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Sample records for accident management measures

  1. Effectiveness of selected accident management measures

    The spectrum of application of accident management measures and the boundary conditions for their performance are discussed. An assessment is made of the feasibility and effectiveness of selected possibilities of intervention for both types of light water reactors. Detailed descriptions are given of accident management measures (bleed and feed) on the secondary and on the primary side. Investigations have revealed that West German light water reactors have a great safety potential by flexible applicaton of the existing systems for controlling events which exceed the design basis. (orig./HP)

  2. Severe Accident Management Measures Introduced in Belgian NPP's

    In response to the Belgian Safety Authorities' request to address the severe accident issue within a decennial safety review, Tractebel, on behalf of the Belgian Utility, Electrabel, examined in detail specific severe accident topics and provided the Utility with several measures that could be implemented to reduce the risk associated with beyond-design accidents. The present paper summarizes the key elements of the approach applied in Belgium: - Presentation of plant-specific studies related to severe accident issues; - Use of PSA results; - Inputs of international R and D projects; - Selection and justification of severe accident measures; - Comparative study between possible mitigative measures; - Definition and justification of implemented severe accident management strategies. The vulnerability to severe accidents as well as the potential causes of containment failures have been identified leading to the study of possible countermeasures taking into account the combination of conservative design and post-TMI measures already implemented . A section of the paper will also be devoted to the specific study made for the selection, the sizing and the implementation of hydrogen control means. After the description of the selected measures implemented, the paper also describes the content of the 'Severe Accident Management Guidelines' developed by Tractebel for the Tihange NPPs and for the Doel NPPs. This project aimed at providing the operators with procedures or guidelines enabling to deal with complex situations not formally considered in the standard Emergency Response Guidelines, including accidents in which a significant portion of the core melts. The objective of these SAMG's programs is to indicate actions that must bring the plant to a controlled stable state and, above all, mitigate any challenges to the fission product barriers. The plant personnel must use the available plant information to determine the best severe accident management measures. Obviously

  3. Investigation on accident management measures for VVER-1000 reactors

    A consequence of a total loss of AC power supply (station blackout) leading to unavailability of major active safety systems which could not perform their safety functions is that the safety criteria ensuring a secure operation of the nuclear power plant would be violated and a consequent core heat-up with possible core degradation would occur. Currently, a study which examines the thermal-hydraulic behaviour of the plant during the early phase of the scenario is being performed. This paper focuses on the possibilities for delay or mitigation of the accident sequence to progress into a severe one by applying Accident Management Measures (AMM). The strategy 'Primary circuit depressurization' as a basic strategy, which is realized in the management of severe accidents is being investigated. By reducing the load over the vessel under severe accident conditions, prerequisites for maintaining the integrity of the primary circuit are being created. The time-margins for operators' intervention as key issues are being also assessed. The task is accomplished by applying the GRS thermal-hydraulic system code ATHLET. In addition, a comparative analysis of the accident progression for a station blackout event for both a reference German PWR and a reference VVER-1000, taking into account the plant specifics, is being performed. (authors)

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

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

  5. EC-sponsored research activities on accident management measures

    The European Commission (EC) is currently funding, via the 1994-1998 R and D Framework Programme, a number of activities in the field of Nuclear Fission Safety (NFS), and particularly in several areas related to 'Reactor Safety Severe Accidents'. This programme continues the research activities of the previous Community Reactor Safety Programme which was carried out as a Reinforced Concerted Action (RCA) during the period 1992-1995. The group of multi-partners projects selected for financial support from the EC under Area B.5.1 of the current NFS Programme, 'Supporting Activities / Accident Management Measures' (known as the 'AMM' cluster) are basically aiming at implementing the results of severe accident research into practical Accident Management (AM) strategies. The generic objective is to exchange information and to develop a common European approach regarding aspects such as phenomena related uncertainties, possible adverse effects of operator actions on the progression of the accident, interpretation of measurements, equipment performance, instrument survival and human error under stress. This paper briefly discusses the objectives and achievements of a completed project of the 1992-1995 RCA, known as 'Accident Management Support' ('AMS'), and also presents the current status of an on-going project of the 1994-1998 NFS Programme, 'Algorithm support for accident identification and Critical safety Functions signal validation' ('ASIA'). The objectives of the 'AMS' project were (i) to define, investigate and develop means and methods to provide reliable information and diagnostics, as well as support tools for accident management, and (ii) investigate the different signal validation methodologies with emphasis on the existing instrumentation rather than on new instrumentation needs. The work started with the writing of two state-of-the-art reports (SOARs) in these two areas. In parallel to the compilation of the SOARs, and later in a second phase, specific

  6. Experimental and analytical verification of accident management measures

    Two complementary test facilities - the Upper Plenum Test Facility and the ''Primaerkreislauf'' test facility were constructed to investigate the thermal hydraulic response of a pressurized water reactor during postulated accidents. The general objective of the experimental programs is to contribute to a better understanding of accident sequences and to provide a detailed data base for the validation of computer codes, i.e. ATHLET and RELAP, the latter being used by Siemens/KWU for reactor safety analyses. A major target of the recent experimental programs has been the verification of accident management procedures, such as secondary and/or primary side bleed-and-feed. The experimental results demonstrate that secondary side bleed-and-feed is a very effective method for removing decay heat without contaminating the containment. Primary side bleed-and-feed was also shown to be a highly effective measure to ensure core cooling under beyond-design-basis conditions. This publication presents results from experiments at the Upper Plenum Test Facility and the ''Primaerkreislauf'' test facility as well as from corresponding RELAP 5/Mod 2 analyses. (orig.)

  7. Modeling and measuring the effects of imprecision in accident management

    This paper presents two approaches for evaluating the uncertainties inherent in accident management strategies. Current PRA methodology uses expert opinion in the assessment of rare event probabilities. The problem is that these probabilities may be difficult to estimate even though reasonable engineering judgement is applied. This occurs because expert opinion under incomplete knowledge and limited data is inherently imprecise. In this case, the concept of uncertainty about a probability value is both intuitively appealing and potentially useful. This analysis considers accident management as a decision problem (i.e. 'applying a strategy' vs. 'do nothing') and uses an influence diagram. Then, the analysis applies two approaches to evaluating imprecise node probabilities in the influence diagram: 'a fuzzy probability' and 'an interval-valued subjective probability'. For the propagation of subjective probabilities, the analysis uses a Monte-Carlo simulation approach. In case of fuzzy probabilities, fuzzy logic is applied to propagate them. We believe that these approaches can allow us to better understand uncertainties associated with severe accident management strategies, because they provide additional information regarding the implications of using imprecise input data

  8. 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)

  9. 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

  10. Accident management measures. Demand for action as seen by the supervising authority

    The various measures taken for accident management in the plant are to be classified into categories of nuclear law, as there are: prevention of hazards, prevention of risks, or non-preventive measures ( management of remaining risk). Screening the various measures for classification shows that most of them belong to the category of preventive action under the Atomic Energy Act. This means that these measures have to be addressed in KTA safety standards. (orig./HP)

  11. 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

  12. Measures against nuclear accidents

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

  13. 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

  14. 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

  15. Analysis simulator, a tool for the evaluation of accident management measures

    The analysis simulator is a manifold and variable engineered tool which permits the interactive handling of very comprehensive model codes and offers the wealth of information calculated by the models in a condensed and uncluttered way by means of graphic displays. The first phase of work on the simulator concentrated on the development of interfaces, interactivity and communication. The experience gathered so far and the case study, in which an accident management measure is taken to prevent a severe accident, show both the advantages of the analysis simulator and its limitations as far as the speed of simulation, its sturdiness and the extent of the models are concerned. The continuation of work on the analysis simulator and the test control room will further extend these limits in order to fully comply with the requirements for the simulation of measures oriented towards certain aims of protection. (orig.)

  16. Proceedings of the workshop on the implementation of severe accident management measures

    The OECD/NEA Workshop on the Implementation of Severe Accident Management (SAM) Measures was hosted by the PSI (Paul Schemer Institut), by two Swiss Utilities (Kernkraftwerk Beznau and Kernkraftwerk Leibstadt), and by Electricite de France. Eighty specialists from fourteen OECD Member countries attended the meeting, as well as specialists from three non-Member economies and the European Commission. Thirty-three papers were presented in four sessions, preceded by a brief Introductory Session (two invited papers) and followed by a General Discussion. The objectives of the meeting were: 1) to exchange information on activities in the area of SAM implementation and on the rationale for such actions, 2) to monitor progress made, 3) to identify cases of agreement or disagreement, 4) to discuss future orientations of work, 5) to make recommendations to the CSNI. Session summaries prepared by the Chairpersons and discussed by the whole writing group are given in Annex. During the first session, 'SAM Programmes Implementation', papers from one regulator and several utilities and national research institutes were presented to outline the status of implementation of SAM programmes in countries like Switzerland, Russia, Spain, Finland, Belgium and Korea. Also, the contribution of SAM to the safety of Japanese plants (in terms of core damage frequency) was quantified in a paper. One paper gave an overview on the situation regarding SAM implementation in Europe. The second session, 'SAM Approach', provided background and bases for Severe Accident Management in countries like Sweden, Japan, Germany and Switzerland, as well as for hardware features in advanced light water reactor designs, such as the European Pressurised Reactor (EPR), regarding Severe Accident Management. The third session, 'SAM Mitigation Measures', was about hardware measures, in particular those oriented towards hydrogen mitigation where fundamentally different approaches have been taken in Scandinavian

  17. The management of accidents

    R. B. Ward

    2009-01-01

    Full Text Available Purpose: This author’s experiences in investigating well over a hundred accident occurrences has led to questioning how such events can be managed - - - while immediately recognising that the idea of managing accidents is an oxymoron, we don’t want to manage them, we don’t want not to manage them, what we desire is not to have to manage not-them, that is, manage matters so they don’t happen and then we don’t have to manage the consequences.Design/methodology/approach: The research will begin by defining some common classes of accidents in manufacturing industry, with examples taken from cases investigated, and by working backwards (too late, of course show how those involved could have managed these sample events so they didn’t happen, finishing with the question whether any of that can be applied to other situations.Findings: As shown that the management actions needed to prevent accidents are control of design and application of technology, and control and integration of people.Research limitations/implications: This paper has shown in some of the examples provided, management actions have been know to lead to accidents being committed by others, lower in the organization.Originality/value: Today’s management activities involve, generally, the use of technology in many forms, varying from simple tools (such as knives to the use of heavy equipment, electric power, and explosives. Against these we commit, in control of those items, the comparatively frail human mind and body, which, again generally, does succeed in controlling these resources, with (another generality by appropriate management. However, sometimes the control slips and an accident occurs.

  18. Workshop proceedings of ISAMM 2009: Implementation of severe accident management measures

    Guentay, S. (ed.) [Paul Scherrer Institute (PSI), Nuclear Energy and Safety Research Department, Laboratory for Thermal Hydraulics, ViIligen (Switzerland)

    2010-10-15

    This comprehensive report published by the Paul Scherrer Institute (PSI) in Switzerland reports on a conference and workshop held in Switzerland in October 2009 dealing with Severe Accidents Management (SAM) in nuclear power stations. The workshop provided an update on the status of severe accident management measures and their implications since the OECD/CSNI workshop held in 2001 at the PSI in Switzerland. Since the 2001 workshop, additional work has been performed to integrate emergency procedures and SAM measures into risk assessments in order to better reflect operator responses to recover a plant from a damaged state. The major focus of the workshop was to address SAM measures for both operational plants and new plant designs. Also, the integration of SAM measures into contemporary/future probabilistic risk assessments was discussed. 41 papers were presented in 8 sessions. The papers addressed the following areas: 1) Current status and insights of SAM (2 sessions); 2) Probabilistic Safety Assessment (PSA) modelling issues; 3) code analysis for supporting Serious Accident Management Guidance (SAMG, 2 sessions); 4) decision making, tools, training, risk-targets and entrance to SAM; 5) design modifications for implementation of SAM; 6) physical phenomena. The last part of the workshop was devoted to the presentation of the most striking highlights of the papers in the above areas, followed by two panellists giving presentations on human and organisational aspects of SAM, their importance in relation to technical issues and the effectiveness of current SAMG implementation. The question of how consequence analyses can be used to improve the effectiveness of SAM is discussed. The contributions were presented by representatives from Austria, Germany, Japan, France, the USA, Korea, Switzerland, Finland, Hungary, Belgium, Canada, Sweden, the Czech republic, the United kingdom, the Netherlands, Spain, Slovenia and Russia. The authors state that the overall picture

  19. Workshop proceedings of ISAMM 2009: Implementation of severe accident management measures

    This comprehensive report published by the Paul Scherrer Institute (PSI) in Switzerland reports on a conference and workshop held in Switzerland in October 2009 dealing with Severe Accidents Management (SAM) in nuclear power stations. The workshop provided an update on the status of severe accident management measures and their implications since the OECD/CSNI workshop held in 2001 at the PSI in Switzerland. Since the 2001 workshop, additional work has been performed to integrate emergency procedures and SAM measures into risk assessments in order to better reflect operator responses to recover a plant from a damaged state. The major focus of the workshop was to address SAM measures for both operational plants and new plant designs. Also, the integration of SAM measures into contemporary/future probabilistic risk assessments was discussed. 41 papers were presented in 8 sessions. The papers addressed the following areas: 1) Current status and insights of SAM (2 sessions); 2) Probabilistic Safety Assessment (PSA) modelling issues; 3) code analysis for supporting Serious Accident Management Guidance (SAMG, 2 sessions); 4) decision making, tools, training, risk-targets and entrance to SAM; 5) design modifications for implementation of SAM; 6) physical phenomena. The last part of the workshop was devoted to the presentation of the most striking highlights of the papers in the above areas, followed by two panellists giving presentations on human and organisational aspects of SAM, their importance in relation to technical issues and the effectiveness of current SAMG implementation. The question of how consequence analyses can be used to improve the effectiveness of SAM is discussed. The contributions were presented by representatives from Austria, Germany, Japan, France, the USA, Korea, Switzerland, Finland, Hungary, Belgium, Canada, Sweden, the Czech republic, the United kingdom, the Netherlands, Spain, Slovenia and Russia. The authors state that the overall picture

  20. 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)

  1. Applicability of Phebus FP results to severe accident safety evaluations and management measures

    The international Phebus FP (Fission Product) programme is the largest research programme in the world investigating core degradation and radioactive product release should a core meltdown accident occur in a light water reactor plant. Three integral experiments have already been performed. The experimental database obtained so far contains a wealth of information to validate the computer codes used for safety and accident management assessment

  2. Opportunities for international cooperation in nuclear accident preparedness and management: Procedural and organizational measures

    In this paper we address a difficult problem: How can we create and maintain preparedness for nuclear accidents? Our research has shown that this can be broken down into two questions: (1) How can we maintain the resources and expertise necessary to manage an accident once it occurs? and (2) How can we develop plans that will help in actually managing an accident once it occurs? It is apparently beyond the means of ordinary human organizations to maintain the capability to respond to a rare event. (A rare event is defined as something like an accident that only happens once every five years or so, somewhere in the world.) Other more immediate pressures tend to capture the resources that should, in a cost/benefit sense, be devoted to maintaining the capability. This paper demonstrates that some of the important factors behind that phenomenon can be mitigated by an international body that promotes and enforces preparedness. Therefore this problem provides a unique opportunity for international cooperation: an international organization promoting and enforcing preparedness could help save us from our own organizational failings. Developing useful accident management plans can be viewed as a human performance problem. It can be restated: how can we support and off-load the accident managers so that their tasks are more feasible? This question reveals the decision analytic perspective of this paper. That is, we look at the problem managing a nuclear accident by focusing on the decision makers, the accident managers: how do we create a decision frame for the accident managers to best help them manage? The decision frame is outlined and discussed. 9 refs

  3. 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

  4. European Union research in safety of LWRs with emphasis on accident management measures

    On April 26th 1994 the European Union (EU) adopted via a Council Decision a multiannual programme for community activities in the field of nuclear research and training for the period 1994 to 1998. This programme continued the EU research activities of the 1992-1995 Reactor Safety Programme which was carried out as a Reinforced Concerted Action (RCA), and which covered mainly research activities in the area of severe accident phenomena, both for the existing and next-generation light water reactors. The 1994-1998 Framework programme includes activities regarding Research and Technological Development (R and TD), such as demonstration projects, international cooperation, dissemination and optimization of results, as well as training, in a wide range of scientific fields, including nuclear fission safety and controlled thermonuclear fusion. The 1994-1998 specific programme for nuclear fission safety has five main activity areas: (i) Exploring Innovative Approaches, (ii) Reactor Safety, (iii) Radioactive Waste Management, Disposal, and Decommissioning, (iv) Radiological Impact on Man and Environment, and (v) Mastering Events of the past. The specific topics included in this work programme were chosen in consultation with the EU Joint Research Centres (JRC), and with experts in the different fields taking into account the needs of the end users of the Community research, i.e. vendors, utilities and licensing and regulators authorities. This paper briefly discusses the objectives and achievements of the 1992-1995 RCA and also describes the projects being (or to be) implemented as part of the 1994-1995 programme in the areas of Reactor Safety/Severe Accidents, particularly those related to Accident Management (AM) Measures. In addition to this, some relevant projects related to AM which have been funded via independent PHARE/TACIS assistance programmes will also be mentioned

  5. Accident management insights after the Fukushima Daiichi NPP accident

    The Fukushima Daiichi nuclear power plant (NPP) accident, that took place on 11 March 2011, initiated a significant number of activities at the national and international levels to reassess the safety of existing NPPs, evaluate the sufficiency of technical means and administrative measures available for emergency response, and develop recommendations for increasing the robustness of NPPs to withstand extreme external events and beyond design basis accidents. The OECD Nuclear Energy Agency (NEA) is working closely with its member and partner countries to examine the causes of the accident and to identify lessons learnt with a view to the appropriate follow-up actions to be taken by the nuclear safety community. Accident management is a priority area of work for the NEA to address lessons being learnt from the accident at the Fukushima Daiichi NPP following the recommendations of Committee on Nuclear Regulatory Activities (CNRA), Committee on the Safety of Nuclear Installations (CSNI), and Committee on Radiation Protection and Public Health (CRPPH). Considering the importance of these issues, the CNRA authorised the formation of a task group on accident management (TGAM) in June 2012 to review the regulatory framework for accident management following the Fukushima Daiichi NPP accident. The task group was requested to assess the NEA member countries needs and challenges in light of the accident from a regulatory point of view. The general objectives of the TGAM review were to consider: - enhancements of on-site accident management procedures and guidelines based on lessons learnt from the Fukushima Daiichi NPP accident; - decision-making and guiding principles in emergency situations; - guidance for instrumentation, equipment and supplies for addressing long-term aspects of accident management; - guidance and implementation when taking extreme measures for accident management. The report is built on the existing bases for capabilities to respond to design basis

  6. 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

  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. Use of PSA and severe accident assessment results for the accident management

    The objectives for this study are to investigate the basic principle or methodology which is applicable to accident management, by using the results of PSA and severe accident research, and also facilitate the preparation of accidents management program in the future. This study was performed as follows: derivation of measures for core damage prevention, derivation of measures for accident mitigation, application of computerized tool to assess severe accident management

  9. 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

  10. Accident management insights from IPE's

    In response to the U.S. Nuclear Regulatory Commission's Generic Letter 88-20, each utility in the U.S.A. has undertaken a probabilistic severe accident study of each plant. This paper provides a high level summary of the generic PWR accident management insights that have been obtained from the IPE reports. More importantly, the paper details some of the limitations of the IPE studies with respect to accident management. The IPE studies and the methodology used was designed to provide a best estimate of the potential for a severe accident and/or for severe consequences from a core damage accident. The accepted methodology employs a number of assumptions to make the objective attainable with a reasonable expenditure of resources. However, some of the assumptions represent limitations with respect to developing an accident management program based solely on the IPE and its results. (author)

  11. 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

  12. Radiation protection management in Fukushima Daiichi NPS and post-accident measures

    Fukushima Daiichi Nuclear Power Station was hit by the big earthquake and tsunami, which caused the station black out and subsequent loss of cooling functions for reactor and spent fuel pools (SFPs). Consequently the fuels were damaged, hydrogen explosion blew off top of the reactor buildings and radioactive materials were released to the atmosphere and the ocean. Tsunami and power loss caused many difficulties of monitoring, dose management, and radiation protection of workers. For example, the radiation management system was down and about 5,000 Alarm Pocket Dosimeters (APDs) and their battery chargers could not be used. Due to the insufficient number of APDs, one representative of each working team had a dosimeter under the limited conditions. Through the accident, we got following lessons learned; (1) Reinforcing monitoring posts, (2) Preparing more radiation protection equipment, (3) Establishing emergency access control centre, and (4) Education and training in radiation protection. (author)

  13. CAMS: Computerized Accident Management Support

    The OECD Halden Reactor Project has initiated a new research programme on computerised accident management support, the so-called CAMS project (CAMS = Computerized Accident Management Support). This work will investigate the possibilities for developing systems which provide more extensive support to the control room staff and technical support centre than the existing SPDS (Safety Parameter Display System) type of systems. The CAMS project will utilize available simulator codes and the capabilities of computerized tools to assist the plant staff during the various accident stages including: identification of the accident state, assessment of the future development of the accident, and planning accident mitigation strategies. This research programme aims at establishing a prototype system which can be used for experimental testing of the concept and serve as a tool for training and education in accident management. The CAMS prototype should provide support to the staff when the plant is in a normal state, in a disturbance sate, and in an accident state. Even though better support in an accident state is the main goal of the project, it is felt to be important that the staff is familiar with the use of the system during normal operation, when they utilize the system during transients

  14. Severe accident management. Prevention and Mitigation

    Effective planning for the management of severe accidents at nuclear power plants can produce both a reduction in the frequency of such accidents as well as the ability to mitigate their consequences if and when they should occur. This report provides an overview of accident management activities in OECD countries. It also presents the conclusions of a group of international experts regarding the development of accident management methods, the integration of accident management planning into reactor operations, and the benefits of accident management

  15. Accident management measures. Demand for action as seen by the supervising authority; Massnahmen des anlageninternen Notfallschutzes - Handlungsbedarf aus behoerdlicher Sicht

    Wolter, W. [Ministerium fuer Finanzen und Energie des Landes Schleswig-Holstein, Kiel (Germany)

    1994-07-01

    The various measures taken for accident management in the plant are to be classified into categories of nuclear law, as there are: prevention of hazards, prevention of risks, or non-preventive measures ( management of remaining risk). Screening the various measures for classification shows that most of them belong to the category of preventive action under the Atomic Energy Act. This means that these measures have to be addressed in KTA safety standards. (orig./HP) [Deutsch] Die rechtliche Einordnung jeder einzelnen Massnahmen des anlageninternen Notfallschutzes in eine der atomrechtlichen Kategorien Gefahrenabwehr, Risikovorsorge oder Nichtvorsorge (Restrisikomassnahme) ist erforderlich. Eine ueberschlaegige Betrachtung fuehrt zu dem Ergebnis, dass zahlreiche technische Massnahmen des anlageninternen Notfallschutzes dem atomrechtlichen Vorsorgebegriff zuzuordnen sind (Risikovorsorge). Sofern Massnahmen des anlageninternen Notfallschutzes der atomrechtlichen Vorsorge zuzuordenen sind, sind sie zwingend auch im KTA-Regelwerk zu verankern. (orig./HP)

  16. Severe accidents at nuclear power plants. Their risk assessment and accident management

    This document is to explain the severe accident issues. Severe Accidents are defined as accidents which are far beyond the design basis and result in severe damage of the core. Accidents at Three Mild Island in USA and at Chernobyl in former Soviet Union are examples of severe accidents. The causes and progressions of the accidents as well as the actions taken are described. Probabilistic Safety Assessment (PSA) is a method to estimate the risk of severe accidents at nuclear reactors. The methodology for PSA is briefly described and current status on its application to safety related issues is introduced. The acceptability of the risks which inherently accompany every technology is then discussed. Finally, provision of accident management in Japan is introduced, including the description of accident management measures proposed for BWRs and PWRs. (author)

  17. Severe accident management concept for LWRS

    Although the advanced built-in engineered safety features and the highly trained personnel have led to extremely low probabilities of core melt accidents, there is a common understanding that even for such very unlikely accidents the plant operators must have the ability and means to mitigate the consequences of such events. This paper outlines a concept for the management of severe accidents based on 1) Computer simulations. 2) Various strategies based on core and containment damage states. 3) Calculational Aids. 4) Procedures. 5) Technical basis report. 6) Training. 7) Drills. The major benefit of this concept is the fact that there is no dedicated operating manual for severe accidents; rather the required mitigative strategies and measures are incorporated into existing accident management manuals leading to truly integrated accident management at the plant. At present this concept is going to be implemented in the NPP Geogen. Although this approach is primarily developed for existing PWRs it is also applicable to other LWRs including new NPP designs. Specific features of the plant can be taken into account by an adaptation of the concept. (authors)

  18. Severe accident management guidelines tool

    Severe Accident is addressed by means of a great number of documents such as guidelines, calculation aids and diagnostic trees. The response methodology often requires the use of several documents at the same time while Technical Support Centre members need to assess the appropriate set of equipment within the adequate mitigation strategies. In order to facilitate the response, TECNATOM has developed SAMG TOOL, initially named GGAS TOOL, which is an easy to use computer program that clearly improves and accelerates the severe accident management. The software is designed with powerful features that allow the users to focus on the decision-making process. Consequently, SAMG TOOL significantly improves the severe accident training, ensuring a better response under a real situation. The software is already installed in several Spanish Nuclear Power Plants and trainees claim that the methodology can be followed easier with it, especially because guidelines, calculation aids, equipment information and strategies availability can be accessed immediately (authors)

  19. Accident management strategy focusing on the software area

    Tokyo Electric Power Company has already conducted individual plant examination (IPE) and worked out specific accident management strategies. In addition to hardware projects which will be carried out in due order from now on, we have studied the software aspects of accident management, including personnel education and training in relevant subjects. Based on the results of these studies, a decision has been made on the work sharing between the main control room (MCR) and technical support center (TSC) in implementing accident management. We have also decided on a improvement of guidelines and manuals, such as emergency operation procedures (EOP) and accident management guidelines (AMG), and on a basic policy on personnel education and training in accident management. Following this decision, our future efforts will be focused on improving software measures in combination with hardware measures to work out a well-balanced accident management program. (author)

  20. The management of radioactive waste from accidents

    Two accident case histories are reviewed - the Three Mile Island (TMI-2) reactor accident in 1979 and the Seveso accident in 1976. The status of the decontamination and radioactive waste management operations at TMI-2 as at 1986 is presented. 1986 estimates of reactor accident and recovery costs are given. 12 refs., 8 tabs

  1. The measurement of accident-proneness

    As, Sicco van

    2001-01-01

    This paper deals with the measurement of accident-proneness. Accidents seem easy to observe, however accident-proneness is difficult to measure. In this paper I first define the concept of accident-proneness, and I develop an instrument to measure it. The research is mainly executed within chemical

  2. Severe accident analysis methodology in support of accident management

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

  3. Verification of accident management strategies at the Forsmark plant

    Due to government requirements severe accident mitigating measures were implemented at the Swedish State Power Board nuclear power plants in 1988. These measures included protection against early containment impairment, highly redundant containment spray and filtered venting of the containment. We also developed accident management strategies and corresponding documents to counteract a severe accident situation. This paper describes the accident management strategies and documents at the Forsmark nuclear power plant, the verification process of the basic approach, and our ongoing program for further development and verification of the accident management program. In summary: From the beginning it was emphasized that it was not only mitigating measures implemented, it was an accident mitigation program, including new EOP's and education and training. This program was implemented, as required by the Swedish government in the end of 1988. Since that time the accident management strategy has been validated, verified and further developed. As a general conclusion, the implemented accident management program has reached a fair degree of completeness at the Forsmark plant. It is expected that in the case a hypothetical accident would occur the envisaged strategy would handle the accident in such a way that the radiological consequences would be insignificant and radiation exposure to the personnel would be within ICRP recommendations. To reach and keep this goal it is imperative that a mental preparedness is always present. This is achieved with a continuous education, training and analyses

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

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

  5. 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)

  6. The screening approach for review of accident management programmes

    In this lecture the screening approach for review of accident management programmes are presented. It contains objective trees for accident management: logic structure of the approach; objectives and safety functions for accident management; safety principles

  7. Containment severe accident management - selected strategies

    The OECD Nuclear Energy Agency (NEA) organized in June 1994, in collaboration with the Swedish Nuclear Power Inspectorate (SKI), a Specialist Meeting on Selected Containment Severe Accident Management Strategies, to discuss their feasibility, effectiveness, benefits and drawbacks, and long-term impact. The meeting focused on water reactors, mainly on existing systems. The technical content covered topics such as general aspects of accident management strategies in OECD Member countries, hydrogen management techniques and other containment accident management strategies, surveillance and protection of the containment function. The main conclusions of the meeting are summarized in the paper. (author)

  8. Strategy generation in accident management support

    An increased interest for research in the field of Accident Management can be noted. Several international programmes have been started in order to be able to understand the basic physical and chemical phenomena in accident conditions. A feasibility study has shown that it would be possible to design and develop a computerized support system for plant staff in accident situations. To achieve this goal the Halden Project has initiated a research programme on Computerized Accident Management Support (CAMS project). The aim is to utilize the capabilities of computerized tools to support the plant staff during the various accident stages. The system will include identification of the accident state, assessment of the future development of the accident and planning of accident mitigation strategies. A prototype is developed to support operators and the Technical Support Centre in decision making during serious accident in nuclear power plants. A rule based system has been built to take care of the strategy generation. This system assists plant personnel in planning control proposals and mitigation strategies from normal operation to severe accident conditions. The ideal of a safety objective tree and knowledge from the emergency procedures have been used. Future prediction requires good state identification of the plant status and some knowledge about the history of some critical variables. The information needs to be validated as well. Accurate calculations in simulators and a large database including all important information form the plant will help the strategy planning. (author). 12 refs, 2 figs

  9. On preparation for accident management in LWR power stations

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

  10. SAMSON: Severe Accident Management System Online Network

    SAMSON, Severe Accident Management System Online Network, is a computational tool used in the event of a nuclear power plant accident by accident managers in the Technical Support Centers (TSC) and Emergency Offsite Facilities (EOF). SAMSON examines over 150 status points monitored by nuclear power plant process computers during a severe accident and makes predictions about when core damage, support plate failure, and reactor vessel failure will occur. These predictions are based on the current state of the plant assuming that all safety equipment not already operating will fail. The status points analyzed include radiation levels, flow rates, pressure levels, temperatures and water levels. SAMSON uses an expert system as well as neural networks trained with the back propagation learning algorithm to make predictions. Previous training on data from accident analysis code allows SAMSON to associate different states in the plant with different times to critical failures. The accidents currently recognized by SAMSON include steam generator tube ruptures (SGTR), with breaks ranging from one tube to eights tubes, and loss of coolant accidents (LOCA), with breaks ranging from 0.001 square feet in size to breaks 3.0 square feet. SAMSON contains several neural networks for each accident type and break size, and chooses the correct network after accident classification by in expert system. SAMSON also provides information concerning the status of plant sensors and recovery strategies

  11. A Methodology for Probabilistic Accident Management

    While techniques have been developed to tackle different tasks in accident management, there have been very few attempts to develop an on-line operator assistance tool for accident management and none that can be found in the literature that uses probabilistic arguments, which are important in today's licensing climate. The state/parameter estimation capability of the dynamic system doctor (DSD) approach is combined with the dynamic event-tree generation capability of the integrated safety assessment (ISA) methodology to address this issue. The DSD uses the cell-to-cell mapping technique for system representation that models the system evolution in terms of probability of transitions in time between sets of user-defined parameter/state variable magnitude intervals (cells) within a user-specified time interval (e.g., data sampling interval). The cell-to-cell transition probabilities are obtained from the given system model. The ISA follows the system dynamics in tree form and braches every time a setpoint for system/operator intervention is exceeded. The combined approach (a) can automatically account for uncertainties in the monitored system state, inputs, and modeling uncertainties through the appropriate choice of the cells, as well as providing a probabilistic measure to rank the likelihood of possible system states in view of these uncertainties; (b) allows flexibility in system representation; (c) yields the lower and upper bounds on the estimated values of state variables/parameters as well as their expected values; and (d) leads to fewer branchings in the dynamic event-tree generation. Using a simple but realistic pressurizer model, the potential use of the DSD-ISA methodology for on-line probabilistic accident management is illustrated

  12. 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

  13. Emergency monitoring strategy and radiation measurements. Working document of the NKS project emergency management and radiation monitoring in nuclear and radiological accidents (EMARAD)

    This report is one of the deliverables of the NKS Project Emergency management and radiation monitoring in nuclear and radiological accidents (EMARAD) (20022005). The project and the overall results are briefly described in the NKS publication 'Emergency Management and Radiation Monitoring in Nuclear and Radiological Accidents. Summary Report on the NKS Project EMARAD' (NKS-137, April 2006). In a nuclear or radiological emergency, all radiation measurements must be performed efficiently and the results interpreted correctly in order to provide the decision-makers with adequate data needed in analysing the situation and carrying out countermeasures. Managing measurements in different situations in a proper way requires the existence of pre-prepared emergency monitoring strategies. Preparing a comprehensive yet versatile strategy is not an easy task to perform because there are lots of different factors that have to be taken into account. The primary objective of this study was to discuss the general problematics concerning emergency monitoring strategies and to describe a few important features of an efficient emergency monitoring system as well as factors affecting measurement activities in practise. Some information concerning the current situation in the Nordic countries has also been included. (au)

  14. Emergency monitoring strategy and radiation measurements document of the NKS project emergency management and radiation monitoring in nuclear and radiological accidents (EMARAD)

    Lahtinen, J. [Radiation and Nuclear Safety Authority (STUK) (Finland)

    2006-04-15

    This report is one of the deliverables of the NKS Project Emergency management and radiation monitoring in nuclear and radiological accidents (EMARAD) (20022005). The project and the overall results are briefly described in the NKS publication 'Emergency Management and Radiation Monitoring in Nuclear and Radiological Accidents. Summary Report on the NKS Project EMARAD' (NKS-137, April 2006). In a nuclear or radiological emergency, all radiation measurements must be performed efficiently and the results interpreted correctly in order to provide the decision-makers with adequate data needed in analysing the situation and carrying out countermeasures. Managing measurements in different situations in a proper way requires the existence of pre-prepared emergency monitoring strategies. Preparing a comprehensive yet versatile strategy is not an easy task to perform because there are lots of different factors that have to be taken into account. The primary objective of this study was to discuss the general problematics concerning emergency monitoring strategies and to describe a few important features of an efficient emergency monitoring system as well as factors affecting measurement activities in practise. Some information concerning the current situation in the Nordic countries has also been included. (au)

  15. Development of integrated accident management assessment technology

    This project aims to develop critical technologies for accident management through securing evaluation frameworks and supporting tools, in order to enhance capabilities coping with severe accidents. For the research goal, firstly under the viewpoint of accident prevention, on-line risk monitoring system and the analysis framework for human error have been developed. Secondly, the training/supporting systems including the training simulator and the off-site risk evaluation system have been developed to enhance capabilities coping with severe accidents. Four kinds of research results have been obtained from this project. Firstly, the framework and taxonomy for human error analysis has been developed for accident management. As the second, the supporting system for accident managements has been developed. Using data that are obtained through the evaluation of off-site risk for Younggwang site, the risk database as well as the methodology for optimizing emergency responses has been constructed. As the third, a training support system, SAMAT, has been developed, which can be used as a training simulator for severe accident management. Finally, on-line risk monitoring system, DynaRM, has been developed for Ulchin 3 and 4 unit

  16. Accident knowledge and emergency management

    Rasmussen, B.; Groenberg, C.D.

    1997-03-01

    The report contains an overall frame for transformation of knowledge and experience from risk analysis to emergency education. An accident model has been developed to describe the emergency situation. A key concept of this model is uncontrolled flow of energy (UFOE), essential elements are the state, location and movement of the energy (and mass). A UFOE can be considered as the driving force of an accident, e.g., an explosion, a fire, a release of heavy gases. As long as the energy is confined, i.e. the location and movement of the energy are under control, the situation is safe, but loss of confinement will create a hazardous situation that may develop into an accident. A domain model has been developed for representing accident and emergency scenarios occurring in society. The domain model uses three main categories: status, context and objectives. A domain is a group of activities with allied goals and elements and ten specific domains have been investigated: process plant, storage, nuclear power plant, energy distribution, marine transport of goods, marine transport of people, aviation, transport by road, transport by rail and natural disasters. Totally 25 accident cases were consulted and information was extracted for filling into the schematic representations with two to four cases pr. specific domain. (au) 41 tabs., 8 ills.; 79 refs.

  17. Accident knowledge and emergency management

    The report contains an overall frame for transformation of knowledge and experience from risk analysis to emergency education. An accident model has been developed to describe the emergency situation. A key concept of this model is uncontrolled flow of energy (UFOE), essential elements are the state, location and movement of the energy (and mass). A UFOE can be considered as the driving force of an accident, e.g., an explosion, a fire, a release of heavy gases. As long as the energy is confined, i.e. the location and movement of the energy are under control, the situation is safe, but loss of confinement will create a hazardous situation that may develop into an accident. A domain model has been developed for representing accident and emergency scenarios occurring in society. The domain model uses three main categories: status, context and objectives. A domain is a group of activities with allied goals and elements and ten specific domains have been investigated: process plant, storage, nuclear power plant, energy distribution, marine transport of goods, marine transport of people, aviation, transport by road, transport by rail and natural disasters. Totally 25 accident cases were consulted and information was extracted for filling into the schematic representations with two to four cases pr. specific domain. (au) 41 tabs., 8 ills.; 79 refs

  18. Fundamentals for reviewing accident managements of reprocessing facilities

    The accident at Fukushima Daiichi Nuclear Power Station insisted a necessity of reconsideration of the defence in depth concept against events exceeding design basis. The insistence suggested a need of practical guidance for reviewing accident management measures for such events. Soon after the accident, Japan Nuclear Energy Safety Organization (JNES) started a preliminary study on the points to be considered in reviewing comprehensiveness and consistency of accident management measures for reprocessing facilities. The results of PSA studies which have been pursued at JNES contributed significantly to the preliminary study, because the contents of the PSA studies have a close relation with subjects to be considered in the review. Based on the insight the paper focuses on such relation and discusses fundamentals for the review in terms of the knowledge derived from the PSA and specific features of reprocessing facilities. The result of the study is also described with touching relations to the fundamentals. (author)

  19. Reconstruction of the Chernobyl emergency and accident management

    Full text of publication follows: on April 26, 1986 the most serious civil technological accident in the history of mankind occurred of the Chernobyl Nuclear Power Plant (ChNPP) in the former Soviet Union. As a direct result of the accident, the reactor was severely destroyed and large quantities of radionuclides were released. Some 800000 persons, also called 'liquidators' - including plant operators, fire-fighters, scientists, technicians, construction workers, emergency managers, volunteers, as well as medical and military personnel - were part of emergency measurements and accident management efforts. Activities included measures to prevent the escalation of the accident, mitigation actions, help for victims as well as activities in order to provide a basic infrastructure for this unprecedented and overwhelming task. The overall goal of the 'Project Chernobyl' of the Institute of Risk Research of the University of Vienna was to preserve for mankind the experience and knowledge of the experts among the 'liquidators' before it is lost forever. One method used to reconstruct the emergency measures of Chernobyl was the direct cooperation with liquidators. Simple questionnaires were distributed among liquidators and a database of leading accident managers, engineers, medical experts etc. was established. During an initial struggle with a number of difficulties, the response was sparse. However, after an official permit had been issued, the questionnaires delivered a wealth of data. Furthermore a documentary archive was established, which provided additional information. The multidimensional problem in connection with the severe accident of Chernobyl, the clarification of the causes of the accident, as well as failures and successes and lessons to be learned from the Chernobyl emergency measures and accident management are discussed. (authors)

  20. Medical response and management of radiation accidents

    An overview is provided of educational programs and principles essential to the appropriate medical management of radiation accident victims. Such an education program will provide details of the physical properties of radiation, of the sources of radiation exposure, of radiation protection standards and of biological radiation effects. The medical management of individuals involved in radiation accidents is discussed. Such management includes emergency medical stabilization, locating and quantitating the level and degree of internal and/or external contamination, wound decontamination, medical surveillance and the evaluation and treatment of local radiation injuries

  1. ATHLET validation using accident management experiments

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

  2. US nuclear industry perspective on accident management

    The Nuclear Management and Resources Council (NUMARC) serves as the United States nuclear power industry's principal mechanism for conveying industry views, concerns, and policies regarding industry wide regulatory issues to the Nuclear Regulatory Commission (NRC) and other government agencies as appropriate. NUMARC and the Electric Power Research Institute (EPRI), in support of the NUMARC Severe Accident Working Group's (SAWG's) efforts with regard to accident management, has developed a framework for evaluation of plant-specific accident management capabilities. These capabilities fall into one of three main categories: (1) personnel resources (organization, training, communications); (2) systems and equipment (restoration and repair, instrumentation, use of alternatives); and (3) information resources (procedures and guidance, technical information, process information). The purpose of this paper is to describe this framework, its objectives, the five major steps involved and areas to consider further

  3. Severe accident management. Optimized guidelines and strategies

    The highest priority for mitigating the consequences of a severe accident with core melt lies in securing containment integrity, as this represents the last barrier against fission product release to the environment. Containment integrity is endangered by several physical phenomena, especially highly transient phenomena following high-pressure reactor pressure vessel failure (like direct containment heating or steam explosions which can lead to early containment failure), hydrogen combustion, quasi-static over-pressure, temperature failure of penetrations, and basemat penetration by core melt. Each of these challenges can be counteracted by dedicated severe accident mitigation hardware, like dedicated primary circuit depressurization valves, hydrogen recombiners or igniters, filtered containment venting, containment cooling systems, and core melt stabilization systems (if available). However, besides their main safety function these systems often have also secondary effects that need to be considered. Filtered containment venting causes (though limited) fission product release into the environment, primary circuit depressurization leads to loss of coolant, and an ex-vessel core melt stabilization system as well as hydrogen igniters can generate high pressure and temperature loads on the containment. To ensure that during a severe accident any available systems are used to their full beneficial extent while minimizing their potential negative impact, AREVA has implemented a severe accident management for German nuclear power plants. This concept makes use of extensive numerical simulations of the entire plant, quantifying the impact of system activations (operational systems, safety systems, as well as dedicated severe accident systems) on the accident progression for various scenarios. Based on the knowledge gained, a handbook has been developed, allowing the plant operators to understand the current state of the plant (supported by computational aids), to predict

  4. Management of foodstuffs after nuclear accidents

    A model for the management of foodstuffs after nuclear accidents is presented. The model is a synthesis of traditions and principles taken from both radioactive protection and management of food. It is based on cooperation between the Nordic countries and on practical experience gained from the Chernobyl accident. The aim of the model is to produce a basis for common plans for critical situations based on criteria for decision making. In the case of radioactive accidents it is important that the protection of the public and of the society is handled in a positive way. The model concerns production, marketing and consumption of food and beverage. The overall aim is that the radiation doses should be as low and harmless to health for individual members of the public. (CLS) 35 refs

  5. Development of severe accident management advisory and training simulator (SAMAT)

    The most operator support systems including the training simulator have been developed to assist the operator and they cover from normal operation to emergency operation. For the severe accident, the overall architecture for severe accident management is being developed in some developed countries according to the development of severe accident management guidelines which are the skeleton of severe accident management architecture. In Korea, the severe accident management guideline for KSNP was recently developed and it is expected to be a central axis of logical flow for severe accident management. There are a lot of uncertainties in the severe accident phenomena and scenarios and one of the major issues for developing a operator support system for a severe accident is the reduction of these uncertainties. In this paper, the severe accident management advisory system with training simulator, SAMAT, is developed as all available information for a severe accident are re-organized and provided to the management staff in order to reduce the uncertainties. The developed system includes the graphical display for plant and equipment status, the previous research results by knowledge-base technique, and the expected plant behavior using the severe accident training simulator. The plant model used in this paper is oriented to severe accident phenomena and thus can simulate the plant behavior for a severe accident. Therefore, the developed system may make a central role of the information source for decision-making for a severe accident management, and will be used as the training simulator for severe accident management

  6. Chernobyl reactor accident: medical management

    Chernobyl reactor accident on 26th April, 1986 is by far the worst radiation accident in the history of the nuclear industry. Nearly 500 plant personnel and rescue workers received doses varying from 1-16 Gy. Acute radiation syndrome (ARS) was seen only in the plant personnel. 499 individuals were screened for ARS symptoms like nausea, vomitting, diarrhoea and fever. Complete blood examination was done which showed initial granulocytosis followed by granulocytopenia and lymphocytopenia. Cytogenetic examinations were confirmatory in classifying the patients on the basis of the doses received. Two hundred and thirty seven cases of ARS were hospitalised in the first 24-36 hrs. No member of general public suffered from ARS. There were two immediate deaths and subsequently 28 died in hospital and one of the cases died due to myocardial infarction, making a total of 31 deaths. The majority of fatal cases had whole body doses of about 6 Gy, besides extensive skin burns. Two cases of radiation burns had thermal burns also. Treatment of ARS consisted of isolation, barrier nursing, replacement therapy with fluid electrolytes, platelets and RBC transfusions and antibiotic therapy for bacterial, fungal and viral infections. Bone marrow transplantations were given to 13 cases out of which 11 died due to various causes. Radiation burns due to beta, gamma radiations were seen in 56 cases and treated with dressings, surgical excision, skin grafting and amputation. Oropharangeal syndrome, producing extensive mucous in the oropharynx, was first seen in Chernobyl. The patients were treated with saline wash of the mouth. The patients who had radioactive contamination due to radioactive iodine were given stable iodine, following wash with soap, water and monitored. Fourteen survivors died subsequently due to other causes. Late health effects seen so far include excess of thyroid cancer in the children and psychological disorders due to stress. No excess leukemia has been reported so

  7. Occupational Radiation Protection in Severe Accident Management

    As an early response to the Fukushima Daiichi NPP accident, the Information System on Occupational Exposure (ISOE) Bureau decided to focus on the following issues as an initial response of the joint program after having direct communications with the Japanese official participants in April 2011: - Management of high radiation area worker doses: It has been decided to make available the experience and information from the Chernobyl accident in terms of how emergency worker / responder doses were legally and practically managed, - Personal protective equipment for highly-contaminated areas: It was agreed to collect information about the types of personnel protective equipment and other equipment (e.g. air bottles, respirators, air-hoods or plastic suits, etc.), as well as high-radiation area worker dosimetry use (e.g. type, number and placement of dosimetry) for different types of emergency and high-radiation work situations. Detailed information was collected on dose criteria which are used for emergency workers /responders and their basis, dose management criteria for high dose/dose rate areas, protective equipment which is recommended for emergency workers / responders, recommended individual monitoring procedures, and any special requirement for assessment from the ISOE participating nuclear utilities and regulatory authorities and made available for Japanese utilities. With this positive response of the ISOE official participants and interest in the situation in Fukushima, the Expert Group on Occupational Radiation Protection in Severe Accident Management (EG-SAM) was established by the ISOE Management Board in May 2011. The overall objective of the EG-SAM is to contribute to occupational exposure management (providing a view on management of high radiation area worker doses) within the Fukushima plant boundary with the ISOE participants and to develop a state-of-the-art ISOE report on best radiation protection management practices for proper radiation

  8. PSA use in accident management studies in Japan

    The safety of NPPs in Japan is secured by stringent safety regulations based on the deterministic method, minimizing the possibility a severe accident to a technologically negligible level. PSA is not required in the current regulatory procedures. Accident management based on PSA is a 'knowledge-based' action dependent on utilities' technical knowledge aimed at further reduction of the risk which is kept small enough by existing measures. The paper discusses the following three kinds of PSAs that have been conducted practically and efficiently on NPPs to provide supplemental information about their safety characteristics in addition to the deterministic evaluation used in the regulatory safety review: PSAs on typical NPPs, PSAs on all NPPs to examine candidates for accident management, and PSAs as part of periodic safety review (PSR). 1 fig., 5 tabs

  9. Development of emergency response support system for accident management

    Specific measures for the accident management (AM) are proposed to prevent the severe accident and to mitigate their effects in order to upgrade the safety of nuclear power plants even further. To ensure accident management effective, it is essential to grasp the plant status accurately. In consideration of the above mentioned background, the Emergency Response Support System (ERSS) was developed as a computer assisted prototype system by a joint study of Japanese BWR group. This system judges and predicts the plant status at the emergency condition in a nuclear power plant. This system displays the results of judgment and prediction. The effectiveness of the system was verified through the test and good prospects for applying the system to a plant was obtained. 7 refs., 10 figs

  10. Artificial intelligence applications in accident management

    For nuclear power plant accident management, there are some addition concerns: linking AI systems to live data streams must be mastered; techniques for processing sensor inputs with varying data quality need to be provided; systems responsiveness to changing plant conditions and multiple user requests should, in general, be improved; there is a need for porting applications from specialized AI machines onto conventional computer hardware without incurring unacceptable performance penalties; human factors guidelines are required for new user interfaces in AI applications; methods for verification and validation of AI-based systems must be developed; and, finally, there is a need for proven methods to evaluate use effectiveness and firmly establish the benefits of AI-based accident management systems. (orig./GL)

  11. Severe accident research and management in Nordic Countries - A status report

    The report describes the status of severe accident research and accident management development in Finland, Sweden, Norway and Denmark. The emphasis is on severe accident phenomena and issues of special importance for the severe accident management strategies implemented in Sweden and in Finland. The main objective of the research has been to verify the protection provided by the accident mitigation measures and to reduce the uncertainties in risk dominant accident phenomena. Another objective has been to support validation and improvements of accident management strategies and procedures as well as to contribute to the development of level 2 PSA, computerised operator aids for accident management and certain aspects of emergency preparedness. Severe accident research addresses both the in-vessel and the ex-vessel accident progression phenomena and issues. Even though there are differences between Sweden and Finland as to the scope and content of the research programs, the focus of the research in both countries is on in-vessel coolability, integrity of the reactor vessel lower head and core melt behaviour in the containment, in particular the issues of core debris coolability and steam explosions. Notwithstanding that our understanding of these issues has significantly improved, and that experimental data base has been largely expanded, there are still important uncertainties which motivate continued research. Other important areas are thermal-hydraulic phenomena during reflooding of an overheated partially degraded core, fission product chemistry, in particular formation of organic iodine, and hydrogen transport and combustion phenomena. The development of severe accident management has embraced, among other things, improvements of accident mitigating procedures and strategies, further work at IFE Halden on Computerised Accident Management Support (CAMS) system, as well as plant modifications, including new instrumentation. Recent efforts in Sweden in this area

  12. Severe accident research and management in Nordic Countries - A status report

    Frid, W. [Swedish Nuclear Power Inspectorate, SKI (Sweden)] (ed.)

    2002-01-01

    The report describes the status of severe accident research and accident management development in Finland, Sweden, Norway and Denmark. The emphasis is on severe accident phenomena and issues of special importance for the severe accident management strategies implemented in Sweden and in Finland. The main objective of the research has been to verify the protection provided by the accident mitigation measures and to reduce the uncertainties in risk dominant accident phenomena. Another objective has been to support validation and improvements of accident management strategies and procedures as well as to contribute to the development of level 2 PSA, computerised operator aids for accident management and certain aspects of emergency preparedness. Severe accident research addresses both the in-vessel and the ex-vessel accident progression phenomena and issues. Even though there are differences between Sweden and Finland as to the scope and content of the research programs, the focus of the research in both countries is on in-vessel coolability, integrity of the reactor vessel lower head and core melt behaviour in the containment, in particular the issues of core debris coolability and steam explosions. Notwithstanding that our understanding of these issues has significantly improved, and that experimental data base has been largely expanded, there are still important uncertainties which motivate continued research. Other important areas are thermal-hydraulic phenomena during reflooding of an overheated partially degraded core, fission product chemistry, in particular formation of organic iodine, and hydrogen transport and combustion phenomena. The development of severe accident management has embraced, among other things, improvements of accident mitigating procedures and strategies, further work at IFE Halden on Computerised Accident Management Support (CAMS) system, as well as plant modifications, including new instrumentation. Recent efforts in Sweden in this area

  13. Precept from the management for the accident of Fukushima daiichi

    At 17 hours after the accident of Fukushima Daiichi Nuclear Power Plant due to the Great East Japan Earthquake, National Institute of Radiological Sciences sent the first REMAT (Radiation Emergency Medical Assistance Team) in the 20 km range from the Plant. The team members were confronted by two issues: (1) Medical activities under the infrastructures destructed by a multiple disaster caused by earthquake, tsunami and nuclear accident, which was not presumed. (2) Radiation protection management for dispatched staff. Measures for this situation worked out by activities on the site are presented. (K.Y.)

  14. The expert assistant in accident management

    In the event of a nuclear accident in proximity to an urban area, the consequences resulting from the complex processes of environmental transport of radioactivity would require complex countermeasures. Emphasis has been placed on either modelling the potential effects of such an event on the population, or on attempting to predict the geographical evolution of the release. Less emphasis has been placed on the development of accident management aids with a in-built data acquisition capability. Given the problems of predicting the evolution of an accidental release of activity, more emphasis should be placed on the development of small regional systems specifically engineered to acquire and display environmental data in the most efficaceous form possible. A wealth of information can be obtained from appropriately-sited outstations which can aid those responsible for countermeasures in their decision making processes. The substantial volume of data which would arrive within the duration and during the aftermath of an accident requires skilled interpretation under conditions of considerable stress. It is necessary that a management aid notonly presents these data in a rapidly assimilable form, but is capable of making intelligent decisions of its own, on such matters as information display priority and the polling frequency of outstations. The requirement is for an expert assistant. The XERSES accident management aid has been designed with the foregoing features in mind. Intended for covering regions up to approximately 100 kms square, it links with between 1 and 64 outstations supplying a variety of environmental data. Under quiescent conditions the system will operate unattended, raising alarms remotely only when detecting abnormal conditions. Under emergency conditions, the system automatically adjusts such operating parameters as data acquisition rate

  15. Simulation of severe accident in reactor core for training and accident management

    An Advanced Real-time Severe Accident Simulation (ARTSAS) train reactor operators and accident management teams for scenarios simulating severe accidents in nuclear reactors. The code has been integrated with the real-time tools and the RAINBO graphic package to provide training and analysis tools on workstations as well as on full-scope simulators. (orig.) (4 refs., 1 fig.)

  16. 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)

  17. Early measurements after the Goiania accident

    During the early, intermediate late phase of the Goiania radiological accident different survey methods were applied involving aerial and terrestrial (using a car and directly in the field) inspections. The present work aims to show how and when they were and the obtained results. Furthermore, the 137Cs concentration in soils were determined using a NaI(Tl) spectrometer during the accident, and also in Rio de Janeiro in a high resolution gamma spectrometry system. The concordance among those results and the validity of the 137Cs measurements in soil with NaI(TI) are demonstrated. (author)

  18. OSSA - An optimized approach to severe accident management: EPR application

    There is a recognized need to provide nuclear power plant technical staff with structured guidance for response to a potential severe accident condition involving core damage and potential release of fission products to the environment. Over the past ten years, many plants worldwide have implemented such guidance for their emergency technical support center teams either by following one of the generic approaches, or by developing fully independent approaches. There are many lessons to be learned from the experience of the past decade, in developing, implementing, and validating severe accident management guidance. Also, though numerous basic approaches exist which share common principles, there are differences in the methodology and application of the guidelines. AREVA/Framatome-ANP is developing an optimized approach to severe accident management guidance in a project called OSSA ('Operating Strategies for Severe Accidents'). There are still numerous operating power plants which have yet to implement severe accident management programs. For these, the option to use an updated approach which makes full use of lessons learned and experience, is seen as a major advantage. Very few of the current approaches covers all operating plant states, including shutdown states with the primary system closed and open. Although it is not necessary to develop an entirely new approach in order to add this capability, the opportunity has been taken to develop revised full scope guidance covering all plant states in addition to the fuel in the fuel building. The EPR includes at the design phase systems and measures to minimize the risk of severe accident and to mitigate such potential scenarios. This presents a difference in comparison with existing plant, for which severe accidents where not considered in the design. Thought developed for all type of plants, OSSA will also be applied on the EPR, with adaptations designed to take into account its favourable situation in that field

  19. Severe accident approach - final report. Evaluation of design measures for severe accident prevention and consequence mitigation.

    Tentner, A. M.; Parma, E.; Wei, T.; Wigeland, R.; Nuclear Engineering Division; SNL; INL

    2010-03-01

    An important goal of the US DOE reactor development program is to conceptualize advanced safety design features for a demonstration Sodium Fast Reactor (SFR). The treatment of severe accidents is one of the key safety issues in the design approach for advanced SFR systems. It is necessary to develop an in-depth understanding of the risk of severe accidents for the SFR so that appropriate risk management measures can be implemented early in the design process. This report presents the results of a review of the SFR features and phenomena that directly influence the sequence of events during a postulated severe accident. The report identifies the safety features used or proposed for various SFR designs in the US and worldwide for the prevention and/or mitigation of Core Disruptive Accidents (CDA). The report provides an overview of the current SFR safety approaches and the role of severe accidents. Mutual understanding of these design features and safety approaches is necessary for future collaborations between the US and its international partners as part of the GEN IV program. The report also reviews the basis for an integrated safety approach to severe accidents for the SFR that reflects the safety design knowledge gained in the US during the Advanced Liquid Metal Reactor (ALMR) and Integral Fast Reactor (IFR) programs. This approach relies on inherent reactor and plant safety performance characteristics to provide additional safety margins. The goal of this approach is to prevent development of severe accident conditions, even in the event of initiators with safety system failures previously recognized to lead directly to reactor damage.

  20. Review of current Severe Accident Management (SAM) approaches for Nuclear Power Plants in Europe

    HERMSMEYER Stephan; Iglesias, R.; Herranz, L; REER B.; SONNENKALB M; NOWACK H.; Stefanova, A.; Raimond, E.; CHATELARD P.; FOUCHER Laurent; BARNAK M.; MATEJOVIC P; PASCAL GHISLAIN; VELA GARCIA MONICA; SANGIORGI MARCO

    2014-01-01

    The Fukushima accidents highlighted that both the in-depth understanding of such sequences and the development or improvement of adequate Severe Accident Management (SAM) measures are essential in order to further increase the safety of the nuclear power plants operated in Europe. To support this effort, the CESAM (Code for European Severe Accident Management) R&D project, coordinated by GRS, started in April 2013 for 4 years in the 7th EC Framework Programme of research and development of th...

  1. Severe accident management program at Cofrentes Nuclear Power Plant

    Cofrentes Nuclear Power Plant (GE BWR/6) has implemented its specific Severe Accident Management Program within this year 2000. New organization and guides have been developed to successfully undertake the management of a severe accident. In particular, the Technical Support Center will count on a new ''Severe Accident Management Team'' (SAMT) which will be in charge of the Severe Accident Guides (SAG) when Control Room Crew reaches the Emergency Operation Procedures (EOP) step that requires containment flooding. Specific tools and training have also been developed to help the SAMT to mitigate the accident. (author)

  2. 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.

  3. Assessment of candidate accident management strategies

    A set of candidate accident management strategies, whose purpose is to prevent or mitigate in-vessel core damage, were identified from various Nuclear Regulatory Commission (NRC) and industry reports. These strategies have been grouped in this report by the challenges they are intended to meet, and assessed to provide information which may be useful to individual licensees for consideration when they perform their Individual Plant Examinations. Each assessment focused on describing and explaining the strategy, considering its relationship to existing requirements and practices as well as identifying possible associated adverse effects. 10 refs

  4. Current state of the technology measures of accident from contamination by the radioactive substance. 2. Overall management of radioactive material contaminated waste in the off-site

    This paper focuses on the disposal standards of the Act on Special Measures Concerning the Handling of Environmental Pollution by Radioactive Materials by the NPS Accident Associated with the Tohoku District - off the Pacific Ocean Earthquake that Occurred on March 11, 2011, which was promulgated on August 30, 2011 as a framework for the management of radioactively contaminated waste and removed soil. It stipulated that the byproducts of water/sewage treatment, major ash, and fly ash up to the radiation of 8,000 Bq/kg can be reclaimed in land. However, fly ash has a limit in landfill conditions, due to very high leaching rate of radioactive cesium. Later, incineration ash with between 8,000 Bq/kg and 100,000 Bq/kg became possible to be buried at disposal sites corresponding to leachate-controlled type. The specified waste with 100,000 Bq/kg or above is reclaimed in land with specified method at a site provided with outer peripheral partition facilities and cut off from the public water and groundwater. In Fukushima Prefecture, the specified waste with 100,000 Bq/kg or above is to be stored in provisional storage facilities, and later sent to final disposal sites outside the prefecture after the volume has been reduced. The decontaminated waste composed of vegetation is covered totally with a breathable waterproof sheet, and stored at a provisional yard. According to the characteristics of each provisional storage yard, there are needs for patrol and management. (A.O.)

  5. Use of probabilistic safety analyses in severe accident management

    An important consideration in the development and assessment of severe accident management strategies is that while the strategies are often built on the knowledge base of Probabilistic Safety Analyses (PSA), they must be interpretable and meaningful in terms of the control room indicators. In the following, the relationships between PSA and severe accident management are explored using ex-vessel accident management at a PWR ice-condenser plant as an example. 2 refs., 1 fig., 3 tabs

  6. Lessons learned from Fukushima accident in relation to emergency management

    The latest accident in Fukushima, Japan, which involved concurrent accidents at multiple nuclear facilities due to the earthquakes and tsunami, as well as station blackouts for an extended period of time, demonstrated the need for an overall review of existing prevention measures. These measures include emergency protection measures for residents beyond the emergency planning zone, the application of radiation protection criteria that consider the release of radioactive materials to the environment over an extended period and the disposal of large-scale radioactive wastes and radiation protection criteria to be applied upon recovery. Accordingly, Japan has taken improvement initiatives in the area of prevention by submitting a government report on the Fukushima accident prior to the IAEA Ministerial Conference on Nuclear Safety in June last year, and the US has devised a regulatory system of its own, including directions for improvement through the NRC, which operated a temporary taskforce specifically for this purpose. This study examined how Japan is responding to the Fukushima accident and investigated directions that countries around the world can take to improve the area of nuclear protection in order to enhance Korea's own radiological emergency management system

  7. Emergency room management of radiation accidents

    Emergency room management of radioactively contaminated patients who have an associated medical injury requiring immediate attention must be handled with care. Radioactive contamination of the skin of a worker is not a medical emergency and is usually dealt with at the plant. Effective preplanning and on-the-scene triage will allow the seriously injured and contaminated patients to get the medical care they need with a minimum of confusion and interference. Immediate medical and surgical priorities always take precedence over radiation injuries and radioactive contamination. Probably the most difficult aspect of emergency management is the rarity of such accidents and hence the unfamiliarity of the medical staff with the appropriate procedures. The authors discuss how the answer to these problems is preplanning, having a simple and workable procedure and finally having 24-h access to experts

  8. A study on the development of framework and supporting tools for severe accident management

    Through the extensive research on severe accidents, knowledge on severe accident phenomenology has constantly increased. Based upon such advance, probabilistic risk studies have been performed for some domestic plants to identify plant-specific vulnerabilities to severe accidents. Severe accident management is a program devised to cover such vulnerabilities, and leads to possible resolution of severe accident issues. This study aims at establishing severe accident management framework for domestic nuclear power plants where severe accident management program is not yet established. Emphasis is given to in-vessel and ex-vessel accident management strategies and instrumentation availability for severe accident management. Among the various strategies investigated, primary system depressurization is found to be the most effective means to prevent high pressure core melt scenarios. During low pressure core melt sequences, cooling of in-vessel molten corium through reactor cavity flooding is found to be effective. To prevent containment failure, containment filtered venting is found to be an effective measure to cope with long-term and gradual overpressurization, together with appropriate hydrogen control measure. Investigation of the availability of Yonggwang 3 and 4 instruments shows that most of instruments essential to severe accident management lose their desired functions during the early phase of severe accident progression, primarily due to the environmental condition exceeded ranges of instruments. To prevent instrument failure, a wider range of instruments are recommended to be used for some severe accident management strategies such as reactor cavity flooding. Severe accidents are generally known to accompany a number of complex phenomena and, therefore, it is very beneficial when severe accident management personnel is aided by appropriately designed supporting systems. In this study, a support system for severe accident management personnel is developed

  9. Assessment of light water reactor accident management programs and experience

    The objective of this report is to provide an assessment of the current light water reactor experience regarding accident management programs and associated technology developments. This assessment for light water reactor (LWR) designs is provided as a resource and reference for the development of accident management capabilities for the production reactors at the Savannah River Site. The specific objectives of this assessment are as follows: 1. Perform a review of the NRC, utility, and industry (NUMARC, EPRI) accident management programs and implementation experience. 2. Provide an assessment of the problems and opportunities in developing an accident management program in conjunction or following the Individual Plant Examination process. 3. Review current NRC, utility, and industry technological developments in the areas of computational tools, severe accident predictive tools, diagnostic aids, and severe accident training and simulation

  10. Assessment of light water reactor accident management programs and experience

    Hammersley, R.J. [Fauske and Associates, Inc., Burr Ridge, IL (United States)

    1992-03-01

    The objective of this report is to provide an assessment of the current light water reactor experience regarding accident management programs and associated technology developments. This assessment for light water reactor (LWR) designs is provided as a resource and reference for the development of accident management capabilities for the production reactors at the Savannah River Site. The specific objectives of this assessment are as follows: 1. Perform a review of the NRC, utility, and industry (NUMARC, EPRI) accident management programs and implementation experience. 2. Provide an assessment of the problems and opportunities in developing an accident management program in conjunction or following the Individual Plant Examination process. 3. Review current NRC, utility, and industry technological developments in the areas of computational tools, severe accident predictive tools, diagnostic aids, and severe accident training and simulation.

  11. A Methodology for Evaluating Severe Accident Management Strategies

    Severe accidents are defined as those which entail at least an initial core damage, in many cases specified as the overcoming of the regulatory fuel. After Fukushima accident, the effectiveness of the severe accident management strategy has been attracted worldwide. There is a typical example of severe accident management strategy like Severe Accident Management and Guideline (SAMG). Unfortunately, suitable method for evaluating the accident management strategy is absence until now. In this study, the evaluation methodology which utilizes the decision tree is developed to evaluate the severe accident management strategies. In addition, we applied the developed methodology to ShinKori nuclear power plant Unit 3, 4 and modeled decision tree for evaluation. In this study, we developed a methodology to evaluate the severe accident management strategy by using decision tree. In addition, the evaluation was carried out by selecting the cavity flooding strategy. Shinkori unit 3, 4 which is APR1400 is selected and analyzed for reference plant. In order to evaluation, decision tree for cavity flooding is modeled. With reliability data, quantification will be conducted. The utility of other severe accident management strategies can be evaluated with proposed methodology in this study. Finally, it is expected that this methodology improves the safety of nuclear power plant

  12. Market-oriented management method of coalmine accident hidden dangers

    LIU Zhao-xia; LI Xing-dong; LU Ying; REN Da-wei

    2007-01-01

    By analyzing the problems which exist currently in the accident hidden dangers management of the coal mine, this paper proposed a new kind of management method-"simulating the market", in which an operation pattern of simulating the market to transact hidden troubles was constructed. This method introduces "Market Mechanism"into safe management, and adopts measurable value to describe the hidden dangers such as" human behavior, technique, environment, equipments etc.". It regards the hidden dangers as "the goods produced by labor" which are found out by the safety managers and the security inspectors, then sells as "commodity". By the process of disposing, counterchecking, re-selling, and redisposing. It forms a set of market-oriented closed-form management pattern of coalmine accident hidden dangers. This kind of management method changes the past traditional methods in which the wageworkers treat safety management passively, but to encourage and restrict them to participate in the check-up and improvement of the hidden dangers.

  13. Assessment of two BWR accident management strategies

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

  14. Accident Management Issues within the ARTIST Project

    An experimental project to be performed in the ARTIST (AeRosol Trapping In a Steam generaTor ) facility is planned at the Paul Scherrer Institut to address aerosol retention in the various parts of the steam generator (SG) following a steam generator tube rupture (SGTR) event. The project will study phenomena at the separate effect and integral levels, and also address accident management (AM) issues. Seven distinct phases are foreseen: 1) Aerosol retention in the tube under dry secondary side conditions, 2) Aerosol retention in the near field close to break under dry conditions, 3) Aerosol retention in the bundle far field under dry conditions, 4) Aerosol retention in the separator and dryer under dry conditions, 5) Aerosol retention in the bundle section under wet conditions, 6) Droplet retention in separator and dryer sections and 7) Integral tests to examine overall retention. The prescribed values of the controlling parameters (aerosol size, aerosol type, gas flow velocity, residence time, etc) cover the range expected in severe accident scenarios. The ARTIST facility is well suited to study phenomena relating to AM. Refilling of the SG might be adopted as an AM measure during an accident in which the SG has dried out. For instance, water injection will establish a pool where the incoming aerosols can be scrubbed to various degrees depending on the aerosol characteristics, water depth and subcooling and steam content in the carrier gas flow. Aerosols are expected to be removed mainly through inertial impaction and diffusiophoresis (condensation) in the vicinity of the break. Away from the break, the remaining gas breaks up in smaller bubbles which rise in the pool, and periodically squirt out through the narrow constrictions of the support plates. In this latter phase, aerosol removal is mainly due to inertial mechanisms. There are many questions that need to be resolved before deciding on the efficacy of flooding the secondary side of a dry SG. These include

  15. Level 2 PSA methodology and severe accident management

    The objective of the work was to review current Level 2-PSA (Probabilistic Safety Assessment) methodologies and practices and to investigate how Level 2-PSA can support severe accident management programmes, i.e. the development, implementation, training and optimisation of accident management strategies and measures. For the most part, the presented material reflects the state in 1996. Current Level 2 PSA results and methodologies are reviewed and evaluated with respect to plant type specific and generic insights. Approaches and practices for using PSA results in the regulatory context and for supporting severe accident management programmes by input from level 2 PSAs are examined. The work is based on information contained in: PSA procedure guides, PSA review guides and regulatory guides for the use of PSA results in risk informed decision making; plant specific PSAs and PSA related literature exemplifying specific procedures, methods, analytical models, relevant input data and important results, use of computer codes and results of code calculations. The PSAs are evaluated with respect to results and insights. In the conclusion section, the present state of risk informed decision making, in particular in the level 2 domain, is described and substantiated by relevant examples

  16. Health protection measures after the Chernobyl accident

    The article describes the nutritional measures introduced to protect health after the Chernobyl accident, and the associated costs. The toal value of the reindeer meat, mutton, lamb and goat meat saved as a result of such measures in 1987 amounted to approx. NOK 250 million. The measures cost approx. NOK 60 million. The resulting reduction in the radiation dose level to which the population was exposed was 450 manSv. In 1988, mutton/lamb and goat meat valued at approx. NOK 310 million was saved from contamination by similar measures, which cost approx. NOK 50 million. The resulting dose level reduction was approx. 200 manSv. The relationship (cost/benefit ratio) between the overall cost of the measures taken to reduce radioactivity levels in food and the dose level reduction achieved was acceptable. 11 refs

  17. Hygienic measures during accidents at nuclear power plants

    Problems of radiation protection in case of large-scale accidents at nuclear power plants are discussed. Aims and purposes of protective measures are shown. Ways of radiation factor effects at various accident stages are described as well as corresponding protective measures. Attention is paid to the criteria of decision adoption at various accident development phases. Examples from the Chernobyl accident experience are presented. 10 refs.; 3 tabs

  18. Using MARS to assist in managing a severe accident

    During an accident, information about the current and possible future states of the plant provides guidance for accident managers in evaluating which actions should be taken. However, depending upon the nature of the accident and the stress levels imposed on the plant staff responding to the accident the current and future plant assessments may be very difficult or nearly impossible to perform without supplemental training and/or appropriate tools. The MAAP Accident Response System (MARS) has been developed as a calculational aid to assist the responsible accident management individuals. Specifically MARS provides additional insights on the current and possible future states of the plant during an accident including the influence of operator actions. In addition to serving as a calculational aid, the MARS software can be an effective means for providing supplemental training. The MARS software uses engineering calculations to perform an integral assessment of the plant status including a consistency assessment of the available instrumentation. In addition, it uses the Modular Accident Analysis Program (MAAP) to provide near term predictions of the plant response if corrective actions are taken. This paper will discuss the types of information that are beneficial to the accident manager and how MARS addresses each. The MARS calculational functions include: instrumentation, validation and simulation, projected operator response based on the EOPs, as well as estimated timing and magnitude of in-plant and off-site radiation dose releases. Each of these items is influential in the management of a severe accident. (author)

  19. 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

  20. 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.

  1. Application of PCTRAN-3/U to studying accident management during PWR severe accident

    In order to improve the safety of nuclear power plant, operator action should be taken into account during a severe accident. While it takes a long time to simulate the plant transient behavior under a severe accident in comparison with the design based accident, a transient simulator should have both high speed calculation capability and interactive functions to model the operating procedures. PCTRAN has been developing to be a simple simulator by using a personal computer to simulate plant behavior under an accident condition. While currently available means usually take relatively long time to simulate plant behavior, using a current high-powered personal computer (PC), PCTRAN-3/U code is designed to operate at a speed significantly faster than real-time. The author describes some results of PCTRAN application in studying the efficiency of accident management for a pressurized water reactor (PWR) during an severe accident

  2. Emergency medical management of radiation accident. Lessons learned from the JCO criticality accident

    A criticality accident occurred at the JCO nuclear fuel processing plant in Tokai-mura, Japan at 10:35 am on September 30, 1999. Three workers while working nearby were exposed to high doses of radiation, especially rich in neutron. They suffered from the acute radiation syndrome and two of them were still under medical treatment. This criticality accident taught us significant lessons of radiation protection for the personnels, e.g. physicians, nurses and firemen who are expected to rescue radiation-exposed patients in radiation accidents. In this article, medical management of radiation accident, e.g. treatment of patient, with high-dosed radiation-exposure and with internal contamination of radioactive nuclides and estimation of individual radiation dose, were briefly explained. The Japanese Association for Medical Management of Radiation Accident was founded on August 29, 1997, in order to promote the mutual communication of physicians who have to be engaged in treatment of radiation-exposed patients. (author)

  3. Fundamental studies into the process and system performance of nuclear power plant, measuring and automation engineering for accident management. Final report

    The application of pretentious methods of signal handling like Observer, Kalman Filter and Fuzzy-Logic in safety-related systems is still limited. The main aim of the project was the improvement of the quality and reliability of the measured signals as well as the determination of non measurable safety-related process paramters with the help of these methods. The investigations were realized on the example of the determination of the safety-related parameter level in pressure vessels with two-phase mixture considering the static and dynamic behaviour of the hydrostatical measuring system during accidents (loss of coolant accident). At the beginning of the project the emphasis of the research work was on the reactor of VVER 440 (horizontal steam generator). Further investigations were expanded to VVER 1000 and BWR (reactor pressure vessel). The method of treatment included the following components: Experimental analysis of single effects at the test facility Pressurizer Model, modelling and simulation with the help of the simulation code ATHLET, development of model-based measurement methods on the basis of adapted models, verification of the developed models and methods. On the basis of the results of the experiments algorithms were developed, realizing the following tasks: - Diagnosis of the process state of pressure vessel and level measuring system with the help of Fuzzy-Logic, - correction of the level indicated by the measuring system, - calculation of the non measurable variable mixture level by Observer and Kalman Filter on the basis of linear state space models, - ATHLET-modules simulating hydrostatic level measurement system (VVER 440, VVER 1000, BWR). (orig.)

  4. The management of individuals involved in radiation accidents

    The author defines the objectives and the coverage of two radiation accident courses presented in 1990 by the US Radiation Emergency Assistance Centre and Training Site of the Oak Ridge Associated Universities together with some Australian Medical institutions. It is estimated that the courses, directed towards physicians, radiotherapists and nurses gave plenty practical advices and details on how to go about radiation accident managements. A manual on handling radiation accidents is also to be prepared after the courses

  5. Strategy generator in computerized accident management support system

    An increased interest for research in the field of accident management of nuclear power plants can be noted. Several international programmes have been started in order to be able to understand the basic physical and chemical phenomena in accident conditions. A feasibility study has shown that it would be possible to design and develop a computerized support system for plant staff in accident situations. To achieve this goal the Halden Project has initiated a research programme on Computerized Accident Management Support (CAMS project). The aim is to utilize the capabilities of computerized tools to support the plant staff during the various accident stages. The system will include identification of the accident state, assessment of the future development of the accident and planning of accident mitigation strategies. A prototype is developed to support operators and the Technical Support Centre in decision making during serious accidents in nuclear power plants. A rule based system has been built to take care of the strategy generation. This system assists plant personnel in planning control proposals and mitigation strategies from normal operation to severe accident conditions. The idea of a safety objective tree and knowledge from the emergency procedures have been used. Future prediction requires good state identification of the plant status and some knowledge about the history of some critical variables. The information needs to be validated as well. Accurate calculations in simulators and a large database including all important information from the plant will help the strategy planning. (orig.). (40 refs., 20 figs.)

  6. 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.)

  7. Unconventional sources of plant information for accident management

    An essential element to accident management is having as clear a picture as is practical of the plant status and thus of the accident and its progress. Effective, appropriate decisions to control and mitigate an accident are dependent on making this assessment of the accident. The objective of this paper is to stimulate consideration of unconventional plant information sources through discussion of specific examples. A plant's condition during an accident can be characterized by plant parameters such as temperatures and pressures and by plant system operational status. For example, core damage is associated with increasing temperatures, pressures, and radiation levels in many different systems and plant areas. Reg. Guide 1.97 instrumentation exists to provide information to allow operators to take specified manual actions (Type A), to indicate whether plant safety functions are being accomplished (Type B), to indicate the potential for breach of barriers to fission product release (Type C), to indicate operability of individual safety systems (Type D), and to indicate the magnitude of radioactive material releases (Type E). Reg. Guide 1.97 instrument range requirements, with the exception of pressure instruments, address conditions up to design basis conditions. Pressure instrument range requirements exceed design basis conditions. During a severe accident, some instruments may not see conditions beyond their design basis. Effective accident management includes the ability to establish a consistent picture of the accident by accumulating information from as many sources as is practical. Operability of systems and components, and non-safety related temperature, radiation, pressure, and water-level indication can be used to directly indicate, measure, or infer plant parameters which confirm, augment or replace those otherwise available. Innovative uses of information sources thus serve to increase the diversity and flexibility of accident data available. Both the

  8. Systematic Review of Accident Management Programs - Principles, Experiences

    Although all plants have some form of accident management, there is not always a proper review of the accident management program neither of its products, i.e. the various procedures and guidelines. Moreover, such reviews are often limited to Emergency Operating Procedures (EOPs) and Severe Accident Management Guidelines (SAMG). More complex events, which include large damage on the site, require additional tools and procedures / guidelines. The present paper describes a new review method that covers this larger area and is capable to identify problems and shortcomings, and offers solutions for those. It basically exists of a three-tier approach: 1. interviews with the national regulator and/or the plant to evaluate the scope of the accident management as required by the national regulation and in comparison with international regulation; 2. interviews with the plant staff to discuss the technical basis of the accident management program and its implementation; and 3. observation of an exercise to test the capability of the plant staff to execute the accident management procedures and guidelines, as well as the value of the exercise for such test. The method is an extension of the IAEA 'Review of Accident Management Program which is limited to review of EOPs and SAMG. It is based on extensive experience with plant reviews. (authors)

  9. 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)

  10. Preliminary severe accident management strategies for Wolsong nuclear power plants

    Severe accident management strategies for Wolsong 2,3,4 Nuclear Power Plants are presented. The defense in depth concept, which limits release of radioactive materials out of containment building, is applied to develop these strategies. These strategies are actions to prevent or to mitigate core damage, rupture of calandria vessel, rupture of calandria vault, rupture of containment building, and release of radioactive materials. These strategies are deduced from the results of level 2 PSA for Wolsong NPPs. These preliminary results will be assessed further and proved to be effective to Wolsong Plants. Then these severe accident management strategies can be used to develop severe accident management program for Wolsong NPPs

  11. 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

  12. Implementation of accident management programmes in nuclear power plants

    According to the generally established defence in depth concept in nuclear safety, consideration in plant operation is also given to highly improbable severe plant conditions that were not explicitly addressed in the original design of currently operating nuclear power plants (NPPs). Defence in depth is achieved primarily by means of four successive barriers which prevent the release of radioactive material (fuel matrix, cladding, primary coolant boundary and containment), and these barriers are primarily protected by three levels of design measures: prevention of abnormal operation and failures (level 1), control of abnormal operation and detection of failures (level 2) and control of accidents within the design basis (level 3). If these first three levels fail to ensure the structural integrity of the core, e.g. due to beyond the design basis multiple failures, or due to extremely unlikely initiating events, additional efforts are made at level 4 to further reduce the risks. The objective at the fourth level is to ensure that both the likelihood of an accident entailing significant core damage (severe accident) and the magnitude of radioactive releases following a severe accident are kept as low as reasonably achievable. Finally, level 5 includes off-site emergency response measures, with the objective of mitigating the radiological consequences of significant releases of radioactive material. The implementation of the emergency response is usually dependent upon the type and magnitude of the accident. Good co-ordination between the operator and the responding organizations is needed to ensure the appropriate response. Accident management is one of the key components of effective defence in depth. In accordance with defence in depth, each design level should be protected individually, independently of other levels. This report focuses on the fourth level of defence in depth, including the transitions from the third level and into the fifth level. It describes

  13. Essential severe accident mitigation measures for operating and future PWR's

    material in the containment atmosphere, on the conditions of the core and for effective accident management decisions. These new in-situ sampling technology was developed and implemented to avoid the strong deposition errors of iodine and aerosols in conventional pipe extraction systems. The venting system is introduced for operating plants and can also be used for future plants although it is not required for the EPR. The Sliding Pressure Venting System consists mainly of a venturi scrubber unit with integrated high efficient metal fiber filter followed by means for super sonic throttling and operation under the sliding containment pressure conditions. Due to this special design and operation the system dimensions could be kept small in spite of obtaining high retention rates for aerosols of >99.99% and that for molecular iodide is >99.5%. For the EPR additional measures for maintaining the containment integrity are foreseen: · use of highly reliable dedicated valves for depressurization which supplement normal bleed valves to eliminate high pressure RPV failure · use of a core melt retention device for melt stabilization by means of spreading of the melt within a large compartment adjacent to the reactor pit, followed by flooding, quenching and cooling of the melt from the top and via a bottom cooling structure; · use of a dedicated active two-train containment heat removal system which needs to operate not earlier than 12h after start of the accident

  14. Applying Functional Modeling for Accident Management of Nuclear Power Plant

    Lind, Morten; Zhang, Xinxin

    2014-01-01

    The paper investigate applications of functional modeling for accident management in complex industrial plant with special reference to nuclear power production. Main applications for information sharing among decision makers and decision support are identified. An overview of Multilevel Flow...

  15. Applying Functional Modeling for Accident Management of Nucler Power Plant

    Lind, Morten; Zhang, Xinxin

    2014-01-01

    The paper investigates applications of functional modeling for accident management in complex industrial plant with special reference to nuclear power production. Main applications for information sharing among decision makers and decision support are identified. An overview of Multilevel Flow...

  16. Development of Krsko Severe Accident Management Guidance (SAMG)

    In this lecture development of severe accident management guidances for Krsko NPP are described. Author deals with the history of severe accident management and implementation of issues (validation, review of E-plan and other aspects SAMG implementation guidance). Methods of Westinghouse owners group, of Combustion Engineering owners group, of Babcock and Wilcox owners group, of the BWR owners group, as well as application of US SAMG methodology in Europe and elsewhere are reviewed

  17. Traffic Accident Prediction Model Implementation in Traffic Safety Management

    Wen, Keyao

    2009-01-01

    As one of the highest fatalities causes, traffic accidents and collisions always requires a large amounteffort to be reduced or prevented from occur. Traffic safety management routines therefore always needefficient and effective implementation due to the variations of traffic, especially from trafficengineering point of view apart from driver education.Traffic Accident Prediction Model, considered as one of the handy tool of traffic safety management,has become of well followed with interest...

  18. 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

  19. Regulatory perspective on accident management issues

    Effective response to reactor accidents requires a combination of emergency operations, technical support and emergency response. The NRC and industry have actively pursued programs to assure the adequacy of emergency operations and emergency response. These programs will continue to receive high priority. By contrast, the technical support function has received relatively little attention from NRC and the industry. The results from numerous PRA studies and the severe accident programs of NRC and the industry have yielded a wealth of insights on prevention and mitigation of severe accidents. The NRC intends to work with the industry to make these insights available to the technical support staffs through a combination of guidance, training and periodic drills

  20. Aerosol measurements and nuclear accidents: a reconsideration

    Within its radioactivity environmental monitoring programme, the Commission of the European Communities and in particular its Joint Research Centre wants to encourage the qualitative improvement of radioactivity monitoring. On 3 and 4 December 1987 an experts' meeting has been organized by the Ispra Joint Research Centre in collaboration with the Gesellschaft fuer Aerosolforschung, in order to discuss measuring techniques for radioactive aerosols in the environment in case of a nuclear accident. During the workshop, current practices in routine monitoring programmes in the near and far field of nuclear power plants were confronted with the latest developments in the metrology of aerosols and radioactivity. The need and feasibility of implementing advanced aerosol and radioactivity techniques in routine monitoring networks have been discussed. This publication gives the full text of 12 presentations and a report of the roundtable discussion being held afterwards. It does not intend to give a complete picture of all activities going on in the field of radioactive aerosol metrology; it rather collects a number of common statements of people who approach the problem from quite different directions

  1. Application of simulation techniques for accident management training in nuclear power plants

    Many IAEA Member States operating nuclear power plants (NPPs) are at present developing accident management programmes (AMPs) for the prevention and mitigation of severe accidents. However, the level of implementation varies significantly between NPPs. The exchange of experience and best practices can considerably contribute to the quality, and facilitate the implementation of AMPs at the plants. Various IAEA activities assist countries in the area of accident management. Several publications have been developed which provide guidance and support in establishing accident management at NPPs. The defence in depth concept in nuclear safety requires that, although highly unlikely, beyond design basis and severe accident conditions should also be considered, in spite of the fact that they were not explicitly addressed in the original design of currently operating nuclear power plants (NPPs). Defence in depth is physically achieved by means of four successive barriers (fuel matrix, cladding, primary coolant boundary, and containment) that prevent the release of radioactive material. These barriers are protected by a set of design measures at three levels, including prevention of abnormal operation and failures (level 1), control of abnormal operation and detection of failures (level 2) and control of accidents within the design basis (level 3). Should these first three levels fail to ensure the structural integrity of the core, additional efforts are made at the fourth level of defence in depth in order to further reduce the risks. The objective at level 4 is to ensure that both the likelihood of an accident entailing significant core damage (severe accident) and the magnitude of radioactive releases following a severe accident are kept as low as reasonably achievable. The term 'accident management' refers to the overall range of capabilities of a NPP and its personnel to both prevent and mitigate accident situations that could lead to severe fuel damage in the reactor

  2. The computer aided education and training system for accident management

    Under severe accident conditions of a nuclear power plant, plant operators and technical support center (TSC) staffs will be under a amount of stress. Therefore, those individuals responsible for managing the plant should promote their understanding about the accident management and operations. Moreover, it is also important to train in ordinary times, so that they can carry out accident management operations effectively on severe accidents. Therefore, the education and training system which works on personal computers was developed by Japanese BWR group (Tokyo Electric Power Co.,Inc., Tohoku Electric Power Co. ,Inc., Chubu Electric Power Co. ,Inc., Hokuriku Electric Power Co.,Inc., Chugoku Electric Power Co.,Inc., Japan Atomic Power Co.,Inc.), and Hitachi, Ltd. The education and training system is composed of two systems. One is computer aided instruction (CAI) education system and the other is education and training system with a computer simulation. Both systems are designed to execute on MS-Windows(R) platform of personal computers. These systems provide plant operators and technical support center staffs with an effective education and training tool for accident management. TEPCO used the simulation system for the emergency exercise assuming the occurrence of hypothetical severe accident, and have performed an effective exercise in March, 2000. (author)

  3. Passive depressurization accident management strategy for boiling water reactors

    Highlights: • We proposed two passive depressurization systems for BWR severe accident management. • Sensitivity analysis of the passive depressurization systems with different leakage area. • Passive depressurization strategies can prevent direct containment heating. - Abstract: According to the current severe accident management guidance, operators are required to depressurize the reactor coolant system to prevent or mitigate the effects of direct containment heating using the safety/relief valves. During the course of a severe accident, the pressure boundary might fail prematurely, resulting in a rapid depressurization of the reactor cooling system before the startup of SRV operation. In this study, we demonstrated that a passive depressurization system could be used as a severe accident management tool under the severe accident conditions to depressurize the reactor coolant system and to prevent an additional devastating sequence of events and direct containment heating. The sensitivity analysis performed with SAMPSON code also demonstrated that the passive depressurization system with an optimized leakage area and failure condition is more efficient in managing a severe accident

  4. Emerging framework of safety management after Fukushima accident

    Since the Fukushima accident onset, concerned organizations and experts have tried to identify the causes and effects of the incident. Many have formulated new national regulatory measures to strengthen nuclear safety in an effort to protect the general public to the extent of probabilistic cases of the most severe or extreme accidents. The Japanese government is set to install a regulatory authority, comparable to the US NRC, which is completely independent from the promotion of nuclear energy. An official report of the National Diet (or Senate) of Japan in June of 2012 laments a lack of safety culture and insists the accident could have been prevented if due consideration and attention had been provided. Both France and other European countries have performed stress tests to their operating units, and have identified many areas for improvement including that of their regulatory framework. The US NRC also conducted special inspections of all operating reactors. In addition, the NRC established both near and long term specific goals, and issued a policy statement for streamlining patch worked regulatory framework. It is also applying the Risk informed Defense in Depth Design which includes the extended design basis requirements. The IAEA General Conference adopted a Nuclear Safety Action Plan in September 2011 and organized an International Expert Meeting in March 2012 in order to analyze all relevant technical aspects from the Japanese incident in order to prevent a reoccurrence. Korea is not an exception to this trend. She was swift to conduct a special inspection of operating reactors and is now implementing many scheduled measures. Numerous facts and insights are now available, not only those gained from the Japanese incident, but also those gleaned from experts worldwide concerning a wide array of information. Therefore, this is an opportunistic time to summarize the insights that have been identified with respect to nuclear safety management and to overview

  5. Emerging framework of safety management after Fukushima accident

    Lee, Joo Sang [TUV SUD KOCEN, Yongin (Korea, Republic of); Rawls, Scott [EXCEL, JP (United States)

    2012-10-15

    Since the Fukushima accident onset, concerned organizations and experts have tried to identify the causes and effects of the incident. Many have formulated new national regulatory measures to strengthen nuclear safety in an effort to protect the general public to the extent of probabilistic cases of the most severe or extreme accidents. The Japanese government is set to install a regulatory authority, comparable to the US NRC, which is completely independent from the promotion of nuclear energy. An official report of the National Diet (or Senate) of Japan in June of 2012 laments a lack of safety culture and insists the accident could have been prevented if due consideration and attention had been provided. Both France and other European countries have performed stress tests to their operating units, and have identified many areas for improvement including that of their regulatory framework. The US NRC also conducted special inspections of all operating reactors. In addition, the NRC established both near and long term specific goals, and issued a policy statement for streamlining patch worked regulatory framework. It is also applying the Risk informed Defense in Depth Design which includes the extended design basis requirements. The IAEA General Conference adopted a Nuclear Safety Action Plan in September 2011 and organized an International Expert Meeting in March 2012 in order to analyze all relevant technical aspects from the Japanese incident in order to prevent a reoccurrence. Korea is not an exception to this trend. She was swift to conduct a special inspection of operating reactors and is now implementing many scheduled measures. Numerous facts and insights are now available, not only those gained from the Japanese incident, but also those gleaned from experts worldwide concerning a wide array of information. Therefore, this is an opportunistic time to summarize the insights that have been identified with respect to nuclear safety management and to overview

  6. The DOE technology development programme on severe accident management

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

  7. 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.)

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

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

  9. The Goiania accident waste management - Reconditioning operation

    As a result of an accidental breakage of a 137Cs radiotherapy source, radioactive waste was generated in Goiania-Brazil. It was collected in different types of packaging and removed to a temporary storage site near Abadia de Goias. After four years in open air storage, corrosion was detected in some packages, especially in the 200 1drums. Measures to ensure a safe interim storage were adopted, until a final disposal plan was to be executed. The objective was to make the waste product suitable for the final disposal requests according to Brazilian standards. These measures were concerned mainly with the waste reconditioning. This paper presents the waste management strategy adopted for this operation

  10. Information processing system and neural network utilization for accident management support

    Tuerkcan, E. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Ciftcioglu, Oe. [Istanbul Technical Univ. (Turkey). Faculty of Electrical and Electronic Engineering; Verhoef, J.P. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Ouden, A.C.B. den [Netherlands Energy Research Foundation (ECN), Petten (Netherlands)

    1996-03-01

    Information processing system with data sensor fusion technology together with potential application of neural network is developed. System is designed for operator in the form of Accident Management Support (AMS) with verification and validation (V and V) for cases of severe accident. To this end, primarily noise analysis techniques are used and their merits are merged for exhaustive information extraction in accident cases where the data from sensors may be obscured by drift, modulation so forth or even incomplete. The information from different methodologies are processed in synergetic form (data sensor fusion) by means of statistical distance measures and neural networks with optimal decisions. (orig.).

  11. Information processing system and neural network utilization for accident management support

    Information processing system with data sensor fusion technology together with potential application of neural network is developed. System is designed for operator in the form of Accident Management Support (AMS) with verification and validation (V and V) for cases of severe accident. To this end, primarily noise analysis techniques are used and their merits are merged for exhaustive information extraction in accident cases where the data from sensors may be obscured by drift, modulation so forth or even incomplete. The information from different methodologies are processed in synergetic form (data sensor fusion) by means of statistical distance measures and neural networks with optimal decisions. (orig.)

  12. Recent Developments in Level 2 PSA and Severe Accident Management

    In 1997, CSNI WGRISK produced a report on the state of the art in Level 2 PSA and severe accident management - NEA/CSNI/R(1997)11. Since then, there have been significant developments in that more Level 2 PSAs have been carried out worldwide for a variety of nuclear power plant designs including some that were not addressed in the original report. In addition, there is now a better understanding of the severe accident phenomena that can occur following core damage and the way that they should be modelled in the PSA. As requested by CSNI in December 2005, the objective of this study was to produce a report that updates the original report and gives an account of the developments that have taken place since 1997. The aim has been to capture the most significant new developments that have occurred rather than to provide a full update of the original report, most of which is still valid. This report is organised using the same structure as the original report as follows: Chapter 2: Summary on state of application, results and insights from recent Level 2 PSAs. Chapter 3: Discussion on key severe accident phenomena and modelling issues, identification of severe accident issues that should be treated in Level 2 PSAs for accident management applications, review of severe accident computer codes and the use of these codes in Level 2 PSAs. Chapter 4: Review of approaches and practices for accident management and SAM, evaluation of actions in Level 2 PSAs. Chapter 5: Review of available Level 2 PSA methodologies, including accident progression event tree / containment event tree development. Chapter 6: Aspects important to quantification, including the use of expert judgement and treatment of uncertainties. Chapter 7: Examples of the use of the results and insights from the Level 2 PSA in the context of an integrated (risk informed) decision making process

  13. Program for accident and incident management support, AIMS

    A prototype of an advisory computer program is presented which could be used in monitoring and analyzing an ongoing incident in a nuclear power plant. The advisory computer program, called the Accident and Incident Management Support (AIMS), focuses on processing a set of data that is to be transmitted from a nuclear power plant to a national or regional emergency center during an incident. The AIMS program will assess the reactor conditions by processing the measured plant parameters. The applied model of the power plant contains a level of complexity that is comparable with the simplified plant model that the power plant operator uses. A standardized decay heat function and a steam water property library is used in the integral balance equations for mass and energy. A simulation of the station blackout accident of the Borssele plant is used to test the program. The program predicts successively: (1) the time of dryout of the steam generators, (2) the time of saturation of the primary system, and (3) the onset of core uncovery. The coolant system with the actual water levels will be displayed on the screen. (orig./HP)

  14. Regulatory requirements on accident management and emergency preparedness - concept of nuclear and radiation safety during beyond-design-basis accidents

    Actual practice the and proposals for further activities in the field of Accident Management (AM) in the member countries of the Co-operation Forum of WWER regulators and in Western countries have been assessed. Further the results of the last working group on AM , the overview of interactions of severe accident research and the regulatory positions in various countries, IAEA reports, practice in Switzerland and Finland, were taken into consideration. From this information, the working group derived recommendations on Accident Management. The general proposals correspond to the present state of the art on AM. They do not describe the whole spectra of recommendations on AM for NPPs with WWER reactors. A basis for the implementation of an AM program is given, which could be extended in a follow-up working group. The developments and research concerning AM have to be continued. The positions of various countries with regard to the 'Interactions of severe accident research and the regulatory positions' are given. On the basis of the working group proposals, the WWER regulators could set regulatory requirements and support further developments of AM strategies, making use of the benefits of common features of NPPs with WWER reactors. Concerted actions in the field of AM between the WWER regulators would bundle the development of a unified concept of recommendations and speed up the implementation of AM measures in order to minimise the risks involved in nuclear power generation

  15. Marine Accidents in Northern Nigeria: Causes, Prevention and Management

    Lawal Bello Dogarawa

    2012-01-01

    Boat mishaps tend to be increasing in Nigeria in spite of all regulatory measures which have been taken to prevent and control marine accidents. Boat mishaps could occur anywhere water transportation takes place. However, there is a general impression that water transportation takes place only in the riverine areas located in Southern Nigeria but, this paper reports about marine accident cases in Northern Nigeria. It evaluates the safety measures put in place by operators and other institutio...

  16. CATHARE Assessment of PACTEL LOCA Experiments with Accident Management

    Luben Sabotinov

    2010-01-01

    Full Text Available This paper summarizes the analysis results of three PACTEL experiments, carried out with the advanced thermal-hydraulic system computer CATHARE 2 code as a part of the second work package WP2 (analytical work of the EC project “Improved Accident Management of VVER nuclear power plants” (IMPAM-VVER. The three LOCA experiments, conducted on the Finnish test facility PACTEL (VVER-440 model, represent 7.4% cold leg breaks with combination of secondary bleed and primary bleed and feed and different actuation modes of the passive safety injection. The code was used for both defining and analyzing the experiments, and to assess its capabilities in predicting the associated complex VVER-related phenomena. The code results are in reasonable agreement with the measurements, and the important physical phenomena are well predicted, although still further improvement and validation might be necessary.

  17. The society's measures against serious accidents

    Methods to obtain better preparedness for accidents leading to release of radioactive material are discussed, and recommendations are made developing a better coordination of the many separate efforts that will be made. More efficient ways for training and education and a modernization of the technology and routines used are also suggested

  18. Severe Accident Management System On-line Network SAMSON

    SAMSON is a computational tool used by accident managers in the Technical Support Centers (TSC) and Emergency Operations Facilities (EOF) in the event of a nuclear power plant accident. SAMSON examines over 150 status points monitored by nuclear power plant process computers during a severe accident and makes predictions about when core damage, support plate failure, and reactor vessel failure will occur. These predictions are based on the current state of the plant assuming that all safety equipment not already operating will fail. SAMSON uses expert systems, as well as neural networks trained with the back propagation learning algorithms to make predictions. Training on data from an accident analysis code (MAAP - Modular Accident Analysis Program) allows SAMSON to associate different states in the plant with different times to critical failures. The accidents currently recognized by SAMSON include steam generator tube ruptures (SGTRs), with breaks ranging from one tube to eight tubes, and loss of coolant accidents (LOCAs), with breaks ranging from 0.0014 square feet (1.30 cm2) in size to breaks 3.0 square feet in size (2800 cm2). (author)

  19. Development of Parameter Network for Accident Management Applications

    When a severe accident happens, it is hard to obtain the necessary information to understand of internal status because of the failure or damage of instrumentation and control systems. We learned the lessons from Fukushima accident that internal instrumentation system should be secured and must have ability to react in serious conditions. While there might be a number of methods to reinforce the integrity of instrumentation systems, we focused on the use of redundant behavior of plant parameters without additional hardware installation. Specifically, the objective of this study is to estimate the replaced value which is able to identify internal status by using set of available signals when it is impossible to use instrumentation information in a severe accident, which is the continuation of the paper which was submitted at the last KNS meeting. The concept of the VPN was suggested to improve the quality of parameters particularly to be logged during severe accidents in NPPs using a software based approach, and quantize the importance of each parameter for further maintenance. In the future, we will continue to perform the same analysis to other accident scenarios and extend the spectrum of initial conditions so that we are able to get more sets of VPNs and ANN models to predict the behavior of accident scenarios. The suggested method has the uncertainty underlain in the analysis code for severe accidents. However, In case of failure to the safety critical instrumentation, the information from the VPN would be available to carry out safety management operation

  20. Development of Parameter Network for Accident Management Applications

    Pak, Sukyoung; Ahemd, Rizwan; Heo, Gyunyoung [Kyung Hee Univ., Yongin (Korea, Republic of); Kim, Jung Taek; Park, Soo Yong; Ahn, Kwang Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    When a severe accident happens, it is hard to obtain the necessary information to understand of internal status because of the failure or damage of instrumentation and control systems. We learned the lessons from Fukushima accident that internal instrumentation system should be secured and must have ability to react in serious conditions. While there might be a number of methods to reinforce the integrity of instrumentation systems, we focused on the use of redundant behavior of plant parameters without additional hardware installation. Specifically, the objective of this study is to estimate the replaced value which is able to identify internal status by using set of available signals when it is impossible to use instrumentation information in a severe accident, which is the continuation of the paper which was submitted at the last KNS meeting. The concept of the VPN was suggested to improve the quality of parameters particularly to be logged during severe accidents in NPPs using a software based approach, and quantize the importance of each parameter for further maintenance. In the future, we will continue to perform the same analysis to other accident scenarios and extend the spectrum of initial conditions so that we are able to get more sets of VPNs and ANN models to predict the behavior of accident scenarios. The suggested method has the uncertainty underlain in the analysis code for severe accidents. However, In case of failure to the safety critical instrumentation, the information from the VPN would be available to carry out safety management operation.

  1. Unconventional sources of plant information for accident management

    Oehlberg, R.; Machiels, A.; Chao, J.; Weiss, J. (Electric Power Research Inst., Palo Alto, CA (United States)); True, D.; James, R. (ERIN Engineering and Research, Walnut Creek, CA (United States))

    1992-01-01

    One phase of accident management covers the actions taken during the course of an accident by the plant operating and technical staff to prevent or minimize off-site radiation releases, gain control, and return the plant to a safe state. Inherent in accomplishing these goals is obtaining a clear picture of the nature of the accident and plant status. Development of a consistent and coherent understanding of the accident and plant status requires plant staff to evaluate and interpret data from a wide range of sources. Plant information during an accident can be obtained from the following sources: (1) plant instrumentation, including Regulatory Guide 1.97 instrumentation; and (2) information sources identified in abnormal operations or emergency operations procedures. Probabilistic risk analyses have shown that events involving the loss of key electrical support systems can be significant contributors to core damage. Such events could jeopardize or degrade instrument availability. Plant-specific accident procedures and interpretation of instruments intended for design-basis events may not be applicable in severe accidents. Information sources such as other nuclear steam supply systems (NSSSs) and balance-of-plant (BOP) instrumentation may be available.

  2. Proceedings of the workshop on operator training for severe accident management and instrumentation capabilities during severe accidents

    This Workshop was organised in collaboration with Electricite de France (Service Etudes et Projets Thermiques et Nucleaires). There were 34 participants, representing thirteen OECD Member countries, the Russian Federation and the OECD/NEA. Almost half the participants represented utilities. The second largest group was regulatory authorities and their technical support organisations. Basically, the Workshop was a follow-up to the 1997 Second Specialist Meeting on Operator Aids for Severe Accident Management (SAMOA-2) [Reports NEA/CSNI/R(97)10 and 27] and to the 1992 Specialist Meeting on Instrumentation to Manage Severe Accidents [Reports NEA/CSNI/R(92)11 and (93)3]. It was aimed at sharing and comparing progress made and experience gained from these two meetings, emphasizing practical lessons learnt during training or incidents as well as feedback from instrumentation capability assessment. The objectives of the Workshop were therefore: - to exchange information on recent and current activities in the area of operator training for SAM, and lessons learnt during the management of real incidents ('operator' is defined hear as all personnel involved in SAM); - to compare capabilities and use of instrumentation available during severe accidents; - to monitor progress made; - to identify and discuss differences between approaches relevant to reactor safety; - and to make recommendations to the Working Group on the Analysis and Management of Accidents and the CSNI (GAMA). The meeting confirmed that only limited information is needed for making required decisions for SAM. In most cases existing instrumentation should be able to provide usable information. Additional instrumentation requirements may arise from particular accident management measures implemented in some plants. In any case, depending on the time frame where the instrumentation should be relied upon, it should be assessed whether it is likely to survive the harsh environmental conditions it will be exposed

  3. A Survey of Implementation of Severe Accident Management in Sweden

    A comprehensive program for severe accident mitigation was completed for all Swedish reactors by the end of 1988. This work included development of new accident management procedures and also training programmes for operators . As a complement to the EOP's, knowledge based handbooks have been written for the reactors in Forsmark and Ringhals. They are intended for the emergency control centre in a late stage of a severe accident, when the procedures in the control room no longer are applicable. In a separate project, the impact from certain actions in a short perspective on the long term scenario has been investigated. Results from that work have been used in the development of knowledge based handbooks as decision support for the emergency control centre. For the PWR's in Ringhals the earlier procedures have been replaced by SAMG from WOG (Westinghouse Owners Group) in a project run by a team in Ringhals with support from Westinghouse. In the ongoing APRI-project (a cooperative effort between the Swedish Nuclear Power Inspectorate, the Swedish power utilities and TVO in Finland), accident management has been addressed in a sub-project with focus on validation of SAM strategies and use of results from the research on severe accidents to improve the SAM strategies. An important part of the program for severe accident mitigation was the development of accident management strategies. This work was documented in EOP's and other documentation to be used by the emergency organisation in case of an accident. Personnel at the utilities took an active part in the work mentioned above and also in later improvements such as the FR1PP project and in the development of handbooks for the emergency control centres in Forsmark and Ringhals. Generally, active participation of the end users in the development of documentation for severe accident management has clear advantages. One is that the staff at the plant will have a better insight in the work. To a certain extent the

  4. Application of probabilistic methods to accident analysis at waste management facilities

    Probabilistic risk assessment is a technique used to systematically analyze complex technical systems, such as nuclear waste management facilities, in order to identify and measure their public health, environmental, and economic risks. Probabilistic techniques have been utilized at the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico, to evaluate the probability of a catastrophic waste hoist accident. A probability model was developed to represent the hoisting system, and fault trees were constructed to identify potential sequences of events that could result in a hoist accident. Quantification of the fault trees using statistics compiled by the Mine Safety and Health Administration (MSHA) indicated that the annual probability of a catastrophic hoist accident at WIPP is less than one in 60 million. This result allowed classification of a catastrophic hoist accident as ''not credible'' at WIPP per DOE definition. Potential uses of probabilistic techniques at other waste management facilities are discussed

  5. Seabrook Station Level 2 PRA Update to Include Accident Management

    A ground-breaking study was recently completed as part of the Seabrook Level 2 PRA update. This study updates the post-core damage phenomena to be consistent with the most recent information and includes accident management activities that should be modeled in the Level 2 PRA. Overall, the result is a Level 2 PRA that fully meets the requirements of the ASME PRA Standard with respect to modeling accident management in the LERF assessment and NRC requirements in Regulatory Guide 1.174 for considering late containment failures. This technical paper deals only with the incorporation of operator actions into the Level 2 PRA based on a comprehensive study of the Seabrook Station accident response procedures and guidance. The paper describes the process used to identify the key operator actions that can influence the Level 2 PRA results and the development of success criteria for these key operator actions. This addresses a key requirement of the ASME PRA Standard for considering SAMG. An important benefit of this assessment was the identification of Seabrook specific accident management insights that can be fed back into the Seabrook Station accident management procedures and guidance or the training provided to plant personnel for these procedures and guidance. (authors)

  6. A framework for assessing severe accident management strategies

    Accident management can be defined as the innovative use of existing and or alternative resources, systems and actions to prevent or mitigate a severe accident. Together with risk management (changes in plant operation and/or addition of equipment) and emergency planning (off-site actions), accident management provides an extension of the defense-in-depth safety philosophy for severe accidents. A significant number of probabilistic safety assessments (PSA) have been completed which yield the principal plant vulnerabilities. For each sequence/threat and each combination of strategy there may be several options available to the operator. Each strategy/option involves phenomenological and operational considerations regarding uncertainty. These considerations include uncertainty in key phenomena, uncertainty in operator behavior, uncertainty in system availability and behavior, and uncertainty in available information (i.e., instrumentation). The objective of this project is to develop a methodology for assessing severe accident management strategies given the key uncertainties mentioned above. Based on Decision Trees and Influence Diagrams, the methodology is currently being applied to two case studies: cavity flooding in a PWR to prevent vessel penetration or failure, and drywell flooding in a BWR to prevent containment failure

  7. U.S. nuclear industry perspective on accident management

    The Nuclear Management and Resources Council (NUMARC) serves as the United States nuclear power industry's principal mechanism for conveying industry views, concerns, and policies regarding industry wide regulatory issues to the Nuclear Regulatory Commission (NRC) and other government agencies as appropriate. NUMARC and the Electric Power Research Institute (EPRI), in support of the NUMARC Severe Accident Working Group's (SAWG's) efforts with regard to accident management, has developed a framework for evaluation of plant-specific accident management capabilities. These capabilities fall into one of three main categories: (1) personnel resources (organization, training, communications); (2) systems and equipment (restoration and repair, instrumentation, use of alternatives); and (3) information resources (procedures and guidance, technical information, process information). The purpose of this paper is to describe this framework, its objectives, the five major steps involved and areas to consider further. (orig.)

  8. Managing major chemical accidents in China: Towards effective risk information

    Chemical industries, from their very inception, have been controversial due to the high risks they impose on safety of human beings and the environment. Recent decades have witnessed increasing impacts of the accelerating expansion of chemical industries and chemical accidents have become a major contributor to environmental and health risks in China. This calls for the establishment of an effective chemical risk management system, which requires reliable, accurate and comprehensive data in the first place. However, the current chemical accident-related data system is highly fragmented and incomplete, as different responsible authorities adopt different data collection standards and procedures for different purposes. In building a more comprehensive, integrated and effective information system, this article: (i) reviews and assesses the existing data sources and data management, (ii) analyzes data on 976 recorded major hazardous chemical accidents in China over the last 40 years, and (iii) identifies the improvements required for developing integrated risk management in China.

  9. Applying Functional Modeling for Accident Management of Nuclear Power Plant

    The paper investigate applications of functional modeling for accident management in complex industrial plant with special reference to nuclear power production. Main applications for information sharing among decision makers and decision support are identified. An overview of Multilevel Flow Modeling is given and a detailed presentation of the foundational means-end concepts is presented and the conditions for proper use in modelling accidents are identified. It is shown that Multilevel Flow Modeling can be used for modelling and reasoning about design basis accidents. Its possible role for information sharing and decision support in accidents beyond design basis is also indicated. A modelling example demonstrating the application of Multilevel Flow Modelling and reasoning for a PWR LOCA is presented

  10. Main post-accident management stakes: IRSN's point of view

    Full text of publication follows: Off site management of a radiological crisis covers two phases which need to be clearly distinguished even if there are links between them: emergency phase and recovery phase (also called late or post-accident phase). The presentation will deal with the latter, rather neglected up until recently, but conveying special attention from now on in France and at the international level. It is clear now that the long term management of a radiological or nuclear crisis cannot be reduced to merely site decontamination. Actually, environmental decontamination considerations would be only one amongst other essential economical, social, health, psychological, cultural, and symbolical concerns. This is why off site management of a radiological crisis requires innovative governance, in order to challenge such a complexity. This need for challenge led IRSN to have on the go technical developments and new governance modes reflection. 1) Technical developments: they deal with implementing an organisation, a set of methods, a platform of technical tools which would allow the stakeholders to carry out efficiently their mission during the recovery phase. For example, countermeasures for agricultural and urban rehabilitation are developed within the framework of the 6. PCRDT EURANOS programme. Teams from several countries are involved in common elaboration of rehabilitation strategies based on the best available knowledge. Besides this, simple operational decision aiding tools for the stakeholders (local administration, elected representatives, professional agricultural groups, etc.) are currently developed by IRSN within the framework of the nuclear post-accident exercises. IRSN is also involved in doctrinal reflections about the respective roles of radioactive measurements in the environment and radiological consequences calculation during emergency and recovery phases. Criteria for emergency countermeasures withdrawal are also currently under

  11. Concern on accident management for the Korea next generation reactor

    The Korean Next Generation Reactor (KNGR) is under development to be built after year 2000 in Korea. To enhance its capability of preventing and/or mitigating severe accidents, various safety features are incorporated in its design. Some of them are designed against severe accidents and can be operated based on accident management program (AMP) for the KNGR. In this study, the potential capability of the Safety Depressurization System (SDS) and the Shutdown Cooling System (SCS) to mitigate the consequence of severe accidents was examined by using the MAAP 4.02 code as a preliminary step of the AMP development for the KNGR. The concerned accident sequences are small break loss of coolant accidents (SB LOCAs) with a failure of high pressure safety injection system (HPSIS) and a total loss of feedwater (TLOFW). In the level 1 Probabilistic Safety Assessment (PSA) of the KNGR, the operation of the SDS and SCS was not considered because the failures of the HPSIS and the aggressive secondary side cooling result in core damage based on the success criteria of the level 1 PSA. The analysis results show that the SDS can depressurize the RCS below the shutoff head of the shutdown cooling system (SCS) prior to reactor vessel failure. Although core uncovery and core damage occur early due to the opening of the SDS valves, the MAAP calculation results show that the SCS can reflood the damaged core and that core damage and reactor vessel failure can be mitigated or prevented by the feed-and-bleed operation with those systems. From the analysis results, therefore, it seems that the operation of the SDS and SCS can provide a means of mitigating accident consequences and can be employed as an effective accident management strategy for the KNGR. 5 refs., 6 figs., 4 tabs

  12. The computer aided education and training system for accident management

    The education and training system for Accident Management was developed by the Japanese BWR group and Hitachi Ltd. The education and training system is composed of two systems. One is computer aided instruction (CAI) education system and the education and training system with computer simulations. Both systems are designed to be executed on personal computers. The outlines of the CAI education system and the education and training system with simulator are reported below. These systems provides plant operators and technical support center staff with the effective education and training for accident management. (author)

  13. A systematic process for developing and assessing accident management plans

    This document describes a four-phase approach for developing criteria recommended for use in assessing the adequacy of nuclear power plant accident management plans. Two phases of the approach have been completed and provide a prototype process that could be used to develop an accident management plan. Based on this process, a preliminary set of assessment criteria are derived. These preliminary criteria will be refined and improved when the remaining steps of the approach are completed, that is, after the prototype process is validated through application. 9 refs., 10 figs., 7 tabs

  14. The evolution of computerized displays in accident management

    Key regulations implemented by the NRC in 1982, which included requirements such as upgraded emergency operating procedures, detailed control room design reviews, the addition of a safety parameter display system, and the inclusion of a degreed shift technical advisor as part of the operating staff, have enabled the use of computerized displays to evolve as an integral part of accident management within each of the four main vendor groups. Problems, however, remain to be resolved in the area of technical content, information reliability, and rules for use in order to achieve the goal of more reliable accident management in nuclear power plants

  15. Populations protection and territories management in nuclear emergency and post-accident situation

    This document gathers the slides of the available presentations given during these conference days. Twenty seven presentations out of 29 are assembled in the document and deal with: 1 - radiological and dosimetric consequences in nuclear accident situation: impact on the safety approach and protection stakes (E. Cogez); 2 - organisation of public authorities in case of emergency and in post-event situation (in case of nuclear accident or radiological terror attack in France and abroad), (O. Kayser); 3 - ORSEC plan and 'nuclear' particular intervention plan (PPI), (C. Guenon); 4 - thyroid protection by stable iodine ingestion: European perspective (J.R. Jourdain); 5 - preventive distribution of stable iodine: presentation of the 2009/2010 public information campaign (E. Bouchot); 6 - 2009/2010 iodine campaign: presentation and status (O. Godino); 7 - populations protection in emergency and post-accident situation in Switzerland (C. Murith); 8 - CIPR's recommendations on the management of emergency and post-accident situations (J. Lochard); 9 - nuclear exercises in France - status and perspectives (B. Verhaeghe); 10 - the accidental rejection of uranium at the Socatri plant: lessons learnt from crisis management (D. Champion); 11 - IRE's radiological accident of August 22, 2008 (C. Vandecasteele); 12 - presentation of the CEA's crisis national organisation: coordination centre in case of crisis, technical teams, intervention means (X. Pectorin); 13 - coordination and realisation of environmental radioactivity measurement programs, exploitation and presentation of results: status of IRSN's actions and perspectives (P. Dubiau); 14 - M2IRAGE - measurements management in the framework of geographically-assisted radiological interventions in the environment (O. Gerphagnon and H. Roche); 15 - post-accident management of a nuclear accident - the CODIRPA works (I. Mehl-Auget); 16 - nuclear post-accident: new challenges of crisis expertise (D. Champion); 17 - aid guidebooks

  16. Effect of guidelines on management of head injury on record keeping and decision making in accident and emergency departments.

    Thomson, R.; Gray, J; Madhok, R; Mordue, A.; Mendelow, A D

    1994-01-01

    OBJECTIVE--To compare record keeping and decision making in accident and emergency departments before and after distribution of guidelines on head injury management as indices of implementation. DESIGN--Before (1987) and after (1990) study of accident and emergency medical records. SETTING--Two accident and emergency departments in England. PATIENTS--1144 adult patients with head injury in department 1 (533 in 1987, 613 in 1990) and 734 in department 2 (370, 364 respectively). MAIN MEASURES--...

  17. Plant specific severe accident management - the implementation phase

    Many plants are in the process of developing on-site guidance for technical staff to respond to a severe accident situation severe accident management guidance (SAMG). Once the guidance is developed, the SAMG must be implemented at the plant site, and this involves addressing a number of additional aspects. In this paper, approaches to this implementation phase are reviewed, including review and verification of plant specific SAMG, organizational aspects and integration with the emergency plan, training of SAMG users, validation and self-assessment and SAMG maintenance. Examples draw on experience from assisting numerous plants to implement symptom based severe accident management guidelines based on the Westinghouse Owners Group approach, in Westinghouse, non-Westinghouse and VVER plant types. It is hoped that it will be of use to those plant operators about to perform these activities.(author)

  18. Proceedings of the specialist meeting on selected containment severe accident management strategies

    Twenty papers were presented at the first specialist meeting on Selected Containment Severe Accident management Strategies, held in Stockholm, Sweden, in 1994, half of them dealing with accident management strategies implementation status, half of them with research aspects. The four sessions were: general aspects of containment accident management strategies, hydrogen management techniques, other containment accident management strategies (spray cooling, core catcher...), surveillance and protection of containment function

  19. The circumstances of severe accident measure implementation and 'the residual risk'

    Time-series sequence and direct and root causes of Fukushima Daiichi accident were up to validation of Hatamura's investigation committee on the accident but it would be clear that measure against tsunamis was not good enough. Based on this unprecedented accident, revision of safety design review guide and regulatory requirements of severe accident (SA) measure were under consideration while SA measure had been implemented as public self-safety management by administrative guidance. History of SA measure preparation including the introduction of 'the residual risk' for expansion and upgrade of SA measure in new review guide of seismic design of nuclear power reactor facilities was looked back to learn lessons for better safety operation of nuclear facilities. Nuclear operators established accident management (AM) incorporating appropriate SA measure extracted from probabilistic safety assessment (PSA) in 2002, which had been expanded and reinforced by periodic safety review (PSR). At the revision of regulation in 2003, PSA became requirement of operational safety program but not mandatory as before and lost the chance of regulatory review at the PSR. Extent of SA measure had not been expanded based on latest knowledge of SA research and PSA technology. Evaluation of 'the residual risk' obtained by seismic PSA could not be reported at seismic back check so far because seismic evaluation against ground motion was obliged to be preferred. Safety regulation system based on safety culture of both nuclear operators and regulators should be established for implementation of advanced AM for a certainty. (T. Tanaka)

  20. Development of Integrated Evaluation System for Severe Accident Management

    The objective of the project is twofold. One is to develop a severe accident database (DB) for the Korean Standard Nuclear Power plant (OPR-1000) and a DB management system, and the other to develop a localized computer code, MIDAS (Multi-purpose IntegrateD Assessment code for Severe accidents). The MELCOR DB has been constructed for the typical representative sequences to support the previous MAAP DB in the previous phase. The MAAP DB has been updated using the recent version of MAAP 4.0.6. The DB management system, SARD, has been upgraded to manage the MELCOR DB in addition to the MAAP DB and the network environment has been constructed for many users to access the SARD simultaneously. The integrated MIDAS 1.0 has been validated after completion of package-wise validation. As the current version of MIDAS cannot simulate the anticipated transient without scram (ATWS) sequence, point-kinetics model has been implemented. Also the gap cooling phenomena after corium relocation into the RPV can be modeled by the user as an input parameter. In addition, the subsystems of the severe accident graphic simulator are complemented for the efficient severe accident management and the engine of the graphic simulator was replaced by the MIDAS instead of the MELCOR code. For the user's convenience, MIDAS input and output processors are upgraded by enhancing the interfacial programs

  1. Development of Integrated Evaluation System for Severe Accident Management

    Kim, Dong Ha; Kim, K. R.; Park, S. H.; Park, S. Y.; Park, J. H.; Song, Y. M.; Ahn, K. I.; Choi, Y

    2007-06-15

    The objective of the project is twofold. One is to develop a severe accident database (DB) for the Korean Standard Nuclear Power plant (OPR-1000) and a DB management system, and the other to develop a localized computer code, MIDAS (Multi-purpose IntegrateD Assessment code for Severe accidents). The MELCOR DB has been constructed for the typical representative sequences to support the previous MAAP DB in the previous phase. The MAAP DB has been updated using the recent version of MAAP 4.0.6. The DB management system, SARD, has been upgraded to manage the MELCOR DB in addition to the MAAP DB and the network environment has been constructed for many users to access the SARD simultaneously. The integrated MIDAS 1.0 has been validated after completion of package-wise validation. As the current version of MIDAS cannot simulate the anticipated transient without scram (ATWS) sequence, point-kinetics model has been implemented. Also the gap cooling phenomena after corium relocation into the RPV can be modeled by the user as an input parameter. In addition, the subsystems of the severe accident graphic simulator are complemented for the efficient severe accident management and the engine of the graphic simulator was replaced by the MIDAS instead of the MELCOR code. For the user's convenience, MIDAS input and output processors are upgraded by enhancing the interfacial programs.

  2. Handbook for medical management of persons exposed in radiation accidents

    The document is intended as a rapid reference handbook for the use of physicians who may be called upon to handle the cases of radiation emergency. It deals mainly with the diagnosis and treatment procedures which should be followed by medical officers. The handbook has following sections : basic radiobiology, classification of radiation accidents and preparedness for medical intervention, management of external radiation exposure, management of radioactive contamination, and action plan for handling radiation facilities. It is advisable to have a separate medical unit for proper management of persons exposed in radiation accidents. Infrastructure and facilities required in such a set-up are described. Names and addresses of : (1) physicians in India who have specialized in medical management of radiation injuries, and (2)medical doctors trained in radiation protection and occupational health in different states of India are listed in an appendix. (M.G.B.). 10 refs., figs., tabs

  3. Recommendations on accident management for NPP with WWER

    The work deals with the analysis of practices in the field of beyond design basis accidents (BDBA) management in countries operating WWER type reactors. The recommendations of the working group are presented. The aim is to cooperate the actions of the regulatory bodies for the development of an unified concept for recommendations and to speed up the DBDA management realization for the decreasing of the risk from the nuclear power plant operation

  4. Nuclear emergency preparedness in Germany - an introduction. Pt. 1. Accident management in NPPs

    For the realization of all safety-relevant requirements of the Atomic Energy Act (Atomgesetz, AtG) and their attached legal and sublegal nuclear regulations the design and operation of nuclear power plants in Germany is based on the 'Multi-Level Defense-in-Depth Safety Concept'. Experiences derived from severe accidents and continuously conducted safety research led to development and implementation of strategies and measures of severe accident management step by step in order to recognize plant states beyond the design basis in good time, to control their course and to limit their on-site and off-site consequences effectively. An overview is provided of the integration of severe accident management into the defense-in-depth concept and the on-site technical, organizational and administrative precautionary measures are described. (orig.)

  5. Neural network-based expert system for severe accident management

    This paper presents the results of the second phase of a three-phase Severe Accident Management expert system program underway. The primary objectives of the second phase were to develop and demonstrate four capabilities of neural networks with respect to nuclear power plant severe accident monitoring and prediction. A second objective of the program was to develop an interactive graphical user interface which presented the system's information in an easily accessible and straightforward manner to the user. This paper describes the technical and regulatory foundation upon which the expert system is based and provides a background on the development of a new severe accident management tool. This tool provides data to assist in; (1) planning and developing priorities for recovery actions, (2) evaluating recovery action feasibility, (3) identifying recovery action options, and (4) assessing the timing and possible effects of potential recovery strategies. These performance characteristics represent the goals identified for the Severe Accident Management Strategies Online Network (SAMSON) which is currently under development. 4 refs, 1 fig., 1 tab

  6. Decision-making guide for management of agriculture in the case of a nuclear accident

    For several years, agricultural and nuclear professionals in France have been working on how to manage the agricultural situation in the event of a nuclear accident. This work resulted in measures at both the national (Aube nuclear safety exercises in 2003, INEX3 in 2005) and international levels (EURATOM Programmes). Following on from the European FARMING (FP5) and EURANOS (FP6) works, ACTA', IRSN and six agricultural technical institutes which are specialized in agricultural production and processing network (arable crop [especially cereals, maize, pulses, potatoes and forage crops], fruits and vegetables, vine and wine, livestock farming [cattle, sheep, goats, pigs, poultry]), created a resource adapted to the French context: the Decision-aiding Tool for the Management of Agriculture in case of a Nuclear Accident. Devised for the Ministry of Agriculture services supporting state officials in a radiation emergency, this manual focuses on the early phase following the accident when the state of emergency would make discussion on countermeasures with a large stakeholder panel impossible. Supported by the Ministry of Agriculture and Fisheries and the French Nuclear Safety Authority, this project increased knowledge of post-accident management strategies and made an important contribution to the national think tank set up within the framework of the French Steering Committee for managing the post-event phase of a nuclear accident (CODIRPA). This article describes how the manual evolved throughout the project and the development of new resources. (authors)

  7. Decision-making guide for management of agriculture in the case of a nuclear accident

    For several years, agricultural and nuclear professionals in France have been working on how to manage the agricultural situation in the event of a nuclear accident. This work resulted in measures at both the national (Aube nuclear safety exercises in 2003, INEX3 in 2005) and international levels (EURATOM Programmes). Following on from the European FARMING (FP5) and EURANOS (FP6) works, ACTA', IRSN and six agricultural technical institutes which are specialized in agricultural production and processing network (arable crop [especially cereals, maize, pulses, potatoes and forage crops], fruits and vegetables, vine and wine, livestock farming [cattle, sheep, goats, pigs, poultry]), created a resource adapted to the French context: the Decision-aiding Tool for the Management of Agriculture in case of a Nuclear Accident. Devised for the Ministry of Agriculture services supporting state officials in a radiation emergency, this manual focuses on the early phase following the accident when the state of emergency would make discussion on countermeasures with a large stakeholder panel impossible. Supported by the Ministry of Agriculture and Fisheries and the French Nuclear Safety Authority, this project increased knowledge of post-accident management strategies and made an important contribution to the national think tank set up within the framework of the French Steering Committee for managing the post-event phase of a nuclear accident (CODIRPA). This article describes how the manual evolved throughout the project and the development of new resources

  8. Beyond Design Basis Severe Accident Management as an Element of DiD Concept Strengthening

    The 4th Level of DiD is ensured by management of beyond design basis accidents which is achieved by implementation of the Beyond Design Basis Accidents Management Guidance (BDBAMG) and, if necessary, by additional technical devices and organizational measures at NPP Unit. BDBAMG is located between Levels 3 and 5 in DiD and is related to them. It is connected with Level 3 by means of conditions generated at this Level and according to which BDBAM should be initiated (Level 4). It is associated with Level 5 by conditions which necessitate implementation of Emergency planning. Both types of conditions should be identified in BDBAMG. BDBAs including the phase of severe damage of fuel and protective barriers (severe accidents) in accordance with Russian regulatory framework are a subset of all BDBAs set. In this connection, such accident scenarios meet the representativeness criterion for further analysis and development of Guidance for their management. BDBAMG availability, as it provides robustness of DiD as a whole, is an obligatory condition for obtaining a NPP operational license. In the process of BDBAMG development and implementation a feedback with technical and organizational measures, comprising Level 1 and, to a less extent, Level 2, comes up. BDBAMG verification is an important final stage of its development. Addressing severe accidents, it is a challenging issue for a full scope simulator and may require its software modernization to make it responsive to severe accident phenomena. The existing BDBAMGs should be updated due to NPP Unit modernizations and in conjunction with the latest knowledge on severe accident phenomenology and lessons learnt from known events (e.g. NPP Fukushima). Thus, improvements incorporated in BDBAMG, enhance the strength of DiD. (author)

  9. Evaluation of RCS injection strategy by normal residual heat removal system in severe accident management

    Highlights: • Integrated severe accident analysis model of ALWR RCS, ESF and containment is built. • Large-break loss of coolant accident and loss of feed water accident are analyzed. • Effectiveness of RNS injection strategy and plant system response are investigated. • Impact of RNS injection on hydrogen generation and distribution is evaluated. • Negative impact induced by different RCS depressurization measures is investigated. - Abstract: Severe Accident Management Guidelines (SAMGs) suggests mitigating the consequence of severe accident scenarios by using the non-safety systems if the safety systems are unavailable. For 1000 MWe advanced passive pressurized water reactor (PWR), the normal residual heat removal system (RNS) is proposed to implement the Reactor Coolant System (RCS) injection strategy during severe accidents if safety systems fail. Therefore, evaluation of the effectiveness and negative impact of RNS injection strategy is performed, in which two typical severe accident sequences are selected, which are the typical low-pressure core melt accident sequence induced by Large-break Loss of Coolant Accident (LLOCA) with double-ended guillotine break at cold leg and the typical high-pressure core melt accident induced by Loss of Feed Water (LOFW), to analyze RCS response using the integrated severe accident analysis code. The plant model, including RCS, Engineering Safety Features (ESF), containment and RNS, is built to evaluate the effectiveness of RNS injection by comparing the sequences with and without RCS injection, which shows that RNS injection can terminate core melt progression and maintain core cooling in these accident sequences. However, hydrogen generated during the core reflooding is investigated for the negative impact, which shows that RNS may increase the hydrogen concentration in the containment. For the sequence induced by LOFW, two different RCS depressurization measurements are compared, which shows that opening ADS

  10. Severe accident analysis to verify the effectiveness of severe accident management guidelines for large pressurized heavy water reactor

    Gokhale, O.S., E-mail: onkarsg@barc.gov.in; Mukhopadhyay, D., E-mail: dmukho@barc.gov.in; Lele, H.G., E-mail: hglele@barc.gov.in; Singh, R.K., E-mail: rksingh@barc.gov.in

    2014-10-15

    Highlights: • The progression of severe accident initiated from high pressure scenario of station black out has been analyzed using RELAP5/SCDAP. • The effectiveness of SAMG actions prescribed has been established through analysis. • The time margin available to invoke the SAMG action has been specified. - Abstract: The pressurized heavy water reactor (PHWR) contains both inherent and engineered safety features that help the reactor become resistant to severe accident and its consequences. However in case of a low frequency severe accident, despite the safety features, procedural action should be in place to mitigate the accident progression. Severe accident analysis of such low frequency event provides insight into the accident progression and basis to develop the severe accident management guidelines (SAMG). Since the order of uncertainty in the progression path of severe accident is very high, it is necessary to study the consequences of the SAMG actions prescribed. The paper discusses severe accident analysis for large PHWRs for multiple failure transients involving a high pressure scenario (initiation event like SBO with loss of emergency core cooling system and loss of moderator cooling). SAMG actions prescribed for such a scenario include water injection into steam generator, calandria vessel or calandria vault at different stages of accident. The effectiveness of SAMG actions prescribed has been investigated. It is found that there is sufficient time margin available to the operator to execute these SAMG actions and the progression of severe accident is arrested in all the three cases.

  11. Severe accident analysis to verify the effectiveness of severe accident management guidelines for large pressurized heavy water reactor

    Highlights: • The progression of severe accident initiated from high pressure scenario of station black out has been analyzed using RELAP5/SCDAP. • The effectiveness of SAMG actions prescribed has been established through analysis. • The time margin available to invoke the SAMG action has been specified. - Abstract: The pressurized heavy water reactor (PHWR) contains both inherent and engineered safety features that help the reactor become resistant to severe accident and its consequences. However in case of a low frequency severe accident, despite the safety features, procedural action should be in place to mitigate the accident progression. Severe accident analysis of such low frequency event provides insight into the accident progression and basis to develop the severe accident management guidelines (SAMG). Since the order of uncertainty in the progression path of severe accident is very high, it is necessary to study the consequences of the SAMG actions prescribed. The paper discusses severe accident analysis for large PHWRs for multiple failure transients involving a high pressure scenario (initiation event like SBO with loss of emergency core cooling system and loss of moderator cooling). SAMG actions prescribed for such a scenario include water injection into steam generator, calandria vessel or calandria vault at different stages of accident. The effectiveness of SAMG actions prescribed has been investigated. It is found that there is sufficient time margin available to the operator to execute these SAMG actions and the progression of severe accident is arrested in all the three cases

  12. Implementation of Severe Accident Management Strategy at the Loviisa NPP

    A comprehensive severe accident management (SAM) strategy has been developed by Fortum for the Loviisa NPP in Finland. The strategy ensures reliable prevention and mitigation of containment - threatening phenomena, and it is built around a set of SAM safety functions. This paper focusses on the implementation status of the new SAM approach. We describe how and to what extent the modifications with regards to containment isolation, primary system depressurization, hydrogen mitigation, in-vessel retention of corium, and long-term containment cooling have been carried out. When implementing SAM, it was also necessary to modify the emergency response organisation to include a SAM support team. SAM guidelines, procedures and a SAM Handbook have been written. The automatic containment isolation function has been studied carefully within the SAM project. A successful isolation function is of paramount importance, when radioactive releases from the core can be expected to occur soon. Certain modifications have been carried out so that it is now possible to manually actuate missing isolation signals and to lock isolation status. New local control centres have been built to enable manual closure of certain isolation valves. Several new containment leak-tightness measurements have been installed. New depressurization valves, manually operated relief valves, were installed in 1996 for primary system depressurization purposes. The modifications to the ice condenser doors have been carried out in the years 2000 and 2001. Passive auto-catalytic recombiners have been successfully field-tested in the Loviisa containment atmosphere. We aim for installation in the year 2002. The locations of the glow plugs are being updated in a currently ongoing project. In-vessel retention of molten corium through external cooling of the reactor pressure vessel required certain plant modifications e.g. in order to guarantee access of water to the RPV wall. Most significantly, the support structures

  13. Management of a radiological emergency. Experience feedback and post-accident management

    In France, the organization of crisis situations and the management of radiological emergency situations are regularly tested through simulation exercises for a continuous improvement. Past severe accidents represent experience feedback resources of prime importance which have led to deep changes in crisis organizations. However, the management of the post-accident phase is still the object of considerations and reflections between the public authorities and the intervening parties. This document presents, first, the nuclear crisis exercises organized in France, then, the experience feedback of past accidents and exercises, and finally, the main aspects to consider for the post-accident management of such events: 1 - Crisis exercises: objectives, types (local, national and international exercises), principles and progress, limits; 2 - Experience feedback: real crises (major accidents, other recent accidental situations or incidents), crisis exercises (experience feedback organization, improvements); 3 - post-accident management: environmental contamination and people exposure, management of contaminated territories, management of populations (additional protection, living conditions, medical-psychological follow up), indemnification, organization during the post-accident phase; 4 - conclusion and perspectives. (J.S.)

  14. Proceedings of the first OECD (NEA) CSNI-Specialist Meeting on Instrumentation to Manage Severe Accidents

    OECD member countries have adopted various accident management measures and procedures. To initiate these measures and control their effectiveness, information on the status of the plant and on accident symptoms is necessary. This information includes physical data (pressure, temperatures, hydrogen concentrations, etc.) but also data on the condition of components such as pumps, valves, power supplies, etc. In response to proposals made by the CSNI - PWG 4 Task Group on Containment Aspects of Severe Accident Management (CAM) and endorsed by PWG 4, CSNI has decided to sponsor a Specialist Meeting on Instrumentation to Manage Severe Accidents. The knowledge-basis for the Specialist Meeting was the paper on 'Instrumentation for Accident Management in Containment'. This technical document (NEA/CSNI/R(92)4) was prepared by the CSNI - Principle Working Group Number 4 of experts on January 1992. The Specialist Meeting was structured in the following sessions: I. Information Needs for Managing Severe Accidents, II. Capabilities and Limitations of Existing Instrumentation, III. Unconventional Use and Further Development of Instrumentation, IV. Operational Aids and Artificial Intelligence. The Specialist Meeting concentrated on existing instrumentation and its possible use under severe accident conditions; it also examined developments underway and planed. Desirable new instrumentation was discussed briefly. The interactions and discussions during the sessions were helpful to bring different perspectives to bear, thus sharpening the thinking of all. Questions were raised concerning the long-term viability of current (or added) instrumentation. It must be realized that the subject of instrumentation to manage severe accidents is very new, and that no international meeting on this topic was held previously. One of the objectives was to bring this important issue to the attention of both safety authorities and experts. It could be seen from several of the presentations and from

  15. Accident management advisor system (AMAS): A Decision Aid for Interpreting Instrument Information and Managing Accident Conditions in Nuclear Power Plants

    Accident management can be characterized as the optimized use of all available plant resources to stop or mitigate the progression of a nuclear power plant accident sequence which may otherwise result i n reactor vessel and containment failure. It becomes important under conditions that have low probability of occurring. However, given that these conditions may lead to extremely severe financial consequences and public health effects, it is now recognized that it is important for the plant owners to develop realistic strategies and guidelines. Recent studies have classified accident management strategies as: - the use of alternative resources (i.e., air, water, power), - the use of alternative equipment (i.e., pumps, water lines, generators), the use of alternative actions (i.e., manual depressurization and injection, 'feed and bleed', etc.) The matching of these alternative actions and resources to an actual plant condition represents a decision process affected by a high degree of uncertainty in several of its fundamental inputs. This uncertainty includes the expected accident progression phenomenology (e.g., the issue of high pressure core ejection from the vessel in a PWR plant with possible 'direct containment heating'), as well as the expected availability and behavior of plant systems and of plant instrumentation. To support the accident management decision process with computer-based decision aids, one needs to develop accident progression models that can be stored in a computer knowledge based and retrieved at will for comparison with actual plant conditions, so that these conditions can be recognized and dealt with accordingly. Recent Probabilistic Safety Assessments (PSAs) [1] show the progression of a severe accident through and beyond the core melt stages via multi-branch accident progression trees. Although these 'accident tree models' were originally intended for accident probability assessment purposes, they do provide a basis of initial information

  16. The role of SKI in the severe accident management programme in Sweden

    The Swedish Nuclear Power Inspectorate (SKI) has responsibilities in all regulatory aspects of the licensing and operation of nuclear reactors. The twelve Swedish reactors have all implemented technical as well as procedural features for the avoidance and mitigation of the consequences of severe accidents. Work is presently in progress to further develop accident management as well as to further reinforce the basis of knowledge in order to verify measures taken. In the event of an accident, SKI has a specific duty to provide an independent assessment of the potential course of the accident in order to assist regional authorities in making decisions on emergency actions. This paper accounts for SKI's past and present efforts in the severe accident management programme. In all parts of reactor operation human factor aspects are essential, and so indeed in severe accident management. The paper brings forward these aspects in the SKI programme. In conclusion: In war it is common sense that you can only trust proven equipment and trained organizations. The same applies to Severe Accident Management. Technical equipment must be adequate and operable. Tools must be logical, clean cut, easy to find, easy to use and if possible easy to learn. The organization should be clear with regard to distribution of authority and responsibility, have short links of communication, be easy to mobilize and be staffed with competent and dedicated people who are well trained to their tasks. Preparedness against nuclear accidents must always be a consideration in daily operational work. Ensuring that good conditions exist for accident management is one important objective of SKI's assessment. Another is the analysis of organizational behaviour in emergency situations. A frequent conclusion of accident analysis is the major role played by the human factor. It is not hard to find examples where accident management decisions have been taken too soon, on the basis of insufficient information

  17. 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.

  18. The Assesment Of Radioactive Accident Management On The RSG-GAS

    In the operational reactor facilities include RSG-GAS, safety factor for radioactive accident very important to be prioritized. Till now the anticipate happening radioactive accident on the RSG-GAS threat only by the RSG-GAS Operation Manual. For increasing the working function need to create radioactive accident management by facility level. From studying result which source IAEA guidebook, can be composed the assessment accident management of radioactive the RSG-GAS.The sketching this accident management of radioactive to be hoped can helping P2TRR organization by handling radioactive accident if this moment happen on the RSG-GAS

  19. Development of the french accident management and procedures - role of operators in accident and incident management

    This paper gives a brief overview of the set of emergency operating procedures for French NPPs and the method used to built and validate these procedures. Particular emphasis is put on the role and organisation of the operating team during an incident or accident. (orig.)

  20. A structured approach to individual plant evaluation and accident management

    The need for long term development of accident management programs is acknowledged and the key tool for that development is identified as the IPE Program. The Edison commitment to build an integrated program is cited and the effect on the IPE effort is considered. Edison's integrated program is discussed in detail. The key benefits, realism and long term savings, are discussed. Some of the highly visible products such as neural network artificial intelligence systems are cited

  1. Role of accident analysis in development of severe accident management guidance for multi-unit CANDU nuclear power plants

    This paper discusses the role of accident analysis in support of the development of Severe Accident Management Guidance for domestic CANDU reactors. In general, analysis can identify what types of challenges can be expected during accident progression but it cannot specify when and to what degree accident phenomena will occur. SAMG overcomes these limitations by monitoring the actual values of key plant indicators that can be used directly or indirectly to infer the condition of the plant and by establishing setpoints beyond which corrective action is required. Analysis can provide a means to correlate observed post-accident plant behavior against predicted behaviour to improve the confidence in and quality of accident mitigation decisions. (author)

  2. Development and validation of Maanshan severe accident management guidelines

    Maanshan is a Westinghouse pressurized water reactor Nuclear Power Plant (NPP) located in south Taiwan. The Severe Accident Management Guideline (SAMG) of Maanshan NPP is developed based on the Westinghouse Owners Group (WOG) SAMG. The Maanshan SAMG is developed at the end of 2002. MAAP4 code is used as tool to validate the SAMG strategies. The development process and characteristics of Maanshan SAMG is described. A Station BlackOut (SBO) accident for Maanshan NPP which occurred in March 2001 is cited as a reference case for SAMG validation. A SBO accident is simulated first. The severe accident progression is simulated and the entry condition of SAMG is described. Mitigation actions are then applied to demonstrate the effect of SAMG. A RCS depressurization, RCS injection, and containment hydrogen reduction strategies are used to restore the system to a stable condition as power is recovered. Hot leg creep rupture is occurs during the mitigation action that is not considered in WOG SAMG. The effect of the RCS depressurization, RCS injection, and containment hydrogen reduction strategies are analyzed with MAAP4 code

  3. Measuring patients' experiences in the Accident and Emergency department

    Bos, N.

    2013-01-01

    Two questionnaires were used to measure patients’ experiences in the Accident and Emergency department (A&E). First, the English A&E department questionnaire used in the English National Survey Programme, and after translation in Dutch used in the Netherlands. The second questionnaire concerned the

  4. Irradiation Accidents in Radiotherapy Analyze, Manage, Prevent

    Why do errors occur? How to minimize them? In a context of widely publicized major incidents, of accelerated technological advances in radiotherapy planning and delivery, and of global communication and information resources, this critical issue had to be addressed by the professionals of the field, and so did most national and international organizations. The ISMP, aware of its responsibility, decided as well to put an emphasis on the topic at the occasion of its annual meeting. In this frame, potential errors in terms of scenarios, pathways of occurrence, and dosimetry, will first be examined. The goal being to prioritize error prevention according to likelihood of events and their dosimetric impact. Then, case study of three incidents will be detailed: Epinal, Glasgow and Detroit. For each one, a description of the incident and the way it was reported, its investigation, and the lessons that can be learnt will be presented. Finally, the implementation of practical measures at different levels, intra- and inter institutions, like teaching, QA procedures enforcement or voluntary incident reporting, will be discussed

  5. The technical requirements concerning severe accident management in nuclear power plants

    The Great East Japan Earthquake with a magnitude of 9.0 (The 2011 off the Pacific coast of Tohoku Earthquake) occurred on March 11, 2011, and the beyond design-basis tsunami descended on the Fukushima Daiichi Nuclear Power Plant by the earthquake. Eventually, the core cooling systems of the units 1, 2 and 3 could not operate stably, they all suffered severe accident, and hydrogen explosions were triggered in the reactor buildings of units 1, 3 and 4. In the light of these circumstances, Atomic Energy Society of Japan (AESJ) decided to establish a standard that consolidates the concept of maintaining and improving severe accident management. In the SAM standard, the combination of hardware and software measures based on the risk assessment enables a scientific and rational approach to apply to scenarios of various severe accidents including low-frequency, high-impact events, and assures safety with functionality and flexibility. The SAM standard is already established in March, 2014. After publication of the SAM standard, with regard to effectiveness assessment for accident management and treatment of the uncertainty of severe accident analysis code, for example, the detailed guideline will be prepared as appendices of the standard. (author)

  6. Strategy adopted for the safe management of the waste arising from the Goiania accident

    The radiological accident in Goiania brought on an unexpected radioactive decontamination problem which generated a large volume of waste. The key to a straightforward management of this waste was the definition of a successful strategy to deal with it. To achieve this, several fundamental aspects were taken into account. Among the most important, one can mention the properties of the waste, the infrastructure available for its collection, the decontamination logistics, the motivation and commitment of the workers of different organizations involved in the cleanup tasks, the politically sensitive definition of handling a different kind of waste and the administrative procedures to set up reliable records on the waste collected. In the aftermath of the accident, management of the waste became complex because of the delay in agreeing on and setting up a disposal facility. Four years after the accident, corrosion was detected in some packages and measures were taken to ensure safe interim storage until final disposal. These measures focused on waste reconditioning, the development and implementation of a database containing a detailed inventory of the waste and the development of a national safety evaluation procedure for the final disposal facility. An overview is presented of the management of the waste derived from the Goiania accident, as well as the solutions adopted for final disposal. (author)

  7. Severe accident management (SAM), operator training and instrumentation capabilities - Summary and conclusions

    for SAM. In most cases existing instrumentation should be able to provide usable information. Additional instrumentation requirements may arise from particular accident management measures implemented in some plants. In any case, depending on the time frame where the instrumentation should be relied upon, it should be assessed whether it is likely to survive the harsh environmental conditions it will be exposed to. Though uncertainties still remain in the understanding of some severe accident phenomena, this should not be considered as a de-facto impediment against using simplified models both as operator aids in the course of an accident and as an option of a simulator severe accident mathematical model. These tools, however, should be based on state-of-the-art physics and calibrated using more sophisticated codes. Having the capability for periodic assessment of trends and predictions against real plant parameter evolution, and subsequent correction, is also advised for such tools. Being prepared for the unexpected is the major objective pursued in training, especially when capabilities extend into severe accident situations. When training for severe accidents is contemplated, skill-oriented sessions should be emphasized as they allow evaluating operator reactions in highly perturbed situations. However, it is also advised to increase operator awareness in case of severe accident situations through tailored sessions stressing knowledge of basic phenomena involved in degraded situations. Though computer-based training could well prevail in the long run, table-top exercises as currently implemented by many utilities also bring extremely valuable results

  8. Westinghouse severe accident management guidance overview and current status

    The Westinghouse Owners Group has completed a major development program in Severe Accident Management. This program draws on all presently available sources of information in the field, including in the field, including NRC, NUMARC and EPRI programs, plant specific Individual Plant Examinations and Probabilistic Safety Assessments, and other international activities. The program has developed a full set of Severe Accident Management Guidance (SAMG) applicable to Westinghouse and Westinghouse licensee PWR plant. The SAMG enhances the capabilities of the plant emergency response team for accident sequences that progress to fuel damage, and therefore beyond the range of applicability of present guidance in the form of Emergency Operating Procedures. Since the first draft of SAMG was transmitted officially to the WOG members and the NRC in July 1993, many activities have been carried out by the different organizations involved, and although no significant changes to the SAMG structure have resulted from these activities, several enhancement have been included, mainly from the comments recorded during the generic SAMG validation exercise at the Point Beach plant. With the issue in June 1994 of the revision 0 SAMG, some plants in the U.S. and abroad are already implementing plant specific guidelines. This paper provides an overview of the SAMG package, and also describe the most important comments and feedback from the validation and review efforts. (author)

  9. PWR accident management realated tests: some Bethsy results

    The BETHSY integral test facility which is a scaled down model of a 3 loop FRAMATOME PWR and is currently operated at the Nuclear Center of Grenoble, forms an important part of the French strategy for PWR Accident Management. In this paper the features of both the facility and the experimental program are presented. Two accident transients: a total loss of feedwater and a 2'' cold leg break in case of High Pressure Safety Injection System failure, involving either Event Oriented - or State Oriented-Emergency Operating Procedures (EO-EOP or SO-EOP) are described and the system response analyzed. CATHARE calculation results are also presented which illustrate the ability of this code to adequately predict the key phenomena of these transients. (authors). 13 figs., 11 refs., 2 tabs

  10. Proceedings of the specialist meeting on severe accident management implementation

    The Niantic Specialist meeting was structured around three main themes, one for each session. During the first session, papers from regulators, research groups, designers/owners groups and some utilities discussed the critical decisions in Severe Accident Management (SAM), how these decisions were addressed and implemented in generic SAM guidelines, what equipment and instrumentation was used, what are the differences in national approaches, etc. During the second session, papers were presented by utility specialists that described approaches chosen to specific implementation of the generic guidelines, the difficulties encountered in the implementation process and the perceived likelihood of success of their SAM program in dealing with severe accidents. The third session was dedicated to discussing what are the remaining uncertainties and open questions in SAM. Experts from several OECD countries presented significant perspectives on remaining open issues

  11. Summary and conclusions: Specialist Meeting on Severe Accident Management Implementation

    During the first session of this meeting, regulators, research groups, designers/owners' groups and some utilities discussed the critical decisions in SAM (Severe Accident Management), how these decisions were addressed and implemented in generic SAM guidelines, what equipment and instrumentation was used, what are the differences in national approaches, etc. During the second session, papers were presented by utility specialists that described approaches chosen for specific implementation of the generic guidelines, the difficulties encountered in the implementation process and the perceived likelihood of success of their SAM programme in dealing with severe accidents. The third and final sessions was dedicated to discussing what are the remaining uncertainties and open questions in SAM. Experts from several OECD countries presented significant perspectives on remaining open issues

  12. RASPLAV, Refine accident management strategies during a reactor core meltdown

    Description: OECD RASPLAV Project. The RASPLAV project aimed to refine accident management strategies during a reactor core meltdown; it was completed in June 2000. Little is known about the complex interactions that take place during a core meltdown, so one of the RASPLAV project's primary goals was to develop an understanding of this process. The information gathered during tests at the Kurchatov Institute have allowed scientists to develop models of a core meltdown. These models can be used in the design of new reactors and in refining the accident procedures for existing ones. Two aspects of the issue were considered. First, for existing reactors, where external cooling may not be practicable, the process and time sequence before melt-through were studied. This was to help develop management strategies for severe accidents. Secondly, for future and some existing reactor designs, the project aimed to determine the heat transfer conditions under which cavity flooding can be a viable accident management option. The project was run in two successive phases. The RASPLAV Phase-2 project investigated the progression of a severe accident and in particular the thermal loading imposed by a corium pool on the lower head of a Light Water Reactor (LWR) vessel. It followed an earlier Phase-1 project dedicated mainly to the build-up of the experimental and analytical infrastructure. The project objectives were to obtain relevant data on the physical and thermal behavior of the corium in large-scale tests, to derive thermal-physical property data for various molten core materials, and to investigate the effects of stratification of molten materials. The programme of work involved the use of the large facilities available at the Kurchatov Institute in Russia. Four large-scale tests were carried out and were complemented by a series of smaller-scale experiments, all involving the use of materials representative of power reactor cores. Experiments with these test materials in

  13. Utilization technique of 'radiation management manual in medical field (2012).' What should be learnt from the Fukushima nuclear accident

    From the abstract of contents of the 'Radiation management manual in medical field (2012),' the utilization technique of the manual is introduced. Introduced items are as follows: (1) Exposure management; exposure management for radiation medical workers, patients, and citizens in the medical field, and exposure management for radiation workers and citizens involved in the emergency work related to the Fukushima nuclear accident, (2) Health management; health management for radiation medical workers, (3) Radiation education: Education/training for radiation medical workers, and radiation education for health care workers, (4) Accident and emergency measures; emergency actions involved in the radiation accidents and radiation medicine at medical facilities

  14. Impact of short-term severe accident management actions in a long-term perspective. Final Report

    The present systems for severe accident management are focused on mitigating the consequences of special severe accident phenomena and to reach a safe plant state. However, in the development of strategies and procedures for severe accident management, it is also important to consider the long-term perspective of accident management and especially to secure the safe state of the plant. The main reason for this is that certain short-term actions have an impact on the long-term scenario. Both positive and negative effects from short-term actions on the accident management in the long-term perspective have been included in this paper. Short-term actions are accident management measures taken within about 24 hours after the initiating event. The purpose of short-term actions is to reach a stable status of the plant. The main goal in the long-term perspective is to maintain the reactor in a stable state and prevent uncontrolled releases of activity. The purpose of this short Technical Note, deliberately limited in scope, is to draw attention to potential long-term problems, important to utilities and regulatory authorities, arising from the way a severe accident would be managed during the first hours. Its objective is to encourage discussions on the safest - and maybe also most economical - way to manage a severe accident in the long term by not making the situation worse through inappropriate short-term actions, and on the identification of short-term actions likely to make long-term management easier and safer. The Note is intended as a contribution to the knowledge base put at the disposal of Member countries through international collaboration. The scope of the work has been limited to a literature search. Useful further activities have been identified. However, there is no proposal, at this stage, for more detailed work to be undertaken under the auspices of the CSNI. Plant-specific applications would need to be developed by utilities

  15. Fundamental study on serious accidents and their management in fuel fabrication/enrichment facilities and reprocessing facilities

    The 'Act for the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors' was amended and issued in June 2012 taking into account the lessons derived from the accident of TEPCO Fukushima Daiichi Nuclear Power Plant occurred in March 2011. The main amendments were as follows; Preparation for the management of serious accidents, Introduction of evaluation system for safety improvement, Application of new standards to existing nuclear facility (back-fitting). Japan Nuclear Energy Safety organization (JNES) started this fundamental study on serious accidents and their management, as a safety studying in fuel fabrication/enrichment facilities and reprocessing facilities, for the purpose to contribute to the implementation of new Rules by Nuclear Regulation Authority. From the technical view to be concerned such as fundamental concept of the Rules and applicability of risk-informed regulation, the following 7 subjects were studied: 1) Application concept of the defense in depth to these facilities. 2) Positioning of serious accidents and their management in the defense in depth. 3) Definition of the serious accidents in these facilities. 4) Postulated external events for the study of the serious accidents and their management. 5) Objectives and requirements of the accident management (assurance of reliability). 6) Confirmation logic flow on sequence of the serious accidents and the accident management measures. 7) Applicability of risk information. During the study on these subjects, features of the facilities were clarified at first. Based on concept of the defense in depth, which is the basic principle in safety, and referring to information related to domestic/foreign serious accidents, JNES conducted the fundamental study and made the following suggestions: 1) Definition of the serious accidents of the facilities. The definition is expected to contribute the discussion on new Rules by Nuclear Regulation Authority. 2) Methodology to examine the

  16. The EPR concept for serious accident management, and accompanying research

    An accident, even if the probability of occurrence is so low that it can practically be excluded, must not require any serious external emergency measures, such as evacuation of human populations outside the immediate neighbourhood of the plant. This demand, which in the meantime has also become part of the German article law, creates a new situation for future light water reactors. In addition to the measures which are to reduce the probability of occurrence of serious accidents, a level is introduced which is designed to control the consequences of serious accidents with postulated core meltdown. The introduction of specific measures and design characteristics is a new challenge which cannot be met by industry alone. It is necessary to resort, to a large extent, to present and future research and development work which has been and will be carried out in this area by large-scale research institutions and universities. As regards the EPR, research and development cooperation in this field has been intensified recently. The CEA research centres and the FZKA signed an agreement on information exchange. (orig./HP)

  17. Specific features of RBMK severe accidents progression and approach to the accident management

    Fundamental construction features of the LWGR facilities (absence of common external containment shell, disintegrated circulation circuit and multichannel reactor core, positive vapor reactivity coefficient, high mass of thermally capacious graphite moderator) predetermining development of assumed heavy non-projected accidents and handling them are treated. Rating the categories of the reactor core damages for non-projected accidents and accident types producing specific grope of damages is given. Passing standard non-projected accidents, possible methods of attack accident consequences, as well as methods of calculated analysis of non-projected accidents are demonstrated

  18. The Role Of Industrial Safety Measures In Combating Occupational Hazards And Accidents In India

    Sharmistha Bhattacharjee

    2012-10-01

    Full Text Available The presence of occupational hazards and industrial accidents de-motivates the worker to contribute their best to the organization. The participation of workers in the workplace which promises safety and security fosters teamwork, quality of product high productivity and a good communication between management and the industrial workers. For combating occupational hazards and accidents in an industrial site, safety is necessary and a challenging issue in an industrial environment. Serious technological accidents happens everyday somewhere in the world, causing deaths, injuries and damages to the environment and to the employees Most accidents are caused by people. People are not aware of how to use protective equipments nor are they aware of industrial hygiene and security measures. This paper provides an overview from the secondary sources of data on occupational hazards and accidents, and focuses on the safety and security services and measures provided by the institutions and government to combat the problems to provide an understanding of the situation in Indian context

  19. Specialist meeting on selected containment severe accident management strategies. Summary and conclusions

    The CSNI Specialist Meeting on Selected Containment Severe Accident Management Strategies held in Stockholm, Sweden in June 1994 was organised by the Task Group on Containment Aspects of Severe Accident Management (CAM) of CSNI's Principal Working Group on the Confinement of Accidental Radioactive Releases (PWG4) in collaboration with the Swedish Nuclear Power Inspectorate (SKI). Conclusions and recommendations are given for each of the sessions of the workshops: Containment accident management strategies (general aspects); hydrogen management techniques and other containment accident management techniques; surveillance and protection of containment function

  20. Nuclear Malaysia Disaster Management-Japan Nuclear Accident

    Japan worst Nuclear Accident tragedy due to the earthquake and tsunami, were shocking the world. Malaysia also feels the impact from this disaster. Nuclear Malaysia personnel was mobilize to perform the radiation and contamination monitoring at Malaysian Airport (KLIA and KKIA), environmental monitoring and sampling at Kudat, Sabah, contamination screening centre at Block 13 and also at National Radiology Emergency Centre at AELB. This paper will discuss how this disaster management being performs and its challenge and also the number or personnel and man-hours involved within 1st month after the tragedy. (author)

  1. Improvement of Severe Accident Analysis Computer Code and Development of Accident Management Guidance for Heavy Water Reactor

    Park, Soo Yong; Kim, Ko Ryu; Kim, Dong Ha; Kim, See Darl; Song, Yong Mann; Choi, Young; Jin, Young Ho

    2005-03-15

    The objective of the project is to develop a generic severe accident management guidance(SAMG) applicable to Korean PHWR and the objective of this 3 year continued phase is to construct a base of the generic SAMG. Another objective is to improve a domestic computer code, ISAAC (Integrated Severe Accident Analysis code for CANDU), which still has many deficiencies to be improved in order to apply for the SAMG development. The scope and contents performed in this Phase-2 are as follows: The characteristics of major design and operation for the domestic Wolsong NPP are analyzed from the severe accident aspects. On the basis, preliminary strategies for SAM of PHWR are selected. The information needed for SAM and the methods to get that information are analyzed. Both the individual strategies applicable for accident mitigation under PHWR severe accident conditions and the technical background for those strategies are developed. A new version of ISAAC 2.0 has been developed after analyzing and modifying the existing models of ISAAC 1.0. The general SAMG applicable for PHWRs confirms severe accident management techniques for emergencies, provides the base technique to develop the plant specific SAMG by utility company and finally contributes to the public safety enhancement as a NPP safety assuring step. The ISAAC code will be used inevitably for the PSA, living PSA, severe accident analysis, SAM program development and operator training in PHWR.

  2. Campfire-2000: Comprehensive Accident Management Program Featuring Innovative Research and Engineering for the Year 2000 and Beyond

    The CAMPFIRE-2000 accident management program is being developed at the Korea Atomic Energy Research Institute symphonizing the proven state-of-the-art technologies and newly proposed innovative research and engineering. The ultimate goal of the program is to resolve the plant-specific accident management issues utilizing a coherent, consistent, pragmatic, methodical approach. The program focuses on the preventive measures to maintain reactor core geometry and the mitigative measures to secure containment integrity, should a severe accident take place in a nuclear power plant. CAMPFIRE-2000 consists of strategy assessment methods, guidance and procedures, instrumentation and information, calculational aids and tools, human and organization factors, handbook of accident management, and technical expert system. In particular, the one most immediate issue involves the simulation of the rather rapid cooling of the core debris and the reactor vessel lower head of be Three Mile Island Unit 2 nuclear plant as has recently been identified from post-accident metallurgical testing of the sample specimens. As a top-notch companion experiment for CAMPFIRE-2000, a large-scale, real-material, high pressure system test SONATA-IV is proposed as a multi-lateral, multi-disciplinary project calling for international collaboration to investigate the potentially inherent, naturally-occurring in-vessel cooling mechanism from the very relevant severe accident management perspective

  3. Contributions to elaboration of concept and measures for optimized management of beyond-design-basis accidents in German LWR power plants

    In the present Project SR 2227 ordered by the Federal Office for Radiation Protection (BfS) within the framework of the Nuclear Regulatory Investigation Program of the Federal Ministry for Environment, Nature Conservation and Nuclear Safety (BMU), major contributions were worked out with regard to the concept and measures for an optimum influencing control of event sequences beyond the design basis of nuclear power plants with light water reactors. The studies dealt with extremely unlikely conditions under which core damage is to be expected due to the boundary conditions postulated or already has occurred. A total of 7 different basic scenarios were analysed with the MELCOR integral code for the PWR reference plant. These concerned LOCAs with small and large leaks in the reactor cooling system (RCS) and transients at low and high RCS pressure. The earliest moment of core destruction was calculated to occur after about half an hour, the latest one after 5.5 hours. The highest rate of H2 formation was determined for cases involving a rather slow progression of core destruction. The retention of released fission products in the RCS strongly depends on the release path into the containment. (orig./GL)

  4. WWER Technical Support Center and Training of its Staff for Severe Accident Management

    The Russian Utility organization Concern Rosenergoatom (REA) has well developed multi-level system of prevention and liquidation of emergency situations at nuclear power plants. This system covers all aspects related to beyond design accidents - from the technical support of the plant personnel to the measures for protection of the population and environment. In case a radiation dangerous situation or accident at a NPP occurred, the urgent help is being performed by the OPAS group, which coordinates the activities of forces and means participating in localization and liquidation of accident. Technical and information needs of the OPAS group is assured by Crisis center of REA (CC) with its Expert group. The task of CC is the development of the technical recommendations for the plant personnel on the accident management measures aimed to prevent the severe accident or to restrict its consequences. This task is being solved by Expert group (EG) of Crisis center in interaction with the Technical support centers (TSC) established in different design and scientific organizations (NSSS General designer, NPP General designer, Scientific leader of NPP design, institutes of Academy of Sciences, etc). Each TSC is being considered as a constituent of Rosenergoatom CC. Such Technical support center for WWER nuclear power plants (WWER TCS) has been established in OKB Gidropress some years ago. Three modes of WWER TSC operation (and, accordingly, its interaction with REA CC) are defined: normal operation, increased readiness and emergency situation. In case of beyond design accident on a plant, WWER TSC under request of REA CC will develop the recommendations for CC Expert group aimed to prevent the accident progression to the severe phase or to restrict the severe accident consequences, if it nevertheless has occurred. In chapter 2 of the present paper, place and role of WWER TSC in general system of emergency response of Rosenergoatom is highlighted. TSC structure, functions of

  5. Construction safety: Can management prevent all accidents or are workers responsible for their own actions?

    The construction industry has struggled for many years with the answer to the question posed in the title: Can Management Prevent All Accidents or Are Workers Responsible for Their Own Actions? In the litigious society that we live, it has become more important to find someone open-quotes at faultclose quotes for an accident than it is to find out how we can prevent it from ever happening again. Most successful companies subscribe to the theme that open-quotes all accidents can be prevented.close quotes They institute training and qualification programs, safe performance incentives, and culture-change-driven directorates such as the Voluntary Protection Program (VPP); yet we still see construction accidents that result in lost time, and occasionally death, which is extremely costly in the shortsighted measure of money and, in real terms, impact to the worker''s family. Workers need to be properly trained in safety and health protection before they are assigned to a job that may expose them to safety and health hazards. A management committed to improving worker safety and health will bring about significant results in terms of financial savings, improved employee morale, enhanced communities, and increased production. But how can this happen, you say? Reduction in injury and lost workdays are the rewards. A decline in reduction of injuries and lost workdays results in lower workers'' compensation premiums and insurance rates. In 1991, United States workplace injuries and illnesses cost public and private sector employers an estimated $62 billion in workers'' compensation expenditures

  6. Identification and evaluation of PWR in-vessel severe accident management strategies

    This reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core degradation and prior to RPV failure. A parallel project at BNL has been considering strategies applicable to the ex-vessel phase of PWR severe accidents

  7. Generalities on nuclear accidents and their short-dated and middle-dated management

    All the nuclear activities present a radiation risk. The radiation exposure of the employees or the public, may occur during normal activity or during an accident. The IRSN realized a document on this radiation risk and the actions of protection. The sanitary and medical aspects of a radiation accident are detailed. The actions of the population protection during an accident and the post accident management are also discussed. (A.L.B.)

  8. Developement of integrated evaluation system for severe accident management

    Kim, Dong Ha; Kim, H. D.; Park, S. Y.; Kim, K. R.; Park, S. H.; Choi, Y.; Song, Y. M.; Ahn, K. I.; Park, J. H

    2005-04-01

    The scope of the project includes four activities such as construction of DB, development of data base management tool, development of severe accident analysis code system and FP studies. In the construction of DB, level-1,2 PSA results and plant damage states event trees were mainly used to select the following target initiators based on frequencies: LLOCA, MLOCA, SLOCA, station black out, LOOP, LOFW and SGTR. These scenarios occupy more than 95% of the total frequencies of the core damage sequences at KSNP. In the development of data base management tool, SARD 2.0 was developed under the PC microsoft windows environment using the visual basic 6.0 language. In the development of severe accident analysis code system, MIDAS 1.0 was developed with new features of FORTRAN-90 which makes it possible to allocate the storage dynamically and to use the user-defined data type, leading to an efficient memory treatment and an easy understanding. Also for user's convenience, the input (IEDIT) and output (IPLOT) processors were developed and implemented into the MIDAS code. For the model development of MIDAS concerning the FP behavior, the one dimensional thermophoresis model was developed and it gave much improvement to predict the amount of FP deposited on the SG U-tube. Also the source term analysis methodology was set up and applied to the KSNP and APR1400.

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

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

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

    Ahn, Kwang Il; Kim, Dong Ha

    2003-01-01

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

  11. Summary and conclusions of the specialist meeting on severe accident management programme development

    The CSNI Specialist meeting on severe accident management programme development was held in Rome and about seventy experts from thirteen countries attended the meeting. A total of 27 papers were presented in four sessions, covering specific aspects of accident management programme development. It purposely focused on the programmatic aspects of accident management rather than on some of the more complex technical issues associated with accident management strategies. Some of the major observations and conclusions from the meeting are that severe accident management is the ultimate part of the defense in depth concept within the plant. It is function and success oriented, not event oriented, as the aim is to prevent or minimize consequences of severe accidents. There is no guarantee it will always be successful but experts agree that it can reduce the risks significantly. It has to be exercised and the importance of emergency drills has been underlined. The basic structure and major elements of accident management programmes appear to be similar among OECD member countries. Dealing with significant phenomenological uncertainties in establishing accident management programmes continues to be an important issue, especially in confirming the appropriateness of specific accident management strategies

  12. Environmental radioactivity measurements at BNL following the Chernobyl accident

    Measurements are reported of the concentrations at Berkeley in Gloucestershire of radioactivity in the air, rainwater, tap water, soil, herbage and fresh vegetables for the period 29 April 1986 to 15 May 1986, following the Chernobyl Power Station accident. Data for up to 18 gamma emitting isotopes are reported, together with some limited actinide-in-air measurements. Deposition velocities are calculated and an assessment is presented of the sensitivity of the techniques employed. Some data are also included on the gaseous composition of the cloud and the isotope dependent dose rate from deposition. (author)

  13. Accident evolution and barrier function and accident evolution management modeling of nuclear power plant incidents

    Every analysis of an accident or an incident is founded on a more or less explicit model of what an accident is. On a general level, the current approach models an incident or accident in a nuclear power plant as a failure to maintain a stable state with all variables within their ranges of stability. There are two main sets of subsystems in continuous interaction making up the analyzed system, namely the human-organizational and the technical subsystems. Several different but related approaches can be chosen to model an accident. However, two important difficulties accompany such modeling: the high level of system complexity and the very infrequent occurrence of accidents. The current approach acknowledges these problems and focuses on modeling reported incidents/accidents or scenarios selected in probabilistic risk assessment analyses to be of critical importance for the safety of a plant

  14. Role of the man-machine interface in accident management strategies

    First, this paper gives a short general review on important safety issues in the field of man-machine interaction as expressed by important nuclear safety organisations. Then follows a summary discussion on what constitutes a modern Man-Machine Interface (MMI) and what is normally meant with accident management and accident management strategies. Furthermore, the paper focuses on three major issues in the context of accident management. First, the need for reliable information in accidents and how this can be obtained by additional computer technology. Second, the use of procedures is discussed, and basic MMI aspects of computer support for procedure presentation are identified followed by a presentation of a new approach on how to computerise procedures. Third, typical information needs for characteristic end-users in accidents, such as the control room operators, technical support staff and plant emergency teams, is discussed. Some ideas on how to apply virtual reality technology in accident management is also presented

  15. Development Process of Plant-specific Severe Accident Management Guidelines for Wolsong Nuclear Power Plants

    A severe accident, which occurred at the TMI in 1979 and Chernobyl in 1986, is an accident that exceeds design basis accidents and leads to significant core damage. The severe accident is the low possibility of occurrence but the high severity. To mitigate the consequences of the severe accidents, Korean Nuclear Safety Committee declared the Severe Accident Policy in 2001, which requested the development of Severe Accident Management Guidelines (SAMGs) for operating plants. SAMG is a symptom-based guidance that takes a set of actions to alleviate the outcomes of severe accidents and to get into the safe stable plant condition. The purpose of this paper is to presents the strategic development process of the PHWR SAMG. The guidelines consist of 5 categories: an emergency guide for the main control room (MCR) operators, a strategy implementing guide for the technical support center (TSC), six mitigation guides, a monitoring guide, and a termination guide

  16. Development of the MIDAS GUI environment for severe accident management and analyses

    MIDAS is being developed at KAERI as an integrated severe accident analysis code with existing model modification and new model addition. Also restructuring of the data transfer scheme is going on to improve user's convenience. In this paper, various MIDAS GUI systems which are input management system IEDIT, variable plotting system IPLOT, severe accident training simulator SATS, and online guidance module HyperKAMG, are introduced. In addition, detail functions and usage of these systems for severe accident management and analyses are described

  17. Radiation accidents and their management: emphasis on the role of nuclear medicine professionals

    Bomanji, Jamshed B.; NOVRUZOV, Fuad; Vinjamuri, Sobhan

    2014-01-01

    Large-scale radiation accidents are few in number, but those that have occurred have subsequently led to strict regulation in most countries. Here, different accident scenarios involving exposure to radiation have been reviewed. A triage of injured persons has been summarized and guidance on management has been provided in accordance with the early symptoms. Types of casualty to be expected in atomic blasts have been discussed. Management at the scene of an accident has been described, with e...

  18. The management of risk to society from potential accidents

    The main report of the United Kingdom Atomic Energy Authority (UKAEA) Working Group on Risks to Society from Potential Major Accidents is presented. It is the outcome of a study by AEA Technology, the trading name of the UKAEA, in support of its own decision-making on risk management of the nuclear plants and laboratories it controls. The principles underlying decisions on social risk are of much broader applicability, however. The report is prefaced by an Executive Summary which is intended to be a stand-alone summary of the results of the study. The topics covered include: an examination of the nature of risk; the distinction to be drawn between individual and societal risk; existing risks; risk estimation; goals and targets as defined in terms of acceptance, tolerability and comparison between risks; regulations relating to risk targets; risk management decisions in theory and practice; societal risk management. A final chapter brings together the conclusions and recommendations from the preceding nine with respect to risk estimation, evaluation, management and overall approach. Two appendices deal with cost benefit analysis and provide a glossary and acronyms. (UK)

  19. Comprehensive Health Risk Management after the Fukushima Nuclear Power Plant Accident.

    Yamashita, S

    2016-04-01

    Five years have passed since the Great East Japan Earthquake and the subsequent Fukushima Daiichi Nuclear Power Plant accident on 11 March 2011. Countermeasures aimed at human protection during the emergency period, including evacuation, sheltering and control of the food chain were implemented in a timely manner by the Japanese Government. However, there is an apparent need for improvement, especially in the areas of nuclear safety and protection, and also in the management of radiation health risk during and even after the accident. Continuous monitoring and characterisation of the levels of radioactivity in the environment and foods in Fukushima are now essential for obtaining informed consent to the decisions on living in the radio-contaminated areas and also on returning back to the evacuated areas once re-entry is allowed; it is also important to carry out a realistic assessment of the radiation doses on the basis of measurements. Until now, various types of radiation health risk management projects and research have been implemented in Fukushima, among which the Fukushima Health Management Survey is the largest health monitoring project. It includes the Basic Survey for the estimation of external radiation doses received during the first 4 months after the accident and four detailed surveys: thyroid ultrasound examination, comprehensive health check-up, mental health and lifestyle survey, and survey on pregnant women and nursing mothers, with the aim to prospectively take care of the health of all the residents of Fukushima Prefecture for a long time. In particular, among evacuees of the Fukushima Nuclear Power Plant accident, concern about radiation risk is associated with psychological stresses. Here, ongoing health risk management will be reviewed, focusing on the difficult challenge of post-disaster recovery and resilience in Fukushima. PMID:26817782

  20. Influence diagrams and decision trees for severe accident management

    A review of relevant methodologies based on Influence Diagrams (IDs), Decision Trees (DTs), and Containment Event Trees (CETs) was conducted to assess the practicality of these methods for the selection of effective strategies for Severe Accident Management (SAM). The review included an evaluation of some software packages for these methods. The emphasis was on possible pitfalls of using IDs and on practical aspects, the latter by performance of a case study that was based on an existing Level 2 Probabilistic Safety Assessment (PSA). The study showed that the use of a combined ID/DT model has advantages over CET models, in particular when conservatisms in the Level 2 PSA have been identified and replaced by fair assessments of the uncertainties involved. It is recommended to use ID/DT models complementary to CET models. (orig.)

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

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

    2003-10-01

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

  2. Measures for reduction of severe accident consequences: Comprehensive evaluation of the results sponsored by the BMI

    A number of analytical studies were initial in the past by the Federal Ministry of Interior (BMI) of FRG, to investigate the potential of additional constructive measures for risk reduction. Those measures were proposed especially against uncontrolled overpressurization of the containment due to continuous gas/steam generation, penetration of the foundation of the reactor building by melt-concrete interaction, and failure of the containment due to violent hydrogen combustion. This report gives an overview about those studies and summarizes their results. Concerning uncontrolled overpressurization, only filtered venting may be a reasonable measure, while it seems to make not much sense, to look at measures against penetration of the foundation like 'core-catcher' in further detail. To prevent hydrogen combustion with severe consequences, several potential possibilities exist, but none of them can be considered as a safe measure. Additional analysis concerning hydrogen distribution and combustion in a multi-compartment containment are necessary. All studies mentioned in this report, deal with additional constructive measures to mitigate the consequences of severe accidents. Up to day in FRG, the potential of accident prevention and mitigation of its consequences by still or again operable and already existing systems of a plant have not been investigated in detail. As indicated by first results, the use of those systems in the frame of an appropriate accident management may have a large potential for risk reduction. (orig.)

  3. Waste management facility accident analysis (WASTE ACC) system: software for analysis of waste management alternatives

    This paper describes the Waste Management Facility Accident Analysis (WASTEunderscoreACC) software, which was developed at Argonne National Laboratory (ANL) to support the US Department of Energy's (DOE's) Waste Management (WM) Programmatic Environmental Impact Statement (PEIS). WASTEunderscoreACC is a decision support and database system that is compatible with Microsoft reg-sign Windows trademark. It assesses potential atmospheric releases from accidents at waste management facilities. The software provides the user with an easy-to-use tool to determine the risk-dominant accident sequences for the many possible combinations of process technologies, waste and facility types, and alternative cases described in the WM PEIS. In addition, its structure will allow additional alternative cases and assumptions to be tested as part of the future DOE programmatic decision-making process. The WASTEunderscoreACC system demonstrates one approach to performing a generic, systemwide evaluation of accident risks at waste management facilities. The advantages of WASTEunderscoreACC are threefold. First, the software gets waste volume and radiological profile data that were used to perform other WM PEIS-related analyses directly from the WASTEunderscoreMGMT system. Second, the system allows for a consistent analysis across all sites and waste streams, which enables decision makers to understand more fully the trade-offs among various policy options and scenarios. Third, the system is easy to operate; even complex scenario runs are completed within minutes

  4. Analytical support for SAMG development as a part of accident management

    The decision to built up and implement a comprehensive Accident Management Program applying best world-wide knowledge made during last year at Temelin. A small group of engineers dedicated to Accident Management was formed at Temelin NPP as a part of the plant organisation scheme. A short summary of these activities performed by this group is presented. (author)

  5. Procedures for field measurements in the case of nuclear accident

    Very simplified, reduced and shorted procedures for main objectives of emergency field monitoring in case of nuclear accident are given only. They could be implemented in Croatia using resources nowadays available. Procedures for gamma/beta dose rates in plume and ground deposition survey and unknown situation evaluation, procedures for alpha and gamma/beta surface contamination measurement, field personnel/equipment contamination and decontamination measurement as well as for in-situ gamma spectrometry measurements are presented. Purpose, short discussion, general precautions and limitations as well as basic equipment and supplies needed are given for all of procedures discussed also. Only measuring steps are given with more details in form of short and clear instructions. (author)

  6. Nuclear Measurement Technologies and Solutions Implemented during Nuclear Accident at Fukushima

    Fukushima accident imposed a stretch to nuclear measurement operational approach requiring in such emergency situation: fast concept development, fast system integration, deployment and start-up in a very short time frame. This paper is describing the Nuclear Measurement that AREVA-BUNM (CANBERRA) has realized and foresight at Fukushima accident site describing the technical solution conceived developed and deployed at Fukushima NPP for the process control of the treatment system of contaminated water. A detailed description of all levels design choices, from detection technologies to system architecture is offer in the paper as well as the read-out and global data management system. This paper describes also the technical choices executed and put in place to overcome the challenges related to the high radiological contamination on site. (authors)

  7. A preliminary study for the implementation of general accident management strategies

    Yang, Soo Hyung; Kim, Soo Hyung; Jeong, Young Hoon; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1997-12-31

    To enhance the safety of nuclear power plants, implementation of accident management has been suggested as one of most important programs. Specially, accident management strategies are suggested as one of key elements considered in development of the accident management program. In this study, generally applicable accident management strategies to domestic nuclear power plants are identified through reviewing several accident management programs for the other countries and considering domestic conditions. Identified strategies are as follows; 1) Injection into the Reactor Coolant System, 2) Depressurize the Reactor Coolant System, 3) Depressurize the Steam Generator, 4) Injection into the Steam Generator, 5) Injection into the Containment, 6) Spray into the Containment, 7) Control Hydrogen in the Containment. In addition, the systems and instrumentation necessary for the implementation of each strategy are also investigated. 11 refs., 3 figs., 3 tabs. (Author)

  8. Unconventional sources of plant information for accident management

    Oehlberg, R. (Electric Power Research Inst., Palo Alto, CA (United States). Nuclear Power Div.); Machiels, A. (Electric Power Research Inst., Palo Alto, CA (United States). Nuclear Power Div.); Chao, J. (Electric Power Research Inst., Palo Alto, CA (United States). Nuclear Power Div.); Weiss, J. (Electric Power Research Inst., Palo Alto, CA (United States). Nuclear Power Div.); True, D. (ERIN Engineering and Research, Inc., Walnut Creek, CA (United States)); James, R. (ERIN Engineering and Research, Inc., Walnut Creek, CA (United States))

    1992-07-01

    The paper highlighted that other information sources can help to augment and confirm data available from dedicated accident instrumentation such as Reg. Guide 1.97 Instrumentation: inferences of plant status are possible from measurements and measurement trends obtained from instruments not expected to function, observations of system or component operability/inoperability, and observations of locally harsh environmental conditions. Detailed plant-specific examples are given, e.g. regarding the reactor pressure and level indication in BWRs, or the reactor cavity temperature indication on WE-type PWRs which the authors speculate may yield information related to vessel and core temperature. The authors advocate that others look at their information sources in a creative way. (orig.)

  9. Proceedings of the Specialist Meeting on Severe Accident Management Programme Development

    Effective Accident Management planning can produce both a reduction in the frequency of severe accidents at nuclear power plants as well as the ability to mitigate a severe accident. The purpose of an accident management programme is to provide to the responsible plant staff the capability to cope with the complete range of credible severe accidents. This requires that appropriate instrumentation and equipment are available within the plant to enable plant staff to diagnose the faults and to implement appropriate strategies. The programme must also provide the necessary guidance, procedures, and training to assure that appropriate corrective actions will be implemented. One of the key issues to be discussed is the transition from control room operations and the associated emergency operating procedures to a technical support team approach (and the associated severe accident management strategies). Following a proposal made by the Senior Group of Experts on Severe Accident Management (SESAM), the Committee on the Safety of Nuclear Installations decided to sponsor a Specialist Meeting on Severe Accident Management Programme Development. The general objectives of the Specialist Meeting were to exchange experience, views, and information among the participants and to discuss the status of severe accident management programmes. The meeting brought together utilities, accident management programme developers, personnel training programme developers, regulators, and researchers. In general, the tone of the Specialist Meeting - designed to promote progress, as contrasted with conferences or symposia where the state-of-the-art is presented - was to be rather practical, and focus on accident management programme development, applications, results, difficulties and improvements. As shown by the conclusions of the meeting, there is no doubt that this objective was widely attained

  10. Measuring and Managing Knowledge

    Housel, Thomas; Bell, Arthur H.

    2001-01-01

    Managing and measuring knowledge teaches through the case method, with extended discussion and investigation of high-interest business scenarios from areas of health management, investment, the Internet, telecommunications, computer technologies, food industry management, heavy industry, and a variety of service industries. In each case, readers learn how new tools of knowledge management can positively impact bottom-line profits and overall business strategy. Readers conclude the businesse...

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

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

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

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

  13. Overview of training methodology for accident management at nuclear power plants

    Many IAEA Member States operating nuclear power plants (NPPs) are at present developing accident management programmes (AMPs) for the prevention and mitigation of severe accidents. However, the level of implementation varies significantly between NPPs. The exchange of experience and best practices can considerably contribute to the quality and facilitate the implementation of AMPs at the plants. The main objective of this publication is to describe available material and technical support tools that can be used to support training of the personnel involved in the accident management (AM), and to highlight the current status of their application. The focus is on those operator aids that can help the plant personnel to take correct actions during an emergency to prevent and mitigate consequences of a severe accident. The second objective is to describe the available material for the training courses of those people who are responsible of the AMP development and implementation of an individual plant. The third objective is to collect a compact set of information on various aspects of AM training into a single publication. In this context, the AM personnel includes both the plant staff responsible for taking the decision and actions concerning preventive and mitigative AM and the persons involved in the management of off-site releases. Thus, the scope of this publication is on the training of personnel directly involved in the decisions and execution of the SAM actions during progression of an accident. The integration of training into the AMP development and implementation is summarized. The technical AM support tools and material are defined as operator aids involving severe accident guidelines, various computational aids and computerized tools. The operator aids make also an essential part of the training tools. The simulators to be applied for the AM training have been developed or are under development by various organizations in order to support the training on

  14. Knowledge data base for severe accident management of nuclear power plants

    For the reinforcement of the safety of NPPs, the continuous efforts are very important to take in the up-to-date scientific and technical knowledge positively and to reflect them into the safety regulation. The purpose of this present study is to gather effectively the scientific and technical knowledge about the severe accident (SA) phenomena and the accident management (AM) for prevention and mitigation of severe accident, and to take in the experimental data by participating in the international cooperative experiments regarding the important SA phenomena and the effectiveness of accident management. Based on those data and knowledge, JNES is developing and improving severe accident analysis models to maintain the severe accident analysis codes and the accident management knowledge base for assessment of the NPPs in Japan. The activities in fiscal year 2010 are as follows; Experimental study on OECD/NEA projects such as MCCI, SERENA, SFP and international cooperative PSI-ARTIST project, and analytical study on accident management review of new plant and making regulation for severe accident. (author)

  15. Knowledge data base for severe accident management of nuclear power plants

    For the reinforcement of the safety of NPPs, the continuous efforts are very important to take in the up-to-date scientific and technical knowledge positively and to reflect them into the safety regulation. The purpose of this present study is to gather effectively the scientific and technical knowledge about the severe accident (SA) phenomena and the accident management (AM) for prevention and mitigation of severe accident, and to take in the experimental data by participating in the international cooperative experiments regarding the important SA phenomena and the effectiveness of accident management. Based on those data and knowledge, JNES is developing and improving severe accident analysis models to maintain the severe accident analysis codes and the accident management knowledge base for assessment of the NPPs in Japan. The activities in fiscal year 2011 are as follows; Experimental study on OECD/NEA projects such as MCCI, SERENA, SFP and international cooperative PSI-ARTIST project, and analytical study on accident management review of new plant and making regulation for severe accident. (author)

  16. Best practice guide for radioactivity measurement laboratories in a post-accident situation

    Published for laboratories likely to be asked to perform radioactivity measurements at the time of or after a radiological or nuclear accident in France or abroad, this guide aims at defining the best practices in terms of laboratory organisation (sample flow management, personnel radioprotection, sample identification and recording, sample cross-contamination risks, result transmission, archiving of data, results and samples, waste dismissal), and in terms of metrology (adaptation to needs in terms of detection limit and measurement uncertainty, preferred use of gamma spectrometry, analysis strategies)

  17. Accident management to ensure containment integrity at Seabrook Station

    This paper reports that PSA results for Seabrook Station have shown capability and strength of the large dry primary containment to withstand early pressure loads that could result from a potential severe core damage event. To build upon a high degree of confidence that containment integrity would be maintained in light of issues such as direct containment heating (DCH) and induced steam generator tube rupture (ISGTR), select accident management strategies have been evaluated for the plant. These strategies include emergency response technical support center procedures and hardware modifications to eliminate the potential for DCH and ISGTR for high pressure core melt scenarios. Operator actions that would result from these strategies include primary system depressurization using the pressurizer power-operated relief valves (PORV) and use of fire water pumps to prevent overheating and thermal creep rupture of the steam generator tubes. The risk management effectiveness of these strategies was quantified with the use of a full-scope Level 3 PSA model of Seabrook Station. A byproduct of this evaluation is a current assessment of the risk significance of DCH and ISGTR for this paper

  18. Developing and validating severe accident management guidelines using SAMPSON-RELAP/SCDAPSIM.MOD3.4

    The development and validation of Severe Accident Management Guidelines (SAMGs) must consider complex thermal-hydraulic and severe accident phenomena. Yet, many of the simplified integral Severe Accident codes, that have been used widely to develop SAMGs in Europe, Asia, and the United States, cannot accurately predict many of these complex interactions. By contrast, detailed codes such as SAMPSON-RELAP/SCDAPSIM have shown, through comparison with the TMI-2 accident and experiments, that they can predict such complex behavior. This paper describes the merger of SAMPSON with RELAP/SCDAPSIM/MOD3.4, reviews the severe accident phenomena important for Severe Accident Management, and then describes the potential impact of using SAMPSON-RELAP/SCDAPSIM on the development and validation of SAMGs. A companion paper, being presented at this conference provides an example of the application of SAMPSON-RELAP/SCDAPSIM for the development and validation of a SAMG for a Nuclear Power Plant. (authors)

  19. Marine Accidents in Northern Nigeria: Causes, Prevention and Management

    Lawal Bello Dogarawa

    2012-11-01

    Full Text Available Boat mishaps tend to be increasing in Nigeria in spite of all regulatory measures which have been taken to prevent and control marine accidents. Boat mishaps could occur anywhere water transportation takes place. However, there is a general impression that water transportation takes place only in the riverine areas located in Southern Nigeria but, this paper reports about marine accident cases in Northern Nigeria. It evaluates the safety measures put in place by operators and other institutional bodies in the areas and assesses the level of infrastructure in terms of quantity, quality and accessibility to boat operators, boat users and institutional staff. Questionnaires were administered through individual and group interviews with boat owners, boat drivers, boat users, boat builders, boat engine mechanics, local government officials, maritime workers union, the marine police, traditional regulators and staff of the federal government agencies for maritime affairs. The paper found that marine transportation is neglected in Northern Nigeria with dilapidated jetties, ill-equipped marine police, non-functional ferries and boast meant to be used by federal officials and wrecks in water channels without removal. Maritime safety is therefore compromised with cases of overloading carrying people, animals, grains and petroleum products in one boat without fire extinguisher and no lifejackets. The paper concludes that there are considerable water transportation activities in Northern Nigeria without a corresponding government attention. It is therefore recommend that government should intervene by providing lifejackets, fire extinguishers, training of surveyors, refurbishing ferries for enforcement as well as creating safety awareness in the region.

  20. EFFICIENCY OF REPEATED AND UNSCHEDULED TRAINING AS THE MEASURES TO PREVENT ACCIDENTS AT SUPPLY DEPOTS AND WAREHOUSES

    Bocharova Irina Nikolaevna

    2013-05-01

    Full Text Available This paper presents the results of the analysis of the state of occupational safety at supply depots and warehouses. It is revealed that most accidents involve the employees who have less than one year’s service. Experience has proven that the preventive activities to avoid occupational traumatism are efficient when a complex of workplace safety measures is implemented. The experts consider the repeated and unscheduled training to be very important events. This is supported by the fact that among the employees of the commercial establishments who underwent repeated and unscheduled training, the number of individuals who suffered an accident is small. The efficient functioning of the occupational safety training system is infeasible without ensuring the motivation for assimilating the knowledge and forming the complete foundation for safe labor. In order to reduce the number of accidents, one should proceed from the principle of responding to accidents to the system for professional risk management.

  1. Management of a severe accident on a pressurised water reactor in France

    This brief document defines what a severe accident is on a nuclear reactor, indicates the different failure modes which have been defined (vapour explosion in the reactor vessel, hydrogen explosion, and so on). It describes the management of a core fusion accident for pressurized water reactors, for which a guide has been designed, the GIAG (intervention guide for a severe accident situation). The principles of such an intervention are described, and then the approach for an EPR reactor

  2. Radiological protection from radioactive waste management in existing exposure situations resulting from a nuclear accident

    Sugiyama, Daisuke; Hattori, Takatoshi

    2012-01-01

    In environmental remediation after nuclear accidents, radioactive wastes have to be appropriately managed in existing exposure situations with contamination resulting from the emission of radionuclides by such accidents. In this paper, a framework of radiation protection from radioactive waste management in existing exposure situations for application to the practical and reasonable waste management in contaminated areas, referring to related ICRP recommendations was proposed. In the proposed...

  3. Measurement of the Portsmouth Gaseous Diffusion Plant criticality accident alarm

    Measurements of the Portsmouth Gaseous Diffusion Plant's nuclear criticality accident radiation alarm signal response time, sound wave frequency, and sound volume levels were made to demonstrate compliance with ANSI/ANS-8.3-1986. A steady-state alarm signal is produced within one-half second of obtaining a two-out-of-three detector trip. The fundamental alarm sound wave frequency is 440 hertz. The sound volume levels are greater than 10 decibels above background and ranged from 100 to 125 A-weighted decibels. The requirements of the standard were met; however the recommended maximum sound volume level of 115 dBA was exceeded. Emergency procedures require immediate evacuation upon initiation of a facility's radiation alarm. Comparison with standards for allowable time of exposure at different noise levels indicate that the elevated noise level at this location does not represent an occupational injury hazard. 8 refs., 5 figs

  4. Measurement of steam condensation on aerosols und LWR accident conditions

    The report summarizes the results of experiments on steam condensation onto aerosol particles. A facility was constructed which allows the direct measurement of the condensation processes. The thermodynamic boundary conditions were typical for a core melt accident. Different aerosol species were used, especially uranium dioxide which constitutes a large fraction of the core melt aerosol. As a general result the condensation process in supersaturated atmospheres causes a drastic change in the shape of the aerosol particles. Originally fluffy chain-like aggregates are compressed to nearly spherical dense particles. A significant simplification of the NAUA-model can be used because the commonly encountered shape factor problems become non-existent. This also leads to a greater reliability of computed results. (orig./HP)

  5. Severe accident instrumentation systems for BWR water level and temperature in primary containment vessel measurements

    The severe accident at TEPCO's Fukushima Daiichi nuclear power station (TF1 accident) in March 2011 brought the lost of the functions of many instrumentation systems. In order to enable the measurements of the important parameters such as reactor water level, temperature and so on even in a case such as the TF1 accident occurs, severe accident instrumentation systems are being developed. In this paper, new system configurations of BWR water level measurement and temperature measurement in primary containment vessels are proposed. Then performance tests for prototype sensors of these measurement systems under high temperature conditions are described. (author)

  6. A Study on Reinforcement of the Accident Management System in Korea

    The aim of this study is to present the status of post-Fukushima actions with respect to accident management and also provides the current status of developing EDMGs and applicability of a FLEX strategy in Korea. As part of the post-Fukushima actions in Korea, SAMGs will be revised to improve the effectiveness of accident management. For this purpose, it is recommended to revise the EOPs and SAMGs and establish the EDMGs with consideration of prolonged SBO, spent fuel pool cooling, using mobile equipment for accident control, feedback of the implementation of the action items of the special safety inspection, multiple severe accidents for all reactors at a site. It is considered that the FLEX strategy may be useful to mitigate the accidents like Fukushima. Therefore, it is recommended to adopt this strategy including provision of the equipment with protection from external events. The Fukushima accident revealed that EOPs and SAMGs were not effectively coping with and mitigating the severe accident caused by extreme natural hazards such as earthquake and tsunami. The accident indicated needs for strengthening the existing accident management procedures such as emergency operating procedures (EOPs) and severe accident management guidelines (SAMGs). In particular, these procedures should address the possibility of extreme natural hazards causing a prolonged SBO condition, which affects multiple-units and Spent Fuel Pools (SFPs) (NTTF Recommendation 9). In addition, in order to prevent and mitigate the potential damage in an extensive scale at a multi-unit site due to external events, fire, various kinds of countermeasures are required by the Regulatory Body. These are the follow-up actions to the special safety inspection carried out just after the Fukushima accident and the stress tests for old plants. Especially, the Extensive Damage Mitigation Guidelines (EDMGs) are being provided by the utility in conjunction with adoption of the FLEX strategy (diverse and

  7. EC Research Contribution to Decision-making Processes Relevant to Severe Accident Management

    As a result of the two well-known civil nuclear accidents and of the consequent increase in safety requirements, the need to properly assess severe accident (SA) scenarios for present and future nuclear power plants (going beyond the traditional three-level defence-in-depth strategy) became evident. In this line, various research activities were launched and are performed within the Euratom Framework Programmes, in particular the completed Fourth one (F P-4, 1994-1998) and the present Fifth one (FP-5, 1998-2002). The initial orientation of the EC research activities was mainly focused on improving the understanding of the phenomena and mechanisms involved in such accidents, in order to contribute to prevent possible final radioactivity releases. A consensus on how to model those SA phenomena in accident safety analyses by means of specific tools (SA codes developed, verified and validated through experimental results provided) is reasonably advanced. Currently, the EC research activities related to severe accidents are balanced between a twofold approach aimed at assessing the risks related with severe accident scenarios and to support the development of severe accident management (SAM) strategies, together with the optimisation of backfitting measures for existing reactors or specific designs for future nuclear power plants. This new orientation is confronting difficulties, inherent to the phenomenological character of several research activities, which make a direct application of the results into SAM measures premature in some cases. In this regard, this paper presents a series of ten selected FP-5 projects with emphasis placed on the applicability of research results towards SAM strategies to be used by decision-makers amongst utilities, the nuclear industry in particular designers, and regulators. The majority of them also contain -further to the SAM approach- supporting elements focused on risk assessment. The revised programme of the key action 'Nuclear

  8. Development of a reactor vessel failure diagnosis system for accident management

    Diagnosis of vessel failure provides for operators and TSC personnel very important information to manage the severe accident in nuclear power plant. However, operators can not diagnose the reactor vessel failure by watching the temporal trends of some parameters because they never have experienced the severe accident. Therefore, this study proposes a method on the diagnosis of the PWR vessel failure using a Spatiotemporal Neural Network (STN). STNs can deal directly with both the spatial and the temporal aspects of input signals and can well identify a time-varying problem. The target patterns are generated from MAAP code. Vessel failure diagnosis has been performed for 8 accidents and the developed STNs have been verified for untrained three severe accidents. STNs identifies the vessel failure time and the initiating events. For example, when large break LOCA (break size = 0.16 m2) is used for input accident scenario, only the output value for the target pattern of LBLOCA is activated greater than the threshold value near the real vessel failure. To validate vessel failure diagnosis system and to train severe accident to operators, extensive severe accident simulator is to be an absolute necessity. Therefore, a simplified severe accident simulator, SIMAAP (severe accident Simulator based on MAAP), has been developed. SIMAAP simulates the various severe accident progress through on-line communication with MAAP

  9. Management of severe pelvic injury following road traffic accident in a resource-limited setting

    A 34 year old woman involved in road traffic accident with severe anterior and posterior pelvic fractures with associated soft tissue injury was referred from Wa Regional Hospital 18 hours after the accident to Tania Specialist Hospital in Tamale. Emergency resuscitative measures such as catheterization and management of pain with analgesics were initiated. Computed tomography (CT scan) or Magnetic resonance imaging has been recommended as the appropriate tools for risk assessment in such cases however none of this was available at the time of the accident. The only assessment tool available was the C-arm machine which was used to X-ray the pelvis in the following plane; anterio - posterior pelvic - inlet and pelvic - outlet. Early internal reduction and stabilization of pelvis was immediately carried out using the procedure of open reduction and internal fixation (ORIF). Approximately 2 weeks after the operation, radiographs showed signs of healing and the patient was discharged on partial body weight bearing. Upon second review 12 weeks post operatively, complete recovery was accomplished.

  10. Severe human factor accidents and their management in a in-service nuclear Power plant

    Human Reliability Analysis (HRA) is an important part of Probabilistic Safety Assessment (PSA) in a nuclear power plant (NPP). It can be used to evaluate and quantify the behaviors of the operators in a post-accident response. The paper picks up the serious human factor event sequences that contribute more than 5% to the overall Core Damage Frequency (CDF) involved in PSA through a HRA analysis on a domestic PWR. The basic human error probabilities (BHEPs) of these human factor event sequences are resulted, on the basis of which the actions of the operators within the main control room (MCR) after the accidents are analyzed and their criticalities are arranged in order. The paper, from the point of engineering management,puts forward the measures to improve the corresponding emergency operating procedures (EOPs) and the MCR surroundings through analyzing serious human factor event sequence arrangement and the actions of operators in the post-accident interferences. With regard to the operator's interferences of high criticality the NPP should enforce training and improve its ability of interferences. (authors)

  11. A study on the use of neural network for severe accident management

    Based on the consensus that the course and consequence of a severe core damage accident can be greatly influenced by the operators' action, there have been extensive efforts to establish severe accident management program. A severe accident management process is essentially a sequence of decision making with a wide variety of available information under the highly uncertain condition, aimed at successful termination of accident progression or consequence minimization. For operators to take correct and timely accident management actions, they should be informed of the accident progression. Some key events, such as onset of core uncovery, core-melt initiation, reactor vessel lower head failure, containment failure, etc., act as landmarks for operators to make decisions in severe accident management process. Thus it is of critical importance to identify the timing at which such events occur in accident management. Unfortunately, it is difficult task partly due to phenomenological complexity and partly due to the lack of instrumentation reliability in severe accident environment, making the traditional procedural or rule-based approach inappropriate to be adopted to this end. Instead a technique, called artificial neural network, has been successfully applied to the similar problem domain out of various disciplines including nuclear industry. This paper presents a study on the application of a special kind of artificial neural network having the capability of recognizing time-varying patterns, called spatiotemporal network (STN), to the event timing prediction which is an important sub function of integrated computer supporting system for severe accident management. As the first trial, concentration was put on the identification of reactor vessel lower head failure which is considered the most critical events discriminating between so called in-vessel and ex-vessel accident management phases. Several sets of seven parameter signals from MAAP-based severe accident

  12. 基于事故发展与控制的隐患分级方法%Risk Classification Method for Accident Potential Based on Development and Control Measures of Accident

    赵东风; 申玉琪; 赵志强; 张佑明; 孟亦飞

    2012-01-01

    For the purpose of making the management of accident potential more scientific, the relationship between accident potential and accident was discussed and studied. The fundamental property of accident potential was brought forward that it can make the accident happen or develop. The mechanism whereby accident potential functions in the accident was revealed clearly, by predicting the impelling and inhibiting factors in the accident process. The accident potential was classified into two types (the first type of accident potential and the second type of accident potential) according to the different time they work. In the risk assessment, the problem of specific accident potential grading was resolved. By introducing the assessment indexes of accident potential exposure frequency, possibility of other factors, corrective actions, the initial value of the consequences of the accident, the correction factor of personnel protection, the correction factor of personnel exposure, the correction factor of emergency measures, the correction factor of property loss, an assessment index system was established. By calculating the risk of accident potential-causing accident, the real risk of accident potential was assessed.%为使隐患管理工作更加科学,对隐患与事故的关系进行研讨,提出隐患的根本属性是能够促使事故发生或发展.通过预估促使和控制(阻碍)事故发展的因素,来揭示隐患在事故过程中的作用机制.根据发生作用的时间将隐患分为第1类隐患和第2类隐患.在风险评估过程中,解决了具体隐患风险分级的问题,提出隐患暴露频率、其他条件的可能性、隐患纠正系数、事故后果初始分值、人员防护修正系数、人员暴露修正系数、应急处理与事故控制修正系数和财产损失修正系数等评价指标.通过隐患致因事故风险的计算,评估隐患的最终风险.

  13. Solid waste accident analysis in support of the Savannah River Waste Management Environmental Impact Statement

    The potential for facility accidents and the magnitude of their impacts are important factors in the evaluation of the solid waste management addressed in the Environmental Impact Statement. The purpose of this document is to address the potential solid waste management facility accidents for comparative use in support of the Environmental Impact Statement. This document must not be construed as an Authorization Basis document for any of the SRS waste management facilities. Because of the time constraints placed on preparing this accident impact analysis, all accident information was derived from existing safety documentation that has been prepared for SRS waste management facilities. A list of facilities to include in the accident impact analysis was provided as input by the Savannah River Technology Section. The accident impact analyses include existing SRS waste management facilities as well as proposed facilities. Safety documentation exists for all existing and many of the proposed facilities. Information was extracted from this existing documentation for this impact analysis. There are a few proposed facilities for which safety analyses have not been prepared. However, these facilities have similar processes to existing facilities and will treat, store, or dispose of the same type of material that is in existing facilities; therefore, the accidents can be expected to be similar

  14. The philosophy of severe accident management in the US

    The US NRC has put forth the initial steps in what is viewed as the resolution of the severe accident issue. Underlying this process is a fundamental philosophy that if followed will likely lead to an order of magnitude reduction in the risk of severe accidents. Thus far, this philosophy has proven cost effective through improved performance. This paper briefly examines this philosophy and the next step in closure of the severe accident issue, the IPE. An example of the authors experience with determinist. (author)

  15. Medical management of radiological accidents in non-specialized clinics: mistakes and lessons

    In 1996-2002 three radiological accidents were developed in Georgia. There were some people injured in those accidents. During medical management of the injured some mistakes and errors were revealed both in diagnostics and scheme of the treatment. The goal of this article is to summarize medical management of the mentioned radiological accidents, to estimate reasons of mistakes and errors, to present the lessons drawn in result of Georgia radiological accidents. There was no clinic with specialized profile and experience. Accordingly due to having no relevant experience late diagnosis can be considered as the main error. It had direct influence on the patients' health and results of treatment. Lessons to be drawn after analyzing Georgian radiological accidents: 1. informing medical staff about radiological injuries (pathogenesis, types, symptoms, clinical course, principles of treatment and etc.); 2. organization of training and meetings in non-specialized clinics or medical institutions for medical staff; 3. preparation of informational booklets and guidelines.(author)

  16. Development of the MIDAS GUI environment for severe accident management and analyses

    Kim, K. R.; Park, S. H.; Kim, D. H. [KAERI, Taejon (Korea, Republic of)

    2004-07-01

    MIDAS is being developed at KAERI as an integrated severe accident analysis code with existing model modification and new model addition. Also restructuring of the data transfer scheme is going on to improve user's convenience. In this paper, various MIDAS GUI systems which are input management system IEDIT, variable plotting system IPLOT, severe accident training simulator SATS, and online guidance module HyperKAMG, are introduced. In addition, detail functions and usage of these systems for severe accident management and analyses are described.

  17. The development and demonstration of integrated models for the evaluation of severe accident management strategies - SAMEM

    This study is concerned with the further development of integrated models for the assessment of existing and potential severe accident management (SAM) measures. This paper provides a brief summary of these models, based on Probabilistic Safety Assessment (PSA) methods and the Risk Oriented Accident Analysis Methodology (ROAAM) approach, and their application to a number of case studies spanning both preventive and mitigative accident management regimes. In the course of this study it became evident that the starting point to guide the selection of methodology and any further improvement is the intended application. Accordingly, such features as the type and area of application and the confidence requirement are addressed in this project. The application of an integrated ROAAM approach led to the implementation, at the Loviisa NPP, of a hydrogen mitigation strategy, which requires substantial plant modifications. A revised level 2 PSA model was applied to the Sizewell B NPP to assess the feasibility of the in-vessel retention strategy. Similarly the application of PSA based models was extended to the Barseback and Ringhals 2 NPPs to improve the emergency operating procedures, notably actions related to manual operations. A human reliability analysis based on the Human Cognitive Reliability (HCR) and Technique For Human Error Rate (THERP) models was applied to a case study addressing secondary and primary bleed and feed procedures. Some aspects pertinent to the quantification of severe accident phenomena were further examined in this project. A comparison of the applications of PSA based approach and ROAAM to two severe accident issues, viz hydrogen combustion and in-vessel retention, was made. A general conclusion is that there is no requirement for further major development of the PSA and ROAAM methodologies in the modelling of SAM strategies for a variety of applications as far as the technical aspects are concerned. As is demonstrated in this project, the

  18. Accident analysis for transuranic waste management alternatives in the U.S. Department of Energy waste management program

    Preliminary accident analyses and radiological source term evaluations have been conducted for transuranic waste (TRUW) as part of the US Department of Energy (DOE) effort to manage storage, treatment, and disposal of radioactive wastes at its various sites. The approach to assessing radiological releases from facility accidents was developed in support of the Office of Environmental Management Programmatic Environmental Impact Statement (EM PEIS). The methodology developed in this work is in accordance with the latest DOE guidelines, which consider the spectrum of possible accident scenarios in the implementation of various actions evaluated in an EIS. The radiological releases from potential risk-dominant accidents in storage and treatment facilities considered in the EM PEIS TRUW alternatives are described in this paper. The results show that significant releases can be predicted for only the most severe and extremely improbable accidents sequences

  19. Initial medical management of criticality accident victim; Conduite a tenir aux victimes d'un accident de criticite

    Miele, A.; Bebaron-Jacobs, L

    2005-07-01

    The extremely severe criticality accidents known to this day, and the subsequent deaths recorded (Sarov 1997 and Tokai Mura 1999), demonstrate the need for sustained surveillance and constant adapted training for the teams in charge of irradiated and/or contaminated victims. The aim of this work group, composed of occupational health services and associated medical biology laboratories, is to present, in leaflet format, the essential data on the documentation and the conduct to be held when facing the victims of a criticality accident. The studies of this work group confirm the difficulties involved in managing this type of accident, both from the dosimetric evaluation point of view and from the therapeutic management point of view. That is why several research themes and perspectives are developed. During the different phases of victim triage, the recommendations given on these leaflets describe the operational conducts to be held. This work will have to be updated according to the evolution in knowledge and means: short and long term effects of exposure to neutrons, multi-competence hospital cooperation, expertise networks related to dosimetric reconstitution. (authors)

  20. Direction Committee for the management of the post-accident phase of a nuclear accident or of a radiological event (CODIRPA). Work group nr 3: 'Assessment of radiological and dose consequences in a post-accident situation'. Final report

    This report first describes how radioactive contamination occurs after a nuclear accident, whether it concerns plants, animals, people, and buildings, how people can be exposed, and how a post-accidental zoning is implemented either to protect population or to control territories. It describes principles and methods for the assessment of the contamination of the environment (radiological values, characterization of radioactive deposits, of agriculture products, and of wastes, materials and manufactured products). It describes how to organise radioactivity measurements in the environment (principles and objectives of measurement programmes, sampling organisation and management, laboratory radioactivity measurements, identification and preparation of radioactivity measurement operators, results management). It describes how to assess doses received by exposed people (measurement techniques, retrospective assessment, proposition of a dose assessment strategy for exposed population)

  1. Help guides for post-accident consequence management: farm activities and exiting the emergency phase

    After having recalled the main actions foreseen in the PPIs (plans particuliers d'intervention, intervention specific plans) in case of radionuclide release in the environment after a nuclear accident, i.e. sheltering and ingestion of steady iodine, and also indicated the different phases of consequence management (preparation, emergency and post-accident phases), this report describes and comments the contents of two guides published by the IRSN (the French Radioprotection and Nuclear Safety Institute) and dealing with the management of post-accident consequences. The first one is a guide to aid to decision-making for the management of the agricultural sector in case of nuclear accident, and the second one is a guide for the preparation of the end of the emergency phase in which actions to be performed during the first week after the end of accidental releases are described

  2. Second Specialist Meeting on operator aids for severe accident management: summary and conclusions

    The second OECD Specialist Meeting on operator aids for severe accident management (SAMOA-2) was held in Lyon, France (1997), and was attended by 33 specialists representing ten OECD member countries. As for SAMOA-1, the scope of SAMOA-2 was limited to operator aids for accident management which were in operation or could be soon. The meeting concentrated on the management of accidents beyond the design basis, including tools which might be extended from the design basis range into the severe accident area. Relevant simulation tools for operator training were also part of the scope of the meeting. 20 papers were presented; there were two demonstrations of computerized systems (the ATLAS analysis simulator developed by GRS, and EDF's 'Simulateur Post Accidentels' (SIPA). The three sessions dealt with operator aids for control rooms, operator aids for technical support centres, and simulation tools for operator training. The various papers for each session are summarized

  3. Objective provision tree application to the effectiveness evaluation of accident management guidelines

    After the Fukushima accident in 2011, various lessons and safety enhancement action items were announced by national regulatory bodies. Among those items, the enforcement of procedural efficiency verification for accidents management guidelines including emergency operating procedures (EOPs), severe accident management guides (SAMGs) and extensive damage mitigating guidelines (EDMG) if applicable, was raised. The Objective Provision Tree (OPT) method is a top down approach which starts from the level of Defense in Depth (DiD), objectives and barriers, safety functions, challenges, mechanisms and finally ends with provisions. The benefit of OPT application to safety concerns includes that the OPT enables the comprehensive review for the verification of consistency and integrity of safety requirements for a specific safety issue. In this study, the preliminary framework for the application of OPT to the effectiveness evaluation of accident management guideline was introduced

  4. The Management of Beyond Design Basis Accidents with Loss of Cooling at NPPs with WWER

    The analysis of Ukrainian guidance on management of beyond design basis accidents at NPP is carried out. International experience on development regulatory documents in this area is considered. Directions for improvements of regulatory documents for NPPs with WWER are determined. For the analysis PSA results of the Ukrainian NPPs are used. It is shown that the primary circuit LOCAs are the dominant contributors of CDF. The set of symptoms for each LOCA group is developed. To develop the management algorithms for each accident group the approach to grouping of accident sequences on the basis of critical configurations of systems is submitted. Examples of necessary changes for improvement of guidance on management of beyond design basis accidents at NPPs with WWER are presented.(author)

  5. Measures for speed management.

    2009-01-01

    Measures for speed management are essential for limiting the negative effects of driving too fast and at inappropriate speeds. To begin with, safe and credible speed limits need to be determined. Dynamic and variable speed limits that take into account the current circumstances, such as weather cond

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

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

  7. Measuring the Impacts of Nuclear Accidents on Energy Policy

    Zsuzsanna Csereklyei

    2013-01-01

    This paper examines the history of nuclear energy, safety developments of reactors and nuclear energy policy from the 1950s on. I investigate the effects of nuclear accidents on energy policy with the help of a panel dataset of 31 countries from 1965-2009, using annual data about the capacity of reactors under construction, primary energy consumption, as well as three nuclear accidents scaled INES five or higher by the International Atomic Energy Agency. After determining the extent of the ac...

  8. Decision making process and emergency management in different phases of a nuclear accident

    EVATECH, Information Requirements and Countermeasure Evaluation Techniques in Nuclear Emergency Management, was a research project in the key action 'Nuclear Fission' of the fifth EURATOM Framework Programme (FP5). The overall objective of the project was to enhance the quality and coherence of response to nuclear emergencies in Europe by improving the decision support methods, models and processes in ways that take into account the expectations and concern of the many different parties involved - stake holders both in managing the emergency response and those who are affected by the consequences of nuclear emergencies. The project had ten partners from seven European countries. The development of the real-time online decision support system RODOS has been one of the major items in the area of radiation protection within the European Commission's Framework Programmes. The main objectives of the RODOS project have been to develop a comprehensive and integrated decision support system that is generally applicable across Europe and to provide a common framework for incorporating the best features of existing decision support systems and future developments. Furthermore the objective has been to provide greater transparency in the decision process to: improve public understanding and acceptance of off-site emergency measures, to facilitate improved communication between countries of monitoring data, predictions of consequences, etc., in the event of any future accident, and to promote, through the development and use of the system, a more coherent, consistent and harmonised response to any future accident that may affect Europe. (authors)

  9. Accident risk and safety measures in the transport sector in Norway; Ulykkesrisiko og sikkerhetstiltak i transportsektoren

    NONE

    1998-12-01

    The scope of the work described in this report was (1) to evaluate methods for risk mapping considering all of the different means of transport, (2) to evaluate the extent to which measures should be taken against various types of accidents, (3) to evaluate cost-benefit assessments of accident-reducing measures irrespective of the different means of transport, (4) to evaluate the preferences of measures/cost effectiveness of different measures within different sectors, and (4) to evaluate the possibility of improving the efficiency of possible measures. It also considers the risk situation for ferry service. In addition to the purely human aspect, traffic accidents constitute an expensive social problem. Yet it would be too costly to meet a potential requirement that traffic accidents should disappear. The resources used by society to combat accidents have to be seen in the light of (1) the profit that can be achieved compared to alternative use of the resources, and (2) the possible negative consequences of different safety measures on, for instance, travel time and the extent of the transport. It is pointed out that when accident risk is compared from one transport means to another, different relative positions are found depending on how risk is quantified. Thus, for instance, on average, per year 5 times as many people die in accidents involving private cars as in motor cycle accidents, while for the number of deaths per billion person kilometers the ratio is almost the opposite,1:6.5. 34 refs., 12 figs., 13 tabs.

  10. Research on the management of the wastes from plant accidents

    The accident in Fukushima Daiichi Nuclear Power Plant released large amount of radio-nuclides and contaminated wide areas within and out of the site. The decontamination, storage, treatment and disposal of generated wastes are now under planning. Though the regulations for radioactive wastes discharged from normal operation and decommissioning of nuclear facilities have been prepared, it is necessary to make amendments of those regulations to deal with wastes from the severe accidents which may have much different features on nuclides contents, or possibility to accompany hazardous chemical materials. Characteristics, treatment and disposal of wastes from accidents were surveyed by literature and the radionuclide migration from the assumed temporally storage yards of the disaster debris was analyzed for consideration of future regulation. (author)

  11. Investigation of the management of the wastes from plant accident

    The accident in Fukushima Daiichi Nuclear Power Plant discharged large amount of radio-nuclides and contaminated wide areas in and out of the site. The decontamination, storage, treatment and disposal of generated wastes are now under planning. Though regulations for the radioactive wastes arisen from normal operation and decommissioning of nuclear facilities have been prepared, it is necessary to make amendment of those regulations to deal with wastes from the severe accident which may have much different features on nuclides contents, or possible accompanying hazardous chemical materials. Characteristics of wastes from accidents in foreign nuclear installations, and the treatment and the disposal of those wastes were surveyed by literature and radionuclide migration from the assumed temporally storage yards of the disaster debris was analyzed for consideration of future regulation. (author)

  12. Example of severe accident management guidelines validation and verification using full scope simulator

    The purpose of Severe Accident Management Guidelines (SAMG) is to provide guidelines to mitigate and control beyond design bases accidents. These guidelines are to be used by the technical support center that is established at the plant within one hour after the beginning of the accident as a technical support for the main control room operators. Since some of the accidents can progress very fast there are also two guidelines provided for the main control room operators. The first one is to be used if the core damage occurs and the TSC is not established yet and the second one after technical support center become operational. After SG replacement and power uprate in year 2000, NPP Krsko developed Rev.1 of these procedures, which have been validated and verified during one-week effort. Plant specific simulator capable of simulating severe accidents was extensively used.(author)

  13. Radioactive waste management after NPP accident: Post-Chernobyl experience

    As a result of the Chernobyl NPP accident a very large amount of so-called 'Chernobyl waste' were generated in the territory of Belarus, which was contaminated much more than all other countries. These wastes relate mainly to two following categories: low-level waste (LLW) and new one 'Conventionally Radioactive Waste' (CRW). Neither regulations nor technology and equipment were sufficiently developed for such an amount and kind of waste before the accident. It required proper decisions in respect of regulations, treatment, transportation, disposal of waste, etc. (author)

  14. Measuring Risk Aversion for Nuclear Power Plant Accident: Results of Contingent Valuation Survey in Korea

    Within the evaluation of the external cost of nuclear energy, the estimation of the external cost of nuclear power plant (NPP) severe accident is one of the major topics to be addressed. For the evaluation of the external cost of NPP severe accident, the effect of public risk averse behavior against the group accidents, such as NPP accident, dam failure, must be addressed. Although the equivalent fatalities from a single group accident are not common and its risk is very small compared to other accidents, people perceive the group accident more seriously. In other words, people are more concerned about low probability/high consequence events than about high probability/low consequence events having the same mean damage. One of the representative method to integrate the risk aversion in the external costs of severe nuclear reactor accidents was developed by Eeckoudt et al., and he used the risk aversion coefficient, mainly based on the analysis of financial risks in the stock markets to evaluate the external cost of nuclear severe accident. However, the use of financial risk aversion coefficient to nuclear severe accidents is not appropriate, because financial risk and nuclear severe accident risk are entirely different. In this paper, the individual-level survey was conducted to measure the risk aversion coefficient and estimate the multiplication factor to integrate the risk aversion in the external costs of NPP severe accident. This study propose an integrated framework on estimation of the external cost associated with severe accidents of NPP considering public risk aversion behavior. The theoretical framework to estimate the risk aversion coefficient/multiplication factor and to assess economic damages from a hypothetical NPP accident was constructed. Based on the theoretical framework, the risk aversion coefficient can be analyzed by conducting public survey with a carefully designed lottery questions. Compared to the previous studies on estimation of the

  15. Measuring Risk Aversion for Nuclear Power Plant Accident: Results of Contingent Valuation Survey in Korea

    Lee, Sang Hun; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    Within the evaluation of the external cost of nuclear energy, the estimation of the external cost of nuclear power plant (NPP) severe accident is one of the major topics to be addressed. For the evaluation of the external cost of NPP severe accident, the effect of public risk averse behavior against the group accidents, such as NPP accident, dam failure, must be addressed. Although the equivalent fatalities from a single group accident are not common and its risk is very small compared to other accidents, people perceive the group accident more seriously. In other words, people are more concerned about low probability/high consequence events than about high probability/low consequence events having the same mean damage. One of the representative method to integrate the risk aversion in the external costs of severe nuclear reactor accidents was developed by Eeckoudt et al., and he used the risk aversion coefficient, mainly based on the analysis of financial risks in the stock markets to evaluate the external cost of nuclear severe accident. However, the use of financial risk aversion coefficient to nuclear severe accidents is not appropriate, because financial risk and nuclear severe accident risk are entirely different. In this paper, the individual-level survey was conducted to measure the risk aversion coefficient and estimate the multiplication factor to integrate the risk aversion in the external costs of NPP severe accident. This study propose an integrated framework on estimation of the external cost associated with severe accidents of NPP considering public risk aversion behavior. The theoretical framework to estimate the risk aversion coefficient/multiplication factor and to assess economic damages from a hypothetical NPP accident was constructed. Based on the theoretical framework, the risk aversion coefficient can be analyzed by conducting public survey with a carefully designed lottery questions. Compared to the previous studies on estimation of the

  16. Bibliography for nuclear criticality accident experience, alarm systems, and emergency management

    Putman, V.L.

    1995-09-01

    The characteristics, detection, and emergency management of nuclear criticality accidents outside reactors has been an important component of criticality safety for as long as the need for this specialized safety discipline has been recognized. The general interest and importance of such topics receives special emphasis because of the potentially lethal, albeit highly localized, effects of criticality accidents and because of heightened public and regulatory concerns for any undesirable event in nuclear and radiological fields. This bibliography lists references which are potentially applicable to or interesting for criticality alarm, detection, and warning systems; criticality accident emergency management; and their associated programs. The lists are annotated to assist bibliography users in identifying applicable: industry and regulatory guidance and requirements, with historical development information and comments; criticality accident characteristics, consequences, experiences, and responses; hazard-, risk-, or safety-analysis criteria; CAS design and qualification criteria; CAS calibration, maintenance, repair, and testing criteria; experiences of CAS designers and maintainers; criticality accident emergency management (planning, preparedness, response, and recovery) requirements and guidance; criticality accident emergency management experience, plans, and techniques; methods and tools for analysis; and additional bibliographies.

  17. Bibliography for nuclear criticality accident experience, alarm systems, and emergency management

    The characteristics, detection, and emergency management of nuclear criticality accidents outside reactors has been an important component of criticality safety for as long as the need for this specialized safety discipline has been recognized. The general interest and importance of such topics receives special emphasis because of the potentially lethal, albeit highly localized, effects of criticality accidents and because of heightened public and regulatory concerns for any undesirable event in nuclear and radiological fields. This bibliography lists references which are potentially applicable to or interesting for criticality alarm, detection, and warning systems; criticality accident emergency management; and their associated programs. The lists are annotated to assist bibliography users in identifying applicable: industry and regulatory guidance and requirements, with historical development information and comments; criticality accident characteristics, consequences, experiences, and responses; hazard-, risk-, or safety-analysis criteria; CAS design and qualification criteria; CAS calibration, maintenance, repair, and testing criteria; experiences of CAS designers and maintainers; criticality accident emergency management (planning, preparedness, response, and recovery) requirements and guidance; criticality accident emergency management experience, plans, and techniques; methods and tools for analysis; and additional bibliographies

  18. Beyond design basis accident management related to Angra 2 nuclear power plant

    Emergency procedures, associated to beyond design basis accident management, are important issues to guarantee the safe operation of nuclear power plants, and they may use plant normal operation systems for controlling such situation. The simulation of accident critical scenarios should be developed with the aid of the most recent versions of advanced tools for thermalhydraulic analysis, in order to verify the effectiveness of plant systems in avoiding or minimizing core damage. This thesis presents the evaluation of the capabilities of a best estimate thermalhydraulic code, RELAP5/MOD3.2, for realistically simulating the application of an emergency procedure, and additional measures, for a total loss of feedwater scenario in Angra 2, a four-loop pressurized water reactor nuclear power plant operating in Brazil. The recently proposed procedure consists of a complete depressurization of the steam generator secondary side. This measure aims at enhancing the passive feed to the steam generator through the feedwater lines and feedwater tank to restore core cooling. An experiment performed at the LOB/MOD2 facility, built with a scale factor of approximately 1/700 relatively to a design similar to that of Angra 2, was the reference data basis for this activity. The results pointed to a large possibility for the control during, at least, six hours after the transient beginning. (author)

  19. A training simulator to support the Loviisa VVER-440 severe accident management programme

    A simulation tool for training operators and technical support personnel for severe accidents is being developed at VTT. The system will be accomplished by implementing severe accident models into the APROS - Advanced Process Simulator - environment, which already includes a model of the Loviisa VVER-440 plant. The system development is closely coupled with the plant severe accident management programme. The Loviisa severe accident management programme consists of four high level actions: primary system depressurization, retention of molten core within the pressure vessel, hydrogen control and containment external spray cooling. The training system will at the first stage simulate in simple terms the key phenomena associated with these actions and their effect on the plant response. The paper describes the system objectives, outline and modelling philosophy

  20. Influence of accident management strategies on source terms of VVER-1000-type reactors

    The source term can be mitigated by effective accident management. The goal of this work is the investigation of the influence of a number of accident management strategies on the source term of a VVER-1000-type reactor. This work is one of a series of studies investigating the behavior of a VVER-1000-type reactor during severe accidents. In particular, it is based on the study in which the pressure rise in the containment and the melt-through of the cavity bottom was investigated, indicating potential mitigation strategies. To rate the usefulness of these strategies, the source terms of selected scenarios are also calculated in the present work. All the calculations were performed using the Source Term Code Package; hydrogen explosions are not considered. For the first time, the source term behavior of these scenarios was simulated up to the very end of the accident the solidification of the melt

  1. Hydrogen management and the metamorphosis of NRC policy on severe nuclear accident risk

    From the early days of light water reactor developments, it was understood that, following a loss-of-coolant accident, hydrogen could accumulate inside the primary reactor containment as a result of: (1) metal-water reaction involving the fuel element cladding; (2) the radiolytic decomposition of the water in the reactor core and the containment sump; (3) the corrosion of certain construction materials by some spray solutions; and (4) possible synergistic effects of chemical, thermal and radiolytic by-products of accidents on containment protective coatings and electric cable insulation. The NRC's policy decisions regarding hydrogen management prior to and in light of the TMI-2 loss of coolant accident are discussed

  2. Requirement analysis of computerized procedures of AP1000 severe accident management guidelines

    Computerized procedures are drawing increased interest for application in nuclear power plants to enhance operator performance, especially in the accident conditions. AP1000 Severe Accident Management Guidelines (SAMG) are established to protect the containment fission product boundaries and to mitigate the accident consequences. This paper introduces the AP1000 SAMG, and according to the functional requirements of the Computerized Procedure System (CPS), some requirements are analyzed. These requirements are special to the Computerized AP1000 SAMG, which need to be especially noticed in the design process. (author)

  3. Facility accident considerations in the US Department of Energy Waste Management Program

    A principal consideration in developing waste management strategies is the relative importance of Potential radiological and hazardous releases to the environment during postulated facility accidents with respect to protection of human health and the environment. The Office of Environmental Management (EM) within the US Department of Energy (DOE) is currently formulating an integrated national program to manage the treatment, storage, and disposal of existing and future wastes at DOE sites. As part of this process, a Programmatic Environmental impact Statement (PEIS) is being prepared to evaluate different waste management alternatives. This paper reviews analyses that have been Performed to characterize, screen, and develop source terms for accidents that may occur in facilities used to store and treat the waste streams considered in these alternatives. Preliminary results of these analyses are discussed with respect to the comparative potential for significant releases due to accidents affecting various treatment processes and facility configurations. Key assumptions and sensitivities are described

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

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

    2011-01-01

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

  5. WASTE-ACC: A computer model for analysis of waste management accidents

    Nabelssi, B.K.; Folga, S.; Kohout, E.J.; Mueller, C.J.; Roglans-Ribas, J.

    1996-12-01

    In support of the U.S. Department of Energy`s (DOE`s) Waste Management Programmatic Environmental Impact Statement, Argonne National Laboratory has developed WASTE-ACC, a computational framework and integrated PC-based database system, to assess atmospheric releases from facility accidents. WASTE-ACC facilitates the many calculations for the accident analyses necessitated by the numerous combinations of waste types, waste management process technologies, facility locations, and site consolidation strategies in the waste management alternatives across the DOE complex. WASTE-ACC is a comprehensive tool that can effectively test future DOE waste management alternatives and assumptions. The computational framework can access several relational databases to calculate atmospheric releases. The databases contain throughput volumes, waste profiles, treatment process parameters, and accident data such as frequencies of initiators, conditional probabilities of subsequent events, and source term release parameters of the various waste forms under accident stresses. This report describes the computational framework and supporting databases used to conduct accident analyses and to develop source terms to assess potential health impacts that may affect on-site workers and off-site members of the public under various DOE waste management alternatives.

  6. WASTE-ACC: A computer model for analysis of waste management accidents

    In support of the U.S. Department of Energy's (DOE's) Waste Management Programmatic Environmental Impact Statement, Argonne National Laboratory has developed WASTE-ACC, a computational framework and integrated PC-based database system, to assess atmospheric releases from facility accidents. WASTE-ACC facilitates the many calculations for the accident analyses necessitated by the numerous combinations of waste types, waste management process technologies, facility locations, and site consolidation strategies in the waste management alternatives across the DOE complex. WASTE-ACC is a comprehensive tool that can effectively test future DOE waste management alternatives and assumptions. The computational framework can access several relational databases to calculate atmospheric releases. The databases contain throughput volumes, waste profiles, treatment process parameters, and accident data such as frequencies of initiators, conditional probabilities of subsequent events, and source term release parameters of the various waste forms under accident stresses. This report describes the computational framework and supporting databases used to conduct accident analyses and to develop source terms to assess potential health impacts that may affect on-site workers and off-site members of the public under various DOE waste management alternatives

  7. Support calculations for management of PRISE leakage accidents

    Matejovic, P.; Vranka, L. [Nuclear Power Plants Research Inst. Vuje, Trnava (Slovakia)

    1997-12-31

    Accidents involving primary-to-secondary leakage (PRISE) caused by rupture of one or a few tubes are well known design basis events in both, western and VVER NPPs. Operating experience and in-service inspections of VVER-440 units have demonstrated also the potential for large PRISE leaks in the case of the steam generator (SG) primary collector cover lift-up (Rovno NPP). Without performing any countermeasure for limitation of SG collector cover lift-up, a full opening results in PRISE leak with an equivalent diameter 107 mm. Although this accident was not considered in the original design, this event is usually analysed as DBA too. Different means are available for detection and mitigation of PRISE leakage in NPPs currently in operation (J.Bohunice V-1 and V-2) or under construction (Mochovce) in Slovakia. 8 refs.

  8. Motor vehicle accidents: How should cirrhotic patients be managed?

    2012-01-01

    Motor vehicle accidents (MVAs) are serious social issues worldwide and driver illness is an important cause of MVAs. Minimal hepatic encephalopathy (MHE) is a complex cognitive dysfunction with attention deficit, which frequently occurs in cirrhotic patients independent of severity of liver disease. Although MHE is known as a risk factor for MVAs, the impact of diagnosis and treatment of MHE on MVA-related societal costs is largely unknown. Recently, Bajaj et al demonstrated valuable findings...

  9. Markov Model of Severe Accident Progression and Management

    Bari, R.A.; Cheng, L.; Cuadra,A.; Ginsberg,T.; Lehner,J.; Martinez-Guridi,G.; Mubayi,V.; Pratt,W.T.; Yue, M.

    2012-06-25

    The earthquake and tsunami that hit the nuclear power plants at the Fukushima Daiichi site in March 2011 led to extensive fuel damage, including possible fuel melting, slumping, and relocation at the affected reactors. A so-called feed-and-bleed mode of reactor cooling was initially established to remove decay heat. The plan was to eventually switch over to a recirculation cooling system. Failure of feed and bleed was a possibility during the interim period. Furthermore, even if recirculation was established, there was a possibility of its subsequent failure. Decay heat has to be sufficiently removed to prevent further core degradation. To understand the possible evolution of the accident conditions and to have a tool for potential future hypothetical evaluations of accidents at other nuclear facilities, a Markov model of the state of the reactors was constructed in the immediate aftermath of the accident and was executed under different assumptions of potential future challenges. This work was performed at the request of the U.S. Department of Energy to explore 'what-if' scenarios in the immediate aftermath of the accident. The work began in mid-March and continued until mid-May 2011. The analysis had the following goals: (1) To provide an overall framework for describing possible future states of the damaged reactors; (2) To permit an impact analysis of 'what-if' scenarios that could lead to more severe outcomes; (3) To determine approximate probabilities of alternative end-states under various assumptions about failure and repair times of cooling systems; (4) To infer the reliability requirements of closed loop cooling systems needed to achieve stable core end-states and (5) To establish the importance for the results of the various cooling system and physical phenomenological parameters via sensitivity calculations.

  10. Markov Model of Severe Accident Progression and Management

    The earthquake and tsunami that hit the nuclear power plants at the Fukushima Daiichi site in March 2011 led to extensive fuel damage, including possible fuel melting, slumping, and relocation at the affected reactors. A so-called feed-and-bleed mode of reactor cooling was initially established to remove decay heat. The plan was to eventually switch over to a recirculation cooling system. Failure of feed and bleed was a possibility during the interim period. Furthermore, even if recirculation was established, there was a possibility of its subsequent failure. Decay heat has to be sufficiently removed to prevent further core degradation. To understand the possible evolution of the accident conditions and to have a tool for potential future hypothetical evaluations of accidents at other nuclear facilities, a Markov model of the state of the reactors was constructed in the immediate aftermath of the accident and was executed under different assumptions of potential future challenges. This work was performed at the request of the U.S. Department of Energy to explore 'what-if' scenarios in the immediate aftermath of the accident. The work began in mid-March and continued until mid-May 2011. The analysis had the following goals: (1) To provide an overall framework for describing possible future states of the damaged reactors; (2) To permit an impact analysis of 'what-if' scenarios that could lead to more severe outcomes; (3) To determine approximate probabilities of alternative end-states under various assumptions about failure and repair times of cooling systems; (4) To infer the reliability requirements of closed loop cooling systems needed to achieve stable core end-states and (5) To establish the importance for the results of the various cooling system and physical phenomenological parameters via sensitivity calculations.

  11. Prevention of the causes and consequences of a criticality accident - measures adopted in France

    The question of safety in regard to criticality accident risks has two aspects: prevention of the cause and limitation of the consequences. These two aspects are closely connected. The effort devoted to prevention of the causes depends on the seriousness of the possible human psychologic and economic consequences of the accident. The criticality accidents which have occurred in the nuclear industry, though few in number, do reveal the imperfect nature of the techniques adopted to prevent the causes, and also constitute the only available realistic basis for evaluating the consequences and developing measures to limit them. The authors give a analysis of the known causes and consequences of past criticality accidents and on this basis make a number of comments concerning: the validity of traditional safety criteria, the probability of accidents for different types of operations, characteristic accidents which can serve as models, and the extent of possible radiological consequences. The measures adopted in France to limit the consequences of a possible criticality accident under the headings: location, design and lay-out of the installations, accident detection, and dosimetry for the exposed personnel, are briefly described after a short account of the criteria used in deciding on them. (author)

  12. The Role Of Industrial Safety Measures In Combating Occupational Hazards And Accidents In India

    Sharmistha Bhattacharjee

    2012-01-01

    The presence of occupational hazards and industrial accidents de-motivates the worker to contribute their best to the organization. The participation of workers in the workplace which promises safety and security fosters teamwork, quality of product high productivity and a good communication between management and the industrial workers. For combating occupational hazards and accidents in an industrial site, safety is necessary and a challenging issue in an industrial environment. S...

  13. Applying of Reliability Techniques and Expert Systems in Management of Radioactive Accidents

    Accidents including radioactive exposure have variety of nature and size. This makes such accidents complex situations to be handled by radiation protection agencies or any responsible authority. The situations becomes worse with introducing advanced technology with high complexity that provide operator huge information about system working on. This paper discusses the application of reliability techniques in radioactive risk management. Event tree technique from nuclear field is described as well as two other techniques from nonnuclear fields, Hazard and Operability and Quality Function Deployment. The objective is to show the importance and the applicability of these techniques in radiation risk management. Finally, Expert Systems in the field of accidents management are explored and classified upon their applications

  14. Traffic accident and emission reduction through intermittent release measures for heavy fog weather

    Shi, Jing; Tan, Jin-Hua

    2015-09-01

    Heavy fog weather can increase traffic accidents and lead to freeway closures which result in delays. This paper aims at exploring traffic accident and emission characteristics in heavy fog, as well as freeway intermittent release measures for heavy fog weather. A driving simulator experiment is conducted for obtaining driving behaviors in heavy fog. By proposing a multi-cell cellular automaton (CA) model based on the experimental data, the role of intermittent release measures on the reduction of traffic accidents and CO emissions is studied. The results show that, affected by heavy fog, when cellular occupancy ρ safety and reduce emissions.

  15. A compendium of the measurements related to the Chernobyl nuclear accident

    Results of radiation measurements performed in Belgium after the Chernobyl accident are presented. Contamination of air, soil, milk, grass, fruit, vegetables and water is studied. The committed effective dose equivalents for the population are estimated. (MCB)

  16. Application of the integral code MELCOR for German NPPs and use within accident management and PSA projects

    The paper summarizes the application of MELCOR to German NPPS with PWR and BWR. A development of different code systems like ATHLET/ATHLET-CD, COCOSYS and ASTEC is done as well at GRS but it is not discussed in this paper. GRS has been using MELCOR since 1990 for real plant calculations. The results of MELCOR analyses are used mainly in PSA level 2 studies and in Accident Management projects for both types of NPPs. MELCOR has been a very useful and robust tool for these analyses. The calculations performed within the PSA level 2 studies for both types of German NPPs have shown that typical severe accident scenarios are characterized by several phases and that the consideration of plant specifics are important not only for realistic source term calculations. An overview of typically severe accident phases together with main accident management measures installed in German NPPs is presented in the paper. Several severe accident sequences have been calculated for both reactor types and some detailed nodalisation studies and code to code comparisons have been prepared in the past, to prove the developed core, reactor circuit and containment/building nodalisation schemes. Together with the compilation of the MELCOR data set, the qualification of the nodalisation schemes has been pursued with comparative calculations with detailed GRS codes for selected phases of severe accidents. The results of these comparative analyses showed in most of the areas a good agreement of essential parameters and of the general description of the plant behaviour during the accident progression. The in general detail of the German plant nodalisation schemes developed for MELCOR contributes significantly to this good agreement between integral and detailed code results. The implementation of MELCOR into the GRS simulator ATLAS was very important for the assessment of the results, not only due to the great detail of the nodalisation schemes used. It is used for training of severe accident

  17. Medical management of two accidents by ionizing radiations

    Two cases of accidents of occupationally exposed personnel are presented, the first one was an accidental sharp exhibition to whole body that a 27 year-old worker suffered when being exposed to a source of Iridium 192 of 94 Curies. For this case it was diagnosed an hematopoiesis syndrome that it was responded to the handling prescribe until him recovery. The second case, a radiologist technical 22 year-old that was irradiated with a source of Iridium 192 of 79 Curies. The treatment consisted on cleaning, antibiotics, analgesic and inert ointment, being achieved recovery after several weeks

  18. Triage and medical management of criticality accident victims

    The criticality accident is the result of an uncontrolled chain fission reaction initiated when the quantities of nuclear materials (uranium or plutonium)present accidentally exceed a given limit called the critical mass. As soon as the critical state is exceeded, the chain reaction increases exponentially. The result is a fast increase in the number of fission events which occur within the fissile medium. This phenomenon results in a release of energy mainly in the form of heat, accompanied by the intense emission of neutron and gamma radiation and the release of fission gases (Barby, 1983)

  19. Review of current status for designing severe accident management support system

    Jeong, Kwang Sub

    2000-05-01

    The development of operator support system (OSS) is ongoing in many other countries due to the complexity both in design and in operation for nuclear power plant. The computerized operator support system includes monitoring of some critical parameters, early detection of plant transient, monitoring of component status, plant maintenance, and safety parameter display, and the operator support system for these areas are developed and are being used in some plants. Up to now, the most operator support system covers the normal operation, abnormal operation, and emergency operation. Recently, however, the operator support system for severe accident is to be developed in some countries. The study for the phenomena of severe accident is not performed sufficiently, but, based on the result up to now, the operator support system even for severe accident will be developed in this study. To do this, at first, the current status of the operator support system for normal/abnormal/emergency operation is reviewed, and the positive aspects and negative aspects of systems are analyzed by their characteristics. And also, the major items that should be considered in designing the severe accident operator support system are derived from the review. With the survey of domestic and foreign operator support systems, they are reviewed in terms of the safety parameter display system, decision-making support system, and procedure-tracking system. For the severe accident, the severe accident management guideline (SAMG) which is developed by Westinghouse is reviewed; the characteristics, structure, and logical flow of SAMG are studied. In addition, the critical parameters for severe accident, which are the basis for operators decision-making in severe accident management and are supplied to the operators and the technical support center, are reviewed, too.

  20. Management for the prevention of accidents from disused sealed radioactive sources

    The objective of this report is to provide advice to sealed radiation source (SRS) users, radioactive waste operators, and other concerned public sectors on the measures to be taken to reduce the risk of accidents associated with disused or spent SRS. The report also explains policies as well as technical and administrative procedures to minimize the risk of accidents and to mitigate the consequences should an accident occur. The report emphasizes areas of high risk in handling disused or spent SRS in any form and condition to help to save health, life and financial resources

  1. The measurement of power reactor stack releases under accident conditions

    The performance of a typical Swedish monitor for ventilation stack radioactivity releases is examined critically with respect to accident generated radioactive particles. The conditions in the stack, particle character, and the monitor design are considered. A large LOCA outside the containment leads to high relative humidity, and high temperature, or mist in the stack. A small external LOCA results in a moderate increase in temperature and humidity, and condensing conditions only with reduced ventilation. Particle size and stickiness are estimated for different types of accident. A particle is sticky if it adheres after contact with a solid, smooth, dry, and clean surface. The monitor performance is concluded to be poor for large, sticky particles, like mist droplets. Dense aerosols, like fire smoke, will plug the sampling filter. Non-sticky particles are generally sampled with acceptable accuracy. (au)

  2. The Fukushima Dai Ichi accident. The narrative of the station manager. Volume 1. The destruction

    While outlining that the Fukushima accident could have been more severe without the courage and action of men who stayed at the controls of the plant under the management of Masao Yoshida, this book proposes a translation of the manager's narrative made for the official inquiry commission. He tells the story of a team of workers facing a disaster foretold. Besides this narrative, the authors propose a discussion on emergency engineering, present the Kan inquiry commission, present the power station and recall the circumstances of the accident and its consequences. Several hearings are reported

  3. Mental health effects from radiological accidents and their social management

    Mental health effects resulting from exposure to radiation have been identified principally in the context of large radiological accidents. They cover an extended scope of manifestations in relation with the notion of stress: increase of some hormones, modifications in mental concentration, symptoms of anxiety and depression, psycho-somatic diseases, deviation behaviours, and, on the long term, a possible post-traumatic stress disorder (PTSD). The main results come from the Three Mile Island, Goiania, and Chernobyl accidents and several modifying factors have been identified. Considering those facts, diverse social responses can be brought to reduce the detriment to affected individuals and communities. Medical treatments are necessary for persons who suffer from pathological diseases. In most cases, a structured public health follow-up is required to establish the seriousness of the health problems, to forecast the extent of medical and psychological assistance, and to inform people who express fears and worries. Social assistance is always valuable under various forms: financial compensations, preferential medical care, and particular advantages concerning working and living conditions. If this social assistance is necessary and helpful, it also induces a loss in personal adjustment capability and initiative capacity. To overcome those negative impacts, some guidelines to authorities' action can be set up. But the best approach, not excluding the previous ones, remains problem solving at the local level through community responsibilization; some instructive examples come from the Chernobyl experience. (author)

  4. Mental health effects from radiological accidents and their social management

    Brenot, J.; Charron, S.; Verger, P. [Institute for Protection and Nuclear Safety, Fontenay-aux-Roses Cedex (France)

    2000-05-01

    Mental health effects resulting from exposure to radiation have been identified principally in the context of large radiological accidents. They cover an extended scope of manifestations in relation with the notion of stress: increase of some hormones, modifications in mental concentration, symptoms of anxiety and depression, psycho-somatic diseases, deviation behaviours, and, on the long term, a possible post-traumatic stress disorder (PTSD). The main results come from the Three Mile Island, Goiania, and Chernobyl accidents and several modifying factors have been identified. Considering those facts, diverse social responses can be brought to reduce the detriment to affected individuals and communities. Medical treatments are necessary for persons who suffer from pathological diseases. In most cases, a structured public health follow-up is required to establish the seriousness of the health problems, to forecast the extent of medical and psychological assistance, and to inform people who express fears and worries. Social assistance is always valuable under various forms: financial compensations, preferential medical care, and particular advantages concerning working and living conditions. If this social assistance is necessary and helpful, it also induces a loss in personal adjustment capability and initiative capacity. To overcome those negative impacts, some guidelines to authorities' action can be set up. But the best approach, not excluding the previous ones, remains problem solving at the local level through community responsibilization; some instructive examples come from the Chernobyl experience. (author)

  5. Fuel performance under transients, and accident management using Geno-Fuzzy concept for nuclear reactors

    Simulation of Pressurized Water Reactor Power Plant (PWR) has been investigated by simulating all components installed in the power plant namely: the reactor core, steam generator, pressurizer, reactor coolant pumps, and turbine. All plant components have been introduced. This simulator is useful for transient analysis studies, engineering designs, safety analysis, and accident management. Accidents in Pressurized Water Reactor Nuclear Power Plant (PWR NPP) may be occurred either due to component failures or human error during maintenance or operation. The main target of accident management is to mitigate accidents if it occurs. The Geno-Fuzzy concept is the way to select some important plant state variables as a gene for the overall plant state chromosome. The selected genes are: reactor power, primary coolant pressure, steam generator water level, and onset boiling on clad surface which has direct impact on fuel behavior. Each of these genes has associated fuzzy level. The main objective of Geno-Fuzzy is turning the plant gene from abnormal states to the normal state by associated control variable using the inference wise fuzzy technique. The Pressurized Water Reactor Nuclear Power Plant simulator has been tested for a typical PWR, for normal transients, Anticipated Transient Without Scram (ATWS), and using the proposed Geno-Fuzzy concept for accident management, which gives very good results in reactor accident mitigation. Some of these tested accidents are; reactor control rod ejection, change in turbine steam load, and loss of coolant flow, which have direct effects on fuel safety and performance. The parameters affecting the behavior of the reactor fuel integrity are analyzed to be considered in future reactor designs. (author)

  6. Is the current management system at Statoil sufficient to prevent potential major accidents from happening at the Snorre A platform?

    Mork, Monica

    2013-01-01

    Only small margins prevented the gas-blow out at one of Statoil's platforms, Snorre A, to develop into a major accident in 2004. The underlying reasons of the accident showed extensive improvement areas, including Statoil's management system. The purpose is to find out whether the current management system at Statoil is sufficient to prevent potential major accidents from happening at the Snorre A platform again. As a guidance, four questions have been deduced. These include if...

  7. Quality assurance in military medical research and medical radiation accident management.

    Hotz, Mark E; Meineke, Viktor

    2012-08-01

    The provision of quality radiation-related medical diagnostic and therapeutic treatments cannot occur without the presence of robust quality assurance and standardization programs. Medical laboratory services are essential in patient treatment and must be able to meet the needs of all patients and the clinical personnel responsible for the medical care of these patients. Clinical personnel involved in patient care must embody the quality assurance process in daily work to ensure program sustainability. In conformance with the German Federal Government's concept for modern departmental research, the international standard ISO 9001, one of the relevant standards of the International Organization for Standardization (ISO), is applied in quality assurance in military medical research. By its holistic approach, this internationally accepted standard provides an excellent basis for establishing a modern quality management system in line with international standards. Furthermore, this standard can serve as a sound basis for the further development of an already established quality management system when additional standards shall apply, as for instance in reference laboratories or medical laboratories. Besides quality assurance, a military medical facility must manage additional risk events in the context of early recognition/detection of health risks of military personnel on deployment in order to be able to take appropriate preventive and protective measures; for instance, with medical radiation accident management. The international standard ISO 31000:2009 can serve as a guideline for establishing risk management. Clear organizational structures and defined work processes are required when individual laboratory units seek accreditation according to specific laboratory standards. Furthermore, international efforts to develop health laboratory standards must be reinforced that support sustainable quality assurance, as in the exchange and comparison of test results within

  8. Occupational Radiation Protection in Severe Accident Management. EG-SAM Interim Report

    As an early response to the Fukushima NPP accident, the ISOE Bureau decided to focus on the following issues as an initial response of the joint program after having direct communications with the Japanese official participants in April 2011; - Management of high radiation area worker doses: It has been decided to make available the experience and information from the Chernobyl accident in terms of how emergency worker / responder doses were legally and practically managed, - Personal protective equipment for highly-contaminated areas: It was agreed to collect information about the types of personnel protective equipment and other equipment (e.g. air bottles, respirators, air-hoods or plastic suits, etc.), as well as high-radiation area worker dosimetry use (e.g. type, number and placement of dosimetry) for different types of emergency and high-radiation work situations. Detailed information was collected on dose criteria which are used for emergency workers/responders and their basis, dose management criteria for high dose/dose rate areas, protective equipment which is recommended for emergency workers / responders, recommended individual monitoring procedures, and any special requirement for assessment from the ISOE participating nuclear utilities and regulatory authorities and made available for Japanese utilities. With this positive response of the ISOE actors and interest in the situation in Fukushima, the Expert Group on Occupational Radiation Protection in Severe Accident Management (EG-SAM) was established by the ISOE Management Board in May 2011. The overall objective of the EG-SAM is to contribute to occupational exposure management (providing a view on management of high radiation area worker doses) within the Fukushima plant boundary with the ISOE participants and to develop a state-of-the- art ISOE report on best radiation protection management practices for proper radiation protection job coverage during severe accident initial response and recovery

  9. Management of accident radioactive waste from Czech nuclear power plants

    A mobile decontamination unit is proposed for the treatment of waste resulting from a LOCA type design accident associated with a compensable or non-compensable primary circuit coolant leakage. The unit comprises a sorption-filtration module for the removal of toxic nuclides, a vitrification module for the solidification of spent inorganic sorbents, and a module for exhaust gases decontamination. The beta-gamma activity of liquid waste is reduced in sorption columns packed with mordenite, to a level enabling its further treatment in the standard decontamination plant of the nuclear power station. The spent inorganic sorbent is mixed with low-melting borosilicate glass and vitrified at 1050 degC, yielding a product suitable for disposal. The result of a long-term cesium leaching rate test of this product, performed according to ISO 6961, is Rn=8.6x10-8 g.cm-2.day-1. (author) 7 tabs., 10 refs

  10. Implementation of the severe accident management in Slovenske Elektrarne, subsidiary of ENEL

    Implementation of the Severe Accident Management (SAM) in Slovenske Elektrarne, subsidiary of ENEL, is a process initiated well before the Fukushima Daiichi accident. The main goal was to cover, comprehensively, level 4 of the Defense in Depth (DiD). The process included development of plant specific severe accident management guidelines (SAMGs) and installation of hardware modifications dedicated to mitigation of severe accidents as an upgrade the original VVER-440/V213. The SAM modifications have been developed with the aim to address all main generic vulnerabilities of VVER-440/V213 containments identified during initial analysis supporting the development of plant specific SAMGs. SAM modifications, in addition to their original purpose, improved plant response also at the level 3 of DiD. SAM modifications installed on VVER-440/V213 units in operation or under construction in Slovakia can be considered as an independent and diverse provision for the main safety functions: core subcriticality, core heat removal and confinement integrity. Basic set of SAM modifications includes independent diesel generator (DG), independent external source of borated water and containment vacuum breaker. Major contribution to safety from SAM modifications has been proved to be the implementation of in-vessel retention, hydrogen management in the containment and reliable depressurization of RCS. The complete set of SAM modifications installed incorporates dedicated SAM I and C to allow for determination and monitoring of plant status via dedicated instrumentation and control of SAM equipment installed at plants during a severe accident. SAM project including updating of SAMGs has been successfully completed on both units of Bohunice NPP and respective activities are continuing on operating units no. 1 and 2 in Mochovce with expected deadline in 2015. The basic design of Mochovce units no. 3 and 4 that are under construction has been modified to incorporate hardware changes

  11. 福岛第一核电厂严重事故管理研究%Research on severe accident management in Fukushima Daiichi Nuclear Power Plant

    刘凯; 王炜

    2013-01-01

    The accident of Fukushima Nuclear Power Plant led to a severe accident of core meltdown, and its process of emergency management exposed various defects which raised great concern about severe accident management in nuclear power plants. In this paper, the specifications of severe accident management that issued by IAEA and Japan were overviewed. Based on Japan specifications, the analysis of sequences and management strategies were presented on severe accident in Fukushima Daiichi Nuclear Power Plant. Following identification of defects on severe accident management, possible corrective measures for current and future plants were discussed. Finally , an approach and a frame model for severe accident management were presented, which may improve nuclear safety in current and future plants.%日本福岛核事故造成了堆芯熔毁的严重事故,应急处置过程暴露出严重事故管理的种种不足,引起对核电厂严重事故管理的关注.简述了国际原子能机构和日本关于核电厂严重事故管理的规范要求,分析了福岛第一核电厂事故序列和严重事故管理策略,讨论了严重事故管理存在的问题及其可能的改进措施,最后提出了改进核电厂严重事故管理的框架模型和方法.

  12. Workshop on iodine aspects of severe accident management. Summary and conclusions

    Following a recommendation of the OECD Workshop on the Chemistry of Iodine in Reactor Safety held in Wuerenlingen (Switzerland) in June 1996 [Summary and Conclusions of the Workshop, Report NEA/CSNI/R(96)7], the CSNI decided to sponsor a Workshop on Iodine Aspects of Severe Accident Management, and their planned or effective implementation. The starting point for this conclusion was the realization that the consolidation of the accumulated iodine chemistry knowledge into accident management guidelines and procedures remained, to a large extent, to be done. The purpose of the meeting was therefore to help build a bridge between iodine research and the application of its results in nuclear power plants, with particular emphasis on severe accident management. Specifically, the Workshop was expected to answer the following questions: - what is the role of iodine in severe accident management? - what are the needs of the utilities? - how can research fulfill these needs? The Workshop was organized in Vantaa (Helsinki), Finland, from 18 to 20 May 1999, in collaboration with Fortum Engineering Ltd. It was attended by forty-six specialists representing fifteen Member countries and the European Commission. Twenty-eight papers were presented. These included four utility papers, representing the views of Electricite de France (EDF), Teollisuuden Voima Oy and Fortum Engineering Ltd (Finland), the Nuclear Energy Institute (USA), and Japanese utilities. The papers were presented in five sessions: - iodine speciation; - organic compound control; - iodine control; - modeling; - iodine management; A sixth session was devoted to a general discussion on iodine management under severe accident conditions. This report summarizes the content of the papers and the conclusions of the workshop

  13. Intervention levels for protective measures in nuclear accidents

    A radiation protection philosophy for exposure situations following an accident has been developed by international organisations such as the ICRP, IAEA, NES/OECD, FAO/WHO, and the CEC during the last decade. After the Chernobyl accident, the application of radiation protection principles for intervention situations such as exposure from accidental contamination or radon in dwellings were further developed and this work is still in progress. The present intervention policy recommended by the international organisations as well as by the Nordic radiation protection authorities is reviewed. The Nordic Intervention levels for foodstuff restrictions, both for the Chernobyl and post-Chernobyl periods, have been based on dose limits and they are therefore in conflict with international intervention policy. Illustrative examples on intervention level setting for relocation and foodstuff restrictions are derived for Nordic conditions from the optimisation principle recommended by the international organisations. Optimised Generic Intervention Levels have been determined to be about 10 mSv x month-1 for relocation/return and 5,000-30,000 Bqxkg-1 for restrictions on various foodstuffs contaminated with 137Cs and 131I. (au) (14 tabs., 1 ill., 16 refs.)

  14. Conclusions of the specialist meeting on operator AIDS for severe accident management and training (SAMOA)

    The scope of the Specialist Meeting was limited to operator aids for accident management which were in operation or could be soon. Moreover, the meeting concentrated on the management of accidents beyond the design basis, including tools which might be extended from the design basis range into the severe accident area. Relevant simulation tools for operator training were also part of the scope of the meeting. The presentations showed that the design and implementation of operator aids were closely related to the organisation adopted by the user, whether it was a utility or a governmental agency. The most common organisation is to share the management of severe accidents among two groups of people: the operating team in the Control Room (CR) and a team of specialists in a Technical Support Centre (TSC). The CR is in charge of the operation of the plant in all conditions using a set of procedures and guidelines, while the experts in the TSC are able to produce in-depth analyses of the plant state and its evolution. The responsibility is shared between the CR and the TSC during accident progression. The TSC acts as a support for the CR for reactor operation and takes charge of the predictions of radioactive releases (source term, accident progression, release and dispersion of radioactive substances, as well as the interaction with public authorities). But this type of organisation is not general and the differences can induce different approaches in the design of operator aids. The first session was dedicated to operator aids for control rooms, the second session to operator aids for technical support centres

  15. Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment failure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects

  16. An examination of the accident and emergency management of deliberate self harm.

    Dennis, M; BEACH, M; Evans, P A; Winston, A.; Friedman, T.

    1997-01-01

    OBJECTIVE: To examine the adequacy of assessment and management of deliberate self harm (DSH) undertaken by accident and emergency (A&E) medical staff. METHODS: The records for attendances to the Leicester Royal Infirmary A&E department with a diagnosis of "self inflicted" injury for the 12 month period April 1994 to March 1995 were scrutinised. If the episode was identified as DSH, then assessment and management were examined, using an instrument based on the Royal College of Psychiatrists' ...

  17. Uncertainty quantification for accident management using ACE surrogates

    The alternating conditional expectation (ACE) regression method is used to generate RELAP5 surrogates which are then used to determine the distribution of the peak clad temperature (PCT) during the loss of feedwater accident coupled with a subsequent initiation of the feed and bleed (F and B) operation in the Zion-1 nuclear power plant. The construction of the surrogates assumes conditional independence relations among key reactor parameters. The choice of parameters to model is based on the macroscopic balance statements governing the behavior of the reactor. The peak clad temperature is calculated based on the independent variables that are known to be important in determining the success of the F and B operation. The relationship between these independent variables and the plant parameters such as coolant pressure and temperature is represented by surrogates that are constructed based on 45 RELAP5 cases. The time-dependent PCT for different values of F and B parameters is calculated by sampling the independent variables from their probability distributions and propagating the information through two layers of surrogates. The results of our analysis show that the ACE surrogates are able to satisfactorily reproduce the behavior of the plant parameters even though a quasi-static assumption is primarily used in their construction. The PCT is found to be lower in cases where the F and B operation is initiated, compared to the case without F and B, regardless of the F and B parameters used. (authors)

  18. Motor vehicle accidents: how should cirrhotic patients be managed?

    Kawaguchi, Takumi; Taniguchi, Eitaro; Sata, Michio

    2012-06-01

    Motor vehicle accidents (MVAs) are serious social issues worldwide and driver illness is an important cause of MVAs. Minimal hepatic encephalopathy (MHE) is a complex cognitive dysfunction with attention deficit, which frequently occurs in cirrhotic patients independent of severity of liver disease. Although MHE is known as a risk factor for MVAs, the impact of diagnosis and treatment of MHE on MVA-related societal costs is largely unknown. Recently, Bajaj et al demonstrated valuable findings that the diagnosis of MHE by rapid screening using the inhibitory control test (ICT), and subsequent treatment with lactulose could substantially reduce the societal costs by preventing MVAs. Besides the ICT and lactulose, there are various diagnostic tools and therapeutic strategies for MHE. In this commentary, we discussed a current issue of diagnostic tools for MHE, including neuropsychological tests. We also discussed the advantages of the other therapeutic strategies for MHE, such as intake of a regular breakfast and coffee, and supplementation with zinc and branched chain amino acids, on the MVA-related societal costs. PMID:22690067

  19. Motor vehicle accidents: How should cirrhotic patients be managed?

    Takumi Kawaguchi; Eitaro Taniguchi; Michio Sata

    2012-01-01

    Motor vehicle accidents (MVAs) are serious social issues worldwide and driver illness is an important cause of MVAs.Minimal hepatic encephalopathy (MHE) is a complex cognitive dysfunction with attention deficit,which frequently occurs in cirrhotic patients independent of severity of liver disease.Although MHE is known as a risk factor for MVAs,the impact of diagnosis and treatment of MHE on MVA-related societal costs is largely unknown.Recently,Bajaj et al demonstrated valuable findings that the diagnosis of MHE by rapid screening using the inhibitory control test (ICT),and subsequent treatment with lactulose could substantially reduce the societal costs by preventing MVAs,Besides the ICT and lactulose,there are various diagnostic tools and therapeutic strategies for MHE.In this commentary,we discussed a current issue of diagnostic tools for MHE,including neuropsychological tests.We also discussed the advantages of the other therapeutic strategies for MHE,such as intake of a regular breakfast and coffee,and supplementation with zinc and branched chain amino acids,on the MVA-related societal costs.

  20. Strategy implemented for a safe management of the waste arising from the Goiania accident

    The management of radioactive waste after the accident is discussed. Several aspects such as properties of the waste, the available infrastructure for its collection, the decontamination logistics, the motivation and commitment of works and the politically sensitive definition of handling different waste as well as the administrative procedure to set up reliable records on the collected waste are studied. Four years after the accident, corrosion was detected in some packages. Waste reconditioning, development and implementation of waste data base and development of a national safety evaluation procedure for the final disposal facility are presented

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

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

  2. Evaluation of BDB accident management in PSA for recent German 1300 MW PWRs (Konvoi)

    The Siemens AG/KWU has been performing the probabilistic safety assessment (PSA) for the nuclear power plants (NPPs) for more than 25 years for purposes of design optimization, safety research and special licensing issues. Focus of the PSA application nowadays is towards development of advanced NPPs such as EPR and 1,000 MW BWR, periodic safety review of operating plants, development and implementation of BDB (beyond design basis)-AM (accident management) measure, and so on. Here were discussed on the last two topics. As a results, PSA gave underline of high safety level on basic design in a plant expressed by the already low hazard states frequency and the balanced design, and it was recognized that efficiency of the BDB emergency measures and procedures expressed reduction of frequency required for plant damage states, importance of the emergency procedures for mitigating damage potential of reactor coolant pressure boundary failure under pressed conditions, and representation of backfitted BDB AM measures for an additional level in multi-level safety concept of the plants. (G.K.)

  3. Development and application of a radioactivity evaluation technique the to obtain radiation exposure dose of radioactivity evaluation technique when a severe accident occurs in the a power station of a severe accident. Accident management guidelines of knowledge-based maintenance

    As a One of the lessons learned from the nuclear accident at the Fukushima Daiichi Nuclear Power Stations of Tokyo Electric Power Company, the was the need for improvement of accident management guidelines is required. In this report study, we developed and applied a dose evaluation technique to evaluated the radiation dose in a nuclear power plant assuming three conditions: employees were evacuation evacuated at the time of a severe accident occurrence; operators carried out the accident management operation; of the operators, and the repair work was carried out for of the trouble damaged apparatuses in a the nuclear power plant using a dose evaluation system. The following knowledge findings were obtained and should to be reflected to in the knowledge base of the guidelines was obtained. (1) By making clearly identifying an areas beforehand becoming the that would receive high radiation doses at the time of a severe accident definitely beforehand, we can employees can be moved to the evacuation places through an areas having of low dose rate and it is also known it how much we long employees can safely stay in the evacuation places. (2) When they circulate CV containment vessel recirculation sump water is recirculated by for the accident management operation and the restoration of safety in the facilities, because the plumbing piping and the apparatuses become radioactive radioactivity sources, the dose evaluation of the shortest access route and detour access routes with should be made for effective the accident management operation is effective. Because the area where a dose rate rises changes which as safety apparatuses are restored, in consideration of a plant state, it is necessary to judge the rightness or wrongness of the work continuation from the spot radioactive dose of the actual apparatus area, with based on precedence of the need to restore with precedence, and to choose a system to be used for accident management. (author)

  4. [Methodological aspects of measuring injuries from traffic accidents at the site of occurrence].

    Híjar-Medina, M C; López-López, M V; Flores-Aldana, M; Anaya, R

    1997-02-01

    Traffic accidents are a well-known public health problem worldwide. In Mexico research into risk factors for motor involving vehicles accidents and their consequences has recently been taken into account. The relevant literature does not normally describe the methodological aspects involved in the collection of primary data, since most studies have used secondary data the good quality and validity of which are assumed. The paper presented seeks to discuss and share with researchers in this field, some of the methodological aspects to be considered in the attempt to recreate the scene of the accident and obtain information approximating to reality. The measurements in situ of, such traffic accident variables as injury, use of seat belt, speed and alcohol intake are discussed. PMID:9430931

  5. Regulatory Research of the PWR Severe Accident. Information Needs and Instrumentation for Hydrogen Control and Management

    Park, Gun Chul; Suh, Kune Y.; Lee, Jin Yong; Lee, Seung Dong [Seoul Nat' l Univ., Seoul (Korea, Republic of)

    2001-03-15

    The current research is concerned with generation of basic engineering data needed in the process of developing hydrogen control guidelines as part of accident management strategies for domestic nuclear power plants and formulating pertinent regulatory requirements. Major focus is placed on identification of information needs and instrumentation methods for hydrogen control and management in the primary system and in the containment, development of decision-making trees for hydrogen management and their quantification, the instrument availability under severe accident conditions, critical review of relevant hydrogen generation model and phenomena In relation to hydrogen behavior, we analyzed the severe accident related hydrogen generation in the UCN 3{center_dot}4 PWR with modified hydrogen generation model. On the basis of the hydrogen mixing experiment and related GASFLOW calculation, the necessity of 3-dimensional analysis of the hydrogen mixing was investigated. We examined the hydrogen control models related to the PAR(Passive Autocatalytic Recombiner) and performed MAAP4 calculation in relation to the decision tree to estimate the capability and the role of the PAR during a severe accident.

  6. Regulatory Research of the PWR Severe Accident. Information Needs and Instrumentation for Hydrogen Control and Management

    The current research is concerned with generation of basic engineering data needed in the process of developing hydrogen control guidelines as part of accident management strategies for domestic nuclear power plants and formulating pertinent regulatory requirements. Major focus is placed on identification of information needs and instrumentation methods for hydrogen control and management in the primary system and in the containment, development of decision-making trees for hydrogen management and their quantification, the instrument availability under severe accident conditions, critical review of relevant hydrogen generation model and phenomena In relation to hydrogen behavior, we analyzed the severe accident related hydrogen generation in the UCN 3·4 PWR with modified hydrogen generation model. On the basis of the hydrogen mixing experiment and related GASFLOW calculation, the necessity of 3-dimensional analysis of the hydrogen mixing was investigated. We examined the hydrogen control models related to the PAR(Passive Autocatalytic Recombiner) and performed MAAP4 calculation in relation to the decision tree to estimate the capability and the role of the PAR during a severe accident

  7. Use of a fuzzy decision-making method in evaluating severe accident management strategies

    In developing severe accident management strategies, an engineering decision would be made based on the available data and information that are vague, imprecise and uncertain by nature. These sorts of vagueness and uncertainty are due to lack of knowledge for the severe accident sequences of interest. The fuzzy set theory offers a possibility of handling these sorts of data and information. In this paper, the possibility to apply the decision-making method based on fuzzy set theory to the evaluation of the accident management strategies at a nuclear power plant is scrutinized. The fuzzy decision-making method uses linguistic variables and fuzzy numbers to represent the decision-maker's subjective assessments for the decision alternatives according to the decision criteria. The fuzzy mean operator is used to aggregate the decision-maker's subjective assessments, while the total integral value method is used to rank the decision alternatives. As a case study, the proposed method is applied to evaluating the accident management strategies at a nuclear power plant

  8. Proceedings of the second OECD specialist meeting on operator aids for severe accident management - SAMOA-2

    The second OECD Specialist Meeting on Operator Aids for Severe Accident Management (SAMOA-2) was organized in Lyon, France from 8 to 10 September 1997 in collaboration with the Thermal and Nuclear Studies and Project Department (SEPTEN) of Electricite de France. It was attended by 33 specialists representing ten OECD Member countries, the OECD Halden Reactor Project, the Commission of the European Communities, and the Russian Federation. The scope of SAMOA-2 was limited to operator aids for accident management which were in operation or could be soon. The meeting concentrated on the management of accidents beyond the design basis, including tools which might be extended from the design basis range into the severe accident area. Relevant simulation tools for operator training were also part of the scope of the meeting. Twenty papers were presented during the meeting, grouped into three sessions. Session 1: operator aids for control rooms; Session 2: operator aids for technical support centres; session 3: simulation tools for operator training. There were two demonstrations of computerized systems: the ATLAS analysis simulator developed by GRS, and EDF's 'Simulateurs Post Accidentels' (SIPA). There was also a video demonstration of the Full Scope Simulator developed by a joint Russian-U.S. team for the Leningrad nuclear power

  9. PCTRAN-3: The third generation of personal computer-based plant analyzer for severe accident management

    PCTRAN is a plant analyzer that uses a personal computer to simulate plant response. The plant model is recently expanded to accommodate beyond design-basis severe accidents. In the event of multiple failures of the plant safety systems, the core may experience heatup and extensive failure. Using a high-powered personal computer (PC), PCTRAN-3 is designed to operate at a speed significantly faster than real-time. A convenient, interactive and user-friendly graphics interface allows full control by the operator. The plant analyzer is intended for use in severe accident management. In this paper the code's component models and sample runs ranging from normal operational transients to severe accidents are reviewed. (author)

  10. Knowledge data base for severe accident management of nuclear power plants

    For the safety enhancement of Nuclear Power Plants (NPPs), continuous efforts are very important to take in the up-to-date scientific and technical knowledge positively and to reflect them into the safety regulation. The purpose of the present study is to gather effectively the scientific and technical knowledge about the severe accident (SA) phenomena and the accident management (AM) for prevention and mitigation of SA, and to take in the experimental data by participating in the international cooperative experiments regarding the important SA phenomena and the effectiveness of AM. Based on those data and knowledge, JNES is developing and improving severe accident analysis models to maintain the SA analysis codes and the AM knowledge base for assessment of the NPPs in Japan. The activities in fiscal year 2012 are as follows; Analytical study on OECD/NEA projects such as MCCI, SERENA and SFP projects, and support in making regulation for SA. (author)

  11. RBMK-1500 accident management for loss of long-term core cooling

    Results of the Level 1 probabilistic safety assessment of the Ignalina NPP has shown that in topography of the risk, transients dominate above the accidents with LOCAs and failure of the core long-term cooling are the main factors to frequency of the core damage. Previous analyses have shown, that after initial event, as a rule, the reactivity control, as well as short-term and intermediate cooling are provided. However, the acceptance criteria of the long-term cooling are not always carried out. It means that from this point of view the most dangerous accident scenarios are the scenarios related to loss of the core long-term cooling. On the other hand, the transition to the core condition due to loss of the long-term cooling specifies potential opportunities for the management of the accident consequences. Hence, accident management for the mitigation of the accident consequences should be considered and developed. The most likely initiating event, which probably leads to the loss of long term cooling accident, is station blackout. The station blackout is the loss of normal electrical power supply for local needs with an additional failure on start-up of all diesel generators. In the case of loss of electrical power supply MCPs, the circulating pumps of the service water system and MFWPs are switched-off. At the same time, TCV of both turbines are closed. Failure of diesel generators leads to the non-operability of the ECCS long-term cooling subsystem. It means the impossibility to feed MCC by water. The analysis of the station blackout for Ignalina NPP was performed using RELAP5 code. (author)

  12. Severe accident management: a summary of the VAHTI and ROIMA projects

    Two severe accident research projects: 'Severe Accident Management' (VAHTI), 1994-96 and 'Reactor Accidents' Phenomena and Simulation (ROIMA) 1997-98. have been conducted at VTT Energy within the RETU research programme. The main objective was to assist the severe accident management programmes of the Finnish nuclear power plants. The projects had several subtopics. These included thermal hydraulic validation of the APROS code, studies of failure mode of the BWR pressure vessel, investigation of core melt progression within a BWR pressure vessel, containment phenomena, development of a computerised severe accident training tool, and aerosol behaviour experiments. The last topic is summarised by another paper in the seminar. The projects have met the objectives set at the project commencement. Calculation tools have been developed and validated suitable for analyses of questions specific for the Finnish plants. Experimental fission product data have been produced that can be used to validate containment aerosol codes. The tools and results have been utilised in plant assessments. One of the main achievements has been the computer code PASULA for analysis of interactions between core melt and pressure vessel. The code has been applied to pressure vessel penetration analysis. The results have shown the importance of the nozzle construction. Modelling possibilities have recently improved by addition of a creep and porous debris models. Cooling of a degraded BWR core has been systematically studied as joint Nordic projects with a set of severe accident codes. Estimates for coolable conditions have been provided. Recriticality due to reflooding of a damaged core has been evaluated. (orig.)

  13. Hindsight Bias in Cause Analysis of Accident

    Atsuo Murata; Yasunari Matsushita

    2014-01-01

    It is suggested that hindsight becomes an obstacle to the objective investigation of an accident, and that the proper countermeasures for the prevention of such an accident is impossible if we view the accident with hindsight. Therefore, it is important for organizational managers to prevent hindsight from occurring so that hindsight does not hinder objective and proper measures to be taken and this does not lead to a serious accident. In this study, a basic phenomenon potentially related to accidents, that is, hindsight was taken up, and an attempt was made to explore the phenomenon in order to get basically insights into the prevention of accidents caused by such a cognitive bias.

  14. Emergency preparedness and measures to prevent severe accidents in the Republic of Korea

    The paper reviews the national programme for prevention and mitigation of severe nuclear plant accidents and emergency preparedness in case of nuclear accidents in the Republic of Korea. The programme has implemented most post-Three Mile Island measures for safety improvements and set up a national system of emergency response to handle any nuclear related accident. The programme has also thoroughly examined safety related equipment and operating procedures of operating reactors in the Republic of Korea. As a result of the safety enhancing activities, Korea Electric Power Corporation (KEPCO) is establishing an emergency response facility, a post-accident sampling system and full scope probablistic risk analysis work. After the Chernobyl accident, the Government of the Republic of Korea went through a safety check-up of the operating plants once again and KEPCO installed a retraining programme for reactor operators and an upgraded safety check-up procedure and schedule. Improvements were made on a number of safety systems including an emergency core cooling system, a fire monitoring system and a quality assurance programme for fire prevention. In addition, the national programme has been setting up an international co-operative system in order to respond quickly to any unexpected accident through rapid mobilization of international experts, equipment and materials. (author). 1 fig., 1 tab

  15. Result of measuring inner radioactive contamination of people due to the Chernobyl accident

    A survey of results of about 140 measurements of internal radioactive contamination of Dutch people in consequence of the Chernobylsk accident is presented. The measurements were performed with total body counters by the Dutch Institute for Radiopathology and Radiation Protection (IRS) and the Radiologic Department TNO (RD-TNO). 4 figs.; 3 tabs

  16. Radiation management at the occurrence of accident and restoration works. Fire and explosion of asphalt solidification processing facility

    Fire and explosion accident in the cell of Asphalt Solidification Processing Facility(ASP) in PNC took placed at March 11 in 1997. Following to the alarm of many radiation monitoring system in the facility, some of workers inhale radioactive materials in their bodies. Indication values of an exhaust monitor installed in the first auxiliary exhaust stack increased suddenly. A large number of windows, doors, and shutters in the facility were raptured by the explosion. A lot of radioactive materials blew up and were released to the outside of the facility. Reinforcement of radiation surveillance function, nose smearing test for the workers and confirmation of contamination situation were implemented on the fire. Investigation of radiation situation, radiation management on the site, exposure management for the workers, surveillance of exhaustion, and restoration works of the damaged radiation management monitoring system were carried out after the explosion. The detailed data of radiation management measures taken during three months after the accident are described in the paper. (M. Suetake)

  17. Radiation management at the occurrence of accident and restoration works. Fire and explosion of asphalt solidification processing facility

    Miyabe, Kenjiro; Jin, K.; Namiki, A.; Mizutani, K.; Horiuchi, N.; Saruta, J. [Power Reactor and Nuclear Fuel Development Corp., Health and Safety Division, Tokai, Ibaraki (Japan); Ninomiya, Kazushige [Power Reactor and Nuclear Fuel Development Corp., Tsuruga, Fukui (Japan). Monju Construction Office

    1998-06-01

    Fire and explosion accident in the cell of Asphalt Solidification Processing Facility(ASP) in PNC took placed at March 11 in 1997. Following to the alarm of many radiation monitoring system in the facility, some of workers inhale radioactive materials in their bodies. Indication values of an exhaust monitor installed in the first auxiliary exhaust stack increased suddenly. A large number of windows, doors, and shutters in the facility were raptured by the explosion. A lot of radioactive materials blew up and were released to the outside of the facility. Reinforcement of radiation surveillance function, nose smearing test for the workers and confirmation of contamination situation were implemented on the fire. Investigation of radiation situation, radiation management on the site, exposure management for the workers, surveillance of exhaustion, and restoration works of the damaged radiation management monitoring system were carried out after the explosion. The detailed data of radiation management measures taken during three months after the accident are described in the paper. (M. Suetake)

  18. Radiation protection measures in the case of incidents and radiation accidents

    Measures to be taken in the case of radiation accidents connected with an unusually high radiation exposure to persons, the amounts of which exceed the limiting values, with depend on whether there has been an external or an internal exposure. In order to give further treatment in the case of whole-body or partial-body irradiation, it is necessary to estimate the exposure dose. In nuclear medicine the accident doses are generally low, i.e. acute radiation damage does not occur here, and immediate measures are not necessary. Therapeutic measures in the case of incorporation accidents are only necessary when the maximum amounts for the nuclide in question recommended by the ICRP has been reached or exceeded in the organism. However, decorporation measures ought to be carried out only by qualified radiation protection physicians. The type of radiation accident which occurs most frequently in nuclear medicine is radiation exposure as a result of contamination. If in the case of contamination of a person the measurement exceeds the radioactivity limit, the decontamination measures are necessary. In the present contribution, these measures for cases without injuries are described in detail. (orig./HP)

  19. Procedural and Organizational Measures to Assist Operations During an Accident in a Nuclear Power Plant in Europe

    Bull, D.; Lathrop, J.W.; Linnerooth, J.; Sinclair, C

    1980-01-01

    This paper is concerned with aspects of organization and procedure in nuclear accident management. Because all accidents can be argued to have common characteristics, a comparative approach is taken here for the discussion of emergency planning for nuclear accidents. This approach reveals several deficiencies in selected European emergency plans, the most important concerning formal and informal communication channels. The most important principle which emerges from this discussion, and which...

  20. Management of radioactive waste during the initial period of eliminating the consequences of the Chornobyl accident. Review and analysis

    This review discusses basic sources of radioactive waste (RAW) formation during the Chornobyl accident, processes of RAW formation, and describes the relevant experience of RAW collection and disposal. Not all sources and materials were available for the research, but the author endeavored to provide the most comprehensive presentation of all the aspects of RAW management during elimination of the Chornobyl accident

  1. Instrumentation Capabilities. Their Influence on Severe Accident Management and How Operator Training can be contemplated

    No currently operating nuclear unit has been explicitly designed to withstand the loads resulting from accident sequences resulting in melting of a very significant portion of the core. As a consequence, instrumentation needs were defined based on what was deemed necessary to control the unit during normal operation and contemplated accident sequences. Detailed requirements for instrumentation were then established based on environmental conditions anticipated during accident sequences addressed in the design, estimation of additional conservatism deemed reasonable for assessing sensor robustness and information reliability, and a realistic understanding of the influence of aging. Though instrument failures could not be excluded, consequences were necessarily limited as adequate redundancy was provided by design for all information needed to adequately control the unit and bring it back to safe shutdown in case of accident could be assumed available. Training programs largely built on this very robust approach and operators were challenged to control situations whose main attributes were: - all systems needed to fulfill essential safety functions are available and have the minimal capability for allowing compliance with otherwise stated acceptance criteria, - information needed to make decisions is available and reliable, - plant evolution, if not easily understandable in all cases, is not confusing to operators as all involved physical phenomena are unambiguous on one side, and can be reasonably well monitored. However, though current plant designs are generally very robust, one cannot exclude that accident sequences involving significant melting of the core can happen. First estimates through risk studies reported in WASH-1400 showed that the risk of core-melt could not be ignored, and the TMI-2 accident in a first step, then Chernobyl confirmed this conclusion. These events gave impetus to the development of Severe Accident Management (SAM) programs, and

  2. Performance Measurement in Property Management

    Aho, Johannes

    2016-01-01

    This research examines the performance measurement of processes in property management and maintenance context. The purpose was to first, develop a framework which divides processes into modes of operation that share managerial similarities. Second, apply this framework in property management performance measurement. Constructive approach was used as the scientific research approach. An extensive literature review was first conducted after which the framework was dev...

  3. Accident management following loss of residual heat removal during mid-loop operation in a Westinghouse two-loop PWR

    Birchley, J. [Paul Scherrer Institut, CH-5232 Villigen (Switzerland)], E-mail: jonathan.birchley@psi.ch; Haste, T.J. [Paul Scherrer Institut, CH-5232 Villigen (Switzerland); Richner, M. [Nordostschweizerische Kraftwerke (NOK) - NPP Beznau, CH-5312 Doettingen (Switzerland)

    2008-09-15

    The possibility of an accident or component failure during mid-loop operation has been identified in probabilistic studies as a major contributor to core melt frequency and source term risk during shutdown conditions. The wide range of plant states encountered and the unavailability of certain safety features make it difficult to guarantee that safety systems operation will always be sufficient to terminate the accident evolution. In this context analyses are performed using MELCOR 1.8.5 for loss of residual heat removal (RHR) at various times during mid-loop operation of a Westinghouse two-loop PWR. In the absence of recovery of RHR or other accident management (AM) measures, the sequences necessarily lead to a long term core uncovery, heat-up and degradation, loss of geometry and eventual failure of the reactor pressure vessel (RPV). The results show an extensive time window before uncovery and additionally before core damage, which increase progressively with increasing time after shutdown at which loss of RHR occurs. Significant oxidation of the cladding may result in concentrations of hydrogen sufficient for deflagration. The slow evolution implies an opportunity for the plant operators to initiate AM measures even after core uncovery has started. The analyses indicate a substantial time window during the uncovery within which the injection can recover the core without damage. The upper end of the window is determined by the temperature at which heat from cladding oxidation becomes a dominant factor, marking a critical point for the effectiveness of this recovery mode. The results provide confidence in the inherent robustness of the plant with respect to accident sequences of this type.

  4. Effectiveness of core exit thermocouple (CET) indication in accident management of light water reactors

    The working group on Analysis and Management of Accidents (WGAMA) of the Committee on the Safety of Nuclear Installations (CSNI) of OECD-NEA had a task on the effectiveness of CET indication in accident management (AM) of light water reactors (LWR). The task collected and reviewed the design basis of CET application for AM procedures through a survey of the CET use in the NEA member countries, and reviewed pertinent experimental results from such test facilities as LOFT, ROSA/LSTF, PKL and PSB-VVER focusing on the time delay in CET from core temperature rise. Scaling issues were discussed considering extrapolation of experimental results to LWR. This paper summarizes major outcomes of the task and indicates possible future work. (author)

  5. The water role in a nuclear accident - Measures to be taken

    In case of nuclear accidents or natural disasters, the contaminated water plays a large part in the environment contamination. This is illustrated by two examples: Agadir earthquake and Chernobyl accident. In Agadir earthquake, the contamination of the water was caused by the mutiple breaking down of the water pipes, and in Chernobyl accident it was derived from: -The reactor cooling water; -The radioactive fallout; -The radioactive clouds. The water concentrates incessantly the radioactivity proceeding by the hydrological cycle: Evaporation, precipitation, flowing. The radio-activity concentration by the water and the atmosphere contamination are explained in this paper. In USSR, the radioactive contamination has affected several Ukranian rivers and the artificial lake of Kiev. The measures that have been taken in USSR and in the next countries to prevent the radioactive contamination propagation by water have been discussed. The reparation of chernobyl accident damages is estimated to three milliard $. Theoretically, every nation, using nuclear energy, has a protection system for the accidental situations but none of them has a second protection system for the accidental situations occuring in the distance. The measures to be taken for the latter situations, particularly in Morocco, have been cited. The lessons learnt from the chernobyl accident have served to broaden the inter-national cooperation fields. 15 figs., 1 tab. (author)

  6. Risk management and role of schools of the Tokai-village radiation accident in 1999. Safety education and risk management before and during the radiation accident from the standpoint of school nurse teachers

    The purpose of this study is to evaluate safety education and risk management in the neighborhood schools before and during the radiation accident in the Tokai-village in 1999 from the standpoint of school nurse teachers. Eighty-six school nurse teachers from 44 elementary, 25 junior-high, 14 high and 3 handicapped children's schools were surveyed within neighboring towns and villages. The main results were as follows: There had been few risk management systems against the potential radiation accidents including safety education, radiological monitoring and protection in all of the neighboring schools. There were no significant difference in risk management systems among the schools before the accident, though the anxiety rates of school children were significantly higher in the schools nearest to the accident site. Some radiation risk management systems must be established in neighboring schools including safety education, radiological monitoring and protection. (author)

  7. Generalities on nuclear accidents and their short-dated and middle-dated management; Generalites sur les accidents nucleaires et leur gestion a court terme et a long terme

    NONE

    2003-03-01

    All the nuclear activities present a radiation risk. The radiation exposure of the employees or the public, may occur during normal activity or during an accident. The IRSN realized a document on this radiation risk and the actions of protection. The sanitary and medical aspects of a radiation accident are detailed. The actions of the population protection during an accident and the post accident management are also discussed. (A.L.B.)

  8. Management of the radioactive wastes arising from the accident in Goiania, Brazil

    The radiological accident that occurred during the month of September, 1987, in Goiania, Brazil, involving a 50,9 TBq (1375 Ci) of a Caesium-137 source has led to the contamination of a large number of individuals and several urban area. The objective of the present article is to describe the waste management program that was implemented in order to deal with the c.a. 3340 m3 of wastes generated as a result of decontamination works performed

  9. Cost per severe accident as an index for severe accident consequence assessment and its applications

    The Fukushima Accident emphasizes the need to integrate the assessments of health effects, economic impacts, social impacts and environmental impacts, in order to perform a comprehensive consequence assessment of severe accidents in nuclear power plants. “Cost per severe accident” is introduced as an index for that purpose. The calculation methodology, including the consequence analysis using level 3 probabilistic risk assessment code OSCAAR and the calculation method of the cost per severe accident, is proposed. This methodology was applied to a virtual 1,100 MWe boiling water reactor. The breakdown of the cost per severe accident was provided. The radiation effect cost, the relocation cost and the decontamination cost were the three largest components. Sensitivity analyses were carried out, and parameters sensitive to cost per severe accident were specified. The cost per severe accident was compared with the amount of source terms, to demonstrate the performance of the cost per severe accident as an index to evaluate severe accident consequences. The ways to use the cost per severe accident for optimization of radiation protection countermeasures and for estimation of the effects of accident management strategies are discussed as its applications. - Highlights: • Cost per severe accident is used for severe accident consequence assessment. • Assessments of health, economic, social and environmental impacts are included. • Radiation effect, relocation and decontamination costs are important cost components. • Cost per severe accident can be used to optimize radiation protection measures. • Effects of accident management can be estimated using the cost per severe accident

  10. Radioactivity measurements of the HMI after the Chernobyl reactor accident

    The report explains the methods applied and the data measured by the HMI campaign. The material is presented so as to be of interest also to readers who in general are not concerned with aspects of radiation protection. The data measured refer to the local dose rate and to radioactivity in the environment (air, rain, surface waters, soil, food, mother's milk. Also, results of measurements of samples from Eastern Europe are given. (orig./HP)

  11. Accident analysis for high-level waste management alternatives in the US Department of Energy Environmental Restoration and Waste Management Programmatic Environmental Impact Statement

    A comparative generic accident analysis was performed for the programmatic alternatives for high-level waste (HLW) management in the US Department of Energy Environmental Restoration and Waste Management Programmatic Environmental Impact Statement (EM PEIS). The key facilities and operations of the five major HLW management phases were considered: current storage, retrieval, pretreatment, treatment, and interim canister storage. A spectrum of accidents covering the risk-dominant accidents was analyzed. Preliminary results are presented for HLW management at the Hanford site. A comparison of these results with those previously advanced shows fair agreement

  12. Combining Neural Methods and Knowledge-Based Methods in Accident Management

    Miki Sirola

    2012-01-01

    Full Text Available Accident management became a popular research issue in the early 1990s. Computerized decision support was studied from many points of view. Early fault detection and information visualization are important key issues in accident management also today. In this paper we make a brief review on this research history mostly from the last two decades including the severe accident management. The author’s studies are reflected to the state of the art. The self-organizing map method is combined with other more or less traditional methods. Neural methods used together with knowledge-based methods constitute a methodological base for the presented decision support prototypes. Two application examples with modern decision support visualizations are introduced more in detail. A case example of detecting a pressure drift on the boiling water reactor by multivariate methods including innovative visualizations is studied in detail. Promising results in early fault detection are achieved. The operators are provided by added information value to be able to detect anomalies in an early stage already. We provide the plant staff with a methodological tool set, which can be combined in various ways depending on the special needs in each case.

  13. Development of Evaluation Technology for Hydrogen Combustion in containment and Accident Management Code for CANDU

    Kim, S. B.; Kim, D. H.; Song, Y. M.; and others

    2011-08-15

    For a licensing of nuclear power plant(NPP) construction and operation, the hydrogen combustion and hydrogen mitigation system in the containment is one of the important safety issues. Hydrogen safety and its control for the new NPPs(Shin-Wolsong 1 and 2, Shin-Ulchin 1 and 2) have been evaluated in detail by using the 3-dimensional analysis code GASFLOW. The experimental and computational studies on the hydrogen combustion, and participations of the OEDE/NEA programs such as THAI and ISP-49 secures the resolving capabilities of the hydrogen safety and its control for the domestic nuclear power plants. ISAAC4.0, which has been developed for the assessment of severe accident management at CANDU plants, was already delivered to the regulatory body (KINS) for the assessment of the severe accident management guidelines (SAMG) for Wolsong units 1 to 4, which are scheduled to be submitted to KINS. The models for severe accident management strategy were newly added and the graphic simulator, CAVIAR, was coupled to addition, the ISAAC computer code is anticipated as a platform for the development and maintenance of Wolsong plant risk monitor and Wolsong-specific SAMG.

  14. Medical and psychological aspects of crisis management during a nuclear accident

    Drottz-Sjoeberg, B.M.

    1993-06-01

    Crisis handling in most kinds of disasters is affected by e.g. the information situation, prior experience and preparedness, availability of resources, efficiency of leadership and coordination, and type of disaster. A nuclear accident creates a situation which differs from many `normal` disasters and natural catastrophes, for example with respects to the invisible nature of radiation and radioactive contamination and thus the dependence on access to specific technical equipment and expertise, and to information about the radiation situation. The scope of the accident, and the existing levels of radiation, define subsequent actions; information policies and existing channels of communication lay the foundation for public reactions. The present paper explores some examples of public reactions, and crisis handling of some previous radiation accidents on the basis of two dimensions, i.e. degree of information availability and degree of impact or `environmental damage`. The examples include the radiation accidents in the Chelyabinsk region in the southern Urals, at Three Mile Island, USA, at Chernobyl in the Ukraine, and in Goiania, Brazil. It is concluded that public reactions differ as a function of existing expectations, and the crisis handling is more affected by the existing organizational and social structures than by needs and reactions of potential victims. Another conclusion is that pre-disaster preparedness regarding public information, and organization of countermeasures, are crucial to the outcome of a successful crisis handling and for enhancing public trust in crisis management. 39 refs, 2 figs.

  15. A procedure to optimize the timing of operator actions of accident management procedures

    The analysis of beyond design basis accidents (BDBA) is an essential component of the safety concept of nuclear power plants (NPP). Goal of the analysis is to achieve a set of actions aimed to prevent the escalation into a severe accident, to mitigate consequences of a severe accident, and to achieve a long term controllable state of the NPP. This paper presents an analytical procedure to optimize the timing of operator interventions. The procedure is demonstrated based on four sets of parameters, first, parameters which define the operator actions are chosen. Second, parameters which define the system availability are chosen. Third, parameters which define in a continuous way the status of the plant are chosen. Finally, one looks for a functional dependency of the accident management (AM)-parameters and the parameters describing the plant status. Once a function could be found, this function is 'optimized' in the sense that the AM-parameters are varied to find a optimal overall condition for the plant. In the first part, the paper presents the analytical procedure in a general way, in the second part, an initiating event is chosen. The procedure is applied to a station black out (SBO) transient, and as operator action secondary side bleed and feed, followed by primary side bleed and feed, is foreseen. As result, the optimal timing to initiate both actions is achieved

  16. Medical and psychological aspects of crisis management during a nuclear accident

    Crisis handling in most kinds of disasters is affected by e.g. the information situation, prior experience and preparedness, availability of resources, efficiency of leadership and coordination, and type of disaster. A nuclear accident creates a situation which differs from many 'normal' disasters and natural catastrophes, for example with respects to the invisible nature of radiation and radioactive contamination and thus the dependence on access to specific technical equipment and expertise, and to information about the radiation situation. The scope of the accident, and the existing levels of radiation, define subsequent actions; information policies and existing channels of communication lay the foundation for public reactions. The present paper explores some examples of public reactions, and crisis handling of some previous radiation accidents on the basis of two dimensions, i.e. degree of information availability and degree of impact or 'environmental damage'. The examples include the radiation accidents in the Chelyabinsk region in the southern Urals, at Three Mile Island, USA, at Chernobyl in the Ukraine, and in Goiania, Brazil. It is concluded that public reactions differ as a function of existing expectations, and the crisis handling is more affected by the existing organizational and social structures than by needs and reactions of potential victims. Another conclusion is that pre-disaster preparedness regarding public information, and organization of countermeasures, are crucial to the outcome of a successful crisis handling and for enhancing public trust in crisis management. 39 refs, 2 figs

  17. Proceedings of the International Workshop on Occupational Radiation Protection in Severe Accident Management 'sharing practices and experiences'

    The objective of the Workshop on Occupational Radiation Protection in Severe Accident Management was to share practices and experiences in approaches to severe accident management. The workshop: provided an international forum for information and experience exchange amongst nuclear electricity utilities and national regulatory authorities on approaches to, and issues in severe accident management, including national and international implications. Focus was placed on sharing practices and experiences in many countries on approaches to severe accident management; identified best occupational radiation protection approaches in strategies, practices, as well as limitations for developing effective management. This included experiences in various countries; identified national experiences to be incorporated into the final version of ISOE EG-SAM report. The workshop included a series of plenary presentations that provided participants with an overview of practices and experiences in severe accident management from various countries. Furthermore, by taking into account the structure of the interim report, common themes and issues were discussed in follow-up breakout sessions. Sessions included invited speakers, moderated by designated experts, allowing participants to discuss their national experiences and possible inputs into the report. The outcomes of the breakout sessions were presented in plenary by the respective moderators followed by an open discussion, with a view towards elaborating ways forward to achieve more effective severe accident management. This document brings together the abstracts and the slides of the available presentations

  18. Doctrinal elements for the post-accidental management of a nuclear accident - Final version

    This report examines and defines the objectives, principles and main actions for the post-accidental management of a nuclear accident. It defines the emergency phase and the post-accidental phase, three basic objectives (to protect the population against the hazards of ionizing radiations, to support populations affected by the accident consequences, to restore affected territories), management principles, key issues for post-accidental management. It defines actions to be undertaken: post-accidental zoning, monitoring of deposited radioactivity, early actions for the protection and taking charge of population, information. It addresses the different aspects of post-accidental management planning in a period of transition: reception of population, reduction of population exposure to deposited radioactivity, treatment of public health problems, improvement of the knowledge on the radiological situation of the environment, improvement of the radiological quality of the different environments, dealing with wastes, empowerment of stakeholders through an adequate governance, support and redeployment of economic activity, help and compensation, information. Appendices more deeply discuss actions to be undertaken just after the emergency phase, for the management of the transition period, and for the management of the long-term period

  19. Results of measurements of internal contamination of persons caused by the Chernobyl accident

    A preliminary summary is presented of c.a. 140 measurements of internal radioactive contamination, motivated by the Chernobyl accident, in Dutch persons, by the Radiological Service TNO, Arnhem and the Institute for Radiopathology and Radiation Protection, Leiden, the Netherlands. 4 figs.; 3 tabs

  20. Measured transfer factors in milk and meat after the Chernobyl reactor accident

    After the nuclear reactor accident at Chernobyl the radioactivity in the environment in Aachen was measured in detail at the Lehrgebiet Strahlenschutz in der Kerntechnik. The change of the different radionuclides in the eco-system made it possible to obtain radioecological parameters especially for iodine and caesium. The knowledge about the transport of iodine into cow's milk could be very much improved

  1. Regulations concerning protective measures against radiation in case of accidents connected with gamma radiography

    Pursuant to the 1938 Act on X-rays and radium, etc and the 1976 regulations on supervision and use of installations and apparatus which release radiation representing a hazard to health, the State Institute of Radiation Health issued these regulations laying down a series of protective measures in case of accidents when using gamma radiography. They entered into force on 1 July 1981

  2. Insights into the control of the release of iodine, cesium, strontium and other fission products in the containment by severe accident management

    atmosphere; - discussion of various possibilities to retain the fission products in the sumps, on surfaces or in filters; - discussion of the possible measures. The first three items had been described in a fair amount of detail in other documents. It was agreed that PWG4's Task Group on Fission Product Phenomena in the Primary Circuit and the Containment (FPC) would be responsible for preparing this part of the report. In a second phase, members of PWG4's task Group on Containment Aspects of Severe Accident Management (CAM) 'enriched' the report with severe accident management considerations. These can be found at the end of Chapters III to VIII. The report does not aim at exhaustiveness, nor at direct applicability to severe accident management situations. Its aim is to give a picture of what is known in the area of fission product sources and how this knowledge can be put to use to mitigate them. Implementation has to be developed on a plant-by-plant basis, taking account of plant specifics. This is the task of the utilities

  3. Status report on severe accident material property measurements

    Measurements of selected material properties of molten reactor core material (corium) were made. The corium used was a mixture of UO2, ZrO2 and Zr, with oxygen content being a parameter to reflect different stages of zirconium oxidation. The mixtures used were representative of typical in-vessel melt sequences. For most measurements, the UO2/ZrO2 mass ratio was 1.51, representative of VVER/440 melt compositions and melt compositions of most US BWRs. Measurements were made of the solidus/liquidus temperatures of corium compositions using a Differential Thermal Analysis technique. Observation of the solubility of unoxidized Zr in the oxide phase was made by metallographic analysis of solidus/liquidus melt samples. The results of laminar flow corium spreading tests in one dimension were used to estimate the viscosity of corium compositions. Measured solidus and liquidus temperatures for compositions representative of Zr oxidation of 30, 50 and 70% were compared with those obtained form a phase diagram provided by Kurchatov Institute. It was found that experimental measurements agreed well with the phase diagram values at 70% oxidation, but the measured solidus temperatures were higher than those on the phase diagram and the measured liquidus temperatures were lower than those on the phase diagram at 30 and 50% oxidation. From a microstructure examination it was determined that there was no global segregation into distinct metal and oxide phases during the cooldown of a sample in which there was initially 70% Zr oxidation. Therefore it is concluded that Zr metal is soluble in the oxide phase under molten conditions. Viscosity estimates were made for compositions representative of Zr oxidation of 30, 50 and 70% by fitting the results of spreading tests to Huppert's equation. It was found that, at a temperature of 2500 C, the viscosity varied by three orders of magnitude over this range of compositions. 10 refs., 39 figs., 16 tabs

  4. Risk indices of an ecological catastrophe because of a severe accident, its insurance, and their measurement units

    The critical analysis of the existing measurement units of the risk of an ecological catastrophe because of severe accidents is performed. The mistake of using the measurement unit 'reactor/year' for estimation of ecological catastrophe's consequences is shown. The complex for risk assessment by costs to ensure the ecological safety is introduced. The index of virtual accident insurance is suggested

  5. Quantifying human and organizational factors in accident management using decision trees: the HORAAM method

    In the framework of the level 2 Probabilistic Safety Study (PSA 2) project, the Institute for Nuclear Safety and Protection (IPSN) has developed a method for taking into account Human and Organizational Reliability Aspects during accident management. Actions are taken during very degraded installation operations by teams of experts in the French framework of Crisis Organization (ONC). After describing the background of the framework of the Level 2 PSA, the French specific Crisis Organization and the characteristics of human actions in the Accident Progression Event Tree, this paper describes the method developed to introduce in PSA the Human and Organizational Reliability Analysis in Accident Management (HORAAM). This method is based on the Decision Tree method and has gone through a number of steps in its development. The first one was the observation of crisis center exercises, in order to identify the main influence factors (IFs) which affect human and organizational reliability. These IFs were used as headings in the Decision Tree method. Expert judgment was used in order to verify the IFs, to rank them, and to estimate the value of the aggregated factors to simplify the quantification of the tree. A tool based on Mathematica was developed to increase the flexibility and the efficiency of the study

  6. Identification and initial assessment of candidate BWR late-phase in-vessel accident management strategies

    Hodge, S.A.

    1991-04-15

    Work sponsored by the United States Nuclear Regulatory Commission (USNRC) to identify and perform preliminary assessments of candidate BWR (boiling water reactor) in-vessel accident management strategies was completed at Oak Ridge National Laboratory (ORNL) during fiscal year 1990. Mitigative strategies for containment events have been the subject of a companion study at Brookhaven National Laboratory. The focus of this Oak Ridge effort was the development of new strategies for mitigation of the late phase events, that is, the events that would occur in-vessel after the onset of significant core damage. The work began with an investigation of the current status of BWR in-vessel accident management procedures and proceeded through a preliminary evaluation of several candidate new strategies. The steps leading to the identification of the candidate strategies are described. The four new candidate late-phase (in-vessel) accident mitigation strategies identified by this study and discussed in the report are: (1) keep the reactor vessel depressurized; (2) restore injection in a controlled manner; (3) inject boron if control blade damage has occurred; and (4) containment flooding to maintain core and structural debris in-vessel. Additional assessments of these strategies are proposed.

  7. Measuring project portfolio management maturity

    Hänninen, Kirsti

    2016-01-01

    The thesis is researching portfolio management maturity in organizations that have project type of work. The objective of the thesis is to define what factors affect portfolio management maturity, how the maturity level can be evaluated and create a method for measuring current level of maturity. The thesis also provides maturity level improvement suggestions. Why is maturity measurement useful? The organizations that have project type of work often have some standardized practices. But t...

  8. Accident Management ampersand Risk-Based Compliance With 40 CFR 68 for Chemical Process Facilities

    A risk-based logic model is suggested as an appropriate basis for better predicting accident progression and ensuing source terms to the environment from process upset conditions in complex chemical process facilities. Under emergency conditions, decision-makers may use the Accident Progression Event Tree approach to identify the best countermeasure for minimizing deleterious consequences to receptor groups before the atmospheric release has initiated. It is concluded that the chemical process industry may use this methodology as a supplemental information provider to better comply with the Environmental Protection Agency's proposed 40 CFR 68 Risk Management Program rule. An illustration using a benzene-nitric acid potential interaction demonstrates the value of the logic process. The identification of worst-case releases and planning for emergency response are improved through these methods, at minimum. It also provides a systematic basis for prioritizing facility modifications to correct vulnerabilities

  9. Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt. [PWR

    Aldrich, D.C.; McGrath, P.E.; Rasmussen, N.C.

    1978-06-01

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment failure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects.

  10. Radiological accidents potentially important to human health risk in the U.S. Department of Energy waste management program

    Human health risks as a consequence of potential radiological releases resulting from plausible accident scenarios constitute an important consideration in the US Department of Energy (DOE) national program to manage the treatment, storage, and disposal of wastes. As part of this program, the Office of Environmental Management (EM) is currently preparing a Programmatic Environmental Impact Statement (PEIS) that evaluates the risks that could result from managing five different waste types. This paper (1) briefly reviews the overall approach used to assess process and facility accidents for the EM PEIS; (2) summarizes the key inventory, storage, and treatment characteristics of the various DOE waste types important to the selection of accidents; (3) discusses in detail the key assumptions in modeling risk-dominant accidents; and (4) relates comparative source term results and sensitivities

  11. Overview of plant specific severe accident management strategies for Kozloduy nuclear power plant, WWER-1000/320

    Andreeva, M. [Institute for Nuclear Research and Nuclear Energy, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)], E-mail: m_andreeva@inrne.bas.bg; Pavlova, M.P. [Institute for Nuclear Research and Nuclear Energy, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)], E-mail: pavlova@inrne.bas.bg; Groudev, P.P. [Institute for Nuclear Research and Nuclear Energy, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)], E-mail: pavlinpg@inrne.bas.bg

    2008-04-15

    This paper focuses on the fourth level of the defence in depth concept in nuclear safety, including the transitions from the third level and into the fifth level. The use of the severe accident management guideline (SAMG) is required when an accident situation is not handled adequately through the use of emergency operating procedures (EOP), thus leading to a partial or a total core melt. In the EOPs, the priority is to save the fuel, whereas, in the SAMG, the priority is to save the containment. Actions recommended in the SAMG aim at limiting the risk of radiologically significant radioactive releases in the short- and mid-term (a few hours to a few days). The paper describes basic severe accident management requirements related to nuclear power plant (NPP), specified by the IAEA and in Republic of Bulgaria Nuclear Legislation. It also surveys plant specific severe accident management (SAM) strategies for the Kozloduy NPP, equipped with WWER-1000 type reactors.

  12. Radiological effects of Chernobylsk-4 reactor accident and preventive measures to decrease its action

    Analysis of radiological effects of Chernobylsk-4 reactor accident in the USSR and preventive measures to decrease its action are given. Systematic medical examination of population and radiation situation in settlements of contaminated area confirmed efficiency of carried out preventive and protective measures. They include decontamination of settlements, removal of children and pregnant women for summer period rest, regular medical examination of local food-stuff, prohibition of conteminated food-stuff usage

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

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

  14. Radioactivity measurements in Europe after the Chernobyl accident. Part 1

    The data base presented is being set up primarily for scientific studies such as the validation of long range transport models, physico-chemical behaviour of radionuclides in air, etc. The data were extracted from the REM Data Bank at the Joint Research Centre of the Commission of the European Communities at Ispra. The data where originally obtained from written reports or were copied directly from tapes or diskettes into the REM Data Bank. In the latter case, verification with written material was persued. The original sources are given in the references. In preparing this report, only those data were retained which were fully specified as far as time, location and measuring techniques are concerned, i.e.: beginning and end of sampling, geographical coordinates and types of filters are known

  15. Safety culture and accident analysis-A socio-management approach based on organizational safety social capital

    One of the biggest challenges for organizations in today's competitive business environment is to create and preserve a self-sustaining safety culture. Typically, Key drivers of safety culture in many organizations are regulation, audits, safety training, various types of employee exhortations to comply with safety norms, etc. However, less evident factors like networking relationships and social trust amongst employees, as also extended networking relationships and social trust of organizations with external stakeholders like government, suppliers, regulators, etc., which constitute the safety social capital in the Organization-seem to also influence the sustenance of organizational safety culture. Can erosion in safety social capital cause deterioration in safety culture and contribute to accidents? If so, how does it contribute? As existing accident analysis models do not provide answers to these questions, CAMSoC (Curtailing Accidents by Managing Social Capital), an accident analysis model, is proposed. As an illustration, five accidents: Bhopal (India), Hyatt Regency (USA), Tenerife (Canary Islands), Westray (Canada) and Exxon Valdez (USA) have been analyzed using CAMSoC. This limited cross-industry analysis provides two key socio-management insights: the biggest source of motivation that causes deviant behavior leading to accidents is 'Faulty Value Systems'. The second biggest source is 'Enforceable Trust'. From a management control perspective, deterioration in safety culture and resultant accidents is more due to the 'action controls' rather than explicit 'cultural controls'. Future research directions to enhance the model's utility through layering are addressed briefly

  16. Emergency Management and Radiation Monitoring in Nuclear and Radiological Accidents. Summary Report on the NKS Project EMARAD

    In order to manage various nuclear or radiological emergencies the authorities must have pre-prepared plans. The purpose of the NKS project EMARAD (Emergency Management and Radiation Monitoring in Nuclear and Radiological Accidents) was to produce and gather various data and information that could be useful in drawing up emergency plans and radiation monitoring strategies. One of the specific objectives of the project was to establish a www site that would contain various radiation-threat and radiation-monitoring related data and documents and that could be accessed by all Nordic countries. Other important objectives were discussing various factors affecting measurements in an emergency, efficient use of communication technology and disseminating relevant information on such topics as urban dispersion and illicit use of radiation. The web server is hosted by the Radiation and Nuclear Safety Authority (STUK) of Finland. The data stored include pre-calculated consequence data for nuclear power plant accidents as well as documents and presentations describing e.g. general features of monitoring strategies, the testing of the British urban dispersion model UDM and the scenarios and aspects related to malicious use of radiation sources and radioactive material. As regards the last item mentioned, a special workshop dealing with the subject was arranged in Sweden in 2005 within the framework of the project. (au)

  17. 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

  18. High temperature measurements in severe accident experiments on the PLINIUS Platform

    Severe accident experiments are conducted on the PLINIUS platform in Cadarache, using prototypic corium. During these experiments, it is essential to measure the temperature to know the thermo-physical state of the corium in static and dynamic conditions or to monitor the concrete ablation phenomenology. Temperature in the corium can reach about 2000 to 3000 K. Such aggressive conditions restrict the type of diagnostics that can be employed to do high temperature measurements during the experiments. We employ both non-intrusive (pyrometers) and intrusive (K-type and C-type thermocouples) diagnostics. In this paper, we present the different high temperature measurements techniques and the results that can be obtained in severe accident experiments as corium heating tests and molten core concrete interaction experiments. (authors)

  19. Construct Measurement in Management Research

    Nielsen, Bo Bernhard

    2014-01-01

    Far too often do management scholars resort to crude and often inappropriate measures of fundamental constructs in their research; an approach which calls in question the interpretation and validity of their findings. Scholars often legitimize poor choices in measurement with a lack of availability...... of better measures and/or that they are simply following existing research in adopting previously published measures without critically assessing the validity, appropriateness, and applicability of such measures in terms of the focal study. Motivated by a recent dialog in Journal of Business Research......, this research note raises important questions about the use of proxies in management research and argues for greater care in operationalizing constructs with particular attention to matching levels of theory and measurement....

  20. Post-test analysis of two accident management experiments performed at the BETHSY test facility using the code ATHLET

    In the framework of the external validation of the thermal-hydraulic code ATHLET, which has been developed by the GRS, post test analyses of two experiments were done, which were performed at the french integral test facility BETHSY. During the experiment 5.2 C the complete loss of steam generator feedwater was simulated. The de-pressurization of the primary circuit and high pressure injection is assumed as an emergency measure. During the experiment 9.3 the break of a steam generator U-tube is simulated. The failure of the high pressure injection is assumed. As accident management measures, the depressurization of the steam generator secondary sides and finally of the primary circuit by opening of the pressurizer valve were investigated. The results show, that the code ATHLET is able to describe the complex scenario in good accordance with the experiment. For both tests the safety related statement could be reproduced. (author)

  1. Assisting emergency operating procedures execution with AMAS, an Accident Management Advisor System

    In an accident situation, because any decisions that the operators make will depend on how instrumentation readings are ultimately interpreted, the issue of instrument uncertainty is of paramount importance. This uncertainty exists because instrument readings may not be available in the desired form - i.e., only indirect readings for a parameter of interest may exist, with uncertainty on which physical models may be used to deduce its value from these indirect indications -, or because readings may be coming from instruments whose accuracy and reliability in the face of the severe conditions produced by the accident are far from what may be expected under normal operating conditions. In following the EOPs, the operators must rely on instrumentation whose readings may not reflect the real situation. The Accident Management Advisor System (AMAS) is a decision aid intended to supplement plant Emergency Operating Procedures (EOPs) by accounting for instrumentation uncertainty, and by alerting the operators if they are on the wrong procedures, or otherwise performing an action that is not optimal in terms of preventing core damage. In AMAS, the availability and reliability of certain important instrument readings is treated in probabilistic, rather than deterministic terms. This issue is discussed in greater detail later in the paper, since it relates to one of the key characteristics of the AMAS decision aid. (author)

  2. The roles of water addition and gas composition in AGR accident management

    Severe Accident Guidelines (SAGs) are being produced in line with best international practice for managing a severe accident in an AGR. Such an accident would be extremely unlikely due to the long timescales for recovery prior to core damage. A number of actions proposed in the SAGs are concerned with the prevention of air ingress and the deliberate injection of water into the core. Air ingress is minimized by sealing breaches and by the controlled injection of inert gases into the vessel, although at high temperatures the change in reactor gas from CO2 to CO provides a more inert atmosphere and reduces any ingressing oxygen. Consideration would be given to the injection of water into the core if the installed cooling systems could not be recovered sufficiently in a direct or improvised manner. In such cases sufficient cooling should be possible by injecting water uniformly across the core into a few tens of channels. Ideally the water would be injected before significant fission product release had occurred and before the graphite had become hot enough to oxidise in steam. This advice is reinforced by experiments with a simulated fuel pin in the presence of reactor graphite which showed that fission products released from the pin could be transported on fine (0.1 μm) graphite aerosols. (author). 4 refs, 6 figs

  3. Depressurization as an accident management strategy to minimize the consequences of direct containment heating

    Hanson, D.J.; Golden, D.W.; Chambers, R.; Miller, J.D.; Hallbert, B.P.; Dobbe, C.A. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1990-10-01

    Probabilistic Risk Assessments (PRAs) have identified severe accidents for nuclear power plants that have the potential to cause failure of the containment through direct containment heating (DCH). Prevention of DCH or mitigation of its effects may be possible using accident management strategies that intentionally depressurize the reactor coolant system (RCS). The effectiveness of intentional depressurization during a station blackout TMLB' sequence was evaluated considering the phenomenological behavior, hardware performance, and operational performance. Phenomenological behavior was calculated using the SCDAP/RELAP5 severe accident analysis code. Two strategies to mitigate DCH by depressurization of the RCS were considered. One strategy, called early depressurization, assumed that the reactor head vent and pressurizer power-operated relief valves (PORVs) were latched open at steam generator dryout. The second strategy, called late depression, assumed that the head vent and PORVs were latched open at a core exit temperature of {approximately}922 K (1200{degree}F). Depressurization of the RCS to a low value that may mitigate DCH was predicted prior to reactor pressure vessel breach for both early and late depressurization. The strategy of late depressurization is preferred over early depressurization because there are greater opportunities to recover plant functions prior to core damage and because failure uncertainties are lessened. 22 refs., 38 figs., 6 tabs.

  4. Depressurization as an accident management strategy to minimize the consequences of direct containment heating

    Probabilistic Risk Assessments (PRAs) have identified severe accidents for nuclear power plants that have the potential to cause failure of the containment through direct containment heating (DCH). Prevention of DCH or mitigation of its effects may be possible using accident management strategies that intentionally depressurize the reactor coolant system (RCS). The effectiveness of intentional depressurization during a station blackout TMLB' sequence was evaluated considering the phenomenological behavior, hardware performance, and operational performance. Phenomenological behavior was calculated using the SCDAP/RELAP5 severe accident analysis code. Two strategies to mitigate DCH by depressurization of the RCS were considered. One strategy, called early depressurization, assumed that the reactor head vent and pressurizer power-operated relief valves (PORVs) were latched open at steam generator dryout. The second strategy, called late depression, assumed that the head vent and PORVs were latched open at a core exit temperature of ∼922 K (1200 degree F). Depressurization of the RCS to a low value that may mitigate DCH was predicted prior to reactor pressure vessel breach for both early and late depressurization. The strategy of late depressurization is preferred over early depressurization because there are greater opportunities to recover plant functions prior to core damage and because failure uncertainties are lessened. 22 refs., 38 figs., 6 tabs

  5. SARNET: Sustainable integration of EU research on severe accident phenomenology and management

    In spite of the accomplishments reached in severe accident research, thanks notably to the EU projects carried out during previous Framework Programmes, a limited number of specific items remain where research activities are still necessary to reduce further uncertainties that are considered of importance for nuclear reactor safety and to consolidate severe accident management plans. Facing and anticipating budget reductions, 52 European R and D organizations, including technical supports of safety authorities, industry, utilities and universities, have decided to join their efforts in a durable way by networking their research activities in the frame of a Network of Excellence proposed as a FP-6 project called SARNET, coordinated by the French Institut de Radioprotection et de Surete Nucleaire. The integral severe accident analysis code ASTEC, developed by IRSN and GRS, will provide the backbone of the integration. Actions are proposed to integrate in ASTEC the current knowledge and all the future knowledge generated within SARNET. In addition, integrating activities will be carried out as the creation of large scientific databases, the elaboration of a research priority index, education and training. (authors)

  6. Analysis of the containment spray effect for severe accident management during Molten Core-Concrete Interaction

    Massive combustible gases generated by MCCI during a severe accident in NPP causes a problem of when we should spray the containment. The increase of hydrogen concentration due to the steam condensation caused by spraying might lead to a hydrogen burning and thus intimidate the containment integrity. In case the containment is designed to be robust enough to sustain the AICC (Adiabatic Isochoric Complete Combustion) load and to prevent DDT (Deflagration to Detonation Transition), it might be effective to spray and thus burn the hydrogen at early phase of MCCI to keep the containment integrity. Spraying the containment at late phase of MCCI might cause the containment to fail because of the increased combustible gases generation. MELCOR analysis for APR1400 shows that spraying the containment at early phase can delay the time to reach containment failure pressure by steam inerting and oxygen depletion. This kind of analysis helps us to better establish a spray actuation time for an accident management procedure against a postulated severe accident

  7. The rise of citizen competence: asset or handicap for the French authorities during post-accident management?

    EDA is a non-profit association of volunteer members aware of the limits of the planet's capacities and motivated to prepare a future based on sustainability and solidarity. The ability to plan in advance the measures to be taken in a post-accident situation, adapting them to the territory concerned and making full use of active and responsible involvement by local participants, is highly valuable. Naturally, this must not take precedence over all the preventive measures to ensure that an accident does not occur. We did not think that after 40 years of practical experience and after 20 years of observation of the seriousness of the consequence of Chernobyl we were so unprepared today. The CLIs (local information committees) attached to basic nuclear installations are indisputably the crucial links to be used to provide the necessary training of selected citizens to help manage the emergency phase and above all the post-accident phases, on condition that they receive financial resources commensurate with the challenges to be faced. It is also important to consider increasing the competence of citizens over the whole territory. Nothing can be achieved under satisfactory conditions if the citizens are not prepared well in advance: this must begin at school, so that the population knows how to behave in these situations. We appreciate being associated with the work of CODIRPA in 'calm' phases and we try to make an active contribution at our level. The accounts from Norway and Belarus are there to prevent us from forgetting that we are all concerned. From 2008 we hope to play a constructive and committed role in confronting the challenges to be faced and thank ASN for allowing us to take part in the work of CODIPPA. (author)

  8. The Risk Assessment on Arbitrary Accidents Orientating in the TSF For LILW Management

    The objective of this study is to conduct the risk assessment on arbitrary accidents originating in the TSF for LILW management through the result of dose assessment. In order to conduct the risk assessment on arbitrary accidents originating in the TSF for LILW management, the result of dose assessment was converted to the risk index. The risk conversion parameter for deriving the risk index was considered in the concept of the total risk factor suggested in the ICRP. After considering each parameter, the total risk factor was represented by the value of 7.3E·5 risk/mSv in terms of risk dimension. And then, the risk-level was also derived with respect to each risk degree. Consequently, the risk-level of all of drums was III regardless of waste stream with respect to the dropping of drums and fire. Especially, the risk originated in dropping of drums could be ignored. In opposition to many of researches on the disposal of LILW, the risk assessment on the TSF has scarcely been conducted. Furthermore, the details in regards of the safety analysis on this facility have not been considered in the preliminary and final safety analysis report because this report focused on the nuclear reactor system rather than this facility. As a consequence of these situations, the number of the researches on the arbitrary accidents occurring in the TSF has not been enough. And then, the numbers of the researches on the predisposal management of LILW have been required for the preparation on new regulatory frame

  9. RISK MANAGEMENT MEASURES IN CMMI

    Mahmoud Khraiwesh

    2012-01-01

    Risk management is a continuous process that could endanger the objectives of a project or application.Risks are handled to reduce and avoid threats effects on the objectives of the project. The sources of riskare both internal and external to the project. This research will identify general measures for the specificgoal and its specific practices of Risk Management Process Area (PA) in Capability Maturity ModelIntegration (CMMI). CMMI is developed by Software Engineering Institute (SEI) in C...

  10. Updated action plan for the implementation of measures as a consequence of the Fukushima reactor accident

    The action plan of the German government concerning the measures following the Fukushima reactor accident include the decision on the future of nuclear power in Germany, safety analyses, investigations and measures for nuclear power plants in a national frame, investigations in an international frame, planning for the implementation of CNS (Convention on nuclear safety) topics 1-3, i.e. measures to increase the robustness in German nuclear power plants, and the planning of implementation of further measures (CNS topics 4-6).

  11. A strategy for the management of milk contaminated as a result of a nuclear accident

    In the context of nuclear accidents, milk is an important foodstuff because it is produced continually in large quantities. However, the availability of both practical advice and policy level guidance on the management of contaminated milk is limited. This report draws together information on the two strategic approaches that need to be considered: waste minimisation and disposal. Data sheets and decision trees are presented to guide the user through a range of potential management options. The practicability of these options is evaluated against a set of well-established criteria. Unsuitable options are also discussed. Finally, a concise, coherent framework on which to base a broad strategy for the management of contaminated milk is proposed which may be of use to senior government advisers. Recommendations for further work are also made so that any remaining uncertainties can be addressed. (author)

  12. Measures introduced in Norway after the Chernobyl accident. A cost benefit analysis

    In the paper, the measures introduced in Norway to alleviate the adverse effects of the Chernobyl accident, and their economic consequences, are discussed. During the three years after the accident almost 20-30% of the sheep and 30-40% of the reindeer each year had activity levels above the action limits. Activity levels above the action limits were also found in goats, cattle and wild freshwater fish. Three main approaches were used in Norway in order to reduce the potential health risk after the Chernobyl accident: decreasing uptake from soil to vegetation and from fodder to animals, lowering unacceptable activity levels in animals by special feeding programmes, and reducing human intake by food condemnation and dietary advice. The total value of mutton, lamb and goat meat saved as a result of such measures in 1987 amounted to approximately 230 million Norwegian kroner (NOK) (US $33 million). The cost of the measures was approximately NOK 40 million ($5.7 million). In 1987, the total reduction in the radiation dose level to which the population was exposed was 450 man.Sv. In 1988, mutton, lamb and goat meat valued at approximately NOK 290 million ($41 million) was saved from condemnation by similar measures, which cost approximately NOK 60 million ($8,5 million). The resulting dose level reduction was approximately 200 man.Sv. The degree to which resources were used during 1987 and 1988 would appear to be justified in light of the reduction in radiation dose achieved. (author). 13 refs, 1 tab

  13. Optimization of the Severe Accident Management Strategy for Domestic Plants and Validation Experiments

    Kim, S. B.; Kim, H. D.; Koo, K. M.; Park, R. J.; Hong, S. H.; Cho, Y. R.; Kim, J. T.; Ha, K. S.; Kang, K. H

    2007-04-15

    nuclear power plants, a technical basis report and computational aid tools were developed in parallel with the experimental and analytical works for the resolution of the uncertain safety issues. ELIAS experiments were carried out to quantify the boiling heat removal rate at the upper surface of a metallic layer for precise evaluations on the effect of a late in-vessel coolant injection. T-HERMES experiments were performed to examine the two-phase natural circulation phenomena through the gap between the reactor vessel and the insulator in the APR1400. Detailed analyses on the hydrogen control in the APR1400 containment were performed focused on the effect of spray system actuation on the hydrogen burning and the evaluation of the hydrogen behavior in the IRWST. To develop the technical basis report for the severe accident management, analyses using SCDAP/RELAP5 code were performed for the accident sequences of the OPR1000. Based on the experimental and analytical results performed in this study, the computational aids for the evaluations of hydrogen flammability in the containment, criteria of the in-vessel corium cooling, criteria of the external reactor vessel cooling were developed. An ASSA code was developed to validate the signal from the instrumentations during the severe accidents and to process the abnormal signal. Since ASSA can perform the signal processing from the direct input of the nuclear power plant during the severe accident, it can be platform of the computational aids. In this study, the ASSA was linked with the computaional aids for the hydrogen flammability.

  14. Quantification of a decision-making failure probability of the accident management using cognitive analysis model

    In the nuclear power plant, much knowledge is acquired through probabilistic safety assessment (PSA) of a severe accident, and accident management (AM) is prepared. It is necessary to evaluate the effectiveness of AM using the decision-making failure probability of an emergency organization, operation failure probability of operators, success criteria of AM and reliability of AM equipments in PSA. However, there has been no suitable qualification method for PSA so far to obtain the decision-making failure probability, because the decision-making failure of an emergency organization treats the knowledge based error. In this work, we developed a new method for quantification of the decision-making failure probability of an emergency organization using cognitive analysis model, which decided an AM strategy, in a nuclear power plant at the severe accident, and tried to apply it to a typical pressurized water reactor (PWR) plant. As a result: (1) It could quantify the decision-making failure probability adjusted to PSA for general analysts, who do not necessarily possess professional human factors knowledge, by choosing the suitable value of a basic failure probability and an error-factor. (2) The decision-making failure probabilities of six AMs were in the range of 0.23 to 0.41 using the screening evaluation method and in the range of 0.10 to 0.19 using the detailed evaluation method as the result of trial evaluation based on severe accident analysis of a typical PWR plant, and a result of sensitivity analysis of the conservative assumption, failure probability decreased about 50%. (3) The failure probability using the screening evaluation method exceeded that using detailed evaluation method by 99% of probability theoretically, and the failure probability of AM in this study exceeded 100%. From this result, it was shown that the decision-making failure probability was more conservative than the detailed evaluation method, and the screening evaluation method satisfied

  15. Evolution of accident management strategies from the present to the next generation nuclear power plants

    The knowledge gained in Accident Management (A.M.) by means of studies and experiments performed for the current NPPs can be largely implemented in the advanced, passive safety NPPs, for which A.M. is still an application step of the defence in depth principle. Obviously, such implementation will take into account the safety philosophy peculiarities of the concerned plants, and the role assigned to their operators, which is largely determined by the objective of drastically reducing human errors and their potential effects on the plant safety. Indeed, in comparison with the current NPPs, the operators of the considered advanced plants are not strictly required to perform safety actions early in accidents progression, but are entrusted to follow-up and support automatic interventions of active and passive systems, and to manage the post accident plant conditions. A preliminary analysis shows that the A.M. implementation in advanced, passive safety plants could undergo as from now recognizable problems, mainly regarding: supervision needs and equipment requirements; utilization strategies for A.M. supporting equipment; staffing and training of operators; technical bases and procedures to cope with severe accidents. The related safety issues should be solved by appropriate analyses, strictly interacting with the design development. Operator's role and needs for (modes of) human intervention should be taken into account in every development stage of the concerned plants designs and should be carefully evaluated in assessing the plant response to the considered events, with deterministic and probabilistic methods. Also, specific studies and experiments should be performed, to support the development of A.M. bases and procedures, and to determine the equipment effectiveness as well. In summary: The knowledge in A.M. gained by studies and experiments performed for the current NPPs must be transferred to the advanced, passive safety NPPs, for which A. M. is still a step of

  16. Regional management of accidents risk level: strategy based on effective feedbacks

    Today the accidents prevention and environmental protection activity in Bashkortostan Republic is regulated by Governmental Programme including risk management as one of the main parts. The authors of the present paper accumulated some experience in risk management system creation because they took part in the investigations according to the mentioned Programme. Their proposal concerns this closed-loop system general structure which is planned to be based on three kinds of feedbacks: internal feedback (it utilizes the special Russian institutions for the plants state observation and limitation such 'Gosgortechnodzor', 'Gossanepidnadzor', etc...; all the noted institutions must be informed of the current situation and fulfill the actions oriented towards risk indices reduction); intermediate feedback (it is represented by the insurance system functioning with respect to insurance agencies investments into the plants operational security); external feedback (it includes the subsystem of HP security declarations analysis mechanism, special HP regional register and the expert commission whose decisions become the foundations for governmental responses). The authors consider all the feedbacks interaction in order to provide the stability of region development. The resulting strategy for accidents risk level management has been confirmed now by some normative documents in Bashkortostan Republic. (authors)

  17. Emergency management in nuclear accident situations the disaster exercise 1995 'Northern Light'

    Emergency management does not only start after something has happened. Initially, a feasibility study usually assesses the risk for technologically critical processes and applications. Preventive strategies will be employed both in the administrative and technical field to minimize risk. Technical solutions will increase inherent safety or provide monitoring of critical components. Administrative action would result e.g. in restricted access, training programs, or detailed operating protocols. A final stage would be preparation for remedial action and defining the groundwork for emergency management in cooperation with civil defense forces. Appropriate precautions will be based on hazard potential, which is inherently substantial when dealing with nuclear accidents. Being the last line of defence, the civil or military defense forces will be involved if a major disaster occurs despite all precautions, overpowering on-site crew capabilities. For major disasters requiring even international assistance, the United Nations Department of Humanitarian Affairs has started to conduct disaster preparedness exercises to improve cooperation and communication among the international relief teams and the local authorities. The EXERCISE '95 was organized by the Russian ministry for disaster management simulating a major accident in an atomic power plant located on the Kola peninsula. (author)

  18. Fitness for accident management through NPP personnel training, simulators and technical support

    The contributions within the context of accident management of the Siemens A G-Power Generation Group Crisis Centre and the Siemens A G Training Centre are described. The Crisis Centre provides direct technical consulting to NPPs from experts in design and engineering. Training of NPP personnel is here outlined with particular emphasis on the use of simulators in getting practice of emergency handling and on development of documentation and operating procedures. It is pointed to projects to the introduction of these services in Eastern NPPs and training facilities

  19. Radioactive Waste Management In The Chernobyl Exclusion Zone - 25 Years Since The Chernobyl Nuclear Power Plant Accident

    Radioactive waste management is an important component of the Chernobyl Nuclear Power Plant accident mitigation and remediation activities of the so-called Chernobyl Exclusion Zone. This article describes the localization and characteristics of the radioactive waste present in the Chernobyl Exclusion Zone and summarizes the pathways and strategy for handling the radioactive waste related problems in Ukraine and the Chernobyl Exclusion Zone, and in particular, the pathways and strategies stipulated by the National Radioactive Waste Management Program. The brief overview of the radioactive waste issues in the ChEZ presented in this article demonstrates that management of radioactive waste resulting from a beyond-designbasis accident at a nuclear power plant becomes the most challenging and the costliest effort during the mitigation and remediation activities. The costs of these activities are so high that the provision of radioactive waste final disposal facilities compliant with existing radiation safety requirements becomes an intolerable burden for the current generation of a single country, Ukraine. The nuclear accident at the Fukushima-1 NPP strongly indicates that accidents at nuclear sites may occur in any, even in a most technologically advanced country, and the Chernobyl experience shows that the scope of the radioactive waste management activities associated with the mitigation of such accidents may exceed the capabilities of a single country. Development of a special international program for broad international cooperation in accident related radioactive waste management activities is required to handle these issues. It would also be reasonable to consider establishment of a dedicated international fund for mitigation of accidents at nuclear sites, specifically, for handling radioactive waste problems in the ChEZ. The experience of handling Chernobyl radioactive waste management issues, including large volumes of radioactive soils and complex structures

  20. RADIOACTIVE WASTE MANAGEMENT IN THE CHERNOBYL EXCLUSION ZONE - 25 YEARS SINCE THE CHERNOBYL NUCLEAR POWER PLANT ACCIDENT

    Farfan, E.; Jannik, T.

    2011-10-01

    Radioactive waste management is an important component of the Chernobyl Nuclear Power Plant accident mitigation and remediation activities of the so-called Chernobyl Exclusion Zone. This article describes the localization and characteristics of the radioactive waste present in the Chernobyl Exclusion Zone and summarizes the pathways and strategy for handling the radioactive waste related problems in Ukraine and the Chernobyl Exclusion Zone, and in particular, the pathways and strategies stipulated by the National Radioactive Waste Management Program. The brief overview of the radioactive waste issues in the ChEZ presented in this article demonstrates that management of radioactive waste resulting from a beyond-designbasis accident at a nuclear power plant becomes the most challenging and the costliest effort during the mitigation and remediation activities. The costs of these activities are so high that the provision of radioactive waste final disposal facilities compliant with existing radiation safety requirements becomes an intolerable burden for the current generation of a single country, Ukraine. The nuclear accident at the Fukushima-1 NPP strongly indicates that accidents at nuclear sites may occur in any, even in a most technologically advanced country, and the Chernobyl experience shows that the scope of the radioactive waste management activities associated with the mitigation of such accidents may exceed the capabilities of a single country. Development of a special international program for broad international cooperation in accident related radioactive waste management activities is required to handle these issues. It would also be reasonable to consider establishment of a dedicated international fund for mitigation of accidents at nuclear sites, specifically, for handling radioactive waste problems in the ChEZ. The experience of handling Chernobyl radioactive waste management issues, including large volumes of radioactive soils and complex structures

  1. 32P measurement and dose conversion factor evaluation of activated human hair by criticality accident

    In order to conduct dose assessment of victims in criticality accidents, a method of fast neutron capture-activated 32P measurement of hair in which samples are treated by a chemical and analytical procedure that takes 9 h and measurement is conducted by liquid scintillation counting is presented. To validate this measurement method, hair samples spiked with a 32P reference source were measured and the results analysed and the optimal sample mass and detection efficiency were determined. To verify the correlation between 32P-specific activity and absorbed dose for spectra with two neutron mean energies, samples collected from three normal individuals were irradiated at various neutron energies and irradiation times using the MC50 Cyclotron of the Korea Institute of Radiological and Medical Sciences. The 32P-specific activity trend of the irradiated hair agreed well with the absorbed doses. Based on the results, dose conversion factors, which were 0.67±0.15 and 0.59±0.06 Gy (Bq g-1)-1 at neutron mean energies of 2.33 and 5.36 MeV, respectively, were calculated as a guide for medical treatment of criticality accident victims. In this study, a method for measuring 32P changes activated by the neutron irradiation of hair samples of criticality accident victims was developed and tested. In addition, a dose conversion factor for two neutron mean energy spectra based on these measurement results was developed. These results agree well with measured absorbed doses from exposure to fast neutron fields. The advantage of the proposed activated hair analysis method based on liquid scintillation counting is that it enables the acquisition of dose information from victims in a short time and with relatively high detection efficiency. In addition, sampling of hair is simpler than it is for other biological samples, and, finally, the conversion factor the authors developed using hair analysis data will be useful for dose assessment in real cases. However, the relation between

  2. Environmental radioactivity measurements at BNL during the year following the Chernobyl accident

    The accident which destroyed Unit 4 of the Chernobyl Nuclear Power Station on 26 April 1986 provided the world's scientists with an opportunity, unique in recent years, to study many of the processes which follow the release of large quantities of radioactivity into the atmosphere. BNL undertook a wide ranging programme of environmental measurements after the accident, the immediate aim being to supply HM Government with data to help assess the radiological consequences to the UK population. As it became clear that the UK dose commitment was relatively low, the thrust of the measurements began to be concentrated on airborne radioactivity and the movement of nuclides in the grass-soil system. The aim of these studies was to assess dispersion and diffusion of radioactivity in these particular compartments of the environment. The measurements have continued over the twelve month period since the Chernobyl accident. This report aims to disseminate the year's data and to offer some initial interpretations of the trends. (U.K.)

  3. Regulatory research of the PWR severe accident information needs and instrumentation availability for hydrogen control and management

    Park, Jae-Hong; Park, Gun-Chul; Suh, Kune Y.; Kang, Yun-Moon; Lee, Un-Jang; Oh, Se-Chul; Lee, Jin-Yong [Seoul Nationl Univ., Seoul (Korea, Republic of)

    1998-03-15

    During the current research period, we have set forth the methodology for identification of a severe accident, developed a framework for hydrogen management decision trees, and analyzed the literature on hydrogen management and experimental data for hydrogen bum. Specifically, we have summarized me results for information needs in a severe accident obtained in the U.S. and other countries, and applied the methodology to the reference plant YGN 3 and 4 as part of severe accident management. We have also examined the existing instruments in terms of their availability and survivability during a severe accident, and identified additionally needed information needs and instruments. We have identified dominant accident sequences for me reference plant YGN 3 and 4 to construct decision trees, and extracted available data from the IPE study of the plant. Based upon the data we have performed preliminary study on the decision tree and decision node. Last, we have examined various mechanisms for hydrogen generation and reIevant experimental data to predict me amount of hydrogen generation and governing factors in me process. We have also reviewed the hydrogen generation related models in the severe accident analysis.

  4. Regulatory research of the PWR severe accident information needs and instrumentation availability for hydrogen control and management

    During the current research period, we have set forth the methodology for identification of a severe accident, developed a framework for hydrogen management decision trees, and analyzed the literature on hydrogen management and experimental data for hydrogen bum. Specifically, we have summarized me results for information needs in a severe accident obtained in the U.S. and other countries, and applied the methodology to the reference plant YGN 3 and 4 as part of severe accident management. We have also examined the existing instruments in terms of their availability and survivability during a severe accident, and identified additionally needed information needs and instruments. We have identified dominant accident sequences for me reference plant YGN 3 and 4 to construct decision trees, and extracted available data from the IPE study of the plant. Based upon the data we have performed preliminary study on the decision tree and decision node. Last, we have examined various mechanisms for hydrogen generation and reIevant experimental data to predict me amount of hydrogen generation and governing factors in me process. We have also reviewed the hydrogen generation related models in the severe accident analysis

  5. Chernobylsk accident: radioactivity measurement of airborn particulate in the AM/CNR network

    Results are presented of the radioactivity measurements performed on samples of airborne particulates taken at the station of the AM/CNR network after the Chernobylsk accident. In particular for the 16 operating stations are reported tables and trends referred to the beta total measurements performed on two samples per day per station and the activity of some gamma emitting radionuclides deriving from the analysis made on concentrations of filters referred to North-Central and Southrn Italy plus Islands. The presentation of data is preceded by a short description of the AM/CNR network and the sampling system as well as the methodology of total beta measurement and gamma analysis

  6. CORRELATION OF CERVICAL LORDOSIS MEASUREMENT WITH INCIDENCE OF MOTOR VEHICLE ACCIDENTS

    Marshall, Dorothy L.; Tuchin, Peter J.

    1996-01-01

    A retrospective analysis of 500 patient radiographs was conducted to measure the clinical correlation of cervical lordosis measurements and incidence of motor vehicle accident (MVA). Five hundred lateral cervical radiographs were selected at random from the practice of one of the authors (DLM). The C1-7 angle of the cervical curve was then measured by two blinded examiners. Inter-examiner reliability had a confidence interval of 95%. Eighty-two percent of patients who have had a MVA had an ab...

  7. Report from the results of measurements of radioactive contaminations after Chernobyl accident

    The results of measurements of radioactive contamination carried out in Cracow during the first days after Chernobyl accident are presented. The particular radioisotopes were determined by gamma spectroscopy. In the period from April 28th to morning hours of May 1st 1986 radiation measurements concerned above all air. After rains considerable contamination of earth's surface was detected and measurements were concentrated on soil contamination. There were also examined water and food samples. The concentration of strontium radioisotopes was determined too. (M. F. W.)

  8. Accident patterns and prevention measures for fatal occupational falls in the construction industry.

    Chi, Chia-Fen; Chang, Tin-Chang; Ting, Hsin-I

    2005-07-01

    Contributing factors to 621 occupational fatal falls have been identified with respect to the victim's individual factors, the fall site, company size, and cause of fall. Individual factors included age, gender, experience, and the use of personal protective equipment (PPE). Accident scenarios were derived from accident reports. Significant linkages were found between causes for the falls and accident events. Falls from scaffold staging were associated with a lack of complying scaffolds and bodily action. Falls through existing floor openings were associated with unguarded openings, inappropriate protections, or the removal of protections. Falls from building girders or other structural steel were associated with bodily actions and improper use of PPE. Falls from roof edges were associated with bodily actions and being pulled down by a hoist, object or tool. Falls through roof surfaces were associated with lack of complying scaffolds. Falls from ladders were associated with overexertion and unusual control and the use of unsafe ladders and tools. Falls down stairs or steps were associated with unguarded openings. Falls while jumping to a lower floor and falls through existing roof openings were associated with poor work practices. Primary and secondary prevention measures can be used to prevent falls or to mitigate the consequences of falls and are suggested for each type of accident. Primary prevention measures would include fixed barriers, such as handrails, guardrails, surface opening protections (hole coverings), crawling boards/planks, and strong roofing materials. Secondary protection measures would include travel restraint systems (safety belt), fall arrest systems (safety harness), and fall containment systems (safety nets). PMID:15892934

  9. Initial integration of accident safety, waste management, recycling, effluent, and maintenance considerations for low-activation materials

    A true low-activation material should ideally achieve all of the following objectives: 1. The possible prompt dose at the site boundary from 100% release of the inventory should be <2 Sv (200 rem); hence, the design would be inherently safe in that no possible accident could result in prompt radiation fatalities. 2. The possible cancers from realistic releases should be limited such that the accident risk is <0.1%/yr of the existing background cancer risk to local residents. This includes consideration of elemental volatility. 3. The decay heat should be limited so that active mitigative measures are not needed to protect the investment from cooling transients; hence, the design would be passively safe with respect to decay heat. 4. Used materials could be either recycled or disposed of as near- surface waste. 5. Hands-on maintenance should be possible around coolant system piping and components such as the heat exchanger. 6. Effluent of activation products should be minor compared to the major challenge of limiting tritium effluents. The most recent studies in these areas are used to determine which individual elements and engineering materials are low activation. Grades from A (best) to G (worst) are given to each element in the areas of accident safety, recycling, and waste management. Structure/fluid combinations are examined for low-activation effluents and out-of-blanket maintenance. The lowest activation structural materials are silicon carbide, vanadium alloys, and ferritic steels. Impurities and minor alloying constituents must be carefully considered. The lowest activation coolants are helium, water, FLiBe, and lithium. The lowest activation breeders are lithium, lithium oxide, lithium silicate, and FLiBe. Designs focusing on these truly low-activation materials will help achieve the excellent safety and environmental potential of fusion energy

  10. Sustainable integration of EU research in severe accident phenomenology and management

    Highlights: → The SARNET network gathers most worldwide actors involved in severe accident research. → It defines common research programmes for resolving the most important pending safety issues. → It optimises the use of the available European resources and constitutes sustainable research groups. → It disseminates the knowledge on severe accidents through education courses. → Knowledge produced is capitalized through physical models in the ASTEC simulation code. - Abstract: In order to optimise the use of the available means and to constitute sustainable research groups in the European Union, the Severe Accident Research NETwork of Excellence (SARNET) has gathered, between 2004 and 2008, 51 organizations representing most of the actors involved in severe accident (SA) research in Europe plus Canada. This project was co-funded by the European Commission (EC) under the 6th Euratom Framework Programme. Its objective was to resolve the most important pending issues for enhancing, in regard of SA, the safety of existing and future nuclear power plants (NPPs). SARNET tackled the fragmentation that existed between the national R and D programmes, in defining common research programmes and developing common computer codes and methodologies for safety assessment. The Joint Programme of Activities consisted in: -Implementing an advanced communication tool for accessing all project information, fostering exchange of information, and managing documents; - Harmonizing and re-orienting the research programmes, and defining new ones; -Analyzing the experimental results provided by research programmes in order to elaborate a common understanding of relevant phenomena; -Developing the ASTEC code (integral computer code used to predict the NPP behaviour during a postulated SA) by capitalizing in terms of physical models the knowledge produced within SARNET; - Developing scientific databases, in which the results of research experimental programmes are stored in a common

  11. Use of an influence diagram and fuzzy probability for evaluating accident management in a BWR

    This paper develops a new approach for evaluating severe accident management strategies. At first, this approach considers accident management as a decision problem (i.e., ''applying a strategy'' vs. ''do nothing'') and uses influence diagrams. This approach introduces the concept of a ''fuzzy probability'' in the evaluation of an influence diagram. When fuzzy logic is applied, fuzzy probabilities in an influence diagram can be easily propagated to obtain results. In addition, the results obtained provide not only information similar to the classical approach using point-estimate values, but also additional information regarding the impact from imprecise input data. The proposed methodology is applied to the evaluation of the drywell flooding strategy for a long-term station blackout sequence in the Peach Bottom nuclear power plant. The results show that the drywell flooding strategy seems to be beneficial for preventing reactor vessel breach. It is also effective for reducing the probability of the containment failure for both liner melt-through and late overpressurization. Even though there exists uncertainty in the results, ''flooding'' is preferred to ''do nothing'' when evaluated in terms of expected consequences, i.e., early and late fatalities

  12. A strategy to the development of a human error analysis method for accident management in nuclear power plants using industrial accident dynamics

    This technical report describes the early progress of he establishment of a human error analysis method as a part of a human reliability analysis(HRA) method for the assessment of the human error potential in a given accident management strategy. At first, we review the shortages and limitations of the existing HRA methods through an example application. In order to enhance the bias to the quantitative aspect of the HRA method, we focused to the qualitative aspect, i.e., human error analysis(HEA), during the proposition of a strategy to the new method. For the establishment of a new HEA method, we discuss the basic theories and approaches to the human error in industry, and propose three basic requirements that should be maintained as pre-requisites for HEA method in practice. Finally, we test IAD(Industrial Accident Dynamics) which has been widely utilized in industrial fields, in order to know whether IAD can be so easily modified and extended to the nuclear power plant applications. We try to apply IAD to the same example case and develop new taxonomy of the performance shaping factors in accident management and their influence matrix, which could enhance the IAD method as an HEA method. (author). 33 refs., 17 tabs., 20 figs

  13. Radiation protection left as myth. Guidelines of nuclear disaster measures left apart from reality of accident

    The handling after the accident of TEPCO Fukushima Daiichi Nuclear Power Station is one step behind in terms of the radiation protection of residents and compensations, due to the continuous lukewarm correspondence, such as the weakness of assumptions against the accident, underestimation of risk, and dwarfing of the damage. This paper points out that the government should make utmost efforts and take various measures in the medical checkup and medical subsidy for residents in Fukushima and neighboring prefectures. For the discretion for the protection of neighboring residents in the event of an emergency, such as a nuclear accident, the 'Nuclear emergency response guidelines' of the Nuclear Regulation Authority is stipulated, in which the criteria for judging the implementation of protective measures are summarized. However, there is no description on the ground for determining initial setting values (standards) that is used in the initial stage of emergency, and this is described in the document of the meeting called 'Study team for nuclear disaster preparedness' of the Nuclear Regulation Authority. Among this, the descriptions in the following standards are picked up and explained: (1) standards for requiring immediate evacuation and temporary relocation, (2) standards for requiring body surface screening and decontamination, and (3) standards for requiring the intake restriction of food and drink. (A.O.)

  14. Level-2 PSA for the prototype fast breeder reactor MONJU applied to the accident management review

    An accident management guideline (AMG) of the prototype fast breeder reactor MONJU has been presented to Nuclear and Industry Safety Agency (NISA) of METI by Japan Atomic Energy Agency (JAEA) with an evaluation result of an effectiveness of the AMG by employing Level-1 and Level-2 PSAs. Japan Nuclear Energy Safety Organization (JNES) evaluated the three events - PLOHS, LORL and ATWS events - and scrutinized the results of the Level-2 PSA carried out by JAEA from the view point of an accident management (AM) review. Regarding ATWS events, we have carried out a qualitative evaluation of the results of JAEA's evaluation and carried out a quantitative evaluation of the containment failure frequency (CFF) in relation to Protected-Loss-of-Heat-Sink (PLOHS) and Loss-of-Reactor-Level (LORL) events. Evaluation of the containment failure probability CFF has been conducted based on the results of the Level-1 PSA by employing the code system developed by JNES. We conducted a close examination of the procedure that JAEA followed to evaluate CFFs in PLOHS and LORL events. It was confirmed that JAEA's Level-2 PSA quantified the phenomenal event trees was expanded in the three processes - the plant response process, the core damage process and the containment vessel response process - based on various analytical and experimental evidence and otherwise followed much the same basic evaluation procedures employed by JNES. As for PLOHS and LORL, quantitative evaluation of CFF was conducted according to the following procedures: Development of an event flow diagram, Development of a phenomenal event tree, Quantification of the phenomenal event tree, Evaluation of containment failure frequencies, and Evaluation of the effectiveness of the AM measures. In the evaluation of the PLOHS and LORL events, the following analytical codes were used; Plant dynamic characteristic analytical code (NALAP-II), Nuclear characteristics analytical system (ARCADIAN-FBR/MVP), Nuclear dynamics analysis code

  15. Analytical evaluation of dose measurement of critical accident at SILENE (Contract research)

    Nakamura, T; Tonoike, K

    2003-01-01

    Institute for Radioprotection and Nuclear Safety (IRSN) and the OECD Nuclear Energy Agency (NEA) jointly organized SILENE Accident Dosimetry Intercomparison Exercise to intercompare the dose measurement systems of participating countries. Each participating country carried out dose measurements in the same irradiation field, and the measurement results were mutually compared. The participated in the exercise to measure the doses of gamma rays and neutron from SILENE by using thermoluminescence dosimeters (TLD's) and an alanine dosimeter. In this examination, the derived evaluation formulae for obtaining a tissue-absorbed dose from measured value (ambient dose equivalent) of TLD for neutron. We reported the tissue-absorbed dose computed using this evaluation formula to OECD/NEA. TLD's for neutron were irradiated in the TRACY facility to verify the evaluation formulae. The results of TLD's were compared with the calculations of MCNP and measurements with alanine dose meter. We found that the ratio of the dose b...

  16. Sustainable integration of EU research in severe accident phenomenology and management (SARNET2 project)

    In order to optimise the use of the available means and to constitute sustainable research groups in the European Union, the Severe Accident Research NETwork of Excellence (SARNET) has gathered 51 organisations representing most of the actors involved in Severe Accident (SA) research in Europe plus Canada. This project was co-funded by the European Commission (EC) under the 6th Euratom Framework Programme. Its objective was to resolve the most important pending issues for enhancing, in regard of SA, the safety of existing and future Nuclear Power Plants (NPPs). SARNET tackled the fragmentation that existed between the national R and D programmes, in defining common research programmes and developing common computer codes for safety assessment. The Joint Programme of Activities consisted in: (i) Implementing an advanced communication tool for accessing all project information, fostering exchange of information, and managing documents; (ii) Harmonizing and re-orienting the research programmes, and defining new ones; (iii) Analyzing the experimental results provided by research programmes in order to elaborate a common understanding of relevant phenomena; (iv) Developing the ASTEC code (integral computer code used to predict the NPP behaviour during a postulated SA) by integrating the knowledge produced within SARNET; (v) Developing Scientific Databases, in which the results of research experimental programmes are stored in a common format; (vi) Developing a common methodology for Probabilistic Safety Assessment of NPPs; (vii) Developing short courses and writing a text book on Severe Accidents for students and researchers; (viii) Promoting personnel mobility amongst various European organizations. This paper presents the major achievements after four and a half years of operation of the network, in terms of knowledge gained, of improvements of the ASTEC reference code, of dissemination of results and of integration of the research programmes conducted by the various

  17. Safety Implementation of Hydrogen Igniters and Recombiners for Nuclear Power Plant Severe Accident Management

    XIAO Jianjun; ZHOU Zhiwei; JING Xingqing

    2006-01-01

    Hydrogen combustion in a nuclear power plant containment building may threaten the integrity of the containment. Hydrogen recombiners and igniters are two methods to reduce hydrogen levels in containment buildings during severe accidents. The purpose of this paper is to evaluate the safety implementation of hydrogen igniters and recombiners. This paper analyzes the risk of deliberate hydrogen ignition and investigates three mitigation measures using igniters only, hydrogen recombiners only or a combination of recombiners and igniters. The results indicate that steam can effectively control the hydrogen flame acceleration and the deflagration-to-detonation transition.

  18. Review of aviation safety measures which have application to aviation accident prevention.

    Doughtery, J D

    1975-01-01

    Introduction of certain human-factors techniques has been followed by market reduction in military and airline accident rates. In this study, these safety measures are analyzed to determine the value of their application to general aviation activity. Some techniques are already in use. They are: 1. medical evaluation of iarcrews; 2. aeronautical innovations which tailor the machine to the man; 3. imporvement of precision navigational air traffic control and flight procedures; 4. standardization of flight training and flight procedures. A remaining field of interest, and one which appears to be underused, is that of supervision. After ending his association with the flight instructor, the general aviation pilot is essentially unsupervised. Accident data gathered over several years show that with increases in the proportion of pilots who have not maintained an association with a flight instructor, the general aviation fatal accident rate is increased. Current regulations, which require revalidation of airman's certificates, provide a method by which this association can be maintained. The flight instructor, or some similar aviation professional, can maintain an element of supervision with otherwise independent general aviation pilots. Data from previous years supports the hypothesis that such a program would make a substantial improvement in general aviation safety. PMID:1115703

  19. Measurement of neutron dose under criticality accident conditions at TRACY using ebonites

    Neutron doses under criticality accident conditions at TRACY were measured using ebonites, which are hard rubber containing sulfur. Ebonites can be easily available and are inexpensive since they are generally used as an insulator. To evaluate a neutron dose, beta rays emitted from 32P induced by 32S(n, p) reaction in an ebonite disc are measured with Geiger-Mueller (GM) counter. Then, a calibration factor (Gy/cpm), which is pre-determined using a 252Cf source, is applied to the count rates to obtain neutron doses. Factors to correct for the difference between responses of 32S(n, p) induced in an ebonite to the spontaneous fission spectrum of 252Cf calibration source and to spectra of TRACY were calculated using MCNP5, and applied to the doses. In the experiments, ebonites placed in free air and on phantom were exposed by TRACY with and without the water reflector. Neutron doses measured with ebonites in TRACY without a reflector were evaluated with an uncertainty of less than about 40%, and were consistently overestimated. On the other hand, average of neutron doses measured with ebonites in TRACY with the water reflector were accurate; however, the disparsion of neutron dose per integrated power of TRACY was large. By these measurements it was found that ebonites can be used as a neutron dosimeter for criticality accidents. (author)

  20. Impact of the Fukushima nuclear accident on background radiation doses measured by control dosimeters in Japan.

    Romanyukha, Alexander; King, David L; Kennemur, Lisa K

    2012-05-01

    After the 9.0 magnitude earthquake and subsequent massive tsunami on 11 March 2011 in Japan, several reactors at the Fukushima Daiichi Nuclear Power Plant suffered severe damage. There was immediate participation of U.S. Navy vessels and other United States Department of Defense (DoD) teams that were already in the area at the time of the disaster or arrived shortly thereafter. The correct determination of occupational dose equivalent requires estimation of the background dose component measured by control dosimeters, which is subsequently subtracted from the total dose equivalent measured by personal dosimeters. The purpose of the control dosimeters is to determine the amount of radiation dose equivalent that has accumulated on the dosimeter from background or other non-occupational sources while they are in transit or being stored. Given the release of radioactive material and potential exposure to radiation from the Fukushima Daiichi Nuclear Power Plant and the process by which the U.S. Navy calculates occupational exposure to ionizing radiation, analysis of pre- and post-event control dosimeters is warranted. Several hundred historical dose records from the Naval Dosimetry Center (NDC) database were analyzed and compared with the post-accident dose equivalent data of control dosimeters. As result, it was shown that the dose contribution of the radiation and released radiological materials from the Fukushima nuclear accident to background radiation doses is less than 0.375 μSv d for shallow and deep photon dose equivalent. There is no measurable effect on neutron background exposure. The latter has at least two important conclusions. First, the NDC can use doses measured by control dosimeters at issuing sites in Japan for determination of personnel dose equivalents; second, the dose data from control dosimeters prior to and after the Fukushima accident may be used to assist in dose reconstruction of non-radiological (non-badged) personnel at these locations

  1. The prediction of the LWR plant accident based on the measured plant data

    In case of accident affecting a nuclear reactor, it is essential to anticipate the possible development of the situation to efficiently succeed in emergency response actions, i.e. firstly to be early warned, to get sufficient information on the plant: and as far as possible. The ASTRID (Assessment of Source Term for Emergency Response based on Installation Data) project consists in developing a methodology: of expertise to; structure the work of technical teams and to facilitate cross competence communications among EP players and a qualified computer tool that could be commonly used by the European countries to reliably predict source term in case of an accident in a light water reactor, using the information available on the plant. In many accident conditions the team of analysts may be located far away from the plant experiencing the accident and their decision making is based on the on-line plant data transmitted into the crisis centre in an interval of 30 - 600 seconds. The plant condition has to be diagnosed based on this information, In the ASTRID project the plant status diagnostics has been studied for the European reactor types including BWR, PWR and VVER plants. The directly measured plant data may be used for estimations of the break size from the primary system and its locations. The break size prediction may be based on the pressurizer level, reactor vessel level, primary pressure and steam generator level in the case of the steam generator tube rupture. In the ASTRID project the break predictions concept was developed and its validity for different plant types and is presented in the paper, when the plant data has been created with the plant specific thermohydraulic simulation model. The tracking simulator attempts to follow the plant behavior on-line based on the measured plant data for the main process parameters and most important boundary conditions. When the plant state tracking fails, the plant may be experiencing an accident, and the tracking

  2. Fission product aerosol removal test by containment spray under accident management conditions (2)

    In order to demonstrate effective fission product (FP) aerosol removal and pressure suppression effects by containment spray under Japanese accident management (AM) conditions, system integral tests simulating typical BWR accident sequences have been carried out using a full-height simulation containment vessel test facility. In case of 10% reduction spray flow rate comparing with a reference test case, aerosol concentration in the entire drywell (D/W) decreased rapidly about 1/10 of initial concentration within 30 min after the spray initiation and remained low through 12 hours test period similar to the reference test case. The maximum pressure was slightly higher in this case. Both the existence of non-condensable gas and the location of aerosol injection did not affect both pressure suppression effect and aerosol removal effect. In case of aerosol injection into the middle D/W, aerosol concentration in the upper D/W was relatively high, but the concentrations in the middle and the lower D/W were extremely low. The degradation of FP removal due to the existence of non-condensable gas was supplemented by FP removal of pool scrubbing in suppression chamber. After the modification of FP removal model in MELCOR, calculated time dependency of CsI aerosol concentration and pressure in the D/W agreed well with the test data. (author)

  3. Quantification of a decision-making failure probability of the accident management using cognitive analysis model

    In a nuclear power plant, much knowledge on severe accidents has been acquired through PSA, and accident management (AM) guidelines are prepared by incorporating that knowledge. In PSA, it is necessary to evaluate the effectiveness of AM using the decision-making failure probability (DFP) of an emergency organization, operation failure probability of operators, success criteria of AM and reliability of AM equipment. However, to date there has been no suitable quantification method for PSA to obtain DFP. In this study, we developed a new method for DFP quantification of an emergency organization using a cognitive analysis model, and tried to apply it to S2DC and TMLF sequence of a typical plant. As a result: (1) The methods enabled to DFP quantification appropriate to level 1.5PSA by choosing the suitable value of a basic failure probability and an error factor. (2) The DFPs of six AMs appeared to be in the range of 0.23 to 0.41 (screening method) and in the range of 0.10 to 0.19 (detailed method), and the DFP decreased about 50% as a result of sensitivity analysis of the conservative assumption. (3) The screening method was more conservative than the detailed method, and it was shown to satisfy the screening performance required by PSA. (author)

  4. Managing Errors to Reduce Accidents in High Consequence Networked Information Systems

    Ganter, J.H.

    1999-02-01

    Computers have always helped to amplify and propagate errors made by people. The emergence of Networked Information Systems (NISs), which allow people and systems to quickly interact worldwide, has made understanding and minimizing human error more critical. This paper applies concepts from system safety to analyze how hazards (from hackers to power disruptions) penetrate NIS defenses (e.g., firewalls and operating systems) to cause accidents. Such events usually result from both active, easily identified failures and more subtle latent conditions that have resided in the system for long periods. Both active failures and latent conditions result from human errors. We classify these into several types (slips, lapses, mistakes, etc.) and provide NIS examples of how they occur. Next we examine error minimization throughout the NIS lifecycle, from design through operation to reengineering. At each stage, steps can be taken to minimize the occurrence and effects of human errors. These include defensive design philosophies, architectural patterns to guide developers, and collaborative design that incorporates operational experiences and surprises into design efforts. We conclude by looking at three aspects of NISs that will cause continuing challenges in error and accident management: immaturity of the industry, limited risk perception, and resource tradeoffs.

  5. Emergency preparedness: medical management of nuclear accidents involving large groups of victims

    The treatment of overexposed individuals implies hospitalisation in a specialized unit applying hematological intense care. If the accident results in a small number of casualties, the medical management does not raise major problems in most of the countries, where specialized units exist, as roughly 7% of the beds are available at any time. But an accident which would involved tens or hundreds of people raises much more problems for hospitalization. Such problems are also completely different and will involve steps in the medical handling, mainly triage, (combined injuries), determination of whole body dose levels, transient hospitalization. In this case, preplanning is necessary, adapted to the system of medical care in case of a catastrophic event in the given Country, with the main basic principles : emergency concerns essentially the classical injuries (burns and trauma) - and contamination problems in some cases - treatment of radiation syndrome is not an emergency during the first days but some essential actions have to be taken such as early blood sampling for biological dosimetry and for HLa typing

  6. An Examination of Commercial Aviation Accidents and Incidents Related to Integrated Vehicle Health Management

    Reveley, Mary S.; Briggs, Jeffrey L.; Thomas, Megan A.; Evans, Joni K.; Jones, Sharon M.

    2011-01-01

    The Integrated Vehicle Health Management (IVHM) Project is one of the four projects within the National Aeronautics and Space Administration's (NASA) Aviation Safety Program (AvSafe). The IVHM Project conducts research to develop validated tools and technologies for automated detection, diagnosis, and prognosis that enable mitigation of adverse events during flight. Adverse events include those that arise from system, subsystem, or component failure, faults, and malfunctions due to damage, degradation, or environmental hazards that occur during flight. Determining the causal factors and adverse events related to IVHM technologies will help in the formulation of research requirements and establish a list of example adverse conditions against which IVHM technologies can be evaluated. This paper documents the results of an examination of the most recent statistical/prognostic accident and incident data that is available from the Aviation Safety Information Analysis and Sharing (ASIAS) System to determine the causal factors of system/component failures and/or malfunctions in U.S. commercial aviation accidents and incidents.

  7. Physical dose reconstruction in case of radiological accidents: an asset for the victims' management

    In most cases of radiological accidents caused by an external source, the irradiation is heterogeneous, even for a whole body irradiation. Therefore, more than a whole body dose, estimating the dose distribution in the victim's organism is essential to assess biological damages. This dose distribution can be obtained by physical dosimetric reconstruction methods. The laboratory has developed several techniques based on experimental and numerical dose reconstruction and retrospective dosimetry by ESR in order to assess as accurately as possible and as quickly as possible the dose received and especially its distribution throughout the organism so that the physicians may fine tune their diagnosis and prescribe the most suitable treatment. These last years, these techniques were applied several times and each time the results obtained proved to be essential for the physicians in charge of the victims in order to define the therapeutic strategy. This article proposes a review of the physical dose reconstructions performed in the laboratory for recent radiological accidents focusing on the complementarity of the methods and the gain for the victims' management. (author)

  8. Lessons learned from the CEOG generic accident management guidelines confirmation (validation) exercise

    In July 1995, the CE Owner's Group completed and issued Revision 0 of the Generic Accident Management Guidelines (AMG's) to the owners group task participants. This guidance provides a structured mechanism for the plant staff at CE utilities to respond to accidents that beyond the plant design basis and, possibly, the Emergency Operating Procedures. Prior to final issue of the generic AMGs, the CEOG conducted an AMG Confirmation Exercise to establish the ability of the AMGs to fulfill this important role. The specific objectives of the AMG Confirmation Exercise were to (1) clarify the interactions and transitions between the AMG/Technical Support Center (TSC) and the EOPS/Operations Personnel (2) validate the adequacy of the AMG data collection and plant condition diagnostic evaluation process and (3) assess the feasibility of the mechanical material and recommendations contained in the AMG's. The purpose of paper is to provide a detailed description of the AMG Confirmation Exercise as well as important lessons learned during the planning and implementation of the exercise. In addition, a discussion will be presented pertaining to the relationship between the AMG's (incumbent to the Technical Support Center) and the plants Emergency Operating Procedures (incumbent to the Control Room Operations Staff)

  9. The Effect of Containment Filtered Venting System on the Severe Accident Management Strategies of the CANDU6 Plant

    In March, 2011, Fukushima daichi nuclear power plants experienced a long term station blackout and severe core damages and released a large amount of radioactive materials outside of the plants. After this accident Nuclear Safety and Security Commission (NSSC) decided to install a filtered containment venting system (CFVS) at all the operating nuclear power plants in Korean. To comply with NSSC's request, Wolsong Unit 1 has installed a CFVS. Current severe accident management guidance, which does not consider a CFVS has 6 severe accident management strategies for CANDU6 plant. These strategies are inject in to the primary heat transport system (PHTS), inject in to the calandria, inject into the calandria vault, reduce fission product releases, control containment conditions, reduce containment hydrogen. The CFVS is designed to open and to close isolation valves by an operator. An operator opens the CFVS isolation valve when the containment pressure exceeds the design pressure (124 kPa(g)) and closes isolation valves when the containment pressure decreases below 50 kPa(g). The operation of the CFVS not only influences the current strategies (adds a means of controlling containment conditions) but also requires the new strategies. This paper discusses the necessity of the new strategies, such as the prevention of containment vacuum and the injection into the containment. The necessity of the additional severe accident management strategies for CANDU6 plants which installed a CFVS is evaluated. The operation of a CFVS affects the water inventory in the basement also, but not significantly. The SBO accident requires the water injection into the containment at least 4 days after an accident initiation if a passive spray system fails. If a spray system operates, then the injection into the containment is required more than 10 days after an accident initiation even though a CFVS operates

  10. Environmental radioactivity measurements in north-western Greece following the Fukushima nuclear accident

    The impact of the Fukushima nuclear accident in north–western Greece was assessed through an environmental monitoring programme activated by the Nuclear Physics Laboratory of the University of Ioannina. Measurements of 131I were carried out in atmospheric particulate, ovine milk and grass samples. In daily aerosol samplings, radioiodine was first detected on March 25-26, 2011 and reached maximum levels, up to 294 μBq m-3, between April 2 and April 4, 2011. In ovine milk samples, 131I concentrations ranged from 2.0 to 2.7 Bq L-1 between April 2 and April 6, 2011, while an average activity of 2.7 Bq kg-1 was measured in grass samples on April 4, 2011. The 134,137Cs isotopes were below detection limits in all samples and could only be determined in the air, by analysis of multiple daily filters. A maximum average activity concentration of 137Cs amounting to 24 μBq m-3 was measured during the period from April 5 to April 9, 2011, with the 134Cs/137Cs activity ratio being close to unity. Activity concentrations were consistent with measurements conducted in other parts of the country and were well below those reported in May 1986 after the Chernobyl accident. The committed effective dose to the whole body and to the thyroid gland from inhalation of 131I was estimated for the adult and infant population and was found to be of no concern for the public health. (author)

  11. Hydrogen combustion management during a severe accident at the plant with the ice condenser type containment

    Watada, M.; Furuta, T.; Ohtani, M. [Kansai Electric Power Co., Inc., Osaka (Japan); Ogino, M. [Mitsubishi Heavy Industry Co., Ltd., Yokohama (Japan)

    1997-03-01

    In Japan, hydrogen mitigation measures inside the containment vessel during a severe accident are taken against the plant with the ice condenser type containment. Ohi Power Station Unit No.1 and 2, which Kansai Electric Power Co.,Inc. owns, are the only plants of this kind in Japan. Kansai has investigated the hydrogen mitigation measures in collaboration with Mitsubishi Heavy Industry Co.,Ltd. As a result of extensive experiments and analyses, the glow plug type igniter was selected as a hydrogen mitigation device. Environmental conditions were investigated for the purpose of selection of the device. To decide the location of installation, Kansai performed analysis of mixing behavior of hydrogen focusing on the results of small scale combustion testing conducted by Nupec (Nuclear Power Engineering Corporation). This paper will introduce the detailed results of Kansai's investigation of hydrogen mitigation measures for Ohi Power Station Unit No.1 and 2. (author)

  12. Hydrogen combustion management during a severe accident at the plant with the ice condenser type containment

    In Japan, hydrogen mitigation measures inside the containment vessel during a severe accident are taken against the plant with the ice condenser type containment. Ohi Power Station Unit No.1 and 2, which Kansai Electric Power Co.,Inc. owns, are the only plants of this kind in Japan. Kansai has investigated the hydrogen mitigation measures in collaboration with Mitsubishi Heavy Industry Co.,Ltd. As a result of extensive experiments and analyses, the glow plug type igniter was selected as a hydrogen mitigation device. Environmental conditions were investigated for the purpose of selection of the device. To decide the location of installation, Kansai performed analysis of mixing behavior of hydrogen focusing on the results of small scale combustion testing conducted by Nupec (Nuclear Power Engineering Corporation). This paper will introduce the detailed results of Kansai's investigation of hydrogen mitigation measures for Ohi Power Station Unit No.1 and 2. (author)

  13. Applications of nano-fluids to enhance LWR accidents management in in-vessel retention and emergency core cooling systems

    Water-based nano-fluid, colloidal dispersions of nano-particles in water; have been shown experimentally to increase the critical heat flux and surface wettability at very low concentrations. The use of nano-fluids to enhance accidents management would allow either to increase the safe margins in case of severe accidents or to upgrade the power of an existing power plant with constant margins. Building on the initial work, computational fluid dynamics simulations of the nano-fluid injection system have been performed to evaluate the feasibility of a nano-fluid injection system for in-vessel retention application. A preliminary assessment was also conducted on the emergency core cooling system of the European Pressurized Reactor (EPR) to implement a nano-fluid injection system for improving the management of loss of coolant accidents. Several design options were compared/or their respective merits and disadvantages based on criteria including time to injection, safety impact, and materials compatibility. (authors)

  14. Water level measurement system in reactor pressure vessel of BWR and hydrogen concentration monitoring system for severe accident

    TEPCO's Fukushima Daiichi Nuclear Power Station Accident caused severe accident to lose functions of many instrumentation systems. As a result, many important plant parameters couldn't be monitored. In order to monitor plant parameters in the case of severe accident, new instrumentation systems available in the severe conditions are being developed. Water level in reactor pressure vessel and hydrogen concentration in primary containment vessel are one of the most important parameters. Performance test results about water level measurement sensor and hydrogen sensor in severe environmental conditions are described. (author)

  15. Final report on Risoe measuring program in connection with Chernobyl accident

    The present report deals with the measurements of Chernobyl debris carried out in Denmark, the Faroe Islands and Greenland in the perioed May-Sept. 1986. The results are presented in details in appendix II, but summarized in tables and figures in the main report, which is in Danish. Appendix I is the samples programme, also in Danish. It is concluded that the dose equivalent commitment to an adult Dane from consumption of foodstuffs in the first year after the accident (May 1986-April 1987) is 17 μ Sv, corresponding to approximately 1% of a years background radiation. (author)

  16. Radionuclides from the Fukushima accident in the air over Lithuania: measurement and modelling approaches

    Analyses of 131I, 137Cs and 134Cs in airborne aerosols were carried out in daily samples in Vilnius, Lithuania after the Fukushima accident during the period of March–April, 2011. The activity concentrations of 131I and 137Cs ranged from 12 μBq/m3 and 1.4 μBq/m3 to 3700 μBq/m3 and 1040 μBq/m3, respectively. The activity concentration of 239,240Pu in one aerosol sample collected from 23 March to 15 April, 2011 was found to be 44.5 nBq/m3. The two maxima found in radionuclide concentrations were related to complicated long-range air mass transport from Japan across the Pacific, the North America and the Atlantic Ocean to Central Europe as indicated by modelling. HYSPLIT backward trajectories and meteorological data were applied for interpretation of activity variations of measured radionuclides observed at the site of investigation. 7Be and 212Pb activity concentrations and their ratios were used as tracers of vertical transport of air masses. Fukushima data were compared with the data obtained during the Chernobyl accident and in the post Chernobyl period. The activity concentrations of 131I and 137Cs were found to be by 4 orders of magnitude lower as compared to the Chernobyl accident. The activity ratio of 134Cs/137Cs was around 1 with small variations only. The activity ratio of 238Pu/239,240Pu in the aerosol sample was 1.2, indicating a presence of the spent fuel of different origin than that of the Chernobyl accident. - Highlights: ► Two observed maxima in radionuclide concentrations were related to air mass transport. ► HYSPLIT backward trajectories were applied for data interpretation. ► 7Be and 212Pb were used to study a vertical transport of air masses. ► The 134Cs/137Cs activity ratio was around 1. ► 238Pu/239,240Pu ratio was different from global fallout and Chernobyl accident.

  17. Investigation of alternative solutions for severe accident management in future reactors

    Since 1991, the CEA/DRN 'Innovations-Future Reactors' Program (IFRP) has been developed in order to elaborate, to evaluate and validate technical options which can be of interest for future reactors. The main objectives of this program are: to improve both the safety and cost of future nuclear power plants, to optimize the fuel cycle and the management of nuclear materials. The present paper is focused on the third R and D theme, i.e., on the 'Innovation-Severe Accident Research Program' (ISARP). This specific CEA long-term program is developed in addition to shorter-term studies conducted in collaboration with the CEA partners (EDF and FRAMATOME), more particularly, for the future European Pressurized Water Reactor (EPR). (J.P.N.)

  18. Human factors issues in severe accident management: Training for decision-making under stress

    Training for operator and other technical positions in the commercial nuclear power industry traditionally has focused on mastery of the formal procedures used to control plant systems and processes. However, there is a growing awareness that the decision-making tasks required for selecting appropriate control actions, in addition to guidance from formal procedures, also involve cognitive activities commonly referred to as judgment or reasoning. A project was completed to address the nature of the cognitive skills that may be important to decision-making in the nuclear power plant environment, especially during severe accident management. The project identified a model of decision-making that could account for both rule-based and knowledge-based decision-making and used it to identify cognitive skills for both individuals and operational crews. This analysis was then used to identify existing training techniques for cognitive skills and the general characteristics of successful training techniques

  19. Nuclear plant analyser, a tool for evaluating actions of accident management

    The nuclear plant analyser currently developed by GRS and Siemens/KWU allows to use detailed models as well as fast and simple models. The thermofluid codes TRAC and ATHLET are used as process models. Both codes are capable of modelling the effects of accident management actions on the course of events on the 'prevention' level. An extension to the 'mitigation' level is possible by combination with the ATHLET-SA version. The analyser allows extensive simulation of thermohydraulic stand by systems and monitoring and safety systems of a plant in order to obtain a realistic verification of preventive actions and their effects by a description of plant behaviour as exact as possible. (orig./DG)

  20. Information and communication technologies, a tool for risk prevention and accident management on sea ice

    Elise Lépy

    2015-06-01

    Full Text Available Marine ice melting topic is a repetitive phenomenon in alarmist speeches on climate change. The present positive evolution of air temperatures has in all probability many impacts on the environment and more or less directly on societies. Face to the temperature elevation, the ice pack is undergone to an important temporal variability of ice growth and melting. Human populations can be exposed to meteorological and ice hazards engendering a societal risk. The purpose of this paper is to better understand how ICT get integrated into the risk question through the example of the Bay of Bothnia in the northern extremity of the Baltic Sea. The study deals with the way that Finnish society, advanced in the ICT field, faces to new technology use in risk prevention and accident management on sea ice.

  1. Body surface monitor for measuring radioactive contamination of the general population after a nuclear accident

    A new body surface monitor for monitoring the surface radioactive contamination of the general population living and working around the site in the early stages of a nuclear accident has been designed. The body surface monitors will be installed in a medium-sized bus with a thyroid counter and moved to the place where measurement is required. The different characteristics needed for the body surface monitor to measure the general population from those of monitors used in nuclear power stations are discussed. The detection sensitivity of the plastic scintillator was measured under various geometric conditions and the Minimum Detectable Activity (MDA) was found to be lower than 1 Bq/cm2 in a 10-second count time. Two body surface monitors can measure 2,880 persons in eight hours. (author)

  2. Quick evaluation of the neutron dose following a criticality accident by measurement of sodium 24 activity

    In order to quickly sort out the irradiated individuals following a criticality accident, the neutron dose can be evaluated quickly by measuring the sodium-24 activity induced in the human body. The report supplies the information necessary for this evaluation from the response of various detectors of current use in radiation protection. The first part describes the method of evaluation of sodium-24 activity (A) given by the reading (M) of each instrument. The second part describes the method of kerma evaluation from the measured sodium-24 activity. The third part is an experimental application of the method of kerma evaluation from the sodium-24 activity measured in a phantom irradiated in the SILENE reactor flux. The results given by radiation protection instruments are in good agreement with the calculated values for a front exposure and demonstrate the usefulness of measuring the induced sodium-24 activity by radiation protection instruments of current use

  3. 'SAMIME' - An EC Concerted Action on Severe Accident Management in Europe

    The EC Concerted Action 'Severe Accident Management Implementation and Expertise' - with acronym 'SAMIME' - was initiated in 1998 with the following main objectives: 1. to determine the status and the extent of the severe accident management (SAM) development and implementation in partner countries/regions/utilities; 2. to determine the extent to which the development is institutional and develop a consensus opinion among partners as to which elements are needed or useful in this respect; 3. to review the tools which are available to support SAM development and implementation and determine the extent to which these tools have a value during a SAM event; 4. to determine in which areas further research may be beneficial for SAM development. Thirteen organisations from all countries with NPPs in the European Union and neighbouring countries, also in Eastern Europe, agreed to participate in this effort. Partners were a balanced group of utilities, vendors and regulators. Two organisations became partners later on. One utility could not be a formal partner, but supported the project with its insights. Two US Owners Groups and the OECD Halden Reactor Programme provided input on specific topics. The Concerted Action was executed in a number of workshops which were preceded by extensive questionnaires to facilitate discussions. The project provided a good overview of what SAM guidance (SAMG) was in place in the various countries and in which direction the development went. After extensive discussions, a consensus opinion was achieved on what partners felt the elements of an adequate SAMG approach should be. Although differences may exist in the way SAMG can be implemented, a common base line could be defined - which was more than a common denominator, as it sometimes exceeded what was in place. Finally, the areas were outlined where additional research work still could enhance the SAMG, taking note of the fact that understanding of severe accidents would never be complete

  4. State of safety improvement measures at European Nuclear Power Plants after the Fukushima accident

    In response to the Fukushima Dai-ichi accident, European Council required comprehensive assessment of safety margins (called stress test) at European Nuclear Power Plants (NPPs). The stress test consisted of the reviews on natural events such as earthquake, floods and extreme weather, loss of safety functions, and severe accident management for each European NPP. Its procedure had 3 steps: (1) nuclear operator performed the stress test and prepared proposals for safety improvements, (2) national regulator performed independent review of the stress test and prepared national report, and (3) the report submitted by national regulator was subjected to peer-review by experts at a European level. The article described outline of the stress tests, their peer-review results verified at the European level and their national action plans. As for state of main countries, activities in France, Germany and Switzerland were introduced. Establishments of Nuclear Rapid Response Force (FARN, Force d'Action Rapide Nuclaire) in France and External Emergency Storage Center in Switzerland were original approaches of their emergency responses. (T. Tanaka)

  5. How to effectively manage crisis situation in the early phase of a radiological accident?

    Full text: In the early phase of accident reasonable decisions should be undertaken in a short time, where often there is still an uncertainty in assessment of radiological situation. The following key factors are of important meaning in a process of working out an optimized decision: (a) accurate information on accident, (b) proper assessment of current status and prognosis of development of radiological situation, (c) proper information on availability of means needed for emergency action like rescue teams, technical and medical equipment, means of transport etc., (d) reliable and fast communication system between decision makers, persons responsible for management of crisis situation, rescue teams and people in affected areas. All these elements can be supported by more or less computerized systems. The first two (a) and (b), depend, in general, on radiological monitoring and decision support systems, using pre-defined scenarios and sometimes sophisticated methods for assessment of radiological situation. The element (c) is very often supported by GIS-like systems available at crisis centres. The last issue however is, to much extent, a question of proper organization of decision making process and management during emergency action. lt seems also that often there are some kind of gaps between items (a)e) and (c)/(d) or maybe particularly between (d) and other factors mentioned above. Hence, there is still a need for more integrated approach. lt should be also mentioned that the last element (d) is usually the weakest point in the whole system. This is often caused by not clear organization and division of responsibilities between persons engaged in the decision making process and management. The problem of communication has also some technical aspects. This can be solved by using more advanced techniques like satellite technologies and centralized computer communication systems of new generation, which allows for fast and reliable sending and receiving messages

  6. Development of a taxonomy of performance influencing factors for human reliability assessment of accident management tasks and its application

    In this study, a new PIF taxonomy for HRA of the tasks during emergency operation and accident management situations. We collected the existing PIF taxonomies as many as possible. Then, we analyzed the trend in the selection of PIFs, the frequency of use between PIFs in HRA methods, and the level of definition of PIFs, in order to reflect these characteristics into the development of a new PIF taxonomy. Next, we analyzed the principal task context during accident management to draw the context specific PIFs. Afterwards, we established several criteria for the selection of the appropriate PIFs for HRA under emergency operation and accident management situations. Finally, the final PIF taxonomy containing the subitems for assessing each PIF was constructed based on the results of the previous steps and the selection criteria. The final result of this study is the new PIF taxonomy for HRA of the tasks during emergency operation and accident management situations. The selected 11 PIFs in the study are as follows: training and experience, availability and quality of information, status and trend of critical parameters, status of safety system/component, time pressure, working environment features, team cooperation and communication, plant policy and safety culture. (author). 35 refs., 23 tabs

  7. Lessons learned from post-accident management at Chernobyl: the P.a.r.e.x. project

    Return of experience on Chernobyl post-accident management: the PAREX study Belarus is the country the most affected by the Chernobyl fallouts and is among the most significant experiences in the nuclear post-accident field. Despite specificities inherent to the political and social situation in Belarus, the experience of post-accidental management in this country holds a wealth of lessons in the perspective of preparation to a post-accidental situation in the French and European context. Through the PAREX project (2005-2006), the French Nuclear Safety Authority analysed the return of experience of Chernobyl post-accident management from 1986 to 2005 in order to draw its lessons in the perspective of a preparation policy. The study was led by a group of experts and involved the participation of a pluralistic group of about thirty participants (public authorities, local governments, NGOs, experts, operators). PAREX highlighted the complexity of a situation of long-lasting radioactive contamination (diversity of stakeholders and of dimensions at stake: health, environment, economy, society...). Beyond traditional public crisis management tools and frameworks, post-accident strategies also involves in the longer term a territorial and social response, which relies on local capacities of initiative. Preparation to such process requires experimenting new modes of operation that allow a diversity of local actors to take part to the response to a situation of contamination and to the surveillance system, with the support of public authorities. The conclusions of PAREX include a set of recommendations in this perspective. (authors)

  8. Development of a taxonomy of performance influencing factors for human reliability assessment of accident management tasks and its application

    Kim, Jae Whan; Jung, Won Dae; Kang, Dae Il; Ha, Jae Joo

    1999-06-01

    In this study, a new PIF taxonomy for HRA of the tasks during emergency operation and accident management situations. We collected the existing PIF taxonomies as many as possible. Then, we analyzed the trend in the selection of PIFs, the frequency of use between PIFs in HRA methods, and the level of definition of PIFs, in order to reflect these characteristics into the development of a new PIF taxonomy. Next, we analyzed the principal task context during accident management to draw the context specific PIFs. Afterwards, we established several criteria for the selection of the appropriate PIFs for HRA under emergency operation and accident management situations. Finally, the final PIF taxonomy containing the subitems for assessing each PIF was constructed based on the results of the previous steps and the selection criteria. The final result ofthis study is the new PIF taxonomy for HRA of the tasks during emergency operation and accident management situations. The selected 11 PIFs in the study are as follows: training and experience, availability and quality of information, status and trend of critical parameters, status of safety system/component, time pressure, working environment features, team cooperation and communication, plant policy and safety culture. (author). 35 refs., 23 tabs.

  9. Public health response to the nuclear accident

    The Act on Special Measures Concerning Nuclear Emergency Preparedness was established in 2000 as a specific act within the broader Disaster Control Measures and Reactor Regulation Act which was written in response to the JCO Criticality Accident of 1999. However, this regulatory system did not address all aspects of the Fukushima Daiichi Nuclear Power Plant Accident. This was especially evident with public health issues. For example, radioactive screening, prophylactic use of potassium iodide, support for vulnerable people, and management of contaminated dead bodies were all requested immediately after the occurrence of the nuclear power plant accident but were not included in these regulatory acts. Recently, the regulatory system for nuclear accidents has been revised in response to this reactor accident. Herein we review the revised plan for nuclear reactor accidents in the context of public health. (author)

  10. Measurement of airborne fission products in Chapel Hill, NC, USA from the Fukushima I reactor accident

    MacMullin, S; Green, M P; Henning, R; Holmes, R; Vorren, K; Wilkerson, J F

    2011-01-01

    We present measurements of airborne fission products in Chapel Hill, NC, USA, from 62 days following the March 11, 2011, accident at the Fukushima I Nuclear Power Plant. Airborne particle samples were collected daily in air filters and radio-assayed with two high-purity germanium (HPGe) detectors. The fission products I-131 and Cs-137 were measured with maximum activities of 4.2 +/- 0.6 mBq/m^2 and 0.42 +/- 0.07 mBq/m^2 respectively. Additional activity from I-131, I-132, Cs-134, Cs-136, Cs-137 and Te-132 were measured in the same air filters using a low-background HPGe detector at the Kimballton Underground Research Facility (KURF).

  11. Consideration of severe accidents in design of advanced WWER reactors

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

  12. Compendium of the Environmental Measurements Laboratory's research projects related to the Chernobyl nuclear accident

    Following the accident at the Chernobyl nuclear reactor power station in the USSR on April 26, 1986, the Environmental Measurements Laboratory (EML) initiated a number of research projects as follows: (1) selected sites in both the Deposition and Surface Air networks were alerted and their sampling protocols adjusted to accommodate the anticipated arrival times and activity concentrations of the Chernobyl debris; (2) a number of cooperative programs involving field work, sampling, analysis and data interpretation were set up with institutions and scientists in other countries; (3) EML's Regional Baseline Station at Chester, NJ, as well as the roof of the Laboratory in New York City, provided bases for sampling and measurements to study the radionuclide concentrations, radiation levels, physical characteristics and potential biological implications of the Chernobyl fallout on the northeastern United States; and (4) the resulting fallout from the Chernobyl accident provided an 'experiment of opportunity' in that it enabled us to study fresh fission product deposition using collection systems resurrected from the 1950's and 1960's for comparison with current state-of-the-art methodology. The 13 reports of this volume have been entered separately into the data base

  13. Compendium of the Environmental Measurements Laboratory's research projects related to the Chernobyl nuclear accident

    Volchok, H L; Chieco, N [comps.

    1986-10-01

    Following the accident at the Chernobyl nuclear reactor power station in the USSR on April 26, 1986, the Environmental Measurements Laboratory (EML) initiated a number of research projects as follows: (1) selected sites in both the Deposition and Surface Air networks were alerted and their sampling protocols adjusted to accommodate the anticipated arrival times and activity concentrations of the Chernobyl debris; (2) a number of cooperative programs involving field work, sampling, analysis and data interpretation were set up with institutions and scientists in other countries; (3) EML's Regional Baseline Station at Chester, NJ, as well as the roof of the Laboratory in New York City, provided bases for sampling and measurements to study the radionuclide concentrations, radiation levels, physical characteristics and potential biological implications of the Chernobyl fallout on the northeastern United States; and (4) the resulting fallout from the Chernobyl accident provided an 'experiment of opportunity' in that it enabled us to study fresh fission product deposition using collection systems resurrected from the 1950's and 1960's for comparison with current state-of-the-art methodology. The 13 reports of this volume have been entered separately into the data base.

  14. Agroindustrial production sphere - radiological consequences of the Chernobyl accident and the chief protective measures

    As a result of the Chernobyl accident, fallout of radionuclides has occurred on farm lands, and the contaminated production of the agroindustrial complex has become a source of additional irradiation of the population. The contribution of the irradiation associated with the consumption of locally produced food products was quite significant, and led to the implementation of protective measures in the agroindustrial production sphere. It should be noted that irradiation of people owing to the consumption of contaminated agricultural products is more easily regulated than external irradiation. For this reason, the decrease in the total dose load is largely determined by the possibilities of restricting the internal irradiation dose to the population from the consumption of food products. The paper discusses radiological conditions in the agroindustrial production sphere in the region of the accident; intake of radionuclides by agricultural plants through leaves; distribution and form of 137Cs in soils; uptake of radionuclides by plants from the soil, animal husbandry aspects of the migration of radionuclides and their biological action; and organizational measures of the USSR for mitigating the consequences

  15. Radioactivity measurements in Krakow surroundings in the aftermath of Chernobyl reactor accident

    A team from different laboratories of the Institute of Nuclear Physics was formed to set a crash program of measurement of water and food contamination after the Chernobyl reactor accident. The main contaminants in the first days were 131I and 132Te which were superseded later on by 104Ru, 137Cs and 134Cs. The highest value of contamination of surface waters by 131I was attained in the Vistula river on the 2-nd of May with 530 Bq/dm3. Also measurements of food contamination by 131I,134Cs, 137Cs and 137Te were carried out. The additional effective dose equivalent related to Chernobyl accident received by the population of Krakow region in May 1986 was estimated at 0.45 mSV (45 rem). Another rise of 134Cs + 137Cs content up to 46 Bq/dm3 in cows milk was observed during March and April 1987 and was probably explicable by the use of hay harvested in June 1986. (author)

  16. Radiocesium contamination in a submediterranean semi-natural ecosystem following the Chernobyl accident: Measurements and models

    Radiocesium dynamics in a Quercus conferta Kit ecosystem in Northern Greece have been extensively studied over the years 1993-1995. Radiocesium distribution in the different parts of the ecosystem was measured. A total 137Cs inventory of 243±66 MBq ha-1 due to the Chernobyl accident was measured in all parts of the ecosystem. Almost 90% of this inventory is still in the upper layers of the soil and the forest floor. In particular 13.4% is in the forest floor, 52.6% in the Ah horizon, and 23.4% in the upper 5 cm of the soil. Only 2.2% of this inventory is in the above ground biomass. The mean total 137Cs deposited on the forest floor from the above ground biomass is 0.18 MBq ha-1 y-1. Cesium leaching from the forest floor is negligible. The radiocesium distribution in soil is fixed and in equilibrium, at least since 1993. Most of radiocesium is not available for migration. Cesium migration in soil was modeled by (a) an open-quotes equivalent diffusionclose quotes model with different initial conditions and (b) a open-quotes compartmentclose quotes model derived from a diffusion-advection model. A compartment model for the contamination of living biomass is proposed. The total absorbed dose rate in air as well as the contribution due to 137Cs from the Chernobyl accident was determined inside the forest, by in-situ gamma spectrometry. 32 refs., 10 figs., 7 tabs

  17. The severe accident research program at KIT

    The understanding of the plant behaviour under beyond design basis accidents as well as the interaction of the operators with the plant is the most important prerequisite to develop proper strategies to both control the accident progression and to minimize the radiological risk that may derive from operating nuclear power plants. In view of the Fukushima accident, a review of many issues important to safety e.g. severe accident analysis methodologies and assumptions, emergency operational procedures, severe accident management procedures (SAM), decision lines of the emergency team, etc. is needed to draw conclusions in order to avoid a repetition of Fukushima-like accidents.In addition, situations like the ‘black control room’ need to be reconsidered and a re-evaluation of the necessary instrumentation for hypothetical severe accident situations is urgently needed. If the real plant state during core meltdown accidents is unknown, no effective measures can be initiated by the emergency team in order to assure the integrity of the safety barriers and hence the release of radioactive material to the environment. The work performed in this area is integrated in the European Networks such as SARNET (Severe Accident Research Network) for the severe accidents, and for emergency management in the NERIS-TP. In future all the activities will be included in the NUGENIA platform. A brief overview of the KIT activities together with the experimental test facilities is given

  18. Accident management following loss-of-coolant accidents during cooldown in a Westinghouse two-loop PWR

    Operation of pressurised water reactors involves shutdown periods for refuelling and maintenance. In preparation for this, the reactor system is cooled down, depressurised and partially drained. Although reactor coolant pressure is lower than during full-power operation, there remains the possibility of a loss-of-coolant accident (LOCA), with a certain but low probability. While the decay heat to be removed is lower than that from a LOCA at full power, the reduced availability of safety systems implies a risk of failing to maintain core cooling, and hence of core damage. This is recognised though probabilistic safety analyses (PSA), which identify low but non-negligible contributions to core damage frequency from accidents during cooldown and shutdown. Analyses are made for a typical two-loop Westinghouse PWR of the consequences of a range of LOCAs during hot and intermediate shutdown, 4 and 5 h after reactor shutdown respectively. The accumulators are isolated, while power to some of the pumped safety injection systems (SIs) is racked out. The study assesses the effectiveness of the nominally assumed SIs in restoring coolant inventory and preventing core damage, and the margin against core damage where their actuation is delayed. The calculations use the engineering-level MELCOR1.8.5 code, supplemented by the SCDAPSIM and SCDAP/RELAP5 codes, which provide a more detailed treatment of coolant system thermal hydraulics and core behaviour. Both treatments show that the core is readily quenched, without damage, by the nominal SI which assumes operation of only one pump. Margins against additional scenario and model uncertainties are assessed by assuming a delay of 900 s (the time needed to actuate the remaining pumps) and a variety of assumptions regarding models and the number of pumps available in conjunction with both MELCOR and versions of SCDAP. Overall, the study provides confidence in the inherent robustness of the plant design with respect to LOCA during

  19. Accident management following loss-of-coolant accidents during cooldown in a Westinghouse two-loop PWR

    Haste, T.J., E-mail: tim.haste@irsn.f [Paul Scherrer Institut (PSI), CH-5232 Villigen (Switzerland); Birchley, J. [Paul Scherrer Institut (PSI), CH-5232 Villigen (Switzerland); Richner, M. [Nordostschweizerische Kraftwerke (NOK) - NPP Beznau, CH-5312 Doettingen (Switzerland)

    2010-06-15

    Operation of pressurised water reactors involves shutdown periods for refuelling and maintenance. In preparation for this, the reactor system is cooled down, depressurised and partially drained. Although reactor coolant pressure is lower than during full-power operation, there remains the possibility of a loss-of-coolant accident (LOCA), with a certain but low probability. While the decay heat to be removed is lower than that from a LOCA at full power, the reduced availability of safety systems implies a risk of failing to maintain core cooling, and hence of core damage. This is recognised though probabilistic safety analyses (PSA), which identify low but non-negligible contributions to core damage frequency from accidents during cooldown and shutdown. Analyses are made for a typical two-loop Westinghouse PWR of the consequences of a range of LOCAs during hot and intermediate shutdown, 4 and 5 h after reactor shutdown respectively. The accumulators are isolated, while power to some of the pumped safety injection systems (SIs) is racked out. The study assesses the effectiveness of the nominally assumed SIs in restoring coolant inventory and preventing core damage, and the margin against core damage where their actuation is delayed. The calculations use the engineering-level MELCOR1.8.5 code, supplemented by the SCDAPSIM and SCDAP/RELAP5 codes, which provide a more detailed treatment of coolant system thermal hydraulics and core behaviour. Both treatments show that the core is readily quenched, without damage, by the nominal SI which assumes operation of only one pump. Margins against additional scenario and model uncertainties are assessed by assuming a delay of 900 s (the time needed to actuate the remaining pumps) and a variety of assumptions regarding models and the number of pumps available in conjunction with both MELCOR and versions of SCDAP. Overall, the study provides confidence in the inherent robustness of the plant design with respect to LOCA during

  20. Effect of Occupational Health and Safety Management System on Work-Related Accident Rate and Differences of Occupational Health and Safety Management System Awareness between Managers in South Korea's Construction Industry

    Yoon, Seok J.; Lin, Hsing K.; Chen, Gang; Yi, Shinjea; Choi, Jeawook; Rui, Zhenhua

    2013-01-01

    Background The study was conducted to investigate the current status of the occupational health and safety management system (OHSMS) in the construction industry and the effect of OHSMS on accident rates. Differences of awareness levels on safety issues among site general managers and occupational health and safety (OHS) managers are identified through surveys. Methods The accident rates for the OHSMS-certified construction companies from 2006 to 2011, when the construction OHSMS became widel...

  1. Cooperative measures to support the Indo-Pak Agreement Reducing Risk from Accidents Relating to Nuclear Weapons.

    Mishra, Sitakanta; Ahmed, Mansoor

    2014-04-01

    In 2012, India and Pakistan reaffirmed the Agreement on Reducing the Risk from Accidents Relating to Nuclear Weapons. Despite a history of mutual animosity and persistent conflict between the two countries, this agreement derives strength from a few successful nuclear confidence building measures that have stood the test of time. It also rests on the hope that the region would be spared a nuclear holocaust from an accidental nuclear weapon detonation that might be misconstrued as a deliberate use of a weapon by the other side. This study brings together two emerging strategic analysts from South Asia to explore measures to support the Agreement and further develop cooperation around this critical issue. This study briefly dwells upon the strategic landscape of nuclear South Asia with the respective nuclear force management structures, doctrines, and postures of India and Pakistan. It outlines the measures in place for the physical protection and safety of nuclear warheads, nuclear materials, and command and control mechanisms in the two countries, and it goes on to identify the prominent, emerging challenges posed by the introduction of new weapon technologies and modernization of the respective strategic forces. This is followed by an analysis of the agreement itself leading up to a proposed framework for cooperative measures that might enhance the spirit and implementation of the agreement.

  2. Traffic Management Systems Performance Measurement: Final Report

    Banks, James H.; Kelly, Gregory

    1997-01-01

    This report documents a study of performance measurement for Transportation Management Centers (TMCs). Performance measurement requirements were analyzed, data collection and management techniques were investigated, and case study traffic data system improvement plans were prepared for two Caltrans districts.

  3. Guidelines for the review of accident management programmes in nuclear power plants. Reference document for the IAEA safety service missions on review of accident management programmes in nuclear power plants

    Similarly as for other IAEA safety services, the objectives of accident management safety service are to assist the Member States in ensuring and enhancing the safety of NPPs. In particular, the objective is to assist at the utility and NPP (i.e. licensee) level in effective plant specific AMP preparation, development and implementation. However, assistance can also be provided to the regulatory body in its reviewing of AMPs. Objectives of the safety service can be summarized as follows: To explain to licensee personnel principles and possible approaches in effective implementation of AMP based on experience world-wide; To give opportunities to experts from the host plant to broaden their experience and knowledge in the field; To perform an objective assessment of the status in various phases of AMP implementation, compared with international experience and practices; To provide the licensee with suggestions and assistance for improvements in various stages of AMP implementation. The objective of the IAEA safety services is to offer two options to respond to individual requirements. These options include missions to review accident analysis needed for accident management and missions to review the whole AMP. Review of accident analysis for accident management (RAAAM): this review is intended to check completeness and quality of accident analysis covering BDBA and severe accidents. The review should be typically performed prior to use of accident analysis for development of AMP. It is considered that 2 experts and 1 IAEA team leader in one-week mission can perform the review. Detailed guidelines for review of analysis are provided in Section 2. Reference is also made to another IAEA Safety Report (Safety Standards Series No. NS-R-1) which is devoted to guidance for accident analysis of nuclear power plants (NPPs). Review of AMP (RAMP): this review of AMP, which is in particular appropriate prior to its implementation, is intended to check its quality, consistency

  4. TL (thermoluminescence) accident dosimetry measurements on samples from the town of Pripyat

    In July 1990, several different types of ceramic samples were collected from the town of Pripyat, situated 3 km NW of the Chernobyl Nuclear Power Plant. The samples were distributed among several laboratories for thermoluminescence (TL) measurement to determine the total absorbed gamma dose at different points within a small area in the most polluted region and to assess the shielding given by the walls of buildings to the people inside apartment blocks. This paper discusses the types of samples and their suitability for accident dosimetry, the TL measurements, minimum limits of detection for various types of samples and the strengths and limitations of the method in this type of situation. The implication of the results are discussed. (author)

  5. Accidental beam loss in superconducting accelerators: Simulations, consequences of accidents and protective measures

    The consequences of an accidental beam loss in superconducting accelerators and colliders of the next generation range from the mundane to rather dramatic, i.e., from superconducting magnet quench, to overheating of critical components, to a total destruction of some units via explosion. Specific measures are required to minimize and eliminate such events as much as practical. In this paper we study such accidents taking the Superconducting Supercollider complex as an example. Particle tracking, beam loss and energy deposition calculations were done using the realistic machine simulation with the Monte-Carlo codes MARS 12 and STRUCT. Protective measures for minimizing the damaging effects of prefire and misfire of injection and extraction kicker magnets are proposed here

  6. Agricultural measures to reduce radiation doses to man caused by severe nuclear accidents

    Agricultural land and products may become contaminated after a severe nuclear accident. If radiation doses to man caused by the ingestion of contaminated agricultural products from such areas will be unacceptably high, measures to reduce this radiation dose will have to be taken. Radiation doses to man can be estimated by using models which describe quantitatively the transfer of radionuclides through the biosphere. The following processes and pathways are described in this study: accidental releases into atmospheric environments and subsequent nearby deposition; contamination of crops by direct deposition and the subsequent short term pathway (e.g. grass-cow-milk-man); contamination of soil and the subsequent long term pathway (e.g. soil-crop-man, soil-grass-cattle-milk/meat-man). Depending on the degree of contamination and on the estimated radiation doses to man, various measures are advised. (Auth.)

  7. Stakeholder involvement in the management of rural areas following a nuclear accident: the farming network

    The importance of the participation of stakeholders in the formulation of strategies for maintaining agricultural production and food safety following a nuclear accident, has been successfully demonstrated by the Agriculture and Food Countermeasures Working Group (AFCWG). This group was set up in the UK by the National Radiological Protection Board (NRPB) and the then Ministry of Agriculture, Fisheries and Food in 1997 (Nisbet and Mondon, 2001). Before this time stakeholder organisations had not collectively considered the implications of contamination of the foodchain in the event of an accidental release of radioactivity. With funding from the European Commission (EC) the UK approach to stakeholder engagement is being taken forward on a European basis during the period 2000-2004 through a project given the acronym FARMING (Food and Agriculture Restoration Management Involving Networked Groups). The overall objective of this project is to create a network of stakeholder working groups in 5 member states (UK, Belgium, Finland, France and Greece) to assist in the development of robust and practicable strategies for restoring and managing contaminated agricultural land and food products in a sustainable way. The initial intention was to involve at least 50 individual stakeholders

  8. Perspectives on Severe Accident Management by Depressurization and External Water Injection under Extended SBO Conditions

    Three major issues of severe accident management guideline (SAMG) after this sort of extended SBO would be depressurization of the primary system, external water injection and hydrogen management inside a containment. Under this situation, typical SAM actions would be depressurization and external water delivery into the core. However, limited amount of external water would necessitate optimization between core cooling, containment integrity and fission product removal. In this paper, effects of SAM actions such as depressurization and external water injection on the reactor and containment conditions after extended SBO are analyzed using MAAP4 code. Positive and negative aspects are discussed with respect to core cooling and fission product retention inside a primary system. Conclusions are made as following: Firstly, early depressurization action itself has two-faces: positive with respect to delay of the reactor vessel failure but negative with respect to the containment failure and fission product retention inside the primary system. Secondly, in order to prevent containment overpressure failure after external water injection, re-closing of PORV later should be considered in SAM, which has never been considered in the previous SAMG. Finally, in case of external water injection, the flow rate should be optimized considering not only the cooling effect but also the long term fission product retention inside the primary system

  9. Applying hierarchical loglinear models to nonfatal underground coal mine accidents for safety management.

    Onder, Mustafa; Onder, Seyhan; Adiguzel, Erhan

    2014-01-01

    Underground mining is considered to be one of the most dangerous industries and mining remains the most hazardous occupation. Categorical analysis of accident records may present valuable information for preventing accidents. In this study, hierarchical loglinear analysis was applied to occupational injuries that occurred in an underground coal mine. The main factors affecting the accidents were defined as occupation, area, reason, accident time and part of body affected. By considering subfactors of the main factors, multiway contingency tables were prepared and, thus, the probabilities that might affect nonfatal injuries were investigated. At the end of the study, important accident risk factors and job groups with a high probability of being exposed to those risk factors were determined. This article presents important information on decreasing the number accidents in underground coal mines. PMID:24934420

  10. Analytical evaluation of dose measurement of critical accident at SILENE (Contract research)

    Institute for Radioprotection and Nuclear Safety (IRSN) and the OECD Nuclear Energy Agency (NEA) jointly organized SILENE Accident Dosimetry Intercomparison Exercise to intercompare the dose measurement systems of participating countries. Each participating country carried out dose measurements in the same irradiation field, and the measurement results were mutually compared. The authors participated in the exercise to measure the doses of gamma rays and neutron from SILENE by using thermoluminescence dosimeters (TLD's) and an alanine dosimeter. In this examination, the authors derived evaluation formulae for obtaining a tissue-absorbed dose from measured value (ambient dose equivalent) of TLD for neutron. We reported the tissue-absorbed dose computed using this evaluation formula to OECD/NEA. TLD's for neutron were irradiated in the TRACY facility to verify the evaluation formulae. The results of TLD's were compared with the calculations of MCNP and measurements with alanine dose meter. We found that the ratio of the dose by the evaluation formula to the measured value by the alanine dosimeter was 0.94 and the formula agreed within 6%. From examination of this TRACY, we can conclude that the value reported to OECD/NEA has equivalent accuracy. (author)

  11. Development of a decision support system for off-site emergency management in the early phase of a nuclear accident

    Full text: Experience gained after the Chernobyl accident clearly demonstrated the importance of improving administrative, organizational and technical emergency management arrangements in India. The more important areas where technical improvements were needed were early warning monitoring, communication networks for the rapid and reliable exchange of radiological and other information and decision support systems for off-site emergency management. A PC based artificial intelligent software has been developed to have a decision support system that can easily implement to manage off-site nuclear emergency and subsequently analyze the off-site consequences of the nuclear accident. A decision support tool, STEPS (source term estimate based on plant status), that provides desired input to the present software was developed. The tool STEPS facilitates meta knowledge of the system. The paper describes the details of the design of the software, functions of various modules, tuning of respective knowledge base and overall its scope in real sense in nuclear emergency preparedness and response

  12. Implications for accident management of adding water to a degrading reactor core

    This report evaluates both the positive and negative consequences of adding water to a degraded reactor core during a severe accident. The evaluation discusses the earliest possible stage at which an accident can be terminated and how plant personnel can best respond to undesired results. Specifically discussed are (a) the potential for plant personnel to add water for a range of severe accidents, (b) the time available for plant personnel to act, (c) possible plant responses to water added during the various stages of core degradation, (d) plant instrumentation available to understand the core condition and (e) the expected response of the instrumentation during the various stages of severe accidents

  13. Development of a national doctrine for the management of the post-accident phase of a radiological emergency situation

    For several years, public Authorities have defined an organization for the management of emergency situations arising from an accident occurring at a nuclear installation. So far, the management of the risk arising from the post accident phase was, in itself, not explored with the same care. What so ever, no format policy on which the action of public Authority could be based is today available. The nuclear safety Authority (ASN), in relation with the other concerned departments, is now in charge, according to the above mentioned directive, to prepare and implement the necessary provisions to respond to a post accident situation. In dune 2005, ASN established the steering committee for the management of post nuclear or radiological emergency situations (CODIRPA). The definition of a national policy related to the management of the radiological risk during a post event situation having to integrate various organization aspects as: lifting of protection emergency provisions and rehabilitation of buildings, life in contaminated rural territories, agriculture and water, dose and radiological consequences, sanitary surveillance of victims and populations, indemnification, waste management of contaminated crops and soils, organization of public Authorities. During the 2. phase of CODIRPA work (2008-2009), the first elements of policy will be consolidated and new scenarios will be studied (one worsened scenario and one scenario with alpha emitting radionuclide). in parallel, a procedure for local actor's consultation should be elaborated. (authors)

  14. International aspects of nuclear accidents

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

  15. Suggestion from young researchers in symposium II conducted by Japan health physics society about Fukushima Daiichi Nuclear Power Plant accident. Focusing on internal exposure management to relate to Fukushima Daiichi Nuclear Power Plant accident

    Fukushima Daiichi Nuclear Power Plants (NPPs) affected by the Great East Japan Earthquake suffered reactor core meltdown and discharged a large amount of radioactive nuclides to the air, which brought about a disorder among the public for internal exposure. Internal exposure management at the accident so as to evaluate internal exposure dose rate of personnel or the public in a quick and optimum way should be standardized with reflecting lessons learned at Fukushima Daiichi NPP accident. Three themes on internal exposure management; (1) thyroid gland screening test, (2) whole-body counters and (3) bioassay, were discussed from young researchers in symposium II conducted by Japan Health Physics Society about Fukushima Daiichi NPP accident. Progression of response to the accident and problems and proposals for each respective theme were presented in the article. (T. Tanaka)

  16. Agro-industrial sphere-radiological consequence of Chernobyl accident and major safety measures

    The early spring radionuclide fall as a consequence of Chernobyl accident caused air contamination of aerial part of agriculture crop - winter crops, natural and seeded permanent grasses. For other plants the soil and wind contamination are prevailing. After radionuclide fall most of them are concentrated in the soil upper layer. Radionuclide uptake depends on the ratio of their concentration in soils, that is varied due to the soil type. The soil development results in variation of radionuclide migration in the crop. In cattle production two main trends are the most significant: estimation of food contamination (primarily of milk and meat) and analysis of physiological state of animals near NPP. Organization and land-improvement measures permitting a stable agro-industrial functionaing on the contaminated territory are considered

  17. Measurement of oxyluminescence of solid waste for speedy dose determination following a radiation accident

    It was the task of the study to show by means of potassium iodide pellets and sugar that chemiluminescence measurement enables quick dose determination on irradiated solid waste, 'quick' meaning within a few hours following a radiation accident. The following results are noteworthy: - The traceability limit of a radiation exposure for potassium iodide is about 0.4 Gy with gamma irradiation (Co-60) and below 0.05 Gy with X-ray irradiation. At present no reasons are known for this great difference of the lower traceability limit. Sugar irradiated with gamma rays can be traced down to about 0.1 Gy. - The intensity of chemiluminescence does not depend on the dose rate in the examined area (0.4 Gy/min to 90 Gy/min). - The radiation exposure can be traced even after a number of months. (orig./HP)

  18. A methodology for supporting decisions on the establishment of protective measures after severe nuclear accidents

    The objective of this report is to demonstrate the use of a methology supporting decisions on protective measures following severe nuclear accidents. A multicriteria decision analysis approach is recommended where value tradeoffs are postponed until the very last stage of the decision process. Use of efficient frontiers is made to exclude all technically inferior solutions and present the decision maker with all nondominated solutions. A choice among these solutions implies a value trade-off among the multiple criteria. An interactive computer packge has been developed where the decision maker can choose a point on the efficient frontier in the consequence space and immediately see the alternative in the decision space resulting in the chosen consequences. The methodology is demonstrated through an application on the choice among possible protective measures in contaminated areas of the former USSR after the Chernobyl accident. Two distinct cases are considered: First a decision is to be made only on the basis of the level of soil contamination with Cs-137 and the total cost of the chosen protective policy; Next the decision is based on the geographic dimension of the contamination ant the total cost. Three alternative countermeasure actions are considered for population segments living on soil contaminated at a certain level or in a specific geographic region: (a) relocation of the population; (b) improvement of the living conditions; and, (c) no countermeasures at all. This is final deliverable of the CEC-CIS Joint Study Project 2, Task 5: Decision-Aiding-System for Establishing Intervention Levels, performed under Contracts COSU-CT91-0007 and COSU-CT92-0021 with the Commission of European Communities through CEPN

  19. Environmental measurements and inspections on imported foods and feedstuffs in Greece after the Fukushima accident.

    Potiriadis, C; Anagnostakis, M J; Clouvas, A; Eleftheriadis, K; Florou, E; Housiadas, C; Ioannides, K; Ioannidou, A; Karangelos, D I; Karfopoulos, K L; Kehagia, K; Kolovou, M; Kritidis, P; Manolopoulou, M; Papastefanou, K; Savva, M I; Simopoulos, S E; Stamoulis, K; Stoulos, S; Xanthos, S; Xarchoulakos, D

    2013-10-01

    The radionuclides released during the accident at the Fukushima Daichii nuclear power plant following the Tōhoku earthquake and tsunami on 11 March 2011 were dispersed in the whole north hemisphere. Traces of (131)I, (134)Cs and (137)Cs reached Greece and were detected in air, grass, sheep milk, ground deposition, rainwater and drainage water. Members of Six Greek laboratories of the national network for environmental radioactivity monitoring have collaborated with the Greek Atomic Energy Commission (GAEC) and carried out measurements during the time period between 11 March 2011 and 10 May 2011 and reported their results to GAEC. These laboratories are sited in three Greek cities, Athens, Thessaloniki and Ioannina, covering a large part of the Greek territory. The concentrations of the radionuclides were studied as a function of time. The first indication for the arrival of the radionuclides in Greece originating from Fukushima accident took place on 24 March 2011. After 28 April 2011', concentrations of all the radionuclides were below the minimum detectable activities (<10 μBq m(-3) for (131)I). The range of concentration values in aerosol particles was 10-520 μBq m(-3) for (131)I, 10-200 μBq m(-3) for (134)Cs and 10-200 μBq m(-3) for (137)Cs and was 10-2200 μBq m(-3) for (131)I in gaseous phase. The ratios of (131)I/(137)Cs and (134)Cs/(137)Cs concentrations are also presented. For (131)I, the maximum concentration detected in grass was 2.2 Bq kg(-1). In the case of sheep milk, the maximum concentration detected for (131)I was 2 Bq l(-1). Furthermore, more than 200 samples of imported foodstuff have been measured in Greece, following the EC directives on the inspection of food and feeding stuffs. PMID:23604742

  20. Work Incapacity and Treatment Costs After Severe Accidents: Standard Versus Intensive Case Management in a 6-Year Randomized Controlled Trial.

    Scholz, Stefan M; Andermatt, Peter; Tobler, Benno L; Spinnler, Dieter

    2016-09-01

    Purpose Case management is widely accepted as an effective method to support medical rehabilitation and vocational reintegration of accident victims with musculoskeletal injuries. This study investigates whether more intensive case management improves outcomes such as work incapacity and treatment costs for severely injured patients. Methods 8,050 patients were randomly allocated either to standard case management (SCM, administered by claims specialists) or intensive case management (ICM, administered by case managers). These study groups differ mainly by caseload, which was approximately 100 cases in SCM and 35 in ICM. The setting is equivalent to a prospective randomized controlled trial. A 6-year follow-up period was chosen in order to encompass both short-term insurance benefits and permanent disability costs. All data were extracted from administrative insurance databases. Results Average work incapacity over the 6-year follow-up, including contributions from daily allowances and permanent losses from disability, was slightly but insignificantly higher under ICM than under SCM (21.6 vs. 21.3 % of pre-accident work capacity). Remaining work incapacity after 6 years of follow-up showed no difference between ICM and SCM (8.9 vs. 8.8 % of pre-accident work incapacity). Treatment costs were 43,500 Swiss Francs (CHF) in ICM compared to 39,800 in SCM (+9.4 %, p = 0.01). The number of care providers involved in ICM was 10.5 compared to 10.0 in ICM (+5.0 %, p reintegration of accident victims. PMID:26687330

  1. Hydrogen-management in beyond design accident conditions in NPP Neckar 2

    Neckar 2 is a 1340 MWE 4-loop pressurized water reactor (PWR) of Siemens KONVOI type, located in the south of Germany. It was first connected to the grid in January 1989. Commercial operation started in April 1989. Task assignment: In Germany it was recommended by the Reactor Safety Commission (RSK) on December 17, 1997, to reequip passive autocatalytic recombiners for the controlling of the hydrogen problem. The removal of the hydrogen is an essential part which guarantees the integrity of the containment. The implementation of the recombiners is a further step for the decrease of the nuclear rest risk. The RSK confirmed, that the implementation of the passive autocatalytic recombiners is a safety measure for the controlled removal of the hydrogen in beyond design accident conditions. Assumption : Failure of the whole residual heat removal system (RHRS) and non sufficient effect of the systems which have been installed for beyond design accident conditions. Effect on the reactor coolant system (RCS): The reactor core will be damaged by non sufficient cooling with the output of hydrogen because all the specified emergency actions have failed. The overheating of the core is responsible for the production of hydrogen by the reaction of zirconium of the fuel-rod cladding with the water vapour. In case of nuclear superheating it would be possible that the reactor vessel would start smelting. The interacting between the core and the concrete, together with the armouring of the biological shield would also produce hydrogen. The hydrogen would escape together with the water vapour out of the leak and would spread out into the whole containment. Results : the number and the position of the different sized recombiners were determined on engineering judgement. the following 4 scenarios are representatively. The 4 scenarios were analyzed for in beyond design accident conditions with the MELCOR-Code: No. 1: Loss of main feedwater supply with primary feed and bleed. No. 2

  2. Implementation of special engineering safety features for severe accident management. New SAMG approach

    Conclusions: As a result of the thermohydraulic analysis conducted the following main conclusions are formulated: The operator actions for accident management are effective and allow reaching conditions for application of the new engineering safety features for SAMG; The new engineering safety features application is effective and prevents severe core damage for Scenario 1. For the Scenario 2 they prevents degradation and relocation of the reactor core for a long period of time (in the analysis this period is 10 h, but the unit could be kept in safe condition for longer time which is not specifically analysed).The maximal fuel cladding temperature for Scenario 1 reaches 558 oC. This low fuel cladding temperature gradient is achieved by applying a complex of operator actions which prevent any core damage. If the additional discharge line with DN 100 mm from the PRZ is not opened then a severe core damage occurs; The maximal fuel cladding temperature for Scenario 2 reaches 1307 oC. One of the possibilities for keeping this temperature below 1200 oC is to mount second line (the first SFP line is between YT12S03.S04) from the SFP to the TQ22 pipeline which is connected to YT14B01 hydroaccumulator line, between the check valves YT14S03.S04

  3. Emergency treatment and nursing management of group patients of traffic accident%成批车祸事故患者的急救与护理管理

    黄宁静

    2015-01-01

    目的 分析成批车祸事故患者的急救与护理.方法 采取2013年7月份至2014年10月份本院收治的4批车祸事故患者共94例,回顾性分析这94例患者的临床资料,总结患者的急救与护理措施.结果 通过采取相应的急救与护理管理后,94例患者中,88例抢救脱险,2例现场死亡,4例抢救无效死亡,死亡率为6.4%.结论 有效的急救与护理管理能有效地降低成批车祸事故中的死亡率,提高车祸事故患者的生存率.%Objective analyze emergency treatment and nursing of group patients of trafifc accident.Method review and analyze 94 patients treated in our hospital from July 2013 to October 2014, 4 groups of patients with trafifc accidents. Summarize their clinical data and emergency treatment and nursing measures.Result after taking corresponding measures of emergency treatment and nursing management, 88 cases of 94 patients survived, 2 cases died on the spot, 4 cases died after emergency treatment, mortality was 6.4%.Conclusion effective emergency management and nursing management can effectively reduce mortality and improve survival rate of patients with trafifc accident.

  4. Organization and implementation of aerial radionuclide measurements after a nuclear accident in Germany

    Full text: Gamma-ray spectrometric systems carried by helicopters prove to be indispensable for the surveillance of environmental radioactivity especially after an accidental release of artificial radionuclides from a nuclear facility. The nuclear accident in Chernobyl has clearly shown that the efforts in performing radioactivity measurements in the environment especially rapid and large-scale nuclide specific determinations of soil contamination had to be improved and intensified. As a result a network with more than 2000 measuring stations has been installed in Germany, which continuously measure the gamma dose rate in the environment. Additionally to these stations, four measuring systems with computerized gamma-ray-spectrometers integrated into helicopters are operated in Germany. These airborne measurement platforms represent an independent method for rapid and large scale detection of soil contamination also in highly polluted or inaccessible areas. The helicopters of type Alouette II are equipped each with a HPGe-detector. Due to its high energy resolution, the HPGe-detector allows to unambiguously identify individual radionuclides. In the case of a nuclear accident, where a great amount of radioactive material is released resulting in a correspondingly high contamination of the environment, the measurement periods of the HPGe-detector can be reduced to a few seconds and thus a high spatial resolution can be obtained. The lower limit of detection for an HPGe-detector for 137Cs at a flight altitude of 100 m above ground and measurement times of 30 s is about 4 kBq m-2. The helicopters are operated by the Federal Border Police who is also responsible for the logistic and the flight regime. The airborne measurement systems are placed at two institutes of the Federal Office of Radiation Protection in Berlin and Muenchen. Both institutes lie close to airfields of the Federal Border Police. In case of emergency the measuring systems can quickly be transported to

  5. Serious and fatal accidents in 2011 in immigrant workers: considerations on the phenomenon and preventive measures

    Innocenzi M

    2012-11-01

    Full Text Available Background: Traumatic events or serious injury, or death occurring to foreign nationals are mostly events of nature work, and the competence to indemnify working accidents (and occupational diseases is attributed to INAIL. An accident at work is defined as a traumatic event which occurred through the intervention of a violent cause during the work, determining a worker's personal injury identified in a temporary incapacity, permanent disability, (allowance for damage between 1 and 5%, with a lump sum payment for the damages of between 6 and 15%, with monthly income for damages equal to or greater than 16% or death. In recent years, Italy has shown a general reduction in the number of harmful events and fatalities, and this was also the case for foreign workers, but in the face of such data has highlighted the persistence of a significant number of serious multiple injuries and deaths. Objectives: To evaluate the possible additional risk factors and possible preventive measures. Methods: The present study investigated the time course of serious and fatal injuries in foreign workers from 2008 to 2011, and in more detail the events for the year 2011, taking into account the business sector, the methods of the event, the spatial distribution and the nationality of the workers, examining the data obtained from the Annual Reports INAIL . For the serious injury is highlighted a progressive decrease in foreign workers during the years 2008 to 2011, in industry and services and, to a lesser extent, in agriculture. Data on fatalities instead show a substantial stability in the number of them, both in percentage and in numerical values. Discussion: It's possible that the factors that contribute to an increased risk of serious and fatal events in foreign workers can be: the imperfect knowledge of the Italian language, the lack of specific training in relation to occupational hazards, the irregularity and uncertainty that often characterizes their work

  6. Nuclear accident dosimetry: Los Alamos measurements at the seventeenth nuclear accident dosimetry intercomparison study at the Oak Ridge National Lab., DOSAR Facility, August 1980

    Teams from various US and foreign organizations participated in the Seventeenth Nuclear Accident Dosimetry Study held at the Oak Ridge National Laboratory's (ORNL) Dosimetry Applications Research (DOSAR) facility August 11 to 15, 1980. Criticality dosimeters were simultaneously exposed to pulses of mixed neutron and gamma radiation from the Health Physics Research Reactor (HPRR). This report summarizes the experimental work conducted by the Los Alamos team. In-air and phantom measurements were conducted by the Los Alamos team using area and personnel dosimeters. Combined blood sodium and sulfur fluence measurements of absorbed dose were also made. In addition, indium foils placed on phantoms were evaluated for the purpose of screening personnel for radiation exposure. All measurements were conducted for unshielded, 5-cm steel and 20-cm concrete shielding configurations. All participant dosimeters were exposed at 3 m from the center of the HPRR core

  7. Hazardous waste storage facility accident scenarios for the U.S. Department of Energy Environmental Restoration and Waste Management Programmatic Environmental Impact Statement

    Policastro, A.; Roglans-Ribas, J.; Marmer, D.; Lazaro, M.; Mueller, C. [Argonne National Lab., IL (United States); Freeman, W. [Univ. of Illinois, Chicago, IL (United States). Dept. of Chemistry

    1994-03-01

    This paper presents the methods for developing accident categories and accident frequencies for internally initiated accidents at hazardous waste storage facilities (HWSFs) at US Department of Energy (DOE) sites. This categorization is a necessary first step in evaluating the risk of accidents to workers and the general population at each of the sites. This risk evaluation is part of the process of comparing alternative management strategies in DOE`s Environmental Restoration and Waste Management (EM) Programmatic Environmental Impact Statement (PEIS). Such strategies involve regionalization, decentralization, and centralization of waste treatment, storage, and disposal activities. Potential accidents at the HWSFs at the DOE sites are divided into categories of spill alone, spill plus fire, and other event combinations including spill plus fire plus explosion, fire only, spill and explosion, and fire and explosion. One or more accidents are chosen to represent the types of accidents for FY 1992 for 12 DOE sites were studied to determine the most representative set of possible accidents at all DOE sites. Each accident scenario is given a probability of occurrence that is adjusted, depending on the throughput and waste composition that passes through the HWSF at the particular site. The justification for the probabilities chosen is presented.

  8. Hazardous waste storage facility accident scenarios for the U.S. Department of Energy Environmental Restoration and Waste Management Programmatic Environmental Impact Statement

    This paper presents the methods for developing accident categories and accident frequencies for internally initiated accidents at hazardous waste storage facilities (HWSFs) at US Department of Energy (DOE) sites. This categorization is a necessary first step in evaluating the risk of accidents to workers and the general population at each of the sites. This risk evaluation is part of the process of comparing alternative management strategies in DOE's Environmental Restoration and Waste Management (EM) Programmatic Environmental Impact Statement (PEIS). Such strategies involve regionalization, decentralization, and centralization of waste treatment, storage, and disposal activities. Potential accidents at the HWSFs at the DOE sites are divided into categories of spill alone, spill plus fire, and other event combinations including spill plus fire plus explosion, fire only, spill and explosion, and fire and explosion. One or more accidents are chosen to represent the types of accidents for FY 1992 for 12 DOE sites were studied to determine the most representative set of possible accidents at all DOE sites. Each accident scenario is given a probability of occurrence that is adjusted, depending on the throughput and waste composition that passes through the HWSF at the particular site. The justification for the probabilities chosen is presented

  9. Level-2 PSA for the Prototype Fast Breeder Reactor MONJU Applied to the Accident Management Review

    JNES independently evaluated the three events it selected - PLOHS, LORL and ATWS events - and reviewed the results of the Level 2 PSA carried out by JAEA. Regarding ATWS events, the organization carried out a qualitative evaluation of the results of JAEA's evaluation and carried out a quantitative evaluation of the containment failure frequency (CFF) in relation to PLOHS and LORL events. In JNES's independent evaluation of PLOHS and LORL events, accident scenarios in the three phases - the plant response phase, the core damage phase and the containment vessel response phase - were analyzed. The phenomenal event trees were quantified by applying the information about phenomena specific to fast reactors, including plant thermal-hydraulic analysis at the time of core damage, boundary structure analysis, analysis of the characteristics of the disrupted core, the results of sodium-concrete reaction tests, and the results of hydrogen diffusion induced combustion tests, to the PRDs. As the result, the total CFF before the preparation of the AM measures was rated at 9.2E-9/reactor year (CDF at 2.7E-7/reactor year), and it has been confirmed that these numerical values are well below the power reactor performance goal indicator values (CDF: 10-4/year or so; CFF: 10-5/year or so) even before the preparation of the AM measures. (author)

  10. Database of meteorological and radiation measurements made in Belarus during the first three months following the Chernobyl accident

    Results of all available meteorological and radiation measurements that were performed in Belarus during the first three months after the Chernobyl accident were collected from various sources and incorporated into a single database. Meteorological information such as precipitation, wind speed and direction, and temperature in localities were obtained from meteorological station facilities. Radiation measurements include gamma-exposure rate in air, daily fallout, concentration of different radionuclides in soil, grass, cow's milk and water as well as total beta-activity in cow's milk. Considerable efforts were made to evaluate the reliability of the measurements that were collected. The electronic database can be searched according to type of measurement, date, and location. The main purpose of the database is to provide reliable data that can be used in the reconstruction of thyroid doses resulting from the Chernobyl accident. - Highlights: ► Meteorological and radiation measurements done after the Chernobyl accident in Belarus were collected. ► Data were verified and incorporated into a single database. ► Results of this study is being used to improve the thyroid dose estimates after the Chernobyl accident.

  11. Analysis of Surface Water Pollution Accidents in China: Characteristics and Lessons for Risk Management

    Yao, Hong; Zhang, Tongzhu; Liu, Bo; Lu, Feng; Fang, Shurong; You, Zhen

    2016-04-01

    Understanding historical accidents is important for accident prevention and risk mitigation; however, there are no public databases of pollution accidents in China, and no detailed information regarding such incidents is readily available. Thus, 653 representative cases of surface water pollution accidents in China were identified and described as a function of time, location, materials involved, origin, and causes. The severity and other features of the accidents, frequency and quantities of chemicals involved, frequency and number of people poisoned, frequency and number of people affected, frequency and time for which pollution lasted, and frequency and length of pollution zone were effectively used to value and estimate the accumulated probabilities. The probabilities of occurrences of various types based on origin and causes were also summarized based on these observations. The following conclusions can be drawn from these analyses: (1) There was a high proportion of accidents involving multi-district boundary regions and drinking water crises, indicating that more attention should be paid to environmental risk prevention and the mitigation of such incidents. (2) A high proportion of accidents originated from small-sized chemical plants, indicating that these types of enterprises should be considered during policy making. (3) The most common cause (49.8 % of the total) was intentional acts (illegal discharge); accordingly, efforts to increase environmental consciousness in China should be enhanced.

  12. A review of post-accident mitigative measures affecting transport and isolation of radionuclides released from the Chernobyl accident

    This paper summarizes the results of eight years of mitigative measures to radioactive contamination within the 30 kilometer exclusion zone surrounding the Chernobyl Nuclear Power Plant. We hope to demonstrate that effectiveness of mitigative measures depends not only on proper application of technology but also on selection of projects offering significant risk reduction potential. In a limited national economy, environmental mitigation projects must maximize risk reduction and cost effectiveness or risk losing funding to more pressing social issues

  13. Role of Laws and Regulations For Nuclear Energy Installation in Developing Safety Measures Against Accident

    The energy industry has been considered as an economic development driver. The fundamental safety policy for nuclear facilities is to protect health and safety of the public and the site personnel against undue risks associated with radiation and radioactive materials resulting from normal operation and abnormal conditions. This policy is implemented, based on the as low as reasonably achievable (ALARA) principle for normal operation and the defense-in-depth principle (prevention of the occurrence of anomalies, prevention of the escalation of anomalies into accidents, and prevention of excessive release of radioactive materials into the environment), through establishment of safety guides and standards. More over the consideration of suitable site selection and safety design, verification by safety evaluation, quality assurance for manufacturing, construction and operation, periodic testing and inspection, confirmation by regulatory bodies, and reflection of experienced troubles to safety countermeasures. Are of these paramount importance concepts are applied variety of nuclear facilities, which is, nuclear reactors, uranium enrichment plants, fuel conversion/fabrication plants, reprocessing plants, radioactive waste management facilities, and so on, considering unique features of each facility.

  14. Dose estimation and evaluation of protector measures for a power plant's accidents scenario, using geographical information system

    Since the initial phase of a project of a nuclear plant several environmental studies are carried out, and a considerable amount of relevant information is generated. Therefore, there is an increasing need of an integrated analysis of this information in order to better evaluate the potential impact associated to hypothetical accident scenarios of such plants. This paper presents a case-study, in which a hypothetical accident scenario is analysed taking into account the environmental and populational information of the Brazilian nuclear power plants region by using a geographical information system. Important areas for planning of protective measures are identified to provide a basis for further analysis. (author)

  15. Measurement of Iodine-129 concentration in water samples in relation with Fukushima Daiichi Nuclear Power Plant accident

    Iodine-129 (129I) concentration in several water samples were measured by means of Accelerator Mass Spectrometry and discussed in relation with Fukushima Daiichi Nuclear Power Plant (FDNPP) accident. 129I concentration of river waters collected Iitate village and Minami-Soma city (North to North-west of FDNPP) showed as high as 1.0x109 atoms/L and had not vary significantly during period from March to October, 2012. The combination of 129I/127I ratio and 127I concentration of these water samples can be explained as mixture of fossil rain water (ground water) and the rain radioactively contaminated by FDNPP accident. (author)

  16. Measurements of the Chernobyl accident fallout in Israel and the assessment of the radiation doses to the population

    Israel is located approximately 2000 km southeast of Chernobyl. The fallout from the accident in Chernobyl reactor no. 4 on April 26, 1986 arrived in Israel on the night of May 2nd. Following the accident, studies of the radiological effects were initiated by many countries some of them many thousands of kilometers away. These studies can be characterized by three periods: a) First months following the accident - Measurements were taken to assess the immediate impact and to propose countermeasures that would reduce doses incurred by the population. b) First years following the accidents - Measurements were taken to validate that radioecological effects are well below any regulatory limits, from both the fallout radioactivity in the country and import of food coming from other affected areas. c) The last years (e.g. 1990-1995) - Measurements were taken within the regular program of environmental radioactivity surveillance. In this paper we have compiled the results of the studies in Israel which have followed the three phases mentioned above. Assessment of the accumulated potential radiation doses to the population in Israel was made based on the results of those measurements covered in the three phases, considering the various possible pathways

  17. Validation of techniques for simulating long range dispersal and deposition of atmospheric pollutants based upon measurements after the Chernobyl accident

    Problem specifications and a time schedule for an international study of computerized simulation of transfrontier atmospheric contamination are presented. Started on the initiative of the Nordic Liaison Committee for Atomic Energy, the study will be based on international measurements after the Chernobyl accident

  18. Manual on the medical management of individuals involved in radiation accidents

    This manual is concerned with accidents or emergencies which involve sources of ionizing radiation. It does not cover other forms of radiation such as non-ionizing radiation (ultra-violet, light, radiofrequency radiations), heat, etc. Most radiation accidents have involved individuals either at the workplace or with medical misadministrations; they have received external exposure from X-ray or gamma-ray sources or have been contaminated with radioactive material. A few members of the public have also been involved through misadventures with radioactive sources although these may not be thought of as accidents; more commonly, they are referred to as 'incidents'. For the purpose of this manual, there is not differentiation between an accident and an incident, as the medical care required is the same in both situations. Some of the reference papers are reprinted at the back of the manual. 17 refs., 12 tabs., 9 figs

  19. Accident management in the case of serious emergencies in nuclear power plant

    On-site emergency planning comprises all action taken in a nuclear power station to identify beyond-design base accidents at an early stage and reliably, to keep it under control and overcome it with the minimum of damage. The individual papers set out the basic terminology, the thermohydraulic processes in the cooling circuits during severe incidents, action to maintain the integrity of the containment, the potential of expert systems, simulator training and new developments for simulating accident conditions. (DG)

  20. Applications of autoassociative neural networks for signal validation in accident management

    The OECD Halden Reactor Project has been working for several years with computer based systems for determination on plant status including early fault detection and signal validation. The method here presented explores the possibility to use a neural network approach to validate important process signals during normal and abnormal plant conditions. In BWR plants, signal validation has two important applications: reliable thermal limits calculation and reliable inputs to other computerized systems that support the operator during accident scenarious. This work shows how a properly trained autoassociative neural network can promptly detect faulty process signal measurements and produce a best estimate of the actual process value. Noise has been artificially added to the input to evaluate the network ability to respond in a very low signal to noise ratio environment. Training and test datasets have been simulated by the real time transient simulator code APROS. Future development addresses the validation of the model through the use of real data from the plant. (author). 5 refs, 17 figs