificity Landscape for Ribonuclease P Processing of Precursor tRNA 5' Leader Sequences.

    Niland, Courtney N; Zhao, Jing; Lin, Hsuan-Chun; Anderson, David R; Jankowsky, Eckhard; Harris, Michael E

    2016-08-19

    Maturation of tRNA depends on a single endonuclease, ribonuclease P (RNase P), to remove highly variable 5' leader sequences from precursor tRNA transcripts. Here, we use high-throughput enzymology to report multiple-turnover and single-turnover kinetics for Escherichia coli RNase P processing of all possible 5' leader sequences, including nucleotides contacting both the RNA and protein subunits of RNase P. The results reveal that the identity of N(-2) and N(-3) relative to the cleavage site at N(1) primarily control alternative substrate selection and act at the level of association not the cleavage step. As a consequence, the specificity for N(-1), which contacts the active site and contributes to catalysis, is suppressed. This study demonstrates high-throughput RNA enzymology as a means to globally determine RNA specificity landscapes and reveals the mechanism of substrate discrimination by a widespread and essential RNA-processing enzyme. PMID:27336323

  16. Nucleotide Sequence of a Chicken Vitellogenin Gene and Derived Amino Acid Sequence of the Encoded Yolk Precursor Protein

    Schip, Fred D. van het; Samallo, John; Broos, Jaap; Ophuis, Jan; Mojet, Mart; Gruber, Max; AB, Geert

    1987-01-01

    The gene encoding the major vitellogenin from chicken has been completely sequenced and its exon-intron organization has been established. The gene is 20,342 base-pairs long and contains 35 exons with a combined length of 5787 base-pairs. They encode the 1850-amino acid pre-peptide of vitellogenin,

  17. Sequence analysis of a cDNA coding for a pancreatic precursor to somatostatin.

    Taylor, W.L.; Collier, K J; Deschenes, R J; Weith, H L; Dixon, J. E.

    1981-01-01

    A synthetic oligonucleotide having the sequence d(T-T-C-C-A-G-A-A-G-A-A) deduced from the amino acid sequence Phe-Phe-Trp-Lys of somatostatin-14 was used to prime the synthesis of a cDNA from channel catfish (Ictalurus punctatus) pancreatic poly(A)-RNA. The major product of this reaction was a cDNA fragment of 565 nucleotides. Chemical sequence analysis of the cDNA fragment revealed that it was complementary to a mRNA coding for somatostatin. The 565-nucleotide cDNA hybridizes strongly with a...

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

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

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

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

    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.

  1. Nucleotide sequence of cloned cDNA for human sphingolipid activator protein 1 precursor

    Two cDNA clones encoding prepro-sphingolipid activator protein 1 (SAP-1) were isolated from a λ gt11 human hepatoma expression library using polyclonal antibodies. These had inserts of ≅ 2 kilobases (λ-S-1.2 and λ-S-1.3) and both were both homologous with a previously isolated clone (λ-S-1.1) for mature SAP-1. The authors report here the nucleotide sequence of the longer two EcoRI fragments of S-1.2 and S-1.3 that were not the same and the derived amino acid sequences of mature SAP-1 and its prepro form. The open reading frame encodes 19 amino acids, which are colinear with the amino-terminal sequence of mature SAP-1, and extends far beyond the predicted carboxyl terminus of mature SAP-1, indicating extensive carboxyl-terminal processing. The nucleotide sequence of cDNA encoding prepro-SAP-1 includes 1449 bases from the assigned initiation codon ATG at base-pair 472 to the stop codon TGA at base-pair 1921. The first 23 amino acids coded after the initiation ATG are characteristic of a signal peptide. The calculated molecular mass for a polypeptide encoded by 1449 bases is ≅ 53 kDa, in keeping with the reported value for pro-SAP-1. The data indicate that after removal of the signal peptide mature SAP-1 is generated by removing an additional 7 amino acids from the amino terminus and ≅ 373 amino acids from the carboxyl terminus. One potential glycosylation site was previously found in mature SAP-1. Three additional potential glycosylation sites are present in the processed carboxyl-terminal polypeptide, which they designate as P-2

  2. Rapid Capture Next-Generation Sequencing in Clinical Diagnostics of Kinase Pathway Aberrations in B-Cell Precursor ALL.

    Stadt, Udo Zur; Escherich, Gabriele; Indenbirken, Daniela; Alawi, Malik; Adao, Manuela; Horstmann, Martin A

    2016-07-01

    Comprehensive next-generation sequencing (NGS) applications have recently identified various recurrent kinase and cytokine receptor rearrangements in Ph-like B-cell precursor (BCP) acute lymphoblastic leukemia (ALL) amenable to tyrosin kinase inhibitor treatment. For rapid diagnostics of kinase pathway aberrations in minimal residual disease (MRD) high-risk BCP-ALL, we developed a PCR-independent NGS custom enrichment capture panel targeting recurrent genomic alterations, which allows for the identification of unknown 5' fusion partner genes and precise mapping of variable genomic breakpoints. Using a standardized bioinformatics algorithm, we identified kinase and cytokine receptor rearrangements in the majority of ALL patients with high burden of postinduction MRD and enrichment of IKZF1 mutation or deletion (IKZF1(del) ). PMID:27007619

  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

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

    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)

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

    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

  6. Large break loss of coolant severe accident sequences at the HFIR [High Flux Isotope Reactor

    An assessment of many potential HFIR severe accident phenomena was conducted during the HFIR design effort, and many severe accident mitigating features were designed into the plant. These evaluation typically incorporated a ''bounding'' or highly conservative analysis approach and employed tools and techniques representative of the state of knowledge in the mid-1960s. Recently, programs to address severe accident issues were initiated at the Oak Ridge National Laboratory (ORNL) to support the HFIR probabilistic risk assessment (PRA) and equipment qualification and accident management studies. This paper presents the results of environment condition calculations conducted to evaluate a response of HFIR's heat exchanger cell environment to a double-ended rupture of a 0.25 m diameter coolant loop downstream of the circulating pump and check valve. The confinement calculations were performed using an atmospheric fission product source for the heat exchanger cell consistent with, but more conservative than that stipulated in Regulatory Guide 1.89. The results of the calculations indicate that the heat exchanger cell atmospheric temperature peaks at 377 K 225 seconds into the transient and then begins decreasing at approximately 1.7 K per minute. 8 refs., 5 figs

  7. A probability risk assessment for MACSTOR/KN-400 during an air inlet blockage accident sequence

    The safety assessment framework for evaluating the spent fuel dry storage facility during the air inlet blockage accident composing of three phases has been established and applied to an interim storage system. They include the analysis of the failure probability of a basket and a cylinder, the accident modeling of spent fuel dry storage facility and the accident consequence assessments. The first phase of the analysis calculated the module failure probability by modeling of the basket and the cylinder, which is major element for containing radioactive substances. The second phase includes a modeling of spent fuel dry storage facility. At this phase, the probability that radioactive substances are released to outside when the initial event happens has been calculated by the construction of the event tree methods against a various elements which affects the air inlet blockage accident. At the third phase of releasing radioactive substances, the radiation damage to affect neighborhood and storage facility worker using MACCS2 code has been evaluated quantitatively. (author)

  8. Large-scale information entropy analysis of important sites in mature and precursor miRNA sequences

    2009-01-01

    In order to find evidence of consistent sequence conservation or the base correlation degree in miRNA,some important sites in the sequences of reported miRNA and their precursors(pre-miRNA) were investigated via information entropy analysis.Twelve different groups of sites were obtained from special locations(head,tail) in miRNAs of different sources according to taxonomy(animal,plant and virus) and then analyzed by measuring the single base information redundancy(D1(L)) and the adjacent base related information redundancy(D2(L)).The results showed that D2(L) has more information than D1(L),though D1(L) changes roughly consistently with D2(L) in each group.Viral pre-miRNAs are more conservative than those belonging to animals or plants.In addition,U is dominant in most sites compared with other nucleotides.It was also found that in the middle of several groups,there were sites where miRNAs were cut down from pre-miRNAs by Enzyme Dicer which were significantly conservative.This phenomenon shows that the conservatism is an aspect of the of miRNA and may be involved in the recognition and cutting by the Dicer.Those results provided another perspective for understanding more about the primary structure of pre-miRNA.

  9. Using BAC transgenesis in zebrafish to identify regulatory sequences of the amyloid precursor protein gene in humans

    Shakes Leighcraft A

    2012-09-01

    Full Text Available Abstract Background Non-coding DNA in and around the human Amyloid Precursor Protein (APP gene that is central to Alzheimer’s disease (AD shares little sequence similarity with that of appb in zebrafish. Identifying DNA domains regulating expression of the gene in such situations becomes a challenge. Taking advantage of the zebrafish system that allows rapid functional analyses of gene regulatory sequences, we previously showed that two discontinuous DNA domains in zebrafish appb are important for expression of the gene in neurons: an enhancer in intron 1 and sequences 28–31 kb upstream of the gene. Here we identify the putative transcription factor binding sites responsible for this distal cis-acting regulation, and use that information to identify a regulatory region of the human APP gene. Results Functional analyses of intron 1 enhancer mutations in enhancer-trap BACs expressed as transgenes in zebrafish identified putative binding sites of two known transcription factor proteins, E4BP4/ NFIL3 and Forkhead, to be required for expression of appb. A cluster of three E4BP4 sites at −31 kb is also shown to be essential for neuron-specific expression, suggesting that the dependence of expression on upstream sequences is mediated by these E4BP4 sites. E4BP4/ NFIL3 and XFD1 sites in the intron enhancer and E4BP4/ NFIL3 sites at −31 kb specifically and efficiently bind the corresponding zebrafish proteins in vitro. These sites are statistically over-represented in both the zebrafish appb and the human APP genes, although their locations are different. Remarkably, a cluster of four E4BP4 sites in intron 4 of human APP exists in actively transcribing chromatin in a human neuroblastoma cell-line, SHSY5Y, expressing APP as shown using chromatin immunoprecipitation (ChIP experiments. Thus although the two genes share little sequence conservation, they appear to share the same regulatory logic and are regulated by a similar set of transcription

  10. Application of MELCOR code to the MCCI analysis in Severe Accident Sequences

    This paper provides some of the technical aspects that can be applied to an analysis of the MCCI phenomena in a severe accident scenario using the current MELCOR version. An application methodology of the MELCOR current version to the analysis of MCCI, the phenomena of which are very uncertain and lack specific knowledge during severe accidents, was introduced. Assumptions based on the experimental results are used instead of the phenomenological detail modeling because of the modeling limitations. In the technical aspects of MCCI, code modification itself is not a big deal, because the code modification is needed for just the user flexibility. The concern will be whether the assumptions made for this analysis are acceptable or not. This paper illustrates the application of a severe accident analysis code, MELCOR, to the analysis of molten corium-concrete interaction (MCCI) phenomena in cases of severe accidents in nuclear power plants. In postulated degraded core accidents, followed by the failure of certain engineered safety features of the reactor system, the reactor core may eventually melt owing to the generation of decay heat. If the safety features of the reactor system fail to arrest the accident within the reactor vessel, the corium (molten core debris) will fall into the reactor cavity and attack the concrete walls and floor. Basemat melt-through refers to the process of concrete decomposition and destruction associated with a corium melt interacting with the reactor cavity basemat. The potential hazard of MCCI is the integrity of the containment building owing to the possibility of a basemat melt-through, containment overpressurization by non-condensible gases, or the oxidation of combustible gases. In the meantime, the MCCI still has large uncertainties in several phenomena such as melt spreading area, debris particulation, and heat transfer between the debris and cooling water. In particular, in the case where the water pool exists in the reactor

  11. Correlation between sequence conservation and structural thermodynamics of microRNA precursors from human, mouse, and chicken genomes

    Wang Shengqi

    2010-10-01

    Full Text Available Abstract Background Previous studies have shown that microRNA precursors (pre-miRNAs have considerably more stable secondary structures than other native RNAs (tRNA, rRNA, and mRNA and artificial RNA sequences. However, pre-miRNAs with ultra stable secondary structures have not been investigated. It is not known if there is a tendency in pre-miRNA sequences towards or against ultra stable structures? Furthermore, the relationship between the structural thermodynamic stability of pre-miRNA and their evolution remains unclear. Results We investigated the correlation between pre-miRNA sequence conservation and structural stability as measured by adjusted minimum folding free energies in pre-miRNAs isolated from human, mouse, and chicken. The analysis revealed that conserved and non-conserved pre-miRNA sequences had structures with similar average stabilities. However, the relatively ultra stable and unstable pre-miRNAs were more likely to be non-conserved than pre-miRNAs with moderate stability. Non-conserved pre-miRNAs had more G+C than A+U nucleotides, while conserved pre-miRNAs contained more A+U nucleotides. Notably, the U content of conserved pre-miRNAs was especially higher than that of non-conserved pre-miRNAs. Further investigations showed that conserved and non-conserved pre-miRNAs exhibited different structural element features, even though they had comparable levels of stability. Conclusions We proposed that there is a correlation between structural thermodynamic stability and sequence conservation for pre-miRNAs from human, mouse, and chicken genomes. Our analyses suggested that pre-miRNAs with relatively ultra stable or unstable structures were less favoured by natural selection than those with moderately stable structures. Comparison of nucleotide compositions between non-conserved and conserved pre-miRNAs indicated the importance of U nucleotides in the pre-miRNA evolutionary process. Several characteristic structural elements were

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

    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

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

    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 can 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

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

    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.)

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

    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)

  16. Benchmarking an expert fault detection and diagnostic system on the Three Mile Island accident event sequence

    Highlights: • Attempt was to use available resources at a nuclear plant in a value added fashion. • Includes plant measurement data and plant training and engineering simulator capabilities. • Correlating fault detection data for systems to develop of a deterministic fault identifications system. • After implementing a host of data manipulation algorithms, the results provided more information on the fault than expected. • TMI benchmark results in value added to the operator and system. - Abstract: Early fault identification systems enable detecting and diagnosing early onset faults or fault causes which allow maintenance planning on the equipment showing signs of deterioration or failure. This includes valve and leaks and small cracks in steam generator tubes usually detected by means of ultrasonic inspection. We have shown (Cilliers and Mulder, 2012) that detecting faults early during transient operation in NPPs is possible when coupled with a reliable reference to compare plant measurements with during transients. We have also shown (Cilliers, 2013) that by correlating the fault detection information as received from distributed systems it is possible to diagnose the faults in terms of location and magnitude. This paper makes use of the techniques and processes developed in the previous papers and apply it to a case study of the Three Mile Island accident. In this way we can determine how the improved information available could present the operator with a better idea to the state of the plant during situations where a combination of faults and transients prevents the operator and conventional systems to recognise the abnormal behaviour

  17. Extreme wind induced accident sequence analysis of the advanced test reactor

    An extreme wind probabilistic risk assessment (PRA) was performed for the Department of Energy (DOE) Advanced Test Reactor (ATR) as part of the external events analysis. The ATR is located at the Idaho National Engineering Laboratory (INEL) in Idaho. The analysis included evaluation of wind fragility of several structures. As part of the analysis the impact of extreme wind on the ATR core fuel damage frequency was evaluated. Loss of commercial power was modeled as an initiating event as a function of wind velocity. Normally, the components located inside the building are not affected directly as a result of wind. However, failure of a structure can eliminate several components as a result of spatial dependency. ATR support systems are located in several structures. Two walkdowns were conducted to collect the information on structures and components and to determine the structural-components interaction. Boolean equations were developed for core fuel damage sequences which included failure of components (structures) from extreme wind, random failures and operator errors. The sequences were quantified using wind hazard curves with wind fragility of the station power, components and non-wind unavailabilities. The result showed that contribution from extreme wind was less than 4%. ATR total core damage frequency from the internal and external events is estimated to be 5.E-5/yr. The system analysis (fault trees) was performed by the EG ampersand G, Idaho Inc. and the structures and components wind fragility and sequence quantification was performed by the EQE Engineering Consultants

  18. Mode of Action of RNase BN/RNase Z on tRNA Precursors: RNase BN DOES NOT REMOVE THE CCA SEQUENCE FROM tRNA*

    Dutta, Tanmay; Deutscher, Murray P.

    2010-01-01

    RNase BN, the Escherichia coli homolog of RNase Z, was previously shown to act as both a distributive exoribonuclease and an endoribonuclease on model RNA substrates and to be inhibited by the presence of a 3′-terminal CCA sequence. Here, we examined the mode of action of RNase BN on bacteriophage and bacterial tRNA precursors, particularly in light of a recent report suggesting that RNase BN removes CCA sequences (Takaku, H., and Nashimoto, M. (2008) Genes Cells 13, 1087–1097). We show that ...

  19. Contribution to the study of the release to the environment of radioactive iodine during an accident sequence type SGTR

    In a Steam Generator Tube Rupture (SGTR) accident occurring to a pressurised nuclear water reactor, a fraction of the radioactive species present in the primary circuit is likely to be transferred to the environment. Particular attention is paid to iodine for two reasons; the first one, it is well known that iodine is a high contributor to the dose at short term and in second, due to possible formation of volatile species, which could be largely sprayed in the environment. In normal operating conditions, the primary circuit is contaminated with some radioactive products flowing through micro-cracks existing in the fuel rod claddings. To better estimate the releases for SGTR sequence, it is crucial to determine the iodine partition between the gas and the liquid phase downstream the tube break as well as the droplet size distribution generated during the flashing. The first part of the PhD presents a heat and mass transfer model developed to predict the two-phase jet behaviour at the break. The steam fraction is calculated as well as the droplet size distribution upstream the break. Experiments available in the literature (tests conducted at the U.S/NRC and INERIS) are used to validate the model. The second part concerns the modelling of the iodine chemical speciation in the primary conditions (irradiation, low concentration and presence of impurities). For each iodine species, the partition coefficient has been determined either in using literature data or with the help of molecular dynamics computations. Last, this global release modelling has been implemented in ASTEC, the IRSN accident simulation software and the releases have been calculated for one SGTR scenario. (author)

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

    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)

  1. The coding sequence of amyloid-beta precursor protein APP contains a neural-specific promoter element.

    Collin, R.W.J.; Martens, G.J.M.

    2006-01-01

    The amyloid-beta precursor protein APP is generally accepted to be involved in the pathology of Alzheimer's disease. Since its physiological role is still unclear, we decided to study the function of APP via stable transgenesis in the amphibian Xenopus laevis. However, the application of constructs

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

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

    1996-12-01

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

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

    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.)

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

    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

  5. Assessment of MELCOR-1.8.5 versus different versions of SCDAP/RELAP5 MOD-3.3 with lower head creep rupture analysis of alternative accident sequences of the Three Mile Island Unit 2

    The objective of this analysis is to assess current versions of MELCOR 1.8.5-RG against 2 different versions of SCDAP/RELAP5 MOD 3.3 (SR5m33kz and SR5m33bf) with the same Henry-Fauske choked flow model. This lower head creep rupture analysis considers 2 TMI-2 (Three Mile Island - Unit 2) alternative accident sequences: (1) TMI-2 alternative accident sequence-1, and (2) TMI-2 alternative accident sequence-2. Significant findings of the analysis include: (1) the TMI-2 lower head wall fails by creep rupture with either deactivations or activations of the High Pressure Injection System; (2) for the TMI-2 alternative accident sequence-1 the time to creep rupture calculated with MELCOR 1.8.5-RG and SCDAP/RELAP5 MOD-3.3 agrees reasonably; (3) the calculation with MELCOR for the TMI-2 alternative accident sequence-1 predicts that the lower head wall failure occurred earlier than penetration failure, while MELCOR predicts the opposite for the TMI-2 alternative accident sequence-2; (4) calculation with MELCOR for TMI-2 alternative accident sequence-2 shows that when the lower head wall fails the temperature calculated with MELCOR is 1810.9 K, which exceeds the melting temperature of 1789 K for carbon steel; (5) calculations with both SR5m33kz and SR5m33bf for both TMI-2 alternative accident sequences indicate that different lower head wall locations fail rapidly one after another by a delay of a few seconds, while this is not the case for MELCOR

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

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

  7. Assessment of accident management measures on early in-vessel station blackout sequence at VVER-1000 pressurized water reactors

    Tusheva, P., E-mail: p.tusheva@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf e.V., Institute of Resource Ecology, Reactor Safety Division, POB 51 01 19, 01314 Dresden (Germany); Schäfer, F., E-mail: f.schaefer@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf e.V., Institute of Resource Ecology, Reactor Safety Division, POB 51 01 19, 01314 Dresden (Germany); Reinke, N., E-mail: nils.reinke@grs.de [Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH, Schwertnergasse 1, 50667 Cologne (Germany); Kamenov, Al., E-mail: alkamenov@npp.bg [Kozloduy NPP Plc., 3321 Kozloduy (Bulgaria); Mladenov, I., E-mail: ivanmladenov@abv.bg [Kozloduy NPP Plc., 3321 Kozloduy (Bulgaria); Kamenov, K., E-mail: k_kamenov@npp.bg [Kozloduy NPP Plc., 3321 Kozloduy (Bulgaria); Kliem, S., E-mail: s.kliem@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf e.V., Institute of Resource Ecology, Reactor Safety Division, POB 51 01 19, 01314 Dresden (Germany)

    2014-10-01

    Highlights: • Accident management procedures for a station blackout scenario are investigated. • Secondary and primary side countermeasures are compared. • In-depth analyses of the plant behaviour and estimation of time margins. • Insights into the physical phenomena which can influence the passive feeding. • Assessment of the effectiveness of the applied bleed and feed procedures. - Abstract: In the process of elaboration and evaluation of severe accident management guidelines, the assessment of the accident management measures and procedures plays an important role. This paper investigates the early in-vessel phase accident progression of a hypothetical station blackout scenario for a generic VVER-1000 pressurized water reactor. The study focuses on the following accident management measures: primary side depressurization with passive safety systems injection, secondary side depressurization with passive feeding from the feedwater system, and a combination of the both procedures. The analyses have been done with the mechanistic computer code ATHLET. The simulations give in-depth analyses of the reactor system behaviour, assessment of the time margins till heating up of the reactor core and insights into physical phenomena which can influence the passive feeding procedures for cooling of the reactor core. The simulation results show that such accident management measures can significantly prolong the time till core degradation. Maximum delay for core heat up can be achieved by sequentially realization of the secondary and primary side bleed and feed strategies. Due to reversed heat transfer in the steam generators or caused by the depressurization itself a part of the injected water is evaporated. Evaporation or flashing in the feedwater system can lead to an intermittent water injection, thus reducing the effectiveness of the feeding procedure.

  8. Assessment of accident management measures on early in-vessel station blackout sequence at VVER-1000 pressurized water reactors

    Highlights: • Accident management procedures for a station blackout scenario are investigated. • Secondary and primary side countermeasures are compared. • In-depth analyses of the plant behaviour and estimation of time margins. • Insights into the physical phenomena which can influence the passive feeding. • Assessment of the effectiveness of the applied bleed and feed procedures. - Abstract: In the process of elaboration and evaluation of severe accident management guidelines, the assessment of the accident management measures and procedures plays an important role. This paper investigates the early in-vessel phase accident progression of a hypothetical station blackout scenario for a generic VVER-1000 pressurized water reactor. The study focuses on the following accident management measures: primary side depressurization with passive safety systems injection, secondary side depressurization with passive feeding from the feedwater system, and a combination of the both procedures. The analyses have been done with the mechanistic computer code ATHLET. The simulations give in-depth analyses of the reactor system behaviour, assessment of the time margins till heating up of the reactor core and insights into physical phenomena which can influence the passive feeding procedures for cooling of the reactor core. The simulation results show that such accident management measures can significantly prolong the time till core degradation. Maximum delay for core heat up can be achieved by sequentially realization of the secondary and primary side bleed and feed strategies. Due to reversed heat transfer in the steam generators or caused by the depressurization itself a part of the injected water is evaporated. Evaporation or flashing in the feedwater system can lead to an intermittent water injection, thus reducing the effectiveness of the feeding procedure

  9. Discovery of precursor and mature microRNAs and their putative gene targets using high-throughput sequencing in pineapple (Ananas comosus var. comosus).

    Yusuf, Noor Hydayaty Md; Ong, Wen Dee; Redwan, Raimi Mohamed; Latip, Mariam Abd; Kumar, S Vijay

    2015-10-15

    MicroRNAs (miRNAs) are a class of small, endogenous non-coding RNAs that negatively regulate gene expression, resulting in the silencing of target mRNA transcripts through mRNA cleavage or translational inhibition. MiRNAs play significant roles in various biological and physiological processes in plants. However, the miRNA-mediated gene regulatory network in pineapple, the model tropical non-climacteric fruit, remains largely unexplored. Here, we report a complete list of pineapple mature miRNAs obtained from high-throughput small RNA sequencing and precursor miRNAs (pre-miRNAs) obtained from ESTs. Two small RNA libraries were constructed from pineapple fruits and leaves, respectively, using Illumina's Solexa technology. Sequence similarity analysis using miRBase revealed 579,179 reads homologous to 153 miRNAs from 41 miRNA families. In addition, a pineapple fruit transcriptome library consisting of approximately 30,000 EST contigs constructed using Solexa sequencing was used for the discovery of pre-miRNAs. In all, four pre-miRNAs were identified (MIR156, MIR399, MIR444 and MIR2673). Furthermore, the same pineapple transcriptome was used to dissect the function of the miRNAs in pineapple by predicting their putative targets in conjunction with their regulatory networks. In total, 23 metabolic pathways were found to be regulated by miRNAs in pineapple. The use of high-throughput sequencing in pineapples to unveil the presence of miRNAs and their regulatory pathways provides insight into the repertoire of miRNA regulation used exclusively in this non-climacteric model plant. PMID:26115767

  10. Accidents - Chernobyl accident

    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.)

  11. A computer code (WETBERAN) for wet sequence behavior of radioactive nuclides in LWR plant at accident conditions

    The WETBERAN code has been developed to simulate the isotopic- and time-dependent behavior fission products (FP) which leak through the multiple paths of liquid and gas flow within an LWR plant under accident conditions. In this code, emphasis is put on the phenomena pertinent to the presence of water. The TMI, SL-1, and Ginna accidents are analyzed to show the code capability. The TMI 40 day analysis gives detailed informations of FP behavior, both leaking from and remaining in the plant, and proves the effectiveness of the network model for describing the multiple leakage paths. The SL-1 analysis is made to study halogen reduction by water, which cannot be taken into account by CORRAL. The Ginna analysis has been made to check iodine transport by droplets usually generated by primary water flashing at SG tube rupture

  12. 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

    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

  13. 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

    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.

  14. 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

    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

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

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

    1994-07-01

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

  16. The Role of the Coprecipitation Sequence of Salt Precursors on the Genesis of Cu-ZnO-Al2O3 Catalysts: Synthesis, Characterization and Activity for Low Temperature Shift Reaction

    1998-01-01

    Cu-ZnO-Al2O3 catalysts for the low-temperature water-gas shift reaction were prepared using methods of direct and reverse coprecipitation. The catalysts obtained were characterized by DRX, TPR, XPS, N2O chemisorption, Hg-Porosimetry and BET surface area. It was observed that the precipitation sequence of the precursors led to significant differences in values of copper dispersion and consequently in the activity of the catalyst for the water-gas shift reaction.

  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

    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. Isolation and sequence of a cDNA encoding the precursor of a bombesinlike peptide from brain and early embryos of Xenopus laevis.

    Wechselberger, C; Kreil, G; Richter, K.

    1992-01-01

    A cDNA encoding the precursor of a bombesinlike peptide was isolated from brain of Xenopus laevis. The predicted end product resembles neuromedin B, which was originally isolated from mammalian spinal cord. The mRNA for this precursor was also present in gastrointestinal tract and in ovaries. Moreover, it could be detected in early embryos (stage 2 and stage 10) of X. laevis. These findings suggest novel roles for peptides of the bombesin family in oocyte maturation and early amphibian develo...

  19. Assessment of accident risks in the CRBRP. Volume 2. Appendices

    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.

  20. Accident management approach in Armenia

    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

  1. Accident management information needs

    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

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

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

  3. Nuclear accidents

    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

  4. Radiation accidents

    Radiation accidents may be viewed as unusual exposure event which provide possible high exposure to a few people and, in the case of nuclear plants events, low exposure to large population. A number of radiation accidents have occurred over the past 50 years, involving radiation machines, radioactive materials and uncontrolled nuclear reactors. These accidents have resulted in number of people have been exposed to a range of internal and external radiation doses and those involving radioactive materials have involved multiple routs of exposure. Some of the more important accidents involving significant radiation doses or releases of radioactive materials, including any known health effects involves in it. An analysis of the common characteristics of accidents is useful resolving overarching issues, as has been done following nuclear power, industrial radiography and medical accidents. Success in avoiding accidents and responding when they do occur requires planning in order to have adequately trained and prepared health physics organization; well defined and developed instrument program; close cooperation among radiation protection experts, local and state authorities. Focus is given to the successful avoidance of accidents and response in the events they do occur. Palomares, spain in late 1960, Goiania, Brazil in 1987, Thule, Greenland in 1968, Rocky flats, Colorado in 1957 and 1969, Three mile island, Pennsylvania in 1979, Chernobyl Ukraine in april 1986, Kyshtym, former Soviet Union in 1957, Windscale, UK in Oct. 1957 Tomsk, Russian Federation in 1993, and many others are the important examples of major radiation accidents. (author)

  5. On the application of near accident data to risk analysis of major accidents

    Major accidents are low frequency high consequence events which are not well supported by conventional statistical methods due to data scarcity. In the absence or shortage of major accident direct data, the use of partially related data of near accidentsaccident precursor data – has drawn much attention. In the present work, a methodology has been proposed based on hierarchical Bayesian analysis and accident precursor data to risk analysis of major accidents. While hierarchical Bayesian analysis facilitates incorporation of generic data into the analysis, the dependency and interaction between accident and near accident data can be encoded via a multinomial likelihood function. We applied the proposed methodology to risk analysis of offshore blowouts and demonstrated its outperformance compared to conventional approaches. - Highlights: • Probabilistic risk analysis is applied to model major accidents. • Two-stage Bayesian updating is used to generate informative distributions. • Accident precursor data are used to develop likelihood function. • A multinomial likelihood function is introduced to model dependencies among data

  6. The Role of the Coprecipitation Sequence of Salt Precursors on the Genesis of Cu-ZnO-Al2O3 Catalysts: Synthesis, Characterization and Activity for Low Temperature Shift Reaction

    R.T. Figueiredo

    1998-06-01

    Full Text Available Cu-ZnO-Al2O3 catalysts for the low-temperature water-gas shift reaction were prepared using methods of direct and reverse coprecipitation. The catalysts obtained were characterized by DRX, TPR, XPS, N2O chemisorption, Hg-Porosimetry and BET surface area. It was observed that the precipitation sequence of the precursors led to significant differences in values of copper dispersion and consequently in the activity of the catalyst for the water-gas shift reaction.

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

    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.

  8. Thermoset precursor

    This invention pertains to a distinctive thermoset precursor which is prepared by mixing a resin composition (A) which can be hardened by ionizing radiation, and a resin composition (B) which can be hardened by heat but cannot be hardened by, or is resistant to, ionizing radiation, and by coating or impregnating a molding or other substrate with a sheet or film of this mixture and irradiating this with an ionizing radiation. The principal components of composition (A) and (B) can be the following: (1) an acrylate or methacrylate and an epoxy resin and an epoxy resin hardener; (2) an unsaturated polyester resin and epoxy resin and an epoxy resin hardener; (3) a diacrylate or dimethacrylate or polyethylene glycol and an epoxy resin; (4) an epoxy acrylates or epoxy methacrylate obtained by the addition reaction of epoxy resin and acrylic or methacrylic acid

  9. Interspecific molecular comparisons between a newly sequenced conotoxin preproregion from Conus geographus and a known ω-conotoxin precursor from Conus magus: an example of essential radioisotope usage in molecular biology

    The promisingly diverse repertoire of conotoxins from the venom of the cone snails (Conus spp.) that researchers are beginning to characterize using techniques in molecular biology represents an arsenal of potential diagnostic tools for neurobiologists in the elucidation of the functioning of neuromuscular junctions and ion channels, particularly in the mammalian central nervous system (CNS). The work reported here is a contribution to the effort of depicting the diverse toxin ligands from Conus geographus and its interspecific comparison with a previously sequenced ω-conotoxin precursor molecule from C. magus. In this investigation, we obtain the cDNA sequence of the preproregion of a conotoxin from C. geographus, translate it into the amino acid sequence, characterize the protein and compare it with a known sequence using computer analysis. A sequence of the conotoxin prepropeptide C71 cDNA was determined to be 93 bp. Translation into its amino acids encodes the sequence: M K L T C V V I V A V L L L T A C Q L T Q L M T Q R Y Q Q T R. The protein code satisfies the prerequisites and characteristics of a signal and a propeptide where the functional domains are shown and discussed. Comparison of C71 peptide with MVIIA, and ω-conotoxin from C. magus, denotes important homologies and variabilities. Secondary structural motifs are also analyzed. The methods briefly described in this paper illustrate the current application of radioisotopes in molecular biology: (1) in the use of radiolabelled adenosine triphosphates (dATP) in DNA sequencing using Sanger's dideoxy method (Sanger et al., 1977), (2) in probing of cDNA clones, and (3) in gel autoadiography. (Author). 18 refs., 2 figs

  10. Accident Statistics

    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. Development of TRAIN for accident management

    Severe accident management can be defined as the use of existing and alternative resources, systems, and actions to prevent or mitigate a core-melt accident in nuclear power plants. TRAIN (Training pRogram for AMP In NPP), developed for training control room staff and the technical group, is introduced in this paper. The TRAIN composes of phenomenological knowledge base (KB), accident sequence KB and accident management procedures with AM strategy control diagrams and information needs. This TRAIN might contribute to training them by obtaining phenomenological knowledge of severe accidents, understanding plant vulnerabilities, and solving problems under high stress. (author)

  12. Worst case reactor accidents: a paradox

    The preliminary results from the application of improved source term methodology indicate a diversity of results for plants of different design, and for different accident sequences postulated for the same plant. While significant reductions from previous estimates are calculated with the new methodology for some accident scenarios, the same methodology predicts release magnitudes of minor difference from those produced with earlier methods for other accident sequences and plants. This divergence of calculated results precludes the adoption of a worst case as a meaningful characterization of severe accident consequences. This situation reinforces the need to consider the consequences of severe accidents only in light of their probability, even in those applications outside the traditional risk assessment process, and may necessitate re-consideration of a probability threshold for extremely low probability events. A practical approach to such a threshold value is discussed, based on NRC's experience with severe accident considerations in environmental impact statements

  13. The vver severe accident management

    The basic approach to the VVER safety management is based on the defence-in-depth principle the main idea of which is the multiplicity of physical barriers on the way of dangerous propagation on the one hand and the diversity of measures to protect each of them on the other hand. The main events of severe accident with loss of core cooling at NPP with WWER can be represented as a sequence of NPP states, in which each subsequent state is more severe than the previous one. The following sequence of states of the accident progression is supposed to be realistic and the most probable: -) loss of efficient core cooling; -) core melting, relocation of the molten core to the lower head and molten pool formation, -) reactor vessel damage, and -) containment damage and fission products release. The objectives of accident management at the design basis stage, the determining factors and appropriate determining parameters of processes are formulated in this paper. The same approach is used for the estimation of processes parameters at beyond design basis accident progression. The accident management goals and the determining factors and parameters are also listed in that case which is characterized by the loss of integrity of the fuel cladding. The accident management goal at the stage of core melt relocation implies the need for an efficient core-catcher

  14. Tchernobyl accident

    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

  15. Accident: Reminder

    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

  16. Aircraft Loss-of-Control Accident Analysis

    Belcastro, Christine M.; Foster, John V.

    2010-01-01

    Loss of control remains one of the largest contributors to fatal aircraft accidents worldwide. Aircraft loss-of-control accidents are complex in that they can result from numerous causal and contributing factors acting alone or (more often) in combination. Hence, there is no single intervention strategy to prevent these accidents. To gain a better understanding into aircraft loss-of-control events and possible intervention strategies, this paper presents a detailed analysis of loss-of-control accident data (predominantly from Part 121), including worst case combinations of causal and contributing factors and their sequencing. Future potential risks are also considered.

  17. Transportation accidents

    Predicting the possible consequences of transportation accidents provides a severe challenge to an analyst who must make a judgment of the likely consequences of a release event at an unpredictable time and place. Since it is impractical to try to obtain detailed knowledge of the meteorology and terrain for every potential accident location on a route or to obtain accurate descriptions of population distributions or sensitive property to be protected (data which are more likely to be more readily available when one deals with fixed-site problems), he is constrained to make conservative assumptions in response to a demanding public audience. These conservative assumptions are frequently offset by very small source terms (relative to a fixed site) created when a transport vehicle is involved in an accident. For radioactive materials, which are the principal interest of the authors, only the most elementary models have been used for assessing the consequences of release of these materials in the transportation setting. Risk analysis and environmental impact statements frequently have used the Pasquill-Gifford/gaussian techniques for releases of short duration, which are both simple and easy to apply and require a minimum amount of detailed information. However, after deciding to use such a model, the problem of selecting what specific parameters to use in specific transportation situations still presents itself. Additional complications arise because source terms are not well characterized, release rates can be variable over short and long time periods, and mechanisms by which source aerosols become entrained in air are not always obvious. Some approaches that have been used to address these problems will be reviewed with emphasis on guidelines to avoid the Worst-Case Scenario Syndrome

  18. The influence of accident measures on accident scenarios for VVER-1000-Type reactors

    For VVER-1000-type reactors severe accident scenarios and possible mitigation strategies are investigated. The Station blackout sequence is chosen as reference case. At first a comparison between the cases with and without working spray systems is discussed. Afterwards the results of a parametric study investigating the influence of different water volumes on the course of the accident are presented. It can be shown that most of these accident mitigation measures will maintain the containment integrity and reduce the source term. (author)

  19. The nature of reactor accidents

    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

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

    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. Severe accident analysis using dynamic accident progression event trees

    Hakobyan, Aram P.

    In present, the development and analysis of Accident Progression Event Trees (APETs) are performed in a manner that is computationally time consuming, difficult to reproduce and also can be phenomenologically inconsistent. One of the principal deficiencies lies in the static nature of conventional APETs. In the conventional event tree techniques, the sequence of events is pre-determined in a fixed order based on the expert judgments. The main objective of this PhD dissertation was to develop a software tool (ADAPT) for automated APET generation using the concept of dynamic event trees. As implied by the name, in dynamic event trees the order and timing of events are determined by the progression of the accident. The tool determines the branching times from a severe accident analysis code based on user specified criteria for branching. It assigns user specified probabilities to every branch, tracks the total branch probability, and truncates branches based on the given pruning/truncation rules to avoid an unmanageable number of scenarios. The function of a dynamic APET developed includes prediction of the conditions, timing, and location of containment failure or bypass leading to the release of radioactive material, and calculation of probabilities of those failures. Thus, scenarios that can potentially lead to early containment failure or bypass, such as through accident induced failure of steam generator tubes, are of particular interest. Also, the work is focused on treatment of uncertainties in severe accident phenomena such as creep rupture of major RCS components, hydrogen burn, containment failure, timing of power recovery, etc. Although the ADAPT methodology (Analysis of Dynamic Accident Progression Trees) could be applied to any severe accident analysis code, in this dissertation the approach is demonstrated by applying it to the MELCOR code [1]. A case study is presented involving station blackout with the loss of auxiliary feedwater system for a

  2. Fukushima accident study using MELCOR

    Randall O Gauntt

    2013-01-01

    The accidents at the Fukushima Daiichi nuclear power station stunned the world as the sequences played out over severals days and videos of hydrogen explosions were televised as they took place.The accidents all resulted in severe damage to the reactor cores and releases of radioactivity to the environment despite heroic measures had taken by the operating personnel.The following paper provides some background into the development of these accidents and their root causes,chief among them,the prolonged station blackout conditions that isolated the reactors from their ultimate heat sink — the ocean.The interpretations given in this paper are summarized from a recently completed report funded by the United States Department of Energy (USDOE).

  3. 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

    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

  4. Criticality Accident

    At a meeting of electric utility presidents in October, 1999, the Federation Power Companies (FEPCO) officially decided to establish a Japanese version of WANO, following the JCO criticality accident. The Japanese WANO is expected to be launched by the end of the year: initially, with some 30 private sector companies concerned with nuclear fuel. It is said that the private sector had to make efforts to ensure that safety was the most important value in management policy throughout the industry, and that comprehensive inspections would be implemented. In anything related to nuclear energy, sufficient safety checks are required even for the most seemingly trivial matters. Therefore, the All-Japan Council of Local Governments with Atomic Power Stations has already proposed to the Japanese government that it should enact the special law for nuclear emergency, providing that the unified responsibility for nuclear disaster prevention should be shifted to the national government, since the nuclear disaster was quite special from the viewpoint of its safety regulation and technical aspects. (G.K.)

  5. Persistence on airline accidents.

    L. A. GIL-ALANA; Barros, C.P. (Carlos P.); J.R. Faria

    2009-01-01

    This paper analyses airline accidents data from 1927-2006. The fractional integration methodology is adopted. It is shown that airline accidents are persistent and (fractionally) cointegrated with airline traffic. Thus, there exists an equilibrium relation between air accidents and airline traffic, with the effect of the shocks to that relationship disappearing in the long run. Policy implications are derived for countering accidents events.

  6. Persistence in Airline Accidents

    Carlos Pestana Barros; João Ricardo Faria; Luis A. Gil-Alana

    2008-01-01

    This paper analyses airline accident data from 1927-2006, through fractional integration. It is shown that airline accidents are persistent and (fractionally) cointegrated with airline traffic. There exists a negative relation between air accidents and airline traffic, with the effect of the shocks to that relationship disappearing in the long run. Policy implications are derived for countering accident events.

  7. Severe accident phenomena

    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.)

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

    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.)

  9. On requirements to environment protection under accident conditions

    Accident situation on nuclear power plant operation is considered. Definition is given of the concept of ''Accident situation'' and recommendations are made for sequence of evaluation of such a situation. Population protection measures at an accident situation are considered depended on the level of radiation hazard. Recommendations are made for functions of accident team emergency evaluation of radiation hazard in the case of accident and recommendations on composition of equipment for mobile field dosimetric groups are also done. Requirements are given for emergency measures plan for nuclear power plant and criterions for radiation hazard estimation

  10. Earthquakes: hydrogeochemical precursors

    Ingebritsen, Steven E.; Manga, Michael

    2014-01-01

    Earthquake prediction is a long-sought goal. Changes in groundwater chemistry before earthquakes in Iceland highlight a potential hydrogeochemical precursor, but such signals must be evaluated in the context of long-term, multiparametric data sets.

  11. Timing of core damage states following severe accidents for the CANDU reactor design

    This paper documents the analytical methodology used to evaluate severe accident sequences. The relevant thermal-mechanical phenomena and the mathematical approach used in calculating the timing of the accident progression are described. An example of a specific accident scenario is provided in order to illustrate the application of the severe accident progression methodology. The postulated sever accidents analyzed mainly differ in the timing to reach and progress through each defined 'core damage state'. (author)

  12. Probabilistic safety assessment precursor studies in support of the NRU research reactor life extension

    A probabilistic safety assessment program is being undertaken in support of a major upgrade planned for the NRU research reactor at the Chalk River Laboratories of Atomic Energy of Canada Limited (AECL) Research. This paper describes the objectives of the program and discusses the safety issues against which the success of the upgrade program is being judged. A licensing basis for the reactor in the form of detailed safety goals and release criteria as a function of event frequency has been proposed. An allowance has been made in the setting of the release criteria as a function of frequency such that the summing of event sequences is not a requirement to demonstrate compliance. The ability to assess each individual sequence that leads to core damage and/or fission-product release from confinement against these release criteria has, to a large extent, dictated the character of the supporting safety analyses. To identify the release frequencies in accident sequences, a probabilistic assessment comprising a number of probabilistic precursor studies is being undertaken. Descriptive event sequences are being used extensively in these studies to complement the more traditional event/fault tree techniques, to - ensure that operating and maintenance staff are involved in their preparation. Details of the program, assessment basis, and the analysis techniques being used in support of the assessment are given. (author). 4 refs, 3 figs

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

    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)

  14. Severe accident insights from the Brunswick IPE

    Miller, G.L. (Carolina Power and Light Company, Raleigh, NC (United States))

    1993-01-01

    Insights gained from the development of the level-2 analysis for a Brunswick individual plant examination (IPE) have led to severe accident insights that take advantage of the unique design of the containment structure. The Brunswick steam electric plant (BSEP) consists of two General Electric BWR-4 boiling water reactors (BWRS) with Mark I containments. The containments are unique among BWR Mark I's because the construction of the drywell and torus is reinforced concrete with steel liners. The typical Mark I is a steel shell construction. Both units are rated at 2436 MW(thermal) and [approximately]760 MW(electric). The Brunswick IPE, representing both units, was submitted to the US Nuclear Regulatory Commission in August 1992 (Ref. 1). The estimated mean core damage frequency (CDF) for the level-1 IPE is 2.7 x 10[sup [minus]5]/yr. Station blackout accident sequences contribute 66% to the overall CDF. Transient initiated sequences that involve loss of decay heat removal contribute 30% to the overall CDF. Accident sequences involving anticipated transients without scram (3%), transients with loss of high-pressure injection (I%), loss-of-coolant accidents (LOCAs) (< 1 %), and interfacing LOCAs (< 1 %) constituted the remainder of the accident sequences, which were above the analytical truncation level of 1 X 10 [sup [minus]8]/yr.

  15. SEVERE ACCIDENT MANAGEMENT TRAINING

    The purpose of this paper is (a) to define the International Atomic Energy Agency's role in the area of severe accident management training, (b) to briefly describe the status of representative severe accident analysis tools designed to support development and validation of accident management guidelines, and more recently, simulate the accident with sufficient accuracy to support the training of technical support and reactor operator staff, and (c) provide an overview of representative design-specific accident management guidelines and training. Since accident management and the development of accident management validation and training software is a rapidly evolving area, this paper is also intended to evolve as accident management guidelines and training programs are developed to meet different reactor design requirements and individual national requirements

  16. Secretion and processing of insulin precursors in yeast.

    Thim, L.; Hansen, M T; Norris, K; Hoegh, I; Boel, E; Forstrom, J; Ammerer, G; Fiil, N P

    1986-01-01

    A series of dibasic insulin precursors including proinsulin was expressed and secreted from Saccharomyces cerevisiae. Recombinant plasmids were constructed to encode fusion proteins consisting of a modified mating factor alpha 1 leader sequence and an insulin precursor. The leader sequence serves to direct the fusion protein into the secretory pathway of the cell and to expose it to the Lys-Arg processing enzyme system. The secreted peptides were purified from the fermentation broth and chara...

  17. Structure and gene organization of bovine neuromedin K precursor.

    Kotani, H.; Hoshimaru, M; Nawa, H; Nakanishi, S.

    1986-01-01

    cDNA and genomic DNA clones for the precursor of a mammalian neuropeptide tachykinin, neuromedin K, have been isolated and characterized by molecular cloning and sequence analysis. The deduced amino acid sequence indicates that the bovine neuromedin K precursor (preprotachykinin B) consists of 126 amino acid residues including a putative signal peptide. There are two preprotachykinin B mRNAs that differ only at the 5' extremity of the untranslated regions. The major mRNA species is encoded by...

  18. Traffic Congestion and Accidents

    Schrage, Andrea

    2006-01-01

    Obstructions caused by accidents can trigger or exacerbate traffic congestion. This paper derives the efficient traffic pattern for a rush hour with congestion and accidents and the corresponding road toll. Compared to the model without accidents, where the toll equals external costs imposed on drivers using the road at the same time, a new insight arises: An optimal toll also internalizes the expected increase in future congestion costs. Since accidents affect more drivers if traffic volumes...

  19. Psychology of nuclear accidents

    Tysoe, M.

    1983-03-31

    Incidents involving nuclear weapons are described, as well as the accident to the Three Mile Island-2 reactor. Methods of assessment of risks are discussed, with particular reference to subjective judgements and the possible role of human error in civil nuclear accidents. Accidents or misunderstandings in communication or human actions which might lead to nuclear war are also discussed.

  20. TMI-2 accident: core heat-up analysis

    This report summarizes NSAC study of reactor core thermal conditions during the accident at Three Mile Island, Unit 2. The study focuses primarily on the time period from core uncovery (approximately 113 minutes after turbine trip) through the initiation of sustained high pressure injection (after 202 minutes). The transient analysis is based upon established sequences of events; plant data; post-accident measurements; interpretation or indirect use of instrument responses to accident conditions

  1. Supervisor's accident investigation handbook

    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

  2. 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

    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.

  3. Framework for accident management

    Accident management is an essential element of the Nuclear Regulatory Commission (NRC) Integration Plan for the closure of severe accident issues. This element will consolidate the results from other key elements; such as the Individual Plant Examination (IPE), the Containment Performance Improvement, and the Severe Accident Research Programs, in a form that can be used to enhance the safety programs for nuclear power plants. The NRC is currently conducting an Accident Management Program that is intended to aid in defining the scope and attributes of an accident management program for nuclear power plants. The accident management plan will ensure that a plant specific program is developed and implemented to promote the most effective use of available utility resources (people and hardware) to prevent and mitigate severe accidents. Hardware changes or other plant modifications to reduce the frequency of severe accidents are not a central aim of this program. To accomplish the outlined objectives, the NRC has developed an accident management framework that is comprised of five elements: (1) accident management strategies, (2) training, (3) guidance and computational aids, (4) instrumentation, and (5) delineation of decision making responsibilities. A process for the development of an accident management program has been identified using these NRC framework elements

  4. Development of training course about Fukushima Daiichi NPPs accident

    The East Japan Great Earthquake and the resulting Tsunamis struck the Fukushima Daiichi Nuclear Power Plants on March 11, 2011, followed by the severe accidents falling into the core damage and the release of radioactive materials. This paper describes the new training program which BTC developed to help operators learn the accident sequence and lessons and the effect of countermeasures. (author)

  5. Study on severe accident mitigation measures for the development of PWR SAMG

    2006-01-01

    In the development of the Severe Accident Management Guidelines (SAMG), it is very important to choose the main severe accident sequences and verify their mitigation measures. In this article, Loss-of-Coolant Accident (LOCA), Steam Generator Tube Rupture (SGTR), Station Blackout (SBO), and Anticipated Transients without Scram (ATWS) in PWR with 300 MWe are selected as the main severe accident sequences. The core damage progressions induced by the above-mentioned sequences are analyzed using SCDAP/RELAP5. To arrest the core damage progression and mitigate the consequences of severe accidents, the measures for the severe accident management (SAM) such as feed and bleed, and depressurizations are verified using the calculation. The results suggest that implementing feed and bleed and depressurization could be an effective way to arrest the severe accident sequences in PWR.

  6. Untangling ENSO Precursors

    Pegion, K.; Alexander, M. A.

    2014-12-01

    There are several proposed precursors to the El Nino-Southern Oscillation (ENSO) that may provide the ability to predict ENSO as much as one year in advance. Some of these precursors are associated with stochastic forcing from extratropical atmospheric variability. Two examples are the seasonal footprinting mechanism (SFM) and the Pacific meridional mode (PMM). Both of these ENSO precursors are thought to be forced by the North Pacific Oscillation (NPO), a north-south sea level pressure dipole in the north Pacific. Additionally, both the PMM and SFM are thought to impact the tropics through wind evaporation SST feedbacks and have a correlation with ENSO up to one year in advance. These two precursors are discussed interchangeably throughout the literature and various indices used to define them co-mingle them. As a result, whether they are independent of each other or are part of the same process has not been investigated. The research presented is focused on untangling the relationship between the PMM, SFM, NPO, and ENSO using observational datasets and model simulations. Observational results demonstrate that these two mechanisms are different, are forced by different atmospheric circulations, and result in different manifestations of ENSO. Modeling results highlight the extent to which climate models can simulate these relationships and their impact on the simulation of ENSO.

  7. Visualization of Traffic Accidents

    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.

  8. Framework for accident management

    A program is being conducted to establish those attributes of a severe accident management plan which are necessary to assure effective response to all credible severe accidents and to develop guidance for their incorporation in a plant's Accident Management Plan. This program is one part of the Accident Management Research Program being conducted by the U. S. Nuclear Regulatory Commission (NRC). The approach used in establishing attributes and developing guidance includes three steps. In the first step the general attributes of an accident management plan were identified based on: (1) the objectives established for the NRC accident management program, (2) the elements of an accident management framework identified by the NRC, and (3) a review of the processes used in developing the currently used approach for classifying and analyzing accidents. For the second step, a process was defined that uses the general attributes identified from the first step to develop an accident management plan. The third step applied the process defined in the second step at a nuclear power plant to refine and develop it into a benchmark accident management plan. Step one is completed, step two is underway and step three has not yet begun

  9. Summary of a workshop on severe accident management for BWRs

    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. For each accident sequence and each combination of 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 instrument behavior during an accident. During the period September 26--28, 1990, a workshop was held at the University of California, Los Angeles, to address these uncertainties for Boiling Water Reactors (BWRs). This report contains a summary of the workshop proceedings

  10. The Chernobyl accident consequences

    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.)

  11. Communication and industrial accidents

    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 concept of role distance, which emphasizes the organizational characteristics. The general hypothesis is that communication failures are a main cause of role distance and accident-proneness within orga...

  12. Laser accidents: Being Prepared

    Barat, K

    2003-01-24

    The goal of the Laser Safety Officer and any laser safety program is to prevent a laser accident from occurring, in particular an injury to a person's eyes. Most laser safety courses talk about laser accidents, causes, and types of injury. The purpose of this presentation is to present a plan for safety offices and users to follow in case of accident or injury from laser radiation.

  13. Analysis of severe accidents in pressurized heavy water reactors

    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

  14. Nuclear accidents and epidemiology

    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

  15. Report of the Ad hoc Committee on the Chernobyl Accident

    The accident, which occurred on April 26 of 1986 at the fourth unit of the Chernobyl Nuclear Power Plant in the Ukrainian Soviet Socialist Republic of the Soviet Union, was the unprecedented accident in terms of, among other things, structural damages given to the reactor, an amount of radioactive materials released to the environment, and a number of casualties resulting from the accident. Investigation and analysis of the accident were conducted at JAERI by forming the Ad hoc Committee on the Chernobyl Accident within the organization under which Task Group A was responsible for the design and characteristics of the reactor and the accident sequence and Task Group B was responsible for behavior of radioactive materials and radiological consequences to the environment. The present report is the summary of the investigations and analyses which were carried out by the committee. (author)

  16. Analysis and research status of severe core damage accidents

    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)

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

    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

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

    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.

  19. SWR-1000 concept on control of severe accidents

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

  20. Computerised severe accident management aids

    The OECD Halden Reactor Project in Norway is running two development projects in the area of computerised accident management in cooperation with the Swedish nuclear plant Forsmark unit 2. Also other nuclear organisations in the Nordic countries take part in the projects. The SAS II system is installed at Forsmark and is now being validated against the plant compact simulator and is later to be installed in the plant control room. It is designed to follow all defined critical safety functions in the same manner as is done in the functionally oriented Emergency Operating Procedures. The shift supervisor thus uses SAS II as a complementary information system after a plant disturbance . The plant operators still use the ordinary instrumentation and the event oriented procedures. This gives to a high extent both redundancy and diversity in information channels and in procedures. Further, a new system is under discussion which goes a step further in accident management than SAS II. It is called the Computerised Accident Management Support (CAMS) system. The objective is to make a computerised tool that can assist both the control room crew and the technical support centre in accident mitigation, especially in the early stages of an accident where the integrity of the core still can be maintained if proper counteractions to the accident sequence are taken. In CAMS another approach is taken than in SAS II by putting the process parameters in focus. A more elaborate signal validation is proposed. The validated signals are input to models that calculates mass and energy balances of the primary system. Among parameters calculated are residual heat. Experiences from these two approaches to computerised accident management support are presented and discussed. In summary: The original project proposal aimed particularly for operator and TSC support during severe accidents. In the CAMS design proposal we have, however, promoted the SMABRE code which is not designed for such

  1. The EM Earthquake Precursor

    Jones, K. B., II; Saxton, P. T.

    2013-12-01

    Many attempts have been made to determine a sound forecasting method regarding earthquakes and warn the public in turn. Presently, the animal kingdom leads the precursor list alluding to a transmission related source. By applying the animal-based model to an electromagnetic (EM) wave model, various hypotheses were formed, but the most interesting one required the use of a magnetometer with a differing design and geometry. To date, numerous, high-end magnetometers have been in use in close proximity to fault zones for potential earthquake forecasting; however, something is still amiss. The problem still resides with what exactly is forecastable and the investigating direction of EM. After the 1989 Loma Prieta Earthquake, American earthquake investigators predetermined magnetometer use and a minimum earthquake magnitude necessary for EM detection. This action was set in motion, due to the extensive damage incurred and public outrage concerning earthquake forecasting; however, the magnetometers employed, grounded or buried, are completely subject to static and electric fields and have yet to correlate to an identifiable precursor. Secondly, there is neither a networked array for finding any epicentral locations, nor have there been any attempts to find even one. This methodology needs dismissal, because it is overly complicated, subject to continuous change, and provides no response time. As for the minimum magnitude threshold, which was set at M5, this is simply higher than what modern technological advances have gained. Detection can now be achieved at approximately M1, which greatly improves forecasting chances. A propagating precursor has now been detected in both the field and laboratory. Field antenna testing conducted outside the NE Texas town of Timpson in February, 2013, detected three strong EM sources along with numerous weaker signals. The antenna had mobility, and observations were noted for recurrence, duration, and frequency response. Next, two

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

    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

  3. Severe accident testing of electrical penetration assemblies

    Clauss, D.B. (Sandia National Labs., Albuquerque, NM (USA))

    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.

  4. Severe accident testing of electrical penetration assemblies

    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

  5. Criticality accident in Argentina

    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)

  6. Chernobyl accident and Danmark

    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)

  7. Radiation accidents in hospitals

    Some of the radiation accidents that have occurred in Indian hospitals and causes that led to them are reviewed. Proper organization of radiation safety minimizes such accidents. It has been pointed out that there must be technical competence and mental preparedness to tackle emergencies when they do infrequently occur. (M.G.B.)

  8. Communication and industrial accidents

    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 conc

  9. Chernobyl accident and Denmark

    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)

  10. Accidents with orphan sources

    The International Atomic Energy Agency has specifically defined statutory functions relating to the development of standards of safety and the provision for their application. It also has responsibilities placed on it by virtue of a number of Conventions, two of which are relevant to nuclear accidents or radiological emergencies - the Convention on Early Notification of a Nuclear Accident and the Convention on Assistance in the Case of a Nuclear Accident or Radiological Emergency. An overview of the way in which these functions are being applied to prevent and respond to radiological accidents, particularly those involving orphan sources, is described in this paper. Summaries of a number of such accidents and of the Agency's Action Plan relating to the safety and security of radiation sources are given. (orig.)

  11. Combining precursor incidents investigations and QRA in oil and gas industry

    Accident investigation is the collection and examination of facts related to an occurred specific incident. Quantitative Risk Analysis (QRA) is the systematic use of available information to identify hazards and probabilities, and to predict the possible consequences to individuals or populations, property or the environment. Traditionally both methods have been used separately; however both accident investigation and QRA describe hazards in a systematic way. The extensive research that is done related to that including human and organisational factors in QRA brings accident investigation and QRA closer together. Every year there are a large number of precursor incidents recorded with the potential to cause major accidents risks in the North Sea oil and gas industry. This article describes how accident investigation and QRA can be combined using available information from a precursor incident as input to QRA-methodology to identify hazards, probabilities, safety barriers and possible consequences. The combined method is shortened as QRA PII (Quantitative Risk Analysis Precursor Incident Investigation) and makes use of well known hazard analysis techniques to produce a more complete cause and risk picture in complex systems. This includes an extended understanding of human and organisational factors in accidents and prevention of these.

  12. Accident report 1975/76

    The statistics previously published on the development of accidents were completed. It is the purpose of this accident report: 1) to present a survey of the development of the number of accidents (no radiation accidents) for the years 1960 - 1976, 2) to break down the accidents by different characteristics in order to be able to recognize the preventive measures to be taken so as to avoid further accidents, 3) to report about accidents experienced and to indicate activities performed with respect to accident prevention and health protection. (orig.)

  13. Database on aircraft accidents

    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

  14. Accidents with sulfuric acid

    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.

  15. Persistence of airline accidents.

    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. PMID:20618386

  16. Management of severe accidents

    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)

  17. Accidents, risks and consequences

    Although the accident at Chernobyl can be considered as the worst accident in the world, it could have been worse. Other far worse situations are considered, such as a nuclear weapon hitting a nuclear reactor. Indeed the accident at Chernobyl is compared to a nuclear weapon. The consequences of Chernobyl in terms of radiation levels are discussed. Although it is believed that a similar accident could not occur in the United Kingdom, that possibility is considered. It is suggested that emergency plans should be made for just such an eventuality. Even if Chernobyl could not happen in the UK, the effects of accidents are international. The way in which nuclear reactor accidents happen is explored, taking the 1957 Windscale fire, Three Mile Island and Chernobyl as examples. Reactor designs and accident scenarios are considered. The different reactor designs are listed. As well as the Chernobyl RBMK design it is suggested that the light water reactors also have undesirable features from the point of view of safety. (U.K.)

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

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

  19. Reverse tracing of precursors and earthquake precursors in Taiwan

    WANG Li-ping; LI Yong; MA Li; ZHANG Shu-mei

    2008-01-01

    In this paper, we investigate the precursors of large earthquakes in the eastern region of Taiwan by means of the reverse tracing of precursors. We discuss the parameters which are suitable for the seismic chains and intermedi-ate-term patterns in this region and obtain the threshold of the patterns. Applying the linear discriminate method to the intermediate-term patterns of seismic chains, we present an approach for exploring the precursors of large earthquakes. The results show that this method can reduce the false alarm rate for large earthquakes in this region, and the reverse tracing of precursors can be applied to the eastern region of Taiwan.

  20. Soviet submarine accidents

    Although the Soviet Union has more submarines than the NATO navies combined, and the technological superiority of western submarines is diminishing, there is evidence that there are more accidents with Soviet submarines than with western submarine fleets. Whether this is due to inadequate crews or lower standards of maintenance and overhaul procedures is discussed. In particular, it is suggested that since the introduction of nuclear powered submarines, the Soviet submarine safety record has deteriorated. Information on Soviet submarine accidents is difficult to come by, but a list of some 23 accidents, mostly in nuclear submarines, between 1966 and 1986, has been compiled. The approximate date, class or type of submarine, the nature and location of the accident, the casualties and damage and the source of information are tabulated. (U.K.)

  1. Accident resistant transport container

    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.

  2. Talking about accidents

    It is argued that the public's emotional fear of the hypothetical, very unlikely, gigantic nuclear accident is partly caused by the nuclear industry's incorrect use of language within its own professional discussions. Improved terminology is suggested. (U.K.)

  3. Boating Accident Statistics

    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...

  4. FATAL ACCIDENT REPORTING SYSTEM (FARS)

    The Fatal Accident Reporting System (FARS) database consist of three relational tables, containing data on automobile accidents on public U.S. roads that resulted in the death of one or more people within 30 days of the accident. Truck and trailer accidents are also included.

  5. The Chernobyl accident

    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

  6. Detection of Chemical Precursors of Explosives

    Li, Jing

    2012-01-01

    Certain selected chemicals associated with terrorist activities are too unstable to be prepared in final form. These chemicals are often prepared as precursor components, to be combined at a time immediately preceding the detonation. One example is a liquid explosive, which usually requires an oxidizer, an energy source, and a chemical or physical mechanism to combine the other components. Detection of the oxidizer (e.g. H2O2) or the energy source (e.g., nitromethane) is often possible, but must be performed in a short time interval (e.g., 5 15 seconds) and in an environment with a very small concentration (e.g.,1 100 ppm), because the target chemical(s) is carried in a sealed container. These needs are met by this invention, which provides a system and associated method for detecting one or more chemical precursors (components) of a multi-component explosive compound. Different carbon nanotubes (CNTs) are loaded (by doping, impregnation, coating, or other functionalization process) for detecting of different chemical substances that are the chemical precursors, respectively, if these precursors are present in a gas to which the CNTs are exposed. After exposure to the gas, a measured electrical parameter (e.g. voltage or current that correlate to impedance, conductivity, capacitance, inductance, etc.) changes with time and concentration in a predictable manner if a selected chemical precursor is present, and will approach an asymptotic value promptly after exposure to the precursor. The measured voltage or current are compared with one or more sequences of their reference values for one or more known target precursor molecules, and a most probable concentration value is estimated for each one, two, or more target molecules. An error value is computed, based on differences of voltage or current for the measured and reference values, using the most probable concentration values. Where the error value is less than a threshold, the system concludes that the target

  7. Accident and emergency management

    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)

  8. Synergy effect in accident simulation

    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

  9. Generation of Nonlinear Vortex Precursors

    Chen, Yue-Yue; Feng, Xun-Li; Liu, Chengpu

    2016-07-01

    We numerically study the propagation of a few-cycle pulse carrying orbital angular momentum (OAM) through a dense atomic system. Nonlinear precursors consisting of high-order vortex harmonics are generated in the transmitted field due to carrier effects associated with ultrafast Bloch oscillation. The nonlinear precursors survive to propagation effects and are well separated with the main pulse, which provides a straightforward way to measure precursors. By virtue of carrying high-order OAM, the obtained vortex precursors as information carriers have potential applications in optical information and communication fields where controllable loss, large information-carrying capacity, and high speed communication are required.

  10. Generation of nonlinear vortex precursors

    Chen, Yue-Yue; Liu, Chengpu

    2016-01-01

    We numerically study the propagation of a few-cycle pulse carrying orbital angular momentum (OAM) through a dense atomic system. Nonlinear precursors consisting of high-order vortex har- monics are generated in the transmitted field due to ultrafast Bloch oscillation. The nonlinear precursors survive to propagation effects and are well separated with the main pulse, which provide a straightforward way of measuring precursors. By the virtue of carrying high-order OAM, the obtained vortex precursors as information carriers have potential applications in optical informa- tion and communication fields where controllable loss, large information-carrying capacity and high speed communication are required.

  11. Historical aspects of radiation accidents

    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

  12. Cyclical Fluctuations in Workplace Accidents

    Boone, J.; van Ours, J.C.

    2002-01-01

    This Paper presents a theory and an empirical investigation on cyclical fluctuations in workplace accidents. The theory is based on the idea that reporting an accident dents the reputation of a worker and raises the probability that he is fired. Therefore a country with a high or an increasing unemployment rate has a low (reported) workplace accident rate. The empirical investigation concerns workplace accidents in OECD countries. The analysis confirms that workplace accident rates are invers...

  13. CANDU safety under severe accidents

    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

  14. Investigation of accident management strategies for VVER-1000-Type reactors

    The goal of this work is the search for an optimal accident management strategy to prevent containment failure and to stop the core/concrete interaction from hindering cavity bottom melt-through on the one hand and from ending the ex-vessel source term increase on the other hand, i.e., to terminate the accident. The work is based on the results of previous studies of physical and chemical phenomena during different accident scenarios for VVER-1000-type reactors. For a TMLB' sequence (an accident caused by a transient in which core melt occurs because the electric power cannot be restored before the pressure vessel melts through), a number of calculations were performed using the source term code package (STCP) to investigate the influence of several accident management measures on the core/concrete interaction and the containment integrity

  15. [Psychogenesis of accidents].

    Giannattasio, E; Nencini, R; Nicolosi, N

    1988-01-01

    After having carried out a historical review of industrial psychology with specific attention to the evolution of the concept of causality in accidents, the Authors formulate their work hypothesis from that research which take into highest consideration the executives' attitudes in the genesis of the accidents. As dogmatism appears to be one of the most negative of executives' attitudes, the Authors administered Rockeach's Scale to 130 intermediate executives from 6 industries in Latium and observed the frequency index for accidents and the morbidity index (absenteeism) of the 2149 workhand. The Authors assumed that to high degree of dogmatism on the executives' side should correspond o a higher level of accidents and absenteeism among the staff. The data processing revealed that, due to the type of machinery employed, three of the industries examined should be considered as High Risk Industrie (HRI), while the remaining three could be considered as Low Risk Industries (LRI): in fact, due to the different working conditions, a significant lower number of accidents occurred in last the three. A statistically significant correlation between the executives' dogmatism and the number of accidents among their workhand in the HRI has been noticed, while this has not been observed in the LRI. This confirms, as had already been pointed out by Gemelli in 1944, that some "objective conditions" are requested so that the accident may actually take place. On the other hand the morbidity index has not shown any difference related to the different kind of industries (HRI, LRI): in both cases statistically significant correlations were obtained between the executives' dogmatism and the staff's absenteeism. absenteeism.(ABSTRACT TRUNCATED AT 250 WORDS) PMID:3154344

  16. Accidents in nuclear ships

    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).

  17. Accidents in nuclear ships

    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)

  18. Methodology for reference accidents definition for ITER

    The safety assessment of ITER presented in the Preliminary safety report (RPrS) for the French Regulator is based on the definition and the study of a limited set of postulated incidental and accidental sequences conservatively selected on deterministic grounds. Ultimate safety margins have also been analysed through hypothetical sequences conservatively extrapolated from the more significant accidents. The rationale for event selection consists firstly of the identification of every radiological source and its confinement barriers; failure of one or several of these barriers may then be presumed and a scenario defined, following a standardized grid; furthermore, the calculations and analysis follow a unique logical scheme to assure consistency and exhaustiveness of the report. Nine accident families have been defined: plasma events, loss of power events, in-vessel events, ex-vessel events, cryostat events, magnet events, maintenance events, tritium plant and fuel events, hot cells events. Calculations with qualified computer codes have shown that the consequences of any postulated accident respect limits defined in the safety guidelines. Moreover, no hypothetical sequence shows any cliff effect, illustrating in this way the robustness of the defence in depth approach used for ITER. It is worth to recall that, in the present status of the ITER project, the only systems subject to the analysis are the tokamak, the tritium plant and the hot cells. (author)

  19. Scoping accident(s) for emergency planning

    At the request of the Conference of State Radiation Control Program Director's (CRCPD), in November 1976 the U.S. Nuclear Regulatory Commission formed a joint Task Force with representatives of the U.S. Environmental Protection Agency to answer a number of questions posed by the States regarding emergency planning. This Task Force held monthly meetings through November 1977. In December 1977 a draft report was prepared for limited distribution for review and comment by selected State and local organizations. The NRC/EPA Task Force deliberations centered on the CRCPD request for '... a determination of the most severe accident basis for which radiological emergency response plans should be developed by offsite agencies...' in the vicinity of nuclear power plants. Federal Interagency guidance to the States in this regard has been that the scoping accident should be the most serious conservatively analyzed accident considered for siting purposes, as exemplified in the Commission's Regulations at 10 CFR Part 100 and the NRC staffs Regulatory Guides 1.3 and 1.4, and as presented in license applicant's Safety Analysis Reports and the USNRC Staffs Safety Evaluation Reports. The draft report of the Task Force amplifies on this recommendation: to present a clearer picture of its import and introduces the concept of protective action zones (PAZs) within which detailed emergency plans should be developed; one zone for the plume exposure pathway and a second, larger zone for contamination pathways. The time dependence of potential releases and atmospheric transport, and important radionuclide groups of possible import are also discussed in the draft Task Force report. A status report regarding this effort, as of June 1978, will be presented. (author)

  20. Analysis of the TMI-2 accident using ATHLET-CD

    One analyzed the simulation of the TMI-2 NPP accident making use of the ATHLET-CD code. One describes the accident sequence, the code structure and performs the comparative analysis of the calculated and the measured data. Simulation of thermohydraulic characteristics was a special success. Application of the codes promotes the NPP optimization, the reactor safety improvement and the risk reduction. The ATHLET-CD system ( the thermohydraulic analysis of leaks and transient processes at the reactor core disruption) will allow to evaluate the adequacy of the models included in the available codes to calculate severe accidents

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

    The behavior of directly electrical heated UO2 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

  2. Fusion of the upstream vpu sequences to the env of simian human immunodeficiency virus (SHIVKU-1bMC33) results in the synthesis of two envelope precursor proteins, increased numbers of virus particles associated with the cell surface and is pathogenic for pig-tailed macaques

    Previous studies have shown that the gene coding for the Vpu protein of the human immunodeficiency virus type 1 (HIV-1) is 5' to the env gene, is in a different reading frame, and overlaps the env by 90 nucleotides. In this study, we examined the processing of the Env protein as well as the maturation and infectivity of a virus (SHIVVpenv) in which a single nucleotide was removed at the vpu-env junction, fusing the first 162 bases of vpu to the env ORF. Pulse-chase analysis revealed that SHIVVpenv-infected cells gave rise to two precursor glycoprotein species (gp160 and gp175). Immune precipitation results also revealed that an anti-Vpu serum could immune precipitate the gp175 precursor, suggesting that the amino-terminal Vpu sequence was fused to the Env protein. Growth curves revealed that the SHIVVpenv-inoculated cultures released approximately three times more p27 into the culture medium than parental SHIVKU-1bMC33. Electron microscopy revealed that while both viruses matured at the cell plasma membrane, significantly higher quantities of virus particles were cell associated on SHIVVpenv-infected cells compared to cultures inoculated with parental SHIVKU-1bMC33. Furthermore, virus was observed maturing into intracellular vesicles of SHIVVpenv-infected cells. To assess the pathogenicity of SHIVVpenv, three pig-tailed macaques were inoculated with the SHIVVpenv and monitored for 6 months for CD4+ T cell levels, viral loads, and the stability of the deletion at the vpu-env junction. Our results indicated that SHIVVpenv caused a severe CD4+ T cell loss in all three macaques within weeks of inoculation. Sequence analysis of the vpu gene analyzed from sequential PBMC samples derived from macaques revealed that this mutation was stable during the period of rapid CD4+ T cell loss. Sequence analysis showed that with increasing time of infection, the one base pair deletion was repaired in all three macaques inoculated with SHIVVpenv with the reversion occurring at 10

  3. Important severe accident research issues after Fukushima accident

    After the Fukushima accident several investigation committees issued reports with lessons learned from the accident in Japan. Among those lessons, several recommendations have been made on severe accident research. Similar to the EURSAFE efforts under EU Program, review of specific severe accident research items was started before Fukushima accident in working group of Atomic Energy Society of Japan (AESJ) in terms of significance of consequences, uncertainties of phenomena and maturity of assessment methodology. Re-investigation has been started since the Fukushima accident. Additional effects of Fukushima accident, such as core degradation behaviors, sea water injection, containment failure/leakage and re-criticality have been covered. The review results are categorized in ten major fields; core degradation behavior, core melt coolability/retention in containment vessel, function of containment vessel, source term, hydrogen behavior, fuel-coolant interaction, molten core concrete interaction, direct containment heating, recriticality and instrumentation in severe accident conditions. Based on these activities and also author's personal view, the present paper describes the perspective of important severe accident research issues after Fukushima accident. Those are specifically investigation of damaged core and components, advanced severe accident analysis capabilities and associated experimental investigations, development of reliable passive cooling system for core/containment, analysis of hydrogen behavior and investigation of hydrogen measures, enhancement of removal function of radioactive materials of containment venting, advanced instrumentation for the diagnosis of severe accident and assessment of advanced containment design which excludes long-term evacuation in any severe accident situations. (author)

  4. Severe accident aerosol research in Finland

    The retention of fission products in the steam generator tubing and in the secondary side is poorly understood at the moment. Most experimental programs have concentrated on the initial stages of deposition. Much less attention has been paid to the situations when deposition-resuspension-revaporisation are important as the deposit layers are getting thicker. The understanding of fission product deposition in realistic steam generator conditions is needed to design efficient accident management procedures. For example if there is large deposition already in the ruptured pipe(s), the accident management procedure is different from the case where most deposition would occur in the secondary side. This is considered very important because steam generator tube rupture sequences are included in the risk dominant sequences. Aerosol deposition has been studied widely in laboratory scale. However, most of the studies have concentrated on situations where the deposit layer is thin and do not significantly affect the process. In severe accident applications the most important deposition studies have been LACE, STORM, TUBA, TRANSAT and AIDA programmes. None of these tests considered steam generator conditions. Thus we can say that there is basic knowledge on aerosol deposition and removal from gas streams in water pools, but it can not be applied directly to steam generator tube rupture cases. At the moment the effectiveness of such accident management procedures as secondary side flooding can not be verified as there is no experimental data and the models in severe accident codes are poor or non-existing. As a results of this work we will get data on deposition in the tubing, in the break location and in the secondary side. Experiments will be performed in horizontal steam generators (VVER reactors). (orig.)

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

    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

  6. Severe Accident Management Strategy for EU-APR1400

    In EU-APR1400, the dedicated instrumentation and mitigation features for SAM are being developed to keep the integrity of containment and to prevent the uncontrolled release of fission products. In this paper, SAM strategy for EU-APR1400 was introduced in stages. It is still under development and finally the Severe Accident Management Guidance will be completed based on this SAM Strategy. Severe accidents in a nuclear power plant are defined as certain unlikely event sequences involving significant core damage with the potential to lead to significant releases according to EUR 2.1.4.4. Even though the probability of severe accidents is extremely low, the radiation release may cause serious effect on people as well as environment. Severe Accident Management (SAM) encompasses those actions which could be considered in recovering from a severe accident and preventing or mitigating the release of fission products to the environment. Whether those actions are successful or not, depending on a progression status of a severe accident to mitigate the consequences of severe accident phenomena to limit the release of radioactive materials keeping the leak tightness of the Primary Containment, and finally to restore transient severe accident progression into a controlled and safe states

  7. Applicability of health physics lessons learned from the Three Mile Island Unit 2 accident to the Fukushima Daiichi accident

    The TMI-2 and Fukushima Daiichi accidents appear to be dissimilar because they involve different reactor types. However, the health physics related lessons learned from TMI-2 are applicable, and can enhance the Fukushima Daiichi recovery effort. - Highlights: ► TMI-2 health physics lessons learned are applicable to Fukushima Daiichi Accident. ► Fuel damage sequence of these accidents are similar. ► On-site recovery actions will be similar, but Fukushima Daiichi is more demanding. ► Offsite recovery actions are significantly more challenging at Fukushima Daiichi.

  8. Radiation accidents and dosimetry

    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.)

  9. Helicopter accident survivability.

    Vyrnwy-Jones, P; Thornton, R

    1984-10-01

    Army Air Corps accident and fatality rates have now reached levels which compare favourably with data from other civilian and military sources. This improvement is the result of enhanced helicopter design and parallel progress in aircrew training. The introduction of new generations of turbine powered rotor craft has largely eliminated mechanical failure as the cause of accident. As a result 75% of Army Air Corps accidents are due to pilot error. This contribution is likely to increase in the future as the pilot's task is made more difficult by the incumberance of personal equipment. Methods whereby occupant protection and aircraft crashworthiness can be improved are reviewed and it is concluded that it would make sound economic sense to implement some of these well proven design features. PMID:6527344

  10. Information at radiation accidents

    This study was undertaken in order to plan an information strategy for possible future accidents involving radioactivity. Six health visitors and six farmers working in the districts of Norway which received the largest amounts of fallout from the Chernobyl accident, were interviewed. The questions were intended to give an indication of their knowledge about radioactivity and radiation, as well as their needs for information in case of a future accident. The results indicate a relatively low educational background in radiation physics and risk estimation. On the other hand the two groups showed a remarkable skill and interest in doing their own evaluation on the background of information that was linked to their daily life. It is suggested that planning of information in this field is done in close cooperation with the potential users of the information