WorldWideScience

Sample records for accident management measures

  1. Accident management

    International Nuclear Information System (INIS)

    Lutz, R.J.; Monty, B.S.; Liparulo, N.J.; Desaedeleer, G.

    1989-01-01

    The foundation of the framework for a Severe Accident Management Program is the contained in the Probabilistic Safety Study (PSS) or the Individual Plant Evaluations (IPE) for a specific plant. The development of a Severe Accident Management Program at a plant is based on the use of the information, in conjunction with other applicable information. A Severe Accident Management Program must address both accident prevention and accident mitigation. The overall Severe Accident Management framework must address these two facets, as a living program in terms of gathering the evaluating information, the readiness to respond to an event. Significant international experience in the development of severe accident management programs exist which should provide some direction for the development of Severe Accident Management in the U.S. This paper reports that the two most important elements of a Severe Accident Management Program are the Emergency Consultation process and the standards for measuring the effectiveness of individual Severe Accident Management Programs at utilities

  2. Accident management for severe accidents

    International Nuclear Information System (INIS)

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

    1988-01-01

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

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

    International Nuclear Information System (INIS)

    Tusheva, P.; Schaefer, F.; Rohde, U.; Reinke, N.

    2009-01-01

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

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

    International Nuclear Information System (INIS)

    2002-01-01

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

  5. Application of the severe accident code ATHLET-CD. Modelling and evaluation of accident management measures (Project WASA-BOSS)

    Energy Technology Data Exchange (ETDEWEB)

    Wilhelm, Polina; Jobst, Matthias; Kliem, Soeren; Kozmenkov, Yaroslav; Schaefer, Frank [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Div. Reactor Safety

    2016-07-01

    The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. Numerical analyses are used to investigate the accident progression and the complex physical phenomena during the core degradation phase, as well as to evaluate the effectiveness of possible countermeasures in the preventive and mitigative domain [1, 2]. The presented analyses have been performed with the computer code ATHLET-CD developed by GRS [3, 4].

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

    International Nuclear Information System (INIS)

    Yu, Donghan

    2002-01-01

    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

  7. Management of severe accidents

    International Nuclear Information System (INIS)

    Jankowski, M.W.

    1987-01-01

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

  8. Management of severe accidents

    International Nuclear Information System (INIS)

    Jankowski, M.W.

    1988-01-01

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

  9. Accident management information needs

    International Nuclear Information System (INIS)

    Hanson, D.J.; Ward, L.W.; Nelson, W.R.; Meyer, O.R.

    1990-04-01

    In support of the US Nuclear Regulatory Commission (NRC) Accident Management Research Program, a methodology has been developed for identifying the plant information needs necessary for personnel involved in the management of an accident to diagnose that an accident is in progress, select and implement strategies to prevent or mitigate the accident, and monitor the effectiveness of these strategies. This report describes the methodology and presents an application of this methodology to a Pressurized Water Reactor (PWR) with a large dry containment. A risk-important severe accident sequence for a PWR is used to examine the capability of the existing measurements to supply the necessary information. The method includes an assessment of the effects of the sequence on the measurement availability including the effects of environmental conditions. The information needs and capabilities identified using this approach are also intended to form the basis for more comprehensive information needs assessment performed during the analyses and development of specific strategies for use in accident management prevention and mitigation. 3 refs., 16 figs., 7 tabs

  10. Accident management information needs

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, D.J.; Ward, L.W.; Nelson, W.R.; Meyer, O.R. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1990-04-01

    In support of the US Nuclear Regulatory Commission (NRC) Accident Management Research Program, a methodology has been developed for identifying the plant information needs necessary for personnel involved in the management of an accident to diagnose that an accident is in progress, select and implement strategies to prevent or mitigate the accident, and monitor the effectiveness of these strategies. This report describes the methodology and presents an application of this methodology to a Pressurized Water Reactor (PWR) with a large dry containment. A risk-important severe accident sequence for a PWR is used to examine the capability of the existing measurements to supply the necessary information. The method includes an assessment of the effects of the sequence on the measurement availability including the effects of environmental conditions. The information needs and capabilities identified using this approach are also intended to form the basis for more comprehensive information needs assessment performed during the analyses and development of specific strategies for use in accident management prevention and mitigation. 3 refs., 16 figs., 7 tabs.

  11. Measures against nuclear accidents

    International Nuclear Information System (INIS)

    1992-01-01

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

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

    International Nuclear Information System (INIS)

    Guentay, S.

    2010-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-10-15

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

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

    International Nuclear Information System (INIS)

    Schwarz, M.; Clement, B.; Jones, A.V.

    2001-01-01

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

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

    International Nuclear Information System (INIS)

    Lathrop, J.

    1989-01-01

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

  16. Accident and emergency management

    International Nuclear Information System (INIS)

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

    1988-06-01

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

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

    International Nuclear Information System (INIS)

    Bermejo, J.M.; Van Goethem, G.

    1998-01-01

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

  18. Communication and Monitoring- Necessary Processes for Managing and Measuring Conflicts, Absenteeism, Fluctuation and Work Accidents

    Directory of Open Access Journals (Sweden)

    Tănăsescu Dorina Antoneta

    2015-06-01

    Full Text Available In any organizations, between employees, between employees and management are inevitabily appearing conflicts, absenteeism, personnel fluctuation and work accidents (CAFA factors. To diminish the impact of these social dysfunctions, organizations must pay attention to the effective managing and monitoring using performance indicators and an effective communication process and implementing a good plan of evaluation.

  19. Joint research project WASA-BOSS: Further development and application of severe accident codes. Assessment and optimization of accident management measures. Project B: Accident analyses for pressurized water reactors with the application of the ATHLET-CD code

    International Nuclear Information System (INIS)

    Jobst, Matthias; Kliem, Soeren; Kozmenkov, Yaroslav; Wilhelm, Polina

    2017-02-01

    Within the framework of the project an ATHLET-CD input deck for a generic German PWR of type KONVOI has been created. This input deck was applied to the simulation of severe accidents from the accident categories station blackout (SBO) and small-break loss-of-coolant accidents (SBLOCA). The complete accident transient from initial event at full power until the damage of reactor pressure vessel (RPV) is covered and all relevant severe accident phenomena are modelled: start of core heat up, fission product release, melting of fuel and absorber material, oxidation and release of hydrogen, relocation of molten material inside the core, relocation to the lower plenum, damage and failure of the RPV. The model has been applied to the analysis of preventive and mitigative accident management measures for SBO and SBLOCA transients. Therefore, the measures primary side depressurization (PSD), injection to the primary circuit by mobile pumps and for SBLOCA the delayed injection by the cold leg hydro-accumulators have been investigated and the assumptions and start criteria of these measures have been varied. The time evolutions of the transients and time margins for the initiation of additional measures have been assessed. An uncertainty and sensitivity study has been performed for the early phase of one SBO scenario with PSD (until the start of core melt). In addition to that, a code -to-code comparison between ATHLET-CD and the severe accident code MELCOR has been carried out.

  20. EPRI research on accident management

    International Nuclear Information System (INIS)

    Oehlberg, R.N.; Chao, J.

    1991-01-01

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

  1. Accident management insights after the Fukushima Daiichi NPP accident

    International Nuclear Information System (INIS)

    Degueldre, Didier; Viktorov, Alexandre; Tuomainen, Minna; Ducamp, Francois; Chevalier, Sophie; Guigueno, Yves; Tasset, Daniel; Heinrich, Marcus; Schneider, Matthias; Funahashi, Toshihiro; Hotta, Akitoshi; Kajimoto, Mitsuhiro; Chung, Dae-Wook; Kuriene, Laima; Kozlova, Nadezhda; Zivko, Tomi; Aleza, Santiago; Jones, John; McHale, Jack; Nieh, Ho; Pascal, Ghislain; ); Nakoski, John; Neretin, Victor; Nezuka, Takayoshi; )

    2014-01-01

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

  2. Severe accident management guidelines

    International Nuclear Information System (INIS)

    Uhle, Jennifer

    2014-01-01

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

  3. Accident management on french PWRS

    International Nuclear Information System (INIS)

    Queniart, D.

    1990-06-01

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

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

    International Nuclear Information System (INIS)

    Jang, S. H.; Kim, H. G.; Jang, H. S.; Moon, S. K.; Park, J. U.

    1993-12-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Jang, S H; Kim, H G; Jang, H S; Moon, S K; Park, J U [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    1993-12-15

    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.

  6. Accident management approach in Armenia

    International Nuclear Information System (INIS)

    Ghazaryan, K.

    1999-01-01

    In this lecture the accident management approach in Armenian NPP (ANPP) Unit 2 is described. List of BDBAs had been developed by OKB Gydropress in 1994. 13 accident sequences were included in this list. The relevant analyses had been performed in VNIIAES and the 'Guidelines on operator actions for beyond design basis accident (BDBA) management at ANPP Unit 2' had been prepared. These instructions are discussed

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

    International Nuclear Information System (INIS)

    Takahira, Shiro

    2014-01-01

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

  8. WASA-BOSS. Development and application of Severe Accident Codes. Evaluation and optimization of accident management measures. Subproject F. Contributions to code validation using BWR data and to evaluation and optimization of accident management measures. Final report

    International Nuclear Information System (INIS)

    Di Marcello, Valentino; Imke, Uwe; Sanchez Espinoza, Victor

    2016-09-01

    The exact knowledge of the transient course of events and of the dominating processes during a severe accident in a nuclear power station is a mandatory requirement to elaborate strategies and measures to minimize the radiological consequences of core melt. Two typical experiments using boiling water reactor assemblies were modelled and simulated with the severe accident simulation code ATHLET-CD. The experiments are related to the early phase of core degradation in a boiling water reactor. The results reproduce the thermal behavior and the hydrogen production due to oxidation inside the bundle until relocation of material by melting. During flooding of the overheated assembly temperatures and hydrogen oxidation are under estimated. The deviations from the experimental results can be explained by the missing model to simulate bore carbide oxidation of the control rods. On basis of a hypothetical loss of coolant accident in a typical German boiling water reactor the effectivity of flooding the partial degraded core is investigated. This measure of mitigation is efficient and prevents failure of the reactor pressure vessel if it starts before molten material is relocated into the lower plenum. Considerable amount of hydrogen is produced by oxidation of the metallic components.

  9. Severe accident management. Prevention and Mitigation

    International Nuclear Information System (INIS)

    1992-01-01

    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

  10. Management of accident risks

    International Nuclear Information System (INIS)

    Compes, P.C.

    1987-01-01

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

  11. [Early management of cerebrovascular accidents].

    Science.gov (United States)

    Libot, Jérômie; Guillon, Benoit

    2013-01-01

    A cerebrovascular accident requires urgent diagnosis and treatment.The management of a stroke must be early and adapted in order to improve the overall clinical outcome and lower the risk of mortality.

  12. The IAEA Accident Management Programme

    International Nuclear Information System (INIS)

    Kabanov, L.; Jankowski, M.; Mauersberger, H.

    1993-01-01

    Accident prevention and mitigation programmes and the Emergency Response System (ERS) are important elements of the Agency's activities in the area of nuclear power plant (NPP) safety. Safety Codes and Guides on siting, design, quality assurance and the operation of NPPs have been produced and are used by NPP operating organizations. Nuclear safety evaluation services are provided by the IAEA. The Emergency Response System and the International Nuclear Event Scale (INES) have been developed. The framework for the development of an accident management programme has been set up. The main goal is to develop an Accident Management Manual to provide a systematic, structured approach to the development and implementation of an accident management programme at NPPs. An outline of the Manual has been distributed and the first draft is available. The component parts are: Co-ordinated research programmes (CRPs) on severe accident management and containment behaviour; the use of vulnerability analysis; mitigation of the effects of hydrogen, and generic symptom oriented emergency operating procedures. The IAEA provides guidance by the dissemination of information on methods for accident management; collates information on approaches in this field in different organizations and countries; and arranges exchange of experience and the promulgation of knowledge through the training of NPP managers and senior technical staff. (orig.)

  13. The IAEA Accident Management Programme

    Energy Technology Data Exchange (ETDEWEB)

    Kabanov, L.; Jankowski, M.; Mauersberger, H. (International Atomic Energy Agency, Vienna (Austria))

    1993-02-01

    Accident prevention and mitigation programmes and the Emergency Response System (ERS) are important elements of the Agency's activities in the area of nuclear power plant (NPP) safety. Safety Codes and Guides on siting, design, quality assurance and the operation of NPPs have been produced and are used by NPP operating organizations. Nuclear safety evaluation services are provided by the IAEA. The Emergency Response System and the International Nuclear Event Scale (INES) have been developed. The framework for the development of an accident management programme has been set up. The main goal is to develop an Accident Management Manual to provide a systematic, structured approach to the development and implementation of an accident management programme at NPPs. An outline of the Manual has been distributed and the first draft is available. The component parts are: Co-ordinated research programmes (CRPs) on severe accident management and containment behaviour; the use of vulnerability analysis; mitigation of the effects of hydrogen, and generic symptom oriented emergency operating procedures. The IAEA provides guidance by the dissemination of information on methods for accident management; collates information on approaches in this field in different organizations and countries; and arranges exchange of experience and the promulgation of knowledge through the training of NPP managers and senior technical staff. (orig.).

  14. Enhanced safety features of CHASHMA NPP UNIT-2 to encounter selected severe accidents, various challenges involved to prove the adequacy of severe accidents prevention/mitigation measures and to write management guidelines with one possible solution to these challenges

    International Nuclear Information System (INIS)

    Iqbal, Z.; Minhaj, A.

    2007-01-01

    This paper describes enhanced safety features of Chashma Nuclear Power Plant Unit-2 (C-2), a 325 MWe PWR to encounter selected severe accidents and discusses various challenges involved to prove the adequacy of severe accidents encountering measures and to write severe accident management guidelines (SAMGs) in compliance with the recently introduced national regulations based on the new IAEA nuclear safety standards. C-2 is being built by China National Nuclear Corporation (CNNC) for Pakistan Atomic Energy Commission (PAEC). Its twin, Unit-1 (C-1) also a 325 MWe PWR, was commissioned in 2000. Nuclear power safety with reference to severe accidents should be treated as a global issue and therefore the developed countries should include the people of developing countries in nuclear power industry's various severe accidents based research and development programs. The implementation of this idea may also deliver few other useful and mutually beneficial byproducts. (author)

  15. NPP Krsko Severe Accident Management Guidelines Upgrade

    International Nuclear Information System (INIS)

    Mihalina, Mario; Spalj, Srdjan; Glaser, Bruno; Jalovec, Robi; Jankovic, Gordan

    2014-01-01

    Nuclear Power Plant Krsko (NEK) has decided to take steps for upgrade of safety measures to prevent severe accidents, and to improve the means to successfully mitigate their consequences. The content of the program for the NEK Safety Upgrade is consistent with the nuclear industry response to Fukushima accident, which revealed many new insights into severe accidents. Therefore, new strategies and usage of new systems and components should be integrated into current NEK Severe Accident Management Guidelines (SAMG's). SAMG's are developed to arrest the progression of a core damage accident and to limit the extent of resulting releases of fission products. NEK new SAMG's revision major changes are made due to: replacement of Electrical Recombiners by Passive Autocatalytic Recombiners (PARs) and the installation of Passive Containment Filtered Vent System (PCFV); to handle a fuel damage situation in Spent Fuel Pool (SFP) and to assess risk of core damage situation during shutdown operation. (authors)

  16. Uncertainties and severe-accident management

    International Nuclear Information System (INIS)

    Kastenberg, W.E.

    1991-01-01

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

  17. Dosimetric management during a criticality accident

    International Nuclear Information System (INIS)

    Lebaron-Jacobs, L.; Fottorino, R.; Racine, Y.; Miele, A.; Barbry, F.; Briot, F.; Distinguin, S.; Le Goff, J.P.; Berard, P.; Boisson, P.; Cavadore, D.; Lecoix, G.; Persico, M.H.; Rongier, E.; Challeton-De Vathaire, C.; Medioni, R.; Voisin, P.; Exmelin, L.; Flury-Herard, A.; Gaillard-Lecanu, E.; Lemaire, G.; Gonin, M.; Riasse, C.

    2008-01-01

    A working group from health occupational and clinical biochemistry services on French sites has issued essential data sheets on the guidelines to follow in managing the victims of a criticality accident. Since the priority of the medical management after a criticality accident is to assess the dose and the distribution of dose, some dosimetric investigations have been selected in order to provide a prompt response and to anticipate the final dose reconstruction. Comparison exercises between clinical biochemistry laboratories on French sites were carried out to confirm that each laboratory maintained the required operational methods for hair treatment and the appropriate equipment for 32 P activity in hair and 24 Na activity in blood measurements, and to demonstrate its ability to rapidly provide neutron dose estimates after a criticality accident. As a result, a relation has been assessed to estimate the dose and the distribution of dose according to the neutron spectrum following a criticality accident. (authors)

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

    International Nuclear Information System (INIS)

    Abe, Kiyoharu.

    1995-05-01

    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)

  19. Application of the accident management information needs methodology to a severe accident sequence

    International Nuclear Information System (INIS)

    Ward, L.W.; Hanson, D.J.; Nelson, W.R.; Solberg, D.E.

    1989-01-01

    The U.S. Nuclear Regulatory Commission is conducting an accident management research program that emphasizes the use of severe accident research to enhance the ability of plant operating personnel to effectively manage severe accidents. Hence, it is necessary to ensure that the plant instrumentation and information systems adequately provide this information to the operating staff during accident conditions. A methodology to identify and assess the information needs of the operating staff of a nuclear power plant during a severe accident has been developed. The methodology identifies (a) the information needs of the plant personnel during a wide range of accident conditions, (b) the existing plant measurements capable of supplying these information needs and minor additions to instrument and display systems that would enhance management capabilities, (c) measurement capabilities and limitations during severe accident conditions, and (d) areas in which the information systems could mislead plant personnel

  20. Application of the accident management information needs methodology to a severe accident sequence

    Energy Technology Data Exchange (ETDEWEB)

    Ward, L.W.; Hanson, D.J.; Nelson, W.R. (Idaho National Engineering Laboratory, Idaho Falls (USA)); Solberg, D.E. (Nuclear Regulatory Commission, Washington, DC (USA))

    1989-11-01

    The U.S. Nuclear Regulatory Commission is conducting an accident management research program that emphasizes the use of severe accident research to enhance the ability of plant operating personnel to effectively manage severe accidents. Hence, it is necessary to ensure that the plant instrumentation and information systems adequately provide this information to the operating staff during accident conditions. A methodology to identify and assess the information needs of the operating staff of a nuclear power plant during a severe accident has been developed. The methodology identifies (a) the information needs of the plant personnel during a wide range of accident conditions, (b) the existing plant measurements capable of supplying these information needs and minor additions to instrument and display systems that would enhance management capabilities, (c) measurement capabilities and limitations during severe accident conditions, and (d) areas in which the information systems could mislead plant personnel.

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  2. Severe accident management guidelines tool

    International Nuclear Information System (INIS)

    Gutierrez Varela, Javier; Tanarro Onrubia, Augustin; Martinez Fanegas, Rafael

    2014-01-01

    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)

  3. Lessons from Chernobyl post-accident management

    International Nuclear Information System (INIS)

    Schneider, T.

    2012-01-01

    The Chernobyl accident has shown that the long-term management of its consequences is not straightforward. The management of the consequences has revealed the complexity of the situation to deal with. The long-term contamination of the environment has affected all the dimensions of the daily life of the inhabitants living in affected territories: health, environment, social life, education, work, distribution of foodstuffs and commodities... The experience from the Chernobyl accident shows 4 key issues that may be beneficial for the populations living in territories affected by the Fukushima accident: 1) the direct involvement of the inhabitants in their own protection, 2) the radiation monitoring system and health surveillance at the local level, 3) to develop a practical radiation protection culture among the population, and 4) the setting up of economic measures to favour the local development. (A.C.)

  4. Severe accident analysis methodology in support of accident management

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  5. Regulatory approach to accident management in Sweden

    International Nuclear Information System (INIS)

    Hoegberg, L.

    1989-01-01

    The Swedish accident management program includes the following components: definition of overall safety and radiation protection objectives for the program; definition of appropriate accident management strategies to reach these objectives, based on plant-specific severe accident analysis; development and installation of appropriate accident management systems and associated management procedure; definition of roles and resposibilities for plant staff involved in accident management and implementation of appropriate training programs. The discussion of these components tries to highlight the basic technical concepts and approaches and the underlying safety philosophy rather than going into design details. 5 figs., 7 refs

  6. The management of severe accidents

    International Nuclear Information System (INIS)

    Pelce, J.; Brignon, P.

    1987-01-01

    In considering severe accidents in water power reactors, a major problem that arises is how to manage them in such a way that the situation can be controlled as well as possible, from the aspects both of preventing serious damage to the core of limiting the discharge of radioactivity. A number of countries have announced provisions in the field of accident management, some already set up, others planned, but these mainly apply to preventing damage to the core. Part of this report deals with this aspect, to show that there is a fairly wide consensus on how problems should be approached. Attitudes vary, on the other hand, in the approach to mitigate radioactive release. In fact, few countries have proposed concrete steps to manage severe accidents in the final stages when the core is seriously damaged. Since it is difficult to compare different approaches, only the French approach is described. This description is however very brief, because in the five or six years since it was defined, the approach has been presented many times. The stress is placed more on the comments which this type of approach suggests, to make the subsequent general discussion easier

  7. Joint research project WASA-BOSS: Further development and application of severe accident codes. Assessment and optimization of accident management measures. Project B: Accident analyses for pressurized water reactors with the application of the ATHLET-CD code; Verbundprojekt WASA-BOSS: Weiterentwicklung und Anwendung von Severe Accident Codes. Bewertung und Optimierung von Stoerfallmassnahmen. Teilprojekt B: Druckwasserreaktor-Stoerfallanalysen unter Verwendung des Severe-Accident-Codes ATHLET-CD

    Energy Technology Data Exchange (ETDEWEB)

    Jobst, Matthias; Kliem, Soeren; Kozmenkov, Yaroslav; Wilhelm, Polina

    2017-02-15

    Within the framework of the project an ATHLET-CD input deck for a generic German PWR of type KONVOI has been created. This input deck was applied to the simulation of severe accidents from the accident categories station blackout (SBO) and small-break loss-of-coolant accidents (SBLOCA). The complete accident transient from initial event at full power until the damage of reactor pressure vessel (RPV) is covered and all relevant severe accident phenomena are modelled: start of core heat up, fission product release, melting of fuel and absorber material, oxidation and release of hydrogen, relocation of molten material inside the core, relocation to the lower plenum, damage and failure of the RPV. The model has been applied to the analysis of preventive and mitigative accident management measures for SBO and SBLOCA transients. Therefore, the measures primary side depressurization (PSD), injection to the primary circuit by mobile pumps and for SBLOCA the delayed injection by the cold leg hydro-accumulators have been investigated and the assumptions and start criteria of these measures have been varied. The time evolutions of the transients and time margins for the initiation of additional measures have been assessed. An uncertainty and sensitivity study has been performed for the early phase of one SBO scenario with PSD (until the start of core melt). In addition to that, a code -to-code comparison between ATHLET-CD and the severe accident code MELCOR has been carried out.

  8. Occupational Accidents And Preventive Measures

    CERN Document Server

    Fassnacht, V

    2006-01-01

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

  9. Application of the accident management information needs methodology to a severe accident sequence

    International Nuclear Information System (INIS)

    Ward, L.W.; Hanson, D.J.; Nelson, W.R.; Solberg, D.E.

    1989-01-01

    The U.S. Nuclear Regulatory Commission (NRC) is conducting an Accident Management Research Program that emphasizes the application of severe accident research results to enhance the capability of plant operating personnel to effectively manage severe accidents. A methodology to identify and assess the information needs of the operating staff of a nuclear power plant during a severe accident has been developed as part of the research program designed to resolve this issue. The methodology identifies the information needs of the plant personnel during a wide range of accident conditions, the existing plant measurements capable of supplying these information needs and what, if any minor additions to instrument and display systems would enhance the capability to manage accidents, known limitations on the capability of these measurements to function properly under the conditions that will be present during a wide range of severe accidents, and areas in which the information systems could mislead plant personnel. This paper presents an application of this methodology to a severe accident sequence to demonstrate its use in identifying the information which is available for management of the event. The methodology has been applied to a severe accident sequence in a Pressurized Water Reactor with a large dry containment. An examination of the capability of the existing measurements was then performed to determine whether the information needs can be supplied

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

    International Nuclear Information System (INIS)

    Girard, P.; Dubreuil, G.H.

    1996-01-01

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

  11. HTR-10 severe accident management

    International Nuclear Information System (INIS)

    Xu Yuanhui; Sun Yuliang

    1997-01-01

    The High Temperature Gas-cooled Reactor (HTR-10) is under construction at the Institute of Nuclear Energy Technology site northwest of Beijing. This 10 MW thermal plant utilizes a pebble bed high temperature gas cooled reactor for a large range of applications such as electricity generation, steam and district heat generation, gas turbine and steam turbine combined cycle and process heat for methane reforming. The HTR-10 is the first high temperature gas cooled reactor to be licensed in China. This paper describes the safety characteristics and design criteria for the HTR-10 as well as the accident management and analysis required for the licensing process. (author)

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  13. SAMEX: A severe accident management support expert

    International Nuclear Information System (INIS)

    Park, Soo-Yong; Ahn, Kwang-Il

    2010-01-01

    A decision support system for use in a severe accident management following an incident at a nuclear power plant is being developed which is aided by a severe accident risk database module and a severe accident management simulation module. The severe accident management support expert (SAMEX) system can provide the various types of diagnostic and predictive assistance based on the real-time plant specific safety parameters. It consists of four major modules as sub-systems: (a) severe accident risk data base module (SARDB), (b) risk-informed severe accident risk data base management module (RI-SARD), (c) severe accident management simulation module (SAMS), and (d) on-line severe accident management guidance module (on-line SAMG). The modules are integrated into a code package that executes within a WINDOWS XP operating environment, using extensive user friendly graphics control. In Korea, the integrated approach of the decision support system is being carried out under the nuclear R and D program planned by the Korean Ministry of Education, Science and Technology (MEST). An objective of the project is to develop the support system which can show a theoretical possibility. If the system is feasible, the project team will recommend the radiation protection technical support center of a national regulatory body to implement a plant specific system, which is applicable to a real accident, for the purpose of immediate and various diagnosis based on the given plant status information and of prediction of an expected accident progression under a severe accident situation.

  14. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1991-01-01

    A recently completed Oak Ridge effort proposes two management strategies for mitigation of the events that might occur in-vessel after the onset of significant core damage in a BWR severe accident. While the probability of such an accident is low, there may be effective yet inexpensive mitigation measures that could be implemented employing the existing plant equipment and requiring only additions to the plant emergency procedures. In this spirit, accident management strategies have been proposed for use of a borated solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and for containment flooding to maintain the core debris within the reactor vessel if injection systems cannot be restored. The proposed strategy for poisoning of the water used for vessel reflood should injection systems be restored after control blade damage has occurred has great promise, using only the existing plant equipment but employing a different chemical form for the boron poison. The dominant BWR severe accident sequence is Station Blackout and without means for mechanical stirring or heating of the storage tank, the question of being able to form the poisoned solution under accident conditions becomes of supreme importance. On the other hand, the proposed strategy for drywell flooding to cool the reactor vessel bottom head and prevent the core and structure debris from escaping to the drywell holds less promise. This strategy does, however, have potential for future plant designs in which passive methods might be employed to completely submerge the reactor vessel under severe accident conditions without the need for containment venting

  15. Use of PSA to support accident management at NPPs

    International Nuclear Information System (INIS)

    Gomez Cobo, A.

    1997-01-01

    The presentation discusses the following: Overview of PSA level 2; Introduction: Framework; Accident Progression Phenomena in the Confinement/containment; Severe Accident Sequences; Examples; Results and Insights. Accident Management: Concepts; Process; Use of PSA to support Accident; Management

  16. NPP Krsko Severe Accident Management Guidelines Implementation

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  17. Approach to accident management in RBMK-1500

    International Nuclear Information System (INIS)

    Kaliatka, A.; Urbonavicius, E.; Uspuras, E.

    2008-01-01

    In order to ensure the safe operation of the nuclear power plants accident management programs are being developed around the world. These accident management programs cover the whole spectrum of accidents, including severe accidents. A lot of work is done to investigate the severe accident phenomena and implement severe accident management in NPPs with vessel-type reactors, while less attention is paid to channel-type reactors CANDU and RBMK. Ignalina NPP with RBMK-1500 reactor has implemented symptom based emergency operation procedures, which cover management of accidents until the core damage and do not extend to core damage region. In order to ensure coverage of the whole spectrum of accidents and meet the requirements of IAEA the severe accident management guidelines have to be developed. This paper presents the basic principles and approach to management of beyond design basis accidents at Ignalina NPP. In general, this approach could be applied to NPPs with RBMK-1000 reactors that are available in Russia, but the design differences should be taken into account

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

    International Nuclear Information System (INIS)

    Sdouz, Gert

    2006-01-01

    demonstrated that the accident management measures have quite lower consequences. In addition it was shown that in the case of a 'Large Break LOCA'-sequence the intact containment retains the nuclides up to a factor of 20 000. This is much more than in the case of a 'Station Blackout'-sequence. Within the frame of the study 17 source terms have been generated to evaluate in detail accident management strategies for VVER-1000 reactors. (authors)

  19. Method of assessing severe accident management strategies

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  20. The screening approach for review of accident management programmes

    International Nuclear Information System (INIS)

    Misak, J.

    1999-01-01

    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

  1. Containment severe accident management - selected strategies

    International Nuclear Information System (INIS)

    Duco, J.; Royen, J.; Rohde, J.; Frid, W.; De Boeck, B.

    1994-01-01

    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

    International Nuclear Information System (INIS)

    Sirola, M.

    1995-01-01

    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. Accident management-defence in depth in Indian PHWRS

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  4. On preparation for accident management in LWR power stations

    International Nuclear Information System (INIS)

    1996-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    International Nuclear Information System (INIS)

    Lahtinen, J.

    2006-04-01

    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)

  7. A Methodology for Probabilistic Accident Management

    International Nuclear Information System (INIS)

    Munteanu, Ion; Aldemir, Tunc

    2003-01-01

    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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  9. Risk evaluation of accident management strategies

    International Nuclear Information System (INIS)

    Dingman, S.; Camp, A.

    1992-01-01

    The use of Probabilistic Risk Assessment (PRA) methods to evaluate accident management strategies in nuclear power plants discussed in this paper. The PRA framework allows an integrated evaluation to be performed to give the full implications of a particular strategy. The methodology is demonstrated for a particular accident management strategy, intentional depressurization of the reactor coolant system to avoid containment pressurization during the ejection of molten debris at vessel breach

  10. Development of integrated accident management assessment technology

    International Nuclear Information System (INIS)

    Jung, Won Dea; Ha, Jae Joo; Jin, Young Ho

    2002-04-01

    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

  11. Severe accident management: radiation dose control, Fukushima Daiichi and TMI-2 nuclear plant accidents

    International Nuclear Information System (INIS)

    Shaw, Roger

    2014-01-01

    This presentation presents valuable dose information related to the Fukushima Daiichi and Three Mile Island Unit 2 (TMI-2) Nuclear Plant accidents. Dose information is provided for what is well known for TMI-2, and what is available for Fukushima Daiichi. Particular emphasis is placed on the difference between the type of reactors involved, overarching plant damage issues, and radiation worker dose outcomes. For TMI-2, more in depth dose data is available for the accident and the subsequent recovery efforts. The comparisons demonstrate the need to understand the wide variation in potential dose management measures and outcomes for severe reactor accidents. (author)

  12. Accident knowledge and emergency management

    Energy Technology Data Exchange (ETDEWEB)

    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.

  13. Accident knowledge and emergency management

    International Nuclear Information System (INIS)

    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

  14. Use of simulators in severe accident management

    International Nuclear Information System (INIS)

    Evans, R.C.

    1994-01-01

    The U.S. nuclear utility industry is moving in a deliberate fashion through a coordinated industry severe accident working group to study and augment, where appropriate, the existing utility organizational and emergency planning structure to address accident and severe accident management. Full-scope simulators are used extensively to train licensed operators for their initial license examinations and continually thereafter in licensed operator requalification training and yearly examinations. The goal of the training (both initial and requalification) is to ensure that operators possess adequate knowledge, skills and abilities to prevent an event from progressing to core damage. The use of full-scope simulators in severe accident management training is in large part viewed by the industry as being premature. The working group study has not progressed to the point where the decision to employ full-scope simulators can be logically considered. It is not however premature to consider part-task or work station simulators as invaluable research tools to support the industry's study. These simulators could be employed, subject to limitations in the current state of knowledge regarding severe accident progression and phenomenological responses, in the validation and verification (V and V) of severe accident models or codes as they are developed. The U.S. nuclear utility industry has made substantial strides in the past 12 years in the accident prevention, mitigation and management arena. These strides are a product of the industry's preference for a logical and systematic approach to change. (orig.)

  15. Jose Cabrera NPP severe accident management activities

    International Nuclear Information System (INIS)

    Blanco, J.; Almeida, P.; Saiz, J.; Sastre, J.L.; Delgado, R.

    1998-01-01

    To prepare a common acting plan with respect to Severe Accident Management, in 1994 was founded the severe accident management ''ad-hoc'' working group from the Spanish Westinghouse PWR Nuclear Power Plant Owners Group. In this group actively collaborated the Jose Cabrera NPP Training Centre and the Department of Nuclear Engineering of UNION FENOSA. From this moment, Jose Cabrera NPP began the planning of its specific Severe Accident Management Program, which main point are Severe Accident Management Guidelines (SAMG). To elaborate this guidelines, the Spanish translation of Westinghouse Owners Group (WOG) Severe Accident Management Guidelines were considered the reference documents. The implementation of this Guidelines to Jose Cabrera NPP started on January 1997. Once the specific guidelines have been implemented to the plant, training activities for the personnel involved in severe accident issues will be developed. To prepare the training exercises MAAP4 code will be used, and with this intention, a specific Jose Cabrera NPP MAAP-GRAAPH screen has been developed. Furthermore, a wide selection of MAAP input files for the simulation of different scenarios and accidental events is available. (Author)

  16. SEVERE ACCIDENT MANAGEMENT STATUS AT Loviisa

    International Nuclear Information System (INIS)

    Kymalainen, O.; Tuomisto, H.

    1997-01-01

    Some of the specific design features of IVO's Loviisa Plant, most notably the ice-condenser containment, strongly affect the plant response in a hypothetical core melt accident. They have together with the relatively stringent Finnish regulatory requirements forced IVO to develop a tailor made severe accident management strategy for Loviisa. The low design pressure of the ice-condenser containment complicates the design of the hydrogen management system. On the other hand, the ice-condensers and the water available from them are facilitating factors regarding in-vessel retention of corium by external cooling of reactor pressure vessel. This paper summarizes the Finnish severe accident requirements, IVO's approach to severe accidents, and its application to the Loviisa Plant

  17. Issues and measures in the design process from the perspective of risk management of construction projects. study of power plant construction projects accident cases

    International Nuclear Information System (INIS)

    Iwahara, Hirohiko; Shiraki, Wataru; Inomo, Hitoshi; Hasegawa, Syuichi

    2015-01-01

    Construction of power plants, foundation work, consisting of a wide variety of construction work, such as plant equipment work. And, civil engineering, technician electrical such as different engineering field, is a comprehensive construction project that works for the design conditions of the structure. However, if the cooperation design conditions is not sufficient, as a construction project, the optimal structures may not be said to have been built. As a result, total cost or increased, including the initial cost of the end construction projects, it is be a cause of the accident. Previous studies, plant equipment construction, is related to safety management and risk of foundation work such as individual construction were many. In this paper, as an example the power plant construction, and performs the following discussion from the point of view of risk management of large-scale construction projects that these individual construction work together with each other. The importance of design conditions cooperation, (1) 'Challenges and countermeasures of ordering method of construction projects', to verify from the (2) 'actually happened substation foundation displacement accident'. And on whether or not the construction project order institutions can be involved in the design from the site preparation stage, we study (3) for 'construction work scope and risk control the construction project ordering institutions to implement' the risk to the natural disaster (earthquake). From these, we describe the challenges and measures in the construction project of the design process. (author)

  18. Aspects of accident management in Cernavoda NPP

    International Nuclear Information System (INIS)

    Dascalu, N.

    1999-01-01

    As a general conclusion, the accident management system as implemented at Cerna voda NPP is expected to be appropriate for handling a severe accident, should it occur, in such a way that the environmental radiological consequences would be insignificant and radiation exposure of the personnel be within recommendations. It is recognized, however, that continued development and verification of the system as well as effective personnel training programs are essential to maintain the safety level achieved. (author)

  19. Reconstruction of the Chernobyl emergency and accident management

    International Nuclear Information System (INIS)

    Schinner, F.; Andreev, I.; Andreeva, I.; Fritsche, F.; Hofer, P.; Lettner, E.; Seidelberger, E.; Kromp-Kolb, H.; Kromp, W.

    1998-01-01

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

  20. Managing Nuclear Reactor Accidents: Issues Raised by Three Mile Island

    OpenAIRE

    Hamilton, G.W.

    1980-01-01

    This paper provides a descriptive account of significant events in the accident at the Three Mile Island nuclear power plant in March, 1979. It is based upon documents collected as background materials for the IIASA workshop: Procedural and Organizational Measures for Accident Management: Nuclear Reactors. In addition to the references listed, information was supplied by John Lathrop, who conducted interviews with government and industry officials involved in the crisis. There have been ...

  1. Medical management of radiation accidents

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1982-12-31

    The film gives advice on actions to be taken in case of a radiation accident. It addresses involving external irradiation of the whole and partial body, very localized exposure, uptake of radioiodine, inhalation of transuranium elements and a wound of a finger. The film is intended to illustrate the Agency`s Safety Series No. 47 entitled ``Manual on Early Medical Treatment of Possible Radiation Injury`` published in 1978

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

    International Nuclear Information System (INIS)

    Nelson, W.R.; Hanson, D.J.; Ward, L.W.; Solberg, D.E.

    1988-01-01

    While it would be very difficult to predetermine all of the actions required to mitigate the consequences of every potential severe accident for a nuclear power plant, development of additional guidance and training could improve the likelihood that the operating crew would implement effective sever-accident management measures. The US Nuclear Regulatory Commission (NRC) is conducting an Accident Management Research Program that emphasizes the application of severe-accident research results to enhance the capability of the plant operating crew to effectively manage severe accidents. One element of this program includes identification of the information needed by the operating crew in severe-accident situations. This paper discusses a method developed for identifying these information needs and its application. The methodology has been applied to a generic reactor design representing a PWR with a large dry containment. The information needs were identified by systematically determining what information is needed to assess the health of the critical functions, identify the presence of challenges, select strategies, and assess the effectiveness of these strategies. This method allows the systematic identification of information needs for a broad range of severe-accident scenarios and can be validated by exercising the functional models for any specific event sequence

  3. ATHLET validation using accident management experiments

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

  4. Use of analytical aids for accident management

    International Nuclear Information System (INIS)

    Ward, L.W.

    1991-01-01

    The use of analytical aids by utility technical support teams can enhance the staff's ability to manage accidents. Since instrumentation is exposed to environments beyond design-basis conditions, instruments may provide ambiguous information or may even fail. While it is most likely that many instruments will remain operable, their ability to provide unambiguous information needed for the management of beyond-design-basis events and severe accidents is questionable. Furthermore, given these limitation in instrumentation, the need to ascertain and confirm current plant status and forecast future behavior to effectively manage accidents at nuclear facilities requires a computational capability to simulate the thermal and hydraulic behavior in the primary, secondary, and containment systems. With the need to extend the current preventive approach in accident management to include mitigative actions, analytical aids could be used to further enhance the current capabilities at nuclear facilities. This need for computational or analytical aids is supported based on a review of the candidate accident management strategies discussed in NUREG/CR-5474. Based on the review of the NUREG/CR-5474 strategies, two major analytical aids are considered necessary to support the implementation and monitoring of many of the strategies in this document. These analytical aids include (1) An analytical aid to provide reactor coolant and secondary system behavior under LOCA conditions. (2) An analytical aid to predict containment pressure and temperature response with a steam, air, and noncondensable gas mixture present

  5. Noticeable aspects of selecting intervention measures for nuclear accidents

    International Nuclear Information System (INIS)

    Guo Yong; Shi Yuanming

    1993-01-01

    Referring to the experience of intervention measures taken for protecting the public after Chernobyl accident and to recent knowledge of source terms of reactor accidents, the noticeable aspects of selecting of intervention measures for nuclear accidents is discussed

  6. ROSA/LSTF Test and RELAP5 Analyses on PWR Cold Leg Small-Break LOCA with Accident Management Measure and PKL Counterpart Test

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Takeshi; Ohtsu, Iwao [Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokaimura (Japan)

    2017-08-15

    An experiment using the Primaerkreislaeufe Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg small-break loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

  7. ROSA/LSTF Test and RELAP5 Analyses on PWR Cold Leg Small-Break LOCA with Accident Management Measure and PKL Counterpart Test

    Directory of Open Access Journals (Sweden)

    Takeshi Takeda

    2017-08-01

    Full Text Available An experiment using the Primӓrkreislӓufe Versuchsanlage (PKL was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF on a cold leg small-break loss-of-coolant accident with an accident management (AM measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

  8. Development of Krsko Severe Accident Management Database (SAMD)

    International Nuclear Information System (INIS)

    Basic, I.; Kocnar, R.

    1996-01-01

    Severe Accident Management is a framework to identify and implement the Emergency Response Capabilities that can be used to prevent or mitigate severe accidents and their consequences. Krsko Severe Accident Management Database documents the severe accident management activities which are developed in the NPP Krsko, based on the Krsko IPE (Individual Plant Examination) insights and Generic WOG SAMGs (Westinghouse Owners Group Severe Accident Management Guidance). (author)

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

    International Nuclear Information System (INIS)

    Golshan, Mina

    2013-01-01

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

  10. Assessment of uncertainties in severe accident management strategies

    International Nuclear Information System (INIS)

    Kastenberg, W.E.; Apostolakis, G.; Catton, I.; Dhir, V.K.; Okrent, D.

    1990-01-01

    Recent progress on the development of Probabilistic Risk Assessment (PRA) as a tool for qualifying nuclear reactor safety and on research devoted to severe accident phenomena has made severe accident management an achievable goal. Severe accident management strategies may involve operational changes, modification and/or addition of hardware, and institutional changes. In order to achieve the goal of managing severe accidents, a method for assessment of strategies must be developed which integrates PRA methodology and our current knowledge concerning severe accident phenomena, including uncertainty. The research project presented in this paper is aimed at delineating uncertainties in severe accident progression and their impact on severe accident management strategies

  11. Concept and objectives of accident management in LWR type plants

    International Nuclear Information System (INIS)

    Herttrich, P.M.; Hicken, E.F.

    1990-01-01

    For the sake of putting the previous protection and prevention concept in its proper place, it is shown, first of all, on which basis the prevention against damages required according to the state of the art in science and technology was proved under the licensing practice applied so far. Secondly, the previous practice of dynamic upgrading of safety engineering and risk prevention is explained. The introduction of accident management measures is a consequent continuation of this practice. Concrete approaches and objectives of accident management are outlined; an overview of scientific and technical foundations for the development, assessment and introduction of accident management measures is given, and finally the most important organizational and procedural aspects are dealt with. (orig./DG) [de

  12. Severe accident management. Optimized guidelines and strategies

    International Nuclear Information System (INIS)

    Braun, Matthias; Löffler, Micha; Plank, Hermann; Asse, Dietmar; Dimmelmeier, Harald

    2014-01-01

    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

  13. Use of NUREG-1150 and IPEs in accident management

    International Nuclear Information System (INIS)

    Mauersberger

    1992-01-01

    The fundamental objective of the accident management program is to assure, in the event of a severe accident at a nuclear plant, that the effectiveness of personnel and equipment is maximized in preventing or mitigating the consequences of the accident. This document studies the use of NUREG-1150 and IPEs in accident management. Figs

  14. Chernobyl reactor accident: medical management

    International Nuclear Information System (INIS)

    Iyer, G.K.

    1996-01-01

    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

  15. Management of foodstuffs after nuclear accidents

    International Nuclear Information System (INIS)

    1991-01-01

    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

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

    International Nuclear Information System (INIS)

    Jeong, K.-S.; Kim, K.-R.; Jung, W.-D.; Ha, J.-J.

    2002-01-01

    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. Evaluation of severe accident environmental conditions taking accident management strategy into account for equipment survivability assessments

    International Nuclear Information System (INIS)

    Lee, Byung Chul; Jeong, Ji Hwan; Na, Man Gyun; Kim, Soong Pyung

    2003-01-01

    This paper presents a methodology utilizing accident management strategy in order to determine accident environmental conditions in equipment survivability assessments. In case that there is well-established accident management strategy for specific nuclear power plant, an application of this tool can provide a technical rationale on equipment survivability assessment so that plant-specific and time-dependent accident environmental conditions could be practically and realistically defined in accordance with the equipment and instrumentation required for accident management strategy or action appropriately taken. For this work, three different tools are introduced; Probabilistic Safety Assessment (PSA) outcomes, major accident management strategy actions, and Accident Environmental Stages (AESs). In order to quantitatively investigate an applicability of accident management strategy to equipment survivability, the accident simulation for a most likely scenario in Korean Standard Nuclear Power Plants (KSNPs) is performed with MAAP4 code. The Accident Management Guidance (AMG) actions such as the Reactor Control System (RCS) depressurization, water injection into the RCS, the containment pressure and temperature control, and hydrogen concentration control in containment are applied. The effects of these AMG actions on the accident environmental conditions are investigated by comparing with those from previous normal accident simulation, especially focused on equipment survivability assessment. As a result, the AMG-involved case shows the higher accident consequences along the accident environmental stages

  18. Managing severe reactor accidents. A review and evaluation of our knowledge on reactor accidents and accident management

    International Nuclear Information System (INIS)

    Gustavsson, Veine

    2002-11-01

    The report gives a review of the results from the last years research on severe reactor accidents, and an opinion on the possibilities to refine the present strategies for accident management in Swedish and Finnish BWRs. The following aspect of reactor accidents are the major themes of the study: 1. Early pressure relief from hydrogen production; 2. Recriticality in re-flooded, degraded core; 3. Melt-through; 4. Steam explosion after melt-through; 5. Coolability of the melt after after melt-through; 6. Hydrogen fire in the reactor containment; 7. Leaking containment; 8. Hydrogen fire in the reactor building; 9. Long-time developments after a severe accident; 10. Accidents during shutdown for overhaul; 11. Information need for remedial actions. Possibilities for improving the strategies in each of these areas are discussed. The review shows that our knowledge is sufficient in the areas 1, 2, 4, 6, 8. For the other areas, more research is needed

  19. Severe accident management at South Africa's Koeberg plant

    International Nuclear Information System (INIS)

    Prior, R.P.; Wolvaardt, F.P.; Holderbaum, D.F.; Lutz, R.J.; Taylor, J.J.; Hodgson, C.D.

    1997-01-01

    Between the middle of 1993 and the end of 1995, Westinghouse and Eskom implemented plant specific Severe Accident Management Guidelines (SAMGs) at the Koeberg Nuclear Power Plant in South Africa. Prior to this project, Koeberg, like many plants, had emergency operating procedures which contain guidance for plant personnel to perform preventive accident management measures in event of an accident. There was, however, no structured guidance on recovery from an event which progresses past core damage -mitigative accident management. The SAMGs meet this need. In this paper, the Westinghouse approach to severe accident management is outlined, and the Koeberg implementation project described. A few key issues which arose during implementation are discussed, including plant instrumentation, flooding of the reactor pit, organisation and training of the Technical Support Centre staff, and impact of SAMG on risk. The means by which both generic and plant-specific SAMG have been validated is also summarised. In the next few years, many LWR owners will be implementing SAMG. In the U.S. all plants are in the process of developing SAMG. The Koeberg project is believed to be the first plant specific implementation of the WOG SAMG worldwide, and this paper has hopefully provided insights into some of the implementation issues for those about to undertake similar projects. (author)

  20. Simulation of operator's actions during severe accident management

    International Nuclear Information System (INIS)

    Viktorov, A.

    2015-01-01

    Implementing accident management counter measures or actions to mitigate consequences of a severe accident is essential to reduce radiological risks to the public and environment. Station-specific severe accident management guidelines (SAMGs) have been developed and implemented at all Canadian nuclear power plants. Following the Fukushima Daiichi nuclear accident certain enhancements were introduced to the SAMG, namely consideration of multi-units accidents, events involving spent fuel pools, incorporation of capability offered by the portable emergency mitigating equipment, and so on. To evaluate the adequacy and usability of the SAMGs, CNSC staff initiated a number of activities including a desktop review of SAMG documentation, evaluation of SAMG implementation through exercises and interviews with station staff, and independent verification of SAMG action effectiveness. This paper focuses on the verification of SAMG actions through analytical simulations. The objectives of the work are two-folds: (a) to understand the effectiveness of SAMG-specified mitigation actions in addressing the safety challenges and (b) to check for potential negative effects of the action. Some sensitivity calculations were performed to help understanding of the impact from actions that rely on the partially effective equipment or limited material resources. The severe accident computer code MAAP4-CANDU is used as a tool in this verification. This paper will describe the methodology used in the verification of SAMG actions and some results obtained from simulations. (author)

  1. Occupational Radiation Protection in Severe Accident Management

    International Nuclear Information System (INIS)

    2015-01-01

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

  2. Early measurements after the Goiania accident

    International Nuclear Information System (INIS)

    Godoy, J.M.; Moreira, M.C.F.; Fonseca, E.S. da

    2000-01-01

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

  3. Analytical measurements of fission products during a severe nuclear accident

    Science.gov (United States)

    Doizi, D.; Reymond la Ruinaz, S.; Haykal, I.; Manceron, L.; Perrin, A.; Boudon, V.; Vander Auwera, J.; tchana, F. Kwabia; Faye, M.

    2018-01-01

    The Fukushima accident emphasized the fact that ways to monitor in real time the evolution of a nuclear reactor during a severe accident remain to be developed. No fission products were monitored during twelve days; only dose rates were measured, which is not sufficient to carry out an online diagnosis of the event. The first measurements were announced with little reliability for low volatile fission products. In order to improve the safety of nuclear plants and minimize the industrial, ecological and health consequences of a severe accident, it is necessary to develop new reliable measurement systems, operating at the earliest and closest to the emission source of fission products. Through the French program ANR « Projet d'Investissement d'Avenir », the aim of the DECA-PF project (diagnosis of core degradation from fission products measurements) is to monitor in real time the release of the major fission products (krypton, xenon, gaseous forms of iodine and ruthenium) outside the nuclear reactor containment. These products are released at different times during a nuclear accident and at different states of the nuclear core degradation. Thus, monitoring these fission products gives information on the situation inside the containment and helps to apply the Severe Accident Management procedures. Analytical techniques have been proposed and evaluated. The results are discussed here.

  4. Analytical measurements of fission products during a severe nuclear accident

    Directory of Open Access Journals (Sweden)

    Doizi D.

    2018-01-01

    Full Text Available The Fukushima accident emphasized the fact that ways to monitor in real time the evolution of a nuclear reactor during a severe accident remain to be developed. No fission products were monitored during twelve days; only dose rates were measured, which is not sufficient to carry out an online diagnosis of the event. The first measurements were announced with little reliability for low volatile fission products. In order to improve the safety of nuclear plants and minimize the industrial, ecological and health consequences of a severe accident, it is necessary to develop new reliable measurement systems, operating at the earliest and closest to the emission source of fission products. Through the French program ANR « Projet d’Investissement d’Avenir », the aim of the DECA-PF project (diagnosis of core degradation from fission products measurements is to monitor in real time the release of the major fission products (krypton, xenon, gaseous forms of iodine and ruthenium outside the nuclear reactor containment. These products are released at different times during a nuclear accident and at different states of the nuclear core degradation. Thus, monitoring these fission products gives information on the situation inside the containment and helps to apply the Severe Accident Management procedures. Analytical techniques have been proposed and evaluated. The results are discussed here.

  5. PSA use in accident management studies in Japan

    International Nuclear Information System (INIS)

    Hirano, Mitsumasa

    1994-01-01

    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

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

    International Nuclear Information System (INIS)

    Tentner, A.M.; Parma, E.; Wei, T.; Wigeland, R.

    2010-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-03-01

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

  8. Management of internal contamination accidents

    International Nuclear Information System (INIS)

    Lipsztein, J.L.; Melo, D.R.

    1997-01-01

    This paper concerns with the techniques for intakes assessment which depend on the mode and level of intake, the type of energy of the radiation emitted, the biokinetic of the contaminant, and the sensitivity and availability of measurement facilities. In vivo and in vitro techniques are used to quantify internal contaminations

  9. Artificial intelligence applications in accident management

    International Nuclear Information System (INIS)

    Cain, D.G.

    1989-01-01

    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)

  10. Application of NUREG-1150 methods and results to accident management

    International Nuclear Information System (INIS)

    Dingman, S.; Sype, T.; Camp, A.; Maloney, K.

    1991-01-01

    The use of NUREG-1150 and similar probabilistic risk assessments in the Nuclear Regulatory Commission (NRC) and industry risk management programs is discussed. Risk management is more comprehensive than the commonly used term accident management. Accident management includes strategies to prevent vessel breach, mitigate radionuclide releases from the reactor coolant system, and mitigate radionuclide releases to the environment. Risk management also addresses prevention of accident initiators, prevention of core damage, and implementation of effective emergency response procedures. The methods and results produced in NUREG-1150 provide a framework within which current risk management strategies can be evaluated, and future risk management programs can be developed and assessed. Examples of the use of the NUREG-1150 framework for identifying and evaluating risk management options are presented. All phases of risk management are discussed, with particular attention given to the early phases of accidents. Plans and methods for evaluating accident management strategies that have been identified in the NRC accident management program are discussed

  11. Application of NUREG-1150 methods and results to accident management

    International Nuclear Information System (INIS)

    Dingman, S.; Sype, T.; Camp, A.; Maloney, K.

    1990-01-01

    The use of NUREG-1150 and similar Probabilistic Risk Assessments in NRC and industry risk management programs is discussed. ''Risk management'' is more comprehensive than the commonly used term ''accident management.'' Accident management includes strategies to prevent vessel breach, mitigate radionuclide releases from the reactor coolant system, and mitigate radionuclide releases to the environment. Risk management also addresses prevention of accident initiators, prevention of core damage, and implementation of effective emergency response procedures. The methods and results produced in NUREG-1150 provide a framework within which current risk management strategies can be evaluated, and future risk management programs can be developed and assessed. Examples of the use of the NUREG-1150 framework for identifying and evaluating risk management options are presented. All phases of risk management are discussed, with particular attention given to the early phases of accidents. Plans and methods for evaluating accident management strategies that have been identified in the NRC accident management program are discussed. 2 refs., 3 figs

  12. Enhancing AP1000 reactor accident management capabilities for long term accidents

    International Nuclear Information System (INIS)

    Jiang Pingting; Liu Mengying; Duan Chengjie; Liao Yehong

    2015-01-01

    Passive safety actions are considered as main measures under severe accident in AP1000 power plant. However, risk is still existed. According to PSA, several probable scenarios for AP1000 nuclear power plant are analyzed in this paper with MAAP the severe accident analysis code. According to the analysis results, several deficiencies of AP1000 severe accident management are found. The long term cooling and containment depressurization capability for AP1000 power plant appear to be most important factors under such accidents. Then, several temporary strategies for AP1000 power plant are suggested, including PCCWST temporary water supply strategy after 72h, temporary injection strategy for IRWST, hydrogen relief action in fuel building, which would improve the safety of AP1000 power plant. At last, assessments of effectiveness for these strategies are performed, and the results are compared with analysis without these strategies. The comparisons showed that correct actions of these strategies would effectively prevent the accident process of AP1000 power plant. (author)

  13. Developing a knowledge base for the management of severe accidents

    International Nuclear Information System (INIS)

    Nelson, W.R.; Jenkins, J.P.

    1986-01-01

    Prior to the accident at Three Mile Island, little attention was given to the development of procedures for the management of severe accidents, that is, accidents in which the reactor core is damaged. Since TMI, however, significant effort has been devoted to developing strategies for severe accident management. At the same time, the potential application of artificial intelligence techniques, particularly expert systems, to complex decision-making tasks such as accident diagnosis and response has received considerable attention. The need to develop strategies for accident management suggests that a computerized knowledge base such as used by an expert system could be developed to collect and organize knowledge for severe accident management. This paper suggests a general method which could be used to develop such a knowledge base, and how it could be used to enhance accident management capabilities

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

    International Nuclear Information System (INIS)

    Frid, W.

    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

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

    International Nuclear Information System (INIS)

    Miyaushiro, Norihiro

    2013-01-01

    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.)

  16. The expert assistant in accident management

    International Nuclear Information System (INIS)

    Goddard, A.J.H.; Cannell, R.J.

    1990-01-01

    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

  17. Development of Human Factor Management Requirements and Human Error Classification for the Prevention of Railway Accident

    International Nuclear Information System (INIS)

    Kwak, Sang Log; Park, Chan Woo; Shin, Seung Ryoung

    2008-08-01

    Railway accident analysis results show that accidents cased by human factors are not decreasing, whereas H/W related accidents are steadily decreasing. For the efficient management of human factors, many expertise on design, conditions, safety culture and staffing are required. But current safety management activities on safety critical works are focused on training, due to the limited resource and information. In order to improve railway safety, human factors management requirements for safety critical worker and human error classification is proposed in this report. For this accident analysis, status of safety measure on human factor, safety management system on safety critical worker, current safety planning is analysis

  18. Measured radioecological parameters after the Chernobyl accident

    International Nuclear Information System (INIS)

    Bonka, H.

    1989-01-01

    After the Chernobyl accident the radioactivity in the environment in Aachen was measured in detail. The change of the different radionuclies in the eco-system made it possible to obtain radioecological parameters especially for iodine and caesium. The most important data obtained like deposition velocity, washout coefficient, retention factor, removal rate constant, and transfer factor food-milk, food-beef, and soil-grass are reported. (orig.)

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

    International Nuclear Information System (INIS)

    Sobajima, M.

    1998-01-01

    The paper describes a proposal for accident management optimization based on the study of accident sequence and source term analyses for a BWR. In Japan, accident management measures are to be implemented in all LWRs by the year 2000 in accordance with the recommendation of the regulatory organization and based on the PSAs carried out by the utilities. Source terms were evaluated by the Japan Atomic Energy Research Institute (JAERI) with the THALES code for all BWR sequences in which loss of decay heat removal resulted in the largest release. Identification of the priority and importance of accident management measures was carried out for the sequences with larger risk contributions. Considerations for optimizing emergency operation guides are believed to be essential for risk reduction. (author)

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

    International Nuclear Information System (INIS)

    Sauvage, E. C.; Prior, R.; Coffey, K.; Mazurkiewicz, S. M.

    2006-01-01

    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

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

    International Nuclear Information System (INIS)

    Endo, Kazuto

    2015-01-01

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

  2. Severe accident management program at Cofrentes Nuclear Power Plant

    International Nuclear Information System (INIS)

    Borondo, L.; Serrano, C.; Fiol, M.J.; Sanchez, A.

    2000-01-01

    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)

  3. The link between off-site-emergency planning and plant-internal accident management

    Energy Technology Data Exchange (ETDEWEB)

    Braun, H.; Goertz, R.

    1995-02-01

    A variety of accident management measures has been developed and implemented in the German nuclear power plants. They constitute a fourth level of safety in the defence-in-depth concept. The containment venting system is an important example. A functioning link with well defined lines of communication between plant-internal accident management and off-site disaster emergency planning has been established.

  4. CANDU severe accident management guidance update

    International Nuclear Information System (INIS)

    Jones, L.; Popov, N.; Gilbert, L.; Weed, J.

    2014-01-01

    The CANDU Owners Group (COG) developed a set of generic and initial station-specific Severe Accident Management Guidance (SAMG) documents to mitigate the consequences to the public in the event of a severe accident. The generic portion of the COG SAMG was completed in 2006; the overall project including the station-specific phase was completed in April 2007. Over the years, the CANDU industry and utilities have continuously increased the knowledge base for SAMG and have incorporated various engineered features based on the knowledge obtained. As a result of the event that occurred at the Fukushima Daiiachi nuclear power plant (NPP) in Japan, the Canadian Nuclear Safety Commission (CNSC) established the CNSC Fukushima Task Force. The results of the task force were documented in INFO-0828, CNSC Staff Action Plan on the CNSC Fukushima Task Force Recommendations. Among the recommendation documented in INFO-828 were Fukushima Action Items (FAIs) directed towards the CANDU utilities in Canada; a portion of which are related to SAMG documentation updates and directed at enhancing SAM response. A COG joint project was established to support the closure of the CNSC FAIs and to revise the current CANDU documentation accordingly. This paper provides a high level summary of the COG project scope and results. It also demonstrates that the CANDU SAMG programs in Canada provide robust protection and mitigation of severe accidents. (author)

  5. CANDU severe accident management guidance update

    Energy Technology Data Exchange (ETDEWEB)

    Jones, L., E-mail: lisa.m.jones@opg.com [Ontario Power Generation, Pickering, ON (Canada); Popov, N., E-mail: nik.popov@rogers.com [Candu Owners Group, Toronto, ON (Canada); Gilbert, L., E-mail: lovell.gilbert@brucepower.com [Bruce Power, Tiverton, ON (Canada); Weed, J., E-mail: jeff.weed@candu.gov [Candu Owners Group, Toronto, ON (Canada)

    2014-07-01

    The CANDU Owners Group (COG) developed a set of generic and initial station-specific Severe Accident Management Guidance (SAMG) documents to mitigate the consequences to the public in the event of a severe accident. The generic portion of the COG SAMG was completed in 2006; the overall project including the station-specific phase was completed in April 2007. Over the years, the CANDU industry and utilities have continuously increased the knowledge base for SAMG and have incorporated various engineered features based on the knowledge obtained. As a result of the event that occurred at the Fukushima Daiiachi nuclear power plant (NPP) in Japan, the Canadian Nuclear Safety Commission (CNSC) established the CNSC Fukushima Task Force. The results of the task force were documented in INFO-0828, CNSC Staff Action Plan on the CNSC Fukushima Task Force Recommendations. Among the recommendation documented in INFO-828 were Fukushima Action Items (FAIs) directed towards the CANDU utilities in Canada; a portion of which are related to SAMG documentation updates and directed at enhancing SAM response. A COG joint project was established to support the closure of the CNSC FAIs and to revise the current CANDU documentation accordingly. This paper provides a high level summary of the COG project scope and results. It also demonstrates that the CANDU SAMG programs in Canada provide robust protection and mitigation of severe accidents. (author)

  6. Risk impact of two accident management strategies

    International Nuclear Information System (INIS)

    Dingman, S.; Camp, A.

    1992-01-01

    This report probabilistic Risk Assessment is used to evaluate two accident management strategies: intentionally depressurizing the reactor coolant system of a pressurized water reactor to prevent containment-pressurization during high pressure melt ejection, and flooding the containment of a boiling water reactor to prevent or delay vessel breach. Sensitivity studies indicated that intentional depressurization would not provide a significant risk reduction at Surry. A preliminary evaluation of the containment flooding strategy indicated that it might prove beneficial for some plants, but that further strategy development would be needed to fully evaluate the strategy-

  7. Accident management strategies for VVER-1000 reactors. Part 1: text

    International Nuclear Information System (INIS)

    Sdouz, G.; Sonneck, G.; Pachole, M.

    1994-10-01

    This report describes the effect of different accident management strategies on the onset, development and end of the core-concrete-interaction as well as on the containment integrity for a TMLB'-type sequence in a Pressurized Water Reactor of the type VVER- 1000. Using the computer code MARCH3 the following strategies were investigated: (1) One or more Spray and LP ECC Systems available with and without coolers after 10 hours (2) Inclusion of the reactor pressure vessel testing facility room to the cavity (3) Containment venting (4) External water supply and (5) Different electric power restoration times. The results show that some of these accident management measures will maintain the containment integrity and reduce the source term drastically, others will reduce the source term rate. For some measures final conclusions can only be given after complete source term calculations have been performed. (authors)

  8. Health protection measures after the Chernobyl accident

    International Nuclear Information System (INIS)

    Strand, P.; Reitan, J.B.; Harbitz, O.; Brynhildsen, L.

    1990-01-01

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

  9. Developing and assessing accident management plans for nuclear power plants

    International Nuclear Information System (INIS)

    Hanson, D.J.; Johnson, S.P.; Blackman, H.S.; Stewart, M.A.

    1992-07-01

    This document is the second of a two-volume NUREG/CR that discusses development of accident management plans for nuclear power plants. The first volume (a) describes a four-phase approach for developing criteria that could be used for assessing the adequacy of accident management plans, (b) identifies the general attributes of accident management plans (Phase 1), (c) presents a prototype process for developing and implementing severe accident management plans (Phase 2), and (d) presents criteria that can be used to assess the adequacy of accident management plans. This volume (a) describes results from an evaluation of the capabilities of the prototype process to produce an accident management plan (Phase 3) and (b), based on these results and preliminary criteria included in NUREG/CR-5543, presents modifications to the criteria where appropriate

  10. WASA-BOSS. Development and application of Severe Accident Codes. Evaluation and optimization of accident management measures. Subproject F. Contributions to code validation using BWR data and to evaluation and optimization of accident management measures. Final report; WASA-BOSS. Weiterentwicklung und Anwendung von Severe Accident Codes. Bewertung und Optimierung von Stoerfallmassnahmen. Teilprojekt F. Beitraege zur Codevalidierung anhand von SWR-Daten und zur Bewertung und Optimierung von Stoerfallmassnahmen. Schlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Di Marcello, Valentino; Imke, Uwe; Sanchez Espinoza, Victor

    2016-09-15

    The exact knowledge of the transient course of events and of the dominating processes during a severe accident in a nuclear power station is a mandatory requirement to elaborate strategies and measures to minimize the radiological consequences of core melt. Two typical experiments using boiling water reactor assemblies were modelled and simulated with the severe accident simulation code ATHLET-CD. The experiments are related to the early phase of core degradation in a boiling water reactor. The results reproduce the thermal behavior and the hydrogen production due to oxidation inside the bundle until relocation of material by melting. During flooding of the overheated assembly temperatures and hydrogen oxidation are under estimated. The deviations from the experimental results can be explained by the missing model to simulate bore carbide oxidation of the control rods. On basis of a hypothetical loss of coolant accident in a typical German boiling water reactor the effectivity of flooding the partial degraded core is investigated. This measure of mitigation is efficient and prevents failure of the reactor pressure vessel if it starts before molten material is relocated into the lower plenum. Considerable amount of hydrogen is produced by oxidation of the metallic components.

  11. Use of probabilistic safety analyses in severe accident management

    International Nuclear Information System (INIS)

    Neogy, P.; Lehner, J.

    1991-01-01

    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

  12. Emergency room management of radiation accidents

    International Nuclear Information System (INIS)

    Rosenberg, R.; Mettler, F.A. Jr.

    1990-01-01

    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

  13. Management of Radioactive Waste after a Nuclear Power Plant Accident

    International Nuclear Information System (INIS)

    Strand, Per; Laurent, Gerard; Rindo, Hiroshi; Georges, Christine; Ito, Eiichiro; Yamada, Norikazu; Iablokov, Iuri; Kilochytska, Tatiana; Jefferies, Nick; Byrne, Jim; Siemann, Michael; Koganeya, Toshiyuki; Aoki, Hiroomi

    2016-01-01

    The NEA Expert Group on Fukushima Waste Management and Decommissioning R and D (EGFWMD) was established in 2014 to offer advice to the authorities in Japan on the management of large quantities of on-site waste with complex properties and to share experiences with the international community and NEA member countries on ongoing work at the Fukushima Daiichi site. The group was formed with specialists from around the world who had gained experience in waste management, radiological contamination or decommissioning and waste management R and D after the Three Mile Island and Chernobyl accidents. This report provides technical opinions and ideas from these experts on post-accident waste management and R and D at the Fukushima Daiichi site, as well as information on decommissioning challenges. Chapter 1 provides general descriptions and a short introduction to nuclear accidents or radiological contaminations; for instance the Chernobyl NPP accident, the Three Mile Island Unit 2 accident and the Windscale fire accident. Chapter 2 provides experiences on regulator-implementer interaction in both normal and abnormal situations, including after a nuclear accident. Chapter 3 provides experiences on stakeholder involvement after accidents. These two chapters focus on human aspects after an accident and provide recommendations on how to improve communication between stakeholders so as to resolve issues arising after unexpected nuclear accidents. Chapters 4, 5 and 6 provide information on technical issues related to waste management after accidents. Chapter 4 focuses on the physical and chemical nature of the waste, Chapter 5 on radiological characterisation, and Chapter 6 on waste classification and categorisation. The persons involved in waste management after an accident should address these issues as soon as possible after the accident. Chapters 7 and 8 also focus on technical issues but with a long-term perspective of the waste direction in the future. Chapter 7 relates

  14. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1994-01-01

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

  15. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1991-01-01

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

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

    International Nuclear Information System (INIS)

    Chang, Hyun Sop

    1996-02-01

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

  17. Accident management strategy in Sweden - implementation and verification

    International Nuclear Information System (INIS)

    Loewenhielm, Gustaf; Engqvist, Alf; Espefaelt, Ralf

    1994-01-01

    A comprehensive program for severe accident mitigation was completed in Sweden by the end of 1988. As described in this paper, this program included plant modifications such as the introduction of filtered containment venting, and an accident management system comprising emergency operating strategies and procedures, training and emergency drills. The accident management system at Vattenfall has been further developed since 1988 and some results and experience from this development are reported in this paper. The main aspects covered concern the emergency organization and the supporting tools developed for use by the emergency response teams, the radiological implications such as accessibility to various locations and the long-term aspects of accident management. ((orig.))

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

    International Nuclear Information System (INIS)

    Klopp, G.T.

    1991-01-01

    The current requirements for the performance of individual plant evaluations (IPE's) include the derivation of accident management insights as and if they occur in the course of finalizing an IPE. The development of formal, structured accident management programs is, however, explicitly excluded from current IPE requirements. The Nuclear Regulatory Commission is following the Nuclear Management and Resources Council (NUMARC) efforts to establish the framework(s) for accident management program development and plants to issue requirements on such development at a later date. The Commonwealth Edison program consists of comprehensive level 2 PRA's which address the requirements for IPE's and which go beyond those requirements. From the start of the IPE efforts, it was firmly held, within Edison, that the best way to fully and economically extract a viable accident management program from an IPE was to integrate the two efforts from the start and include the accident management program development as a required IPE product

  19. Big Rock Point severe accident management strategies

    International Nuclear Information System (INIS)

    Brogan, B.A.; Gabor, J.R.

    1996-01-01

    December 1994, the Nuclear Energy Institute (NEI) issued guidance relative to the formal industry position on Severe Accident Management (SAM) approved by the NEI Strategic Issues Advisory Committee on November 4, 1994. This paper summarizes how Big Rock Point (BRP) has and continues to address SAM strategies. The historical accounting portion of this presentation includes a description of how the following projects identified and defined the current Big Rock Point SAM strategies: the 1981 Level 3 Probabilistic Risk Assessment performance; the development of the Plant Specific Technical Guidelines from which the symptom oriented Emergency Operating Procedures (EOPs) were developed; the Control Room Design Review; and, the recent completion of the Individual Plant Evaluation (IPE). In addition to the historical presentation deliberation, this paper the present activities that continue to stress SAM strategies

  20. Level 2 PSA methodology and severe accident management

    International Nuclear Information System (INIS)

    1997-01-01

    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

  1. Using MARS to assist in managing a severe accident

    International Nuclear Information System (INIS)

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

    2004-01-01

    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)

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

    International Nuclear Information System (INIS)

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

    1993-09-01

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

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

    International Nuclear Information System (INIS)

    Frid, W.; Sandervaag, O.

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-09-01

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

  5. PWR severe accident mitigation measures, the french point of view

    International Nuclear Information System (INIS)

    Duco, J.; L'Homme, A.; Queniart, D.

    1990-01-01

    French studies have early considered the fact that, despite all the precautions taken, the possibility of severe accidents cannot be absolutely excluded; these accidents include core meltdown and a more or less significant loss, at an early or later stage, of the confinement of the radioactive substances in the containment. For a given scenario, one can almost always imagine a more severe scenario by postulating additional failures, but it is obvious that, as the severity of the imagined scenario increases, the probability of its occurrence tends towards zero. However, it does not appear reasonable to attempt to set a probability threshold below which the scenarios should be excluded. First of all, the higher the improbability of the scenarios, the greater the uncertainty in the calculation of their probability, with the result that the calculation is not very meaningful. Secondly, and more importantly, this approach ignores the essential problem of accident situation management. From the outset, French studies have been focused on controlling the development of these situations and mitigating their consequences by means of a series of appropriate actions involving, on the one hand, optimum use of the resources available in the installation during the course of the accident and, on the other hand, the taking of protective measures for the population. To attempt to prevent an initial event to degenerate into a severe accident leading to core meltdown if the proper actions are not taken, Electricite de France has proposed a new operating procedure based on the characterization of every possible cooling state of the core

  6. German Phase B [risk study] highlights the role of [reactor] accident management

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    Phase B of the German probabilistic risk assessment study, now scheduled for publication this month, suggests that reactor accident management measures can prevent or mitigate about 90 per cent of event sequences. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

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

  8. Strategy generator in computerized accident management support system

    International Nuclear Information System (INIS)

    Sirola, M.

    1994-02-01

    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.)

  9. The management of individuals involved in radiation accidents

    Energy Technology Data Exchange (ETDEWEB)

    Swindon, T N [Australian Radiation Lab., Melbourne (Australia)

    1991-09-01

    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.

  10. Development of an accident management expert system for containment assessment

    International Nuclear Information System (INIS)

    Nelson, W.R.; Sebo, D.E.; Haney, L.N.

    1987-01-01

    The United States Nuclear Regulatory Commission (USNRSC) is sponsoring a program at the Idaho National Engineering Laboratory (INEL) to develop an accident management expert system. The intended users of the system are the personnel of the NRC Operations Center in Washington, D.C. The expert system will be used to help NRC personnel monitor and evaluate the status and management of the containment during a severe reactor accident. The knowledge base will include severe accident knowledge regarding the maintenance of the critical safety functions, especially containment integrity, during an accident. This paper summarizes the concepts that have been developed for the accident management expert system, and the plans that have been developed for its implementation

  11. Environmental measurements during the TMI-2 accident

    International Nuclear Information System (INIS)

    Hull, A.P.

    1988-01-01

    Although the environmental consequences of the TMI accident were relatively insignificant, it was a major test of the ability of the involved state and federal radiological agencies to make a coordinated environmental monitoring response. This was accomplished largely on an ad hoc basis under the leadership of DOE. With some fine tuning, it is the basis for today's integrated FRMAP monitoring plan, which would be put into operation should another major accident occur at a US nuclear facility

  12. Aerosol measurements and nuclear accidents: a reconsideration

    International Nuclear Information System (INIS)

    Raes, F.

    1988-01-01

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

  13. Accident management for PWRs in France and Germany

    International Nuclear Information System (INIS)

    Heili, F.; Lecomte, C.; L'Homme, A.

    1991-11-01

    The results of risk analyses, research and particularly the two severe accidents in the nuclear power plants TMI-2 and Chernobyl let to a worldwide re-examination of all aspects dealing with the capability to cope with severe accidents. Strategies have been developed or are under development providing actions that can be taken to prevent severe accidents or to mitigate their consequences. Those strategies are investigated and discussed using the term 'accident management'. The purpose of this report is to present the respective views in France and Germany and to point out differences and commonalties of the approaches. This report also includes proposals for further work

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  15. Implementation of accident management programmes in nuclear power plants

    International Nuclear Information System (INIS)

    2004-01-01

    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

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

    Science.gov (United States)

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

    2015-06-01

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

  17. Methodological aspects to elaborate the management and procedure guides of severe accidents

    International Nuclear Information System (INIS)

    Gonzalez Gonzalez, F.; Jimenez Fernandez, A.

    1995-01-01

    The management guides in severe accidents are very important to know the procedures in these accidents. The present articles summarizes the methodological aspects to elaborate the management guides, in order to prevent the severe accidents

  18. Severe accident management guidance for third Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Su Changsong

    2010-01-01

    The paper describes the background, document structure and the summaries of Severe Accident Management Guidance (SAMG) for Third Qinshan Nuclear Power Plant (TQNPP), and also introduces briefly some design features and their impacts on SAMG. (authors)

  19. Strategy-oriented display concept to assist severe accident management

    International Nuclear Information System (INIS)

    Jeong, Kwangsub; Ha, Jaejoo

    2000-01-01

    The Critical Function Monitoring System (CFMS) is a typical Safety Parameter Display System (SPDS) to assist the operation of Korean Standard Nuclear Power Plants during normal and emergency operation, and SPDS for severe accident is being developed in Korea. When the existing CFMS is used under a severe accident situation, some problems are expected from: (1) different design basis, i.e. prevention of core melt vs. protection of radiation release to environment, (2) different parameters for decision-making, and (3) different domain and depth of information to restore the plant. To resolve the above problems, a concept, 'Strategy-Oriented Information Display' concept, for displaying information for severe accident management is developed in this paper. Whereas the existing SPDS structure is based on the critical safety function, the developed concept is based on the severe accident management strategy. The display for each strategy includes the plant parameters to check the status of plant and component with the logical or graphical views necessary for executing the strategy. As the application of the proposed concept, KAERI is developing a display system, the prototype severe accident SPDS, Severe Accident Management Display System (SAMDIS), to assist plant personnel for executing Korean Severe Accident Management Guidelines. CFMS is developed for a general display suitable to all situations with various displays. On the contrary, SAMDIS provides all the relevant information on one screen based on the proposed concept. The SAMDIS screen shows more extensive area than CFMS and thus plant personnel can recognize the overall plant status at a glance. This concept is quite effective when used with severe accident management guidelines because of the relatively macroscopic characteristics of a severe accident management strategy. (author)

  20. Chernobyl accident: Causes, consequences and problems of radiation measurements

    International Nuclear Information System (INIS)

    Kortov, V.; Ustyantsev, Yu.

    2013-01-01

    General description of Chernobyl accident is given in the review. The accident causes are briefly described. Special attention is paid to radiation situation after the accident and radiation measurements problems. Some data on Chernobyl disaster are compared with the corresponding data on Fukushima accident. It is noted that Chernobyl and Fukushima lessons should be taken into account while developing further measures on raising nuclear industry safety. -- Highlights: ► The short comparative analysis of accidents at Chernobyl and Fukushima is given. ► We note the great effect of β-radiation on the radiation situation at Chernobyl. ► We discuss the problems of radiation measurements under these conditions. ► The impact of shelter on the radiation situation near Chernobyl NPS is described

  1. Discussion on several issues of the accidents management of nuclear power plants in operation

    International Nuclear Information System (INIS)

    Cao Xuewu; Wang Zhe; Zhang Yingzhen

    2009-01-01

    This article discusses several issues of the accident management of nuclear power plants in operation, for example: the necessity, implementation principle of accident management and accident management program etc. For conducting accident management for beyond design basis accidents, this article thinks that the accident management program should be developed and implemented to ensure that the plant and its personnel with responsibilities for accident management are adequately prepared to take effective on-site actions to prevent or mitigate the consequences of severe accident. (authors)

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

    International Nuclear Information System (INIS)

    Cizel, F.

    1999-01-01

    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

  3. Recent Perspective on the Severe Accident Management Programme for Nuclear Power Plant

    International Nuclear Information System (INIS)

    Kim, Manwoong; Lee, Sukho; Lee, Jungjae; Chung, Kuyoung

    2017-01-01

    Severe Accident Management Guidelines (SAMGs), has been developed to help operators to prevent or mitigate the impacts of accidents at nuclear power plants. Severe accident management was first introduced in the 1990s with the creation of SAMGs following recognition that post-Three Mile Island Emergency Operating Procedures (EOPs) did not adequately address severe core damage conditions. Establishing and maintaining multiple layers of defence against any internal/external hazards is an important measure to reduce radiological risks to the public and environment. This study is intended to suggest future regulatory perspectives to strengthen the prevention and mitigation strategies for severe accidents by review of the current status of revision of IAEA Safety Standard on Severe Accident Management Programmes for Nuclear Power Plants and the combined PWR SAMG. This new IAEA Safety Guide will address guidelines for preparation, development, implementation and review of severe accident management programs during all operating conditions for both reactor and spent fuel pool. This Guide is used by operating organizations of nuclear power plants and their support organizations. It may also be used by national regulatory bodies and technical support organizations as a reference for developing their relevant safety requirements and for conducting reviews and safety assessments for SAMP including SAMG. The Pressurized Water Reactor Owner’s Group (PWROG) is upgrading the original generic Severe Accident Management Guidelines (SAMGs) into single Severe Accident Guidelines (SAGs) for the PWR SAMG aims to consolidate the advantages of each of the separate vendor severe accident (SA) mitigation methods. This new PWROG SAGs changes the SAMG process to be made that can improve SA response. Changes have been made that guidance is available for control room operators when the TSC is not activated thus allowing for timely accident response. Other changes were made to the guidance

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

    International Nuclear Information System (INIS)

    Klopp, G.T.; Silverman, E.B.

    1992-01-01

    This paper presents the results of the second phase of a three-phase Severe Accident Management expert system program underway at Commonwealth Edison Company (CECo). Phase I successfully demonstrated the feasibility of Artificial Neural Networks to support several of the objectives of severe accident management. Simulated accident scenarios were generated by the Modular Accident Analysis Program (MAAP) code currently in use by CECo as part of their Individual Plant Evaluations (IPE)/Accident Management Program. 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. The results of this work would form the foundation of a demonstration system which included expert system performance features. These capabilities included the ability to: (1) Predict the time available prior to support plate (and reactor vessel) failure; (2) Calculate the time remaining until recovery actions were too late to prevent core damage; (3) Predict future parameter values of each of the MAAP parameter variables; and (4) Detect simulated sensor failure and provide best-value estimates for further processing in the presence of a sensor failure. A variety of accident scenarios for the Zion and Dresden plants were used to train and test the neural network expert system. These included large and small break LOCAs as well as a range of transient events. 3 refs., 1 fig., 1 tab

  5. Regulatory perspective on accident management issues

    International Nuclear Information System (INIS)

    Barrett, R.J.

    1988-01-01

    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

  6. Accidents - Chernobyl accident

    International Nuclear Information System (INIS)

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    Yoneyama, Mitsuru; Masuda, Takahiro; Kubota, Ryuji; Fujiwara, Tadashi; Sakuma, Hitoshi

    2000-01-01

    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)

  8. Passive depressurization accident management strategy for boiling water reactors

    International Nuclear Information System (INIS)

    Liu, Maolong; Erkan, Nejdet; Ishiwatari, Yuki; Okamoto, Koji

    2015-01-01

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

  9. Passive depressurization accident management strategy for boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Maolong, E-mail: liuml@vis.t.u-tokyo.ac.jp [Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo (Japan); Erkan, Nejdet [Nuclear Professional School, School of Engineering, The University of Tokyo (Japan); Ishiwatari, Yuki [Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo (Japan); Hitachi-GE Nuclear Energy, Ltd. (Japan); Okamoto, Koji [Nuclear Professional School, School of Engineering, The University of Tokyo (Japan)

    2015-04-01

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

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

    International Nuclear Information System (INIS)

    2003-05-01

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

  11. Written instructions for the transport of hazardous materials: Accident management instruction sheets

    International Nuclear Information System (INIS)

    Ridder, K.

    1988-01-01

    In spite of the regulations and the safety provisions taken, accidents are not entirely avoidable in the transport of hazardous materials. For managing an accident and preventing further hazards after release of dangerous substances, the vehicle drivers must carry with them the accident management instruction sheets, which give instructions on immediate counter measures to be taken by the driver, and on information to be given to the police and the fire brigades. The article in hand discusses the purpose, the contents, and practice-based improvement of this collection of instruction sheets. Particular reference is given to the newly revised version of June 15, 1988 (Verkehrsblatt 1/88) of the 'Directives for setting up accident management instruction sheets - written instructions - for road transport of hazardous materials', as issued by the Federal Ministry of Transport. (orig./HP) [de

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

    International Nuclear Information System (INIS)

    Kikuchi, Toru

    2014-01-01

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

  13. The DOE technology development programme on severe accident management

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  14. Usage of geotechnologies for risk management in radiation accidents

    International Nuclear Information System (INIS)

    Silva, T.A.A.; Marques, F.A.P.; Murta, Y.L.

    2017-01-01

    Through the use of geotechnologies an important tool can be created for risk management in radiation accidents. With the use of the QGIS software (Las Palmas version), it is shown its applicability in situations of radiological emergency, as in the case of the accident with cesium-137 in Goiânia. The work analyses the risk of a possible accident with the deposit of cesium wastes that still remains in the region, aiming to protect the population with the best exit routes and forms of allocation of the residents

  15. Extension of emergency operating procedures for severe accident management

    International Nuclear Information System (INIS)

    Chiang, S.C.

    1992-01-01

    To enhance the capability of reactor operators to cope with the hypothetical severe accident its the key issue for utilities. Taiwan Power Company has started the enhancement programs on extension of emergency operating procedures (EOPs). It includes the review of existing LOPs based on the conclusions and recommendations of probabilistic risk assessment studies to confirm the operator actions. Then the plant specific analysis for accident management strategy will be performed and the existing EOPs will be updated accordingly

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, W.R.; Hanson, D.J.; Ward, L.W.; Solberg, D.E.

    1988-01-01

    While it would be very difficult to predetermine all of the actions required to mitigate the consequences of every potential severe accident for a nuclear power plant, development of additional guidance and training could improve the likelihood that the operating crew would implement effective sever-accident management measures. The US Nuclear Regulatory Commission (NRC) is conducting an Accident Management Research Program that emphasizes the application of severe-accident research results to enhance the capability of the plant operating crew to effectively manage severe accidents. One element of this program includes identification of the information needed by the operating crew in severe-accident situations. This paper discusses a method developed for identifying these information needs and its application. The methodology has been applied to a generic reactor design representing a PWR with a large dry containment. The information needs were identified by systematically determining what information is needed to assess the health of the critical functions, identify the presence of challenges, select strategies, and assess the effectiveness of these strategies. This method allows the systematic identification of information needs for a broad range of severe-accident scenarios and can be validated by exercising the functional models for any specific event sequence.

  18. Management of older patients presenting after a fall - an accident ...

    African Journals Online (AJOL)

    Background. It is common for older patients to present to accident and emergency (AE) departments after a fall. Management should include assessment and treatment of the injuries and assessment and correction of underlying risk factors in order to prevent recurrent falls. Objectives. To determine management of older ...

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

    International Nuclear Information System (INIS)

    Ang, Ming Leang; Shepherd, Charles; Gauntt, Randall; Landgren, Vickie; Van Dorsselaere, Jean Pierre; Chaumont, Bernard; Raimond, Emmanuel; Magallon, Daniel; Prior, Robert; Mlady, Ondrej; Khatib-Rahbar, Mohsen; Lajtha, Gabor; Tinkler, Charles; Siu, Nathan

    2007-01-01

    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

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

    International Nuclear Information System (INIS)

    Yanke, R.

    2002-01-01

    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

  1. Program for accident and incident management support, AIMS

    International Nuclear Information System (INIS)

    Putra, M.A.

    1993-12-01

    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)

  2. Severe Accident Management System On-line Network SAMSON

    International Nuclear Information System (INIS)

    Silverman, Eugene B.

    2004-01-01

    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 cm 2 ) in size to breaks 3.0 square feet in size (2800 cm 2 ). (author)

  3. Benchmarking MARS (accident management software) with the Browns Ferry fire

    International Nuclear Information System (INIS)

    Dawson, S.M.; Liu, L.Y.; Raines, J.C.

    1992-01-01

    The MAAP Accident Response System (MARS) is a userfriendly computer software developed to provide management and engineering staff with the most needed insights, during actual or simulated accidents, of the current and future conditions of the plant based on current plant data and its trends. To demonstrate the reliability of the MARS code in simulatng a plant transient, MARS is being benchmarked with the available reactor pressure vessel (RPV) pressure and level data from the Browns Ferry fire. The MRS software uses the Modular Accident Analysis Program (MAAP) code as its basis to calculate plant response under accident conditions. MARS uses a limited set of plant data to initialize and track the accidnt progression. To perform this benchmark, a simulated set of plant data was constructed based on actual report data containing the information necessary to initialize MARS and keep track of plant system status throughout the accident progression. The initial Browns Ferry fire data were produced by performing a MAAP run to simulate the accident. The remaining accident simulation used actual plant data

  4. Development of Parameter Network for Accident Management Applications

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. CATHARE Assessment of PACTEL LOCA Experiments with Accident Management

    Directory of Open Access Journals (Sweden)

    Luben Sabotinov

    2010-01-01

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

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

    International Nuclear Information System (INIS)

    2001-01-01

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

  7. Application of NUREG-1150 methods and results to accident management

    International Nuclear Information System (INIS)

    Dingman, S.E.; Sype, T.T.; Camp, A.L.

    1990-01-01

    The risk from five nuclear power plants was examined during the NUREG-1150 program. When the analyses of the plants were complete, an effort was undertaken to examine the implications of NUREG-1150 for accident management initiatives. The framework provided by the NUREG-1150 analysis presented a means within which current accident management strategies could be evaluated and future accident management strategies could be developed and assessed. Five separate but interrelated phases of risk management were considered: (1) prevention of accident initiators, (2) prevention of core damage, (3) implementation of an effective emergency response, (4) prevention of vessel breach and mitigation of radionuclide releases from the reactor coolant system, and (5) retention of fission products in the containment and other surrounding buildings. A risk-based methodology was developed to identify and evaluate risk management options for each of these five phases. The methodology was demonstrated through quantitative examples for the first two phases of risk management listed above. In addition, the reduction in risk for several currently implemented risk management strategies at operating plants was quantified

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

    International Nuclear Information System (INIS)

    Lutz, Robert; Lucci, Melissa; Kiper, Kenneth; Henry, Robert

    2006-01-01

    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)

  9. A framework for assessing severe accident management strategies

    International Nuclear Information System (INIS)

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

    1991-01-01

    Accident management can be defined as the innovative use of existing and or alternative resources, systems and actions to prevent or mitigate a severe accident. 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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

    Accident management can be defined as the innovative use of existing and or alternative resources, systems and actions to prevent or mitigate a severe accident. 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 pressurized water reactor to prevent vessel penetration or failure, and drywell flooding in a boiling water reactor to prevent containment failure

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

    International Nuclear Information System (INIS)

    He Guizhen; Zhang Lei; Lu Yonglong; Mol, Arthur P.J.

    2011-01-01

    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.

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

    International Nuclear Information System (INIS)

    Banz, I.

    1986-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lind, Morten; Zhang Xinxin [Harbin Engineering University, Harbin (China)

    2014-08-15

    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.

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

    International Nuclear Information System (INIS)

    Andre Oudiz

    2006-01-01

    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

  15. OSSA. A second generation of severe accident management

    International Nuclear Information System (INIS)

    Sauvage, E.C.; Musoyan, G.; Ducros, V.D.

    2009-01-01

    Nowadays the severe accident and their management are an integrated part of the new generation of power plants. The EPR, as the third generation of nuclear plants, includes both systems and instrumentation to mitigate a severe accident, but also a new generation of severe accident management guidelines: the OSSA. Severe accident management guidelines are highly dependent on human means available: emergency organization actors, training and knowledge shall be taken in consideration in an innovative way. Their impacts on ergonomy and content of the document lead to a new generation of guidelines with several innovative features. This second generation of severe accident management guidelines was developed in parallel with the PSA level 2, the human reliability analyses, the validation and verification process, the severe accident simulator progresses. By taking in consideration this variety of input the OSSA were developed in a user aspect orientation. For example in the OSSA a larger responsibility is given to the operational crew to better support the technical support group evaluation. Their existing knowledge of the plant and of the systems and instrumentation is used. This collaboration work implies a strong communication tool that has been developed to enhance the permanent communication within the emergency organization, but although to ensure the main up-to-date information for evaluation will be available where required. The entry condition is based on a strong and stand alone diagnostic for all plant states, that uses in particular a curve of core exit temperature as a function of primary pressure for a fixed core cladding temperature, or its equivalent in term of containment conditions. It ensures relatively consistent core conditions on entry. A first criterion for ultimate final primary depressurization is provided, ensuring all attempts to reflood the core with the available means have been ensured before the OSSA entry condition is reached. This

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

    International Nuclear Information System (INIS)

    Yoneyama, Mitsuru; Kubota, Ryuji; Fujiwara, Tadashi; Sakuma, Hitoshi

    1999-01-01

    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)

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

    International Nuclear Information System (INIS)

    Hanson, D.J.; Blackman, H.S.; Meyer, O.R.; Ward, L.W.

    1991-04-01

    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

  18. The evolution of computerized displays in accident management

    International Nuclear Information System (INIS)

    DeBor, J.

    1988-01-01

    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

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

    International Nuclear Information System (INIS)

    Bourrel, M.; Calmon, Ph.; Calvez, M.; Chambrette, V.; Champion, D.; Devin, P.; Godino, O.; Lombard, J.; Rzepka, J.P.; Schneider, Th.; Verhaeghe, B.; Cogez, E.; Kayser, O.; Guenon, C.; Jourdain, J.R.; Bouchot, E.; Murith, Ch.; Lochard, J.; Cluchier, A.; Vandecasteele, Ch.; Pectorin, X.; Dubiau, Ph.; Gerphagnon, O.; Roche, H.; Cessac, B.; Cochard, A.; Machenaud, G.; Jourdain, J.R.; Pirard, Ph.; Leger, M.; Bouchot, E.; Demet, M.; Charre, J.P.; Poumadere, M.; Cogez, E.

    2010-01-01

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

  20. Plant specific severe accident management - the implementation phase

    International Nuclear Information System (INIS)

    Prior, R.

    1999-01-01

    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)

  1. A systems approach to the management of radiation accidents

    International Nuclear Information System (INIS)

    Richter, L.L.; Berk, H.W.; Teates, C.D.; Larkham, N.E.; Friesen, E.J.; Edlich, R.F.

    1980-01-01

    Management of radiation accident patients should have a multidisciplinary approach that includes all health professionals as well as members of public safety agencies. Emergency plans for radiation accidents include detection of the ionizing radiation, patient evacuation, resuscitation, and decontamination. The resuscitated patient should be transported to a radiation control area located outside but adjacent to the emergency department. Ideally this area is accessed through an entrance separate from that used for the main flow of daily emergency department patients. The hospital staff, provided with protective clothing, dosimeters, and preprinted guidelines, continues the resuscitation and definitive care of the patient. This system approach to the management of radiation accidents may be tailored to meet the specific needs of other emergency medical systems

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

    International Nuclear Information System (INIS)

    1995-07-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-15

    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

  4. Development of Integrated Evaluation System for Severe Accident Management

    Energy Technology Data Exchange (ETDEWEB)

    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

    International Nuclear Information System (INIS)

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

    2007-06-01

    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. NPP accident scenario. Which emergency measures are planned in Switzerland?

    International Nuclear Information System (INIS)

    Flury, Christoph

    2016-01-01

    As a consequence of the reactor accident in Fukushima the Swiss government has ordered an extensive analysis of emergency planning in case of a NPP accident Switzerland. A special working group has analyzed the possible improvements of Swiss emergency planning based on the experiences in Japan. Under the special direction of the Bundesamt fuer Bevoelkerungsschutz (BABS) the agreed improvements were integrated into the emergency concept. The reference scenarios have been re-assessed and the zone concept adapted. The emergency measures include shelter-type rooms (basement or window-less rooms), the preventive distribution of iodine pills, measures concerning agriculture, aquatic systems, preventive evacuation, traffic regulations, and delayed evacuation.

  7. Radioactivity measurement for emergency or post-accident situations

    International Nuclear Information System (INIS)

    Champion, D.

    2010-01-01

    Specific objectives have to be achieved by radioactivity measurements during emergency or post-accident situations, which are different from those in normal situation. At the beginning of a nuclear emergency, few radioactivity data will be available, mainly from automatic monitoring systems implemented on the site or in its surrounding. Progressively, measurement programmes will be performed, in priority to get information on dose rate, atmospheric radionuclides and surface activities. In order to avoid excessive exposure of the measure teams, these programmes should be optimized. During early post-accident phase, different types of measurements will be done, following two main objectives: 1) to improve the assessment of the environmental contamination and people exposure; 2) for control purpose, to check the contamination of urban places, foodstuff and other products, compared to specific reference levels. The samples measurement in laboratories would be a challenge: usually, the laboratories involved in routine monitoring have to deal with very low level of radioactivity and a poor diversity of artificial radionuclides; after a reactor accident, the environmental samples to be measured would be more active and with a mixture of radionuclides (mainly with short or middle half-life) difficult to be characterized. So theses laboratories have to be trained and organised before any severe accident. (author)

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

    International Nuclear Information System (INIS)

    Fourrie, Laetitia; Grosjean, Francois; Adam, Didier; Pretet, Caroline; Michel, Aurelie; Fostier, Bernard; Bertrand, Sophie; Cessac, Bruno; Reales, Nicolas IRSN; Aubert, Claude

    2007-05-01

    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

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

    International Nuclear Information System (INIS)

    Reales, N.; Fourrie, L.; Quinio, C.; Grastilleur, Ch.

    2008-01-01

    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. Use of decision trees for evaluating severe accident management strategies in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Jae, Moosung [Hanyang Univ., Seoul (Korea, Republic of). Dept. of Nuclerar Engineering; Lee, Yongjin; Jerng, Dong Wook [Chung-Ang Univ., Seoul (Korea, Republic of). School of Energy Systems Engineering

    2016-07-15

    Accident management strategies are defined to innovative actions taken by plant operators to prevent core damage or to maintain the sound containment integrity. Such actions minimize the chance of offsite radioactive substance leaks that lead to and intensify core damage under power plant accident conditions. Accident management extends the concept of Defense in Depth against core meltdown accidents. In pressurized water reactors, emergency operating procedures are performed to extend the core cooling time. The effectiveness of Severe Accident Management Guidance (SAMG) became an important issue. Severe accident management strategies are evaluated with a methodology utilizing the decision tree technique.

  11. First aid and subsequent measures after radiation accidents

    International Nuclear Information System (INIS)

    Flach, H.D.

    1980-01-01

    An organisation schedule and first aid measures after accidents involving ionizing radiation are presented, both in accordance with the current practice of the responsible professional associations. Optimum care also of persons with radiation injuries will be assured by cooperation between voluntary lay helpers, physicians, and regional centres in which failure symptons of the haematopoietic system can be treated. (DG) [de

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

    Energy Technology Data Exchange (ETDEWEB)

    Kuznetsov, M., E-mail: kuznetsov_mv@vosafety.ru [FSUE VO “Safety”, Moscow (Russian Federation)

    2014-10-15

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

  13. An operational centre for managing major chemical industrial accidents.

    Science.gov (United States)

    Kiranoudis, C T; Kourniotis, S P; Christolis, M; Markatos, N C; Zografos, K G; Giannouli, I M; Androutsopoulos, K N; Ziomas, I; Kosmidis, E; Simeonidis, P; Poupkou, N

    2002-01-28

    The most important characteristic of major chemical accidents, from a societal perspective, is their tendency to produce off-site effects. The extent and severity of the accident may significantly affect the population and the environment of the adjacent areas. Following an accident event, effort should be made to limit such effects. Management decisions should be based on rational and quantitative information based on the site specific circumstances and the possible consequences. To produce such information we have developed an operational centre for managing large-scale industrial accidents. Its architecture involves an integrated framework of geographical information system (GIS) and RDBMS technology systems equipped with interactive communication capabilities. The operational centre was developed for Windows 98 platforms, for the region of Thriasion Pedion of West Attica, where the concentration of industrial activity and storage of toxic chemical is immense within areas of high population density. An appropriate case study is given in order to illuminate the use and necessity of the operational centre.

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

    International Nuclear Information System (INIS)

    Dubiau, Ph.

    2007-01-01

    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.)

  15. Guide on medical management of persons exposed in radiation accidents

    International Nuclear Information System (INIS)

    1990-01-01

    The present guide has been prepared in order to provide guidance to medical and para-medical personnel regarding medical management of the different types of radiation accidents. It discusses briefly the physical aspects and biological effect of radiation, for the benefit of those who have not specialised in radiation medicine. The diagnosis, medical management and follow-up of persons involved in different types of radiation accidents are also dealt with. The implementation of the procedures described calls for organisation of appropriate facilities and provision of requisite equipment as well as education and training of the staff. It is emphasised that major radiation accidents are rare events and the multi-disciplinary nature of the response required to deal with them calls for proper planning and continuous liaison among plant management, radiation protection personnel, first-aid assistants and medical and paramedical staff. The organisation and conduct of emergency drills may help in maintaining preparedness of the medical facilities for efficient management of radiation casualities. (original). 64 refs., tabs., figs

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

    Science.gov (United States)

    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.

  17. Development of Severe Accident Management Strategies for Shin-Kori 3 and 4

    International Nuclear Information System (INIS)

    Lee, Youngseung; Kim, Hyeongtaek; Shin, Jungmin

    2013-01-01

    Shin-Kori units 3 and 4 are new reactors under construction as an APR 1400 type reactor. The plants which considered coping with severe accident from design phase are different from other operating plants in view of severe accident management strategies. The purpose of this paper is to establish optimal strategies for Shin-Kori 3 and 4. A scheme for optimized severe accident management was drawn up with the object of achieving core cooling, containment integrity, and decreased release of fission product. Shin-Kori units 3 and 4 are a new reactor and designed to add mitigating systems for coping with severe accident such as ECSBS, PAR, and CFS. Also the plants are reflected as a part of Fukushima followup measures The strategies of SAMG for Shin-Kori 3 and 4 were developed. The strategic approach was based on the concept of defense in depth. Firstly, strategies for core cooling were chosen such as RCS depressurization, injection to SG, injection to RCS, and injection to reactor cavity. Secondly, the plans for containment integrity were developed for controlling pressure and hydrogen in containment. Lastly, reduced release of fission product was considered for protection of the public after containment failure. The achieved strategies meet the needs of effective methods for severe accident management and enhancement of safety

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

    International Nuclear Information System (INIS)

    Sonnenkalb, Martin

    1992-07-01

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

  19. Examination of offsite emergency protective measures for core melt accidents

    International Nuclear Information System (INIS)

    Aldrich, D.C.; McGrath, P.E.; Ericson, D.M. Jr.; Jones, R.B.; Rasmussen, N.C.

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

  20. Contamination measurements on persons after a nuclear accident

    International Nuclear Information System (INIS)

    Maushart, R.

    1992-01-01

    The purpose of contamination measurements after accidents is threefold: to detect and localize contaminations; to determine the level of contamination as a base for medical decisions; and to check the scope and efficiency of the decontamination measures. Persons involved in accidents should never measure their contamination themselves. The radiation protection personnel, whether it belongs to the medical team or to the disaster control squads, must be familiar with the measuring instruments and experienced in handling these instruments. A high priority has to be placed on training and constant practice. The monitors used must meet the special requirements of an emergency situation. This includes resistance against environmental factors - temperature, humidity, vibrations - as well as the simplicity of handling the instrument, and the clear presentation of the results. (author)

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

    International Nuclear Information System (INIS)

    Soejoedi, Agoes; Karmana, Endang

    2000-01-01

    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

  2. Development of the severe accident management guidance module for the SATS training simulator

    International Nuclear Information System (INIS)

    Kim, K. R.; Park, S. H.; Kim, D. H.

    2004-01-01

    Recently KAERI has developed severe accident management guidance to establish Korea standard severe accident management system. On the other hand PC-based severe accident training simulator SATS has been developed, which uses MELCOR computing code as the simulation engine. SATS graphically displays and simulates the severe accident progression with interactive user inputs. The control capability of SATS makes a severe accident training course more interesting and effective. In this paper the development and functions of HyperKAMG module are explained. Furthermore easiness and effectiveness of the HyperKAMG-SATS system in severe accident management are described

  3. Major accident prevention through applying safety knowledge management approach.

    Science.gov (United States)

    Kalatpour, Omid

    2016-01-01

    Many scattered resources of knowledge are available to use for chemical accident prevention purposes. The common approach to management process safety, including using databases and referring to the available knowledge has some drawbacks. The main goal of this article was to devise a new emerged knowledge base (KB) for the chemical accident prevention domain. The scattered sources of safety knowledge were identified and scanned. Then, the collected knowledge was formalized through a computerized program. The Protégé software was used to formalize and represent the stored safety knowledge. The domain knowledge retrieved as well as data and information. This optimized approach improved safety and health knowledge management (KM) process and resolved some typical problems in the KM process. Upgrading the traditional resources of safety databases into the KBs can improve the interaction between the users and knowledge repository.

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

    International Nuclear Information System (INIS)

    Klopp, G.T.

    1992-01-01

    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. WASA-BOSS. Development and application of Severe Accident Codes. Evaluation and optimization of accident management measures. Subproject E. Improvement of the lower head model in MELCOR and calculations in connection with the FUKUSHIMA accident. Final report; WASA-BOSS. Weiterentwicklung und Anwendung von Severe Accident Codes. Bewertung und Optimierung von Stoerfallmassnahmen. Teilprojekt E. Verbesserung des Lower Head-Modelles fuer MELCOR und MELCOR-Rechnungen zu Fukushima. Schlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Kretzschmar, Frank; Dietrich, Philipp; Gabriel, Stephan; Miassoedov, Alexei

    2016-12-15

    The knowledge of the key phenomena, which govern the chronological sequence of a core melt accident, is a crucial precondition for the development of SAMGs (Severe Management Guides) to avoid and mitigate the radiological consequences for the population and the environment. In the frame of a dissertation a new model has been coupled with MELCOR, which describes the thermal interaction of a core melt with the RPV (reactor pressure vessel) wall in the lower plenum. This model allows a better description of this phenomenon. The method to couple extern programs with MELCOR had been already developed and used in a former dissertation at KIT-IKET. The model has been validated recalculating according experiments in the LIVE facility. Afterwards a defined accident scenario has been calculated for a German generic KONVOI power plant. 12 months after the start of the project a MELCOR input has been developed using data delivered by the Ruhr university of Bochum (subproject ''Simulation des Unfalls in Fukushima-Daichi zur Bewertung des Stoerfall-Analysecodes ATHLET-CD''). The results of this simulation have made a contribution to review the current understanding of the FUKUSHIMA sequence. HZDR and KIT-IKET have agreed in the course of the project, that KIT-IKET will develop a MELCOR input of a german generic KONVOI power plant following an ATHLET-CDinput of HZDR. Using this MELCOR input, a comparative analysis has been performed.

  6. Application of high-order uncertainty for severe accident management

    International Nuclear Information System (INIS)

    Yu, Donghan; Ha, Jaejoo

    1998-01-01

    The use of probability distribution to represent uncertainty about point-valued probabilities has been a controversial subject. Probability theorists have argued that it is inherently meaningless to be uncertain about a probability since this appears to violate the subjectivists' assumption that individual can develop unique and precise probability judgments. However, many others have found the concept of uncertainty about the probability to be both intuitively appealing and potentially useful. Especially, high-order uncertainty, i.e., the uncertainty about the probability, can be potentially relevant to decision-making when expert's judgment is needed under very uncertain data and imprecise knowledge and where the phenomena and events are frequently complicated and ill-defined. This paper presents two approaches for evaluating the uncertainties inherent in accident management strategies: 'a fuzzy probability' and 'an interval-valued subjective probability'. At first, 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 above to evaluate imprecise node probabilities in the influence diagram. For the propagation of subjective probabilities, the analysis uses the Monte-Carlo simulation. In case of fuzzy probabilities, the fuzzy logic is applied to propagate them. We believe that these approaches can allow us to understand uncertainties associated with severe accident management strategy since they offer not only information similar to the classical approach using point-estimate values but also additional information regarding the impact from imprecise input data

  7. Pending issues for severe accident management in Wolsong plants

    International Nuclear Information System (INIS)

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

    2015-01-01

    While the fraction of electric power supplied from a PHWR is more than 10% in Korea, the establishment of PHWR safety enhancement based on the SAM (Severe Accident Management) technology is still weak. The final approval on the extended operation and a stress test of Wolsong-1 were made under the condition that SAM is to be enhanced. Under this situation, the current research at KAERI of Korea has a vision to strengthen the unique value of a PHWR by resolving the pending SAM issues devaluating the PHWRs’ original value. Research activities in this area will be presented. This presentation will include: The operating strategy of CFVS (Containment Filtered Vent System) for Wolsong in which vent size and closure pressure are treated because some peak spikes (at failure times of calandria and calandria vault) are difficult to be controlled; Reactor Building failure pressure at which failure probability is treated for different modes such as global and leak failures; the adequacy of DCRV (Degasser Condenser tank Relief Valve) steam relief capacity with severe SGTR source term, and Hydrogen generation and control issue which is specific to CANDU. Furthermore, current SAM guidance has a lack of information on accident diagnostic and prognostic analyses, which is difficult for the TSC (Technical Service Center) emergency staff members to deal with under real accident conditions. Thus, prototypic technologies (such as an accident inferring engine and simulator) together with SAM updates are being developed as key elements to SAM supporting tools called SAMEX-CANDU

  8. A Bayesian ensemble of sensitivity measures for severe accident modeling

    Energy Technology Data Exchange (ETDEWEB)

    Hoseyni, Seyed Mohsen [Department of Basic Sciences, East Tehran Branch, Islamic Azad University, Tehran (Iran, Islamic Republic of); Di Maio, Francesco, E-mail: francesco.dimaio@polimi.it [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Vagnoli, Matteo [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Zio, Enrico [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Chair on System Science and Energetic Challenge, Fondation EDF – Electricite de France Ecole Centrale, Paris, and Supelec, Paris (France); Pourgol-Mohammad, Mohammad [Department of Mechanical Engineering, Sahand University of Technology, Tabriz (Iran, Islamic Republic of)

    2015-12-15

    Highlights: • We propose a sensitivity analysis (SA) method based on a Bayesian updating scheme. • The Bayesian updating schemes adjourns an ensemble of sensitivity measures. • Bootstrap replicates of a severe accident code output are fed to the Bayesian scheme. • The MELCOR code simulates the fission products release of LOFT LP-FP-2 experiment. • Results are compared with those of traditional SA methods. - Abstract: In this work, a sensitivity analysis framework is presented to identify the relevant input variables of a severe accident code, based on an incremental Bayesian ensemble updating method. The proposed methodology entails: (i) the propagation of the uncertainty in the input variables through the severe accident code; (ii) the collection of bootstrap replicates of the input and output of limited number of simulations for building a set of finite mixture models (FMMs) for approximating the probability density function (pdf) of the severe accident code output of the replicates; (iii) for each FMM, the calculation of an ensemble of sensitivity measures (i.e., input saliency, Hellinger distance and Kullback–Leibler divergence) and the updating when a new piece of evidence arrives, by a Bayesian scheme, based on the Bradley–Terry model for ranking the most relevant input model variables. An application is given with respect to a limited number of simulations of a MELCOR severe accident model describing the fission products release in the LP-FP-2 experiment of the loss of fluid test (LOFT) facility, which is a scaled-down facility of a pressurized water reactor (PWR).

  9. Molten Corium-Concrete Interaction Behavior Analyses for Severe Accident Management in CANDU Reactor

    International Nuclear Information System (INIS)

    Choi, Y.; Kim, D. H.; Song, Y. M.

    2014-01-01

    After the last few severe accidents, the importance of accident management in nuclear power plants has increased. Many countries, including the United States (US) and Canada, have focused on understanding severe accidents in order to identify ways to further improve the safety of nuclear plants. It has been recognized that severe accident analyses of nuclear power plants will be beneficial in understanding plant-specific vulnerabilities during severe accidents. The objectives of this paper are to describe the molten corium behavior to identify a plant response with various concrete specific components. Accident analyses techniques using ISSAC can be useful tools for MCCI behavior in severe accident mitigation

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

    International Nuclear Information System (INIS)

    Okamoto, Koji; Sugiyama, Tomoyuki; Kamata, Shinya

    2014-01-01

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

  11. Early measurements in urban areas after the Chernobyl accident

    International Nuclear Information System (INIS)

    Likhtarev, I.

    2000-01-01

    This paper summarises the experience on the radioactive monitoring of the environment and population dose assessment provided in urban areas, mainly in Kiev, after the Chernobyl accident. It emphasises the need of several radiological teams, of the support from several institutions and of preparedness for a consistent database, dose assessment and criteria for decision making. Main results of measurements of gamma exposure rates, air, grass and food radioactive contamination are presented. (author)

  12. Analysis of the reasons of recently some radioactive source accidents and suggestions for management countermeasures

    International Nuclear Information System (INIS)

    Su Yongjie; Feng Youcai; Song Chenxiu; Gao Huibin; Xing Jinsong; Pang Xinxin; Wang Xiaoqing; Wei Hong

    2007-01-01

    The article introduces recently some radioactive source accidents in China, and analyses the reasons of the accidents. Some important issues existed in the process of implementing new regulation were summarized, and some suggestions for managing radioactive sources are made. (authors)

  13. A Study on the Requisite Information for Severe Accident Management

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sunhee; Ahn, Kwang-Il; Kim, Jae-Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Related this research on arranging the requisite information for severe accident management, the documents of various forms in each country as well as the domestic literature are secured and analyzed. The analyzed information is arranged up to a detailed level. For the secured documents, the issued organizations and the issued purpose are diverse. Thus, the contents of the secured documents are also diverse according to the reactor type, and the purpose and standards of the classification are also diverse. Moreover, terminologies with same meaning are not unified. These various documents are analyzed to arrange the requisite information for severe accident management. Based on the documents of a related severe accident, the major information was analyzed. The information is different according to the reactor type, classification standard, and classification standard of the safety function. Thus the information is classified variously. In this study, based on the analysis results of the documents described these information, the major information and parameters are examined as safety function. And the results of parameters and information including the safety function and the detail information are induced.

  14. Influence of boron reduction strategies on PWR accident management flexibility

    International Nuclear Information System (INIS)

    Papukchiev, Angel Aleksandrov; Liu, Yubo; Schaefer, Anselm

    2007-01-01

    In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. Design changes to reduce boron concentration in the reactor coolant are of general interest regarding three aspects - improved reactivity feedback properties, lower impact of boron dilution scenarios on PWR safety and eventually more flexible accident management procedures. In order to assess the potential advantages through the introduction of boron reduction strategies in current PWRs, two low boron core configurations based on fuel with increased utilization of gadolinium and erbium burnable absorbers have been developed. The new PWR designs permit to reduce the natural boron concentration in reactor coolant at begin of cycle to 518 ppm and 805 ppm. For the assessment of the potential safety advantages of these cores a hypothetical beyond design basis accident has been simulated with the system code ATHLET. The analyses showed improved inherent safety and increased accident management flexibility of the low boron cores in comparison with the standard PWR. (author)

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

    International Nuclear Information System (INIS)

    2002-01-01

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

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

    International Nuclear Information System (INIS)

    1995-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    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

  18. PWR accident management realated tests: some Bethsy results

    International Nuclear Information System (INIS)

    Clement, P.; Chataing, T.; Deruaz, R.

    1993-01-01

    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

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

    International Nuclear Information System (INIS)

    1995-01-01

    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

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

    International Nuclear Information System (INIS)

    Vasilevskij, V.P.; Nikitin, Yu.M.; Petrov, A.A.; Potapov, A.A.; Cherkashov, Yu.M.

    2001-01-01

    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 [ru

  1. Environmental radioactivity measurements at BNL following the Chernobyl accident

    International Nuclear Information System (INIS)

    Hancock, R.; Woollam, P.B.

    1986-06-01

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

  2. Consideration of Command and Control Performance during Accident Management Process at the Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Ahmed, Nisrene M. [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Kim, Sok Chul [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-10-15

    The accident at the Fukushima Daiichi nuclear power plants shifted the nuclear safety paradigm from risk management to on-site management capability during a severe accident. The kernel of on-site management capability during an accident at a nuclear power plant is situation awareness and agility of command and control. However, little consideration has been given to accident management. After the events of September 11, 2001 and the catastrophic Fukushima nuclear disaster, agility of command and control has emerged as a significant element for effective and efficient accident management, with many studies emphasizing accident management strategies, particularly man-machine interface, which is considered a key role in ensuring nuclear power plant safety during severe accident conditions. This paper proposes a conceptual model for evaluating command and control performance during the accident management process at a nuclear power plant. Communication and information processing while responding to an accident is one of the key issues needed to mitigate the accident. This model will give guidelines for accurate and fast communication response during accident conditions.

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

    International Nuclear Information System (INIS)

    2000-03-01

    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

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

    International Nuclear Information System (INIS)

    Tayloe, R.W. Jr.; D'Aquila, D.M.; McGinnis, R.B.

    1991-01-01

    The nuclear criticality accident radiation alarm system installed at the Portsmouth Gaseous Diffusion Plant was tested extensively at critical facilities located at the Los Alamos National Laboratory. The ability of the neutron scintillator radiation detection units to respond to a minimum accident of concern as defined in Standard ANSI/ANS-83.-1986 was demonstrated. Detector placement and the established trip point are based on shielding calculations performed by the Oak Ridge National Laboratory and criticality specialists at the Portsmouth plant. Based on these experiments and calculations, a detector trip point of 5 mrad/h in air is used. Any credible criticality accident is expected to produce neutron radiation fields >5 mrad/h in air at one or more radiation alarm locations. Each radiation alarm location has a cluster of three detectors that employs a two-out-of-three alarm logic. Earlier work focused on testing the alarm logic latching circuitry. This work was directed toward measurements involving the actual audible alarm signal delivered

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

    International Nuclear Information System (INIS)

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

    2005-03-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. Investigation of safety measures to severe accident of Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    So as to plan the accident management to severe accident of Fast Breeder Reactor (FBR), it is primary important to understand the progression of severe accident (SA) precisely. In this study, it has been aimed to reveal two items that work as keys in the evaluation of SA in sodium cooled FBR. One is the cool-ability of degraded core on the core support plate by sodium natural circulation in the post accident heat removal (PAHR) phase. An obstacle that hinders the smooth heat transfer from fuel debris to coolant is the formation of sodium-uranate by chemical reaction between sodium and fuel. Following the measurement of physical values of sodium-uranate in FY 2011, experiments has been performed to reveal the conditions for sodium-uranate formation on fuel debris in sodium pool simulating the actual situation of the degraded core. The cool-ability of the debris bed was analyzed using the Lipinski 1-D model. Another research performed in this study is the measurement of fission product (cesium and antimony) evaporation rates from FBR fuel as a function of temperature, because presently the fission product evaporation rates data for LWR is also temporarily used for FBR SA analysis. The measurement was performed using the irradiated fuels in the Test Reactor JOYO. (author)

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

    International Nuclear Information System (INIS)

    1994-01-01

    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

  9. Technical, organizational and human-centered requirements for the purpose of accident management

    International Nuclear Information System (INIS)

    Berning, A.; Fassmann, W.; Preischl, W.

    1998-01-01

    A catalog of ergonomic recommendations for organizational measures and design of paper documented work aids for accident management situations in nuclear power plants was developed. Attention was given to provide recommendations meeting practical needs and being sufficiently flexible to allow plant specific [aptation. A weight was assigned to each recommendation indicating its importance. The development of the recommendations was based on the state of the art concerning research and practical experience. Results from walk-/talk-through experiments, training and exercises, discussions with on-site experts, and investigations of emergency manuals from German and foreign nuclear power plants were taken into account. The catalog is founded on a bro[ knowledge base covering important aspects. The catalog is intended for qualitative evaluation and design of organizational measures and procedures. The catalog shall assure high quality. The project further provides an important contribution to the standardization of organizational and human centered demands concerning accident management procedures. Thus it can contribute to develop general regulations regarding ergonomic design of accident management measures. (orig.) [de

  10. The PSI Artist Project: Aerosol Retention and Accident Management Issues Following a Steam Generator Tube Rupture

    International Nuclear Information System (INIS)

    Guntay, Salih; Dehbi, Abdel; Suckow, Detlef; Birchley, Jon

    2002-01-01

    Steam generator tube rupture (SGTR) incidents, such as those, which occurred in various operating pressurized, water reactors in the past, are serious operational concerns and remain among the most risk-dominant events. Although considerable efforts have been spent to understand tube degradation processes, develop improved modes of operation, and take preventative and corrective measures, SGTR incidents cannot be completely ruled out. Under certain conditions, high releases of radionuclides to the environment are possible during design basis accidents (DBA) and severe accidents. The severe accident codes' models for aerosol retention in the secondary side of a steam generator (SG) have not been assessed against any experimental data, which means that the uncertainties in the source term following an un-isolated SGTR concurrent with a severe accident are not currently quantified. The accident management (AM) procedures aim at avoiding or minimizing the release of fission products from the SG. The enhanced retention of activity within the SG defines the effectiveness of the accident management actions for the specific hardware characteristics and accident conditions of concern. A sound database on aerosol retention due to natural processes in the SG is not available, nor is an assessment of the effect of management actions on these processes. Hence, the effectiveness of the AM in SGTR events is not presently known. To help reduce uncertainties relating to SGTR issues, an experimental project, ARTIST (Aerosol Trapping In a Steam generator), has been initiated at the Paul Scherrer Institut to address aerosol and droplet retention in the various parts of the SG. The test section is comprised of a scaled-down tube bundle, a full-size separator and a full-size dryer unit. The project will study phenomena at the separate effect and integral levels and address AM issues in seven distinct phases: Aerosol retention in 1) the broken tube under dry secondary side conditions, 2

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

    International Nuclear Information System (INIS)

    2003-03-01

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

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

    International Nuclear Information System (INIS)

    Dukelow, J.S.; Harrison, D.G.; Morgenstern, M.

    1992-03-01

    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

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

    International Nuclear Information System (INIS)

    Barisic, D.; Lulic, S.

    2000-01-01

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

  14. Accident management: What is it and how do you do it?

    International Nuclear Information System (INIS)

    Henry, Robert E.; Hammersley, Robert J.

    2004-01-01

    Accident management is the composite of those actions that would prevent, stop and/or mitigate a severe accident in a nuclear power plant. Since they act to prevent core damage, the Emergency Operating Procedures (EOPs) are an integral part of accident management. Each of the Owners Groups have developed EOPs that are well thought out for instructing the operator to respond to accident conditions which could threaten the core. However, for those very low probability events in which the core could be uncovered and damaged, accident management actions arise from a logical evaluation of possible actions (strategies) for recovering from the accident state and protecting the public health and safety. To understand the character of accident management it is first necessary to define: 1. What is threatened as a result of the accident? 2. Fundamentally, what needs to be protected? 3. What is known during an accident? 4. What have we learned from the TMI-2 accident? 5. What have we learned from the plant specific IPEs? Once these subjects are reviewed on a utility specific and plant specific basis, accident management actions become relatively straightforward and likely can be effectively addressed using the total capability available in a given design. This paper discusses these five questions in a global manner with the aim being to aid plant specific implementation. (author)

  15. Developement of integrated evaluation system for severe accident management

    Energy Technology Data Exchange (ETDEWEB)

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

    International Nuclear Information System (INIS)

    Cotten, G.B.; Jenkins, S.L.

    1997-01-01

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

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

    International Nuclear Information System (INIS)

    Bracht, K.; Friedrichs, H.G.

    1989-01-01

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

  18. Measures for preventing and mitigating severe accidents of nuclear power plants

    International Nuclear Information System (INIS)

    Lin Chengge

    1993-01-01

    Safety goals, integrity of the containment, accident management, functions of existing equipment and measures and emergency preparedness are discussed as technical basis for implementing the new safety code on the nuclear power plant safety design (HAF-0200(91)). The main quantitative safety goals are presented as core melt frequency -5 /ry for new plants and -4 /ry for existing or constructed plants, and 0.1% I, Cs release frequency -6 /ry. To keep the integrity of the containment, main efforts should be placed on the prevention of early failure of the containment and by pass or isolation failures. Should a late failure of the containment occur at a high probability, measures such as filtering vent should be considered. The leak rate of the containment could be higher than the previous 0.1-0.5 wt%/day, depending on the source term and dose results. But, a limiting leak rate of 1 wt%/day is defined. Accident management involves emergency operating procedures, training and retraining for the AM and adding some supporting equipment and display and diagnostic system for the AM. Those requirements are described. Emergency preparedness and measures can reduced the risk significantly. In the most case of accidents, sheltering is preferred as an effective protective actions

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

    International Nuclear Information System (INIS)

    Ahn, Kwang Il; Kim, Dong Ha

    2003-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Kwang Il; Kim, Dong Ha

    2003-01-01

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

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

    International Nuclear Information System (INIS)

    Morichi, M.; Toubon, H.; Bronson, Frazier; Venkataraman, Ram; Beaujoin, J.; Dubart, P.

    2013-06-01

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

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

    International Nuclear Information System (INIS)

    1992-01-01

    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

  3. Analysis of emergency response to fukushima nuclear accident in Japan and suggestions for China's nuclear emergency management

    International Nuclear Information System (INIS)

    Li Wei; Ding Qihua; Wu Haosong

    2014-01-01

    On March 11, 2011, the Fukushima Dai-ichi Nuclear Power Station of the Tokyo Electric Power Company ('TEPCO') was hit and damaged by a magnitude 9 earthquake and accompanying tsunami. The accident is determined to be of the highest rating on the International Nuclear Event Scale. The Government of Japan and TEPCO have taken emergency response actions on-site and off-site at the accident. It became clear through the investigation that the accident had been initiated on the occasion of a natural disaster of an earthquake and tsunami, but there have been various complex problems behind this very serious and large scale accident. For an example, the then-available accident preventive measures and disaster preparedness of TEPCO were insufficient against tsunami and severe accidents; inadequate TEPCO emergency responses to the accident at the site were also identified. The accident rang the alarm for the nuclear safety of nuclear power plants. It also taught us a great of lessons in nuclear emergency management. (authors)

  4. Development of PSA module for computerized accident management support (CAMS)

    International Nuclear Information System (INIS)

    Iguchi, Yukihiro

    1996-10-01

    CAMS (Computerised Accident Management Support) is a system that will provide assistance in case of the accidents in a nuclear power plant. The PSA module was developed in order to give useful information in this situation applying the PSA method, which is a comprehensive source of safety knowledge. This module contains plant-specific PSA data, comprising event trees, failure probabilities etc. It has several event trees categorised according to the initiating events. Each event tree has an initiating event frequency and branching probabilities. The various support systems for branches are considered and their dependencies are calculated logically. This module can be activated by data from the state identification (SI) module of CAMS. If an initiating event occurs, the event tree is re-calculated and the PSA module shows which systems of the plant should be activated to bring the plant to a safe state. If the plant responds to the event in the normal way, the plant will be shut down and come to a safe state. However, if some functions do not work, the PSA module generates another path and gives information about the critical systems. If the state of the plant is changed, either by the operators or automatically by the control system, the PSA module follows the new path. Because the estimation of the core damage frequency should be very quick in the accident situation, a simplified model of the event tree and fault trees was adopted. It enabled the PSA module to calculates the CDF within 5 seconds on a standard type work station. The development of the module has been successful. However, further development of the functionality of the module is suggested like real connection to a plant and to the strategy generator module of CAMS, applications for operational support, low power operation optimisation, etc. (author)

  5. Characteristics of Hydrogen Monitoring Systems for Severe Accident Management at a Nuclear Power Plant

    Science.gov (United States)

    Petrosyan, V. G.; Yeghoyan, E. A.; Grigoryan, A. D.; Petrosyan, A. P.; Movsisyan, M. R.

    2018-02-01

    One of the main objectives of severe accident management at a nuclear power plant is to protect the integrity of the containment, for which the most serious threat is possible ignition of the generated hydrogen. There should be a monitoring system providing information support of NPP personnel, ensuring data on the current state of a containment gaseous environment and trends in its composition changes. Monitoring systems' requisite characteristics definition issues are considered by the example of a particular power unit. Major characteristics important for proper information support are discussed. Some features of progression of severe accident scenarios at considered power unit are described and a possible influence of the hydrogen concentration monitoring system performance on the information support reliability in a severe accident is analyzed. The analysis results show that the following technical characteristics of the combustible gas monitoring systems are important for the proper information support of NPP personnel in the event of a severe accident at a nuclear power plant: measured parameters, measuring ranges and errors, update rate, minimum detectable concentration of combustible gas, monitoring reference points, environmental qualification parameters of the system components. For NPP power units with WWER-440/270 (230) type reactors, which have a relatively small containment volume, the update period for measurement results is a critical characteristic of the containment combustible gas monitoring system, and the choice of monitoring reference points should be focused not so much on the definition of places of possible hydrogen pockets but rather on the definition of places of a possible combustible mixture formation. It may be necessary for the above-mentioned power units to include in the emergency operating procedures measures aimed at a timely heat removal reduction from the containment environment if there are signs of a severe accident phase

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

    International Nuclear Information System (INIS)

    1998-06-01

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

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

    International Nuclear Information System (INIS)

    Oewre, Fridtjov

    2001-01-01

    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

  8. Accident beyond the design basis management with the coolant loss at the NPP with WWER

    International Nuclear Information System (INIS)

    Skalozubov, V.I.; Klyuchnikov, A.A.; Kolykhanov, V.N.

    2010-01-01

    The analysis of status and experience of development on modelling and accident beyond the design basis management, including the severe accidents, at the nuclear power plants is carried out. The methodical providing of manuals on the accident beyond the design basis management with the coolant loss on the basis of simulated critical system configurations providing the necessary safety function performance on reactor unit is proposed. The project of symptom-oriented manuals on accident beyond the design basis management with the coolant loss on the serial power unit with WWER-1000 on the basis of developed methodical providing and well known results of deepened safety analysis is presented.

  9. Development of severe accident management guidance for Younggwang units 5 and 6

    International Nuclear Information System (INIS)

    Lee, K. W.; Beon, C. S.; Kim, M. K.; Hong, S. Y.; Park, K. S.

    2001-01-01

    Severe Accident Management Guidance (SAMG) has been developed for Younggwang Units 5 and 6. It is consisted of Severe Accident Control Room Guideline, Diagnostic Flow Chart, Severe Accident Guideline, Severe Challenge Guideline, TSC Long Term Monitoring, SAMG Termination. Severe Accident Control Room Guideline, which deals with severe accident after finishing Emergency Operation Procedure, consists of acitions before and after TSC actuation. Seven servere accident management strategies are developed. Diagnostic Flow Chart, Severe Accident Guideline, and Severe Challenge Guideline are developed for each strategy, which enables the users to the implementation of strategy easily and systematically. TSC Long Term Monitoring is also developed to monitor long term activities after a particular strategy. Total of 45 set points are developed for decision making during the implementation of the SAMG

  10. The management of risk to society from potential accidents

    International Nuclear Information System (INIS)

    Allen, F.R.; Garlick, A.R.; Hayns, M.R.; Taig, A.R.

    1992-01-01

    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)

  11. Influence diagrams and decision trees for severe accident management

    Energy Technology Data Exchange (ETDEWEB)

    Goetz, W.W.J.

    1996-09-01

    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.).

  12. Influence diagrams and decision trees for severe accident management

    International Nuclear Information System (INIS)

    Goetz, W.W.J.; Seebregts, A.J.; Bedford, T.J.

    1996-08-01

    A review of relevent 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 as complementary to CET models. (orig.)

  13. Influence diagrams and decision trees for severe accident management

    International Nuclear Information System (INIS)

    Goetz, W.W.J.

    1996-09-01

    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.)

  14. Protection measures taken by the Japanese Authorities following the Fukushima accident

    International Nuclear Information System (INIS)

    Gallay, F.

    2013-01-01

    This article reviews 6 different sides of the Fukushima accident: 1) the emergency measures like massive evacuation of people and the taking of iodine tablets, 2) the emergency medical aid for contaminated people and the medical monitoring of workers and population, 3) the setting of restricted areas in the region around the Fukushima plant, 4) the monitoring of food, 5) the implementation of measures to curb the contamination levels (decontamination of buildings, the removal of the superficial layer of the soil, and the ploughing of some plots of land), and 6) the management of wastes

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

    International Nuclear Information System (INIS)

    2011-01-01

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

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

    Science.gov (United States)

    Yamashita, S

    2016-04-01

    Five years have passed since the Great East Japan Earthquake and the subsequent Fukushima Daiichi Nuclear Power Plant accident on 11 March 2011. Countermeasures aimed at human protection during the emergency period, including evacuation, sheltering and control of the food chain were implemented in a timely manner by the Japanese Government. However, there is an apparent need for improvement, especially in the areas of nuclear safety and protection, and also in the management of radiation health risk during and even after the accident. Continuous monitoring and characterisation of the levels of radioactivity in the environment and foods in Fukushima are now essential for obtaining informed consent to the decisions on living in the radio-contaminated areas and also on returning back to the evacuated areas once re-entry is allowed; it is also important to carry out a realistic assessment of the radiation doses on the basis of measurements. Until now, various types of radiation health risk management projects and research have been implemented in Fukushima, among which the Fukushima Health Management Survey is the largest health monitoring project. It includes the Basic Survey for the estimation of external radiation doses received during the first 4 months after the accident and four detailed surveys: thyroid ultrasound examination, comprehensive health check-up, mental health and lifestyle survey, and survey on pregnant women and nursing mothers, with the aim to prospectively take care of the health of all the residents of Fukushima Prefecture for a long time. In particular, among evacuees of the Fukushima Nuclear Power Plant accident, concern about radiation risk is associated with psychological stresses. Here, ongoing health risk management will be reviewed, focusing on the difficult challenge of post-disaster recovery and resilience in Fukushima. Copyright © 2016 The Royal College of Radiologists. Published by Elsevier Ltd. All rights reserved.

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-10-01

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

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

    International Nuclear Information System (INIS)

    Honcarenko, R.

    1999-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Bocharova Irina Nikolaevna

    2013-05-01

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

  1. Coal Mine Accidents in Bangladesh: Its Causes and Remedial Measures

    Directory of Open Access Journals (Sweden)

    MD. MINHAJ UDDIN MONIR

    2012-12-01

    C, water flow 600 m3/hr and humidity 100%. Due to roof-fall, toxic gas and very high temperature and humidity in the face 1105 and 1108, some miners were highly injured and died. Therefore, taking necessary safety and risk measures during development of face and roadway in BCM, production will be sustained by controlling roof-fall. Strata pressure of every faces is monitor regularly through electromagnetic emission detector as well as changing sufficient ventilation system, accidents of BCM could be controlled. Adding earth filling materials (e.g. sand, silt etc. in to the vacuum goaf area, air blast accident should be reduced and surface subsidence would also be controlled. Considering all the problems of 1st slice and initiating extra caution for face development accident in BCM may reduce in 2nd slice and overall production may be increased significantly.

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

    International Nuclear Information System (INIS)

    Tayloe, R.W. Jr.; McGinnis, B.

    1990-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

    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)

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

    International Nuclear Information System (INIS)

    Yang, Soo Hyung; Kim, Soo Hyung; Jeong, Young Hoon; Chang, Soon Heung

    1997-01-01

    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

  6. Chernobyl post-accident management: the ETHOS project.

    Science.gov (United States)

    Dubreuil, G H; Lochard, J; Girard, P; Guyonnet, J F; Le Cardinal, G; Lepicard, S; Livolsi, P; Monroy, M; Ollagnon, H; Pena-Vega, A; Pupin, V; Rigby, J; Rolevitch, I; Schneider, T

    1999-10-01

    ETHOS is a pilot research project supported by the radiation protection research program of the European Commission (DG XII). The project provides an alternative approach to the rehabilitation of living conditions in the contaminated territories of the CIS in the post-accident context of Chernobyl. Initiated at the beginning of 1996, this 3-y project is currently being implemented in the Republic of Belarus. The ETHOS project involves an interdisciplinary team of European researchers from the following institutions: the Centre d'etude sur l'Evaluation de la Protection dans le domaine Nucleaire CEPN (radiological protection, economics), the Institute National d'Agronomie de Paris-Grignon INAPG (agronomy, nature & life management), the Compiegne University of Technology (technological and industrial safety, social trust), and the Mutadis Research Group (sociology, social risk management), which is in charge of the scientific co-ordination of the project. The Belarussian partners in the ETHOS project include the Ministry of Emergencies of Belarus as well as the various local authorities involved with the implementation site. The ETHOS project relies on a strong involvement of the local population in the rehabilitation process. Its main goal is to create conditions for the inhabitants of the contaminated territories to reconstruct their overall quality of life. This reconstruction deals with all the day-to-day aspects that have been affected or threatened by the contamination. The project aims at creating a dynamic process whereby acceptable living conditions can be rebuilt. Radiological security is developed in the ETHOS project as part of a general improvement in the quality of life. The approach does not dissociate the social and the technical dimensions of post-accident management. This is so as to avoid radiological risk assessment and management being reduced purely to a problem for scientific experts, from which local people are excluded, and to take into

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-04-15

    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.

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

    International Nuclear Information System (INIS)

    1992-04-01

    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

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

    International Nuclear Information System (INIS)

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    Asmolov, V.G.

    1997-01-01

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

  11. Medical management of the victims of Goiana 137 Cs accident

    International Nuclear Information System (INIS)

    Brandao-Mello, C.E.; Farina, R.; Oliveira, A.R.; Valverde, N.J.

    1989-01-01

    In September 1987, a radiation accident happened in Gioania due to the theft of a shielded source of 137 Cs. Fifty persons were involved showing whole-body and local acute irradiation and also external and/or internal contamination mainly due to ingestion or absorption of 137 Cs. Fourteen of fifty developed severe bone marrow depression characterized by neutropenia and thrombocytopenia. None were submitted to bone marrow transplantation. Eight of fourteen received intravenously GM-CSF. Four of the fourteen died due to hemorrhage and infection. For those with significant internal contamination, Prussian blue was administered (1.5 to 10.0 gm/per day). Besides blue, other measures were taken to increase decorporation of 137 Cs including diuretics, water overload and ergometric exercises. Radiation induced skin injuries were observed in 28 patients and were treated conservatively. 6 patients required surgical intervention. (author)

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

    International Nuclear Information System (INIS)

    2005-04-01

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

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

    International Nuclear Information System (INIS)

    Ogino, Masao; Kawabe, Ryuhei; Nagasaka, Hideo; Sumida, Susumu; Fukasawa, Masanori; Muta, Hitoshi

    2011-01-01

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

  14. Assessment of generic accident management strategies considered for near term implementation

    International Nuclear Information System (INIS)

    Lehner, J.R.; Luckas, W.J.; Vandenkieboom, J.J.

    1989-01-01

    The US Nuclear Regulatory Commission (NRC) and the industry are both participating in the identification of measures that can prevent the progression of a severe accident or mitigate its consequences. Information important for evaluating these accident management strategies for specific plants is expected to result from the ongoing Individual Plant Evaluation (IPE) program. However, NRC staff have identified a number of generic strategies which may not have to await the results of the IPE program and therefore can be considered for earlier implementation. The NRC requested two of its contractors, Brookhaven National Laboratory (BNL) and Battelle Pacific Northwest Laboratories (PNL) to evaluate these strategies. The twenty one candidate strategies fall under three broad global strategies: (1) conserving and replenishing limited resources, (2) use of systems/components in innovative applications, and (3) defeating interlocks and component protective trips in emergencies. Some strategies apply to BWRs or PWRs only, other apply to both types of plants. This paper describes the evaluation of the strategies performed by Brookhaven National Laboratory. Brookhaven National Laboratory assessed the proposed strategies by first detailing the objective of the strategy and listing the actions involved in the implementation. A description of the plant systems associated with the strategy was given. Next, the applicability of existing rules or plant procedures to a particular strategy was investigated. This was accomplished by a fairly detailed, but by no means exhaustive review of the emergency operating procedures of several plants, as well as utility and NRC reports related to accident management

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

    DEFF Research Database (Denmark)

    Lind, Morten; Zhang, Xinxin

    2014-01-01

    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...

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

    DEFF Research Database (Denmark)

    Lind, Morten; Zhang, Xinxin

    2014-01-01

    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....

  17. Nuclear accidents and safety measures of domestic nuclear power plants

    International Nuclear Information System (INIS)

    Song Zurong; Che Shuwei; Pan Xiang

    2012-01-01

    Based on the design standards for the safety of nuclear and radiation in nuclear power plants, the three accidents in the history of nuclear power are analyzed. And the main factors for these accidents are found out, that is, human factors and unpredicted natural calamity. By combining the design and operation parameters of domestic nuclear plants, the same accidents are studied and some necessary preventive schemes are put forward. In the security operation technology of domestic nuclear power plants nowadays, accidents caused by human factors can by prevented completely. But the safety standards have to be reconsidered for the unpredicted neutral disasters. How to reduce the hazard of nuclear radiation and leakage to the level that can be accepted by the government and public when accidents occur under extreme conditions during construction and operation of nuclear power plants must be considered adequately. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Bittermann, Dietmar; Eckardt, Bernd A.; Lechleuthner, Michael [Framatome ANP GmbH, Erlangen (Germany)

    2003-04-01

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

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

    International Nuclear Information System (INIS)

    Wang Yuexing

    1999-01-01

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

  20. Aspects of risk analysis application to estimation of nuclear accidents and tests consequences and intervention management

    International Nuclear Information System (INIS)

    Demin, V.F.; Hedemann-Jensen, P.; Rolevich, I.V.; Schneider, T.S.; Sobolev, B.G.

    1996-01-01

    For assessment of accident consequences and a post-accident management a risk analysis methodology and data bank (BARD) with allowance for radiation and non-radiation risk causes should be developed and used. Aspects of these needs and developments are considered. Some illustrative results of health risk estimation made with BARD for the Bryansk region territory with relatively high radioactive contamination from the Chernobyl accident are presented

  1. Main results of assessing integrity of RNPP-3 steam generator heat exchange tubes in accident management

    International Nuclear Information System (INIS)

    Shugajlo, Al-j P.; Mustafin, M.A.; Shugajlo, Al-r P.; Ryzhov, D.I.; Zhabin, O.I.

    2017-01-01

    Tubes integrity evaluation under accident conditions considering drain of SG and current technical state of steam exchange tubes is an important question regarding SG long-term operation and improvement of accident management strategy.The main investigation results prepared for heat exchange surface of RNPP-3 steam generator are presented in this research aimed at assessing integrity of heat exchange tubes under accident conditions, which lead to full or partial drain of heat exchange surface, in particular during station blackout.

  2. Management, administrative and operational causes of the accident: Chernobyl nuclear power station

    International Nuclear Information System (INIS)

    Anastas, G.

    1996-01-01

    Full text: The Chernobyl accident, which occurred in April 1986, was the result of management, administrative, operational, technical and design flaws. The accident released millions of curies of mixed fission products (including 70-100 P Bq of 137 Cs). The results of this study strongly suggest that the cultural, political, managerial and operational attributes of the Soviet 'system' performed in a synergistic manner to significantly contribute to the initiation of the accident. At the time of the accident, science, engineering and safety in the former Soviet Union were dominated by an atmosphere of politics, group think and 'dingoes tending the sheep'

  3. Management, administrative and operational causes of the accident: Chernobyl nuclear power station

    International Nuclear Information System (INIS)

    Anastas, G.

    1996-01-01

    The Chernobyl accident, which occurred in April 1986, was the result of management, administrative, operational, technical and design flaws. The accident released millions of curies of mixed fission products including 70-100 PBq of 137 Cs. At the time of the accident, science, engineering and safety in the former Soviet Union were dominated by an atmosphere of politics, group think and 'dingoes tending the sheep'. This corrupted safety culture exacerbated the poor design of the reactor. The results of this study strongly suggest that the cultural, political, managerial and operational attributes of the Soviet 'system' performed in a synergistic manner to significantly contribute to the initiation of the accident. (authors)

  4. US nuclear industry approach to severe accident management guidance development and implementation

    International Nuclear Information System (INIS)

    Modeen, D.; Walsh, L.; Oehlberg, R.

    1992-01-01

    The purpose of this paper is to discuss the US nuclear industry activities, occurring under the auspices of Nuclear Management and Resources Council (NUMARC), to define, develop and implement enhancements to utility accident management capabilities. This effort consists of three major parts: (1) Development of a practical framework for evaluation of plant-specific accident management capabilities and the subsequent implementation of selected enhancements. (2) Development of specific technical guidance that address arresting core damage if it begins, either in-vessel or ex-vessel, and maintaining containment integrity. Preventing inadequate core cooling or minimizing the consequences of offsite releases, while considered to be candidate areas for accident management enhancements, have been the subject of intense previous study and development. (3) Plant-specific implementation of accident management enhancements in three areas: (a) personnel resources (organization, training, communications); (b) systems and equipment (restoration and repair, instrumentation, use of alternatives); and (c) information resources (procedures and guidance, technical information, process information)

  5. U.S. nuclear industry approach to severe accident management guidance, development and implementation

    International Nuclear Information System (INIS)

    Modeen, D.; Walsh, L.; Oehlberg, R.

    1991-01-01

    The purpose of this paper is to discuss the US nuclear industry activities, occurring under the auspices of Nuclear Management and Resources Council (NUMARC), to define, develop and implement enhancements to utility accident management capabilities. This effort consists of three major parts: (1) Development of a practical framework for evaluation of plant-specific accident management capabilities and the subsequent implementation of selected enhancements. (2) Development of specific technical guidance that address arresting core damage if it begins, either in-vessel or ex-vessel, and maintaining containment integrity. Preventing inadequate core cooling or minimizing the consequences of offsite releases, while considered to be candidate areas for accident management enhancements, have been the subject of intense previous study and development. (3) Plant-specific implementation of accident management enhancements in three areas: (a) personnel resources (organization, training, communications); (b) systems and equipment (restoration and repair, instrumentation, use of alternatives); and (c) information resources (procedures and guidance, technical information, process information)

  6. Development of A Methodology for Assessing Various Accident Management Strategies Using Decision Tree Models

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Nam Yeong; Kim, Jin Tae; Jae, Moo Sung [Hanyang University, Seoul (Korea, Republic of); Jerng, Dong Wook [Chung-Ang University, Seoul (Korea, Republic of)

    2016-05-15

    The purpose of ASP (Accident Sequence Precursor) analysis is to evaluate operational accidents in full power and low power operation by using PRA (Probabilistic Risk Assessment) technologies. The awareness of the importance of ASP analysis has been on rise. The methodology for ASP analysis has been developed in Korea, KINS (Korea Institute of Nuclear Safety) has managed KINS-ASP program since it was developed. In this study, we applied ASP analysis into operational accidents in full power and low power operation to quantify CCDP (Conditional Core Damage Probability). To reflect these 2 cases into PRA model, we modified fault trees and event trees of the existing PRA model. Also, we suggest the ASP regulatory system in the conclusion. In this study, we reviewed previous studies for ASP analysis. Based on it, we applied it into operational accidents in full power and low power operation. CCDP of these 2 cases are 1.195E-06 and 2.261E-03. Unlike other countries, there is no regulatory basis of ASP analysis in Korea. ASP analysis could detect the risk by assessing the existing operational accidents. ASP analysis can improve the safety of nuclear power plant by detecting, reviewing the operational accidents, and finally removing potential risk. Operator have to notify regulatory institute of operational accident before operator takes recovery work for the accident. After follow-up accident, they have to check precursors in data base to find similar accident.

  7. An evaluation of the Davis-Besse loss of feedwater event (June 1985) from an accident management perspective

    International Nuclear Information System (INIS)

    Di Salvo, R.; Leonard, M.T.; Wreathall, J.

    1986-01-01

    An accident management perspective is used to analyze events associated with a total loss-of-feedwater at the Davis-Besse nuclear power plant in June 1985. The relationships of accident management to the closely associated concepts of risk management and emergency management are delineated. The analysis shows that the principal contributors to the event's occurrence were shortcomings in risk management. Successful performance by the operators in accident management was principally responsible for terminating the event without consequence to public health

  8. Comparison of Management Oversight and Risk Tree and Tripod-Beta in Excavation Accident Analysis

    Directory of Open Access Journals (Sweden)

    Mohamadfam

    2015-01-01

    Full Text Available Background Accident investigation programs are a necessary part in identification of risks and management of the business process. Objectives One of the most important features of such programs is the analysis technique for identifying the root causes of accidents in order to prevent their recurrences. Analytical Hierarchy Process (AHP was used to compare management oversight and risk tree (MORT with Tripod-Beta in order to determine the superior technique for analysis of fatal excavation accidents in construction industries. Materials and Methods MORT and Tripod-Beta techniques were used for analyzing two major accidents with three main steps. First, these techniques were applied to find out the causal factors of the accidents. Second, a number of criteria were developed for the comparison of the techniques and third, using AHP, the techniques were prioritized in terms of the criteria for choosing the superior one. Results The Tripod-Beta investigation showed 41 preconditions and 81 latent causes involved in the accidents. Additionally, 27 root causes of accidents were identified by the MORT analysis. Analytical hierarchy process (AHP investigation revealed that MORT had higher priorities only in two criteria than Tripod-Beta. Conclusions Our findings indicate that Tripod-Beta with a total priority of 0.664 is superior to MORT with the total priority of 0.33. It is recommended for future research to compare the available accident analysis techniques based on proper criteria to select the best for accident analysis.

  9. Measures for speed management.

    NARCIS (Netherlands)

    2009-01-01

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

  10. ASTEC code development, validation and applications for severe accident management within the CESAM European project - 15392

    International Nuclear Information System (INIS)

    Van Dorsselaere, J.P.; Chatelard, P.; Chevalier-Jabet, K.; Nowack, H.; Herranz, L.E.; Pascal, G.; Sanchez-Espinoza, V.H.

    2015-01-01

    ASTEC, jointly developed by IRSN and GRS, is considered as the European reference code since it capitalizes knowledge from the European research on the domain. The CESAM project aims at its enhancement and extension for use in severe accident management (SAM) analysis of the nuclear power plants (NPP) of Generation II-III presently under operation or foreseen in near future in Europe, spent fuel pools included. Within the CESAM project 3 main types of research activities are performed: -) further validation of ASTEC models important for SAM, in particular for the phenomena being of importance in the Fukushima-Daichi accidents, such as reflooding of degraded cores, pool scrubbing, hydrogen combustion, or spent fuel pools behaviour; -) modelling improvements, especially for BWR or based on the feedback of validation tasks; and -) ASTEC applications to severe accident scenarios in European NPPs in order to assess prevention and mitigation measures. An important step will be reached with the next major ASTEC V2.1 version planned to be delivered in the first part of 2015. Its main improvements will concern the possibility to simulate in details the core degradation of BWR and PHWR and a model of reflooding of severely degraded cores. A new user-friendly Graphic User Interface will be available for plant analyses

  11. The philosophy of severe accident management in the US

    International Nuclear Information System (INIS)

    Baratta, A.J.

    1990-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-12-01

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

  13. Medical management of patients overexposed to irradiation in acute and subacute accidents

    Energy Technology Data Exchange (ETDEWEB)

    Genyao, Ye; Guilin, Wang; Shimin, Huang; Xiyuan, Cheng; Bingzhi, Mao; Yingqi, Li [North Taiping Road Hospital, Beijing, BJ (China)

    1991-11-01

    The authors summarize the experiences in the medical management of 7 radiation accidents resulting in 5 cases of acute radiation sickness, 3 cases of subacute radiation sickness and 8 cases of overexposed persons with estimated physical doses below 1 Gy in the past 5 years. The therapeutic measures including the application of antiradiation drugs such as estriol and herbal medicine '208', measures to improve the microcirculation and transfusion of fetal liver cells, etc., were emphasized in the treatment of acute radiation sickness. For subacute radiation sickness, large doses of stanozolum (18 mg/d) in combination with 654-2 (60 mg/d) were administered as the principal therapeutic agents. The advances in the medical management of acute accidentally overexposed cases in China are briefly reviewed and discussed.

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

    International Nuclear Information System (INIS)

    Copeland, W.J.; Crumm, A.T.; Kearnaghan, D.P.; Rabin, M.S.; Rossi, D.E.

    1994-07-01

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

  15. Accident management information needs for a BWR with a MARK I containment

    Energy Technology Data Exchange (ETDEWEB)

    Chien, D.N.; Hanson, D.J. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1991-05-01

    In support of the US Nuclear Regulatory Commission Accident Management Research Program, information needs during severe accidents have been evaluated for Boiling Water Reactors (BWRs) with MARK 1 containments. This evaluation was performed using a methodology that identifies plant information needs necessary for personnel to: (a) diagnose that an accident is in progress, (b) select and implement strategies to prevent or mitigate the accident, and (c) monitor the effectiveness of these strategies. The information needs and capabilities identified are intended to form a basis for more comprehensive information needs assessments. The assessments will be performed during the analysis and development of specific strategies, which will be used in accident management prevention and mitigation. 3 refs., 4 figs., 2 tabs.

  16. Accident management information needs for a BWR with a MARK I containment

    International Nuclear Information System (INIS)

    Chien, D.N.; Hanson, D.J.

    1991-05-01

    In support of the US Nuclear Regulatory Commission Accident Management Research Program, information needs during severe accidents have been evaluated for Boiling Water Reactors (BWRs) with MARK 1 containments. This evaluation was performed using a methodology that identifies plant information needs necessary for personnel to: (a) diagnose that an accident is in progress, (b) select and implement strategies to prevent or mitigate the accident, and (c) monitor the effectiveness of these strategies. The information needs and capabilities identified are intended to form a basis for more comprehensive information needs assessments. The assessments will be performed during the analysis and development of specific strategies, which will be used in accident management prevention and mitigation. 3 refs., 4 figs., 2 tabs

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

    International Nuclear Information System (INIS)

    Jikia, D.

    2009-01-01

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

  18. Simulator drills for the management of severe accidents

    International Nuclear Information System (INIS)

    Hoffmann, E.

    1989-01-01

    The present state of deliberations on the simulation of severe accidents is presented and applied to a training philosophy. The special characteristics of 'severe' accidents are addressed and, falling under this category, the 'psychological structure of the man-machine-situation' is examined. The valid rules for drilling 'post-RESA-conduct' (RESA = fast reactor shut down) and the monitoring of safety goals are introduced. 2 figs., 1 tab

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

    International Nuclear Information System (INIS)

    Lee, Sang Hun; Kang, Hyun Gook

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  1. The five essential ('key') elements of severe accident management. To be developed as part of a SAMG industry standard

    International Nuclear Information System (INIS)

    Vayssier, George

    2017-01-01

    The Fukushima-Daiichi accident has caused a renewed interest in tools and guidelines to mitigate severe accidents. Notably, industry approaches have been reviewed and features added from the lessons learned. The various severe accident management approaches vary considerably: they have different measures, different priorities for the various actions, different staff responsibilities and different sorts of communication to the off-site authorities. It appears that there is no common basis from which the approaches have been developed. In this paper, the five elements are treated which the author considers essential for proper tools to terminate severe accidents and mitigate their consequences. These five elements should be trained in well-developed drills/exercises, involving all functions of accident management. An industrial standard to define a minimum common basis, to which individual approaches should adhere and so decrease the large scatter in these approaches present now.

  2. The five essential ('key') elements of severe accident management. To be developed as part of a SAMG industry standard

    Energy Technology Data Exchange (ETDEWEB)

    Vayssier, George [NSC Netherlands, Hansweert (Netherlands)

    2017-07-15

    The Fukushima-Daiichi accident has caused a renewed interest in tools and guidelines to mitigate severe accidents. Notably, industry approaches have been reviewed and features added from the lessons learned. The various severe accident management approaches vary considerably: they have different measures, different priorities for the various actions, different staff responsibilities and different sorts of communication to the off-site authorities. It appears that there is no common basis from which the approaches have been developed. In this paper, the five elements are treated which the author considers essential for proper tools to terminate severe accidents and mitigate their consequences. These five elements should be trained in well-developed drills/exercises, involving all functions of accident management. An industrial standard to define a minimum common basis, to which individual approaches should adhere and so decrease the large scatter in these approaches present now.

  3. Post-Accident Waste Management in Ukraine : Challenges and Steps Needed to Resolve the Accident Waste Problem

    International Nuclear Information System (INIS)

    Kilochytska, T.; Zinkevich, L.; Shybetskyi, I.; Krone, J.

    2016-01-01

    Conclusions: • Solving the problem of management of “Chernobyl waste” requires: - a systematic review of the existing norms and regulations with respect to best international practice of radioactive waste management; - improvement of the existing classification / characterization of radioactive waste with a focus on disposal; - improvement of the safety assessment approaches for licensing the disposal of accident waste, taking into account relevant safety features and site specific conditions; - involving of international experience and support to plan and perform safety related activity on the Shelter Object transformation

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

    International Nuclear Information System (INIS)

    Ang, M.L.; Peers, K.; Kersting, E.; Fassmann, W.; Tuomisto, H.; Lundstroem, P.; Helle, M.; Gustavsson, V.; Jacobsson, P.

    2001-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Miele, A; Bebaron-Jacobs, L

    2005-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Miele, A.; Bebaron-Jacobs, L

    2005-07-01

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

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

    International Nuclear Information System (INIS)

    Nabelssi, B.; Mueller, C.; Roglans-Ribas, J.; Folga, S.; Tompkins, M.; Jackson, R.

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    1997-01-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-08-15

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

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

    International Nuclear Information System (INIS)

    Mikhalevich, A.; Grebenkov, A.

    2000-01-01

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

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

    International Nuclear Information System (INIS)

    Duranova, T.

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Krajnc, B.; Basic, I.; Spiler, J.

    2001-01-01

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

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

    International Nuclear Information System (INIS)

    Fruchard, Y.; Lavie, J.M.

    1966-01-01

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

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

    NARCIS (Netherlands)

    He, G.; Zhang, L.; Lu, Y.; Mol, A.P.J.

    2011-01-01

    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

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

    International Nuclear Information System (INIS)

    Putman, V.L.

    1995-09-01

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

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

    International Nuclear Information System (INIS)

    Deworm, J.P.

    1987-01-01

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

  19. Description of Various Factors Contributing to Traffic Accidents in Youth and Measures Proposed to Alleviate Recurrence

    Directory of Open Access Journals (Sweden)

    Ludovic Gicquel

    2017-06-01

    Full Text Available Traffic accidents are the leading cause of hospitalization in adolescence, with the 18–24-year-old age group accounting for 23% of deaths by traffic accidents. Recurrence rate is also high. One in four teenagers will have a relapse within the year following the first accident. Cognitive impairments known in adolescence could cause risky behaviors, defined as repetitive engagement in dangerous situations such as road accidents. Two categories of factors seem to be associated with traffic accidents: (1 factors specific to the traffic environment and (2 “human” factors, which seem to be the most influential. Moreover, the establishment of a stronger relation to high speed driving increases traffic accident risks and can also be intensified by sensation seeking. Other factors such as substance use (alcohol, drugs, and “binge drinking” are also identified as risk factors. Furthermore, cell phone use while driving and attention deficit disorder with or without hyperactivity also seem to be important risk factors for car accidents. The family environment strongly influences a young person’s driving behavior. Some interventional driving strategies and preventive measures have reduced the risk of traffic accidents among young people, such as the graduated driver licensing program and advertising campaigns. So far, few therapeutic approaches have been implemented. Reason why, we decided to set up an innovative strategy consisting of a therapeutic postaccident group intervention, entitled the ECARR2 protocol, to prevent recurrence among adolescents and young adults identified at risk, taking into account the multiple risk factors.

  20. Description of Various Factors Contributing to Traffic Accidents in Youth and Measures Proposed to Alleviate Recurrence.

    Science.gov (United States)

    Gicquel, Ludovic; Ordonneau, Pauline; Blot, Emilie; Toillon, Charlotte; Ingrand, Pierre; Romo, Lucia

    2017-01-01

    Traffic accidents are the leading cause of hospitalization in adolescence, with the 18-24-year-old age group accounting for 23% of deaths by traffic accidents. Recurrence rate is also high. One in four teenagers will have a relapse within the year following the first accident. Cognitive impairments known in adolescence could cause risky behaviors, defined as repetitive engagement in dangerous situations such as road accidents. Two categories of factors seem to be associated with traffic accidents: (1) factors specific to the traffic environment and (2) "human" factors, which seem to be the most influential. Moreover, the establishment of a stronger relation to high speed driving increases traffic accident risks and can also be intensified by sensation seeking. Other factors such as substance use (alcohol, drugs, and "binge drinking") are also identified as risk factors. Furthermore, cell phone use while driving and attention deficit disorder with or without hyperactivity also seem to be important risk factors for car accidents. The family environment strongly influences a young person's driving behavior. Some interventional driving strategies and preventive measures have reduced the risk of traffic accidents among young people, such as the graduated driver licensing program and advertising campaigns. So far, few therapeutic approaches have been implemented. Reason why, we decided to set up an innovative strategy consisting of a therapeutic postaccident group intervention, entitled the ECARR2 protocol, to prevent recurrence among adolescents and young adults identified at risk, taking into account the multiple risk factors.

  1. The post-accident protective measures in the region of the Chernobyl catastrophe

    International Nuclear Information System (INIS)

    Pasechnikov, A.

    1993-01-01

    The strategy of protective measures and the system of intervention levels are considered. It makes it possible to limit the region where post-accident measures must be taken. The concept of 'safe living' for life and surface contamination concept are discussed. The measurement results of a surface contamination are pointed out in the maps for series settlements of the 30 km zone. The features of severe accidents to nuclear power plants are that damage is caused not only by destruction and downtime of power installations, but also by radioactive contamination of the environment. Therefore, the term 'severe accident' is accepted to mean an accident with an off-site impact, which requires to perform large-scale and expensive work on elimination of the consequences of the accident. The whole off-site damage due to the Chernobyl accident is caused exclusively by contamination, as no destruction was observed beyond the site. As a result of the Chernobyl accident the greatest short-term releases of radioactive materials to the atmosphere occurred from a single source. Four elements from all the materials released from the core have determined the short-term and long-term radiological situation in the affected areas. These are iodine, cesium, strontium and plutonium. Moreover, in the releases there were highly - radioactive fragments of fuel (hot particles). 7 figs

  2. Measures taken by electric power companies for the accident at JCO

    International Nuclear Information System (INIS)

    Taniguchi, Kazufumi

    2000-01-01

    Activities for the JCO accident were done by 700 electric company personnel/day at the maximum, who were mainly the experts of radiation management and were from the whole Japanese regions. The major activities were as follows. Monitoring of body surface contamination of people at the refuge, which was done on request from Tokaimura, showed no radioactive contamination over detection limit. Activities based on the request from the national counter-planning headquarters involved environmental monitoring, which only showed the higher ambient dose rate than usual at the boundary of JCO site. Monitoring of toy equipments in kindergartens and primary and junior schools showed no contamination over detection limit. Measurement of dose equivalent of the personnel showed 0.0 mSv exposure by γ-, β-ray and neutron. In summary, questions and requests raised by people mainly concerned the influence of radiation exposure. (K.H.)

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

    International Nuclear Information System (INIS)

    Mueller, C.

    1994-01-01

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

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

    Science.gov (United States)

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

    2011-01-01

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

  5. Support calculations for management of PRISE leakage accidents

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  6. Support calculations for management of PRISE leakage accidents

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  7. The design of PSB-VVER experiments relevant to accident management

    International Nuclear Information System (INIS)

    Del Nevo, Alessandro; D'auria, Francesco; Mazzini, Marino; Bykov, Michael; Elkin, Ilya V.; Suslov, Alexander

    2008-01-01

    Experimental programs carried-out in integral test facilities are relevant for validating the best estimate thermal-hydraulic codes, which are used for accident analyses, design of accident management procedures, licensing of nuclear power plants, etc. The validation process, in fact, is based on well designed experiments. It consists in the comparison of the measured and calculated parameters and the determination whether a computer code has an adequate capability in predicting the major phenomena expected to occur in the course of transient and/or accidents. University of Pisa was responsible of the numerical design of the 12 experiments executed in PSB-VVER facility, operated at Electrogorsk Research and Engineering Center (Russia), in the framework of the TACIS 2.03/97 Contract 3.03.03 Part A, EC financed. The paper describes the methodology adopted at University of Pisa, starting form the scenarios foreseen in the final test matrix until the execution of the experiments. This process considers three key topics: a) the scaling issue and the simulation, with unavoidable distortions, of the expected performance of the reference nuclear power plants; b) the code assessment process involving the identification of phenomena challenging the code models; c) the features of the concerned integral test facility (scaling limitations, control logics, data acquisition system, instrumentation, etc.). The activities performed in this respect are discussed, and emphasis is also given to the relevance of the thermal losses to the environment. This issue affects particularly the small scaled facilities and has relevance on the scaling approach related to the power and volume of the facility. (author)

  8. The Design of PSB-VVER Experiments Relevant to Accident Management

    Science.gov (United States)

    Nevo, Alessandro Del; D'Auria, Francesco; Mazzini, Marino; Bykov, Michael; Elkin, Ilya V.; Suslov, Alexander

    Experimental programs carried-out in integral test facilities are relevant for validating the best estimate thermal-hydraulic codes(1), which are used for accident analyses, design of accident management procedures, licensing of nuclear power plants, etc. The validation process, in fact, is based on well designed experiments. It consists in the comparison of the measured and calculated parameters and the determination whether a computer code has an adequate capability in predicting the major phenomena expected to occur in the course of transient and/or accidents. University of Pisa was responsible of the numerical design of the 12 experiments executed in PSB-VVER facility (2), operated at Electrogorsk Research and Engineering Center (Russia), in the framework of the TACIS 2.03/97 Contract 3.03.03 Part A, EC financed (3). The paper describes the methodology adopted at University of Pisa, starting form the scenarios foreseen in the final test matrix until the execution of the experiments. This process considers three key topics: a) the scaling issue and the simulation, with unavoidable distortions, of the expected performance of the reference nuclear power plants; b) the code assessment process involving the identification of phenomena challenging the code models; c) the features of the concerned integral test facility (scaling limitations, control logics, data acquisition system, instrumentation, etc.). The activities performed in this respect are discussed, and emphasis is also given to the relevance of the thermal losses to the environment. This issue affects particularly the small scaled facilities and has relevance on the scaling approach related to the power and volume of the facility.

  9. Development of a site-wide accident management center for the Savannah River Site

    International Nuclear Information System (INIS)

    Heal, D.W.; Britt, T.E.

    1992-01-01

    In 1990, the Safety Analysis Group at the Savannah River Site (SRS) began development of an Accident Management program. The program was designed to provide a total system which would meet the Department of Energy (DOE) Safety Performance Criteria, in regard to severe accident management, in the most effective manner. This paper will present two significant changes in the current SRS Accident Management program which will be used to meet these expanded needs. The first and most significant change will be to expand the diversity of the groups involved in the Accident Management process. In the future, organizations such as Environmental Safety, Health ampersand Quality Assurance, Emergency Planning, Site Management, Human Factors, Risk Assessment, and many others will work as an integrated team to solve facility problems. Organizations such as Materials Technology, Equipment Engineering and many of the laboratories on site will be utilized as support groups to increase the technical capability for specific accident analyses. This phase of the program is currently being structured, and should be operational by January of 1993

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

    International Nuclear Information System (INIS)

    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

  11. Beyond-design-basis accident management in the RF regulation documents

    International Nuclear Information System (INIS)

    Bukrinskij, A.M.

    2010-01-01

    The article observes the issues of the management of beyond-design-basis accidents (BDBA) in the existing regulations in Russia. The ideology of the approach to the definition of the BDBA list to formulate the management guidelines has been proposed [ru

  12. Intervention levels for protective measures in nuclear accidents

    International Nuclear Information System (INIS)

    Hedemann Jensen, P.

    1992-12-01

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

  13. The role of systems availability and operator actions in accident management

    International Nuclear Information System (INIS)

    Lutz, R.J. Jr.; Scobel, J.H.

    1988-01-01

    Traditional analyses of severe accidents, such as those presented in Probabilistic Risk Assessment (PRA) studies of nuclear power stations, have generally been performed on the assumption that all means of cooling the reactor core are lost and that no operator actions to mitigate the consequences or progression of the severe accident are performed. The assumption to neglect the availability of safety systems and operator actions which do not prevent core melting can lead to erroneous conclusions regarding the plant severe accident profile. Recent work in severe accident management has identified the need to perform analyses which consider all systems availabilities and operator actions, irrespective of their contribution to the prevention of core melting. These new analyses have far reaching conclusions. The analysis results indicate an unacceptably high degree of simplicity in the present severe accident analyses for Probabilistic Risk Assessment studies; the simplicity is in the assumption that systems availabilities and operator actions which do not impact core melt frequency can be neglected in the severe accident analyses. This results in overly pessimistic predictions of the time of core melting and the subsequent potential for recovery of core cooling prior to core melting. This simplicity can have a considerable impact on severe accident decision making, particularly in the evaluation of alternate plant design features and the priorities for research studies

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

    International Nuclear Information System (INIS)

    Bava, G.; Zambardi, F.

    1992-01-01

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

  15. The use of influence diagrams for evaluating severe accident management strategies

    International Nuclear Information System (INIS)

    Jae, M.; Apostolakis, G.E.

    1992-01-01

    In this paper, the influence diagram, a new analytical tool for developing and evaluating severe accident management strategies, is presented. Influence diagrams are much simpler than decision trees because they do not lead to the large number of branches that are generated when decision trees are used in realistic problems; furthermore, they show explicitly the dependencies between the variables of the problem. One of the accident management strategies proposed for light water reactors, flooding the reactor cavity as a means of preventing vessel breach during a short-term station blackout sequence, is presented. The influence diagram associated with this strategy is constructed. Finally, the advantages of using influence diagrams in accident management are explored

  16. Proceedings of the International conference on nuclear accidents and crisis management

    International Nuclear Information System (INIS)

    Stefenson, B.; Landahl, P.A.; Ritchey, T.

    1993-06-01

    This booklet presents the proceedings of the international conference on nuclear accidents and crisis management, held in Stockholm 16-18 March, 1993. It consists of a collection of lectures and discussion notes. The overall purpose of the conference was to promote a greater awareness of crisis management problems during a nuclear accident of potential international scope. Emphasis was placed on information and cooperation, and on experience of different forms of emergency planning and crisis management. The foreign participants in the conference were scientists and representatives from different levels of authority in Denmark, Finland, Germany, Latvia, Lithuania, Norway, Russia, and USA. The second half of the conference was reserved for Swedish national issues. Several additional themes were discussed here, inter alia: *problems of local, regional and central government cooperation. *the need for special laws and directives concerning nuclear accidents. *the need for more research. The lectures and discussion notes from the second part of the conference are in Swedish

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

    International Nuclear Information System (INIS)

    Aldaihan, S.; Alhbaib, A.; Alrushudi, S.; Karazaitri, C.

    1998-01-01

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

  18. Medical management of two accidents by ionizing radiations

    International Nuclear Information System (INIS)

    Castaneda C, G.

    1991-01-01

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

  19. Triage and medical management of criticality accident victims

    International Nuclear Information System (INIS)

    Lebaron-Jacobs, L.; Flury-Herard, A.; Cavadore, D.

    2002-01-01

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

  20. Risk management: Role of societal factors in major industrial accidents

    International Nuclear Information System (INIS)

    Hovden, J.; Rausand, M.; Sergeev, G.

    1995-01-01

    The paper discusses factors influencing the occurrence of major accidents in complex technological systems. Societal factors are identified as most significant in this context. Important types of societal factors are pin-pointed and discussed. The safety situation in the former Soviet Union and in today's Russian is described. The calamities at Chernobyl, Three Mile Island, and partly also Bhopal are discussed, and the role of societal factors identified. A main point of view is that it is not surprising that these catastrophes happened in the then existing conditions. What is surprising is that they did not happen earlier exclamation point

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

    International Nuclear Information System (INIS)

    2001-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Andre Oudiz [Institut de Radioprotection et de Surete Nucleaire (France)

    2006-07-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kwang Sub

    2000-05-01

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

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

    International Nuclear Information System (INIS)

    Jeong, Kwang Sub

    2000-05-01

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

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

    International Nuclear Information System (INIS)

    Sonnenkalb, Martin

    2006-01-01

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

  6. Management Innovation and Its Measurement

    OpenAIRE

    Kraśnicka, Teresa; Głód, Wojciech; Wronka-Pośpiech, Martyna

    2016-01-01

    The aim of this paper is to explain the nature of management innovation, as well as to propose its measurement instrument. The paper offers a review of key publications on management innovation published in research journals within the last two decades. The critical analysis – primarily focused on definitions, the proposed dimensions of management innovation and the scales used for their measurement – has allowed for the development of an original tool for measuring management ...

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

    International Nuclear Information System (INIS)

    Brenot, J.; Charron, S.; Verger, P.

    2000-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

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

    International Nuclear Information System (INIS)

    Guarnieri, Franck; Travadel, Sebastien; Martin, Christophe; Portelli, Aurelien; Afrouss, Aissame; Takesada, Tomoko

    2015-01-01

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

  10. Information needs and instrument availability for accident management : Application to YGN 3 and 4

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Whan; Park, Rae Jun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Suh, Kune Yull [Seoul Nationsl University, Seoul (Korea, Republic of)

    1996-12-01

    This paper introduces the five-step methodology for identifying information needs and assessing instrument availability during the course of severe in nuclear power plants. The methodology is applied to the Yonggwang (YGN) 3 and 4 to shed light on accident management. It constructs three safety objective trees to prevent the reactor vessel failure, to prevent the containment failure, and to mitigate the fission product release from the containment. The study assesses information needs and instrument availability under severe conditions for preventing the reactor vessel failure of YGN 3 and 4, and recommends additional instruments that may prove to be vital importance in managing the accident. 6 refs., 6 figs., 3 tabs. (author).

  11. Five Years Progress on Waste Management of Fukushima-Daiichi Nuclear Accident

    International Nuclear Information System (INIS)

    Nomura, Shigeo; Katoh, Kazuyuki; Okano, Kenta

    2016-01-01

    Conclusions: • A huge amount of off-site specified waste is planned to be managed by constructing and operating interim storage facilities. However, there still needs a lot of initiatives to recover the 1F nuclear accident. • On-site management of solid waste generated by the accident should be sustained as long-term key activities, such as safe storage, characterization, processing and disposal of various wastes. • Effective collaborations among NDF, TEPCO, IRID, JAEA, other domestic and international organizations and companies are strongly requested to tackle challenging projects on 1F decommissioning.

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

    Science.gov (United States)

    Sugiyama, Daisuke; Hattori, Takatoshi

    2013-01-01

    In environmental remediation after nuclear accidents, radioactive wastes have to be appropriately managed in existing exposure situations with contamination resulting from the emission of radionuclides by such accidents. In this paper, a framework of radiation protection from radioactive waste management in existing exposure situations for application to the practical and reasonable waste management in contaminated areas, referring to related ICRP recommendations was proposed. In the proposed concept, intermediate reference levels for waste management are adopted gradually according to the progress of the reduction in the existing ambient dose in the environment on the basis of the principles of justification and optimisation by taking into account the practicability of the management of radioactive waste and environmental remediation. It is essential to include the participation of relevant stakeholders living in existing exposure situations in the selection of reference levels for the existing ambient dose and waste management.

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

    International Nuclear Information System (INIS)

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

    1978-06-01

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

  14. How to manage forest environments after a nuclear accident? Lessons learned from the Chernobyl and Fukushima accidents

    International Nuclear Information System (INIS)

    2016-03-01

    Based on several published studies, this report proposes a synthetic overview of observations made on the fate of radionuclides in contaminated forests, like in forest environments which represent a great part of highly contaminated areas about Chernobyl and Fukushima. It appears that the main characteristics of forest ecosystems impacted by radioactive fallouts are different (there is no 'red' (dead) forest around Fukushima), that processes governing the fate of radionuclides in forest ecosystems imply a high remanence of radioactive contamination in these environments. It also appears that the interception of radioactive fallouts by the canopy and radionuclide transfers towards the litter and the soil are the most important processes during the early phase and during the first months after the accident. Thus, the soil becomes the main reservoir in which radio-caesium can be found. Some studies outline that the management of contaminated forest ecosystems after the Fukushima accident differs from that applied in the Chernobyl exclusion zone. Others notice that the fire risk is higher in the Chernobyl exclusion zone

  15. iROCS: Integrated accident management framework for coping with beyond-design-basis external events

    International Nuclear Information System (INIS)

    Kim, Jaewhan; Park, Soo-Yong; Ahn, Kwang-Il; Yang, Joon-Eon

    2016-01-01

    Highlights: • An integrated mitigating strategy to cope with extreme external events, iROCS, is proposed. • The strategy aims to preserve the integrity of the reactor vessel as well as core cooling. • A case study for an extreme damage state is performed to assess the effectiveness and feasibility of candidate mitigation strategies under an extreme event. - Abstract: The Fukushima Daiichi accident induced by the Great East Japan earthquake and tsunami on March 11, 2011, poses a new challenge to the nuclear society, especially from an accident management viewpoint. This paper presents a new accident management framework called an integrated, RObust Coping Strategy (iROCS) to cope with beyond-design-basis external events (BDBEEs). The iROCS approach is characterized by classification of various plant damage conditions (PDCs) that might be impacted by BDBEEs and corresponding integrated coping strategies for each of PDCs, aiming to maintain and restore core cooling (i.e., to prevent core damage) and to maintain the integrity of the reactor pressure vessel if it is judged that core damage may not be preventable in view of plant conditions. From a case study for an extreme damage condition, it showed that candidate accident management strategies should be evaluated from the viewpoint of effectiveness and feasibility against accident scenarios and extreme damage conditions of the site, especially when employing mobile or portable equipment under BDBEEs within the limited time available to achieve desired goals such as prevention of core damage as well as a reactor vessel failure.

  16. Analysis of labour accidents in tunnel construction and introduction of prevention measures.

    Science.gov (United States)

    Kikkawa, Naotaka; Itoh, Kazuya; Hori, Tomohito; Toyosawa, Yasuo; Orense, Rolando P

    2015-01-01

    At present, almost all mountain tunnels in Japan are excavated and constructed utilizing the New Austrian Tunneling Method (NATM), which was advocated by Prof. Rabcewicz of Austria in 1964. In Japan, this method has been applied to tunnel construction since around 1978, after which there has been a subsequent decrease in the number of casualties during tunnel construction. However, there is still a relatively high incidence of labour accidents during tunnel construction when compared to incidence rates in the construction industry in general. During tunnel construction, rock fall events at the cutting face are a particularly characteristic of the type of accident that occurs. In this study, we analysed labour accidents that possess the characteristics of a rock fall event at a work site. We also introduced accident prevention measures against rock fall events.

  17. Analysis of labour accidents in tunnel construction and introduction of prevention measures

    Science.gov (United States)

    KIKKAWA, Naotaka; ITOH, Kazuya; HORI, Tomohito; TOYOSAWA, Yasuo; ORENSE, Rolando P.

    2015-01-01

    At present, almost all mountain tunnels in Japan are excavated and constructed utilizing the New Austrian Tunneling Method (NATM), which was advocated by Prof. Rabcewicz of Austria in 1964. In Japan, this method has been applied to tunnel construction since around 1978, after which there has been a subsequent decrease in the number of casualties during tunnel construction. However, there is still a relatively high incidence of labour accidents during tunnel construction when compared to incidence rates in the construction industry in general. During tunnel construction, rock fall events at the cutting face are a particularly characteristic of the type of accident that occurs. In this study, we analysed labour accidents that possess the characteristics of a rock fall event at a work site. We also introduced accident prevention measures against rock fall events. PMID:26027707

  18. Six Decades of Nuclear Accidents, Nuclear Compensation, and Issues of Radioactive Waste Management

    International Nuclear Information System (INIS)

    Boonsuwan, P.; Songjakkeaw, A.

    2011-11-01

    Thailand has made a serious aim to employ nuclear power by adopting five 1,000 MWt in the 2010 national Power Development Plan (PDP 2010) with the first NPP coming online in 2020. However, after the Fukushima nuclear disaster in March 2011, the National Energy Policy Committee had made the resolution to postpone the plan by 3 years. The post-Fukushima atmosphere does not bode well for the public sentiment towards the proposed programme, especially with regards to safety of an NPP. Nonetheless, during the six decades that NPPs have been in operation in 32 countries worldwide, there are only 19 serious accidents involving fatalities and/or damage to properties in excess of 100 million USD. Out of the three significant accidents - Fukushima nuclear accident (2011), Chernobyl nuclear accident (1986), and Three Miles Island nuclear accident (1979) - only the accident at Three Miles Island occurs during normal operation. Such can be implied that the operation of NPPs does maintain a high level of safety. The current technology on nuclear safety has been advancing greatly to the point that the new NPP design claims to render the possibility of a severe accident resulting in core melting insignificant. Along with the technical improvements, laws and regulations have also be progressing in parallel to adequately compensate and limit the liability of operators in case of a nuclear accident. The international agreements such as the Vienna Convention on Civil Liability for Nuclear Damage and the Convention of the Third Party Liability in the Field of Nuclear Energy had also been established and also the national laws of countries such as the United States and Japan have been implemented to address such issues to the point that victims of a nuclear accidents are adequately and justly compensated. In addition to the issues of nuclear accident, the dilemma in nuclear waste management, especially with regards to the High Level Waste which is highly radioactive while having very

  19. Protective measures adopted in OECD member countries in response to the Chernobyl accident

    International Nuclear Information System (INIS)

    Yoshida, Yoshikazu

    1988-01-01

    The report outlines the measures for exposure prevention taken in West European countries following the Chernobyl power plant accident. In particular, the radioactivity regulation levels for foods (derived intervention levels) adopted in these countries are described in detail, citing from the reports of the Committee on Radiation Protection and Public Health of OECD/NEA (The Radiological Impact of the Chernobyl Accident in OECD Countries) and an scientific seminar held by EC (International Scientific Seminar on Foodstuffs Intervention Levels Following a Nuclear Accident). It is pointed out that these countries rather largely vary in measures taken and the derived intervention levels adopted although the principles for radiation protection which provide the basis for emergency protection measures must be nearly the same in all of the countries. It is necessary to establish consistent standards in each country in consideration of an accident, like the one at Chernobyl, that may have global effects. The ICRP recommendations and IAEA safety guidelines so far are centered on ''near-field'' measures to be taken in areas near an accident site. Thus, studies should be made to establish measures to be taken in areas far from the site. (Nogami, K.)

  20. Analysis of occupational accidents: prevention through the use of additional technical safety measures for machinery

    Science.gov (United States)

    Dźwiarek, Marek; Latała, Agata

    2016-01-01

    This article presents an analysis of results of 1035 serious and 341 minor accidents recorded by Poland's National Labour Inspectorate (PIP) in 2005–2011, in view of their prevention by means of additional safety measures applied by machinery users. Since the analysis aimed at formulating principles for the application of technical safety measures, the analysed accidents should bear additional attributes: the type of machine operation, technical safety measures and the type of events causing injuries. The analysis proved that the executed tasks and injury-causing events were closely connected and there was a relation between casualty events and technical safety measures. In the case of tasks consisting of manual feeding and collecting materials, the injuries usually occur because of the rotating motion of tools or crushing due to a closing motion. Numerous accidents also happened in the course of supporting actions, like removing pollutants, correcting material position, cleaning, etc. PMID:26652689

  1. Analysis of occupational accidents: prevention through the use of additional technical safety measures for machinery.

    Science.gov (United States)

    Dźwiarek, Marek; Latała, Agata

    2016-01-01

    This article presents an analysis of results of 1035 serious and 341 minor accidents recorded by Poland's National Labour Inspectorate (PIP) in 2005-2011, in view of their prevention by means of additional safety measures applied by machinery users. Since the analysis aimed at formulating principles for the application of technical safety measures, the analysed accidents should bear additional attributes: the type of machine operation, technical safety measures and the type of events causing injuries. The analysis proved that the executed tasks and injury-causing events were closely connected and there was a relation between casualty events and technical safety measures. In the case of tasks consisting of manual feeding and collecting materials, the injuries usually occur because of the rotating motion of tools or crushing due to a closing motion. Numerous accidents also happened in the course of supporting actions, like removing pollutants, correcting material position, cleaning, etc.

  2. Post-Chernobyl accident radioactivity measurements in the Comunidad Autonoma de Valencia, Spain.

    Science.gov (United States)

    Ortiz, J; Ballesteros, L; Serradell, V

    1992-03-01

    Increased atmospheric radioactivity after the accident in Chernobyl was first detected on air filters. Measurements were begun in Valencia on May 2, 1986, with the maximum activity being observed around May 3-4, 1986. As a consequence of this accident, annual campaigns of measurements on migrating birds (several species of aquatic birds and song-thrushes) were started. The data corresponding to the campaign immediately after the accident (1986/87) show a generalized contamination (approximately 50% of the measured specimens). Significant levels of 134Cs, 137Cs and 110Agm were found. It is important to note that 110Agm is only present in Aythya ferina. In the successive campaigns in 1988/89 and 1989/91 few samples were found to be contaminated and only 137Cs was identified. Strontium-90 was measured and identified in some specimens, mainly in their bones.

  3. An Am-Be neutron source Accident and its management

    International Nuclear Information System (INIS)

    Bai Guang; Wang Xinyong; Wu Zhenghan

    1988-01-01

    An 241 Am-Be neutron source for inaustrial use was lost in a county of Guangdong Province in April, 1982. A school boy picked up and brought it to his home. The source was broken and 10 people were contaminated with radioactive substance. The boy (X) received the highest external irradiation, with chest dose of 0.12 Sv and hand dose of 0.32 Sv. His brother (Y) incurred the heaviest internal contamination by 241 Am, about 3.3 x 10 3 Bq. Decorporation was carried out in four persons including Y, and the excretion of 241 Am in stools and urine was increased significantly. With the medical examination performed upon these persons one and half years after the accident, no positive findings induced by radiation were found except the increase of chromosomal aberration rate in lymphocytes

  4. Tenth Warren K. Sinclair keynote address-the Fukushima nuclear power plant accident and comprehensive health risk management.

    Science.gov (United States)

    Yamashita, Shunichi

    2014-02-01

    Just two years have passed since the Tokyo Electric Power Company-Fukushima Daiichi Nuclear Power Plant (NPP) accident, a multidimensional disaster that combined to destroy the local infrastructure on which the safety system depended and gave a serious impact to the world. Countermeasures including evacuation, sheltering, and control of the food chain were implemented in a timely manner by the Japanese government. However, there is a clear need for improvement, especially in the areas of nuclear safety and protection and also in the management of the radiation health risk during and even after the accident. To date there have been no acute radiation injuries. The radiation-related physical health consequences to the general public, including evacuees, are likely to be much lower than those arising from the Chernobyl nuclear reactor accident, because the radiation fallout and the subsequent environmental contamination were much more limited. However, the social, psychological, and economic impacts of the Fukushima NPP accident are expected to be considerable. Currently, continued monitoring and characterization of the levels of radioactivity in the environment and foods in Fukushima are vital for obtaining informed consent to the decisions on living in the areas already radiocontaminated and returning back to the evacuated areas once re-entry is permitted; it is also important to perform a realistic assessment of the radiation doses on the basis of measurements. We are currently implementing the official plans of the Fukushima Health Management Survey, which includes a basic survey for the estimation of the external doses that were received during the first 4 mo after the accident and four more detailed surveys (thyroid ultrasound examination, comprehensive health check-up, mental health and life-style survey, and survey of pregnant women and nursing mothers), with the aim to take care of the health of all of the residents of the Fukushima Prefecture for a long time

  5. Severe Accident Management Guidance: Lessons Still to be Learned after Fukushima

    International Nuclear Information System (INIS)

    Vayssier, G.

    2016-01-01

    After the accidents in Three Mile Island (TMI) and Chernobyl, many countries decided to develop and implement guidelines specifically directed to mitigate accidents with core damage, so-called severe accidents. The guidelines are usually named Severe Accident Management Guidelines (SAMG). In the USA, all operating plants had these guidelines in place at the end of 1998. Most other countries followed later, but today, it can be said that many nuclear power plants in the world have such guidelines in place. Typically, however, the guidelines were constructed under the assumption that many plant systems still will be available, i.e. there will be DC to feed the instruments, AC to feed equipment and water to restore cooling to the core. Typically, this was basically the situation at TMI: most equipment was functional, only the insight of what had happened had been lost and operators did not know how to respond. At Fukushima-Daiichi, a Site Disruptive Accident (SDA) occurred and it appeared that the situation was much more complex: much of the needed supportive equipment needed was unavailable, which greatly complicated the handling of the event. In this paper, the major shortcomings of the present existing SAMG are discussed, both from a technical, and an organisational viewpoint. It is concluded that, where proper regulation still is missing, the development of an industrial standard is recommended to define adequate tools and guidelines to mitigate severe accidents, including SDAs. (author).

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

    Science.gov (United States)

    Hotz, Mark E; Meineke, Viktor

    2012-08-01

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

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

    International Nuclear Information System (INIS)

    2014-01-01

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

  8. TL detectors for gamma ray dose measurements in criticality accidents

    International Nuclear Information System (INIS)

    Miljanic, S.; Zorko, B.; Gregori, B.; Knezevic, Z.

    2007-01-01

    Determination of gamma ray dose in mixed neutron + gamma ray fields is still a demanding task. Dosemeters used for gamma ray dosimetry are usually in some extent sensitive to neutrons and their response variations depend on neutron energy i.e., on neutron spectra. Besides, it is necessary to take into account the energy dependence of dosemeter responses to gamma rays. In this work, several types of thermoluminescent detectors (TLD) placed in different holders used for gamma ray dose determination in the mixed fields were examined. Dosemeters were from three different institutions: Ruder Boskovic Inst. (RBI), Croatia, Jozef Stefan Inst. (JSI), Slovenia and Autoridad Regulatoria Nuclear (ARN), Argentina. All dosemeters were irradiated during the International Intercomparison of Criticality Accident Dosimetry Systems at the SILENE Reactor, Valduc, June 2002. Three accidental scenarios were reproduced and in each irradiation the dosemeters were exposed placed on the front of phantom and 'free in air'. Following types of TLDs were used: 7 LiF (TLD-700), CaF 2 :Mn and Al2 O3 :Mg,Y - all from RBI; CaF 2 :Mn from JSI and 7 LiF (TLD-700) from ARN. Reported doses were compared with the reference values as well as with the values obtained from the results of all participants. The results show satisfactory agreement with other dosimetry systems used in the Intercomparison. The influence of different types of holders and applied corrections of dosemeters' readings are discussed. (authors)

  9. TL detectors for gamma ray dose measurements in criticality accidents.

    Science.gov (United States)

    Miljanić, Saveta; Zorko, Benjamin; Gregori, Beatriz; Knezević, Zeljka

    2007-01-01

    Determination of gamma ray dose in mixed neutron+gamma ray fields is still a demanding task. Dosemeters used for gamma ray dosimetry are usually in some extent sensitive to neutrons and their response variations depend on neutron energy i.e., on neutron spectra. Besides, it is necessary to take into account the energy dependence of dosemeter responses to gamma rays. In this work, several types of thermoluminescent detectors (TLD) placed in different holders used for gamma ray dose determination in the mixed fields were examined. Dosemeters were from three different institutions: Ruder Bosković Institute (RBI), Croatia, JoZef Stefan Institute (JSI), Slovenia and Autoridad Regulatoria Nuclear (ARN), Argentina. All dosemeters were irradiated during the International Intercomparison of Criticality Accident Dosimetry Systems at the SILENE Reactor, Valduc, June 2002. Three accidental scenarios were reproduced and in each irradiation the dosemeters were exposed placed on the front of phantom and 'free in air'. Following types of TLDs were used: 7LiF (TLD-700), CaF2:Mn and Al2O3:Mg,Y-all from RBI; CaF2:Mn from JSI and 7LiF (TLD-700) from ARN. Reported doses were compared with the reference values as well as with the values obtained from the results of all participants. The results show satisfactory agreement with other dosimetry systems used in the Intercomparison. The influence of different types of holders and applied corrections of dosemeters' readings are discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

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

    International Nuclear Information System (INIS)

    2000-03-01

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

  12. Geographic Information System (GIS) capabilities in traffic accident information management: a qualitative approach

    Science.gov (United States)

    Ahmadi, Maryam; Valinejadi, Ali; Goodarzi, Afshin; Safari, Ameneh; Hemmat, Morteza; Majdabadi, Hesamedin Askari; Mohammadi, Ali

    2017-01-01

    Background Traffic accidents are one of the more important national and international issues, and their consequences are important for the political, economical, and social level in a country. Management of traffic accident information requires information systems with analytical and accessibility capabilities to spatial and descriptive data. Objective The aim of this study was to determine the capabilities of a Geographic Information System (GIS) in management of traffic accident information. Methods This qualitative cross-sectional study was performed in 2016. In the first step, GIS capabilities were identified via literature retrieved from the Internet and based on the included criteria. Review of the literature was performed until data saturation was reached; a form was used to extract the capabilities. In the second step, study population were hospital managers, police, emergency, statisticians, and IT experts in trauma, emergency and police centers. Sampling was purposive. Data was collected using a questionnaire based on the first step data; validity and reliability were determined by content validity and Cronbach’s alpha of 75%. Data was analyzed using the decision Delphi technique. Results GIS capabilities were identified in ten categories and 64 sub-categories. Import and process of spatial and descriptive data and so, analysis of this data were the most important capabilities of GIS in traffic accident information management. Conclusion Storing and retrieving of descriptive and spatial data, providing statistical analysis in table, chart and zoning format, management of bad structure issues, determining the cost effectiveness of the decisions and prioritizing their implementation were the most important capabilities of GIS which can be efficient in the management of traffic accident information. PMID:28848627

  13. Geographic Information System (GIS) capabilities in traffic accident information management: a qualitative approach.

    Science.gov (United States)

    Ahmadi, Maryam; Valinejadi, Ali; Goodarzi, Afshin; Safari, Ameneh; Hemmat, Morteza; Majdabadi, Hesamedin Askari; Mohammadi, Ali

    2017-06-01

    Traffic accidents are one of the more important national and international issues, and their consequences are important for the political, economical, and social level in a country. Management of traffic accident information requires information systems with analytical and accessibility capabilities to spatial and descriptive data. The aim of this study was to determine the capabilities of a Geographic Information System (GIS) in management of traffic accident information. This qualitative cross-sectional study was performed in 2016. In the first step, GIS capabilities were identified via literature retrieved from the Internet and based on the included criteria. Review of the literature was performed until data saturation was reached; a form was used to extract the capabilities. In the second step, study population were hospital managers, police, emergency, statisticians, and IT experts in trauma, emergency and police centers. Sampling was purposive. Data was collected using a questionnaire based on the first step data; validity and reliability were determined by content validity and Cronbach's alpha of 75%. Data was analyzed using the decision Delphi technique. GIS capabilities were identified in ten categories and 64 sub-categories. Import and process of spatial and descriptive data and so, analysis of this data were the most important capabilities of GIS in traffic accident information management. Storing and retrieving of descriptive and spatial data, providing statistical analysis in table, chart and zoning format, management of bad structure issues, determining the cost effectiveness of the decisions and prioritizing their implementation were the most important capabilities of GIS which can be efficient in the management of traffic accident information.

  14. Military Curricula for Vocational and Technical Education. Traffic Management and Accident Investigation, 17-8.

    Science.gov (United States)

    Air Force, Washington, DC.

    This teaching guide and student workbook for a postsecondary level course in traffic management and accident investigation is one of a number of military-developed curriculum packages selected for adaptation to vocational instruction and curriculum development in a civilian setting. Purpose stated for the 132-hour course is to expose students to…

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

    1994-01-01

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

  17. Uncertainty quantification for accident management using ACE surrogates

    International Nuclear Information System (INIS)

    Varuttamaseni, A.; Lee, J. C.; Youngblood, R. W.

    2012-01-01

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

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

    Institute of Scientific and Technical Information of China (English)

    Takumi Kawaguchi; Eitaro Taniguchi; Michio Sata

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    Herzberg, B.

    1976-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Miaw, Sophia T.W. [International Atomic Energy Agency, Vienna (Austria). Safety Co-ordination Section; Mezhari, Arnaldo; Shu, Jane; Xavier, Ana Maria [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil). Superintendencia de Licenciamento e Controle

    1997-12-31

    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 11 refs., 5 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

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

    International Nuclear Information System (INIS)

    Miaw, Sophia T.W.; Mezhari, Arnaldo; Shu, Jane; Xavier, Ana Maria

    1997-01-01

    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

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

    International Nuclear Information System (INIS)

    Kawasaki, Ikuo; Yoshida, Yoshitaka

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Behrend, U.; Dulski, P.; Friedland, E.M.; Gawlik, D.; Kirschfeld, K.E.; Schubert, P.; Steinmetz, K.H.

    1987-01-01

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

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

    International Nuclear Information System (INIS)

    Ambroggi, R.

    1988-01-01

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

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

    International Nuclear Information System (INIS)

    Park, Gun Chul; Suh, Kune Y.; Lee, Jin Yong; Lee, Seung Dong

    2001-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    Jae, M.; Moon, J.H.

    2002-01-01

    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

  9. Learning Lessons from TMI to Fukushima and Other Industrial Accidents: Keys for Assessing Safety Management Practices

    International Nuclear Information System (INIS)

    Dechy, N.; Rousseau, J.-M.; Dien, Y.; Montmayeul, R.; Llory, M.

    2016-01-01

    The main objective of the paper is to discuss and to argue about transfer, from an industrial sector to another industrial sector, of lessons learnt from accidents. It will be achieved through the discussion of some theoretical foundations and through the illustration of examples of application cases in assessment of safety management practices in Nuclear Power Plant (NPP). The nuclear energy production industry has faced three big ones in 30 years (TMI, Chernobyl, Fukushima) involving three different reactor technologies operated in three quite different cultural, organizational and regulatory contexts. Each of those accident has been the origin of questions, but also generator of lessons, some changing the worldview (see Wilpert and Fahlbruch, 1998) of what does cause an accident in addition to the engineering view about the importance of technical failures (human error, safety culture, sociotechnical interactions). Some of their main lessons were implemented such as improvements of human-machine interfaces ergonomics, recast of some emergency operating procedures, severe accident mitigation strategies and crisis management. Some lessons did not really provide deep changes. It is the case for organizational lessons such as, organizational complexity, management of production pressures, regulatory capture, and failure to learn, etc.

  10. Construction Project Administration and Management for Mitigating Work Zone Accidents and Fatalities: An Integrated Risk Management Model

    Science.gov (United States)

    2009-10-01

    The goal of this research is to mitigate the risk of highway accidents (crashes) and fatalities in work zones. The approach of this research has been to address the mitigation of work zone crashes through the creation of a formal risk management mode...

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

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

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

    International Nuclear Information System (INIS)

    Li-Chi Cliff Po; Link, John M.

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    2013-01-01

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

  14. Accident management to prevent containment failure and reduce fission product release

    International Nuclear Information System (INIS)

    Lehner, J.R.; Lin, C.C.; Luckas, W.J.; Pratt, W.T.

    1991-01-01

    Brookhaven National Laboratory, under the auspices of the US Nuclear Regulatory Commission, is investigating accident management strategies which could help preserve containment integrity or minimize releases during a severe accident. The strategies considered 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 sizable uncertainties exist regarding some of the phenomena involved. The identification and assessment process for containment and release strategies is described, and some insights derived from its application to specific containment types are presented. 2 refs., 5 figs., 2 tabs

  15. Developing a Minimum Data Set for an Information Management System to Study Traffic Accidents in Iran.

    Science.gov (United States)

    Mohammadi, Ali; Ahmadi, Maryam; Gharagozlu, Alireza

    2016-03-01

    Each year, around 1.2 million people die in the road traffic incidents. Reducing traffic accidents requires an exact understanding of the risk factors associated with traffic patterns and behaviors. Properly analyzing these factors calls for a comprehensive system for collecting and processing accident data. The aim of this study was to develop a minimum data set (MDS) for an information management system to study traffic accidents in Iran. This descriptive, cross-sectional study was performed in 2014. Data were collected from the traffic police, trauma centers, medical emergency centers, and via the internet. The investigated resources for this study were forms, databases, and documents retrieved from the internet. Forms and databases were identical, and one sample of each was evaluated. The related internet-sourced data were evaluated in their entirety. Data were collected using three checklists. In order to arrive at a consensus about the data elements, the decision Delphi technique was applied using questionnaires. The content validity and reliability of the questionnaires were assessed by experts' opinions and the test-retest method, respectively. An (MDS) of a traffic accident information management system was assigned to three sections: a minimum data set for traffic police with six classes, including 118 data elements; a trauma center with five data classes, including 57 data elements; and a medical emergency center, with 11 classes, including 64 data elements. Planning for the prevention of traffic accidents requires standardized data. As the foundation for crash prevention efforts, existing standard data infrastructures present policymakers and government officials with a great opportunity to strengthen and integrate existing accident information systems to better track road traffic injuries and fatalities.

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

    International Nuclear Information System (INIS)

    Uspuras, E.; Kaliatka, A.

    2001-01-01

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

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

    International Nuclear Information System (INIS)

    Sairanen, R.

    1998-01-01

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

  18. Status report on severe accident material property measurements

    International Nuclear Information System (INIS)

    Farmer, M.T.; McUmber, L.; Spencer, B.W.; Aeschlimann, R.W.

    1997-06-01

    Measurements of selected material properties of molten reactor core material (corium) were made. The corium used was a mixture of UO 2 , ZrO 2 and Zr, with oxygen content being a parameter to reflect different stages of zirconium oxidation. The mixtures used were representative of typical in-vessel melt sequences. For most measurements, the UO 2 /ZrO 2 mass ratio was 1.51, representative of VVER/440 melt compositions and melt compositions of most US BWRs. Measurements were made of the solidus/liquidus temperatures of corium compositions using a Differential Thermal Analysis technique. Observation of the solubility of unoxidized Zr in the oxide phase was made by metallographic analysis of solidus/liquidus melt samples. The results of laminar flow corium spreading tests in one dimension were used to estimate the viscosity of corium compositions. Measured solidus and liquidus temperatures for compositions representative of Zr oxidation of 30, 50 and 70% were compared with those obtained form a phase diagram provided by Kurchatov Institute. It was found that experimental measurements agreed well with the phase diagram values at 70% oxidation, but the measured solidus temperatures were higher than those on the phase diagram and the measured liquidus temperatures were lower than those on the phase diagram at 30 and 50% oxidation. From a microstructure examination it was determined that there was no global segregation into distinct metal and oxide phases during the cooldown of a sample in which there was initially 70% Zr oxidation. Therefore it is concluded that Zr metal is soluble in the oxide phase under molten conditions. Viscosity estimates were made for compositions representative of Zr oxidation of 30, 50 and 70% by fitting the results of spreading tests to Huppert's equation. It was found that, at a temperature of 2500 C, the viscosity varied by three orders of magnitude over this range of compositions. 10 refs., 39 figs., 16 tabs

  19. Evaluation of a severe accident management strategy for boiling water reactors -- Drywell flooding

    International Nuclear Information System (INIS)

    Yu, D.; Xing, L.; Kastenberg, W.E.; Okrent, D.

    1994-01-01

    Flooding of the drywell has been suggested as a strategy to prevent reactor vessel and containment failure in boiling water reactors. To evaluate the candidate strategy, this study considers accident management as a decision problem (''drywell flooding'' versus ''do nothing'') and develops a decision-oriented framework, namely, the influence diagram approach. This analysis chooses the long-term station blackout sequence for a Mark 1 nuclear power plant (Peach Bottom), and an influence diagram with a single decision node is constructed. The node probabilities in the influence diagram are obtained from US Nuclear Regulatory Commission reports or estimated by probabilistic risk assessment methodology. In assessing potential benefits compared with adverse effects, this analysis uses two consequence measures, i.e., early and late fatalities, as decision criteria. The analysis concludes that even though potential adverse effects exist, such as ex-vessel steam explosions and containment isolation failure, the drywell flooding strategy is preferred to ''do nothing'' when evaluated in terms of these consequence measures

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

    International Nuclear Information System (INIS)

    Bonka, H.; Kueppers, J.; Maqua, M.

    1988-01-01

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

  1. Construct Measurement in Management Research

    DEFF Research Database (Denmark)

    Nielsen, Bo Bernhard

    2014-01-01

    Far too often do management scholars resort to crude and often inappropriate measures of fundamental constructs in their research; an approach which calls in question the interpretation and validity of their findings. Scholars often legitimize poor choices in measurement with a lack of availability......, this research note raises important questions about the use of proxies in management research and argues for greater care in operationalizing constructs with particular attention to matching levels of theory and measurement....

  2. Distribution and ecological risk assessment of cadmium in water and sediment in Longjiang River, China: Implication on water quality management after pollution accident.

    Science.gov (United States)

    Zhao, Xue-Min; Yao, Ling-Ai; Ma, Qian-Li; Zhou, Guang-Jie; Wang, Li; Fang, Qiao-Li; Xu, Zhen-Cheng

    2018-03-01

    In early January 2012, the Longjiang River was subjected to a serious cadmium (Cd) pollution accident, which led to negatively environmental and social impacts. A series of measures of emergency treatment were subsequently taken to reduce water Cd level. However, little information was available about the change of Cd level in environmental matrices and long-term effect of this pollution accident to aquatic ecosystem. Thus, this study investigated the distribution of Cd in water and sediment of this river for two years since pollution accident, as well as assessed its ecological risk to aquatic ecosystem of Longjiang River. The results showed that it was efficient for taking emergency treatment measures to decrease water Cd concentration to below the threshold value of national drinking water quality standard of China. There was high risk (HQ > 1) to aquatic ecosystem in some of reaches between February and July 2012, but low or no risk (HQ polluted reaches increased after pollution accident and emergency treatments in 2012, but decreased in 2013. During flood period, the sediment containing high concentration of Cd in Longjiang River was migrated to downstream Liujiang River. Cd content in sediment was reduced to background level after two years of the pollution accident occurrence. The study provides basic information about Cd levels in different media after pollution accident, which is helpful in evaluating the effectiveness of emergency treatments and the variation of ecological risk, as well as in conducting water management and conservation. Copyright © 2017 Elsevier Ltd. All rights reserved.

  3. On-line fission products measurements during a PWR severe accident: the French DECA-PF project

    Energy Technology Data Exchange (ETDEWEB)

    Ducros, G.; Allinei, P.G.; Roure, C. [CEA, DEN, F-13108 Saint-Paul-lez-Durance, (France); Rozel, C. [EDF SEPTEN, 12-14 Avenue Dutrievoz, F-69628, Villeurbanne, (France); Blanc De Lanaute, N. [CANBERRA, 1 rue des Herons, F-78182, Saint Quentin en Yvelines, (France); Musoyan, G. [AREVA, Tour AREVA, 1 place Jean Millier, F-92084 Paris La Defense Cedex, (France)

    2015-07-01

    Following the Fukushima accident, a lot of recommendations was drawn by international organizations (IAEA, OECD, NUGENIA network...) in order to improve the safety in such accidental conditions and mitigate their consequences. One of these recommendations was to improve the robustness of the instrumentation, which was dramatically lacking at Fukushima, as well as to better determine the Source Term involved in nuclear accident. The DECA-PF project (Diagnosis of a degraded reactor core through Fission Product measurements) was elaborated in this context and selected as one of 21 collaborative R and D projects in the field of nuclear safety and radioprotection, funded in May 2013 by the French National Research Agency. Over the months following the Fukushima accident, a CEA crisis team was held in order to analyze on-line the situation taking into account the data delivered by TEPCO and other organizations. Despite the difficulties encountered concerning the reliability of these data, the work performed showed the high capacity of Fission Products (FP) measurements to get a diagnosis relative to the status of the reactors and the spent fuel pools (SFP). Based on these FP measurements, it was possible to conclude that the main origin of the releases was coming from the cores and not from the SFP, in particular for SFP-4 which was of high concern, and that the degradation level of the reactors was very large, including probably an extensive core melting. To improve the reliability of this kind of diagnosis, the necessity to get such measurements as soon as possible after the accident and as near as possible from the reactor was stressed. In this way the present DECA-PF project intends to develop a new and innovative instrumentation taking into account the design of the French nuclear power plants on which sand bed filters have been implemented for severe accident management. Three complementary techniques, devoted to measure the FP release on-line, are being studied

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-06-01

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

  5. Conservative Management of Theca Lutein Cyst Accident: A Case Report

    Directory of Open Access Journals (Sweden)

    Radha Bai Prabhu T

    2017-10-01

    Full Text Available Theca lutein cysts can occur in 20-25% of molar pregnancies. These cysts can undergo complications such as torsion, rupture, and haemorrhage. As these are functional cysts, when there are complications such as torsion they can be managed conservatively by aspirating the cysts under ultrasound guidance or by detorsion at the time of laparoscopy. By simple detorsion, ovaries can be preserved in 80-90% of cases. In order to prevent recurrence adnexal fixation can be undertaken by plicating the ovarian ligament.

  6. Severe accident management; the approach in the USA. Applications of US methods in Europe. Other approaches in Europe

    International Nuclear Information System (INIS)

    Vayssier, G.

    1999-01-01

    In this lecture severe accident management, applications of US methods in Europe are presented. Author deals with historical perspective, US industry position to core melt accidents, method of Westinghouse owners group, method of Combustion Engineering owners group, method of Babcock and Wilcox Owners group, interaction with/inspection by the USNRC and with assessment of US SAMG methods

  7. Accident Management System Based on Vehicular Network for an Intelligent Transportation System in Urban Environments

    Directory of Open Access Journals (Sweden)

    Yusor Rafid Bahar Al-Mayouf

    2018-01-01

    Full Text Available As cities across the world grow and the mobility of populations increases, there has also been a corresponding increase in the number of vehicles on roads. The result of this has been a proliferation of challenges for authorities with regard to road traffic management. A consequence of this has been congestion of traffic, more accidents, and pollution. Accidents are a still major cause of death, despite the development of sophisticated systems for traffic management and other technologies linked with vehicles. Hence, it is necessary that a common system for accident management is developed. For instance, traffic congestion in most urban areas can be alleviated by the real-time planning of routes. However, the designing of an efficient route planning algorithm to attain a globally optimal vehicle control is still a challenge that needs to be solved, especially when the unique preferences of drivers are considered. The aim of this paper is to establish an accident management system that makes use of vehicular ad hoc networks coupled with systems that employ cellular technology in public transport. This system ensures the possibility of real-time communication among vehicles, ambulances, hospitals, roadside units, and central servers. In addition, the accident management system is able to lessen the amount of time required to alert an ambulance that it is required at an accident scene by using a multihop optimal forwarding algorithm. Moreover, an optimal route planning algorithm (ORPA is proposed in this system to improve the aggregate spatial use of a road network, at the same time bringing down the travel cost of operating a vehicle. This can reduce the incidence of vehicles being stuck on congested roads. Simulations are performed to evaluate ORPA, and the results are compared with existing algorithms. The evaluation results provided evidence that ORPA outperformed others in terms of average ambulance speed and travelling time. Finally, our

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

    International Nuclear Information System (INIS)

    Pampuro, V.I.

    2001-01-01

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

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

    International Nuclear Information System (INIS)

    Raes, F.; Graziani, G.; Grossi, L.; Marciano, L.; Peirs, D.; Pedersen, B.; Stanners, D.; Zarimpas, N.

    1989-01-01

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

  10. The management of victims of collective accidents. Methodological guide

    International Nuclear Information System (INIS)

    Meadel, Juliette; Marie, Eugenie; Badorc, Yves; Riviere, Anne; Chalaux, Marion; Heisserer, Julie; Tosello-Orsola, Romain; Gelli, Robert; Boura, Florent; Sookahet, Stephanie; Deparis, Benjamin; Tarabeux, Xavier; Besson, Jean-Paul; Porterie, Frederique; Paris-Muller, Gaelle; Marchelli, Sylvie; Gesret, Samuel; Guthleben, Sandra; Sauvee, Marie-Line; Percie Du Sert, Xavier; Bossuyt, Yves; Viteau, Thierry; Etienne-Diener, Annick; Le Caro, Sylvie; Taffin, Monique; Deluz, Damien; Cremniter, Didier; Philippe, Jean-Marc; Ouvrard, Anne-Emmanuelle; Andre, Jean-Marc; Lansman, Patrick; Jouty, Remi; Delbono, Martine; Cebe, Philippe; Le Divenah, Jean-Paul; Vandoorne, Pierre-Jean; Bertin, Jerome; Sadowski, Isabelle; Bovis, Marcel; Gicquel, Stephane; Seco, Sophia; Ieche, Clara; Bertrand, Paul; Tricot, Sandrine; Gomes, Thierry; Marissal, Jean-Luc; Voss, Beatrice; Mescam, Marie; Le Chevalier, Elisabeth; Bouchet, Loic; Lienhard, Claude; Dantchev, Nicolas; Delval, Christophe; Naudascher, Michel; Sublet, Xavier; Cutajar, Chantal; Trestard, Jean-Francois; Quentin, Cecile

    2017-11-01

    This publication proposes several sets of sheets. A first set concerns the crisis phase itself and sheets present or describe the coordination during an emergency phase, the role of the public prosecutor, the role of the prefect, the crisis and support Centre of the Ministry for Foreign Affairs, the health management of victims, the mobilisation of association for the support to victims, the role of associations of victims, crisis information telephone numbers, the family welcome centre, information meetings for victims and families, relationships with media. The second set concerns the post-crisis phase: coordination during the follow-up phase, the intervention of insurance providers, and the intervention of lawyers. A last sheet concerns the legal phase with a presentation of the legal procedure

  11. Medical measures in case of nuclear power plant accidents

    International Nuclear Information System (INIS)

    1986-01-01

    The Laender governments of the Federal Republic of Germany are of the opinion that within the framework of precautionary disaster control, plans have to be set up dealing with nuclear disasters that seem improbable but cannot be completely excluded. Accordingly, recommendations presented by the Federal Government and the Laender governments have been combined into a framework disaster control scheme where the competencies for activities and measures lie with the several Laender governments, as given by the Basic Law. A further recommendation deals with the medical care and service in case of a nuclear disaster, and the practical guide presented here is intended to give the information and instructions needed in order to comply with the legal framework. A working group has been set up in order to work out the rules and facts for optimum medical care. The activities are planned to be based on an emergency station responsible for medical examination, treatment, and transfer of victims. The practical guide has been discussed by the 'Committee for disaster control in the vicinity of nuclear installations' of the SSK, has been approved of by the supreme Land authorities of the Laender concerned, and has been passed by the SSK at its 63rd meeting. With 5 figs., 6 tabs [de

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

    International Nuclear Information System (INIS)

    Silva, Kampanart; Ishiwatari, Yuki; Takahara, Shogo

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-01

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

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

    International Nuclear Information System (INIS)

    1996-08-01

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

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

    International Nuclear Information System (INIS)

    Akisaka, Masafumi; Nakamura, Tomoko; Satake, Tsuyoshi

    2002-01-01

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

  17. A system of safety management practices and worker engagement for reducing and preventing accidents: an empirical and theoretical investigation.

    Science.gov (United States)

    Wachter, Jan K; Yorio, Patrick L

    2014-07-01

    The overall research objective was to theoretically and empirically develop the ideas around a system of safety management practices (ten practices were elaborated), to test their relationship with objective safety statistics (such as accident rates), and to explore how these practices work to achieve positive safety results (accident prevention) through worker engagement. Data were collected using safety manager, supervisor and employee surveys designed to assess and link safety management system practices, employee perceptions resulting from existing practices, and safety performance outcomes. Results indicate the following: there is a significant negative relationship between the presence of ten individual safety management practices, as well as the composite of these practices, with accident rates; there is a significant negative relationship between the level of safety-focused worker emotional and cognitive engagement with accident rates; safety management systems and worker engagement levels can be used individually to predict accident rates; safety management systems can be used to predict worker engagement levels; and worker engagement levels act as mediators between the safety management system and safety performance outcomes (such as accident rates). Even though the presence of safety management system practices is linked with incident reduction and may represent a necessary first-step in accident prevention, safety performance may also depend on mediation by safety-focused cognitive and emotional engagement by workers. Thus, when organizations invest in a safety management system approach to reducing/preventing accidents and improving safety performance, they should also be concerned about winning over the minds and hearts of their workers through human performance-based safety management systems designed to promote and enhance worker engagement. Copyright © 2013 The Authors. Published by Elsevier Ltd.. All rights reserved.

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

    International Nuclear Information System (INIS)

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

    2011-08-01

    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

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

    International Nuclear Information System (INIS)

    Drottz-Sjoeberg, B.M.

    1993-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Drottz-Sjoeberg, B M

    1993-06-01

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

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

    International Nuclear Information System (INIS)

    Bouyer, V.; Cassiaut-Louis, N.; Fouquart, P.; Journeau, C.; Piluso, P.; Parga, C.

    2013-06-01

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

  2. Comparing recall vs. recognition measures of accident under-reporting: A two-country examination.

    Science.gov (United States)

    Probst, Tahira M; Petitta, Laura; Barbaranelli, Claudio

    2017-09-01

    A growing body of research suggests that national injury surveillance data significantly underestimate the true number of non-fatal occupational injuries due to employee under-reporting of workplace accidents. Given the importance of accurately measuring such under-reporting, the purpose of the current research was to examine the psychometric properties of two different techniques used to operationalize accident under-reporting, one using a free recall methodology and the other a recognition-based approach. Moreover, in order to assess the cross-cultural generalizability of these under-reporting measures, we replicated our psychometric analyses in the United States (N=440) and Italy (N=592). Across both countries, the results suggest that both measures exhibited similar patterns of relationships with known antecedents, including job insecurity, production pressure, safety compliance, and safety reporting attitudes. However, the recall measures had more severe violations of normality and were less correlated with self-report workplace injuries. Considerations, implications, and recommendations for using these different types of accident measures are discussed. Copyright © 2017 Elsevier Ltd. All rights reserved.

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

    International Nuclear Information System (INIS)

    2014-06-01

    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

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

    International Nuclear Information System (INIS)

    2000-03-01

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

  5. An Approach to Description of Accident Causation and Identification of Preconditions for Effective Risk Management

    DEFF Research Database (Denmark)

    Rasmussen, Jens; Svedung, Inge

    "safety control" must be established from a control theory point of view? The report discusses these issues on the basis of the present rapid evolution of new cognitive approaches to the study of decision making in action and dynamic, learning organizations, and the rapid change of modern information......The ultimate objectives of the present study is to better understand the mechanisms of major accidents in the present dynamic and technological society. From this understanding, guides to improved strategies for industrial risk management is sought. In this approach, risk management is considered...

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

    International Nuclear Information System (INIS)

    2014-01-01

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

  7. The reactor accident at Chernobyl, U.S.S.R. Radiation measurements in Denmark. 3. report

    International Nuclear Information System (INIS)

    1986-01-01

    In continuation of the reporting of 4 May and 11 May 1986 this report summarizes the radioactivity measurements made during the third and fourth week after the accident at Chernobyl. The data have been collated by the Inspectorate of Nuclear Installations from measurements made by Risoe National Laboratory and the National Institute of Radiation Hygiene. The radioactivity remaining in the air after the first two weeks shows daily variations at low levels without significant contribution to the fall out levels on the ground surfaces. The ground contamination shows a decreasing trend according to radioactive decay and for the plants also according to natural cleaning mechanisms. The radioactive data from the third and fourth week after the accident confirm the previous estimate that the total radiation impact on the Danish area from the accident, including future radiation exposures from the contamination experienced up to now, corresponds at most to approximately one month of natural background radiation. For the time to come the measuring programme and data reporting arrangements will be reorganized with a view to the future long term follow-up of the situation. Thus, this report is expected to be the last in the series of ad hoc reports for prompt dissemination of data on the Danish radioactivity measurements. (author)

  8. Plant state identification using fuzzy logic in the framework of computerized accident management support (CAMS)

    International Nuclear Information System (INIS)

    Van Dyck, Claude

    1997-05-01

    CAMS (computerized accident management support) is a system that will provide assistance in case of accident in a nuclear power plant. In order to support the user in evaluating the plant state, it contains a state identification module. The state identification module provides high-level, qualitative information about the status of critical safety functions, about the availability of safety systems and about the occurrence of initiating events. This information is sent to the man-machine interface and to other CAMS modules. The state identification module is developed using a specific tool: GPS (Goal Processing System) which is based on the Goal Tree - Success Tree formalism. GPS is a tool designed to manage ''process related'' knowledge and aimed at process supervision via real-time acquisition of process variables. Fuzzy logic has been introduced in GPS in order to have smoother transitions between different states of critical safety functions and systems changes and to have a truth value associated to each piece of information provided to the user. The whole system has been tested, integrated with the rest of CAMS, on several accident scenarios. The test results are satisfactory. A brief comparison is made between the present work and previous related work at the HRP. (author)

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

    International Nuclear Information System (INIS)

    Baumont, G.; Menage, F.; Schneiter, J.R.; Spurgin, A.; Vogel, A.

    2000-01-01

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

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

    International Nuclear Information System (INIS)

    Hodge, S.A.

    1991-01-01

    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

  11. Evaluation of Coolant Injection Procedure in the Severe Accident Management Strategy of APR1400

    International Nuclear Information System (INIS)

    Cho, Yongjin; Lim, Kukhee; Song, Sungchu; Lee, Sukho; Hwang, Taesuk

    2013-01-01

    A coolant injection strategy in the severe accident management guideline (SAMG) of APR1400 relates to immediate coolant injection into RCS (Reactor Coolant System) or injection following the recovery of secondary coolant inventory. This strategy could play important role in accident mitigation and radiological consequences. In this study, appropriateness of the strategy was evaluated using MELCOR1.8.6 and several sensitivity studies of the key parameters were performed. Analysis for APR1400 using MELCOR 1.8.6 was performed to evaluate the effectiveness of accident management strategies and the following conclusions were identified. Sequential operation of secondary and RCS injection may not be the best strategy and the simultaneous injection of secondary and RCS injection could be more preferable. At least, the RCS injection should start before complete drainage of water in the safety injection tank using mobile pumps. In this study, the effectiveness of timing of operator action has been examined and the amount of injection flowrate needs to be studied in the future

  12. State of Level 2 analyses and severe accident management in Spanish nuclear power plants

    International Nuclear Information System (INIS)

    Otero, R.

    1998-01-01

    The state of the PSA/IPE studies in the Spanish NPPs is presented in this report, as well as the plans to implement the severe accident management strategy both in the Spanish BWRs and PWRs. First, the Spanish LWRs are introduced, and the scope of the IPE analyses required by the Spanish Regulatory Commission (CSN) is given. The general overview is completed with the current degree of development for the IPE studies in each plant. In the second part the methods and tools are shown which are used by the Spanish plants to develop their Level 2 analysis. The different strategies for severe accident management adopted by the BWPs and PWRs in Spain are also outlined. The sources and implementation of the Severe Accident Guidelines (SAG) are described. More detail is given in the following chapters to the containment analysis of Trillo (PWR, KWU design) and Cofrentes (BWR/6, GE design) NPPs, whose development is being carried out by IBERDROLA. The analysis which has been performed up to date for Trillo is limited to the Plant Damage State (PDS) definition. However, the main phenomena challenging its containment performance have been identified, and they are summarized here. The Level 2 analysis for Cofrentes is comparatively more developed. The main phenomena and the key equipment affecting its containment behaviour are presented. Finally the conclusions of this report are elaborated. (author)

  13. Development of criticality accident detector measuring neutrons and gamma-rays

    International Nuclear Information System (INIS)

    Tsujimura, Norio; Yoshida, Tadayoshi; Ishii, Masato

    2005-01-01

    The authors developed a new criticality accident detector measuring neutrons and gamma-rays. The detector is a cylindrical plastic scintillator coupled to a current-mode operated photomultiplier, and is covered by an inner cadmium shell, acting as a neutron to gamma-ray converter, and a 5cm thick outer polyethylene moderator in order to respond to the same threshold triggering dose regardless of whether it was exposed to neutrons, gamma-rays or a mixture of the two radiations. (author)

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

    International Nuclear Information System (INIS)

    O'Kula, K.R.; Taylor, R.P. Jr.; Ashbaugh, S.G.

    1995-01-01

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

  15. Nuclear accidents

    International Nuclear Information System (INIS)

    1987-01-01

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

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

    International Nuclear Information System (INIS)

    Hancock, R.; Woollam, P.B.

    1987-07-01

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

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

    International Nuclear Information System (INIS)

    Mueller, C.; Roglans-Ribas, J.; Folga, S.; Nabelssi, B.; Jackson, R.

    1995-01-01

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

  18. Nuclear accident dosimetry measurements at the fourth IAEA intercomparison, Harwell, United Kingdom, April 1975

    International Nuclear Information System (INIS)

    Majborn, B.

    1975-11-01

    During the fourth IAEA intercomparison at Harwell, teams from several countries compared their systems for assessing doses in criticality accidents. The dosimeters were exposed to mixed pulses of neutron and gamma radiation produced by the VIPER reactor at AWRE, Aldermaston. Risoe participated in this intercomparison with the routine personnel dosimeters employed in 'criticality areas' at Risoe. These include the UKAEA criticality dosimeter for the measurement of neutron doses and the Risoe TLD badge for the measurement of gamma doses. The final results of the Risoe measurements are presented in this report in a form designed for ease of comparison with results of other pparticipants. (author)

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

    International Nuclear Information System (INIS)

    Rao, Suman

    2007-01-01

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

  20. Study on entry criteria for severe accident management during hot leg LBLOCAs in a PWR

    International Nuclear Information System (INIS)

    Zhang, Longfei; Zhang, Dafa; Wang, Shaoming

    2007-01-01

    The risk of Large Break Loss of Coolant Accidents (LBLOCA) has been considered an important safety issue since the beginning of the nuclear power industry. The rapid depressurization occurs in the primary coolant circuit when a large break appears in a Pressurized Water Reactors (PWR).Then the coolant temperature reaches saturation at a very low pressure. The core outlet fluid temperatures maybe not reliable indicators of the core damage states at a such lower pressure. The problem is how to decide the time for water injection in the SAM (Severe Accident Management). An alternative entry criterion is the fluid temperature just above the hot channel in which the fluid temperature showed maximum among all the channels. For that reason, a systematic study of entry criterion of SAM for different hot leg break sizes in a 3-loop PWR has been started using the detailed system thermal hydraulic and severe accident analysis code package, RELAP/SCDAPSIM. Best estimate calculations of the large break LOCA of 15 cm, 20 cm and 25 cm without accident managements and in the case of high-pressure safety injection as the accident management were performed in this paper. The analysis results showed that the core exit temperatures are not reliable indicators of the peak core temperatures and core damage states once peak core temperatures reach 1500 K, and the proposed entry criteria for SAM at the time when the core outlet temperature reaches 900 K is not effective to prevent core melt. Then other analyses were performed with a parameter of fluid temperature just above the hot channel. The latter analysis showed that earlier water injection when the fluid temperature just above the hot channel reaches 900 K is effective to prevent further core melt. Since fuel surface and hot channel have spatial distribution and depend on a period of cycle operation, a series of thermocouples are required to install just above the fuel assembly. The maximum exit temperature of 900 K that captured by

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

    International Nuclear Information System (INIS)

    Lahtinen, J.

    2006-04-01

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

  2. Emergency Management and Radiation Moni-toring in Nuclear and Radiological Accidents. Summary Report on the NKS Project EMARAD

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-04-15

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

  3. Use of an influence diagram and fuzzy probability for evaluating accident management in a boiling water reactor

    International Nuclear Information System (INIS)

    Yu, D.; Kastenberg, W.E.; Okrent, D.

    1994-01-01

    A new approach is presented for evaluating the uncertainties inherent in severe accident management strategies. At first, this analysis considers accident management as a decision problem (i.e., applying a strategy compared with do nothing) and uses an influence diagram. To evaluate imprecise node probabilities in the influence diagram, the analysis introduces the concept of a fuzzy probability. When fuzzy logic is applied, fuzzy probabilities are easily propagated to obtain results. In addition, the results obtained provide not only information similar to the classical approach, which uses point-estimate values, but also additional information regarding the impact of using imprecise input data. As an illustrative example, the proposed methodology is applied to the evaluation of the drywell flooding strategy for a long-term station blackout sequence at the Peach Bottom nuclear power plant. The results show that the drywell flooding strategy is beneficial for preventing reactor vessel breach. It is also effective for reducing the probability of containment failure for both liner melt-through and late overpressurization. Even though uncertainty exists in the results, flooding is preferred to do nothing when evaluated in terms of two risk measures: early and late fatalities

  4. Examination of off-site emergency protective measures for core melt accidents

    International Nuclear Information System (INIS)

    Aldrich, D.C.; Ericson, D.M. Jr.; Jones, R.B.

    1978-01-01

    Results from the Reactor Safety Study (RSS) have shown that to cause significant impacts off-site, i.e., sufficient quantities of biologically important radionuclides released, it is necessary to have a core melt accident. To mitigate the impact of such potential accidents, the design of appropriate emergency response actions requires information as to the relative merit of publicly available protective measures. In order to provide this information, a study using the consequence model developed for the RSS is being conducted to evaluate (in terms of reduced public health effects and dose exposure) potential off-site protective strategies. The paper describes the methods being used in the study as well as the results and conclusions obtained

  5. Environmental radioactivity measurements in greece following the Fukushima Daichi nuclear accident

    International Nuclear Information System (INIS)

    Potiriadis, C.; Kolovou, M.; Clouvas, A.; Xanthos, S.

    2008-01-01

    Since the double disaster of the 9.0 magnitude earthquake and tsunami that affected hundreds of thousands of people and seriously damaged the Fukushima Daichi power plant in Japan on 11 March 2011, traces of radioactive emissions from Fukushima have spread across the entire northern hemisphere. The radioactive isotope of iodine 131 I that was generated by the nuclear accident in Fukushima arrived in Greece on 24 March 2011. Radioactive iodine is present in the air either as gas or bound to particles (aerosols). The maximum 131 I concentrations were measured between 3 and 5 April 2011. In aerosols the maximum 131 I values measured in Southern Greece (Athens) and Northern Greece (Thessaloniki) were 585±70 and 408±61 μBq m -3 , respectively 131 I concentrations in gas were about 3.5 times higher than in aerosols. Since 29 April 2011, the 131 I concentration has been below detection limits. Traces of 137 Cs and 134 Cs were also measured in the air filters with an activity ratio of 137 Cs/ 134 Cs equal to 1 and 131 I/ 137 Cs activity ratio of about 3. Since 16 May 2011, the 137 Cs concentration in air has been determined to be about the same as before the Fukushima accident. Traces of 131 I were also measured in grass and milk. The maximum measured activity of 131 I in sheep milk was about 2 Bq l -1 which is 5000 times less than that measured in Greece immediately after the Chernobyl accident. The measured activity concentrations of artificial radionuclides in Greece due to the Fukushima release, have been very low, with no impact on human health. (authors)

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

    International Nuclear Information System (INIS)

    Hanson, D.J.; Golden, D.W.; Chambers, R.; Miller, J.D.; Hallbert, B.P.; Dobbe, C.A.

    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 ∼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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    Guarro, S.; Milici, T.; Wu, J.S.; Apostolakis, G.

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    Villers, A.

    2008-01-01

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

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

    International Nuclear Information System (INIS)

    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.

    2007-04-01

    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

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

    International Nuclear Information System (INIS)

    Yoshida, Yoshitaka; Ohtani, Masanori; Fujita, Yushi

    2002-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, Yoshitaka; Ohtani, Masanori [Institute of Nuclear Safety System, Inc., Mihama, Fukui (Japan); Fujita, Yushi [TECNOVA Corp., Tokyo (Japan)

    2002-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-04-15

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

  14. Severe accident management at nuclear power plants - emergency preparedness and response actions

    International Nuclear Information System (INIS)

    Pawar, S.K.; Krishnamurthy, P.R.

    2015-01-01

    This paper describes the current level of emergency planning and preparedness and also improvement in the emergency management programme over the years including lessons learned from Fukushima accident, hazard analysis and categorization of nuclear facilities into hazard category for establishing the emergency preparedness class, classification of emergencies based on the Emergency Action Levels (EAL), development of EAL’s for PHWR, Generic Criteria in terms of projected dose for initiating protective actions (precautionary urgent protective actions, urgent protective actions, early protective actions), operational intervention levels (OIL), Emergency planning zones and distances, protection strategy and reference levels, use of residual dose for establishing reference levels for optimization of protection strategy, criteria for termination of emergency, transition of emergency exposure situation to existing exposure situation or planned exposure situation, criteria for medical managements of exposed persons and guidance for controlling the dose of emergency workers. This paper also highlights the EALs for typical PHWR type reactors for all types of emergencies (plant, site and offsite), transition from emergency operating procedures (EOP) to accident management guidelines (AMG) to emergency response actions and proposed implementation of guidelines

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

    International Nuclear Information System (INIS)

    Nisbet, A.

    2002-01-01

    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)

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

    CERN Document Server

    Nakamura, T; Tonoike, K

    2003-01-01

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

  17. On-line measurements of RuO{sub 4} during a PWR severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Reymond-Laruinaz, S.; Doizi, D. [CEA, DEN, Departement de Physico-chimie, CEA/Saclay, 91191 Gif sur Yvette Cedex, (France); Manceron, L. [Societe Civile Synchrotron SOLEIL, L' Orme des Merisiers, St-Aubin BP48, 91192 Gif-sur-Yvette Cedex, (France); MONARIS, UMR 8233, Universite Pierre et Marie Curie, 4 Place Jussieu, case 49, F-75252 Paris Cedex 05, (France); Boudon, V. [Laboratoire Interdisciplinaire Carnot de Bourgogne, UMR 6303 CNRS-Universite de Bourgogne, 9 avenue Alain Savary, BP 47870, F-21078 Dijon Cedex, (France); Ducros, G. [CEA, DEN, Departement d' Etudes des Combustibles, CEA/Cadarache, 13108 Saint-Paul-lez-Durance cedex, (France)

    2015-07-01

    After the Fukushima accident, it became essential to have a way to monitor in real time the evolution of a nuclear reactor during a severe accident, in order to react efficiently and minimize the industrial, ecological and health consequences of the accident. Among gaseous fission products, the tetroxide of ruthenium RuO{sub 4} is of prime importance since it has a significant radiological impact. Ruthenium is a low volatile fission product but in case of the rupture of the vessel lower head by the molten corium, the air entering into the vessel oxidizes Ru into gaseous RuO{sub 4}, which is not trapped by the Filtered Containment Venting Systems. To monitor the presence of RuO{sub 4} allows making a diagnosis of the core degradation and quantifying the release into the atmosphere. To determine the presence of RuO{sub 4}, FTIR spectrometry was selected. To study the feasibility of the monitoring, high-resolution IR measurements were realized at the French synchrotron facility SOLEIL on the infrared beam line AILES. Thereafter, theoretical calculations were done to simulate the FTIR spectrum to describe the specific IR fingerprint of the molecule for each isotope and based on its partial pressure in the air. (authors)

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

    Science.gov (United States)

    Lujanienė, G; Byčenkienė, S; Povinec, P P; Gera, M

    2012-12-01

    Analyses of (131)I, (137)Cs and (134)Cs in airborne aerosols were carried out in daily samples in Vilnius, Lithuania after the Fukushima accident during the period of March-April, 2011. The activity concentrations of (131)I and (137)Cs ranged from 12 μBq/m(3) and 1.4 μBq/m(3) to 3700 μBq/m(3) and 1040 μBq/m(3), respectively. The activity concentration of (239,240)Pu in one aerosol sample collected from 23 March to 15 April, 2011 was found to be 44.5 nBq/m(3). The two maxima found in radionuclide concentrations were related to complicated long-range air mass transport from Japan across the Pacific, the North America and the Atlantic Ocean to Central Europe as indicated by modelling. HYSPLIT backward trajectories and meteorological data were applied for interpretation of activity variations of measured radionuclides observed at the site of investigation. (7)Be and (212)Pb activity concentrations and their ratios were used as tracers of vertical transport of air masses. Fukushima data were compared with the data obtained during the Chernobyl accident and in the post Chernobyl period. The activity concentrations of (131)I and (137)Cs were found to be by 4 orders of magnitude lower as compared to the Chernobyl accident. The activity ratio of (134)Cs/(137)Cs was around 1 with small variations only. The activity ratio of (238)Pu/(239,240)Pu in the aerosol sample was 1.2, indicating a presence of the spent fuel of different origin than that of the Chernobyl accident. Copyright © 2011 Elsevier Ltd. All rights reserved.

  19. Helping HSE Team in Learning from Accident by Using the Management Oversight and Risk Tree Analysis Method

    Directory of Open Access Journals (Sweden)

    Iraj Mohammadfam

    2016-09-01

    Conclusion: The analysis using MORT method helped the organization with learning lessons from the accident especially at the management level. In order to prevent the similar and dissimilar accidents, the inappropriate informational network within the organization, inappropriate operational readiness, lack of proper implementation of work permit, the inappropriate and lack of updated technical information systems regarding equipments and working process, and the inappropriate barriers should be considered in a special way.

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

    International Nuclear Information System (INIS)

    Miettinen, J.; Schmuck, P.

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Farfan, E.; Jannik, T.

    2011-01-01

    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 t