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Sample records for accident management strategy

  1. 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)

  2. 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

  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. 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.))

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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.

  11. 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)

  12. 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.)

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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)

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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)

  7. 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.

  8. 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

  9. 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

  10. 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

  11. 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.

  12. 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

  13. 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)

  14. 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

  15. 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

  16. 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)

  17. 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)

  18. 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.

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. Application of Core Exit Temperature for Effective Safety Injection Strategy of Severe Accident Management Guidance

    Due to limited time for operator's action under the postulated severe accident, immediate and short term actions are needed and relevant strategies are constructed in the SAMG. Therefore, the SAMG includes a variety of information to assist the proper operator actions. Among these, pre-calculated graphs and formulas facilitate understanding of plant status and operator's action needed for accident mitigation. These are essential for ease of application and regarded as Computational Aids (CA). The representative example is the estimation of injection flow rates for removing decay heat and oxidation heat of core, and hydrogen generation rate, to mention a few. Most of all, calculation of the necessary injection flow rate is important in order to mitigate and/or terminate core damages. In estimating the flow rate for accident mitigation, Core Exit Temperature (CET) is utilized as a key variable. CET is considered most effective and reliable means for diagnosing core state. As such, CET has been adopted as a criterion transitioning from EOPs to SAMG. In this study, the necessary flow rate is calculated utilizing simple model with CET for RCS injection in mitigation strategy of SAMG. MELCOR simulation results are introduced for the calculation. A simple model of flow rate necessary for core heat removal is developed using CET data obtained from MELCOR simulations of OPR1000. The suggested model is expected to contribute on judging the core state in its coolability and required flow injection due to ease of application. More detailed analyses are needed to normalize by including additional accident scenarios

  5. 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

  6. External cooling: The SWR 1000 severe accident management strategy. Part 1: motivation, strategy, analysis: melt phase, vessel integrity during melt-water interaction

    This paper provides the description of the basics behind design features for the severe accident management strategy of the SWR 1000. The hydrogen detonation/deflagration problem is avoided by containment inertization. In-vessel retention of molten core debris via water cooling of the external surface of the reactor vessel is the severe accident management concept of the SWR 1000 passive plant. During postulated bounding severe accidents, the accident management strategy is to flood the reactor cavity with Core Flooding Pool water and to submerge the reactor vessel, thus preventing vessel failure in the SWR 1000. Considerable safety margins have determined by using state of the art experiment and analysis: regarding (a) strength of the vessel during the melt relocation and its interaction with water; (b) the heat flux at the external vessel wall; (c) the structural resistance of the hot structures during the long term period. Ex-vessel events are prevented by preserving the integrity of the vessel and its penetrations and by assuring positive external pressure at the predominant part of the external vessel in the region of the molten corium pool. Part 1 describes the motivation for selecting this strategy, the general description of the strategy and the part of the analysis associated with the vessel integrity during the melt-water interaction. (author)

  7. 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.

  8. 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.

  9. 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)

  10. 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)

  11. Development of TRAIN for accident management

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

  12. In-vessel melt retention as a severe accident management strategy for the Loviisa Nuclear Power Plant

    Kymaelaeinen, O.; Tuomisto, H. [IVO International Ltd., Vantaa (Finland); Theofanous, T.G. [Univ. of California, Santa Barbara, CA (United States)

    1997-02-01

    The concept of lower head coolability and in-vessel retention of corium has been approved as a basic element of the severe accident management strategy for IVO`s Loviisa Plant (VVER-440) in Finland. The selected approach takes advantage of the unique features of the plant such as low power density, reactor pressure vessel without penetrations at the bottom and ice-condenser containment which ensures flooded cavity in all risk significant sequences. The thermal analyses, which are supported by experimental program, demonstrate that in Loviisa the molten corium on the lower head of the reactor vessel is coolable externally with wide margins. This paper summarizes the approach and the plant modifications being implemented. During the approval process some technical concerns were raised, particularly with regard to thermal loadings caused by contact of cool cavity water and hot corium with the reactor vessel. Resolution of these concerns is also discussed.

  13. Computerised severe accident management aids

    operational regimes. If a severe accident situation occurs, CAMS should be able to switch to use and present pre-calculated transients from severe accident codes such as MAAP. Another approach would be to switch from using the SMABRE code to using the on-line MAAP-code [6] for detailed calculations during severe accident states. The switch could be governed by a very simple rule such as 'Switch to the MAAP-code when core temperature > 700 deg. C'. A third approach could be to use the on-line MARS code for the whole programme, but there are several arguments against this. The on-line MARS code is not available at present, and the system is assumed to be expensive. There is not very much information available about the system, and it is unclear how difficult it will be to include MARS in the CAMS design. The MARS code will presumably not be useful for other operational regimes than the most serious ones. It will therefore not meet the stated objectives. In CAMS the possibility to assess different accident management strategies is achieved by means of the strategy generator and the predictive simulator based on the SMABRE code. By means of interaction with the strategy generator and some stored information on accident management principles, it will be possible to formulate a first control strategy. This strategy is given to the predictive simulator that calculates forward in time. The outcome can be studied by the operator /TSC. Such a calculation can be done assuming no operator actions or with operator actions. A realisation of the proposed CAMS system design will accordingly meet the stated objectives both with respect to normal operation, accident and severe accident operational regimes. The modularized flexible design will allow a smooth expansion of the system with new modules. The system is designed around a central database. The inclusion of a more or less complete plant simulator in the plant monitoring system is adding a new dimension to the plant monitoring task. Any

  14. Effectiveness and adverse effects of reactor coolant system depressurization strategy with various severe accident management guidance entry conditions for OPR1000

    Severe accident analysis for Korean OPR1000 with MELCOR 1.8.6 was performed by adapting a mitigation strategy under different entry conditions of Severe Accident Management Guidance (SAMG). The analysis was focused on the effectiveness of the mitigation strategy and its adverse effects. Four core exit temperatures (CETs) were selected as SAMG entry conditions, and Small Break Loss of Coolant Accident (SBLOCA), Station Blackout (SBO), and Total Loss of Feed Water (TLOFW) were selected as postulated scenarios that may propagate into severe accidents. In order to delay reactor pressure vessel (RPV) failure, entering the SAMG when the CET reached 923 K, 923 K, and 753 K resulted in the best results for SBLOCA, SBO, and TLOFW scenarios, respectively. This implies that using event-based diagnosis for severe accidents may be more beneficial than using symptom-based diagnosis. There is no significant difference among selected SAMG entry conditions in light of the operator's available action time before the RPV failure. Potential vulnerability of the RPV due to hydrogen generation was analyzed to investigate the foreseeable adverse effects that act against the accident mitigation strategies. For the SBLOCA cases, mitigation cases generated more hydrogen than the base case. However, the amount of hydrogen generated was similar between the base and mitigation cases for SBO and TLOFW. Hydrogen concentrations of containment were less than 5% before RPV failure for most cases. (author)

  15. 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

  16. 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)

  17. Strategies for operation of containment related ESFs in managing activity release to the environment during accident conditions

    In Indian PHWR design, a double containment concept with passive vapour suppression pool (to limit peak pressure) system has been adopted. In addition to it, various Engineered Safety Features (ESFs) have been incorporated to limit the release of radioactivity to the environment. They are: Reactor building emergency coolers for cooling which results in fast reduction of overpressure; Primary Containment Filtration and Pump Back System (PCFPBS) for reduction in iodine concentration inside RB atmosphere during post LOCA period; and, Primary Containment Controlled Discharge System (PCCDS) for the rapid reduction of over-pressure tail. Due to operation of secondary containment purge system, which maintain negative pressure in the annulus, the ground level release is negligibly small. However, if non- availability of negative pressure in secondary containment space is assumed, then operation of PCFPBS and PCCDS system reduces the ground level release significantly. In this situation, depending upon time of operation of the PCFPBS, it can effectively reduce the iodine release, both in stack level and ground level by trapping it in charcoal filters. It is seen that delay time of PCFPBS operation in conjunction with prevailing weather condition can be manipulated to reduce the effect of stack level release of iodine. In this paper the containment related ESFs used in Indian PHWR is discussed in brief and the effectiveness of operator actions and management strategies in actuation of the ESFs in reducing the activity release to environment (during postulated accident conditions) will be brought out. (author)

  18. 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

  19. 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

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

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

    2005-04-15

    Main components and structures in nuclear power plants generally use materials having superior resistance to corrosion.Since the damages related to corrosion have become a menace to the safety of NPPs as well as economical loss and the steam generator tubing forming a boundary between the primary and secondary sides of NPPs is one of the main components that are most damaged by corrosion, it is strongly required to verify the mechanisms of the steam generator tubing degradations, to develop remedial techniques for the degradations, to manage the damages, and to develop techniques for the extension of the plant's life. In this study, the PWSCC characteristics of the archived steam generator tube materials in the domestic NPPs were evaluated and the databases of the obtained results were established. Also, the PWSCC characteristics of the welding material, Alloy 182, for Alloy 600, were evaluated. To verify the damage mechanisms of the circumferential SCC occurring in the expansion transition region of the tubes in the Korean standard NPPS, the evaluation technique for the residual stresses in the expanded region was acquired. A procedure of the inhibition technique for the SCC occurring in the secondary side of steam generators and a model for estimating the safety of damaged tubes by the structural leakage were developed, by which the fundamental technologies for the safe operations of NPPs, the management of the damages, and the expansion of the plant life were acquired. The material improvement technique for the integrity enhancement of tubes was developed. Along with the development of the Ni-coating technique the evaluation of the properties such as mechanical and SCC properties of the coated film was performed.

  1. 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

  2. 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)

  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. 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)

  5. 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)

  6. Alternative evacuation strategies for nuclear power accidents

    In the U.S., current protective-action strategies to safeguard the public following a nuclear power accident have remained largely unchanged since their implementation in the early 1980s. In the past thirty years, new technologies have been introduced, allowing faster computations, better modeling of predicted radiological consequences, and improved accident mapping using geographic information systems (GIS). Utilizing these new technologies, we evaluate the efficacy of alternative strategies, called adaptive protective action zones (APAZs), that use site-specific and event-specific data to dynamically determine evacuation boundaries with simple heuristics in order to better inform protective action decisions (rather than relying on pre-event regulatory bright lines). Several candidate APAZs were developed and then compared to the Nuclear Regulatory Commission’s keyhole evacuation strategy (and full evacuation of the emergency planning zone). Two of the APAZs were better on average than existing NRC strategies at reducing either the radiological exposure, the population evacuated, or both. These APAZs are especially effective for larger radioactive plumes and at high population sites; one of them is better at reducing radiation exposure, while the other is better at reducing the size of the population evacuated. - Highlights: • Developed framework to compare nuclear power accident evacuation strategies. • Evacuation strategies were compared on basis of radiological and evacuation risk. • Current strategies are adequate for smaller scale nuclear power accidents. • New strategies reduced radiation exposure and evacuation size for larger accidents

  7. Accident management information needs

    The tables contained in this Appendix A describe the information needs for a pressurized water reactor (PWR) with a large, dry containment. To identify these information needs, the branch points in the safety objective trees were examined to decide what information is necessary to (a) determine the status of the safety functions in the plant, i.e., whether the safety functions are being adequately maintained within predetermined limits, (b) identify plant behavior (mechanisms) or precursors to this behavior which indicate that a challenge to plant safety is occurring or is imminent, and (c) select strategies that will prevent or mitigate this plant behavior and monitor the implementation and effectiveness of these strategies. The information needs for the challenges to the safety functions are not examined since the summation of the information needs for all mechanisms associated with a challenge comprise the information needs for the challenge itself

  8. Strategy for the Development of Severe Accident Analysis Technology

    To ensure the safety of people living near the nuclear power plants during the postulated events of severe accidents, a severe accident management strategy is prepared for the operating reactors and dedicated engineered features for the severe accidents are under research and development for the new reactors, such as GEN-III reactors. To accomplish these tasks, not only a proper understanding of fundamental physics of severe accident phenomena but also reliable computer codes for analyzing the severe accident phenomena is very necessary. This report deals with a strategic plan for a development and provision of computer code system for analyzing the severe accidents. This reports includes a summary of major phenomena of severe accidents, an peer review of the computer codes for analyzing the integral behavior of severe accident scenario and computer codes for analyzing the specific phenomena. Finally, a strategic plan for an equipment of severe accident computer codes either by use of already available computer codes or a development of our own computer codes, which could be competitive with world class foreign computer codes

  9. 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

  10. Analysis of Hydrogen Control Strategy Using Igniter during Severe Accident

    The Severe Accident Management Guidelines (SAMGs) for the operating pressurized water reactor (PWR) have been completed within 2006. Among the SAMG strategies, mitigation-07 is the most important strategy for managing a severe accident of a PWR in order to reduce containment hydrogen. The fastest way to reduce the containment hydrogen concentration is to intentionally ignite the hydrogen. For this strategy, igniters exist in Optimized Power Reactor 1000 (OPR 1000) to burn hydrogen for a severe accident. For using the igniters during a severe accident, the adverse effects such as the explosion of the hydrogen mixture should be considered for containment integrity. However, an applicable discrimination method to activate the igniters does not exist, so that the hydrogen control strategy using the igniters cannot be chosen during a severe accident. Thus, this study focused on suggesting an applicable discrimination method to carry out the strategy of using the igniters. In this study, the specific plant used for this analysis is Ulchin Unit 5 and 6, OPR 1000 plant, in Korea

  11. 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

  12. 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

  13. Occupational Radiation Protection in Severe Accident Management

    application. Chapter 6 discusses monitoring and management strategies for the radioactive releases and contamination control during the emergency phase. Appendix-1 addresses key lessons learned from past accidents, including TMI, Chernobyl and Fukushima Daiichi and Appendix-2 includes information on the international workshop, which was organized in June 2014 to finalize this ISOE expert group report

  14. 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)

  15. 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

  16. Depressurization as an accident management strategy for Jose Cabrera nuclear plant loss of feedwater and station blackout events

    This paper reports on an evaluation of the efficiency of the operator initiated depressurization in the Spanish Westinghouse one loop Jose Cabrera nuclear power plant that has been developed. This operation is recommended in the present emergency procedure for the total loss of feedwater event in the bleed and feed mode. RELAP5/MOD2 analyses show that this is an effective measure to bring the plant to a cold and stable condition in a design-based accident scenario

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. Strategies for dealing with resistance to recommendations from accident investigations

    Lundberg, J.; Rollenhagen, C.; Hollnagel, E.;

    2012-01-01

    Accident investigation reports usually lead to a set of recommendations for change. These recommendations are, however, sometimes resisted for reasons such as various aspects of ethics and power. When accident investigators are aware of this, they use several strategies to overcome the resistance....... This paper describes strategies for dealing with four different types of resistance to change. The strategies were derived from qualitative analysis of 25 interviews with Swedish accident investigators from seven application domains. The main contribution of the paper is a better understanding of...... effective strategies for achieving change associated with accident investigation. (C) 2011 Elsevier Ltd. All rights reserved....

  4. 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

  5. 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

  6. 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.

  7. Code strategy for simulating Severe Accident Scenario

    Severe accident scenarios of Sodium-cooled fast reactors involves various phenomena: core degradation, melt progression towards the core catcher, corium behaviour on the core catcher, energetic corium/sodium interactions, structure mechanical behaviour during expansion phase, containment behaviour, and fission production release and transport. In order to simulate the complete accident scenarios, CEA strategy relies on two sets of calculation codes: a reference set of codes and a set of simplified coupled models dedicated to Probabilistic Risk Assessment analyses. Concerning the reference set, that includes SAS-SFR, SIMMER, CONTAIN, EUROPLEXUS, and TOLBIAC, CEA started, with JAEA and KIT, a validation process based on existing experimental results such as CABRI and SCARABEE programs, and recently against the EAGLE1&2 program results, in the frame of a specific contract with JAEA. Furthermore, CEA is preparing additional experimental programs including in-pile experiments in IGR (NNC reactor), and out-of-pile experiments in the future experimental FOURNAISE facility to be built in CEA Cadarache (France). (author)

  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. 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

  10. 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

  11. 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)

  12. 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

  13. 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)

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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.

  1. 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

  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. 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

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

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

  5. 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)

  6. 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

  7. 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

  8. 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.

  9. 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

  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. 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

  12. 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.)

  13. 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

  14. 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

  15. 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.

  16. 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

  17. "Defence-in-Depth" Strategy in Transport Risk Management

    Szymanek, Andrzej

    Safety management is a kind of system management, that is management by purposes. Taking "defence-in-depth" strategy, DDS - there can be defined four main aims and four method groups of risk management in transport: 1. minimizing transport accidents risk; 2. minimizing number of undesirable transport events (incidents, conflicts, collisions, accidents). Above purposes relate stages of safety management in transport. At each level of management should be elaborated methods, procedures and technologies of minimizing transport accidents risk. According to DDS any management system of transport safety should have a structure of multilevel chain protections which supervise main transport processes. About those problems in the paper.

  18. Strategies for psychosocial risk management in manufacturing

    Guadix Martín, José; Carrillo Castrillo, Jesús Antonio; Onieva Giménez, Luis Gerardo; Lucena, David

    2015-01-01

    Psychosocial risk is a concern for employers across Europe. Psychosocial risk management, however, is younger than other risk management fields such as safety, hygiene, and ergonomics. Psychosocial risk control prevents accidents and absenteeism. This study examines strategies for psychosocial risk management in manufacturing organizations. The study employs structural equation modeling to analyze results of the European Survey of Enterprises onNewand Emerging Risks (ESENER), a survey that fi...

  19. 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)

  20. 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.

  1. 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

  2. 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

  3. 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

  4. Investment Strategy Based on Aviation Accidents: Are there abnormal returns?

    Marcos Rosa Costa

    2013-06-01

    Full Text Available This article investigates whether an investment strategy based on aviation accidents can generate abnormal returns. We performed an event study considering all the aviation accidents with more than 10 fatalities in the period from 1998 to 2009 and the stock market performance of the respective airlines and aircraft manufacturers in the days after the event. The tests performed were based on the model of Campbell, Lo & MacKinlay (1997 for definition of abnormal returns, by means of linear regression between the firms’ stock returns and the return of a market portfolio used as a benchmark. This enabled projecting the expected future returns of the airlines and aircraft makers, for comparison with the observed returns after each event. The result obtained suggests that an investment strategy based on aviation accidents is feasible because abnormal returns can be obtained in the period immediately following an aviation disaster.

  5. Speed Management Strategies; A Systematic Review

    Homayoun Sadeghi-Bazargani

    2016-07-01

    Full Text Available Objective: To systematically identify the various methods of speed management and their effects. Methods: A systematic search was performed in Science Direct, Ovid Medline, Scopus, PubMed and ProQuest databases from April to June 2015. Hand searching and reference of selected articles were used to improve article identification. Articles published after 1990 which had reported on efficacy/effectiveness of speed management strategies were included. Data were extracted using pre-defined extraction table. Results: Of the 803 retrieved articles, 22 articles were included in this review. Most of the included articles (63% had before-after design and were done in European countries. Speed cameras, engineering schemes, intelligent speed adaption (ISA, speed limits and zones, vehicle activated sign and integrated strategies were the most common strategies reported in the literature. Various strategies had different effects on mean speed of the vehicles ranging from 1.6 to 10 km/h. Moreover, 8-65% and 11-71% reduction was reported in person injured accidents and fatal accidents, respectively as a result of employing various strategies. Conclusion: Literature revealed positive effects of various speed management strategies. Using various strategies was mostly dependent on road characteristics, driver’s attitude about the strategy as well as economic and technological capabilities of the country. Political support is considered as a main determinant in selecting speed management strategies.

  6. Speed Management Strategies; A Systematic Review

    Sadeghi-Bazargani, Homayoun; Saadati, Mohammad

    2016-01-01

    Objective: To systematically identify the various methods of speed management and their effects. Methods: A systematic search was performed in Science Direct, Ovid Medline, Scopus, PubMed and ProQuest databases from April to June 2015. Hand searching and reference of selected articles were used to improve article identification. Articles published after 1990 which had reported on efficacy/effectiveness of speed management strategies were included. Data were extracted using pre-defined extraction table. Results: Of the 803 retrieved articles, 22 articles were included in this review. Most of the included articles (63%) had before-after design and were done in European countries. Speed cameras, engineering schemes, intelligent speed adaption (ISA), speed limits and zones, vehicle activated sign and integrated strategies were the most common strategies reported in the literature. Various strategies had different effects on mean speed of the vehicles ranging from 1.6 to 10 km/h. Moreover, 8-65% and 11-71% reduction was reported in person injured accidents and fatal accidents, respectively as a result of employing various strategies. Conclusion: Literature revealed positive effects of various speed management strategies. Using various strategies was mostly dependent on road characteristics, driver’s attitude about the strategy as well as economic and technological capabilities of the country. Political support is considered as a main determinant in selecting speed management strategies.

  7. 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

  8. 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

  9. 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

  10. Management strategies for fibromyalgia

    Le Marshall KF; Littlejohn GO

    2011-01-01

    Kim Francis Le Marshall, Geoffrey Owen LittlejohnDepartments of Rheumatology and Medicine, Monash Medical Centre and Monash University, Victoria, AustraliaDate of preparation: 14 June 2011Clinical question: What are the effective, evidence-based strategies available for the management of fibromyalgia?Conclusion: There are a number of management strategies available with robust evidence to support their use in clinical practice.Definition: Fibromyalgia is a complex pain syndrome characterized ...

  11. 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

  12. 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

  13. 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.

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

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

  15. 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)

  16. 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

  17. 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)

  18. 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

  19. 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

  20. 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)

  1. 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.)

  2. 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.)

  3. 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...

  4. 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...

  5. 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

  6. 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...

  7. 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

  8. Remediation strategies for contaminated territories resulting from the Chernobyl accident

    The Directorate General for Environment of the European Commission has supported two projects on the issue of remediation strategies for contaminated territories resulting from the Chernobyl accident. The first one aimed at identifying and costing a set of additional countermeasures that would enable the reduction of the annual exposure of the inhabitants down to 1 mSv. The second one (still running) is developing a new rehabilitation approach based on the involvement of the local population in the decision taking process concerning the type of countermeasures to be applied (the ETHOS approach). (author)

  9. Severe accident mitigation strategy for the generation II PWRs in France. Some outcomes of the on-going periodic safety review of the French 1300 MWe PWR series

    The 3rd Periodic Safety Review of the French 1300 MWe PWRs series includes some modifications to increase their robustness in case of a severe accident. Their review is based on both deterministic and probabilistic approaches, keeping in mind that severe accidents frequencies and radiological consequences should be as low as reasonably practicable, severe accidents management strategies should be as safe as possible and the robustness of equipment used for severe accident management should be ensured. Consequently, the IRSN level 2 probabilistic safety assessment (L2 PSA) studies for the 1300 MWe reactors have been used to re-assess the results of the utility's L2 PSA and rank them to identify the containment failure modes contributing the most to the global risk. This ranking helped the review of plant modifications. Regarding strategies for accident management, the EDF management of water in the reactor cavity during a severe accident for the 1300 MWe PWRs is presented as well as the IRSN position on this strategy: this is an example where the optimal severe accident management strategy choice is not so easy to define. Regarding the robustness of equipment used for severe accident management, the interest of a diversification or redundancy of the French emergency filtered containment venting opening is one example among many others. (orig.)

  10. 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

  11. Formalizing a Debt Management Strategy

    Magnusson, Tomas I.

    2005-01-01

    In short, central government debt management can be defined as the process of establishing and executing a strategy in order to meet the debt management objectives. Undoubtedly, the development of the strategy is the most important debt management decision. Given the market constraints, it is the strategy document that decides on issues such as the level of exposure to foreign currency ris...

  12. 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

  13. 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

  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. Strategies for operator response in mitigating loss of containment heat removal accident scenarios

    In anticipation of the US Nuclear Regulatory Commission generic letter regarding accident management, the Boiling Water Reactor Owners' Group (BWROG) has commissioned the development of Accident Management Guidelines (AMGs). One outgrowth of the industry performance of individual plant examinations (IPES) is the development of more effective accident management guidance to prevent or mitigate the effects of severe accidents. The BWROG is determining a process for integrating these insights into a coherent format that can be implemented by BWR owners as part of accident management

  16. 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

  17. 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)

  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. 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

  20. 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.

  1. 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)

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

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

  3. 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

  4. Immunosuppressive strategies and management

    Shi-hui PAN

    2008-01-01

    Advances in immunosuppressive therapy have significantly improved short-term allograft and patient survival.However,chronic allograft failure,antibody mediated rejection,recurrent diseases and immunosuppressive drug associated adverse effects remain serious barriers to long-term survival and quality of life.New immunosuppressive agents and protocols are being evaluated to combat these problems.Importantly,clinicians must work to manage post-transplant complications and avoid complex medication regimens,which will potentiate drug interactions and non.compliance.Different organs have different immunogenicities and each recipient has a unique clinical and immunologic profile.The clinician must recognize these variations and customize the immunosuppressive regimens and treatment protocols based on the individual condition.The general principles of an individualized immunosuppressive protocol should take the following factors into account:organ type,donor and recipient characteristics,quality of the donor organ,recipienVs medical history,recipient's undedying disease,immunologic risk for acute rejection,potential co-morbidity related to immunosuppression,significant druginteractions,medication costs and patient compliance.In addition,the combination of immunosuppressive drugs must have a pharmacologic rationale to achieve the desired goal of suppressing the individual's immune system to render the patient tolerant to the allograft while minimizing co-morbidities.For the past few years,many clinical strategies have been applied in an attempt to improve graft survival or to reduce immunsuppressants induced side-effects.Specific protocols include steroid or CNI avoidance,minimization or withdraw,desensitization,and treatment for antibody mediated rejection,disease specific,and pediatric specific.The short-term outcomes from these different strategies are promising but the long-term results remain to be determined.Unfortunately,current immunosuppressive agents or strategies

  5. 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)

  6. 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

  7. 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.

  8. 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.

  9. 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)

  10. 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

  11. 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)

  12. 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.

  13. 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.)

  14. 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

  15. 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.)

  16. 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

  17. 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)

  18. 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

  19. Identification and assessment of containment and release management strategies

    Brookhaven National Laboratory, under the auspices of the U.S. Nuclear Regulatory Commission, is investigating accident management strategies which could help preserve containment integrity or minimize releases during a severe accident. The objective is to make use of existing plant systems and equipment in innovative ways to reduce the likelihood of containment failure or to mitigate the release of fission products to the environment if failure cannot be prevented. Many of these strategies would be implemented during the later stages of a severe accident (i.e., after vessel breach) and sizeable uncertainties exist regarding some of the phenomena involved. A majority of the strategies identified go well beyond existing procedures and often depend on the specific containment type. Strategies for all of the five different containments used in the U.S. are being considered: BWR Mark I, Mark II, and Mark III, as well as PWR ice condenser and large dry containments. Accident management strategies related to the in-vessel phase of a severe core melt accident are being dealt with under another NRC program. For each containment type the most likely challenges are identified and existing emergency guidelines and procedures are reviewed as to how they address these challenges

  20. 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

  1. 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

  2. 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)

  3. Identification and assessment of containment and release management strategies

    Lehner, J.R. (Dept. of Nuclear Energy, Brookhaven National Lab., Upton, NY (United States)); Lin, C.C. (Dept. of Nuclear Energy, Brookhaven National Lab., Upton, NY (United States)); Neogy, P. (Dept. of Nuclear Energy, Brookhaven National Lab., Upton, NY (United States))

    1993-08-01

    Brookhaven National Laboratory, under the auspices of the U.S. Nuclear Regulatory Commission, is investigating accident management strategies which could help preserve containment integrity or minimize the release of radioactivity during a severe accident in a nuclear reactor. The objective is to make use of existing plant systems and equipment in innovative ways to reduce the likelihood of containment failure or to mitigate the release of fission products to the environment if failure cannot be prevented. Many of these strategies would be implemented during the later stages of a severe accident, i.e. after the molten core penetrates the reactor vessel. Significant uncertainties exist regarding some of the phenomena involved with this phase of a severe accident. The identification and assessment process for containment and release strategies is described, and some insights derived from its application to a BWR Mark I plant are presented. A station blackout accident for this kind of plant is considered. The challenges encountered are identified and existing emergency guidelines are reviewed, where needed and when possible, new strategies are devised. The feasibility and effectiveness of these new strategies are assessed, making due allowances for the complicated phenomena and associated uncertainties involved. Both beneficial and adverse effects of the suggested strategies are considered. (orig.)

  4. 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.

  5. 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

  6. 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

  7. Effectiveness of talent management strategies

    Bethke-Langenegger, Pamela; Mahler, Philippe (collab.); Staffelbach, Bruno

    2011-01-01

    This paper investigates the effects of different types of talent management strategies on organisational performance. We introduce four different strategies and show how they affect organisational performance. For this purpose, we use a particularly detailed dataset of 138 Swiss companies. We find that talent management focusing on retaining and developing talents as job satisfaction, motivation, commitment and trust in leaders. Moreover, talent management practices with a strong focus on cor...

  8. 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

  9. A framework for assessing hydrogen management strategies involving multiple decisions

    An accident management framework consisting of multiple and sequential decisions is developed and applied to a hydrogen control strategy for a reference plant. The compact influence diagrams including multiple decisions are constructed and evaluated with MAAP4 calculations. Each decision variable, represented by a node in the influence diagrams, has an uncertainty distribution. Using the values from the IPE (Individual Plant Examinations) report for the reference plant (UCN 3 and 4), the hydrogen control and accident management strategies are assessed. In this paper, a problem with two decisions is modeled for a simple illustration of the process involved. One decision is whether or not to actuate igniters at the time of core uncovery. Another decision is whether or not to turn on the containment sprays. We chose a small-break loss-of-coolant accident (LOCA) sequence, which was one of the dominant accident sequences in the reference plant. The framework involves the modeling of the decision problem by using decision-making tools, data analysis, and the MAAP4 calculations. It is shown that the proposed framework with a new measure for assessing hydrogen control is flexible enough to be applied to various accident management strategies. (author)

  10. 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.)

  11. Marketing Management and Strategy

    with practical information from the cases, the reader is introduced to issues relating to marketing strategy formulation, managerial actions in designing and implementing marketing decisions, as well as the operational contexts within which these actions are taken. The book is essential reading for both...... undergraduate and graduate students in marketing, international strategy and international business who require an understanding of African business...

  12. 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

  13. 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

  14. 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.

  15. Hydrogen management strategies using the igniters and recombiners

    Hydrogen, which is generated by cladding oxidation, can diffuse to all subcompartments in the containment so that high hydrogen concentration may be induced only locally. If the hydrogen concentration in containment reaches to the flammability limit and enough hydrogen is allowed to build up, the hydrogen can ignite and cause a spike in containment pressure that exceeds the containment ultimate design pressure, thereby failing the containment. It is thus important to analyze the accident sequence inducing high hydrogen concentration locally and to develop adequate control and management strategies. In this paper, a framework for evaluating hydrogen control and management strategies involving multiple decisions is presented. The compact influence diagrams including multiple decisions are constructed and evaluated with the MAAP4 calculations. Each decision variable, represented by a node in the influence diagrams, has an uncertainty distribution. Using the values from standard safety analysis report of the reference plant, Advanced Power Reactor 1400MWe (APR1400), the hydrogen control and accident management strategies are assessed. The strategies are ranked with respect to a new measure in terms of hydrogen concentration. The MAAP4 code calculations are performed to generate data for hydrogen concentrations and to identify the important severe accident phenomena in containment for the decision-making analysis as well. In this paper, a problem with two decisions is modeled for a simplistic illustration. One decision is whether or not to actuate the igniters at the time of core uncovery; another decision is to use the passive autocatalytic recombiners (PARs). We chose a small-break loss-of-coolant accident (LOCA) sequence, which was one of the dominant accident sequences in the reference plant, as the reference case. The framework ling of the decision problem by using the decision-making tools, data analysis, and the MAAP4 calculations. It is shown that the proposed

  16. Accident Monitoring Systems for Nuclear Power Plants

    In the Fukushima Daiichi accident, the instrumentation provided for accident monitoring proved to be ineffective for a combination of reasons. The accident has highlighted the need to re-examine criteria for accident monitoring instrumentation. This publication covers all relevant aspects of accident monitoring in NPPs. The critical issues discussed reflect the lessons learned from the Fukushima Daiichi accident, involve accident management and accident monitoring strategies for nuclear power plants, selection of plant parameters for monitoring plant status, establishment of performance, design, qualification, display, and quality assurance criteria for designated accident monitoring instrumentation, and design and implementation considerations. Technology needs and techniques for accident monitoring instrumentation are also addressed

  17. 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.)

  18. 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

  19. 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.

  20. 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

  1. 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

  2. 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.)

  3. 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.

  4. 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

  5. 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)

  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. Strategy of fuel management

    The management of nuclear fuels in PWR type reactors has been adapted to improve the safety and the competitiveness of brackets. The economic optimum, at the park level, depends on many parameters, variable with time and in function of them, we favour the annual campaigns and the economy won on the cost of cycle, or long campaigns with benefit on availability. The reduction of the number of stopping improves the availability, limits the doses integrated by the personnel of intervention and reduces the number of incidents during the stopping. An other determining factor is connected to the policy of closed cycle with the the principle of equality between the reprocessing flux and the valorization of reprocessed fuels: plutonium and reprocessed uranium. The progress of fuel have allowed significant improvements in the managements of cores. With the safety, the aim is also to keep if not improve the competitiveness of the Nuclear park by valorizing the matter coming from reprocessing. (N.C.)

  8. Spent fuel management strategies

    Nuclear fuel cycle is divided into two sections; front end and back end of the fuel cycle. Front end of the fuel cycle, which covers all the activities of the fuel cycle before the fuel goes into the reactor has better developed and well-defined technologies. For storage of the spent fuel which are subjects of the back end of the fuel cycle, the waste management policies are not so well defined. There are three approaches that exist today for management of spent fuel. 1. For once through or open fuel cycles direct disposal of spent fuel in a deep geological repository, 2. For closed fuel cycles reprocessing of spent fuel and recycling of the recovered plutonium and uranium in new mixed oxide (MOX) fuels, 3. The spent fuel is placed in long term interim storage pending a decision as to its ultimate reprocessing or disposal. There are so large scale geological repositories for the final disposal of spent fuel in operation. Studies on suitable site selection, design, construction and licensing take about 30-40 years. Reprocessing, on the other hand, produces plutonium and is therefore under close inspection because of the Non Proliferation Treaty. Today more countries are delaying their final decision about the spent fuel management approach and using the long term interim storage approach

  9. 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

  10. 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

  11. Strategies for password management

    2013-01-01

    Passwords are a critical part of information and network security. Passwords serve as primary authentication method to protect user accounts but a poorly chosen password, if compromised, could put the entire network at risk. Many users do not understand why good passwords and password management are important for information systems. “We‟re secure! We use passwords!” How many of us have heard this claim? Or even – “We‟re secure! We have a password policy!” Using a password or having a passwor...

  12. 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)

  13. 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

  14. 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

  15. 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

  16. 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

  17. Strategies for managing margins.

    2012-08-01

    Potential Medicare and Medicaid reimbursement cuts have made it critical for home health agencies to manage their gross and net operating profit margins. Agencies need to develop tools to analyze their margins and make sure they are following best practices. Try as you may, your agency might still face the question, "Why am I not meeting my budget?" Get some answers in this session from David Berman and Andrea L. Devoti. Berman is a principal at Simione Healthcare Consultants in Hamden, CT, where he is responsible for merchant acquisitions, business valuation due diligence, and oversight of the financial monitor benchmarking tool besides serving as interim chief financial officer. Devoti is chairman of the NAHC board and President & CEO of Neighborhood Health Visiting Nurse Association in West Chester PA. PMID:23074756

  18. 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)

  19. 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

  20. Managing incontinence: women's normalizing strategies.

    Skoner, M M; Haylor, M J

    1993-01-01

    Women's strategies for managing urinary incontinence were examined in a grounded-theory study. The women's basic social concern was dealing with incontinence in a manner that enabled them to feel normal. Feeling normal meant being able to do what they wanted to do and needed to do to have a normal life-style as they perceived it. This goal was accomplished by normalizing incontinence and its management. Normalization was achieved by directing its course through self-management, accounting for it in terms of personal history and life experiences, and delaying medical counsel. These strategies are described. The findings provide fresh insights about women's response to incontinence and their practice of self-managing its consequences. PMID:8138472

  1. 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

  2. Strategie podniku a management jakosti

    BOČKOVÁ, Jiřina

    2012-01-01

    The aim of the thesis was to analyze the selected project management company, including a review of the objectives of the strategy. Market environment has been investigated and followed the SWOT analysis, from which I draw the appropriate consequences. The result landed proposed solution in the form of an investment plan. The development of a Project Management Plan that defines the philosophy of the investment project, an integral part of the timetable for the progress of work and cost analy...

  3. Passive Decay Heat Removal Strategy of Integrated Passive Safety System (IPSS) for SBO-combined Accidents

    Kim, Sang Ho; Chang, Soon Heung; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-10-15

    The weak points of nuclear safety would be in outmoded nuclear power plants like the Fukushima reactors. One of the systems for the safety enhancement is integrated passive safety system (IPSS) proposed after the Fukushima accidents. It has the five functions for the prevention and mitigation of a severe accident. Passive decay heat removal (PDHR) strategy using IPSS is proposed for coping with SBO-combined accidents in this paper. The two systems for removing decay heat before core-melt were applied in the strategy. The accidents were simulated by MARS code. The reference reactor was OPR1000, specifically Ulchin-3 and 4. The accidents included loss-of-coolant accidents (LOCA) because the coolant losses could be occurred in the SBO condition. The examples were the stuck open of PSV, the abnormal open of SDV and the leakage of RCP seal water. Also, as LOCAs with the failure of active safety injection systems were considered, various LOCAs were simulated in SBO. Based on the thermal hydraulic analysis, the probabilistic safety analysis was carried out for the PDHR strategy to estimate the safety enhancement in terms of the variation of core damage frequency. AIMS-PSA developed by KAERI was used for calculating CDF of the plant. The IPSS was applied in the PDHR strategy which was developed in order to cope with the SBO-combined accidents. The estimation for initiating SGGI or PSIS was based on the pressure in RCS. The simulations for accidents showed that the decay heat could be removed for the safety duration time in SBO. The increase of safety duration time from the strategy provides the increase of time for the restoration of AC power.

  4. 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

  5. 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.)

  6. 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.

  7. Conflict Management Strategies in Workplace

    黄玉霞

    2011-01-01

    As we all know, it is inevitable to be confronted with verbal aggressiveness by employees, peers, and supervisors in the workplace. In order to avoid these conflict with others in the future workplace, this paper is to discuss about the management strategies dealing with these conflicts.

  8. 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)

  9. Management strategies for encouraging creativity.

    Preston, P

    1998-01-01

    Change, chaos, and uncertainty touch every part of every institution. The laboratory is not immune. Managers content to continue on their familiar path soon will find themselves bypassed. To meet today's challenges, directors of technical operations, laboratory directors, team leaders, and coordinators need plenty of creativity--from everyone on their staff. It is no longer just "nice" to improve group output and problem-solving skills while staying within a "shoestring" budget. It is absolutely necessary. In this article, we explore strategies laboratory managers can use to tap the creative potential and commitment of their people. These strategies work. Whether it involves using humor, creating "idea centers," or "deconstructing the bureaucracy," the goal is the same: to encourage clinical managers to think beyond their technical and managerial experience. The examples in this article may not suit the needs, situations, or tastes of all laboratory managers. They are "food for thought." The concepts and strategies these examples illustrate are every laboratory manager's keys to adapting successfully to future challenges. PMID:10181491

  10. RADIATION ACCIDENTS: EXPERIENCE OF MEDICAL PROTECTION AND MODERN STRATEGY OF PHARMACOLOGICAL MAINTENANCE

    A. N. Grebenyuk

    2012-01-01

    Full Text Available Experience of medical protection at radiation accidents is analyzed. It is shown, that medicines that have been in the arsenal of medical service during the liquidation of consequences of the Chernobyl nuclear power plant accident satisfied their predestination in a whole and were rather effective for radiation protection. The modern strategy of pharmacological maintenance based on use of means and methods, allowing to keeping a life, health and professional serviceability of people in conditions of amazing action of a complex of factors of radiation accidents, is submitted.

  11. RADIATION ACCIDENTS: EXPERIENCE OF MEDICAL PROTECTION AND MODERN STRATEGY OF PHARMACOLOGICAL MAINTENANCE

    A. N. Grebenyuk; V. I. Legeza; V. V. Zatsepin

    2012-01-01

    Experience of medical protection at radiation accidents is analyzed. It is shown, that medicines that have been in the arsenal of medical service during the liquidation of consequences of the Chernobyl nuclear power plant accident satisfied their predestination in a whole and were rather effective for radiation protection. The modern strategy of pharmacological maintenance based on use of means and methods, allowing to keeping a life, health and professional serviceability of people in condit...

  12. 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.)

  13. Study of Containment Vent Strategies During Severe Accident Progression for the CANDU6 Plant

    Jin, Youngho; Ahn, K. I. [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    In March, 2011, Fukushima daichi nuclear power plants experienced a long term station blackout. Severe core damage occurred and a large amount of radioactive materials are released outside of the plants. After this terrible accident Nuclear Safety and Security Commission (NSSC) enforced to increase nuclear safety for all operating plants in Korea. To increase plant safety, both hardware reinforcement and software improvement are encouraged. Hardware reinforcement includes the preparation of the external water injection paths to the RCS and the spent fuel pool, a filtered containment venting system (CFVS), and AC power generating truck. Software improvement includes the increase of the effectiveness of the severe accident management guidance (SAMG) and plant staff training. To comply with NSSC's request, Wolsong Unit 1 has fulfilled the hardware reinforcement including the installation of a CFVS and started the extension of a SAMG to the low power and shutdown operation mode. Current SAMG deals accident occurred during full power operation only. The CFVS is designed to open and to close isolation valves manually. It does not require AC power. The operation of the CFVS prevents the reactor containment building failure due to the over-pressurization but it may release radioactive materials out of the reactor containment building. This paper discusses the radiological source terms for the containment vent strategy during severe accident progression which occurred during shutdown operation mode. This work is a part of the development of shutdown SAMG.. The CFVS is an effective means to control the containment pressure when the local air coolers are unavailable. Radioactive materials may release through the CFVS, but their amounts are reduced significantly. The alternative means, i.e., containment vent through the ventilation system which does not have an effective filter, is not a good choice to control the containment condition. It can maintain the containment

  14. Study of Containment Vent Strategies During Severe Accident Progression for the CANDU6 Plant

    In March, 2011, Fukushima daichi nuclear power plants experienced a long term station blackout. Severe core damage occurred and a large amount of radioactive materials are released outside of the plants. After this terrible accident Nuclear Safety and Security Commission (NSSC) enforced to increase nuclear safety for all operating plants in Korea. To increase plant safety, both hardware reinforcement and software improvement are encouraged. Hardware reinforcement includes the preparation of the external water injection paths to the RCS and the spent fuel pool, a filtered containment venting system (CFVS), and AC power generating truck. Software improvement includes the increase of the effectiveness of the severe accident management guidance (SAMG) and plant staff training. To comply with NSSC's request, Wolsong Unit 1 has fulfilled the hardware reinforcement including the installation of a CFVS and started the extension of a SAMG to the low power and shutdown operation mode. Current SAMG deals accident occurred during full power operation only. The CFVS is designed to open and to close isolation valves manually. It does not require AC power. The operation of the CFVS prevents the reactor containment building failure due to the over-pressurization but it may release radioactive materials out of the reactor containment building. This paper discusses the radiological source terms for the containment vent strategy during severe accident progression which occurred during shutdown operation mode. This work is a part of the development of shutdown SAMG.. The CFVS is an effective means to control the containment pressure when the local air coolers are unavailable. Radioactive materials may release through the CFVS, but their amounts are reduced significantly. The alternative means, i.e., containment vent through the ventilation system which does not have an effective filter, is not a good choice to control the containment condition. It can maintain the containment

  15. 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.

  16. 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

  17. The rehabilitation strategies in agriculture in the long term after the Chernobyl NPP accident

    The experience gained in the aftermath of the severe radiation accidents shows that in the case of large-scaled radionuclide contamination the limitation of internal radiation doses to people by means of restoration of agricultural lands is more realistic than reduction of levels of external irradiation. Therefore, the problems connected with the optimal restoration strategies of agricultural land subjected to radioactive contamination after the Chernobyl accident are of crucial importance. The justification of the approach for the estimation of the effectiveness of countermeasure strategies in the long term after the Chernobyl accident, based on the classification of farms by contamination density and risk of the exceeding of radiological standards, restricting the use of agricultural products, is presented. For each class of the farms the ranking of rehabilitation options and the time periods when their application would be of importance are given. Comparative analysis of the rehabilitation strategies, which are different in their effectiveness and cost, is provided. (author)

  18. 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.

  19. 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

  20. 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

  1. 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

  2. Methodology for rehabilitation strategies evaluation aid after an accident: application and results

    Post-accidental management is a matter for an optimisation process in a multi-criteria context, which has to combine quantitative parameters as well as qualitative parameters. In order to fit with this problematic, a methodological work has been realised at IPSN, in the context of the Becquerel national exercise, which has Simulated a Borax type accident on an experimental reactor. The methodology consists in making a systematic evaluation of indicators in order to provide inter-comparing data for many rehabilitation strategies. Thus, each strategy is evaluated on the one hand in terms of radiological benefit and on the other hand, in terms of incidences such as duration of realisation, materials and human means, workers doses, amount of generated wastes. In this context, calculations highlighted the importance of counter-measures aiming at reducing the external exposure, which is the dominating exposure pathway. The counter-measures aimed at reducing the dose by ingestion lead to a reduction of specific activity are then useful mainly in regard to the limits of commercialization. This work is a first and useful contribution to the clarification of the post-accidental rehabilitation problem. It has also put the light on three important needs: formalization of counter-measures and gathering into a data base, adaptation of calculation tools to make their use easier, and improvement of the knowledge of the operational considerations for the interventions. (authors)

  3. 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

  4. Reactor accidents. Public health strategies and their medical implications

    In a reactor accident with release of radioactivity, the major potential public health risks are likely to result from radioiodines, particularly iodine 131, which can be readily volatilized and dispersed. They are rapidly absorbed and concentrated by the thyroid, which could result in substantial thyroidal radiation. Although some forms of radiation can cause thyroid cancer in humans, 131I has not been shown to do so, and 40 years of safe experience with this radionuclide in routine clinical thyroid testing and treatment has been reassuring. Nevertheless, since 131I has been shown to cause thyroid neoplasms in animals, efforts to minimize unwarranted exposure seem advisable. Potassium iodide, administered at an appropriate time, will effectively block thyroid uptake, but it has potential toxicity and may be difficult to distribute effectively and safely on a large scale in an emergency. Evaluation of the risks and benefits of potassium iodide use is essential to establishing sound public health policy but awaits additional scientific information

  5. Nuclear accidents. Optimisation in the planning of strategies for the introduction of precautionary countermeasures

    The profile of a nuclear accident commences with an initiator and may progress, through a number of stages, to a major release to the environment. Emergency planners, in drawing up contingency arrangements must take into account the range of possible accidents, the available countermeasures and their effectiveness in achieving averted dose for comparison against a range of national or international intervention levels. When precautionary countermeasures are considered it is often difficult to demonstrate that the initial response proposed provides for optimised protection. This paper aims to provide a simple method which may be used as an input to deciding optimum countermeasure strategies and planning zones. The method requires a knowledge of the frequencies and consequences of a representative range of accident sequences together with details of suitable warning times and release durations for each sequence. Local knowledge of the likely effectiveness of a countermeasure is also required in order to estimate the averted dose benefit for a given implementation strategy. The method takes into account the likely development of each sequence and its frequency weighting. Finally, a site specific assessment of the disbenefit of implementation of a countermeasure is incorporated so that the net benefit for each strategy may be derived as a function of distance from the accident site. The paper develops an example showing how the net benefit of a range of strategies may be compared leading to a choice of optimum strategy within a range of possible planning zones. (author)

  6. 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)

  7. 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

  8. 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...

  9. 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...

  10. 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)