Processing Expert Judgements in Accident Consequence Modelling (invited paper)

International Nuclear Information System (INIS)

Kraan, B.C.P.; Cooke, R.M.

2000-01-01

In performing uncertainty analysis a distribution on the code input parameters is required. The construction of the distribution on the code input parameters for the joint CEC/USNRC Accident Consequence Code Uncertainty Analysis using Expert Judgement is discussed. An example from the food chain module is used to illustrate the construction. Different mathematical techniques have been developed to transform the expert judgements into the required format. Finally, the effect of taking account of correlations in performing uncertainty analysis is investigated. (author)

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

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)

The Fukushima Daiichi Accident. Technical Volume 1/5. Description and Context of the Accident. Annexes

International Nuclear Information System (INIS)

2015-08-01

The Fukushima Daiichi Accident consists of a Report by the IAEA Director General and five technical volumes. It is the result of an extensive international collaborative effort involving five working groups with about 180 experts from 42 Member States with and without nuclear power programmes and several international bodies. It provides a description of the accident and its causes, evolution and consequences, based on the evaluation of data and information from a large number of sources available at the time of writing. The Fukushima Daiichi Accident will be of use to national authorities, international organizations, nuclear regulatory bodies, nuclear power plant operating organizations, designers of nuclear facilities and other experts in matters relating to nuclear power, as well as the wider public. The set contains six printed parts and five supplementary CD-ROMs. Contents: Report by the Director General; Technical Volume 1/5, Description and Context of the Accident; Technical Volume 2/5, Safety Assessment; Technical Volume 3/5, Emergency Preparedness and Response; Technical Volume 4/5, Radiological Consequences; Technical Volume 5/5, Post-accident Recovery; Annexes. The Report by the Director General is available separately in Arabic, Chinese, English, French, Russian, Spanish and Japanese

Undiagnosed post-traumatic stress disorder following motor vehicle accidents.

Science.gov (United States)

Green, M M; McFarlane, A C; Hunter, C E; Griggs, W M

1993-10-18

To determine the pattern of emergence of post-traumatic stress disorder (PTSD) among motor vehicle accident victims and to examine the influence of PTSD on subsequent levels of disability. A longitudinal study of motor vehicle accident victims one month and 18 months after the accident. Twenty-four motor vehicle accident victims admitted by the trauma team at the Royal Adelaide Hospital. A 52% response rate was achieved. Post-traumatic stress disorder as diagnosed by the Diagnostic Interview Schedule and disability as measured with the Sickness Impact Profile. Eighteen months after their accidents, six of the 24 subjects had clinically significant PTSD and one was considered borderline. None had been previously diagnosed or treated. The group with PTSD had higher scores on all measures of psychological distress one month after the accident and were more likely to use immature psychological defences. There was no association between physical outcome (measured with the modified Glasgow Outcome Scale) at six months and subsequent diagnosis of PTSD. However, the group with PTSD had higher levels of disability on assessment with the Sickness Impact Profile, particularly in the domain of social functioning. The results suggest PTSD was associated with work-related dysfunction equal to that associated with severe physical handicap. The data from this pilot study suggest that PTSD after motor vehicle accidents is an important cause of disability, which may also become the focus for damages in litigation. Thus, there is a need for further investigation of the early patterns of distress and to design preventive programs for victims of road accidents.

Safety Research Opportunities Post-Fukushima. Initial Report of the Senior Expert Group

International Nuclear Information System (INIS)

Baek, Won-Pil; Yang, Joon-Eon; Ball, Joanne; Glowa, Glenn; Bisconti, Giulia; Peko, Damian; Bolshov, Leonid; Burgazzi, Luciano; De Rosa, Felice; Conde, Jose M.; Cook, Gary; Evrard, Jean-Michel; Jacquemain, Didier; Funaki, Kentaro; Uematsu, Mari Marianne; Miyoshi, Katsumasa; Tatematsu, Atsushi; Hirano, Masashi; Hoshi, Harutaka; Kawaragi, Chie; Kobayashi, Youko; Sakamoto, Kazunobu; Journeau, Christophe; Kim, Han-Chul; Klein-Hessling, Walter; Sonnenkalb, Martin; Koganeya, Toshiyuki; White, Andrew; ); Lind, Terttaliisa; Zimmermann, Martin; Lindholm, Ilona; Castelo Lopez, Carlos; Nagase, Fumihisa; Washiya, Tadahiro; Oima, Hirofumi; Okada, Hiro; Richards, Stuart; West, Steven; Sandberg, Nils; Suzuki, Shunichi; Vitanza, Carlo; Yamanaka, Yasunori

2017-02-01

One of the imperatives following the accident at the Fukushima Daiichi nuclear power station is for the nuclear science and industry communities to ensure that knowledge gaps in nuclear safety are identified and that research programs to address these gaps are being instituted. In recognition of broad international interest in additional information that could be gained from post-accident examinations related to Fukushima Daiichi, Japan recommended to the Committee on the Safety of Nuclear Installations (CSNI) in June 2013 that a process be developed to identify and follow up on opportunities to address safety research gaps. Consequently, a Senior Expert Group (SEG) on Safety Research Opportunities post-Fukushima (SAREF) was formed. The members of the group are senior technical experts from technical support organisations, nuclear regulatory authorities and Japanese organisations responsible for planning and execution of Fukushima Daiichi decommissioning. The domain of interest for the group is activities that address safety research knowledge gaps and also the needs of Fukushima Daiichi decommissioning. SEG on SAREF identified areas where these two interests intersect or overlap, and activities that could be undertaken to generate information of common benefit. The group's output is documented in this report; Chapter 2 describes the current status of the damaged units at Fukushima Daiichi NPS; Chapter 3 summarises safety research areas of common interest; Chapter 4 summarises the safety research activities recommended as short-term projects; Chapter 5 summarises those as long-term considerations; Chapter 6 supplies conclusions and recommendations. The appendix contains detailed information compiled by the SEG members on all safety research areas of interest

Status Report on Activities of the Systems Assessment Task Force, OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs

Energy Technology Data Exchange (ETDEWEB)

Bragg-Sitton, Shannon Michelle [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

The Organization for Economic Cooperation and Development /Nuclear Energy Agency (OECD/NEA) Nuclear Science Committee approved the formation of an Expert Group on Accident Tolerant Fuel (ATF) for LWRs (EGATFL) in 2014. Chaired by Kemal Pasamehmetoglu, INL Associate Laboratory Director for Nuclear Science and Technology, the mandate for the EGATFL defines work under three task forces: (1) Systems Assessment, (2) Cladding and Core Materials, and (3) Fuel Concepts. Scope for the Systems Assessment task force (TF1) includes definition of evaluation metrics for ATF, technology readiness level definition, definition of illustrative scenarios for ATF evaluation, and identification of fuel performance and system codes applicable to ATF evaluation. The Cladding and Core Materials (TF2) and Fuel Concepts (TF3) task forces will identify gaps and needs for modeling and experimental demonstration; define key properties of interest; identify the data necessary to perform concept evaluation under normal conditions and illustrative scenarios; identify available infrastructure (internationally) to support experimental needs; and make recommendations on priorities. Where possible, considering proprietary and other export restrictions (e.g., International Traffic in Arms Regulations), the Expert Group will facilitate the sharing of data and lessons learned across the international group membership. The Systems Assessment task force is chaired by Shannon Bragg-Sitton (Idaho National Laboratory [INL], U.S.), the Cladding Task Force is chaired by Marie Moatti (Electricite de France [EdF], France), and the Fuels Task Force is chaired by a Masaki Kurata (Japan Atomic Energy Agency [JAEA], Japan). The original Expert Group mandate was established for June 2014 to June 2016. In April 2016 the Expert Group voted to extend the mandate one additional year to June 2017 in order to complete the task force deliverables; this request was subsequently approved by the Nuclear Science Committee. This

Status Report on Activities of the Systems Assessment Task Force, OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs

International Nuclear Information System (INIS)

Bragg-Sitton, Shannon Michelle

2016-01-01

The Organization for Economic Cooperation and Development /Nuclear Energy Agency (OECD/NEA) Nuclear Science Committee approved the formation of an Expert Group on Accident Tolerant Fuel (ATF) for LWRs (EGATFL) in 2014. Chaired by Kemal Pasamehmetoglu, INL Associate Laboratory Director for Nuclear Science and Technology, the mandate for the EGATFL defines work under three task forces: (1) Systems Assessment, (2) Cladding and Core Materials, and (3) Fuel Concepts. Scope for the Systems Assessment task force (TF1) includes definition of evaluation metrics for ATF, technology readiness level definition, definition of illustrative scenarios for ATF evaluation, and identification of fuel performance and system codes applicable to ATF evaluation. The Cladding and Core Materials (TF2) and Fuel Concepts (TF3) task forces will identify gaps and needs for modeling and experimental demonstration; define key properties of interest; identify the data necessary to perform concept evaluation under normal conditions and illustrative scenarios; identify available infrastructure (internationally) to support experimental needs; and make recommendations on priorities. Where possible, considering proprietary and other export restrictions (e.g., International Traffic in Arms Regulations), the Expert Group will facilitate the sharing of data and lessons learned across the international group membership. The Systems Assessment task force is chaired by Shannon Bragg-Sitton (Idaho National Laboratory [INL], U.S.), the Cladding Task Force is chaired by Marie Moatti (Electricite de France [EdF], France), and the Fuels Task Force is chaired by a Masaki Kurata (Japan Atomic Energy Agency [JAEA], Japan). The original Expert Group mandate was established for June 2014 to June 2016. In April 2016 the Expert Group voted to extend the mandate one additional year to June 2017 in order to complete the task force deliverables; this request was subsequently approved by the Nuclear Science Committee. This

Research on alarm triggered fault-diagnosis expert system for U-shaped tube breaking accident of steam generators

International Nuclear Information System (INIS)

Qian Hong; Luo Jianbo; Jin Weixiao; Zhou Jinming; Wang Du

2015-01-01

According to the U-shaped tube breaking accident of steam generator (SGTR), this paper designs a fault-diagnosis expert system based on the alarm triggering. By analyzing the fault mechanism of SGTR accidents, the fault symptom is obtained. The parameters of the belief rule are set up based on the simulation experiment. The information fusion is conducted on the fault-diagnosis results from multiple expert systems to obtain the final diagnose result. The test result shows that the expert system can diagnose the SGTR accident accurately and rapidly, and provide with the operation guidance. (authors)

The post-accident nuclear issue: the new crisis expertise challenges for the IRSN; Post-accidentel nucleaire: les nouveaux challenges de l'expertise de crise a L'IRSN

Energy Technology Data Exchange (ETDEWEB)

Champion, D. [Institut de Radioprotection et de Surete Nucleaire (IRSN), Direction de l' environnement et de l' intervention, 78 - Le Vesinet (France)

2010-07-01

The author reports the work performed by two work groups conducted by the IRSN (the French Radioprotection and Nuclear Safety Institute), the first one on the issue of assessment of radiological and dosimetric consequences in a post-accident situation, and the second one on hypotheses to be used to perform predictive assessments of these consequences. First dealing with the end of the emergency phase, he describes how to anticipate actions of protection against immediate post-accident consequences: orientation of the expertise strategy based on the CODIRPA's doctrine, post-accident zoning based on predictive indicators, use of reasonably prudent hypotheses for the first predictive assessments, importance of initial radioactive deposits to perform predictive assessments. Then, the author presents an iterative method of assessment of post-accident consequences: organization of environment radioactivity measurement programmes, periodic update of mapping of initial deposit and of actual deposits at a given time

Psychological and social impacts of post-accident situations: lessons from the Chernobyl accident

International Nuclear Information System (INIS)

Lochard, J.

1996-01-01

This paper presents the main features, from the psychological and social points of view, of the post-accident situation in the contaminated areas around Chernobyl. This is based on a series of surveys performed in the concerned territories of the CIS republics. The high level of stress affecting a large segment of the population is related to the perception of the situation by those living in a durably contaminated environment but also to the side-effects of some of the countermeasures adopted to mitigate the radiological consequences or to compensate the affected population. The distinction between the accident and the post-accident phase is enlarged to take into account the various phases characterizing the dynamics of the social response. Although the size of the catastrophe as well as the economic and political conditions that were prevailing at the time and after the accident have resulted in a maximal intensity of the reactions of the population, many lessons can be drawn for the management of potential post-accident situations. (author)

Risk assessment for long-term post-accident sequences

International Nuclear Information System (INIS)

Ellia-Hervy, A.; Ducamp, F.

1987-11-01

Probabilistic risk analysis, currently conducted by the CEA (French Atomic Energy Commission) for the French replicate series of 900 MWe power plants, has identified accident sequences requiring long-term operation of some systems after the initiating event. They have been named long-term sequences. Quantification of probabilities of such sequences cannot rely exclusively on equipment failure-on-demand data: it must also take into account operating failures, the probability of which increase with time. Specific studies have therefore been conducted for a number of plant systems actuated during these long-term sequences. This has required: - Definition of the most realistic equipment utilization strategies based on existing emergency procedures for 900 MWe French plants. - Evaluation of the potential to repair failed equipment, given accessibility, repair time, and specific radiation conditions for the given sequence. - Definition of the event bringing the long-term sequence to an end. - Establishment of an appropriate quantification method, capable of taking into account the evolution of assumptions concerning equipment utilization strategies or repair conditions over time. The accident sequence quantification method based on realistic scenarios has been used in the risk assessment of the initiating event loss of reactor coolant accident occurring at power and at shutdown. Compared with the results obtained from conventional methods, this method redistributes the relative weight of accident sequences and also demonstrates that the long term can be a significant contribution to the probability of core melt

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

Post-accident fuel relocation and heat removal in the LMFBR

International Nuclear Information System (INIS)

Kazimi, M.S.; Tsai, S.S.; Gasser, R.D.

1976-08-01

Assessment of the dynamics of post-accident fuel relocation and heat removal is an important aspect of the evaluation of the consequences of a hypothetical accident in an LMFBR. Such an assessment is of particular importance in the evaluation of the post-accident radiological doses around the reactor site. In the present evaluation particular attention is given to the design features of the Clinch River Breeder Reactor Plant (CRBR). Fuel relocation and heat removal, assuming certain conditions have resulted in core disruption, are discussed. The discussion of events and phenomena involved in the relocation processes is centered around the resulting patterns of heat source distribution. The factors influencing fuel relocation and distribution in the inlet and outlet plena of the reactor vessel are discussed. The current technology of in-vessel heat removal is applied to the design of the CRBR reactor. Both fuel debris cooling limits and overall coolant flow in the reactor under natural convection conditions are explored. Some of the uncertainties in ex-vessel fuel behavior are addressed. In particular, the effect of melting the cavity bed on the rate of growth of a molten fuel pool is investigated

Use of expert judgement in NUREG-1150

International Nuclear Information System (INIS)

Ortiz, N.R.; Wheeler, T.A.; Breeding, R.J.; Hora, S.; Meyer, M.A.; Kenney, R.L.

1991-01-01

The explicit expert judgment process used in NUREG-1150, 'Severe Accident Risks: An Assessment for Five US Nuclear Plants', is discussed in this paper. The main steps of the process are described, including selection of issues and experts, elicitation training, presentation of issues to the experts, preparation of issue analyses by the experts, discussion of issue analyses and elicitation, and recomposition and aggregation of results. To demonstrate the application of the expert judgment process to NUREG-1150, two issues are summarized: one from the accident frequency analysis, and one from the accident progression analysis. Recommendations and insights are provided to improve the use of explicit expert judgment in complex technical issues. (orig.)

A Human Reliability Analysis of Post- Accident Human Errors in the Low Power and Shutdown PSA of KSNP

Energy Technology Data Exchange (ETDEWEB)

Kang, Daeil; Kim, J. H.; Jang, S. C

2007-03-15

Korea Atomic Energy Research Institute, using the ANS low power and shutdown (LPSD) probabilistic risk assessment (PRA) Standard, evaluated the LPSD PSA model of the KSNP, Yonggwang Units 5 and 6, and identified the items to be improved. The evaluation results of human reliability analysis (HRA) of the post-accident human errors in the LPSD PSA model for the KSNP showed that 10 items among 19 items of supporting requirements for those in the ANS PRA Standard were identified as them to be improved. Thus, we newly carried out a HRA for post-accident human errors in the LPSD PSA model for the KSNP. Following tasks are the improvements in the HRA of post-accident human errors of the LPSD PSA model for the KSNP compared with the previous one: Interviews with operators in the interpretation of the procedure, modeling of operator actions, and the quantification results of human errors, site visit. Applications of limiting value to the combined post-accident human errors. Documentation of information of all the input and bases for the detailed quantifications and the dependency analysis using the quantification sheets The assessment results for the new HRA results of post-accident human errors using the ANS LPSD PRA Standard show that above 80% items of its supporting requirements for post-accident human errors were graded as its Category II. The number of the re-estimated human errors using the LPSD Korea Standard HRA method is 385. Among them, the number of individual post-accident human errors is 253. The number of dependent post-accident human errors is 135. The quantification results of the LPSD PSA model for the KSNP with new HEPs show that core damage frequency (CDF) is increased by 5.1% compared with the previous baseline CDF It is expected that this study results will be greatly helpful to improve the PSA quality for the domestic nuclear power plants because they have sufficient PSA quality to meet the Category II of Supporting Requirements for the post-accident

A Human Reliability Analysis of Post- Accident Human Errors in the Low Power and Shutdown PSA of KSNP

International Nuclear Information System (INIS)

Kang, Daeil; Kim, J. H.; Jang, S. C.

2007-03-01

Korea Atomic Energy Research Institute, using the ANS low power and shutdown (LPSD) probabilistic risk assessment (PRA) Standard, evaluated the LPSD PSA model of the KSNP, Yonggwang Units 5 and 6, and identified the items to be improved. The evaluation results of human reliability analysis (HRA) of the post-accident human errors in the LPSD PSA model for the KSNP showed that 10 items among 19 items of supporting requirements for those in the ANS PRA Standard were identified as them to be improved. Thus, we newly carried out a HRA for post-accident human errors in the LPSD PSA model for the KSNP. Following tasks are the improvements in the HRA of post-accident human errors of the LPSD PSA model for the KSNP compared with the previous one: Interviews with operators in the interpretation of the procedure, modeling of operator actions, and the quantification results of human errors, site visit. Applications of limiting value to the combined post-accident human errors. Documentation of information of all the input and bases for the detailed quantifications and the dependency analysis using the quantification sheets The assessment results for the new HRA results of post-accident human errors using the ANS LPSD PRA Standard show that above 80% items of its supporting requirements for post-accident human errors were graded as its Category II. The number of the re-estimated human errors using the LPSD Korea Standard HRA method is 385. Among them, the number of individual post-accident human errors is 253. The number of dependent post-accident human errors is 135. The quantification results of the LPSD PSA model for the KSNP with new HEPs show that core damage frequency (CDF) is increased by 5.1% compared with the previous baseline CDF It is expected that this study results will be greatly helpful to improve the PSA quality for the domestic nuclear power plants because they have sufficient PSA quality to meet the Category II of Supporting Requirements for the post-accident

Psychical and social effects related to post-accident situations: some training of Chernobyl accident

International Nuclear Information System (INIS)

Lochand, J.

1995-01-01

Some preliminary considerations on the psychic and societal dimensions related to post-accident situations connected to large scale and heavy land contamination are presented. This is done with the objective of exploring the role that these dimensions could play in the elaboration of new radiological protection principles and concepts in order to restore confidence among affected populations after a nuclear accident. It is important to facilitate the return to normal or, at least, acceptable living conditions, as soon as reasonably achievable, and to prevent the possible emergence of a post-accident crisis. A scheme is proposed for understanding the dynamics of the various phases after an accident, taking into account the collective response to the consequences as well as, the response to the countermeasures. (Author)

Chernobyl accident: causes and consequences (expert conclusion). Part 3. Chernobyl accident effect on Belarus

International Nuclear Information System (INIS)

Nesterenko, V.B.

1992-01-01

Expert conclusion is presented on the Chernobyl accident effect on Belarus. Problems of ground and food contamination, medical and biological radiation effects on the population are considered. Attention is paid to the radiation monitoring and radiometric gages. Scale of the damage for forestry and agriculture is described and recommendations on the agriculture is described and recommendations on the agricultural production and forest utilization at contaminated areas are given. 24 refs.; 4 figs.; 24 tabs

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.)

The role of post accident chemistry data in nuclear safety

International Nuclear Information System (INIS)

Bradshaw, R.W.; Caruthers, G.F.

1982-01-01

The NRC instituted the NUREG-0737 requirements as implementation of the Post-TMI Action Plan in October, 1980. Among these requirements was the capability to obtain chemistry samples of the reactor coolant and containment building atmosphere under post accident conditions. The quantitative criteria were, in general, beyond the capabilities of existing plant systems. As a consequence the nuclear industry expended substantial efforts to design and install the post-accident sampling systems necessary to comply with these criteria. With such efforts essentially complete, the task remains to establish the role that data provided by these systems would play in mitigating the consequences of a nuclear plant accident. This role definition must include a characterization of the timing and priority for the post accident chemistry data. This paper defines that role using the Safety Level and Safety Function concepts as a matrix

Status Report on Activities of the Systems Assessment Task Force, OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs

Energy Technology Data Exchange (ETDEWEB)

Bragg-Sitton, Shannon Michelle [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-09-01

The Organization for Economic Cooperation and Development /Nuclear Energy Agency (OECD/NEA) Nuclear Science Committee approved the formation of an Expert Group on Accident Tolerant Fuel (ATF) for LWRs (EGATFL) in 2014. Chaired by Kemal Pasamehmetoglu, INL Associate Laboratory Director for Nuclear Science and Technology, the mandate for the EGATFL defines work under three task forces: (1) Systems Assessment, (2) Cladding and Core Materials, and (3) Fuel Concepts. Scope for the Systems Assessment task force includes definition of evaluation metrics for ATF, technology readiness level definition, definition of illustrative scenarios for ATF evaluation, parametric studies, and selection of system codes. The Cladding and Core Materials and Fuel Concepts task forces will identify gaps and needs for modeling and experimental demonstration; define key properties of interest; identify the data necessary to perform concept evaluation under normal conditions and illustrative scenarios; identify available infrastructure (internationally) to support experimental needs; and make recommendations on priorities. Where possible, considering proprietary and other export restrictions (e.g., International Traffic in Arms Regulations), the Expert Group will facilitate the sharing of data and lessons learned across the international group membership. The Systems Assessment Task Force is chaired by Shannon Bragg-Sitton (INL), while the Cladding Task Force will be chaired by a representative from France (Marie Moatti, Electricite de France [EdF]) and the Fuels Task Force will be chaired by a representative from Japan (Masaki Kurata, Japan Atomic Energy Agency [JAEA]). This report provides an overview of the Systems Assessment Task Force charter and status of work accomplishment.

Proceedings of the Second Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs, 23-25 September 2014, OECD-NEA HQ

International Nuclear Information System (INIS)

Massara, S.; ); Bragg-Sitton, Shannon; Pasamehmetoglu, K.; Yang, Jae Ho; Dolley, Evan J.; Rebak, Raul B.; Sowder, Andrew; Cheng, Bo; Kurata, Masaki; Van Nieuwenhove, Rudi; Li, R.; McClellan, Ken; Nelson, Andy; Carmack, Jon; Harp, Jason; Finck, Phillip; ); Kakicuhi, K.

2014-09-01

Under the guidance of the OECD-NEA Nuclear Science Committee, the expert group acts as a forum for scientific and technical information exchange on advanced light water reactor (LWR) fuels with enhanced accident tolerance. The expert group focusses on the fundamental properties and behaviour under normal operations and accident conditions for advanced core materials and components (fuels, cladding, control rods, etc.). The materials considered are applicable to Gen II and Gen III Light Water Reactors, as well as Gen III+ reactors under construction. The objective of the expert group is to define and coordinate a programme of work to help advance the scientific knowledge needed to provide the technical underpinning for the development of advanced LWR fuels with enhanced accident tolerance compared to currently used zircaloy/UO 2 fuel systems, as well as other non-fuel core components with important roles in LWR performance under accident conditions. This document brings together the available presentations (slides) given at the Second Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs. Content: 1 - Proposed Agenda; 2 - Expert Group meeting - 23 September 2014: - Introduction and background (S. Massara, OECD-NEA) - Expected outcomes from the TFs meetings scheduled on 24-25 September (K. Pasamehmetoglu, EG Chair, INL); 3 - Task Force 1 (Systems assessment) meeting - 24 September 2014: - Metrics for the Evaluation of LWR Accident Tolerant Fuel (S. Bragg-Sitton, INL); 4 - Task Force 2 (Cladding/core materials) meeting - 24 September 2014: - Summary on SiC Task Force 2 (Clad) meeting (J.H. Yang, KAERI); - Accident Tolerant Advanced Steels Cladding for Commercial Light Water Reactors (E. Dolley, GE); - Molybdenum-Alloy Fuel Cladding Development and Testing - Update from April 2014 NEA ATF Meeting (A. Sowder, EPRI); - Accident Tolerant Control Rod Development in Japan (M. Kurata, JAEA); - IFA-774: The first in-pile test with coated fuel rods (R. Van

International Experts' Meeting on Decommissioning and Remediation after a Nuclear Accident. Presentations

International Nuclear Information System (INIS)

2013-01-01

Against the backdrop of the accident at TEPCO's Fukushima Daiichi nuclear power plant in March 2011, the Director General of the International Atomic Energy Agency (IAEA) convened the IAEA Ministerial Conference on Nuclear Safety in Vienna, Austria, in June 2011. The Conference adopted a Ministerial Declaration which, inter alia, requested the Director General to prepare a draft Action Plan covering all the relevant aspects relating to nuclear safety, emergency preparedness and response, and radiation protection of people and the environment, as well as the relevant international legal framework. On 22 September 2011, the IAEA General Conference unanimously endorsed the draft IAEA Action Plan on Nuclear Safety approved by the Board of Governors. The Action Plan sets out a comprehensive programme of work, in 12 major areas, to strengthen nuclear safety worldwide. Under one of these areas, headed 'Enhance transparency and effectiveness of communication and improve dissemination of information', the IAEA Secretariat was requested to organize an International Experts' Meeting (IEM) on decommissioning, cleanup and remediation of nuclear facilities and contaminated lands after a nuclear accident. This IEM was organized in response to that request. The IEM focussed on the complex technical, societal, environmental and economic issues that need to be considered for decommissioning and remediation activities after a nuclear accident, specifically after the emergency exposure situation of an accident has been declared ended. The objective of the IEM is to assist Member States to prepare for and to be able to manage the consequences resulting from a nuclear accident. The meeting highlighted the specific short term and long term issues that may need to be addressed during decommissioning of facilities and remediation of the off-site environment affected by a nuclear accident. It is of interest to a wide range of experts, such as decision makers, regulators, operators

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

Accident Sequence Evaluation Program: Human reliability analysis procedure

International Nuclear Information System (INIS)

Swain, A.D.

1987-02-01

This document presents a shortened version of the procedure, models, and data for human reliability analysis (HRA) which are presented in the Handbook of Human Reliability Analysis With emphasis on Nuclear Power Plant Applications (NUREG/CR-1278, August 1983). This shortened version was prepared and tried out as part of the Accident Sequence Evaluation Program (ASEP) funded by the US Nuclear Regulatory Commission and managed by Sandia National Laboratories. The intent of this new HRA procedure, called the ''ASEP HRA Procedure,'' is to enable systems analysts, with minimal support from experts in human reliability analysis, to make estimates of human error probabilities and other human performance characteristics which are sufficiently accurate for many probabilistic risk assessments. The ASEP HRA Procedure consists of a Pre-Accident Screening HRA, a Pre-Accident Nominal HRA, a Post-Accident Screening HRA, and a Post-Accident Nominal HRA. The procedure in this document includes changes made after tryout and evaluation of the procedure in four nuclear power plants by four different systems analysts and related personnel, including human reliability specialists. The changes consist of some additional explanatory material (including examples), and more detailed definitions of some of the terms. 42 refs

Post-accident core coolability of light water reactors

International Nuclear Information System (INIS)

Michio, I.; Teruo, I.; Tomio, Y.; Tsutao, H.

1983-01-01

A study on post-accident core coolability of LWR is discussed based on the practical fuel failure behavior experienced in NSRR, PBF, PNS and others. The fuel failure behavior at LOCA, RIA and PCM conditions are reviewed, and seven types of fuel failure modes are extracted as the basic failure mechanism at accident conditions. These are: cladding melt or brittle failure, molten UO 2 failure, high temperature cladding burst, low temperature cladding burst, failure due to swelling of molten UO 2 , failure due to cracks of embrittled cladding for irradiated fuel rods, and TMI-2 core failure. The post-accident core coolability at each failure mode is discussed. The fuel failures caused actual flow blockage problems. A characteristic which is common among these types is that the fuel rods are in the conditions violating the present safety criteria for accidents, and UO 2 pellets are in melting or near melting hot conditions when the fuel rods failed

Science Communication in the Post-Expert Digital Age

Science.gov (United States)

Luers, A.

2014-12-01

The digital age has given rise to a post-expert world, which is poses challenges for science communication. Mass communication is shifting from a "broadcast" to "conversation" model, while audiences increasingly are finding information with search tools that create personalized filters showing only news they want to see. Such changes dilute expert voices and strengthen insular "tribal" discourse. We argue that these changes in communication pose particular challenges for science communication around politically charged issues such as climate change, because they create mini-echo chambers that can feed cultural wars. To overcome these challenges the scientific community must rethink how we engage the public. In the post-expert world, we need to shift our mindset from reporting the facts to joining diverse conversations.

Probabilistic accident consequence uncertainty analysis: Food chain uncertainty assessment. Volume 1: Main report

Energy Technology Data Exchange (ETDEWEB)

Brown, J. [National Radiological Protection Board (United Kingdom); Goossens, L.H.J.; Kraan, B.C.P. [Delft Univ. of Technology (Netherlands)] [and others

1997-06-01

This volume is the first of a two-volume document that summarizes a joint project conducted by the US Nuclear Regulatory Commission and the European Commission to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. These codes were developed primarily for estimating the risks presented by nuclear reactors based on postulated frequencies and magnitudes of potential accidents. This document reports on an ongoing project to assess uncertainty in the MACCS and COSYMA calculations for the offsite consequences of radionuclide releases by hypothetical nuclear power plant accidents. A panel of sixteen experts was formed to compile credible and traceable uncertainty distributions for food chain variables that affect calculations of offsite consequences. The expert judgment elicitation procedure and its outcomes are described in these volumes. Other panels were formed to consider uncertainty in other aspects of the codes. Their results are described in companion reports. Volume 1 contains background information and a complete description of the joint consequence uncertainty study. Volume 2 contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures for both panels, (3) the rationales and results for the panels on soil and plant transfer and animal transfer, (4) short biographies of the experts, and (5) the aggregated results of their responses.

Probabilistic accident consequence uncertainty analysis: Food chain uncertainty assessment. Volume 1: Main report

International Nuclear Information System (INIS)

Brown, J.; Goossens, L.H.J.; Kraan, B.C.P.

1997-06-01

This volume is the first of a two-volume document that summarizes a joint project conducted by the US Nuclear Regulatory Commission and the European Commission to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. These codes were developed primarily for estimating the risks presented by nuclear reactors based on postulated frequencies and magnitudes of potential accidents. This document reports on an ongoing project to assess uncertainty in the MACCS and COSYMA calculations for the offsite consequences of radionuclide releases by hypothetical nuclear power plant accidents. A panel of sixteen experts was formed to compile credible and traceable uncertainty distributions for food chain variables that affect calculations of offsite consequences. The expert judgment elicitation procedure and its outcomes are described in these volumes. Other panels were formed to consider uncertainty in other aspects of the codes. Their results are described in companion reports. Volume 1 contains background information and a complete description of the joint consequence uncertainty study. Volume 2 contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures for both panels, (3) the rationales and results for the panels on soil and plant transfer and animal transfer, (4) short biographies of the experts, and (5) the aggregated results of their responses

Accidents and undetermined deaths: re-evaluation of nationwide samples from the Scandinavian countries.

Science.gov (United States)

Tøllefsen, Ingvild Maria; Thiblin, Ingemar; Helweg-Larsen, Karin; Hem, Erlend; Kastrup, Marianne; Nyberg, Ullakarin; Rogde, Sidsel; Zahl, Per-Henrik; Østevold, Gunvor; Ekeberg, Øivind

2016-05-27

National mortality statistics should be comparable between countries that use the World Health Organization's International Classification of Diseases. Distinguishing between manners of death, especially suicides and accidents, is a challenge. Knowledge about accidents is important in prevention of both accidents and suicides. The aim of the present study was to assess the reliability of classifying deaths as accidents and undetermined manner of deaths in the three Scandinavian countries and to compare cross-national differences. The cause of death registers in Norway, Sweden and Denmark provided data from 2008 for samples of 600 deaths from each country, of which 200 were registered as suicides, 200 as accidents or undetermined manner of deaths and 200 as natural deaths. The information given to the eight experts was identical to the information used by the Cause of Death Register. This included death certificates, and if available external post-mortem examinations, forensic autopsy reports and police reports. In total, 69 % (Sweden and Norway) and 78 % (Denmark) of deaths registered in the official mortality statistics as accidents were confirmed by the experts. In the majority of the cases where disagreement was seen, the experts reclassified accidents to undetermined manner of death, in 26, 25 and 19 % of cases, respectively. Few cases were reclassified as suicides or natural deaths. Among the extracted accidents, the experts agreed least with the official mortality statistics concerning drowning and poisoning accidents. They also reported most uncertainty in these categories of accidents. In a second re-evaluation, where more information was made available, the Norwegian psychiatrist and forensic pathologist increased their agreement with the official mortality statistics from 76 to 87 %, and from 85 to 88 %, respectively, regarding the Norwegian and Swedish datasets. Among the extracted undetermined deaths in the Swedish dataset, the two experts

77 FR 10666 - Pipeline Safety: Post Accident Drug and Alcohol Testing

Science.gov (United States)

2012-02-23

... 199 [Docket No. PHMSA-2011-0335] Pipeline Safety: Post Accident Drug and Alcohol Testing AGENCY... operators of Liquefied Natural Gas (LNG) facilities to conduct post- accident drug and alcohol tests of..., operators must drug and alcohol test each covered employee whose performance either contributed to the...

The Fukushima Daiichi Accident. Technical Volume 4/5. Radiological Consequences. Annexes

International Nuclear Information System (INIS)

2015-08-01

The Fukushima Daiichi Accident consists of a Report by the IAEA Director General and five technical volumes. It is the result of an extensive international collaborative effort involving five working groups with about 180 experts from 42 Member States with and without nuclear power programmes and several international bodies. It provides a description of the accident and its causes, evolution and consequences, based on the evaluation of data and information from a large number of sources available at the time of writing. The Fukushima Daiichi Accident will be of use to national authorities, international organizations, nuclear regulatory bodies, nuclear power plant operating organizations, designers of nuclear facilities and other experts in matters relating to nuclear power, as well as the wider public. The set contains six printed parts and five supplementary CD-ROMs. Contents: Report by the Director General; Technical Volume 1/5, Description and Context of the Accident; Technical Volume 2/5, Safety Assessment; Technical Volume 3/5, Emergency Preparedness and Response; Technical Volume 4/5, Radiological Consequences; Technical Volume 5/5, Post-accident Recovery; Annexes. The Report by the Director General is available separately in Arabic, Chinese, English, French, Russian, Spanish and Japanese

Post-modern career assessment for traditionally disadvantaged ...

African Journals Online (AJOL)

Post-modern career assessment for traditionally disadvantaged South African learners: Moving away from the 'expert opinion' ... Perspectives in Education ... This article explores the perceptions of learners from a disadvantaged community regarding the limitations and advantages of traditional and post-modern career ...

Accident Sequence Evaluation Program: Human reliability analysis procedure

Energy Technology Data Exchange (ETDEWEB)

Swain, A.D.

1987-02-01

This document presents a shortened version of the procedure, models, and data for human reliability analysis (HRA) which are presented in the Handbook of Human Reliability Analysis With emphasis on Nuclear Power Plant Applications (NUREG/CR-1278, August 1983). This shortened version was prepared and tried out as part of the Accident Sequence Evaluation Program (ASEP) funded by the US Nuclear Regulatory Commission and managed by Sandia National Laboratories. The intent of this new HRA procedure, called the ''ASEP HRA Procedure,'' is to enable systems analysts, with minimal support from experts in human reliability analysis, to make estimates of human error probabilities and other human performance characteristics which are sufficiently accurate for many probabilistic risk assessments. The ASEP HRA Procedure consists of a Pre-Accident Screening HRA, a Pre-Accident Nominal HRA, a Post-Accident Screening HRA, and a Post-Accident Nominal HRA. The procedure in this document includes changes made after tryout and evaluation of the procedure in four nuclear power plants by four different systems analysts and related personnel, including human reliability specialists. The changes consist of some additional explanatory material (including examples), and more detailed definitions of some of the terms. 42 refs.

A quick look at the post accident review meeting (PARM)

International Nuclear Information System (INIS)

Dastur, A.; Osborne, R.; Pendergast, D.; Primeau, D.B.; Snell, V.G.; Torgerson, D.

1986-01-01

The IAEA sponsored Post-Accident Review Meeting (PARM) for Chernobyl took place in Vienna from August 25 to 29, 1986. A report by the International Nuclear Safety Advisory Group (INSAG) and the group of experts supporting them is under preparation and will be available later in September. In view of the importance of the meeting, this report has been prepared by the AECL Conference delegation to provide more rapid feedback. Since digestion of the information will take much more time, we have attempted to stick to what we heard, avoiding speculation unless absolutely necessary. Since the Chernobyl event was so serious, it is essential that the root causes be identified, in order to avoid a related occurrence. It is necessary that the design and operation of CANDU reactors be carefully reviewed to identify, if any, of the lessons have any relevance. This must be done thoroughly and without complacency

Probabilistic safety assessment based expert systems in support of dynamic risk assessment

International Nuclear Information System (INIS)

Varde, P.V.; Sharma, U.L.; Marik, S.K.; Raina, V.K.; Tikku, A.C.

2006-01-01

Probabilistic Safety Assessment (PSA) studies are being performed, world over as part of integrated risk assessment for Nuclear Power Plants and in many cases PSA insight is utilized in support of decision making. Though the modern plants are built with inherent safety provisions, particularly to reduce the supervisory requirements during initial period into the accident, it is always desired to develop an efficient user friendly real-time operator advisory system for handling of plant transients/emergencies which would be of immense benefit for the enhancement of operational safety of the plant. This paper discusses an integrated approach for the development of operator support system. In this approach, PSA methodology and the insight obtained from PSA has been utilized for development of knowledge based or rule based experts system. While Artificial Neural Network (ANN) approach has been employed for transient identification, rule-base expert system shell environment was used for the development of diagnostic module in this system. Attempt has been made to demonstrate that this approach offers an efficient framework for addressing requirements related to handling of real-time/dynamic scenario. (author)

Progress summary of the Chernobyl accident

International Nuclear Information System (INIS)

Iddekinge, F.W. van

1986-01-01

Based on two IAEA documents (the report of the USSR State Committee on the Utilization of Atomic Energy named 'The accident at the Chernobyl nuclear power plant and its consequences' prepared for the IAEA Experts Meeting held in Vienna on 25-29 August, 1986 and the INSAG (International Nuclear Safety Advisory Group) summary report on the Post-accident review meeting on the Chernobyl accident, drawn up in Vienna from August 30 until September 5, 1986, this publication tries to present a logic relation between the special features of the RMBK-1000 LWGR, the cause of the accident, and the technical countermeasures. (Auth.)

Use of questionnaires and an expert panel to judge the environmental consequences of chemical spills for the development of an environment-accident index.

Science.gov (United States)

Andersson, Asa Scott; Stjernström, Olof; Fängmark, Ingrid

2005-05-01

Assessing the environmental consequences of a chemical accident is a complex task. To date, the methods used to evaluate the environmental effects of an acute release of a chemical have often been based on measurements of chemical and physical variables deemed to be important, such as the concentration of the chemical. However, a broader strategy is needed to predict the environmental consequences of potential accidents during the planning process. An Environment-Accident Index (EAI), a simple tool based on such a strategy, has been developed to facilitate the consideration of a multitude of influential variables. The objectives of this study were to evaluate whether questionnaire-based expert panel's judgements could provide useful data on the environmental consequences of chemical spills, and an effective basis for further development of the EAI. As expected, the judgements did not agree perfectly, but they do give rough indications of the environmental effects, and highlight consistent trends that should be useful inputs for planning, prevention and decontamination processes. The different accidents were also judged to have caused everything from minor to very major effects in the environment, implying that a wide range of accident scenarios were represented in the material and covered by the EAI. Therefore, questionnaires and expert panel judgements can be used to collect useful data for estimating the likely environmental consequences of chemical accidents and for further development of the EAI.

The work programme of NERIS in post-accident recovery.

Science.gov (United States)

Schneider, T; Andronopoulos, S; Camps, J; Duranova, T; Gallego, E; Gering, F; Isnard, O; Maître, M; Murith, C; Oughton, D; Raskob, W

2018-01-01

NERIS is the European platform on preparedness for nuclear and radiological emergency response and recovery. Created in 2010 with 57 organisations from 28 different countries, the objectives of the platform are to: improve the effectiveness and coherency of current approaches to preparedness; identify further development needs; improve 'know how' and technical expertise; and establish a forum for dialogue and methodological development. The NERIS Strategic Research Agenda is now structured with three main challenges: (i) radiological impact assessments during all phases of nuclear and radiological events; (ii) countermeasures and countermeasure strategies in emergency and recovery, decision support, and disaster informatics; and (iii) setting up a multi-faceted framework for preparedness for emergency response and recovery. The Fukushima accident has highlighted some key issues for further consideration in NERIS research activities, including: the importance of transparency of decision-making processes at local, regional, and national levels; the key role of access to environmental monitoring; the importance of dealing with uncertainties in assessment and management of the different phases of the accident; the use of modern social media in the exchange of information; the role of stakeholder involvement processes in both emergency and recovery situations; considerations of societal, ethical, and economic aspects; and the reinforcement of education and training for various actors. This paper emphasises the main issues at stake for NERIS for post-accident management.

Industrial disasters - the expert systems solution

International Nuclear Information System (INIS)

Sachs, P.

1986-01-01

Six mistakes by the operators led to the accident at the Cherobyl nuclear reactor. These have been studied. It is suggested that an expert systems approach could prevent similar accidents. The expert system is a new approach to software programming where programs are required to perform intelligent analyses of complex situations. It separates the knowledge of a problem from the procedural code that performs the decision. An expert system will evaluate data and indicate a priority on alarms in real time. Now software systems can detect the cause of a problem in a process plant and present their findings to the operators in the control room. This should enable operators to make the correct decisions as they will know which underlying process faults are causing the alarms to operate. The Chernobyl post-mortem meeting made 13 proposals for improving safety. Two in particular are noted as relevant to expert advice systems; international collaboration on man-reactor relationships and a conference to explore the balance of automation and human action to minimise operating errors. (U.K.)

Populations protection and territories management in nuclear emergency and post-accident situation; Protection des populations et gestion des territoires en situation d'urgence nucleaire et post-accidentelle

Energy Technology Data Exchange (ETDEWEB)

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.

2010-07-01

. Champion); 17 - aid guidebooks for the management of post-accident consequences: agricultural activities and exit of the emergency stage (B. Cessac); 18 - presentation of the operational means of the GIE Groupe Intra (A. Cochard); 19 - synthetical presentation of the studies of post-accident management conditions in the milk industry (G. Machenaud); 20 - reconstruction of doses received by exposed populations (J.R. Jourdain); 21 - organisation of the health monitoring of populations after a radiological accident - presentation taken from the recommendations of the Codirpa's working group 'health stakes' (P. Pirard); 22 - losses indemnification in case of nuclear accident (M. Leger); 23 - post nuclear accident communication - reflections of the Codirpa's working group 'communication' (E. Bouchot); 24 - the ANCCLI's permanent group 'territories - post-accident' (M. Demet); 25 - experience feedback from the local information commission (CLI) 'Marcoule-Gard' in the management of a post-accident situation (J.P. Charre); 26 - commitment of intervening parties in the preparation to accident and post-accident situations: European experiences (T. Schneider); 27 - socio-cognitive exposure and risk assessment - the radio waves case (M. Poumadere). (J.S.)

Post-accident heat removal research: A state of the art review

International Nuclear Information System (INIS)

Mueller, U.; Schulenberg, T.

1983-11-01

For a realistic assessment of the consequence of extremely unlikely reactor accidents resulting in core degradation or core meltdown key questions are how to remove the decay heat from the reactor system and how to retain the radioactive core debris within the containment. Usually, this complex of questions is referred to as Post-Accident Heat Removal (PAHR). In this article the research work on PAHR performed by various institutions during the last decade has been reviewed. The main results have been summarized under the chapter headings ''Accident Scenarios,'' - ''Core Debris Accommodation Concepts,'' and ''PAHR Topics.'' Particular emphasis has been placed on the presentation of the following problems: characteristics and coolability of solid core debris in the vector vessel, heat removal from molten pools of core material, and core-melt interaction with structural materials. Some unresolved or insufficiently answered questions relating to special ''PAHR Topics'' have been mentioned or discussed at the end of the particular Chapter. Problem areas of major uncertainty have been identified and listed at the end of the review article. They include the following subjects: formation of debris beds and bed characteristics, post dryout behaviour of particle beds, long-term availability and proper location of heat sinks, creep rupture of structures under high thermal loads. (orig.) [de

Direction Committee for the management of the post-accident phase of a nuclear accident or of a radiological event (CODIRPA). Work group 'Hypotheses'. Contextual data and hypotheses to perform predictive assessments of radiological and dose consequences at the beginning of a post-accidental transition phase. 2007-2009 work report

International Nuclear Information System (INIS)

2010-01-01

This report first describes how to examine the various exposure ways of a person present on a contaminated territory and formulates hypotheses for the calculation of radioactive doses received by ingestion of contaminated food products, by external irradiation, or by involuntary inhalation of radioactive particles. It identifies factors which may influence the contamination of food products, and gives recommendations for the predictive calculation of their contamination during the first month following the accident. It indicates available methods for the predictive assessment of radioactive deposits at the beginning of the transition phase. It proposes an expertise method to assess the post-accident consequences

Risk assessment model for nuclear accident emergency protection countermeasure based on fuzzy matter-element analysis

International Nuclear Information System (INIS)

Xin Jing; Tang Huaqing; Zhang Yinghua; Zhang Limin

2009-01-01

A risk assessment model of nuclear accident emergency protection countermeasure based on fuzzy matter-element analysis and Euclid approach degree is proposed in the paper. The weight of assessed index is determined by information entropy and the scoring by experts, which could not only make full use of the inherent information of the indexes adequately, but reduce subjective assumption in the course of assessment effectively. The applied result shows that it is reasonable that the model is adopted to make risk assessment for nuclear accident emergency protective countermeasure,and it could be a kind of effective analytical method and decision making basis to choose the optimum protection countermeasure. (authors)

Probabilistic accident consequence uncertainty analysis: Food chain uncertainty assessment. Volume 2: Appendices

Energy Technology Data Exchange (ETDEWEB)

Brown, J. [National Radiological Protection Board (United Kingdom); Goossens, L.H.J.; Kraan, B.C.P. [Delft Univ. of Technology (Netherlands)] [and others

1997-06-01

This volume is the second of a two-volume document that summarizes a joint project by the US Nuclear Regulatory and the Commission of European Communities to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. These codes were developed primarily for estimating the risks presented by nuclear reactors based on postulated frequencies and magnitudes of potential accidents. This two-volume report, which examines mechanisms and uncertainties of transfer through the food chain, is the first in a series of five such reports. A panel of sixteen experts was formed to compile credible and traceable uncertainty distributions for food chain transfer that affect calculations of offsite radiological consequences. Seven of the experts reported on transfer into the food chain through soil and plants, nine reported on transfer via food products from animals, and two reported on both. The expert judgment elicitation procedure and its outcomes are described in these volumes. This volume contains seven appendices. Appendix A presents a brief discussion of the MAACS and COSYMA model codes. Appendix B is the structure document and elicitation questionnaire for the expert panel on soils and plants. Appendix C presents the rationales and responses of each of the members of the soils and plants expert panel. Appendix D is the structure document and elicitation questionnaire for the expert panel on animal transfer. The rationales and responses of each of the experts on animal transfer are given in Appendix E. Brief biographies of the food chain expert panel members are provided in Appendix F. Aggregated results of expert responses are presented in graph format in Appendix G.

Probabilistic accident consequence uncertainty analysis: Food chain uncertainty assessment. Volume 2: Appendices

International Nuclear Information System (INIS)

Brown, J.; Goossens, L.H.J.; Kraan, B.C.P.

1997-06-01

This volume is the second of a two-volume document that summarizes a joint project by the US Nuclear Regulatory and the Commission of European Communities to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. These codes were developed primarily for estimating the risks presented by nuclear reactors based on postulated frequencies and magnitudes of potential accidents. This two-volume report, which examines mechanisms and uncertainties of transfer through the food chain, is the first in a series of five such reports. A panel of sixteen experts was formed to compile credible and traceable uncertainty distributions for food chain transfer that affect calculations of offsite radiological consequences. Seven of the experts reported on transfer into the food chain through soil and plants, nine reported on transfer via food products from animals, and two reported on both. The expert judgment elicitation procedure and its outcomes are described in these volumes. This volume contains seven appendices. Appendix A presents a brief discussion of the MAACS and COSYMA model codes. Appendix B is the structure document and elicitation questionnaire for the expert panel on soils and plants. Appendix C presents the rationales and responses of each of the members of the soils and plants expert panel. Appendix D is the structure document and elicitation questionnaire for the expert panel on animal transfer. The rationales and responses of each of the experts on animal transfer are given in Appendix E. Brief biographies of the food chain expert panel members are provided in Appendix F. Aggregated results of expert responses are presented in graph format in Appendix G

The Fukushima Daiichi Accident. Technical Volume 3/5. Emergency Preparedness and Response. Annexes

International Nuclear Information System (INIS)

2015-08-01

The Fukushima Daiichi Accident consists of a Report by the IAEA Director General and five technical volumes. It is the result of an extensive international collaborative effort involving five working groups with about 180 experts from 42 Member States with and without nuclear power programmes and several international bodies. It provides a description of the accident and its causes, evolution and consequences, based on the evaluation of data and information from a large number of sources available at the time of writing. The Fukushima Daiichi Accident will be of use to national authorities, international organizations, nuclear regulatory bodies, nuclear power plant operating organizations, designers of nuclear facilities and other experts in matters relating to nuclear power, as well as the wider public. The set contains six printed parts and five supplementary CD-ROMs. Contents: Report by the Director General; Technical Volume 1/5, Description and Context of the Accident; Technical Volume 2/5, Safety Assessment; Technical Volume 3/5, Emergency Preparedness and Response; Technical Volume 4/5, Radiological Consequences; Technical Volume 5/5, Post-accident Recovery; Annexes. The Report by the Director General is available separately in Arabic, Chinese, English, French, Russian, Spanish and Japanese

Proceedings of the Start-up Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs, 28-29 April 2014, OECD-NEA HQ

International Nuclear Information System (INIS)

Kurata, Masaki; Bragg-Sitton, Shannon; Pasamehmetoglu, K.; Sowder, Andrew; Koo, Yang-Hyun; Yang, Jae-Ho; Kim, Hyun-Gil; Zhou, Y.; Forgeron, T.; Guedeney, Ph.; Brachet, J.C.; Michaux, A.; Chauvin, Nathalie; Waeckel, N.; Ambard, A.; Blanpain, P.; Bischoff, J.; Zvonarev, Yu.; Verwerft, M.; Weber, M.; Lambrinou, K.; Koonen, E.; Van Dyck, S.; PETIT, Marc; Cornet, Stephanie; ); YAMAJI, Akifumi; ); Inozemtsev, V.; )

2014-04-01

-based alloys for LWR fuel cladding (K.A. Terrani et al.); - Post-quench ductility evaluation of Zircaloy-4 and select iron alloys under design basis and extended LOCA conditions (Y. Yan et al.); - High temperature oxidation of molybdenum in water vapor environments (A.T. Nelson et al.); - Continuous SiC fiber, CVI SiC matrix composites for nuclear applications: Properties and irradiation effects (Yutai Katoh et al.); - Thermo-mechanical analysis of LWR SiC/SiC composite cladding (M. Ben-Belgacem et al.); - Effects of neutron irradiation on mechanical properties of silicon carbide composites fabricated by nano-infiltration and transient eutectic-phase process (Takaaki Koyanagi et al.); - Irradiation creep of 3C-SiC and microstructural understanding of the underlying mechanisms (Sosuke Kondo et al.); - Radiation-tolerant joining technologies for silicon carbide ceramics and composites (Yutai Katoh et al.); - The effect of fuel thermal conductivity on the behavior of LWR cores during loss-of-coolant accidents (Kurt A. Terrani et al.); - Preliminary assessment of accident-tolerant fuels on LWR performance during normal operation and under DB and BDB accident conditions (L.J. Ott et al.); - Scoping assessments of ATF impact on late-stage accident progression including molten core-concrete interaction (M.T. Farmer et al.); - Thermo-mechanical analysis of LWR SiC/SiC composite cladding (M. Ben-Belgacem et al.); - High temperature oxidation of fuel cladding candidate materials in steam-hydrogen environments (B.A. Pint et al.)

Accidents and emergency conditions: Tasks of the radiation protection expert

International Nuclear Information System (INIS)

Hacke, J.

1985-01-01

This paper reviews and explains the tasks of the radiation protection expert at a given site in the event of accidents or emergencies involving a radiation hazard to the personnel. The various measures recommended discriminate between the main two types of hazards, namely external radiation or internal radiation. The paper discusses the first-aid and emergency measures recommended in various publications (BG, 1982; ICRP, 1980; MO, 1972; ME, 1980) and also cites recommendations contained therein, referring to preventive means and measures and to communications to the press and the general public. (DG) [de

NEA international peer reviews of post-accident protection policy

International Nuclear Information System (INIS)

Lazo, T.

2011-01-01

For many years, the NEA has offered international peer reviews of national, high-level radioactive waste management policies and approaches. Until recently, this service had not been requested in the area of radiological protection. However, the 3. International Nuclear Emergency Exercise (INEX-3, 2005-2006) addressed post-accident consequence management for the first time in a broad, international sense, and helped generate significant national reflections in this area. In particular, in 2005 the French government began an extensive programme of post-emergency consequence management planning, resulting in a draft national policy to address such situations. The Finnish government used the INEX-3 exercise as a vehicle to discuss post-emergency consequence management with a broad group of governmental and private stakeholders, and also began to develop national policy in this area. In order to further refine national efforts, the French Nuclear Safety Authority (ASN) invited the NEA to perform in April 2011 its first international peer review in the radiological protection area focusing on its post-emergency consequence management policy under development. Finnish experts participated in this peer review team, and as a result, subsequently invited the NEA to perform an international peer review of its developing policy in this area in September 2011. These draft national policies and their international peer reviews are briefly presented in this paper. Feedback from both the French ASN and the Finnish STUK suggests that the detailed, external input provided by the international peer review teams have been extremely valuable in refining the content of the guides so that they are more clear, concise, understandable and implementable. It should be recalled that both national policy documents reviewed are far more detailed and extensive than described here. The intent of this article was not to provide a review of the national policies themselves, but rather to give an

The post-accidents exploitation of Chernobyl NPP

International Nuclear Information System (INIS)

Nosovskij, A.V.

1996-01-01

Problems facing the Chernobyl NPP personnel since its commissioning and during its post-accident operation are considered. The accomplished measures on improving the power units safety, normalization of the radiation situation as well as the impact of psychological factor at the safety culture level are discussed

A review of severe accident assessment

International Nuclear Information System (INIS)

Kawashima, Kei

2000-01-01

One of the most difficult problems on evaluation of external costs on nuclear power generation is value on a severe accident risk. Once forming a severe accident, its effect is very important and extends to a wide range, to give a lot of damages. It is a main area of study on externality of energy to compare various risks by means of price conversion at unit kWh. Here was outlined on research examples on main severe accident risks before then. A common fact on estimation cost such research examples is to limit it to direct cost (mainly to health damage) at accident phenomenon. As an actual problem, it is very difficult to substantially quantify such parameters because of basically belonging to social psychology. It is due to no finding out decisive evaluation method on this problem to be adopted conventional EED (Expert Expected Damages) approach in the ExternE Phase III, either. (G.K.)

The Fukushima Daiichi Accident. Report by the Director General [Chinese Version

International Nuclear Information System (INIS)

2015-08-01

This report presents an assessment of the causes and consequences of the accident at the Fukushima Daiichi nuclear power plant in Japan, which began on 11 March 2011. Caused by a huge tsunami that followed a massive earthquake, it was the worst accident at a nuclear power plant since the Chernobyl disaster in 1986. The report considers human, organizational and technical factors, and aims to provide an understanding of what happened, and why, so that the necessary lessons learned can be acted upon by governments, regulators and nuclear power plant operators throughout the world. Measures taken in response to the accident, both in Japan and internationally, are also examined. The Fukushima Daiichi Accident consists of a Report by the IAEA Director General and five technical volumes. It is the result of an extensive international collaborative effort involving five working groups with about 180 experts from 42 Member States with and without nuclear power programmes and several international bodies. It provides a description of the accident and its causes, evolution and consequences, based on the evaluation of data and information from a large number of sources available at the time of writing. The Fukushima Daiichi Accident will be of use to national authorities, international organizations, nuclear regulatory bodies, nuclear power plant operating organizations, designers of nuclear facilities and other experts in matters relating to nuclear power, as well as the wider public. The set contains six printed parts and five supplementary CD-ROMs. Contents: Report by the Director General; Technical Volume 1/5, Description and Context of the Accident; Technical Volume 2/5, Safety Assessment; Technical Volume 3/5, Emergency Preparedness and Response; Technical Volume 4/5, Radiological Consequences; Technical Volume 5/5, Post-accident Recovery; Annexes. The Report by the Director General is also available separately in Arabic, Chinese, English, French, Russian, Spanish and

The Fukushima Daiichi Accident. Report by the Director General [Japanese Version

International Nuclear Information System (INIS)

2015-08-01

This report presents an assessment of the causes and consequences of the accident at the Fukushima Daiichi nuclear power plant in Japan, which began on 11 March 2011. Caused by a huge tsunami that followed a massive earthquake, it was the worst accident at a nuclear power plant since the Chernobyl disaster in 1986. The report considers human, organizational and technical factors, and aims to provide an understanding of what happened, and why, so that the necessary lessons learned can be acted upon by governments, regulators and nuclear power plant operators throughout the world. Measures taken in response to the accident, both in Japan and internationally, are also examined. The Fukushima Daiichi Accident consists of a Report by the IAEA Director General and five technical volumes. It is the result of an extensive international collaborative effort involving five working groups with about 180 experts from 42 Member States with and without nuclear power programmes and several international bodies. It provides a description of the accident and its causes, evolution and consequences, based on the evaluation of data and information from a large number of sources available at the time of writing. The Fukushima Daiichi Accident will be of use to national authorities, international organizations, nuclear regulatory bodies, nuclear power plant operating organizations, designers of nuclear facilities and other experts in matters relating to nuclear power, as well as the wider public. The set contains six printed parts and five supplementary CD-ROMs. Contents: Report by the Director General; Technical Volume 1/5, Description and Context of the Accident; Technical Volume 2/5, Safety Assessment; Technical Volume 3/5, Emergency Preparedness and Response; Technical Volume 4/5, Radiological Consequences; Technical Volume 5/5, Post-accident Recovery; Annexes. The Report by the Director General is also available separately in Arabic, Chinese, English, French, Russian, Spanish and

The Fukushima Daiichi Accident. Report by the Director General [Spanish Version

International Nuclear Information System (INIS)

2015-08-01

This report presents an assessment of the causes and consequences of the accident at the Fukushima Daiichi nuclear power plant in Japan, which began on 11 March 2011. Caused by a huge tsunami that followed a massive earthquake, it was the worst accident at a nuclear power plant since the Chernobyl disaster in 1986. The report considers human, organizational and technical factors, and aims to provide an understanding of what happened, and why, so that the necessary lessons learned can be acted upon by governments, regulators and nuclear power plant operators throughout the world. Measures taken in response to the accident, both in Japan and internationally, are also examined. The Fukushima Daiichi Accident consists of a Report by the IAEA Director General and five technical volumes. It is the result of an extensive international collaborative effort involving five working groups with about 180 experts from 42 Member States with and without nuclear power programmes and several international bodies. It provides a description of the accident and its causes, evolution and consequences, based on the evaluation of data and information from a large number of sources available at the time of writing. The Fukushima Daiichi Accident will be of use to national authorities, international organizations, nuclear regulatory bodies, nuclear power plant operating organizations, designers of nuclear facilities and other experts in matters relating to nuclear power, as well as the wider public. The set contains six printed parts and five supplementary CD-ROMs. Contents: Report by the Director General; Technical Volume 1/5, Description and Context of the Accident; Technical Volume 2/5, Safety Assessment; Technical Volume 3/5, Emergency Preparedness and Response; Technical Volume 4/5, Radiological Consequences; Technical Volume 5/5, Post-accident Recovery; Annexes. The Report by the Director General is also available separately in Arabic, Chinese, English, French, Russian, Spanish and

The Fukushima Daiichi Accident. Report by the Director General [Russian Version

International Nuclear Information System (INIS)

2015-08-01

This report presents an assessment of the causes and consequences of the accident at the Fukushima Daiichi nuclear power plant in Japan, which began on 11 March 2011. Caused by a huge tsunami that followed a massive earthquake, it was the worst accident at a nuclear power plant since the Chernobyl disaster in 1986. The report considers human, organizational and technical factors, and aims to provide an understanding of what happened, and why, so that the necessary lessons learned can be acted upon by governments, regulators and nuclear power plant operators throughout the world. Measures taken in response to the accident, both in Japan and internationally, are also examined. The Fukushima Daiichi Accident consists of a Report by the IAEA Director General and five technical volumes. It is the result of an extensive international collaborative effort involving five working groups with about 180 experts from 42 Member States with and without nuclear power programmes and several international bodies. It provides a description of the accident and its causes, evolution and consequences, based on the evaluation of data and information from a large number of sources available at the time of writing. The Fukushima Daiichi Accident will be of use to national authorities, international organizations, nuclear regulatory bodies, nuclear power plant operating organizations, designers of nuclear facilities and other experts in matters relating to nuclear power, as well as the wider public. The set contains six printed parts and five supplementary CD-ROMs. Contents: Report by the Director General; Technical Volume 1/5, Description and Context of the Accident; Technical Volume 2/5, Safety Assessment; Technical Volume 3/5, Emergency Preparedness and Response; Technical Volume 4/5, Radiological Consequences; Technical Volume 5/5, Post-accident Recovery; Annexes. The Report by the Director General is also available separately in Arabic, Chinese, English, French, Russian, Spanish and

Design considerations for post accident monitoring system of a research reactor

International Nuclear Information System (INIS)

Jang, Gwi Sook; Park, Je Yun; Kim, Young Ki

2012-01-01

The Post Accident Monitoring System (PAMS) provides primary information for operators to assess the plant conditions and perform their role in bringing the plant to a safe condition during an accident. The PAMS of NPP (Nuclear Power Plant) in KOREA provides the continuous display of the PAM category 1 parameters specified in R.G 1.97, Rev. 03. Recently the PAMS of NPP has been designed according to R.G 1.97, Rev. 04. There is no PAMS at the HANARO in KOREA, but recently RRs (Research Reactors) around the world are going to have PAMS for various multi purposes. We should determine the design considerations for PAMS in a Korean RR based on the design state analysis. Thus, this paper proposes strategies on the design considerations for the PAMS of a Korean RR

Dose assessment in radiological accidents

International Nuclear Information System (INIS)

Donkor, S.

2013-04-01

The applications of ionizing radiation bring many benefits to humankind, ranging from power generation to uses in medicine, industry and agriculture. Facilities that use radiation source require special care in the design and operation of equipment to prevent radiation injury to workers or to the public. Despite considerable development of radiation safety, radiation accidents do happen. The purpose of this study is therefore to discuss how to assess doses to people who will be exposed to a range of internal and external radiation sources in the event of radiological accidents. This will go a long way to complement their medical assessment thereby helping to plan their treatment. Three radiological accidents were reviewed to learn about the causes of those accidents and the recommendations that were put in place to prevent recurrence of such accidents. Various types of dose assessment methods were discussed.(au)

Psycho-social and Mental Variables and Post-Traumatic Stress Disorder in Traffic Accident Survivors in Northern Iran.

Science.gov (United States)

Khodadadi-Hassankiadeh, Naema; Dehghan-Nayeri, Nahid; Shahsavari, Hooman; Yousefzadeh-Chabok, Shahrokh; Haghani, Hamid

2017-07-01

To assess the psycho-social and mental variables associated with post-traumatic stress disorder (PTSD) in a series of Iranian patients. A total of 528 eligible accident survivors in pre-sampling of a randomized controlled trial targeting PTSD were included in this cross-sectional study. Psycho-social characteristics associated to PTSD were explored in these survivors in an outpatient clinic. They completed the questionnaires via interview between six weeks to six months after accident. Data collection tools were PSS (DSM-V version) for PTSD and BDI-II for depression and a researcher-made questionnaire for psycho-social variables. There was a significant association between PTSD and the following variables; family communication, current depression, return to work, history of death of relatives, witnessed the death, length of amnesia, hospitalization, injured situation, and accident severity. Multivariate logistic regression indicated that some variables were associated with PTSD such as accident severity, ( p severe accident and poor family communication who do not return to work. Thus, routine assessment of PTSD, depression and psycho-social variables after traffic accidents must be taken into account.

Improvement of the assessment of the external costs of severe nuclear accidents

International Nuclear Information System (INIS)

Markandya, A.; Dale, N.; Schneider, T.

1998-12-01

The first part of this document presents a bibliographic study on the accidents costs. The second part is devoted to an empirical study realized in Spain, concerning the risk assessment by experts. The third part proposes an approach in terms of hope of utility for the aversion calculation facing the major risks. The last part presents the probabilities transformations taking into account the human perception of the risk. (A.L.B.)

Post-accident monitoring systems in Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

Suriya Murthy, N.; Sivasailanathan, Vidhya; Ananth, Allu; Roy, Kallol

2018-01-01

PFBR is a 500 MW(e) MOX fueled and sodium cooled fast reactor (SFR) under advanced stage of commissioning at Kalpakkam. Currently, the main vessel is preheated and sodium has been charged into two secondary loops that are operated in recirculation mode. In order to characterize the radiation field and contamination, the workplace monitoring is undertaken using installed monitors that are commissioned and made operational. This helps to ensure radiological protection during normal operating conditions. On the other hand, radiological monitoring in emergency conditions is quite different. For undertaking the mitigative accident management, a set of specialized nuclear instruments called post-accident monitoring systems (PAMS) which include radiation monitors are stipulated. The Fukushima Daiichi accident emphasized the importance and need for reliable accident monitoring instrumentation to indicate the safety functions during the progression and aftermath of accident in NPP. In PFBR, the PAMS are integrated with other monitoring systems in design stage itself to manage the measurements and indicating the safety functions for implementing EOP and SAMG

Assessing economic consequences of radiation accidents

International Nuclear Information System (INIS)

Rowe, M.D.; Lee, J.C.; Grimshaw, C.A.; Kalb, P.D.

1987-01-01

A recent review of existing models and methods for assessing potential consequences of accidents in the high-level radioactive waste (HLW) disposal system identifies economic consequence assessment methods as a weak point. Existing methods have mostly been designed to assess economic consequences of reactor accidents, the possible scale of which can be several orders of magnitude greater than anything possible in the HLW disposal system. There is therefore some question about the applicability of these methods, their assumptions, and their level of detail to assessments of smaller accidents. The US Dept. of Energy funded this study to determine needs for code modifications or model development for assessing economic costs of accidents in the HLW disposal system. The objectives of the study were as follows: (1) review the literature on economic consequences of accidents to determine the availability of assessment methods and data and their applicability to the HLW disposal system before closure. (2) Determine needs for expansion, revision, or adaptation of methods and data for modeling economic consequences of accidents of the scale projected for the disposal system. (3) Gather data that might be useful for the needed revisions for modeling economic impacts on this scale

Comparative risk assessment of severe accidents in the energy sector

International Nuclear Information System (INIS)

Burgherr, Peter; Hirschberg, Stefan

2014-01-01

Comparative assessment of accident risks in the energy sector is a key aspect in a comprehensive evaluation of sustainability and energy security concerns. Safety performance of energy systems can have important implications on the environmental, economic and social dimensions of sustainability as well as availability, acceptability and accessibility aspects of energy security. Therefore, this study provides a broad comparison of energy technologies based on the objective expression of accident risks for complete energy chains. For fossil chains and hydropower the extensive historical experience available in PSI's Energy-related Severe Accident Database (ENSAD) is used, whereas for nuclear a simplified probabilistic safety assessment (PSA) is applied, and evaluations of new renewables are based on a combination of available data, modeling, and expert judgment. Generally, OECD and EU 27 countries perform better than non-OECD. Fatality rates are lowest for Western hydropower and nuclear as well as for new renewables. In contrast, maximum consequences can be by far highest for nuclear and hydro, intermediate for fossil, and very small for new renewables, which are less prone to severe accidents. Centralized, low-carbon technology options could generally contribute to achieve large reductions in CO 2 -emissions; however, the principal challenge for both fossil with Carbon Capture and Storage and nuclear is public acceptance. Although, external costs of severe accidents are significantly smaller than those caused by air pollution, accidents can have disastrous and long-term impacts. Overall, no technology performs best or worst in all respects, thus tradeoffs and priorities are needed to balance the conflicting objectives such as energy security, sustainability and risk aversion to support rationale decision making. - Highlights: • Accident risks are compared across a broad range of energy technologies. • Analysis of historical experience was based on the

10 CFR 76.85 - Assessment of accidents.

Science.gov (United States)

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Assessment of accidents. 76.85 Section 76.85 Energy... Assessment of accidents. The Corporation shall perform an analysis of potential accidents and consequences to... postulated accidents which include internal and external events and natural phenomena in order to ensure...

Environmental consequences of the Chernobyl accident and their remediation: Twenty years of experience. Report of the UN Chernobyl Forum Expert Group 'Environment' (EGE). Working material

International Nuclear Information System (INIS)

2005-08-01

remediation and special health-care programmes, and to suggest areas where further research is required; and to accept the following Terms of Reference (TOR) of the Forum. The objectives of the Chernobyl Forum were defined as follows: To explore and refine the current scientific assessments on the long-term health and environmental consequences of the Chernobyl accident, with a view to producing authoritative consensus statements focusing on: the health effects attributable to radiation exposure caused by the accident, the environmental consequences induced by the radioactive materials released due to the accident, e.g., contamination of foodstuffs, and additionally to address the consequences attributable to the accident although not directly related to the radiation exposure or radioactive contamination; To identify gaps in scientific research relevant to the radiation-induced or radioactive contamination-induced health and environmental impacts of the accident, and suggest areas where further work is required based on an assessment of the work done in the past, and bearing in mind ongoing work and projects; To provide advice on, and to facilitate implementation of scientifically sound programmes on mitigation of the accident consequences, including possible joint actions of the organizations participating in the Forum, such as: agricultural, economic and social life under safe conditions, special health care of the affected population, monitoring of the long-term human exposure to radiation, and addressing the environmental issues pertaining to the decommissioning of the Shelter and management of radioactive waste originating from the Chernobyl accident. The Chernobyl Forum itself continued as a high-level organisation of senior officials from UN agencies and the three more affected countries. The actual work has been accomplished by two expert groups: Expert Group -Environment - (EGE) and Expert Group 'Health' (EGH). Members of each of these two groups consisted of

Environmental consequences of the Chernobyl accident and their remediation: Twenty years of experience. Report of the UN Chernobyl Forum Expert Group 'Environment' (EGE). Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

2005-08-01

remediation and special health-care programmes, and to suggest areas where further research is required; and to accept the following Terms of Reference (TOR) of the Forum. The objectives of the Chernobyl Forum were defined as follows: To explore and refine the current scientific assessments on the long-term health and environmental consequences of the Chernobyl accident, with a view to producing authoritative consensus statements focusing on: the health effects attributable to radiation exposure caused by the accident, the environmental consequences induced by the radioactive materials released due to the accident, e.g., contamination of foodstuffs, and additionally to address the consequences attributable to the accident although not directly related to the radiation exposure or radioactive contamination; To identify gaps in scientific research relevant to the radiation-induced or radioactive contamination-induced health and environmental impacts of the accident, and suggest areas where further work is required based on an assessment of the work done in the past, and bearing in mind ongoing work and projects; To provide advice on, and to facilitate implementation of scientifically sound programmes on mitigation of the accident consequences, including possible joint actions of the organizations participating in the Forum, such as: agricultural, economic and social life under safe conditions, special health care of the affected population, monitoring of the long-term human exposure to radiation, and addressing the environmental issues pertaining to the decommissioning of the Shelter and management of radioactive waste originating from the Chernobyl accident. The Chernobyl Forum itself continued as a high-level organisation of senior officials from UN agencies and the three more affected countries. The actual work has been accomplished by two expert groups: Expert Group -Environment - (EGE) and Expert Group 'Health' (EGH). Members of each of these two groups consisted of

Management of a radiological emergency. Experience feedback and post-accident management; Gestion d'une urgence radiologique. Retour d'experience et gestion post-accidentelle

Energy Technology Data Exchange (ETDEWEB)

Dubiau, Ph. [Institut de Radioprotection et de Surete Nucleaire (IRSN), 92 - Clamart (France)

2007-07-15

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.)

Source term assessment, containment atmosphere control systems, and accident consequences. Report to CSNI by an OECD/NEA Group of experts

International Nuclear Information System (INIS)

1987-04-01

CSNI Report 135 summarizes the results of the work performed by CSNI's Principal Working Group No. 4 on the Source Term and Environmental Consequences (PWG4) during the period extending from 1983 to 1986. This document contains the latest information on some important topics relating to source terms, accident consequence assessment, and containment atmospheric control systems. It consists of five parts: (1) a Foreword and Executive Summary prepared by PWG4's Chairman; (2) a Report on the Technical Status of the Source Term; (3) a Report on the Technical Status of Filtration and Containment Atmosphere Control Systems for Nuclear Reactors in the Event of a Severe Accident; (4) a Report on the Technical Status of Reactor Accident Consequence Assessment; (5) a list of members of PWG4

A first assessment of the psychic and social effects of the Chernobyl accident

International Nuclear Information System (INIS)

Heriard Dubreuil, G.

1994-01-01

A synthesis has been made of a series of surveys carried out in Ukraine in 1992 and 1993 on the psychic and social consequences of the Chernobyl accident, within the framework of the ''Evaluation programme of the consequences of the Chernobyl nuclear accident'' of the Commission of the European communities. The main results demonstrate the strength of the post-accident dynamics of the accident, more than 7 years later. Some 3 millions people were directly affected in their everyday life by the post-accident management which resulted in many perverse effects on the social and psychic levels. Economically, each year, financing of the post-accident management system requires nearly 1/6 of the Ukraine budget. Politically speaking, Chernobyl is still a major stake for the various actors of the institutional transition process underway since the disappearance of the soviet system. The article shows the systemic complexity of the local situation and the many explanatory factors (physical, sanitary, political, cultural, historical) at the origin of the post-accident dynamics. A systemic modelling of the interactions between these factors is presented. It makes it possible to better define the contributions of both accident and post-accident stages to the process that has led to the present situation. It shows out the close connections between the different accident stages and the need, from the very beginning of an accident, to take into account the mid-and long-term consequences arising from the accident management. (author). 11 refs., 3 figs

Study on the code system for the off-site consequences assessment of severe nuclear accident

Energy Technology Data Exchange (ETDEWEB)

Kim, Sora; Mn, Byung Il; Park, Ki Hyun; Yang, Byung Mo; Suh, Kyung Suk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-12-15

The importance of severe nuclear accidents and probabilistic safety assessment (PSA) were brought to international attention with the occurrence of severe nuclear accidents caused by the extreme natural disaster at Fukushima Daiichi nuclear power plant in Japan. In Korea, studies on level 3 PSA had made little progress until recently. The code systems of level 3 PSA, MACCS2 (MELCORE Accident Consequence Code System 2, US), COSYMA (COde SYstem from MAria, EU) and OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents, JAPAN), were reviewed in this study, and the disadvantages and limitations of MACCS2 were also analyzed. Experts from Korea and abroad pointed out that the limitations of MACCS2 include the following: MACCS2 cannot simulate multi-unit accidents/release from spent fuel pools, and its atmospheric dispersion is based on a simple Gaussian plume model. Some of these limitations have been improved in the updated versions of MACCS2. The absence of a marine and aquatic dispersion model and the limited simulating range of food-chain and economic models are also important aspects that need to be improved. This paper is expected to be utilized as basic research material for developing a Korean code system for assessing off-site consequences of severe nuclear accidents.

Study on the code system for the off-site consequences assessment of severe nuclear accident

International Nuclear Information System (INIS)

Kim, Sora; Mn, Byung Il; Park, Ki Hyun; Yang, Byung Mo; Suh, Kyung Suk

2016-01-01

The importance of severe nuclear accidents and probabilistic safety assessment (PSA) were brought to international attention with the occurrence of severe nuclear accidents caused by the extreme natural disaster at Fukushima Daiichi nuclear power plant in Japan. In Korea, studies on level 3 PSA had made little progress until recently. The code systems of level 3 PSA, MACCS2 (MELCORE Accident Consequence Code System 2, US), COSYMA (COde SYstem from MAria, EU) and OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents, JAPAN), were reviewed in this study, and the disadvantages and limitations of MACCS2 were also analyzed. Experts from Korea and abroad pointed out that the limitations of MACCS2 include the following: MACCS2 cannot simulate multi-unit accidents/release from spent fuel pools, and its atmospheric dispersion is based on a simple Gaussian plume model. Some of these limitations have been improved in the updated versions of MACCS2. The absence of a marine and aquatic dispersion model and the limited simulating range of food-chain and economic models are also important aspects that need to be improved. This paper is expected to be utilized as basic research material for developing a Korean code system for assessing off-site consequences of severe nuclear accidents

Hierarchical Model of Assessing and Selecting Experts

OpenAIRE

Chernysheva, Tatiana Yurievna; Korchuganova, Mariya Anatolievna; Borisov, V. V.; Minkov, S. L.

2016-01-01

Revealing experts' competences is a multi-objective issue. Authors of the paper deal with competence assessing methods of experts seen as objects, and criteria of qualities. An analytic hierarchy process of assessing and ranking experts is offered, which is based on paired comparison matrices and scores, quality parameters are taken into account as well. Calculation and assessment of experts is given as an example.

Tako-tsubo cardiomyopathy and post-traumatic stress disorder after a skiing accident: a case report.

Science.gov (United States)

Harb, Birgit Maria; Wonisch, Manfred; Fruhwald, Friedrich; Fazekas, Christian

2015-03-01

Symptoms of a post-traumatic stress disorder can follow Tako-tsubo cardiomyopathy. This vignette describes such a linkage and exemplifies the risk that these symptoms may remain undetected. After a skiing accident that had evoked existential fear of suffocation, a post-menopausal woman was diagnosed with Tako-tsubo syndrome and myocardial contusion. Symptoms of post-traumatic stress disorder appeared 2 weeks after remission of the cardiomyopathy. Two months later, a psychological assessment was conducted during cardiac rehabilitation. A post-traumatic stress disorder was diagnosed and successfully treated by narrative exposure. This case report suggests that these patients should be informed during the initial hospital stay that post-traumatic stress symptoms could appear. It also suggests including a screening for post-traumatic stress disorder in the follow-up of these patients.

Preventive radioecological assessment of territory for optimization of monitoring and countermeasures after radiation accidents.

Science.gov (United States)

Prister, B S; Vinogradskaya, V D; Lev, T D; Talerko, M M; Garger, E K; Onishi, Y; Tischenko, O G

2018-04-01

A methodology of a preventive radioecological assessment of the territory has been developed for optimizing post-emergency monitoring and countermeasure implementation in an event of a severe radiation accident. Approaches and main stages of integrated radioecological zoning of the territory are described. An algorithm for the assessment of the potential radioecological criticality (sensitivity) of the area is presented. The proposed approach is validated using data of the dosimetric passportization in Ukraine after the Chernobyl accident for the test site settlements. Copyright © 2018 Elsevier Ltd. All rights reserved.

Methods for the elicitation and use of expert opinion in risk assessment: Phase 1, A critical evaluation and directions for future research

International Nuclear Information System (INIS)

Mosleh, A.; Bier, V.M.; Apostolakis, G.

1987-08-01

The purpose of this work is to critically review and evaluate the elicitation and use of expert opinion in probabilistic risk assessment (PRA) in light of the available empirical and theoretical results on expert opinion use. PRA practice is represented by five case studies selected to represent a variety of aspects of the problem: assessments of component failure rates and maintenance data; recent assessments of seismic hazard rates; assessments of containment phenomenology; assessments of human error rates; and accident precursor studies. The review has yielded mixed results. On the negative side, there appears to be little reliance on normative expertise in structuring the process of expert opinion elicitation and use; most applications instead rely primarily on the common sense of the experts involved in the analysis, which is not always an adequate guide. On the positive side, however, there is evidence that expert opinions can in fact be used well in practical settings. Suggestions are given for Phase II work to enhance the applicability and use of appropriate expert opinion methods

Proceedings of the Third Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs, 3-5 March 2015, OECD-NEA HQ

International Nuclear Information System (INIS)

Bischoff, Jeremy; Gandrille, Pascal; Forgeron, Thierry; Brachet, Jean-Christophe; Lorrette, Christophe; Valot, C.; Freyss, M.; Braun, J.; Sauder, C.; Moatti, Marie; Waeckel, Nicolas; Ambard, Antoine; Pasamehmetoglu, Kemal; Johnston, Emma; Bragg-Sitton, Shannon M.; Kurata, M.; Hallstadius, Lars; Ohta, H.; Ogata, T.; Besmann, T.; Chauvin, Nathalie; Cornet, Stephanie; Massara, S.; Kohyama, Akira; Kishimoto, Hirotatsu; Park, Joon Soo; Nakazato, Naofumi; Hayasaka, Daisuke; Asakura, Yuuki; Kanda, Chisato; Kohyama, Akira; Terrani, Kurt; Katoh, Yutai; Yamamoto, Yukinori; Field, Kevin; Snead, Lance; Hu, Xunxiang; Dryepondt, Sebastien; Unocic, Kinga A.; Hoelzer, David T.; Pint, Bruce A.; Besmann, T.; Steinbrueck, M.; Grosse, M.; Jianu, A.; Weisenburger, A.; Avincola, V.; Ahmad, S.; Tang, C.; Heuser, Brent J.; Sickafus, Kurt; Kim, Hyun-Gil; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jung-Hwan; Park, Jeong-Yong; Koo, Yang-Hyun; Lee, B.O.; Van Nieuwenhove, Rudi; Kim, Young; Rebak, Raul; Dolly, Evan; Dolley, E.J.; Rebak, R.B.; Maloy, Stu; Yang, Jae-Ho; Kim, Dong-Joo; Kim, Keon-Sik; Koo, Yang-Hyun; Lee, Won Jae; Tulenko, James S.; Puide, Mattias; Liu, T.; Gueneau, C.; Gosse, S.; Dupin, N.; Barber, D.; Corcoran, E.; Dumas, J.C.; Hania, R.; Kaye, M.; Turchi, P.

2015-03-01

Under the guidance of the OECD-NEA Nuclear Science Committee, the expert group acts as a forum for scientific and technical information exchange on advanced light water reactor (LWR) fuels with enhanced accident tolerance. The expert group focusses on the fundamental properties and behaviour under normal operations and accident conditions for advanced core materials and components (fuels, cladding, control rods, etc.). The materials considered are applicable to Gen II and Gen III Light Water Reactors, as well as Gen III+ reactors under construction. The objective of the expert group is to define and coordinate a programme of work to help advance the scientific knowledge needed to provide the technical underpinning for the development of advanced LWR fuels with enhanced accident tolerance compared to currently used zircaloy/UO 2 fuel systems, as well as other non-fuel core components with important roles in LWR performance under accident conditions. This document brings together the available presentations (slides) given at the Third Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs. Content: 1 - Task Force 1 (Systems assessment) meeting, 3-4 March 2015: - French evaluation of ATF Concepts (J. Bischoff, AREVA); - Technology Readiness Levels - TRL - for Fuels (K. Pasamehmetoglu, INL); - TRL-definition for advanced fuel concept applied for commercial LWRs in Japan (M. Kurata, JAEA); - Application of TRLs in NNL (E. Johnston, NNL); - Technology Readiness Levels for Advanced Nuclear Fuel and Materials (S. Bragg-Sitton, INL); 1a - Definition of the illustrative scenarios: - AREVA's proposal concerning scenario for Accident Tolerant Fuel studies (P. Gandrille); - A Simplified Accident Scenario (L. Hallstadius); - Accident Scenarios for ATF Performance Evaluation of BWR and PWR in Japan (H. Ohta, CRIEPI); 1b - Related NEA activities: - Working Party on Multi-scale Modelling of Fuels and Structural Materials for Nuclear Systems - WPMM, Expert

Environmental consequences of the Chernobyl accident and their remediation: Twenty years of experience. Report of the Chernobyl Forum Expert Group 'Environment'

International Nuclear Information System (INIS)

2006-01-01

Chernobyl Forum in 2003. The mission of the Forum was - through a series of managerial and expert meetings - to generate 'authoritative consensual statements' on the environmental consequences and health effects attributable to radiation exposure arising from the accident, as well as to provide advice on environmental remediation and special health care programmes, and to suggest areas in which further research is required. The Forum was created as a contribution to the United Nations' ten year strategy for Chernobyl, launched in 2002 with the publication of Human Consequences of the Chernobyl Nuclear Accident - A Strategy for Recovery. Over a two year period, two groups of experts from 12 countries, including Belarus, the Russian Federation and Ukraine, and from relevant international organizations, assessed the accident's environmental and health consequences. In early 2005 the Expert Group 'Environment', coordinated by the IAEA, and the Expert Group 'Health', coordinated by the WHO, presented their reports for the consideration of the Chernobyl Forum. Both reports were considered and approved by the Forum at its meeting on 18-20 April 2005. This meeting also decided, inter alia, 'to consider the approved reports... as a common position of the Forum members, i.e., of the eight United Nations organizations and the three most affected countries, regarding the environmental and health consequences of the Chernobyl accident, as well as recommended future actions, i.e., as a consensus within the United Nations system.' This report presents the findings and recommendations of the Chernobyl Forum concerning the environmental effects of the Chernobyl accident. The Forum's report considering the health effects of the Chernobyl accident is being published by the WHO

Mentality of nuclear energy and industry experts from post-socialist and post-capitalist countries and a problem of cooperation

International Nuclear Information System (INIS)

Gavrilov, S.D.; Kremnev, V.A.

1996-01-01

The disintegration of the Socialist Community and thereafter Yugoslavia, the Soviet Union and Czechoslovakia has led to the establishment of more than 20 post-socialist states. Their economies differ widely, and yet a mentality of the overwhelming majority of various strata of the population, including scientists and engineers, is almost invariable so far. The processes have had an effect on the mentality of nuclear experts, especially those who deal with nuclear weapon. The superposition of general crises of post-socialist countries and an availability of nuclear facilities and installations and nuclear technologies have already had an adverse effect on the specialists' mentality. On the other hand, experts from the post-capitalist countries have other mentalities and their socioeconomic position is cardinally distinguished. Thus, at most they may barely perceive the post-socialist specialists' envelope of mentality. It doesn't only prevent individuals' collaboration and organizations' cooperation by the lack of socio-culture-psychological misunderstanding,. In some instances it has already resulted in the interruption or cancellation of joint projects at the early stages of their implementation. The report is devoted to the problem of mutually beneficial interaction between experts and organizations from recently antagonistic systems, promoting better understanding, and a different rank specialists mentality convergence. The mentality and economic situation for experts from both the post-socialist countries and the Western industrial economies are discussed

Health effects of the Chernobyl accident and special health care programmes. Report of the UN Chernobyl Forum Expert Group 'Health' (EGH). Working draft

International Nuclear Information System (INIS)

2005-01-01

This report has been prepared by three WHO expert committees convened under auspices of the Chernobyl Forum's Expert Group 'Health' (EGH), and by WHO staff. It provides an updated assessment of the health consequences of the Chernobyl accident, and follows a detailed report on this topic published by the United Nations Scientific Committee on the Effects of Atomic Radiation in 2000 (UNSCEAR, 2000). The accident occurred at the Chernobyl nuclear power plant in northern Ukraine on April 26, 1986 and released large amounts of radioactivity, primarily radioactive isotopes of caesium and iodine. These releases contaminated large areas of Belarus, the Russian Federation and Ukraine and other countries to a lesser extent, These releases exposed sizable populations to internal and external radiation doses. The Chernobyl accident caused the deaths of 30 power plant employees and firemen within a few days or weeks (including 28 deaths that were due to radiation exposure). In addition, about 240,000 recovery operation workers (also called 'liquidators' or 'clean-up workers') were called upon in 1986 and 1987 to take part in major mitigation activities at the reactor and within the 30-km zone surrounding the reactor. Residual mitigation activities continued on a relatively large scale until 1990. All together, about 600,000 persons (civilian and military) have received special certificates confirming their status as liquidators, according to laws promulgated in Belarus, the Russian Federation, and Ukraine (UNSCEAR, 2000). In addition, massive releases of radioactive materials into the atmosphere brought about the evacuation of about 116,000 people from areas surrounding the reactor during 1986, and the relocation, after 1986, of about 220,000 people from what are at this time three independent republics of the former Soviet Union: Belarus, the Russian Federation, and Ukraine. Vast territories of those three republics were contaminated to a substantial level. The population of

The application of the assessment of nuclear accident status in emergency decision-making during nuclear accident

International Nuclear Information System (INIS)

Yang Ling

2011-01-01

Nuclear accident assessment is one of the bases for emergency decision-making in the situation of nuclear accident in NPP. Usually, the assessment includes accident status and consequence assessment. It is accident status assessment, and its application in emergency decision-making is introduced here. (author)

ASSESSING ACCIDENT HOTSPOTS BY USING VOLUNTEERED GEOGRAPHIC INFORMATION

Directory of Open Access Journals (Sweden)

Golnoosh

2017-11-01

Full Text Available Due to the ever-increasing number of vehicles, transportation issues, especially transportation safety have gained great importance. One of the social problems in the world, and particularly in developing countries, which each year imposes great casualties, and economic, social and cultural costs on society, is traffic accidents. Traffic accidents cause waste of time and assets and loss of human resources in society, therefore studies and measures to reduce accidents and damage caused by them, particularly in recent decades, has become important. One of the suggested ways to deal with the problem of car accidents is the modeling of accident-prone points, as by identifying these points, factors affecting accidents can be identified, and elimination of these factors leads to a reduction in accidents. Numerous studies have been conducted in this respect, using official police data to identify these points and performing necessary analysis on them. Official data has gaps and shortcomings. Using Volunteered Geographic Information to determine accident-prone venues can be a suitable answer to the problems of using official data. The aim of this study is the use of volunteered geographic information in relation to the accidents and their causes. By taking into account factors affecting traffic accidents in the study area, and determining the importance of each factor, as well as the severity-of-accidents parameter, and using the Expert Choice software, a decision-making software based on the hierarchical analysis, high-risk venues are determined, and the accident-prone points of the study area are specified.

Accident diagnosis system based on real-time decision tree expert system

Science.gov (United States)

Nicolau, Andressa dos S.; Augusto, João P. da S. C.; Schirru, Roberto

2017-06-01

Safety is one of the most studied topics when referring to power stations. For that reason, sensors and alarms develop an important role in environmental and human protection. When abnormal event happens, it triggers a chain of alarms that must be, somehow, checked by the control room operators. In this case, diagnosis support system can help operators to accurately identify the possible root-cause of the problem in short time. In this article, we present a computational model of a generic diagnose support system based on artificial intelligence, that was applied on the dataset of two real power stations: Angra1 Nuclear Power Plant and Santo Antônio Hydroelectric Plant. The proposed system processes all the information logged in the sequence of events before a shutdown signal using the expert's knowledge inputted into an expert system indicating the chain of events, from the shutdown signal to its root-cause. The results of both applications showed that the support system is a potential tool to help the control room operators identify abnormal events, as accidents and consequently increase the safety.

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

Assessment of accident energetics in LMFBR core-disruptive accidents

International Nuclear Information System (INIS)

Fauske, H.K.

1977-01-01

An assessment of accident energetics in LMFBR core-disruptive accidents is given with emphasis on the generic issues of energetic recriticality and energetic fuel-coolant interaction events. Application of a few general behavior principles to the oxide-fueled system suggests that such events are highly unlikely following a postulated core meltdown event

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

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.

Hierarchical Model of Assessing and Selecting Experts

Science.gov (United States)

Chernysheva, T. Y.; Korchuganova, M. A.; Borisov, V. V.; Min'kov, S. L.

2016-04-01

Revealing experts’ competences is a multi-objective issue. Authors of the paper deal with competence assessing methods of experts seen as objects, and criteria of qualities. An analytic hierarchy process of assessing and ranking experts is offered, which is based on paired comparison matrices and scores, quality parameters are taken into account as well. Calculation and assessment of experts is given as an example.

Utilization of accident databases and fuzzy sets to estimate frequency of HazMat transport accidents

International Nuclear Information System (INIS)

Qiao Yuanhua; Keren, Nir; Mannan, M. Sam

2009-01-01

Risk assessment and management of transportation of hazardous materials (HazMat) require the estimation of accident frequency. This paper presents a methodology to estimate hazardous materials transportation accident frequency by utilizing publicly available databases and expert knowledge. The estimation process addresses route-dependent and route-independent variables. Negative binomial regression is applied to an analysis of the Department of Public Safety (DPS) accident database to derive basic accident frequency as a function of route-dependent variables, while the effects of route-independent variables are modeled by fuzzy logic. The integrated methodology provides the basis for an overall transportation risk analysis, which can be used later to develop a decision support system.

What are the factors that contribute to road accidents? An assessment of law enforcement views, ordinary drivers' opinions, and road accident records.

Science.gov (United States)

Rolison, Jonathan J; Regev, Shirley; Moutari, Salissou; Feeney, Aidan

2018-06-01

What are the main contributing factors to road accidents? Factors such as inexperience, lack of skill, and risk-taking behaviors have been associated with the collisions of young drivers. In contrast, visual, cognitive, and mobility impairment have been associated with the collisions of older drivers. We investigated the main causes of road accidents by drawing on multiple sources: expert views of police officers, lay views of the driving public, and official road accident records. In Studies 1 and 2, police officers and the public were asked about the typical causes of road traffic collisions using hypothetical accident scenarios. In Study 3, we investigated whether the views of police officers and the public about accident causation influence their recall accuracy for factors reported to contribute to hypothetical road accidents. The results show that both expert views of police officers and lay views of the driving public closely approximated the typical factors associated with the collisions of young and older drivers, as determined from official accident records. The results also reveal potential underreporting of factors in existing accident records, identifying possible inadequacies in law enforcement practices for investigating driver distraction, drug and alcohol impairment, and uncorrected or defective eyesight. Our investigation also highlights a need for accident report forms to be continuously reviewed and updated to ensure that contributing factor lists reflect the full range of factors that contribute to road accidents. Finally, the views held by police officers and the public on accident causation influenced their memory recall of factors involved in hypothetical scenarios. These findings indicate that delay in completing accident report forms should be minimised, possibly by use of mobile reporting devices at the accident scene. Copyright © 2018 The Authors. Published by Elsevier Ltd.. All rights reserved.

Chairperson's Summary, International Expert Meeting on Decommissioning and Remediation after a Nuclear Accident

International Nuclear Information System (INIS)

Larsson, Carl Magnus

2013-01-01

History has clearly demonstrated, and it has repeated itself in the events and aftermath of March 2011, that a major nuclear accident, just as any other major accident, not only affects public and environmental health, but in addition causes a wide range of direct and indirect effects. These include evacuation and relocation; social unrest; indirect health effects related to anxiety, radiological stigma and symptoms of a post-traumatic nature; as well as effects on property, the economy, public policy and politics. All of these factors influence the setting of targets for decommissioning and remediation; this is often an iterative process involving consideration of the legal framework, finances, processes and methodology and technology. Importantly, decommissioning and remediation is carried out in close interaction with stakeholders, of which the public (affected by both the accident and the recovery from its consequences) form an important part. Today, we have substantial knowledge about the impact of major nuclear accidents as well as a wealth of experience - good and sometimes less so - from a range of decommissioning and remediation projects following nuclear accidents. There are also a number of lessons to be learned from decommissioning and remediation of other legacy sites that have not originated from nuclear accidents but where the problems encountered are of similar nature. Experiences have over the years been discussed in many fora, including a number of IAEA initiatives and UNSCEAR reviews of scientific information on health and environmental effects of accidents and legacies of comparable nature. It is timely to discuss this knowledge in relation to the Fukushima Daiichi accident, to provide guidance to future actions aimed at strengthening our understanding of the exposure situation, and our ability to successfully carry out decommissioning of facilities and environmental remediation after a nuclear accident. (author)

Risk assessment of complex accident scenarios

International Nuclear Information System (INIS)

Kluegel, Jens-Uwe

2012-01-01

The use of methods of risk assessment in accidents in nuclear plants is based on an old tradition. The first consistent systematic study is considered to be the Rasmussen Study of the U.S. Nuclear Regulatory Commission, NRC, WASH-1400. Above and beyond the realm of nuclear technology, there is an extensive range of accident, risk and reliability research into technical-administrative systems. In the past, it has been this area of research which has led to the development of concepts of safety precautions of the type also introduced into nuclear technology (barrier concept, defense in depth, single-failure criterion), where they are now taken for granted as trivial concepts. Also for risk analysis, nuclear technology made use of methods (such as event and fault tree analyses) whose origins were outside the nuclear field. One area in which the use of traditional methods of probabilistic safety analysis is encountering practical problems is risk assessment of complex accident scenarios in nuclear technology. A definition is offered of the term 'complex accident scenarios' in nuclear technology. A number of problems are addressed which arise in the use of traditional PSA procedures in risk assessment of complex accident scenarios. Cases of complex accident scenarios are presented to demonstrate methods of risk assessment which allow robust results to be obtained even when traditional techniques of risk analysis are maintained as a matter of principle. These methods are based on the use of conditional risk metrics. (orig.)

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

Nuclear-station post-accident liquid-sampling system: developed by Duke Power Company

International Nuclear Information System (INIS)

Burton, D.A.; Birch, M.L.; Orth, W.C.

1981-01-01

The accident at Three Mile Island showed that means must be provided to determine the radioactivity levels in high activity liquid and gaseous systems of a nuclear power plant without undue radiation exposure to personnel. The Duke Power Post Accident Liquid Sampling System provides the means for obtaining diluted liquid samples and diluted dissolved gas samples following a reactor accident involving substantial core damage. Their approach yields a straightforward engineering solution at a fraction of the cost of other systems. A description of the system, general design criteria, and color coded flow diagrams are included

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

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)

International seminar on Post nuclear accident from May 5 and 6, 2011 - The CODIRPA progresses

International Nuclear Information System (INIS)

Birraux, Claude; Lacoste, Andre-Claude; Lachaume, Jean-Luc; Mehl-Auget, Isabelle; Godet, Jean-Luc; Dandrieux, Geraldine; Delmestre, Alain; Gallay, Florence; Niel, Jean-Christophe; Bataille, Christian; Dederen, Guillaume; Guenon, Catherine; Repussard, Jacques; Cessac, Bruno; Champion, Didier; Miniere, Dominique; Andrieux, Jean-Luc; Bernard, Herve; Delalonde, Jean-Claude; Calafat, Alexis; Demet, Michel; Barbey, Pierre; Janssens, Augustin; Lazo, Ted; Ahier, Brian; Ugletveit, Finn; Guerson, Nathalie; Riou, Jeanine; Robert, Joel; Pirard, Philippe; Gerbeaux, Jerome; Foix, Olivier; Le Gac, Alain; Lahaye, Thierry; Wiest, Annick; Crouail, Pascal; Lochard, Jacques; Villers, Anita; Eberbach, Friedrich; Murith, Christophe; Samain, Jean-Paul; Javanni, Jean; Averin, Viktor; Trafimchik, Zoia; Liland, Astrid; Durand, Francois

2011-01-01

The objective of the seminar is the anticipation of the radiological emergency situations to limit the consequences of nuclear accidents. The Steering Committee to manage the post-accident of a Nuclear Accident or Radiological Emergency (CODIRPA) was started in June 2005 by the French nuclear safety authority (ASN), and was charged with developing French policy for the management of the post-accident phase of a nuclear or radiological accident situation. An ambitious program mobilising more than 200 people was put in place, including representatives of relevant national administrations and their local representatives, utility and industrial representatives, technical service organisations, nuclear safety authorities from bordering countries to France, NGOs and local elected officials. On December 2007, the previous seminar had constructively challenged the work of CODIRPA with international experiences and analysis. The 2011 seminar, organized by ASN with the support of the Parliamentary Office for the Evaluation of Scientific Choices and Technologies (OPECST), reports on the progress of the national doctrine's construction. The National Preparedness Guide for Exiting the Emergency Phase and drawn lessons concerning its local adaptation by regional and local community governmental organizations was presented at this occasion. The parliamentary mission headed by OPECST on the safety of nuclear installations holds its first public hearing on May 5, 2011. The seminar provides an opportunity for members of the parliamentary mission to learn about crisis management and post-accident of the major nuclear accident occurred in Japan at Fukushima. This document comprises a complete FR/EN detailed synthesis of the seminar followed by the recommendations of CODIRPA (in French) and by the slides of the available presentations for each session: Opening of the meeting; Session 1: National Preparedness Guide for Exiting the Emergency Phase; Session 2: Development of the

Commitment of involved actors in the preparation of accidental and post-accident situations: European experiments

International Nuclear Information System (INIS)

Schneider, Th.

2010-01-01

The author briefly describes some approaches developed within the EURANOS European research programme between 2004 and 2009 which aims at promoting the building up of a European network (NERIS) for the management of nuclear accidental and post-accident situations. Notably, he comments the experiment which took place in the Montbeliard district where two types of radiological events have been modelled and simulated: an accident in the Fessenheim nuclear power plant with two scenarios of release, and a transportation accident with a release of radioactive caesium 137. He also evokes the Norwegian experience and some other actions in Finland, Great-Britain, Spain and Slovakia where reflections on the management of accidental and post-accident situations or crisis exercises have been organized

Post-accident cleanup and decommissioning of a reference pressurized-water reactor

International Nuclear Information System (INIS)

Murphy, E.S.; Holter, G.M.

1982-10-01

This paper summarizes the results of a conceptual study to evaluate the technical requirements, costs, and safety impacts of the cleanup and decommissioning of a large pressurized water reactor (PWR) involved in an accident. The costs and occupational doses for post-accident cleanup and dcommissioning are estimated to be substantially higher than those for decommissioning following the orderly shutdown of a reactor. A major factor in these cost and occupational dose increases is the high radiation environment that exists in the containment building following an accident which restricts worker access and increases the difficulty of performing certain tasks. Other factors which influence accident cleanup and decommissioning costs are requirements for the design and construction of special tools and equipment, increased requirements for regulatory approvals, and special waste management needs. Radiation doses to the public from routine accident cleanup and decommissioning operations are estimated to be below permissible radiation dose levels in unrestricted areas and within the range of annual doses from normal background

Assessment of Mobile Accident Response Capability

International Nuclear Information System (INIS)

1983-03-01

This report presents the results of a DOE-sponsored assessment of nuclear accident response resources. It identifies the mobile resources that could be required to respond to different types of nuclear accidents including major ones like TMI-2, identifies the resources currently available and makes recommendations for the design and construction of additional mobile accident response resources to supplement those already in existence. This project is referred to as the Mobile Accident Response Capability (MARC) program

An expert system approach for safety diagnosis

International Nuclear Information System (INIS)

Erdmann, R.C.; Sun, B.K.H.

1988-01-01

An expert system was developed with the intent to provide real-time information about an accident to an operator who is in the process of diagnosing and bringing that accident under control. Explicit use was made of probabilistic risk analysis techniques and plant accident response information in constructing this system. The expert system developed contains 70 logic rules and provides contextual messages during simulated accident sequences and logic sequence information on the entire sequence in graphical form for accident diagnosis. The present analysis focuses on integrated control system-related transients with Babcock and Wilcox-type reactors. While the system developed here is limited in extent and was built for a composite reactor, it demonstrates that an expert system may enhance the operator's capability in the control room

Post-accident radiation monitors

International Nuclear Information System (INIS)

Laughlin, G.J.; Kathren, R.L.

1982-01-01

Under contract to the Nuclear Safety Analysis Center of the Electric Power Research Institute, technical information and specifications were obtained for commercially available radiological monitoring instrumentation designed for use as post-accident monitors. The information was collated and published in the NSAC Handbook of Postaccident Instrumentation (Kathren and Laughlin 1981), and included such data as range, accuracy, precision, sensitivity, and energy dependence of the detector, environmental and seismic limitations of the equipment, the testing program performed to evaluate the equipment, a list of references where the instrumentation is currently installed, and a list of features and accessories available with the monitoring systems. The information presented in this section reveals that, even though a number of vendors claim to be able to meet the guidance of Regulatory Guide 1.97 (USNRC 1980), few have actually conducted tests to verify that their equipment does indeed satisfy the guidance of this Regulatory Guide, and that some of the guidance may in fact be unrealistic

[Experience in organization of joint actions of expert divisions during the accident at P.S. Podporozniy Sayano-Shushenskaya hydroelectric power station].

Science.gov (United States)

Kolkutin, V V; Ivanov, P L; Fetisov, V A; Afanas'ev, S A; Dorozhkin, O A; Vognerubov, R N; Kuznetsov, T L

2010-01-01

The authors illustrate positive experience in organization and coordination of joint actions of expert divisions of different sectors during the accident at P.S. Podporozniy Sayano-Shushenskaya hydroelectric power station in August 2009. Special emphasis is laid on the participation of experts of quick-reaction teams formed by territorial forensic medical bureaus, mobile and supporting forces from the adjacent regions.

Relocation tabletop exercise: federal radiological response in the post-accident phase

International Nuclear Information System (INIS)

Grant, K.; Adler, M.V.; Wolff, W.F.

1986-01-01

The federal Radiological Emergency Response Plan (FRERP) was developed to provide the framework for coordinating federal radiological assistance to states and to local authorities faced with a large radiological accident. The Relocation Tabletop Exercise was conducted on December 9-11, 1985 at the Beaver Valley Power Station, the site of the simulated accident. The exercise scenario had postulated a substantial release of radioactive materials from a fuel handling accident at the Beaver Valley Power Station in Shippingport, Pennsylvania, leaving radioactive materisls deposited over part of the surrounding area. The exercise was structured as a sequential series of nice mini-scenarios, each of which focused on one problems. The exercise was intended to identify issues and problems which needed consideration or procedures which might need to be developed for this post-accident phase. It was a ''no-fault'' excercise

Expert judgement in performance assessment

International Nuclear Information System (INIS)

Wilmot, R.D.; Galson, D.A.

2000-01-01

This report is a pilot study that systematically describes the various types of expert judgement that are made throughout the development of a PA, and summarizes existing tools and practices for dealing with expert judgements. The report also includes recommendations for further work in the area of expert judgement. Expert judgements can be classified in a number of ways, including classification according to why the judgements are made and according to how the judgements are made. In terms of why judgements are made, there is a broad distinction between: Judgements concerning data that are made because alternatives are not feasible; and Judgements about the conduct of a PA that are made because there are no alternative approaches for making the decision. In the case of how judgements are made, the report distinguishes between non-elicited judgements made by individuals, non-elicited judgements made by groups, and elicited judgements made by individuals or groups. These types of judgement can generally be distinguished by the extent of the associated documentation, and hence their traceability. Tools for assessing judgements vary depending on the type of judgements being examined. Key tools are peer review, an appropriate QA regime, documentation, and elicitation. Dialogue with stake holders is also identified as important in establishing whether judgements are justified in the context in which they are used. The PA process comprises a number of stages, from establishing the assessment context, through site selection and repository design, to scenario and model development and parametrisation. The report discusses how judgements are used in each of these stages, and identifies which of the tools and procedures for assessing judgements are most appropriate at each stage. Recommendations for further work include the conduct of a trial expert elicitation to gain experience in the advantages and disadvantages of this technique, the development of guidance for peer

Expert judgement in performance assessment

Energy Technology Data Exchange (ETDEWEB)

Wilmot, R.D.; Galson, D.A. [Galson Sciences Ltd, Oakham (United Kingdom)

2000-01-01

This report is a pilot study that systematically describes the various types of expert judgement that are made throughout the development of a PA, and summarizes existing tools and practices for dealing with expert judgements. The report also includes recommendations for further work in the area of expert judgement. Expert judgements can be classified in a number of ways, including classification according to why the judgements are made and according to how the judgements are made. In terms of why judgements are made, there is a broad distinction between: Judgements concerning data that are made because alternatives are not feasible; and Judgements about the conduct of a PA that are made because there are no alternative approaches for making the decision. In the case of how judgements are made, the report distinguishes between non-elicited judgements made by individuals, non-elicited judgements made by groups, and elicited judgements made by individuals or groups. These types of judgement can generally be distinguished by the extent of the associated documentation, and hence their traceability. Tools for assessing judgements vary depending on the type of judgements being examined. Key tools are peer review, an appropriate QA regime, documentation, and elicitation. Dialogue with stake holders is also identified as important in establishing whether judgements are justified in the context in which they are used. The PA process comprises a number of stages, from establishing the assessment context, through site selection and repository design, to scenario and model development and parametrisation. The report discusses how judgements are used in each of these stages, and identifies which of the tools and procedures for assessing judgements are most appropriate at each stage. Recommendations for further work include the conduct of a trial expert elicitation to gain experience in the advantages and disadvantages of this technique, the development of guidance for peer

Cosyma a new programme package for accident consequence assessment

International Nuclear Information System (INIS)

Kelly, G.N.

1991-01-01

This report gives details of a new programme package for accident consequence assessment, prepared under the CEC's Maria programme (Methods for assessing the radiological impact of accidents) initiated in 1982 to review and build on the nuclear accident consequence assessment methods in use within the European Community

The accident at the Chernobyl' nuclear power plant and its consequences

International Nuclear Information System (INIS)

1986-08-01

The material is taken from the conclusions of the Government Commission on the causes of the accident at the fourth unit of the Chernobyl' nuclear power plant and was prepared by a team of experts appointed by the USSR State Committee on the Utilization of Atomic Energy. It contains general material describing the accident, its causes, the action taken to contain the accident and to alleviate its consequences, the radioactive contamination and health of the population and some recommendations for improving nuclear power safety. 7 annexes are devoted to the following topics: water-graphite channel reactors and operating experience with RBMK reactors, design of the reactor plant, elimination of the consequences of the accident and decontamination, estimate of the amount, composition and dynamics of the discharge of radioactive substances from the damaged reactor, atmospheric transport and radioactive contamination of the atmosphere and of the ground, expert evaluation and prediction of the radioecological state of the environment in the area of the radiation plume from the Chernobyl' nuclear power station, medical-biological problems. A separate abstract was prepared for each of these annexes. The slides presented at the post-accident review meeting are grouped in two separate volumes

An approach to build a knowledge base for reactor accident diagnostic expert system

International Nuclear Information System (INIS)

Yoshida, K.; Fujii, M.; Fujiki, K.; Yokobayashi, M.; Kohsaka, A.; Aoyagi, T.; Hirota, Y.

1987-01-01

In the development of a rule based expert system, one of the key issues is how to acquire knowledge and to build knowledge base (KB). On building the KB of DISKET, which is an expert system for nuclear reactor accident diagnosis developed in JAERI, several problems have been experienced as follows. To write rules is a time consuming task, and it is difficult to keep the objectivity and consistency of rules as the number of rules increase. Further, certainty factors (CFs) must be often determined according to engineering judgment, i.e., empirically or intuitively. A systematic approach was attempted to handle these difficulties and to build an objective KB efficiently. The approach described in this paper is based on the concept that a prototype KB, colloquially speaking an initial guess, should first be generated in a systematic way and then is to be modified and/or improved by human experts for practical use. Statistical methods, principally Factor Analysis, were used as the systematic way to build a prototype KB for the DISKET using a PWR plant simulator data. The source information is a number of data obtained from the simulation of transients, such as the status of components and annunciator etc., and major process parameters like pressures, temperatures and so on

Post-accident cleanup and decommissioning of a reference pressurized water reactor

International Nuclear Information System (INIS)

Murphy, E.S.; Holter, G.M.

1982-01-01

This paper summarizes the results of a conceptual study to evaluate the technical requirements, costs, and safety impacts of the cleanup and decommissioning of a large pressurized water reactor (PWR) involved in an accident. The costs and occupational doses for post-accident cleanup and decommissioning are estimated to be substantially higher than those for decommissioning following the orderly shutdown of a reactor. A major factor in these cost and occupational dose increases is the high radiation environment that exists in the containment building following an accident which restricts worker access and increases the difficulty of performing certain tasks. Other factors which influence accident cleanup and decommissioning costs are requirements for the design and construction of special tools and equipment, increased requirements for regulatory approvals, and special waste management needs. Radiation doses to the public from routine accident cleanup and decommissioning operations are estimated to be below permissible radiation dose levels in unrestricted areas and within the range of annual doses from normal background. 6 references, 1 figure, 7 tables

CHOOZ-A expert assessment program

International Nuclear Information System (INIS)

Bouat, M.; Godin, R.

1993-01-01

CHOOZ-A Nuclear Power Plant, the first French-Belgian PWR unit (300 MWe) was definitively shut down at the end of October 1991, after 24 years in operation. Since summer 1991, the steering committee of the French (EDF) Lifetime Project has initiated a large inquiry to the different technical specialists of EDF and external organizations, trying to define a wide expert assessment program on this plant. The aim is to improve the knowledge of aging mechanisms such as those observed on the 52 PWR French nuclear power plants (900 and 1,300 MWe), and contribute to the validation of non-destructive in-service testing methods. This paper presents the retained CHOOZ-A expert assessment program and technical lines followed during its set up. First major project stages are described, then technical choices are explained, and at last the final program is presented with the specific content of each expert assessment. The definitive program is scheduled for a three year period starting at the moment of final shutdown license acquisition, with a provisional total budget of more than US $10 million

Environmental consequences of the Chernobyl accident and their remediation: Twenty years of experience. Report of the Chernobyl Forum Expert Group 'Environment'

International Nuclear Information System (INIS)

2008-01-01

countries were also affected as a result of the atmospheric transfer of radioactive material. These countries also encountered problems in the radiation protection of their populations, but to a lesser extent than the three most affected countries. Although the accident occurred nearly two decades ago, controversy still surrounds the real impact of the disaster. Therefore the IAEA, in cooperation with the Food and Agriculture Organization of the United Nations (FAO), the United Nations Development Programme (UNDP), the United Nations Environment Programme (UNEP), the United Nations Office for the Coordination of Humanitarian Affairs (OCHA), the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR), the World Health Organization (WHO) and the World Bank, as well as the competent authorities of Belarus, the Russian Federation and Ukraine, established the Chernobyl Forum in 2003. The mission of the Forum was - through a series of managerial and expert meetings - to generate 'authoritative consensual statements' on the environmental consequences and health effects attributable to radiation exposure arising from the accident, as well as to provide advice on environmental remediation and special health care programmes, and to suggest areas in which further research is required. The Forum was created as a contribution to the United Nations' ten year strategy for Chernobyl, launched in 2002 with the publication of Human Consequences of the Chernobyl Nuclear Accident - A Strategy for Recovery. Over a two year period, two groups of experts from 12 countries, including Belarus, the Russian Federation and Ukraine, and from relevant international organizations, assessed the accident's environmental and health consequences. In early 2005 the Expert Group 'Environment', coordinated by the IAEA, and the Expert Group 'Health', coordinated by the WHO, presented their reports for the consideration of the Chernobyl Forum. Both reports were considered and approved by

Expert Judgement Assessment & SCENT Ontological Analysis

Directory of Open Access Journals (Sweden)

NICHERSU Iulian

2018-05-01

Full Text Available This study aims to provide insights in the starting point of the Horizon 2020 ECfunded project SCENT (Smart Toolbox for Εngaging Citizens into a People-Centric Observation Web Citizen Observatory (CO in terms of existing infrastructure, existing monitoring systems and some discussion on the existing legal and administrative framework that relate to flood monitoring and management in the area of Danube Delta. The methodology used in this approach is based on expert judgement and ontological analysis, using the information collected from the identified end-users of the SCENT toolbox. In this type of analysis the stages of flood monitoring and management that the experts are involved in are detailed. This is done through an Expert Judgement Assessment analysis. The latter is complemented by a set of Key Performance Indicators that the stakeholders have assessed and/or proposed for the evaluation of the SCENT demonstrations, for the impact of the project and finally for SCENT toolbox performance and usefulness. The second part of the study presents an analysis that attempts to map the interactions between different organizations and components of the existing monitoring systems in the Danube Delta case study. Expert Judgement (EJ allows to gain information from specialists in a specific field through a consultation process with one or more experts that have experience in similar and complementary topics. Expert judgment, expert estimates, or expert opinion are all terms that refer to the contents of the problem; estimates, outcomes, predictions, uncertainties, and their corresponding assumptions and conditions are all examples of expert judgment. Expert Judgement is affected by the process used to gather it. On the other hand, the ontological analysis comes to complete this study, by organizing and presenting the connections behind the flood management and land use systems in the three phases of the flood event.

Application of expert elicitation techniques in human reliability, assessment

International Nuclear Information System (INIS)

Sanyasi Rao, V.V.S.; Saraf, R.K.; Ghosh, A.K.; Kushwaha, H.S.

2006-01-01

Expert elicitation techniques are being used, in the area of technological forecasting, in estimating data needed for analysis when it is either difficult to arrive at the data by experimental means or when it is quite involved to plan and conduct the experiment. In this study, expert elicitation techniques are applied to the evaluation of the frequencies of the various accident sequences that can result from the initiating event (IE) 'High Pressure Process Water (HPPW) system failure' in typical Indian Pressurised Heavy Water Reactor (IPHWR) of the older generation. The Operating Procedure under Emergency Conditions (OPEC) for this IE involves human actions according to a pre-defined procedure. The Human Error Probabilities for all these human actions are obtained using expert elicitation techniques. These techniques aim at eliciting the opinion of the experts in the area of interest with regard to the issue in question. The uncertainty is analysed by employing the measure of dissonance and the most probable range of human error probabilities are arrived at by maximizing this measure. These values are combined using the same procedures mentioned above to yield a distribution representing the uncertainty associated with the predictions. (author)

Assessment of the effectiveness of the post-Fukushima food monitoring campaign in the first year after the nuclear accident: A hypothesis

International Nuclear Information System (INIS)

Steinhauser, Georg

2016-01-01

The purpose of this study was to assess whether or not the food monitoring campaign after the Fukushima nuclear accident has been successful in reducing the number of above-limit-food from reaching the consumers. The hypothesis of this study is that the fraction of “post-market” food can be used for this purpose, when the post-market fraction in the above-limit (p′) items is compared to the post-market fraction in the entity of food items (p) that have been screened for radionuclides ("1"3"4Cs and "1"3"7Cs). Indeed the post-market fraction in most vegetarian produce decreased significantly in the above-limit food items (p′/p   1), indicating a much lower effectiveness of the monitoring action for beef. The author speculates that, by following the governmental monitoring manual (which gives “meat” only second priority), the sudden exceedances caught the monitoring agencies unprepared and triggered a much higher density monitoring of beef with the delay of several weeks. Even then, many above limit items reached the market (mainly because the monitoring during this period had focused on the post-market). Therefore, it is likely that some above-limit beef has been consumed by the public. For other meat products, the fraction of post-market samples was very low, which does not allow for the validation of the effectiveness of the monitoring campaign. Overall, the monitoring seemed to have been more effective for vegetarian produce than for meat. - Highlights: • The fraction of post-market food above the regulatory indicates the effectiveness of the monitoring campaign. • Monitoring of vegetables has been more effective than of meat. • Beef had the highest fraction of exceedances in the post-market fraction.

Post-accident cooling capacity analysis of the AP1000 passive spent fuel pool cooling system

International Nuclear Information System (INIS)

Su Xia

2013-01-01

The passive design is used in AP1000 spent fuel pool cooling system. The decay heat of the spent fuel is removed by heating-boiling method, and makeup water is provided passively and continuously to ensure the safety of the spent fuel. Based on the analysis of the post-accident cooling capacity of the spent fuel cooling system, it is found that post-accident first 72-hour cooling under normal refueling condition and emergency full-core offload condition can be maintained by passive makeup from safety water source; 56 hours have to be waited under full core refueling condition to ensure the safety of the core and the spent fuel pool. Long-term cooling could be conducted through reserved safety interface. Makeup measure is available after accident and limited operation is needed. Makeup under control could maintain the spent fuel at sub-critical condition. Compared with traditional spent fuel pool cooling system design, the AP1000 design respond more effectively to LOCA accidents. (authors)

The risk of major nuclear accident: calculation and perception of probabilities

International Nuclear Information System (INIS)

Leveque, Francois

2013-01-01

Whereas before the Fukushima accident, already eight major accidents occurred in nuclear power plants, a number which is higher than that expected by experts and rather close to that corresponding of people perception of risk, the author discusses how to understand these differences and reconcile observations, objective probability of accidents and subjective assessment of risks, why experts have been over-optimistic, whether public opinion is irrational regarding nuclear risk, and how to measure risk and its perception. Thus, he addresses and discusses the following issues: risk calculation (cost, calculated frequency of major accident, bias between the number of observed accidents and model predictions), perceived probabilities and aversion for disasters (perception biases of probability, perception biases unfavourable to nuclear), the Bayes contribution and its application (Bayes-Laplace law, statistics, choice of an a priori probability, prediction of the next event, probability of a core fusion tomorrow)

Microbial aerosol generation during laboratory accidents and subsequent risk assessment.

Science.gov (United States)

Bennett, A; Parks, S

2006-04-01

To quantify microbial aerosols generated by a series of laboratory accidents and to use these data in risk assessment. A series of laboratory accident scenarios have been devised and the microbial aerosol generated by them has been measured using a range of microbial air samplers. The accident scenarios generating the highest aerosol concentrations were, dropping a fungal plate, dropping a large bottle, centrifuge rotor leaks and a blocked syringe filter. Many of these accidents generated low particle size aerosols, which would be inhaled into the lungs of any exposed laboratory staff. Spray factors (SFs) have been calculated using the results of these experiments as an indicator of the potential for accidents to generate microbial aerosols. Model risk assessments have been described using the SF data. Quantitative risk assessment of laboratory accidents can provide data that can aid the design of containment laboratories and the response to laboratory accidents. A methodology has been described and supporting data provided to allow microbiological safety officers to carry out quantitative risk assessment of laboratory accidents.

Appraisals and Cognitive Coping Styles Associated with Chronic Post-Traumatic Symptoms in Child Road Traffic Accident Survivors

Science.gov (United States)

Stallard, Paul; Smith, Elisabeth

2007-01-01

Background: Comparatively little is known about the cognitive appraisals and coping styles of child road traffic accident (RTA) survivors that are associated with chronic post-traumatic reactions. Methods: Seventy-five children and young people aged 7-18 who were involved in a road traffic accident and attended an accident and emergency department…

Volumes of radionuclide into the basins of water while the accident at the Chernobyl nuclear power station and a specifics of radiation situation development in the post-accidents periods

International Nuclear Information System (INIS)

Standritchuk, O.Z.; Maksin, V.I.; Goncharuk, V.V.

1996-01-01

There was stated total content of radionuclide pollution, rejected to the environment in consequence of the accident at the Chernobyl nuclear power station, specifics of qualitative and quantitative change which supposes the division of post-accident period into five conventional post-accident periods. There were given the data about the levels of main fragmentation radionuclide activity in river water, atmospheric precipitation and sewage of the objects of sanitary treatment in May 1986. According to these data there were estimated the volumes of radioactive pollution rejection to the Kiev basins of water (1.56 centre dot 10 10 Ku, that is equal to 144,57 kg of radionuclides or 3,67 % of their mass in reactor) and their going into the Dnieper river. There was shown an interconnection of all season state of water basins which are near to Chernobyl nuclear power station, with specific development of radiation situation in them after the accident. There was proposed a probated variant of improvement of the traditional technology of drinking water preparation from the open water source within 1-2 post-accident periods

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

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

Analysis of Construction Accidents in Turkey and Responsible Parties

Science.gov (United States)

GÜRCANLI, G. Emre; MÜNGEN, Uğur

2013-01-01

Construction is one of the world’s biggest industry that includes jobs as diverse as building, civil engineering, demolition, renovation, repair and maintenance. Construction workers are exposed to a wide variety of hazards. This study analyzes 1,117 expert witness reports which were submitted to criminal and labour courts. These reports are from all regions of the country and cover the period 1972–2008. Accidents were classified by the consequence of the incident, time and main causes of the accident, construction type, occupation of the victim, activity at time of the accident and party responsible for the accident. Falls (54.1%), struck by thrown/falling object (12.9%), structural collapses (9.9%) and electrocutions (7.5%) rank first four places. The accidents were most likely between the hours 15:00 and 17:00 (22.6%), 10:00–12:00 (18.7%) and just after the lunchtime (9.9%). Additionally, the most common accidents were further divided into sub-types. Expert-witness assessments were used to identify the parties at fault and what acts of negligence typically lead to accidents. Nearly two thirds of the faulty and negligent acts are carried out by the employers and employees are responsible for almost one third of all cases. PMID:24077446

Analysis of construction accidents in Turkey and responsible parties.

Science.gov (United States)

Gürcanli, G Emre; Müngen, Uğur

2013-01-01

Construction is one of the world's biggest industry that includes jobs as diverse as building, civil engineering, demolition, renovation, repair and maintenance. Construction workers are exposed to a wide variety of hazards. This study analyzes 1,117 expert witness reports which were submitted to criminal and labour courts. These reports are from all regions of the country and cover the period 1972-2008. Accidents were classified by the consequence of the incident, time and main causes of the accident, construction type, occupation of the victim, activity at time of the accident and party responsible for the accident. Falls (54.1%), struck by thrown/falling object (12.9%), structural collapses (9.9%) and electrocutions (7.5%) rank first four places. The accidents were most likely between the hours 15:00 and 17:00 (22.6%), 10:00-12:00 (18.7%) and just after the lunchtime (9.9%). Additionally, the most common accidents were further divided into sub-types. Expert-witness assessments were used to identify the parties at fault and what acts of negligence typically lead to accidents. Nearly two thirds of the faulty and negligent acts are carried out by the employers and employees are responsible for almost one third of all cases.

Should evacuation conditions after a nuclear accident be revised?

International Nuclear Information System (INIS)

Nifenecker, H.

2011-01-01

The author proposes to draw lessons from the Fukushima accident, notably in the field of post-accident management. He discusses the definition of an as widely understandable as possible method of description of risks related to irradiations after a nuclear accident. As these irradiations are mainly low dose ones which have a carcinogenic effect, he proposes to assess the average life expectancy loss due to an irradiation. Then, this risk can be easily compared with other risks like air pollution, smoking and passive smoking, and so on. Then, once this risk assessment method is well defined, it is possible to associate the inhabitants of contaminated areas to the post-accident management. They could then decide to go back to their homes or not with full knowledge of the facts

IAEA Sends International Fact-finding Expert Mission to Japan

International Nuclear Information System (INIS)

2011-01-01

Full text: The International Atomic Energy Agency will dispatch an international expert fact-finding mission to Japan. Based upon the agreement between the IAEA and the Government of Japan, the mission, comprising nearly 20 international and IAEA experts from a dozen countries, will visit Japan between 24 May and 2 June 2011. Under the leadership of Mr. Mike Weightman, HM Chief Inspector of Nuclear Installations of the United Kingdom, the mission will conduct fact-finding activities at Tokyo Electric Power Company's Fukushima Dai-ichi Nuclear Power Station (NPS) site and in other locations. The expert mission will make a preliminary assessment of the safety issues linked with TEPCO's Fukushima Dai-ichi NPS accident following the Great East Japan Earthquake and Tsunami. During the mission, areas that need further exploration or assessment based on the IAEA safety standards will also be identified. In the course of the IAEA mission, the international experts will become acquainted with the Japanese lessons learned from the accident and will share their experience and expertise in their fields of competence with the Japanese authorities. Mr. Weightman will present the mission's report at the Ministerial Conference on Nuclear Safety organised by the IAEA in Vienna from 20 to 24 June 2011, as an important input in the process of reviewing and strengthening the global nuclear safety framework that will be launched by the Conference. (IAEA)

Optimization guidance to post-accident intervention: a specific case

International Nuclear Information System (INIS)

Garcia-Ramirez, J.E.; Reyes-Sanchez, M.A.

1996-01-01

ICRP recommends the application of the system of protection to intervention situations, i.e. those in which exposure pathways are already present; e.g the public exposure following an accident. This implies that intervention must be justified and optimized, being the optimization the process of deciding the nature of protective action in order to obtain the maximum net benefit. This paper provides an example of one optimization model to guide a decision making process in a specific case of post-accident intervention. The involved scenario postulates the contamination of big quantities of reinforcing steel bars used in construction industry, and a lot of them present in the structure of several dwellings. Inhabitants of these dwellings must be protected and the proposed action is to demolish those homes exceeding some intervention criterion. The objective of this study is to reach such intervention level trough an optimization process economically focused. (author)

Risks of insufficient information communication during the post-accident period of the Chernobyl accident

International Nuclear Information System (INIS)

Joukovskaia, Olga V.; Rolevich, Igor V.

2000-01-01

The modified psychological climate and increased social-psychological pressure in the population, affected as a result of the Chernobyl accident, emerged partially because of insufficient information provided to the population with respect to the radiation and ecological conditions. Such situation resulted in development of chronic psychological stress in the majority of the population residing on the affected areas. The post-accidental stress, which appeared in many people, is characterized by its extraordinary stability. Up to 74% of the affected population were subjected to stress. In 1986 the depressing condition of anxiety was observed in 50% of those examined. By 1998 this number increased up to 76%. Aggravation of health condition still remains in the center of anxiety reasons for the majority of those examined, when in the areas contaminated greater the number of those anxious is much higher than in others. Besides, the urban population is more concerned in unsatisfactory solution of the problem of liquidation of the Chernobyl accident consequences, than village inhabitants (88,5 and 79,70/o accordingly). Noteworthy, that 43% of the urban population and only 25,20/6 of the village settlers is concerned in small efficiency of rehabilitation activities on the radioactive contaminated territories. Respondents-women 86,1%) are more anxious than men 84,2%). Besides, almost three quarters of the respondents 74,5%) for last three years became more anxious for their future and future of their children, which leads to greater worries. At the same time it is necessary to take into account, that 7 of the respondents expressed apathy and indifference to everything, and at 75% have the feeling of hopelessness. Another negative tendency exposed in the population, affected by the Chernobyl accident is the reduction of trust to the authorities and governmental bodies, reduction of satisfaction by the activity of local authorities. Only 60,6% of the interrogated

Probabilistic accident consequence uncertainty analysis -- Late health effects uncertain assessment. Volume 2: Appendices

Energy Technology Data Exchange (ETDEWEB)

Little, M.P.; Muirhead, C.R. [National Radiological Protection Board (United Kingdom); Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Harper, F.T. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States)

1997-12-01

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA late health effects models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the expert panel on late health effects, (4) short biographies of the experts, and (5) the aggregated results of their responses.

Experimental applications of an expert system to operator problem solving in process control

International Nuclear Information System (INIS)

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

1985-01-01

The United States Nuclear Regulatory Commission (USNRC) has sponsored a program to assess the effectiveness of expert systems for nuclear reactor operators. The project has included two human factors experimental evaluations of the Response Tree expert system, a prototype expert system for helping nuclear reactor operators respond to emergency conditions. This paper discusses the Response Tree expert system, the experiments which have been performed to test its effectiveness, and the results of the experiments. Reference is made to the accident at TMI. 12 refs

Internal dose assessment in radiation accidents

International Nuclear Information System (INIS)

Toohey, R.E.

2003-01-01

Although numerous models have been developed for occupational and medical internal dosimetry, they may not be applicable to an accident situation. Published dose coefficients relate effective dose to intake, but if acute deterministic effects are possible, effective dose is not a useful parameter. Consequently, dose rates to the organs of interest need to be computed from first principles. Standard bioassay methods may be used to assess body contents, but, again, the standard models for bioassay interpretation may not be applicable because of the circumstances of the accident and the prompt initiation of decorporation therapy. Examples of modifications to the standard methodologies include adjustment of biological half-times under therapy, such as in the Goiania accident, and the same effect, complicated by continued input from contaminated wounds, in the Hanford 241 Am accident. (author)

Post-glacial faulting in the Lansjaerv area, Northern Sweden. Comments from the expert group on a field visit at the Molberget post-glacial fault area, 1991

International Nuclear Information System (INIS)

Stanfors, R.; Ericsson, L.O.

1993-05-01

Post-glacial faults have been recognized in the northern Baltic shield for several decades. It is important to evaluate whether such neotectonic movements can lead to new fracturing or decisively alter the geohydrological or geohydrochemical situation around a final repository for spent nuclear fuel. The post-glacial Lansjaerv fault was chosen for an interdisciplinary study because of its relative accessibility. The goals of the study were to assess the mechanisms that caused present day scraps, to clarify the extent of any recent fracturing and to clarify the extent of any ongoing movements. All these objectives were reasonably met through a series of studies, which have been performed by SKB during 1986-1992 in two phases. This report gives a summary of the first phase of the Lansjaerv study (1986-1989) and describes achievements that have been gained during the second phase of the study. As a final of the field-work in the Lansjaerv area a meeting combined with a field excursion was arranged by SKB in June 1991 for a group of international experts. Comments from the expert group on the excursion and the overall Lansjaerv project are presented. One of the major conclusions is that the Lansjaerv post-glacial fault reactivated pre-existing old structures and that the causes of the post-glacial movements is a combination of plate tectonics and deglaciation

Instrumentation and equipment for monitoring and controlling NPP post-accident situations. Working material. Proceedings of a specialists' meeting held in Dimitrovgrad, Russian Federation, 12-15 September 1995

International Nuclear Information System (INIS)

1995-01-01

The objectives of the Specialist's Meeting were: To provide an international forum for presentation and discussion on experience with instrumentation and equipment for monitoring and controlling NPP post-accident situations; to share experience on the design and improvements in the subject area; to identify and describe advanced features for safety and post-accident management improvements. In order to facilitate a structure discussion and not to omit important problem areas, papers on the following subjects were considered to be within the scope of the Specialist's Meeting: Analysis of existing post-accident instrumentation (in terms of range, qualification and redundancy) to support post-accident management. Requirements for post accident instrumentation and equipment - operating experience, regulations, standards; experience in upgrading of the existing instrumentation; instrumentation and equipment for emergency management systems (emergency rooms and national centers); post-accident computerized aids to assist operators; post-accident operating procedures; R and D in post-accident instrumentation, future trends. The present volume contains: (1) the papers presented by national delegates; (2) programme of the meeting; (3) description of the RIAR research facilities; and (4) list of participants. Refs, figs and tabs

Learning lessons from Natech accidents - the eNATECH accident database

Science.gov (United States)

Krausmann, Elisabeth; Girgin, Serkan

2016-04-01

When natural hazards impact industrial facilities that house or process hazardous materials, fires, explosions and toxic releases can occur. This type of accident is commonly referred to as Natech accident. In order to prevent the recurrence of accidents or to better mitigate their consequences, lessons-learned type studies using available accident data are usually carried out. Through post-accident analysis, conclusions can be drawn on the most common damage and failure modes and hazmat release paths, particularly vulnerable storage and process equipment, and the hazardous materials most commonly involved in these types of accidents. These analyses also lend themselves to identifying technical and organisational risk-reduction measures that require improvement or are missing. Industrial accident databases are commonly used for retrieving sets of Natech accident case histories for further analysis. These databases contain accident data from the open literature, government authorities or in-company sources. The quality of reported information is not uniform and exhibits different levels of detail and accuracy. This is due to the difficulty of finding qualified information sources, especially in situations where accident reporting by the industry or by authorities is not compulsory, e.g. when spill quantities are below the reporting threshold. Data collection has then to rely on voluntary record keeping often by non-experts. The level of detail is particularly non-uniform for Natech accident data depending on whether the consequences of the Natech event were major or minor, and whether comprehensive information was available for reporting. In addition to the reporting bias towards high-consequence events, industrial accident databases frequently lack information on the severity of the triggering natural hazard, as well as on failure modes that led to the hazmat release. This makes it difficult to reconstruct the dynamics of the accident and renders the development of

Accident Assessment

International Nuclear Information System (INIS)

Tripputi, Ivo; Lund, Ingemar

2002-01-01

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

NPP post-accident monitoring system based on unmanned aircraft vehicle:concept, design principles

International Nuclear Information System (INIS)

Sachenko, A.A.; Kochan, V.V.; Kharchenko, V.S.; Yanovskij, M.Eh.; Yastrebenetskij, M.A.; Fesenko, G.V.

2016-01-01

The paper presents a concept of designing the post-accident system for monitoring the equipment and territory of nuclear power plant after a severe accident based on unmanned aircraft vehicle (UAVs). Wired power and communications networks are found out as the most vulnerable ones during the accident monitoring, and informativity, reliability and veracity are recognized as system basic parameters. It is proposed to equip measurement and control modules with backup wireless communication channels and deploy the repeaters network based on UAVs to ensure the informativity. Modules possess the backup power battery, and repeaters appear in the appropriate places after the accident to provide the survivability. Moreover, an optimization of UAVs' location is proposed according to the minimum energy consumption criterion. To ensure the veracity, it is expected to design the noise-immune protocol for message exchange and archiving and self-diagnostics of all system components

The TEX-I real-time expert system, applied to situation assessment for the SNR-300 reactor

International Nuclear Information System (INIS)

Schmal, N.; Leder, H.J.; Schade, H.J.

1988-01-01

Interatom, a subsidiary company of Siemens, is developing expert systems for the technical domain. These systems are operating in various industrial applications like flexible manufacturing or plant configuration, based on a domain-specific expert system shell, developed by Interatom. Additional projects are focusing on real-time diagnostics, e.g., for nuclear power plants. The authors report in this paper about a diagnosis expert system for the liquid-metal fast breeder reactor SNR-300, which uses new real-time tools, developed within the German TEX-I project (technical expert systems for data interpretation, diagnosis, and process control). The purpose of the system is to support the reactor operators in assessing plant status in real time, based on readings from many sensors. By on-line connection to the process control computer, it can monitor all incoming signal values, check the consistency of data, continuously diagnose the current plant status, detect unusual trends prior to accidents, localize faulty components, and recommended operator response in abnormal conditions. In the present knowledge acquisition and test phase, the expert system is connected to a real-time simulation of the reactor. The simulator is based on a thermohydraulic code for simulation of the transient behavior of temperatures and flow rates in the reactor core, plena, pipes, pumps, valves, intermediate heat exchangers, and cooling components. Additionally, the system's response to an asynchronous operator interaction can be simulated

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

Predictors of Post-Traumatic Stress Disorder among Victims of Serious Motor Vehicle Accidents

Directory of Open Access Journals (Sweden)

Naema Khodadadi-Hassankiadeh

2017-10-01

Full Text Available Background: Compelling evidence has shown that motor vehicle accidents have an enormous impact on mental health. Post-traumatic Stress Disorder (PTSD is one of the most common psychological consequences in adult survivors of accidents, so it is important to understand the prevalence and predictors of this issue since delay causes damage to crucial daily functioning. This study aimed at investigating the prevalence and predictors of PTSD after motor vehicle accident. Methods: This cross-sectional study was conducted on 528 injured patients six weeks to six months after motor vehicle accident in Imam Reza Clinic of Poursina hospital, Rasht in 2015. Data collection tools were three questionnaires including post-traumatic stress-self report (PSS, Beck Depression Inventory (BDI-II, and the Numeric Rating Scale (NRS for pain. The data were analyzed in SPSS (Version 19 using Chi-square, Fischer’s exact test and multivariate logistic regression. Significance level was considered P≤0.05. Results: The prevalence of PTSD and depression was 30.49% and 19.89% in participants, respectively. Chi-square test indicated a significant relationship among age (P=0.02, sex (P<0.001, education level (P<0.001, work status (P<0.001 and PTSD. Participants who reported pain (P<0.001 and depression (P<0.001 were more likely to have high score of PTSD than the others. Multivariate logistic regression showed this significance in sex, depression, age, educational status and pain, as constant risk factors in developing PTSD after accident. Conclusion: This study suggests that primary care setting should be readily prompted for diagnosis of these disorders in non-treatment seeking individuals in the community.

Should evacuation conditions after a nuclear accident be revised?; Faut-il revoir les conditions d'evacuation a la suite d'un accident nucleaire?

Energy Technology Data Exchange (ETDEWEB)

Nifenecker, H.

2011-07-01

The author proposes to draw lessons from the Fukushima accident, notably in the field of post-accident management. He discusses the definition of an as widely understandable as possible method of description of risks related to irradiations after a nuclear accident. As these irradiations are mainly low dose ones which have a carcinogenic effect, he proposes to assess the average life expectancy loss due to an irradiation. Then, this risk can be easily compared with other risks like air pollution, smoking and passive smoking, and so on. Then, once this risk assessment method is well defined, it is possible to associate the inhabitants of contaminated areas to the post-accident management. They could then decide to go back to their homes or not with full knowledge of the facts

System Design Strategies of Post-Accident Monitoring System for a PGSFR in Korea

International Nuclear Information System (INIS)

Jang, Gwi-sook; Jeong, Kwang-il; Keum, Jong-yong; Seong, Seung-hwan

2013-06-01

Monitoring systems of a PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor) in Korea provide alarms, integrity information in the reactor building, sodium-water reaction information in the steam generator, fuel failure information, and supporting information for maintenance and inspection. In particular, a Post-Accident Monitoring System (PAMS) provides primary information for operators to assess the plant conditions and perform their role in bringing the plant to a safe condition during an accident. Some PAM variables can be allocated as more two types. It is important for system designers to confirm the suitability of the selection of PAM variables. In addition, the PAMS is a position 4 display against common cause failures of safety I and C systems. The position 4 display should be independent and diverse from the safety I and C systems. The diversity of safety I and C equipment has led to an increase in the design and verification and validation cost. Thus, this paper proposes the system design strategies on the PAMS design problems of the PGSFR in KOREA. The results will be input into a conceptual system design for the PAMS of the PGSFR in KOREA. (authors)

Assessing economic consequences of radiation accidents

International Nuclear Information System (INIS)

Rowe, M.D.; Lee, J.C.; Grimshaw, C.A.; Kalb, P.D.

1987-01-01

This project reviewed the literature on the economic consequences of accidents to determine the availability of assessment methods and data and their applicability to the high-level radioactive waste (HLW) disposal system before closure; determined needs for expansion, revision, or adaptation of methods and data for modeling economic consequences of accidents of the scale projected for the disposal system; and gathered data that might be useful for the needed revisions. 8 refs., 1 tab

Assessment of CRBR core disruptive accident energetics

International Nuclear Information System (INIS)

Theofanous, T.G.; Bell, C.R.

1984-03-01

The results of an independent assessment of core disruptive accident energetics for the Clinch River Breeder Reactor are presented in this document. This assessment was performed for the Nuclear Regulatory Commission under the direction of the CRBR Program Office within the Office of Nuclear Reactor Regulation. It considered in detail the accident behavior for three accident initiators that are representative of three different classes of events; unprotected loss of flow, unprotected reactivity insertion, and protected loss of heat sink. The primary system's energetics accommodation capability was realistically, yet conservatively, determined in terms of core events. This accommodation capability was found to be equivalent to an isentropic work potential for expansion to one atmosphere of 2550 MJ or a ramp rate of about 200 $/s applied to a classical two-phase disassembly

Accident assessment under emergency situation in Daya Bay nuclear power station

International Nuclear Information System (INIS)

Yang Ling; Chen Degan; Lin Shumou; Fu Guohui

2004-01-01

The accident assessment under emergency situation includes the accident status evaluation and its consequence estimation. This paper introduces evaluation methods for accident status and its assistant computer system (SESAME-GNP) utilized during the emergency situation in Guangdong Daya Bay Nuclear Power Station (GNPS) in detail. At the same time, an improved accident consequence estimation system in GNPS (RACAS-GNP) is briefly described. With the improvement of the accident assessment systems, the capability of emergency response in GNPS is strengthened

Toward the Development of Expert Assessment Systems.

Science.gov (United States)

Hasselbring, Ted S.

1986-01-01

The potential application of "expert systems" to the diagnosis and assessment of special-needs children is examined and existing prototype systems are reviewed. The future of this artificial intelligence technology is discussed in relation to emerging development tools designed for the creation of expert systems by the lay public. (Author)

Accident consequence assessment code development

International Nuclear Information System (INIS)

Homma, T.; Togawa, O.

1991-01-01

This paper describes the new computer code system, OSCAAR developed for off-site consequence assessment of a potential nuclear accident. OSCAAR consists of several modules which have modeling capabilities in atmospheric transport, foodchain transport, dosimetry, emergency response and radiological health effects. The major modules of the consequence assessment code are described, highlighting the validation and verification of the models. (author)

Philosophy of safety assurance after the Fukushima Daiichi accident. From views of experts

International Nuclear Information System (INIS)

Hisada, Tsukasa

2014-01-01

Knowledge incorporating meetings were held to exchange views of experts in order to learn respective safety concept and philosophy of safety assurance except nuclear area, how should be prepared for beyond expectation and what was needed to build social credibility, and how to upgrade safety measures of nuclear power station after the Fukushima Daiichi accident. Meeting had been held twice a year since FY2012 and two lecturers were invited at each meeting to give a lecture on the specified theme such as safety assurance in aviation area and chemical plants, and 'safety target of engineering system'. Common or different views on safety assurance between nuclear and other areas were identified, and risk concept and sincere attitude of explaining engineer were mentioned quite important for preparation for beyond expectation and building social credibility. (T. Tanaka)

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

Energy Technology Data Exchange (ETDEWEB)

None

1977-03-01

Appendices to Volume I include core-related accident-sequence definition, CRBRP risk-assessment sequence-probability determinations, failure-probability data, accident scenario evaluation, radioactive material release analysis, ex-core accident analysis, safety philosophy and design features, calculation of reactor accident consequences, sensitivity study, and risk from fires.

Post accident training program design at Three Mile Island

International Nuclear Information System (INIS)

Lawyer, L.L.

1981-01-01

The TMI preaccident training staff typically consisted of 9 professional and 3 administrative support persons. Procedures were prepared and facilities designated for operator training. The thrust of the post accident effort was directed to expanding the training function to include all other personnel while modifying the operator training to address lessons learned. Significant experiences were encountered in part task simulation, job and task analysis, decision analysis and with various external committees. These experiences led to specific opinions on industry needs in the areas of staffing, regulation, importance of training and contractor assistance

Post-processing activities after Chernobyl accident in Ukraine and lesson learned to the response Fukushima Dai-ichi accident

International Nuclear Information System (INIS)

Fujii, Yuzo

2012-01-01

After the accident of Chernobyl NPP no.4 1986, various activities including the construction of the shelter, prevention of the release of radioactive dust and liquid from the shelter, monitoring the condition of the damaged core, and disposal of radioactive waste have been implemented in the Chernobyl site for mitigating the nuclear and radioactive risks of damaged nuclear facilities, and the reducing radiation dose of working personnel. The construction of new shelter started for the decommissioning of the damaged unit no.4. facility. For reducing the radiation dose to the inhabitants from the contaminated land and feedstuff, the countermeasures including the set of the exclusive zone and permissible level of radionuclide in the foodstuff have been conducted for the countrywide. These activities include many valuable information about how to recover the condition of the site and maintain the social activities after the severe accident of NPP, and it would be important to learn the above activities in conducting the post-processing activities on the Fukushima-Daiichi accident successfully. (author)

Usefulness of post mortem computed tomography versus conventional forensic autopsy of road accident victims (drivers and passengers).

Science.gov (United States)

Moskała, Artur; Woźniak, Krzysztof; Kluza, Piotr; Romaszko, Karol; Lopatin, Oleksiy

2017-01-01

Aim of the study: Deaths of in-vehicle victims (drivers and passengers) of road accidents represent a significant group of issues addressed by forensic medicine. Expressing opinions in this regard involves first of all the determination of the cause of death and the forensic pathologist's participation in the process of road accident reconstruction through defining the mechanism of bodily harm. The scope of the opinion as well as its accuracy and degree of detail largely depend on the scope of forensic autopsy. In this context, techniques that broaden the capabilities of standard autopsy are of particular importance. This paper compares the results of post mortem computed tomography (PMCT) of road accident victims (drivers and passengers) against the results of standard examination in order to determine the scope to which PMCT significantly enhances autopsy capabilities. Material and methods: The analysis covers 118 in-vehicle victims (drivers and passengers) examined from 2012 to 2014. In each case, post-mortem examination was preceded by PMCT examination using Somatom Emotion 16 (Siemens AG, Germany). Results: The results are presented in a tabular form. Conclusions: In most road accident victims (drivers and passengers), post mortem computed tomography significantly increases the results' degree of detail, particularly with regard to injuries of bones and gas collections.

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

The Fukushima Daiichi Accident. Technical Volume 2/5. Safety Assessment

International Nuclear Information System (INIS)

2015-08-01

Technical Volume 1 of this report has described what happened during the accident at the Fukushima Daiichi nuclear power plant (NPP). This volume begins (Section 2.1) with a review of how the design basis of the site for external events was assessed initially and then reassessed over the life of the NPP. The section also describes the physical changes that were made to the units as a result. The remainder of the volume describes the treatment of beyond design basis events in the safety assessment of the site, the accident management provisions, the effectiveness of regulatory programmes, human and organizational factors and the safety culture, and the role of operating experience. Further background information is contained in three annexes included on the CD-ROM of this Technical Volume which describe analytical investigations of the accident along with information on topics such as system performance, defence in depth and severe accident phenomena. Section 2.2 provides an assessment of the systems that failed, resulting in a failure to maintain the fundamental safety functions in Units 1–3, which were in operation at the time of the tsunami and in which the reactor pressure vessels (RPV) and containment vessels failed. The section also describes Units 4-6, which were shut down at the time of the tsunami, and the site’s central spent fuel storage facility. Section 2.3 discusses the probabilistic and deterministic safety assessments of beyond design basis accidents (BDBAs) that had been performed for the plant and the insights from these assessments that had led to changes in the plant’s design. The section pays particular attention to the assessment of extreme natural hazards, such as the one which led to the total loss of AC power supply on the site. The additional loss of DC power supply in Units 1 and 2 played a key role in the progression of the accident because it impeded the diagnosis of plant conditions and made the operators unaware of the status of

Unusually extensive head trauma in a hydraulic elevator accident: post-mortem MSCT findings, autopsy results and scene reconstruction.

Science.gov (United States)

Jacobsen, Christina; Schön, Corinna A; Kneubuehl, Beat; Thali, Michael J; Aghayev, Emin

2008-10-01

Accidental or intentional falls from a height are a form of blunt trauma and occur frequently in forensic medicine. Reports describing elevator accidents as a small subcategory of falls from heights are rare in the medical literature and no report on injury patterns or scene reconstruction of such an accident was found. A case of an accident in a hydraulic elevator with a man falling 3m was examined using post-mortem multi-slice computed tomography (MSCT) and autopsy. The man suffered an unusually extensive trauma and died at the scene. Post-mortem MSCT examination showed a comminute fracture of the skull, the right femur and the first lumbar vertebra. Severe lacerations of the brain with epidural, subdural and subarachnoidal haemorrhages over both hemispheres were diagnosed. Autopsy confirmed these findings. To reconstruct the accident we used radiological and autopsy results as well as findings at the scene.

EPRI nuclear fuel-cycle accident risk assessment

International Nuclear Information System (INIS)

Anon.

1981-01-01

The present results of the nuclear fuel-cycle accident risk assessment conducted by the Electric Power Research Institute show that the total risk contribution of the nuclear fuel cycle is only approx. 1% of the accident risk of the power plant; hence, with little error, the accident risk of nuclear electric power is essentially that of the power plant itself. The power-plant risk, assuming a very large usage of nuclear power by the year 2005 is only approx. 0.5% of the radiological risk of natural background. The smallness of the fuel-cycle risk relative to the power-plant risk may be attributed to the lack of internal energy to drive an accident and the small amount of dispersible material. This work aims at a realistic assessment of the process hazards, the effectiveness of confinement and mitigation systems and procedures, and the associated likelihood of errors and the estimated size of errors. The primary probabilistic estimation tool is fault-tree analysis, with the release source terms calculated using physicochemical processes. Doses and health effects are calculated with CRAC (Consequences of Reactor Accident Code). No evacuation or mitigation is considered; source terms may be conservative through the assumption of high fuel burnup (40,000 MWd/t) and short cooling period (90 to 150 d); high-efficiency particulate air filter efficiencies are derived from experiments

Health effects of the Chernobyl accident and special health care programmes. Report of the UN Chernobyl Forum Expert Group 'Health'

International Nuclear Information System (INIS)

Bennett, B.; Repacholi, M.; Carr, Z.

2006-01-01

Twenty years have passed since the worst nuclear reactor accident in the world occurred at the Chernobyl nuclear power plant in Ukraine. The radioactive contamination which resulted from the explosion and fire in the first few days spread over large areas of neighbouring Belarus and the Russian Federation, with most of the fallout in Belarus. While national and local authorities did not immediately disclose the scale of the accident, the mitigation measures, such as distribution of potassium iodine pills, food restriction, and mass evacuation from areas where the radioactive contamination was greatest, undoubtedly reduced the health impact of the radiation exposure and saved many lives. The accident caused severe social and economic disruption and had significant environmental and health impact. This was aggravated by the political and economical changes in the three affected states related to the break-down of the Soviet Union. In the aftermath of the accident the international scientific and medical community collaborated closely with national experts dealing with health effects of the accident in the affected countries. There is a substantial body of international collaborative projects on the situation, which should lead to advancement in radiation sciences. However, considerable speculation and disinformation remains about the possible health impact of the accident for the millions of affected people. To address the health, environmental and socioeconomic consequences of the Chernobyl accident, the United Nations in 2003 launched an Inter-Agency initiative, the Chernobyl Forum. The Forum's Secretariat, led by the International Atomic Energy Agency (IAEA), the World Health Organization (WHO), the United Nations Development Programme (UNDP), and several other international organizations collaborated with the governments of the affected countries. The purpose of the Chernobyl Forum was to review the consequences of the accident, issue technical reports and, based

Experimental evaluation of an expert system for nuclear reactor operators

International Nuclear Information System (INIS)

Nelson, W.R.

1984-10-01

The United States Nuclear Regulatory Commission (USNRC) is supporting a program for the experimental evaluation of an expert system for nuclear reactor operators. A prototype expert system, called the Response Tree System, has been developed and implemented at INEL. The Response Tree System is designed to assess the status of a reactor system following an accident and recommend corrective actions to reactor operators. The system is implemented using color graphic displays and is driven by a computer simulation of the reactor system. Control of the system is accomplished using a transparent touch panel. Controlled experiments are being conducted to measure performance differences between operators using the Response Tree System and those not using it to respond to simulated accident situations. This paper summarizes the methodology and results of the evaluation of the Response Tree System, including the quantitative results obtained in the experiments thus far. Design features of the Response Tree System are discussed, and general conclusions regarding the applicability of expert systems in reactor control rooms are presented

Estimation of performance shaping factors for overtime and shift length using expert judgment based on related assessments

International Nuclear Information System (INIS)

Vickroy, S.C.

1986-01-01

This paper presents the results of a study to estimate human performance and error rate under several amounts of overtime and different shifts without the use of human subjects. Ten chronobiology, fatigue, and shift scheduling experts were administered a questionnaire to rate the effects that several shifts and overtime amounts might have on the performance of individuals working under their constraints. The data from the ratings were transformed to generate performance shaping factors used for sensitivity analyses on three previously published probabilistic risk assessments. This procedure was performed to determine the effect that different shift schedules and amounts of overtime would have on overall plant performance. The results of the analysis suggest that the risk due to human errors caused by the shift scheduling and overtime could increase the risk of accidents at a nuclear power plant caused by humans by up to a factor of five. This could increase the chance for an accident at a nuclear power plant by a factor of about three

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

Program for rapid dose assessment in criticality accident, RADAPAS

International Nuclear Information System (INIS)

Takahashi, Fumiaki

2006-09-01

In a criticality accident, a person near fissile material can receive extremely high dose which can cause acute health effect. For such a case, medical treatment should be carried out for the exposed person, according to severity of the exposure. Then, radiation dose should be rapidly assessed soon after an outbreak of an accident. Dose assessment based upon the quantity of induced 24 Na in human body through neutron exposure is expected as one of useful dosimetry techniques in a criticality accident. A dose assessment program, called RADAPAS (RApid Dose Assessment Program from Activated Sodium in Criticality Accidents), was therefore developed to assess rapidly radiation dose to exposed persons from activity of induced 24 Na. RADAPAS consists of two parts; one is a database part and the other is a part for execution of dose calculation. The database contains data compendiums of energy spectra and dose conversion coefficients from specific activity of 24 Na induced in human body, which had been derived in a previous analysis using Monte Carlo calculation code. Information for criticality configuration or characteristics of radiation in the accident field is to be interactively given with interface displays in the dose calculation. RADAPAS can rapidly derive radiation dose to the exposed person from the given information and measured 24 Na specific activity by using the conversion coefficient in database. This report describes data for dose conversions and dose calculation in RADAPAS and explains how to use the program. (author)

Expert judgement models in quantitative risk assessment

Energy Technology Data Exchange (ETDEWEB)

Rosqvist, T. [VTT Automation, Helsinki (Finland); Tuominen, R. [VTT Automation, Tampere (Finland)

1999-12-01

Expert judgement is a valuable source of information in risk management. Especially, risk-based decision making relies significantly on quantitative risk assessment, which requires numerical data describing the initiator event frequencies and conditional probabilities in the risk model. This data is seldom found in databases and has to be elicited from qualified experts. In this report, we discuss some modelling approaches to expert judgement in risk modelling. A classical and a Bayesian expert model is presented and applied to real case expert judgement data. The cornerstone in the models is the log-normal distribution, which is argued to be a satisfactory choice for modelling degree-of-belief type probability distributions with respect to the unknown parameters in a risk model. Expert judgements are qualified according to bias, dispersion, and dependency, which are treated differently in the classical and Bayesian approaches. The differences are pointed out and related to the application task. Differences in the results obtained from the different approaches, as applied to real case expert judgement data, are discussed. Also, the role of a degree-of-belief type probability in risk decision making is discussed.

Expert judgement models in quantitative risk assessment

International Nuclear Information System (INIS)

Rosqvist, T.; Tuominen, R.

1999-01-01

Expert judgement is a valuable source of information in risk management. Especially, risk-based decision making relies significantly on quantitative risk assessment, which requires numerical data describing the initiator event frequencies and conditional probabilities in the risk model. This data is seldom found in databases and has to be elicited from qualified experts. In this report, we discuss some modelling approaches to expert judgement in risk modelling. A classical and a Bayesian expert model is presented and applied to real case expert judgement data. The cornerstone in the models is the log-normal distribution, which is argued to be a satisfactory choice for modelling degree-of-belief type probability distributions with respect to the unknown parameters in a risk model. Expert judgements are qualified according to bias, dispersion, and dependency, which are treated differently in the classical and Bayesian approaches. The differences are pointed out and related to the application task. Differences in the results obtained from the different approaches, as applied to real case expert judgement data, are discussed. Also, the role of a degree-of-belief type probability in risk decision making is discussed

Severe accidents risk assessment as a basis for emergency preparedness

International Nuclear Information System (INIS)

Sinka, D.; Mikulicic, V.

2000-01-01

The paper demonstrates, by example of the Republic of Croatia, the possibilities of implementing risk assessment as basis for nuclear accident emergency preparedness development. Individual risks of severe accidents for citizens of the biggest Croatian population centers, as well as collective risk for entire population have been assessed using the PRONEL method. The assessment covered 90 power reactors located at a distance up to 1.000 km. The conducted assessment shows the risks for various regions of the Republic of Croatia, and comparison between them. If risk would be taken as basic criterion in nuclear emergency planning, the results of assessment would directly indicate the necessary preparation level for each region. Furthermore, the assessment of risks from individual power plants and power plant types indicates to which facilities the greatest attention should be paid in nuclear accidents preparedness development. Risks from groups of power plants formed in accordance with their respective distance from exposure location shows what kind of tools for determining consequences and protective actions during a nuclear accident should be made available. (author)

Socioeconomic consequences of nuclear reactor accidents

International Nuclear Information System (INIS)

Tawil, J.J.; Callaway, J.W.; Coles, B.L.; Cronin, F.J.; Currie, J.W.; Imhoff, K.L.; Lewis, P.M.; Nesse, R.J.; Strenge, D.L.

1984-06-01

This report identifies and characterizes the off-site socioeconomic consequences that would likely result from a severe radiological accident at a nuclear power plant. The types of impacts that are addressed include economic impacts, health impacts, social/psychological impacts and institutional impacts. These impacts are identified for each of several phases of a reactor accident - from the warning phase through the post-resettlement phase. The relative importance of the impact during each accident phase and the degree to which the impact can be predicted are indicated. The report also examines the methods that are currently used for assessing nuclear reactor accidents, including development of accident scenarios and the estimating of socioeconomic accident consequences with various models. Finally, a critical evaluation is made regarding the use of impact analyses in estimating the contribution of socioeconomic consequences to nuclear accident reactor accident risk. 116 references, 7 figures, 15 tables

JAERI's activities in JCO accident

International Nuclear Information System (INIS)

2000-09-01

The Japan Atomic Energy Research Institute (JAERI) was actively involved in a variety of technical supports and cooperative activities, such as advice on terminating the criticality condition, contamination checks of the residents and consultation services for the residents, as emergency response actions to the criticality accident at the uranium processing facility operated by the JCO Co. Ltd., which occurred on September 30, 1999. These activities were carried out in collaborative ways by the JAERI staff from the Tokai Research Establishment, Naka Fusion Research Establishment, Oarai Research Establishment, and Headquarter Office in Tokyo. As well, the JAERI was engaged in the post-accident activities such as identification of accident causes, analyses of the criticality accident, and dose assessment of exposed residents, to support the Headquarter for Accident Countermeasures of the Science and Technology Agency (STA), the Accident Investigation Committee and the Health Control Committee of the Nuclear Safety Commission of Japan (NSC). This report compiles the activities, that the JAERI has conducted to date, including the discussions on measures for terminating the criticality condition, evaluation of the fission number, radiation monitoring in the environment, dose assessment, analyses of criticality dynamics. (author)

Commitment of involved actors in the preparation of accidental and post-accident situations: European experiments; Engagement des parties prenantes a la preparation aux situations accidentelles et post-accidentelles: experiences Europeennes

Energy Technology Data Exchange (ETDEWEB)

Schneider, Th. [CEPN, 28 rue de la Redoute, 92260 Fontenay-aux-Roses (France)

2010-07-01

The author briefly describes some approaches developed within the EURANOS European research programme between 2004 and 2009 which aims at promoting the building up of a European network (NERIS) for the management of nuclear accidental and post-accident situations. Notably, he comments the experiment which took place in the Montbeliard district where two types of radiological events have been modelled and simulated: an accident in the Fessenheim nuclear power plant with two scenarios of release, and a transportation accident with a release of radioactive caesium 137. He also evokes the Norwegian experience and some other actions in Finland, Great-Britain, Spain and Slovakia where reflections on the management of accidental and post-accident situations or crisis exercises have been organized

[Construction of indicators for assessing the policy of reducing accidents and violence for the elderly care].

Science.gov (United States)

de Souza, Edinilsa Ramos; Correia, Bruna Soares Chaves

2010-09-01

The follow article presents the methodology used to construct indicators to assess the implementation of the National Policy of Mortality Reduction by Accidents and Violence, of public health policies aimed at the elderly and the Mental Health Policy developed in the research entitled Diagnostic Analysis of Local Health Systems to Meet the Problems Caused by Accidents and Violence against Elderly. These indicators were applied in health services that meet elderly victims of accidents and violence in five Brazilian cities: Brasília, Curitiba, Manaus, Recife and Rio de Janeiro. It started with 124 indicatives to assistance level pre-hospital, hospital, rehabilitation and CAPS. There was a selection phase where indicatives without discriminatory capability were eliminated. It was also decided by the relaxation of some criteria are creating new categories. After this step, a group of the experts validate the questionnaires created with these indicators by using Nominal Technical Group. We performed the Kruskal-Wallis test and a graphical analysis. In the final round, the indicators were grouped by similarity, building synthetic indices, 60 indicatives left. These methods can be used in other organizations to evaluate and adjust their health care based on public policies.

Comparative assessment of severe accident risks in the energy sector

International Nuclear Information System (INIS)

Hirschberg, S.; Spiekerman, G.; Dones, R.

1997-01-01

This paper addresses one of the major limitations of the current comparative studies of environmental and health impacts of energy systems, i.e. the treatment of severe accidents. The work covers technical aspects of severe accidents and thus primarily reflects an engineering perspective on the energy-related risk issues. The assessments concern full energy chains associated with fossil sources (coal, oil and gas), nuclear power and hydro power. A comprehensive severe accidents database has been established. Thanks to the variety of information sources used, it exhibits in comparison with other corresponding databases a far more extensive coverage of the energy-related accidents. For hypothetical nuclear accidents the probabilistic approach has been employed and extended to cover the economic consequences of power reactor accidents. Results of comparisons between the various energy chains are shown and discussed along with a number of current issues in comparative assessment of severe accidents. As opposed to the previous studies, the aim of the present work has been, to cover whenever possible, a relatively broad spectrum of damage categories of interest. (author) 5 figs., 1 tab., 18 refs

Nuclear post-accident communication: reflections of CODIR-PA's 'Communication' work group (GT8)

International Nuclear Information System (INIS)

2010-01-01

This document briefly presents the approach adopted in France to define a communication doctrine to be applied in a post-accident situation, to make practical recommendations to actors, and to develop practical tools (speech elements). It evokes the main themes dealt with (such as safety culture, communication, etc.) by a specific work group

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

International Nuclear Information System (INIS)

Olast, M.; Sinnaeve, J.

1988-01-01

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

Probabilistic accident consequence uncertainty analysis: Dispersion and deposition uncertainty assessment, main report

International Nuclear Information System (INIS)

Harper, F.T.; Young, M.L.; Miller, L.A.; Hora, S.C.; Lui, C.H.; Goossens, L.H.J.; Cooke, R.M.; Paesler-Sauer, J.; Helton, J.C.

1995-01-01

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the risks presented by nuclear installations based on postulated frequencies and magnitudes of potential accidents. In 1991, the US Nuclear Regulatory Commission (NRC) and the Commission of the European Communities (CEC) began a joint uncertainty analysis of the two codes. The ultimate objective of the joint effort was to develop credible and traceable uncertainty distributions for the input variables of the codes. Expert elicitation was identified as the best technology available for developing a library of uncertainty distributions for the selected consequence parameters. The study was formulated jointly and was limited to the current code models and to physical quantities that could be measured in experiments. Experts developed their distributions independently. To validate the distributions generated for the wet deposition input variables, samples were taken from these distributions and propagated through the wet deposition code model. Resulting distributions closely replicated the aggregated elicited wet deposition distributions. To validate the distributions generated for the dispersion code input variables, samples from the distributions and propagated through the Gaussian plume model (GPM) implemented in the MACCS and COSYMA codes. Project teams from the NRC and CEC cooperated successfully to develop and implement a unified process for the elaboration of uncertainty distributions on consequence code input parameters. Formal expert judgment elicitation proved valuable for synthesizing the best available information. Distributions on measurable atmospheric dispersion and deposition parameters were successfully elicited from experts involved in the many phenomenological areas of consequence analysis. This volume is the first of a three-volume document describing the project

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

Key risk indicators for accident assessment conditioned on pre-crash vehicle trajectory.

Science.gov (United States)

Shi, X; Wong, Y D; Li, M Z F; Chai, C

2018-08-01

Accident events are generally unexpected and occur rarely. Pre-accident risk assessment by surrogate indicators is an effective way to identify risk levels and thus boost accident prediction. Herein, the concept of Key Risk Indicator (KRI) is proposed, which assesses risk exposures using hybrid indicators. Seven metrics are shortlisted as the basic indicators in KRI, with evaluation in terms of risk behaviour, risk avoidance, and risk margin. A typical real-world chain-collision accident and its antecedent (pre-crash) road traffic movements are retrieved from surveillance video footage, and a grid remapping method is proposed for data extraction and coordinates transformation. To investigate the feasibility of each indicator in risk assessment, a temporal-spatial case-control is designed. By comparison, Time Integrated Time-to-collision (TIT) performs better in identifying pre-accident risk conditions; while Crash Potential Index (CPI) is helpful in further picking out the severest ones (the near-accident). Based on TIT and CPI, the expressions of KRIs are developed, which enable us to evaluate risk severity with three levels, as well as the likelihood. KRI-based risk assessment also reveals predictive insights about a potential accident, including at-risk vehicles, locations and time. Furthermore, straightforward thresholds are defined flexibly in KRIs, since the impact of different threshold values is found not to be very critical. For better validation, another independent real-world accident sample is examined, and the two results are in close agreement. Hierarchical indicators such as KRIs offer new insights about pre-accident risk exposures, which is helpful for accident assessment and prediction. Copyright © 2018 Elsevier Ltd. All rights reserved.

Comparative Assessment of Severe Accidents in the Chinese Energy Sector

Energy Technology Data Exchange (ETDEWEB)

Hirschberg, S; Burgherr, P; Spiekerman, G; Cazzoli, E; Vitazek, J; Cheng, L

2003-03-01

This report deals with the comparative assessment of accidents risks characteristic for the various electricity supply options. A reasonably complete picture of the wide spectrum of health, environmental and economic effects associated with various energy systems can only be obtained by considering damages due to normal operation as well as due to accidents. The focus of the present work is on severe accidents, as these are considered controversial. By severe accidents we understand potential or actual accidents that represent a significant risk to people, property and the environment and may lead to large consequences. (author)

Semi-passive pad (post-accident dilution) system combined with recombiners or igniters for e.g., multiple-unit VVER

International Nuclear Information System (INIS)

Eckardt, B.A.

1997-01-01

In order to reduce the risk associated with hypothetical severe nuclear accidents, it has been necessary to develop means for hydrogen control which are capable of handling high H 2 production rates of several m 3 /s. These systems and components should be of simple and, as far as possible, passive design. Moreover, their efficiency must not be affected by the H 2 production scenario. To meet these requirements, Siemens has supplemented its dual process for hydrogen reduction using recombiners and igniters by developing a new process: the semi-passive Post-Accident Dilution (PAD) System. The objective of the work presented in this paper was to design an alternative approach for hydrogen control comprising post-accident CO 2 dilution combined with recombiners or igniters for postulated severe accidents that result in significant hydrogen releases. The combined technique was to keep the dynamic loads arising inside the containment to within a range of only a few millibars, even in the event of transient hydrogen release and distribution processes accompanied by combustion, and was thus to prevent any hazard for safety-related equipment. (author)

Probabilistic accident consequence uncertainty analysis: Dispersion and deposition uncertainty assessment. Volume 3, Appendices C, D, E, F, and G

Energy Technology Data Exchange (ETDEWEB)

Harper, F.T.; Young, M.L.; Miller, L.A. [Sandia National Labs., Albuquerque, NM (United States)] [and others

1995-01-01

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, completed in 1990, estimate the risks presented by nuclear installations based on postulated frequencies and magnitudes of potential accidents. In 1991, the US Nuclear Regulatory Commission (NRC) and the Commission of the European Communities (CEC) began a joint uncertainty analysis of the two codes. The objective was to develop credible and traceable uncertainty distributions for the input variables of the codes. Expert elicitation, developed independently, was identified as the best technology available for developing a library of uncertainty distributions for the selected consequence parameters. The study was formulated jointly and was limited to the current code models and to physical quantities that could be measured in experiments. To validate the distributions generated for the wet deposition input variables, samples were taken from these distributions and propagated through the wet deposition code model along with the Gaussian plume model (GPM) implemented in the MACCS and COSYMA codes. Resulting distributions closely replicated the aggregated elicited wet deposition distributions. Project teams from the NRC and CEC cooperated successfully to develop and implement a unified process for the elaboration of uncertainty distributions on consequence code input parameters. Formal expert judgment elicitation proved valuable for synthesizing the best available information. Distributions on measurable atmospheric dispersion and deposition parameters were successfully elicited from experts involved in the many phenomenological areas of consequence analysis. This volume is the third of a three-volume document describing the project and contains descriptions of the probability assessment principles; the expert identification and selection process; the weighting methods used; the inverse modeling methods; case structures; and summaries of the consequence codes.

Probabilistic accident consequence uncertainty analysis: Dispersion and deposition uncertainty assessment. Volume 3, Appendices C, D, E, F, and G

International Nuclear Information System (INIS)

Harper, F.T.; Young, M.L.; Miller, L.A.

1995-01-01

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, completed in 1990, estimate the risks presented by nuclear installations based on postulated frequencies and magnitudes of potential accidents. In 1991, the US Nuclear Regulatory Commission (NRC) and the Commission of the European Communities (CEC) began a joint uncertainty analysis of the two codes. The objective was to develop credible and traceable uncertainty distributions for the input variables of the codes. Expert elicitation, developed independently, was identified as the best technology available for developing a library of uncertainty distributions for the selected consequence parameters. The study was formulated jointly and was limited to the current code models and to physical quantities that could be measured in experiments. To validate the distributions generated for the wet deposition input variables, samples were taken from these distributions and propagated through the wet deposition code model along with the Gaussian plume model (GPM) implemented in the MACCS and COSYMA codes. Resulting distributions closely replicated the aggregated elicited wet deposition distributions. Project teams from the NRC and CEC cooperated successfully to develop and implement a unified process for the elaboration of uncertainty distributions on consequence code input parameters. Formal expert judgment elicitation proved valuable for synthesizing the best available information. Distributions on measurable atmospheric dispersion and deposition parameters were successfully elicited from experts involved in the many phenomenological areas of consequence analysis. This volume is the third of a three-volume document describing the project and contains descriptions of the probability assessment principles; the expert identification and selection process; the weighting methods used; the inverse modeling methods; case structures; and summaries of the consequence codes

[Perspective on the nuclear power plant accident caused by the great east Japan earthquake and tsunami: health impairment risks due to pollution by radioactive materials from the damaged plant as recognized by experts and by the general population and role of the experts].

Science.gov (United States)

Sugita, Minoru; Miyakawa, Michiko

2012-01-01

Large amounts of radioactive materials were leaked into the environment from the Fukushima Daiichi Nuclear Power Plant (FDNPP) of the Tokyo Electric Power Company damaged by the 2011 Great East Japan Earthquake and accompanying tsunami. Increased health impairment risks due to the leaked radioactive materials are of concern over a long period of time and over a wide geographical area. From the results of epidemiologic studies conducted after the Chernobyl accident, the health risks are not anticipated to be very marked. The purpose of the present study is to examine (i) the elevated health risks as viewed by the general population, (ii) tolerance to the risks that the general population suffer from their viewpoint, and (iii) the overall picture as seen by researchers and experts in specialized areas of study after the accident. Information was obtained from articles in print and on the Internet and by interviewing a psychologist and tens of employees of several corporations. Epidemiologic studies conducted after the severe accident of the nuclear power plant in Chernobyl revealed an elevated risk of thyroid cancer in children due to (131)I while elevated risks due to radioactive materials other than (131)I were not detected. The amount of radioactive materials leaked into the environment from the FDNPP was less than that in Chernobyl. Therefore, it is possible to estimate that health impairment risks due to the leaked radioactive materials from the FDNPP are low. However, it is impossible to conclude a zero risk. It is likely that the general population does not fully understand the health impairment risks due to the leaked radioactive materials from the FDNPP. Although no increased incidences of diseases other than thyroid cancer of children were scientifically shown en masse from studies in Chernobyl, individual risks and results in the future caused by the severe accident of FDNPP cannot be denied. Much of the general population is apt to demand the security of

Perspective on the nuclear power plant accident caused by the great east Japan earthquake and tsunami. Health impairment risks due to pollution by radioactive materials from the damaged plant as recognized by experts and by the general population and role of the experts

International Nuclear Information System (INIS)

Sugita, Minoru; Miyakawa, Michiko

2012-01-01

Large amounts of radioactive materials were leaked into the environment from the Fukushima Daiichi Nuclear Power Plant (FDNPP) of the Tokyo Electric Power Company damaged by the 2011 Great East Japan Earthquake and accompanying tsunami. Increased health impairment risks due to the leaked radioactive materials are of concern over a long period of time and over a wide geographical area. From the results of epidemiologic studies conducted after the Chernobyl accident, the health risks are not anticipated to be very marked. The purpose of the present study is to examine the elevated health risks as viewed by the general population, tolerance to the risks that the general population suffer from their viewpoint, and the overall picture as seen by researchers and experts in specialized areas of study after the accident. Information was obtained from articles in print and on the Internet and by interviewing a psychologist and tens of employees of several corporations. Epidemiologic studies conducted after the severe accident of the nuclear power plant in Chernobyl revealed an elevated risk of thyroid cancer in children due to 131 I while elevated risks due to radioactive materials other than 131 I were not detected. The amount of radioactive materials leaked into the environment from the FDNPP was less than that in Chernobyl. Therefore, it is possible to estimate that health impairment risks due to the leaked radioactive materials from the FDNPP are low. However, it is impossible to conclude a zero risk. It is likely that the general population does not fully understand the health impairment risks due to the leaked radioactive materials from the FDNPP. Although no increased incidences of diseases other than thyroid cancer of children were scientifically shown en masse from studies in Chernobyl, individual risks and results in the future caused by the severe accident of FDNPP cannot be denied. Much of the general population is apt to demand the security of a zero risk from

Determining cutoff distances for assessing risks from transportation accident radiation releases

International Nuclear Information System (INIS)

Sandquist, G.M.; Slaughter, D.M.; Kimura, C.Y.; Brumburgh, G.

1995-01-01

The transportation of radioactive materials throughout the United States and the world is a ubiquitous and sometimes controversial activity. Almost universally, these transportation activities have been performed without major incident, and the safety record for transportation of radioactive material is outstanding compared with the transportation of other hazardous materials. Nevertheless, concerns still exist regarding adequate regulation of radioactive material transportation and accurate assessment of the health risks associated with accidents. These concerns are addressed through certification by the cognizant regulatory authority over the transportation container or the performance of a transportation risk assessment. In a transportation risk assessment, accident situations are examined, frequencies are estimated, and consequences resulting from the accident are analyzed and evaluated for acceptance. A universal question with any transportation risk assessment that examines the radiological consequences from release accidents is, At what distance may the dispersion analysis be terminated? This paper examines cutoff distances and their consequences for assessing health risks from radiological transportation releases

The radiological accident in Cochabamba

International Nuclear Information System (INIS)

2004-07-01

In April 2002 an accident involving an industrial radiography source containing 192 Ir occurred in Cochabamba, Bolivia, some 400 km from the capital, La Paz. A faulty radiography source container had been sent back to the headquarters of the company concerned in La Paz together with other equipment as cargo on a passenger bus. This gave rise to a potential for serious exposure for the bus passengers as well as for the company employees who were using and transporting the source. The Government of Bolivia requested the assistance of the IAEA under the terms of the Convention on Assistance in the Case of a Nuclear Accident or Radiological Emergency. The IAEA in response assembled and sent to Bolivia a team composed of senior radiation safety experts and radiation pathology experts from Brazil, the United Kingdom and the IAEA to investigate the accident. The IAEA is grateful to the Government of Bolivia for the opportunity to report on this accident in order to disseminate the valuable lessons learned and help prevent similar accidents in the future

Probabilistic accident consequence uncertainty analysis -- Early health effects uncertainty assessment. Volume 2: Appendices

Energy Technology Data Exchange (ETDEWEB)

Haskin, F.E. [Univ. of New Mexico, Albuquerque, NM (United States); Harper, F.T. [Sandia National Labs., Albuquerque, NM (United States); Goossens, L.H.J.; Kraan, B.C.P. [Delft Univ. of Technology (Netherlands)

1997-12-01

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA early health effects models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the panel on early health effects, (4) short biographies of the experts, and (5) the aggregated results of their responses.

Probabilistic accident consequence uncertainty analysis -- Uncertainty assessment for internal dosimetry. Volume 2: Appendices

Energy Technology Data Exchange (ETDEWEB)

Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Harrison, J.D. [National Radiological Protection Board (United Kingdom); Harper, F.T. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States)

1998-04-01

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA internal dosimetry models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the panel on internal dosimetry, (4) short biographies of the experts, and (5) the aggregated results of their responses.

Identification and assessment of BWR in-vessel severe accident mitigation strategies

International Nuclear Information System (INIS)

Hodge, S.A.; Kress, T.S.; Cleveland, J.C.; Petek, M.

1992-01-01

This paper briefly describes the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to evaluate the effectiveness and feasibility of current and proposed strategies for BWR severe accident management. These results are described in detail in the just-released report Identification and Assessment of BWR In-Vessel Severe Accident Mitigation Strategies, NUREG/CR-5869, which comprises three categories of findings. First, an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences is combined with a review of the BWR Owners' Group Emergency Procedure Guidelines (EPGs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, two of the four candidate strategies identified by this effort are assessed in detail. These are (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored

The Development of Marine Accidents Human Reliability Assessment Approach: HEART Methodology and MOP Model

Directory of Open Access Journals (Sweden)

Ludfi Pratiwi Bowo

2017-06-01

Full Text Available Humans are one of the important factors in the assessment of accidents, particularly marine accidents. Hence, studies are conducted to assess the contribution of human factors in accidents. There are two generations of Human Reliability Assessment (HRA that have been developed. Those methodologies are classified by the differences of viewpoints of problem-solving, as the first generation and second generation. The accident analysis can be determined using three techniques of analysis; sequential techniques, epidemiological techniques and systemic techniques, where the marine accidents are included in the epidemiological technique. This study compares the Human Error Assessment and Reduction Technique (HEART methodology and the 4M Overturned Pyramid (MOP model, which are applied to assess marine accidents. Furthermore, the MOP model can effectively describe the relationships of other factors which affect the accidents; whereas, the HEART methodology is only focused on human factors.

SIPA, a PWR simulator for post-accident training and studies

International Nuclear Information System (INIS)

Peltier, J.; Poizat, F.

1990-01-01

SIPA (Simulator for Post-Accident conditions) which is now under development will be operated by EDF and CEA. Each organization will have its own version, SIPA 1 for EDF and SIPA 2 for CEA. The three main purposes will meet the needs of EDF and CEA as described below: - training of the EDF's ISR (Ingenieurs de Surete et Radioprotection = Shift Safety Advisors) which needs physical relevance, real time during accidental transients and visualisation of two-phase flow phenomena to well understand what could physically happen, - studies for EDF's designs which require calculation of a lot of points or scenarios. Quality Assurance of the models and data package, interactivity for procedure finalisation, availability of resources to all engineers, and possibility of creation of new models, - safety analysis requirements for CEA/IPSN (technical support of the French safety authority, the Central Service for the Safety of Nuclear Installations) which includes the actual safety analysis (analysis of procedures, design basis accidents, probabilistic safety analysis, real incidents studies, reactor tests...), the preparation and the execution of safety drills and training of engineer analysts

Assessment of chemical processes for the post-accident decontamination of reactor-coolant systems. Final report

International Nuclear Information System (INIS)

Munson, L.F.; Card, C.J.; Divine, J.R.

1983-02-01

Previously used chemical decontamination processes and potentially useful new decontamination processes were examined for the usefulness following a reactor accident. Both generic fuel damage accidents and the accident at TMI-2 were considered. A total of fourteen processes were evaluated. Process evaluation included data in the following categories: technical description of the process, recorded past usage, effectiveness, process limitation, safety consideration, and waste management. These data were evaluated, and cost considerations were presented along with a description of the applicability of the process to TMI-2 and development and demonstration needs. Specific recommendations regarding a primary-system decontamination development program to support TMI-2 recovery were also presented

Insights into location dependent loss-of-coolant-accident (LOCA) frequency assessment for GSI-191 risk-informed applications

Energy Technology Data Exchange (ETDEWEB)

Fleming, K.N., E-mail: KarlFleming@comcast.net [KNF Consulting LLC, Spokane, WA (United States); Lydell, B.O.Y. [SIGMA-PHASE INC., Vail, AZ (United States)

2016-08-15

Highlights: • Role of operating experience in loss-of-coolant-accident (LOCA) frequency assessment. • Plant-to-plant variability in calculated LOCA frequency. • Frequency of double-ended-guillotine-break (DEGB). • Uncertainties in LOCA frequencies. • Risk management insights. - Abstract: As a tribute to the published work by S.H. Bush, S. Beliczey and H. Schulz, this paper assesses the progress with methods and techniques for quantifying the reliability of piping systems in commercial nuclear power plants on the basis of failure rate estimates derived from field experience data in combination with insights and results from probabilistic fracture mechanics analyses and expert elicitation exercises. This status assessment is made from a technical perspective obtained through development of location-specific loss-of-coolant-accident (LOCA) frequencies for input to risk-informed resolution of the generic safety issue (GSI) 191. The methods and techniques on which these GSI-191 applications are based build on a body of work developed by the authors during a period spanning more than two decades. The insights that are presented and discussed in this paper cover today’s knowledge base concerning how to utilize a risk-informed approach to the assessment of piping reliability in the context of probabilistic risk assessment (PRA) in general and the resolution of GSI-191 in particular. Specifically the paper addresses the extent to which LOCA frequencies vary from location to location within a reactor coolant system pressure boundary (RCPB) for a given plant as well as vary from plant to plant, and the reasons for these variabilities. Furthermore, the paper provides the authors’ perspectives on interpretations and applications of information extracted from an expert elicitation process to obtain LOCA frequencies as documented in NUREG-1829 and how to apply this information to GSI-191. Finally, this paper documents technical insights relative to mitigation of

Insights into location dependent loss-of-coolant-accident (LOCA) frequency assessment for GSI-191 risk-informed applications

International Nuclear Information System (INIS)

Fleming, K.N.; Lydell, B.O.Y.

2016-01-01

Highlights: • Role of operating experience in loss-of-coolant-accident (LOCA) frequency assessment. • Plant-to-plant variability in calculated LOCA frequency. • Frequency of double-ended-guillotine-break (DEGB). • Uncertainties in LOCA frequencies. • Risk management insights. - Abstract: As a tribute to the published work by S.H. Bush, S. Beliczey and H. Schulz, this paper assesses the progress with methods and techniques for quantifying the reliability of piping systems in commercial nuclear power plants on the basis of failure rate estimates derived from field experience data in combination with insights and results from probabilistic fracture mechanics analyses and expert elicitation exercises. This status assessment is made from a technical perspective obtained through development of location-specific loss-of-coolant-accident (LOCA) frequencies for input to risk-informed resolution of the generic safety issue (GSI) 191. The methods and techniques on which these GSI-191 applications are based build on a body of work developed by the authors during a period spanning more than two decades. The insights that are presented and discussed in this paper cover today’s knowledge base concerning how to utilize a risk-informed approach to the assessment of piping reliability in the context of probabilistic risk assessment (PRA) in general and the resolution of GSI-191 in particular. Specifically the paper addresses the extent to which LOCA frequencies vary from location to location within a reactor coolant system pressure boundary (RCPB) for a given plant as well as vary from plant to plant, and the reasons for these variabilities. Furthermore, the paper provides the authors’ perspectives on interpretations and applications of information extracted from an expert elicitation process to obtain LOCA frequencies as documented in NUREG-1829 and how to apply this information to GSI-191. Finally, this paper documents technical insights relative to mitigation of

Site restoration: Estimation of attributable costs from plutonium-dispersal accidents

International Nuclear Information System (INIS)

Chanin, D.I.; Murfin, W.B.

1996-05-01

A nuclear weapons accident is an extremely unlikely event due to the extensive care taken in operations. However, under some hypothetical accident conditions, plutonium might be dispersed to the environment. This would result in costs being incurred by the government to remediate the site and compensate for losses. This study is a multi-disciplinary evaluation of the potential scope of the post-accident response that includes technical factors, current and proposed legal requirements and constraints, as well as social/political factors that could influence decision making. The study provides parameters that can be used to assess economic costs for accidents postulated to occur in urban areas, Midwest farmland, Western rangeland, and forest. Per-area remediation costs have been estimated, using industry-standard methods, for both expedited and extended remediation. Expedited remediation costs have been evaluated for highways, airports, and urban areas. Extended remediation costs have been evaluated for all land uses except highways and airports. The inclusion of cost estimates in risk assessments, together with the conventional estimation of doses and health effects, allows a fuller understanding of the post-accident environment. The insights obtained can be used to minimize economic risks by evaluation of operational and design alternatives, and through development of improved capabilities for accident response

Site restoration: Estimation of attributable costs from plutonium-dispersal accidents

Energy Technology Data Exchange (ETDEWEB)

Chanin, D.I.; Murfin, W.B. [Technadyne Engineering Consultants, Inc., Albuquerque, NM (United States)

1996-05-01

A nuclear weapons accident is an extremely unlikely event due to the extensive care taken in operations. However, under some hypothetical accident conditions, plutonium might be dispersed to the environment. This would result in costs being incurred by the government to remediate the site and compensate for losses. This study is a multi-disciplinary evaluation of the potential scope of the post-accident response that includes technical factors, current and proposed legal requirements and constraints, as well as social/political factors that could influence decision making. The study provides parameters that can be used to assess economic costs for accidents postulated to occur in urban areas, Midwest farmland, Western rangeland, and forest. Per-area remediation costs have been estimated, using industry-standard methods, for both expedited and extended remediation. Expedited remediation costs have been evaluated for highways, airports, and urban areas. Extended remediation costs have been evaluated for all land uses except highways and airports. The inclusion of cost estimates in risk assessments, together with the conventional estimation of doses and health effects, allows a fuller understanding of the post-accident environment. The insights obtained can be used to minimize economic risks by evaluation of operational and design alternatives, and through development of improved capabilities for accident response.

Development of Safety Significance Evaluation Program for Accidents and Events in NPPs

International Nuclear Information System (INIS)

Yang, Hui Chang; Hong, Seok Jin; Cho, Nam Chul; Chung, Dae Wook; Lee, Chang Joo

2010-01-01

To evaluate the significance in terms of safety for the accidents and events occurred in nuclear power plants using probabilistic safety assessment techniques can provide useful insights to the regulator. Based on the quantified risk information of accident or event occurred, regulators can decide which regulatory areas should be focused than the others. To support these regulatory analysis activities, KINS-ASP program was developed. KINS-ASP program can supports the risk increase due to the occurred accidents or events by providing the graphic interfaces and linked quantification engines for the PSA experts and non- PSA acquainted regulators both

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

Predicting Consequences of Technological Disasters from Natural Hazard Events: Challenges and Opportunities Associated with Industrial Accident Data Sources

Science.gov (United States)

Wood, M.

2009-04-01

The increased focus on the possibility of technological accidents caused by natural events (Natech) is foreseen to continue for years to come. In this case, experts in prevention, mitigation and preparation activities associated with natural events will increasingly need to borrow data and expertise traditionally associated with the technological fields to carry out the work. An important question is how useful is the data for understanding consequences from such natech events. Data and case studies provided on major industrial accidents tend to focus on lessons learned for re-engineering the process. While consequence data are reported at least nominally in most reports, their precision, quality and completeness is often lacking. Consequences that are often or sometimes available but not provided can include severity and type of injuries, distance of victims from the source, exposure measurements, volume of the release, population in potentially affected zones, and weather conditions. Yet these are precisely the type of data that will aid natural hazard experts in land-use planning and emergency response activities when a Natech event may be foreseen. This work discusses the results of a study of consequence data from accidents involving toxic releases reported in the EU's MARS accident database. The study analysed the precision, quality and completeness of three categories of consequence data reported: the description of health effects, consequence assessment and chemical risk assessment factors, and emergency response information. This work reports on the findings from this study and discusses how natural hazards experts might interact with industrial accident experts to promote more consistent and accurate reporting of the data that will be useful in consequence-based activities.

Semi-portable whole body counter deployable during post-radiological accident

International Nuclear Information System (INIS)

Panchal, C.G.; Bansode, P.Y.; Vinod, M.; Sarade, Bhagyashree; Jain, R.K.; Jakati, R.K.; Pithawa, C.K.

2012-01-01

A versatile whole body counter with the state of art electronics has been indigenously designed and developed which can be readily deployed for use during nuclear emergency. This instrument is designed to quickly identify and quantify the activity of high energy photon emitters accumulated particularly in vital organs like thyroid, lungs besides the body of the victims of the radiological incident or accident. Special features of the instruments are swivel type detectors assembly mountable on a wall or table top and detachable collimator configurable to assess the internal contamination selectively to meet protective measures of the radiological accidents, mechanically rugged and functionally reliable to perform in contaminated environmental field conditions. (author)

Development of a national doctrine for the management of the post-accident phase of a radiological emergency situation

International Nuclear Information System (INIS)

Niel, J.Ch.; Godet, J.L.

2008-01-01

For several years, public Authorities have defined an organization for the management of emergency situations arising from an accident occurring at a nuclear installation. So far, the management of the risk arising from the post accident phase was, in itself, not explored with the same care. What so ever, no format policy on which the action of public Authority could be based is today available. The nuclear safety Authority (ASN), in relation with the other concerned departments, is now in charge, according to the above mentioned directive, to prepare and implement the necessary provisions to respond to a post accident situation. In dune 2005, ASN established the steering committee for the management of post nuclear or radiological emergency situations (CODIRPA). The definition of a national policy related to the management of the radiological risk during a post event situation having to integrate various organization aspects as: lifting of protection emergency provisions and rehabilitation of buildings, life in contaminated rural territories, agriculture and water, dose and radiological consequences, sanitary surveillance of victims and populations, indemnification, waste management of contaminated crops and soils, organization of public Authorities. During the 2. phase of CODIRPA work (2008-2009), the first elements of policy will be consolidated and new scenarios will be studied (one worsened scenario and one scenario with alpha emitting radionuclide). in parallel, a procedure for local actor's consultation should be elaborated. (authors)

Thermo-physical properties of corium: development of an assessed data base for severe accident applications

Energy Technology Data Exchange (ETDEWEB)

Strizhov, V.F.; Galimov, R.G.; Ozrin, V.D. [Nuclear Safety Institute of the Russian Academy of Sciences, Moscow (Russian Federation); Yu Zitserman, V.; Kobzev, G.I.; Fokin, L.R. [Institute of high temperatures, Russian Academy of Sciences, Moscow (Russian Federation); Piluso, P. [CEA Cadarache (DEN/DTN/STRI), Lab. d' essais pour la Maitrise des Accidents graves, 13 - Saint Paul lez Durance (France); Chalaye, H. [CEA Saclay, Dir. de l' Energie Nucleaire, 91 - Gif sur Yvette (France)

2007-07-01

In a hypothetical case of a core melt-down scenarios a very high temperature would be reached (up to 3000 K). In this case, the materials of the core and structural materials (fuel, cladding, metallic alloys, concrete, etc.) could melt to form complex and aggressive mixtures called corium. Modelling of severe accident phenomena, code development and assessments of nuclear safety require a reliable knowledge of the thermophysical properties of corium at wide temperature range (below solidus temperature, between solidus and liquidus temperature and above the liquidus temperature). Common Russian-French project ISTC 3078, has been devoted to the development, assessment and recommendation for the establishment of a reliable thermophysical data base for severe accident applications. The project consists of two tasks related to properties of pure metallic (U, Zr, Fe, Cr, Ni) and oxide (UO{sub 2}, U{sub 3}O{sub 8}, U{sub 4}O{sub 9}, NiO, ZrO{sub 2}, Cr{sub 2}O{sub 3}, FeO, Fe{sub 2}O{sub 3}, Fe{sub 3}O{sub 4}, Al{sub 2}O{sub 3}, CaO, MgO, SiO{sub 2}, HfO{sub 2}, CeO{sub 2}) components, and mixtures relevant to severe accident conditions. Three categories of data (on UPAK classification) were considered: experimental data, critically evaluated data, and predicted data. The data of the first category is a result of specific experiment, data of the second category is a result of the analysis of data consistency and co-processing (expert and statistical) obtained in several experiments, data of the third category are based on model estimates, using correlations between different physical properties. The process of assessing, review and development of recommendation is described in the paper and illustrated by examples on thermophysical properties. (authors)

GOTHIC analysis of post-accident hydrogen mixing behaviour in CANDU fuelling machine vault

International Nuclear Information System (INIS)

Yim, K.; Wong, R.C.; Fluke, R.J.

1997-01-01

The GOTHIC code was used to assess the post-accident hydrogen gas mixing patterns in a CANDU reactor containment and demonstrate the acceptability of Ontario Hydro Nuclear's hydrogen mitigation methods. The fuelling machine vault, being a small volume room containing major reactor piping, is the room of most concern with respect to hydrogen concentrations. Detailed three dimensional modelling of the gas mixing patterns in the fuelling machine vault was completed. Results showed that, even without forced air circulation, there is enough dispersion of hydrogen to other parts of containment to preclude the build-up of sensitive mixtures in the vault. For a brief time during the peak period of hydrogen release, hydrogen concentrations rise to close to the lower ignition limit in a small portion of the vault, but these hydrogen-steam-air mixtures are considered acceptable. Natural mixing alone is sufficient to preclude damaging hydrogen burns. (author)

Lessons learned from on-site safety assessments performed by DOE in response to the Tomsk accident

International Nuclear Information System (INIS)

Witmer, F.E.

1995-01-01

In response to the accident, in April 1993, at the nuclear fuel reprocessing plant of the Siberian chemical Combine, Tomsk, Russia, the U.S. Department of Energy (DOE) initiated concurrent efforts to understand the causes for the accident and to review potential vulnerabilities for similar occurrences across the DOE radiochemical complex. Because the accident occurred in the feed adjustment stage of a Purex type process, US facilities which contained significant inventories of TBP, organic diluent and nitric acid were evaluated by expert teams. From accident conditions, prior experience, modeling and experimental programs and confirmatory dialogue with the Russians, enhanced understanding was achieved and vulnerabilities (e.g., lack of safety analysis, organic layering, inadvertent acid addition, use of aromatic diluents, uncertain venting capability, no mitigative/emergency procedures, etc.) were identified and corrected

Post-accident TMI-2 (Three Mile Island-Unit 2) decontamination and defueling

International Nuclear Information System (INIS)

Hofstetter, K.J.; Baston, V.F.

1986-01-01

Following the accident at Three Mile Island-Unit 2 (TMI-2), a substantial quantity of fission products was distributed throughout various plant systems and areas. The control of further migration of these radionuclides was accomplished by various physical and chemical means. The decontamination and defueling activities have proceeded within specific regulatory, administrative, and hardware restrictions. A summary of the post-accident status of the plant systems, the decontamination methods used, and the end-point criteria will be discussed. The hardware installed or utilized to perform the cleanup operations will be described. The methods and progress of defueling will also be presented. The development of detailed water chemistry requirements and their effects on systems and decontamination efforts are discussed. The planning, scheduling, and performance of specific recovery tasks are presented along with a general overview of water and chemical management at TMI-2

Expert assessments of retrofitting coal-fired power plants with carbon dioxide capture technologies

International Nuclear Information System (INIS)

Chung, Timothy S.; Patino-Echeverri, Dalia; Johnson, Timothy L.

2011-01-01

A set of 13 US based experts in post-combustion and oxy-fuel combustion CO 2 capture systems responded to an extensive questionnaire asking their views on the present status and future expected performance and costs for amine-based, chilled ammonia, and oxy-combustion retrofits of coal-fired power plants. This paper presents the experts' responses for technology maturity, ideal plant characteristics for early adopters, and the extent to which R and D and deployment incentives will impact costs. It also presents the best estimates and 95% confidence limits of the energy penalties associated with amine-based systems. The results show a general consensus that amine-based systems are closer to commercial application, but potential for improving performance and lowering costs is limited; chilled ammonia and oxy-combustion offer greater potential for cost reductions, but not without greater uncertainty regarding scale and technical feasibility. - Highlights: → Study presents experts' views on CCS retrofit costs and performance. → Experts commented on amine-based systems, chilled ammonia, and oxy-fuel combustion. → Estimates of future energy penalty show uncertainty for the three technologies. → These estimates under an aggressive RD and D policy scenario narrow significantly. → The experts' judgments support the need for enhanced RD and D for post-combustion CCS.

Expert assessments of retrofitting coal-fired power plants with carbon dioxide capture technologies

Energy Technology Data Exchange (ETDEWEB)

Chung, Timothy S. [Eastern Research Group, Inc., 110 Hartwell Avenue 1, Lexington, MA 02421-3134l (United States); Patino-Echeverri, Dalia, E-mail: dalia.patino@duke.edu [Nicholas School of the Environment, Box 90328, Duke University, Durham NC 27708 (United States); Johnson, Timothy L. [Nicholas School of the Environment, Box 90328, Duke University, Durham NC 27708 (United States)

2011-09-15

A set of 13 US based experts in post-combustion and oxy-fuel combustion CO{sub 2} capture systems responded to an extensive questionnaire asking their views on the present status and future expected performance and costs for amine-based, chilled ammonia, and oxy-combustion retrofits of coal-fired power plants. This paper presents the experts' responses for technology maturity, ideal plant characteristics for early adopters, and the extent to which R and D and deployment incentives will impact costs. It also presents the best estimates and 95% confidence limits of the energy penalties associated with amine-based systems. The results show a general consensus that amine-based systems are closer to commercial application, but potential for improving performance and lowering costs is limited; chilled ammonia and oxy-combustion offer greater potential for cost reductions, but not without greater uncertainty regarding scale and technical feasibility. - Highlights: > Study presents experts' views on CCS retrofit costs and performance. > Experts commented on amine-based systems, chilled ammonia, and oxy-fuel combustion. > Estimates of future energy penalty show uncertainty for the three technologies. > These estimates under an aggressive RD and D policy scenario narrow significantly. > The experts' judgments support the need for enhanced RD and D for post-combustion CCS.

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

Accident frequency and unrealistic optimism: Children's assessment of risk.

Science.gov (United States)

Joshi, Mary Sissons; Maclean, Morag; Stevens, Claire

2018-02-01

Accidental injury is a major cause of mortality and morbidity among children, warranting research on their risk perceptions. Three hundred and seven children aged 10-11 years assessed the frequency, danger and personal risk likelihood of 8 accidents. Two social-cognitive biases were manifested. The frequency of rare accidents (e.g. drowning) was overestimated, and the frequency of common accidents (e.g. bike accidents) underestimated; and the majority of children showed unrealistic optimism tending to see themselves as less likely to suffer these accidents in comparison to their peers, offering superior skills or parental control of the environment as an explanation. In the case of pedestrian accidents, children recognised their seriousness, underestimated the frequency of this risk and regarded their own road crossing skill as protection. These findings highlight the challenging task facing safety educators who, when teaching conventional safety knowledge and routines, also need to alert children to the danger of over-confidence without disabling them though fear. Copyright © 2017 Elsevier Ltd. All rights reserved.

The Development of Marine Accidents Human Reliability Assessment Approach: HEART Methodology and MOP Model

OpenAIRE

Ludfi Pratiwi Bowo; Wanginingastuti Mutmainnah; Masao Furusho

2017-01-01

Humans are one of the important factors in the assessment of accidents, particularly marine accidents. Hence, studies are conducted to assess the contribution of human factors in accidents. There are two generations of Human Reliability Assessment (HRA) that have been developed. Those methodologies are classified by the differences of viewpoints of problem-solving, as the first generation and second generation. The accident analysis can be determined using three techniques of analysis; sequen...

Chernobyl accident: Assessing the data

Energy Technology Data Exchange (ETDEWEB)

Soerensen, B

1986-01-01

Data presented in the official Soviet report to the IAEA on the Chernobyl reactor accident are critically assessed. Special attention is given to the derivation of release fractions from fallout measurements, a procedure which is demonstrated to involve large elements of uncertainty. Further comments relate to estimates of plume rise and deposition velocity. A comparison is made with the predictions of previously published theoretical reactor safety studies.

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.)

JAERI's activities in JCO accident

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-09-01

The Japan Atomic Energy Research Institute (JAERI) was actively involved in a variety of technical supports and cooperative activities, such as advice on terminating the criticality condition, contamination checks of the residents and consultation services for the residents, as emergency response actions to the criticality accident at the uranium processing facility operated by the JCO Co. Ltd., which occurred on September 30, 1999. These activities were carried out in collaborative ways by the JAERI staff from the Tokai Research Establishment, Naka Fusion Research Establishment, Oarai Research Establishment, and Headquarter Office in Tokyo. As well, the JAERI was engaged in the post-accident activities such as identification of accident causes, analyses of the criticality accident, and dose assessment of exposed residents, to support the Headquarter for Accident Countermeasures of the Science and Technology Agency (STA), the Accident Investigation Committee and the Health Control Committee of the Nuclear Safety Commission of Japan (NSC). This report compiles the activities, that the JAERI has conducted to date, including the discussions on measures for terminating the criticality condition, evaluation of the fission number, radiation monitoring in the environment, dose assessment, analyses of criticality dynamics. (author)

How does the medical graduates' self-assessment of their clinical competency differ from experts' assessment?

Science.gov (United States)

2013-01-01

Background The assessment of the performance of medical school graduates during their first postgraduate years provides an early indicator of the quality of the undergraduate curriculum and educational process. The objective of this study was to assess the clinical competency of medical graduates, as perceived by the graduates themselves and by the experts. Methods This is a hospital based cross-sectional study. It covered 105 medical graduates and 63 experts selected by convenient sampling method. A self-administered questionnaire covering the different areas of clinical competency constructed on a five-point Likert scale was used for data collection. Data processing and analysis were performed using the Statistical Package for Social Science (SPSS) 16.0. The mean, frequency distribution, and percentage of the variables were calculated. A non-parametric Kruskal Wallis test was applied to verify whether the graduates' and experts' assessments were influenced by the graduates' variables such as age, gender, experience, type of hospital, specialty and location of work at a (p ≤ 0.05) level of significance. Results The overall mean scores for experts' and graduates' assessments were 3.40 and 3.63, respectively (p= 0.035). Almost 87% of the graduates perceived their competency as good and very good in comparison with only 67.7% by experts. Female and male graduates who rated themselves as very good were 33.8% and 25% respectively. More than 19% of the graduates in the age group > 30 years perceived their clinical competency as inadequate in contrast with only 6.2% of the graduates in the youngest age group. Experts rated 40% of the female graduates as inadequate versus 20% of males, (p= 0.04). More than 40% of the graduates in younger age group were rated by experts as inadequate, versus 9.7% of the higher age group >30 years (p = 0.03). Conclusion There was a wide discrepancy between the graduates' self-assessment and experts' assessment, particularly in the level

77 FR 61446 - Proposed Revision Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors

Science.gov (United States)

2012-10-09

... Severe Accident Evaluation for New Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Standard... its Standard Review Plan (SRP), Section 19.0, ``Probabilistic Risk Assessment and Severe Accident... assessment (PRA) information and severe accident assessments for new reactors submitted to support design...

Assessment of geomechanical properties of intact Opalinus Clay - Expert report

International Nuclear Information System (INIS)

Amann, F.; Vogelhuber, M.

2015-11-01

This comprehensive report published by the Swiss Federal Nuclear Safety Inspectorate ENSI presents an expert report published on the assessment of the geomechanical properties of intact Opalinus Clay. This review report addresses the conceptual constitutive framework for repositories in Opalinus Clay. The author addresses the geomechanical fundamentals that are necessary in order to adequately judge experiments on intact Opalinus Clay and the interpretation of the results. The report assesses in detail the various test series on intact Opalinus Clay carried out along with the interpretations made by experts and NAGRA. Further assessments are quoted including those on sample geometries tested, effective strength properties, undrained shear strength properties and elastic properties. The results of work done by other experts are also presented and discussed. The report is completed with a list of relevant literature

Assessment of geomechanical properties of intact Opalinus Clay - Expert report

Energy Technology Data Exchange (ETDEWEB)

Amann, F. [Eidgenössische Technische Hochschule ETHZ, Zürich (Switzerland); Vogelhuber, M. [Dr. von Moos AG, Geotechnisches Büro, Zürich (Switzerland)

2015-11-15

This comprehensive report published by the Swiss Federal Nuclear Safety Inspectorate ENSI presents an expert report published on the assessment of the geomechanical properties of intact Opalinus Clay. This review report addresses the conceptual constitutive framework for repositories in Opalinus Clay. The author addresses the geomechanical fundamentals that are necessary in order to adequately judge experiments on intact Opalinus Clay and the interpretation of the results. The report assesses in detail the various test series on intact Opalinus Clay carried out along with the interpretations made by experts and NAGRA. Further assessments are quoted including those on sample geometries tested, effective strength properties, undrained shear strength properties and elastic properties. The results of work done by other experts are also presented and discussed. The report is completed with a list of relevant literature.

Assessment of risk of accident at work as an indicator of safe behaviour of workers

Energy Technology Data Exchange (ETDEWEB)

Pisiewicz, K

1978-10-01

In 1977 the Psychology and Sociology Research Development Unit of the Central Mining Institute carried out research on the influence of assessment of the accident risk on the safe behaviour of workers. 450 workers employed at the longwall faces in 6 coal mines with various accident rates were questioned. It was found that a low assessment of risk favours hazardous operations, contrary to the principles of work safety, while a high assessment of the risk does not favour hazardous operations. Miners employed in coal mines with high accident rates tend to a low assessment of accident risk (arithmetic mean x 48.54) in comparison to miners from mines with low accident rates (arithmetic mean x 53.68). It was also found that the arithmetic mean of assessment of risks among workers who had had an accident at work is lower (x 50.3) than among workers who had not yet had an accident at work (x 55.32).

French regulatory requirements for the occupational radiation protection in severe accident situations and post-accident recovery

International Nuclear Information System (INIS)

Couasnon, Olivier

2014-01-01

Workers of the concerned company and other persons and teams called 'intervention personnel' (specialized firemen, first aider, etc.) are to be involved in radiological emergency situations. Radiation protection provisions for workers and for intervention personnel complement one another because they cover persons with different statutes (workers under the responsibility of an employer and persons acting within the framework of agreements with the public authorities or within the framework of the requisitions). Work or operations exposing workers to ionizing radiation in radiological emergency situations can be assigned only to workers satisfying all of the following conditions: classification in category A worker; free of any medical unfitness; on a list drawn up in advance for this purpose; having received appropriate information on the risks and the precautions to take during the work or the operation; not having received, during the preceding twelve months, a dose greater than one of the annual limit values for exposures subject to special authorization. In addition, the worker must be a volunteer to carry out the work or the operations concerned in radiological emergency situations and have individual dosimetry means appropriate for the situation. Intervention personnel are possibly composed of personnel from responding organizations, such as police officers, fire-fighters, medical personnel, drivers and crews of evacuation vehicles, or of workers employed by the head of the damaged plant. In order to determine their selection, training and medical and radiological monitoring conditions, intervention personnel are classified into two groups: personnel forming the special technical, medical and health intervention teams readied in advance to deal with radiological emergency situations and persons not belonging to special teams but intervening as part of the tasks within the scope of their competence. In case of an existing exposure situation (post-accident

Accident diagnosis, recovery, and prognosis aid

International Nuclear Information System (INIS)

Touchton, R.A.

1987-01-01

This paper describes an investigation that was conducted to assess and demonstrate the feasibility of using artificial intelligence (AI) techniques to develop an expert system serving as a nuclear plant accident diagnosis, recovery, and prognosis aid. This effort was sponsored under a contract with the Department of Energy as a part of their Small Business Innovation Research Program. The interest in such a system is based upon on-going industry and regulatory commitment to improved nuclear plant performance and safety

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

Assessment and management of agitation in psychiatry: Expert consensus.

Science.gov (United States)

Garriga, Marina; Pacchiarotti, Isabella; Kasper, Siegfried; Zeller, Scott L; Allen, Michael H; Vázquez, Gustavo; Baldaçara, Leonardo; San, Luis; McAllister-Williams, R Hamish; Fountoulakis, Konstantinos N; Courtet, Philippe; Naber, Dieter; Chan, Esther W; Fagiolini, Andrea; Möller, Hans Jürgen; Grunze, Heinz; Llorca, Pierre Michel; Jaffe, Richard L; Yatham, Lakshmi N; Hidalgo-Mazzei, Diego; Passamar, Marc; Messer, Thomas; Bernardo, Miquel; Vieta, Eduard

2016-01-01

Psychomotor agitation is associated with different psychiatric conditions and represents an important issue in psychiatry. Current recommendations on agitation in psychiatry are not univocal. Actually, an improper assessment and management may result in unnecessary coercive or sedative treatments. A thorough and balanced review plus an expert consensus can guide assessment and treatment decisions. An expert task force iteratively developed consensus using the Delphi method. Initial survey items were based on systematic review of the literature. Subsequent surveys included new, re-worded or re-rated items. Out of 2175 papers assessing psychomotor agitation, 124 were included in the review. Each component was assigned a level of evidence. Integrating the evidence and the experience of the task force members, a consensus was reached on 22 statements on this topic. Recommendations on the assessment of agitation emphasise the importance of identifying any possible medical cause. For its management, experts agreed in considering verbal de-escalation and environmental modification techniques as first choice, considering physical restraint as a last resort strategy. Regarding pharmacological treatment, the "ideal" medication should calm without over-sedate. Generally, oral or inhaled formulations should be preferred over i.m. routes in mildly agitated patients. Intravenous treatments should be avoided.

Comparison of accident risks in different energy systems: Comments from Russian specialists

International Nuclear Information System (INIS)

2000-01-01

Many articles on accident risk analysis of different energy systems in comparison with nuclear power share certain stereotypical features. For example: When assessing the risks associated with the operation of such facilities, they ignore the effects of the upgrading of RBMK reactors which was carried out after the Chernobyl accident. In their integrated assessment of the radiological consequences of the Chernobyl accident they use numerous studies which frequently contain unreliable source data and unfounded predictions, and they ignore many socio-political factors which considerably increased the damage caused by the accident. Unfortunately, the study in question, despite its topicality and originality of approach, is also not without such shortcomings. After the Chernobyl accident, reconstruction and safety enhancement measures were implemented at nuclear power plants with RBMK reactors which were without precedent in world practice and have continued to this day. According to probabilistic safety assessments (PSA) carried out with the assistance of international experts, the probability of serious accidents at RBMKs has decreased by a factor of two or more thanks to the above mentioned measures. The mean weighted safety index for all operational RBMK reactors is 10 -4 l/year and is decreasing thanks to the ongoing and planned reconstruction of all units. All operational nuclear power plants with RBMK reactors are thus on a par with the successfully operating Soviet WWERs and western boiling water reactors (BWRs) and pressurized water reactors (PWRs), and satisfy the IAEA recommendations regarding the risk level of older generation nuclear power plants. The authors of the IAEA Bulletin article give estimates of the remote radiological consequences of the Chernobyl accident which range from an estimated 10,000 to 30,000 fatal cases of radiation-induced cancer, and the literature on the subject contains even more extreme estimates. However, our 14 years

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

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

A radioecological risk assessment tool for post-accidental situations. Application in the Toulon marine area (South of France)

International Nuclear Information System (INIS)

Duffa, Celine; Thebault, Herve

2010-01-01

To estimate any post-accidental consequences in the environment and to propose adapted management plans, data concerning both radionuclides spatial dispersion and environmental sensitivity to this contamination are necessary. IRSN develops integrated tools to support experts and decision makers in post-accident situation concerning this area. This management tool aims to combine radionuclides dispersion modelling and radioecological sensitivity local maps to outline vulnerable areas front of a known release. An example of application to a maritime environment is given: the naval base of Toulon. Calculation of radionuclides dispersion is based on the existing MARS 3D hydrodynamic model. The radioecological sensitivity of the marine environment is based on the ecosystem intrinsic characteristics and the socio-economical resources. Mapping radionuclides dispersion results and sensitivity of defined zones of the studies area, we show that the Little Bay and the western part of Toulon marine area are the most vulnerable in case of accidental release occuring in the harbour. (orig.)

Identification and assessment of BWR in-vessel severe accident mitigation strategies

Energy Technology Data Exchange (ETDEWEB)

Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M. [Oak Ridge National Lab., TN (United States)

1992-10-01

This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners` Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored.

Identification and assessment of BWR in-vessel severe accident mitigation strategies

International Nuclear Information System (INIS)

Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M.

1992-10-01

This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners' Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored

Chinese academic experts' assessment for forest bio-energy development in China

International Nuclear Information System (INIS)

Qu Mei; Ahponen, Pirkkoliisa; Tahvanainen, Liisa; Pelkonen, Paavo

2010-01-01

The aim of this study was to assess the current situation of the forest bio-energy development in China. This assessment is based on opinions of Chinese academic experts. Key drivers and uncertainties regarding the implementation, and the strategies for the future practices in the development of forest bio-energy were investigated. In addition, the purpose of this study was also to determine whether there is a consensus among the experts concerning forest bio-energy and if this consensus agrees with policy-makers in China. A thorough assessment was conducted using a two-round Delphi survey of sixty-one bio-energy experts in China. The results revealed the advantages, potential problems, and the experts' recommendations for the future development. Furthermore, the experts agreed that the Chinese government plays a dominant role in the development process of forest bio-energy in the country. The experts recognized that the process of developing forest bio-energy is a challenging task both domestically and globally. At the same time they also highlighted the potential benefits of developing forest bio-energy in China during the next ten years. The outcomes of this study could be used to give advice to policy-makers and to support the implementation of the future forest bio-energy policies in China.

Probabilistic accident consequence uncertainty analysis: Dispersion and deposition uncertainty assessment, appendices A and B

International Nuclear Information System (INIS)

Harper, F.T.; Young, M.L.; Miller, L.A.; Hora, S.C.; Lui, C.H.; Goossens, L.H.J.; Cooke, R.M.; Paesler-Sauer, J.; Helton, J.C.

1995-01-01

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, completed in 1990, estimate the risks presented by nuclear installations based on postulated frequencies and magnitudes of potential accidents. In 1991, the US Nuclear Regulatory Commission (NRC) and the Commission of the European Communities (CEC) began a joint uncertainty analysis of the two codes. The objective was to develop credible and traceable uncertainty distributions for the input variables of the codes. Expert elicitation, developed independently, was identified as the best technology available for developing a library of uncertainty distributions for the selected consequence parameters. The study was formulated jointly and was limited to the current code models and to physical quantities that could be measured in experiments. To validate the distributions generated for the wet deposition input variables, samples were taken from these distributions and propagated through the wet deposition code model along with the Gaussian plume model (GPM) implemented in the MACCS and COSYMA codes. Resulting distributions closely replicated the aggregated elicited wet deposition distributions. Project teams from the NRC and CEC cooperated successfully to develop and implement a unified process for the elaboration of uncertainty distributions on consequence code input parameters. Formal expert judgment elicitation proved valuable for synthesizing the best available information. Distributions on measurable atmospheric dispersion and deposition parameters were successfully elicited from experts involved in the many phenomenological areas of consequence analysis. This volume is the second of a three-volume document describing the project and contains two appendices describing the rationales for the dispersion and deposition data along with short biographies of the 16 experts who participated in the project

Probabilistic accident consequence uncertainty analysis: Dispersion and deposition uncertainty assessment, appendices A and B

Energy Technology Data Exchange (ETDEWEB)

Harper, F.T.; Young, M.L.; Miller, L.A. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States); Lui, C.H. [Nuclear Regulatory Commission, Washington, DC (United States); Goossens, L.H.J.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Paesler-Sauer, J. [Research Center, Karlsruhe (Germany); Helton, J.C. [and others

1995-01-01

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, completed in 1990, estimate the risks presented by nuclear installations based on postulated frequencies and magnitudes of potential accidents. In 1991, the US Nuclear Regulatory Commission (NRC) and the Commission of the European Communities (CEC) began a joint uncertainty analysis of the two codes. The objective was to develop credible and traceable uncertainty distributions for the input variables of the codes. Expert elicitation, developed independently, was identified as the best technology available for developing a library of uncertainty distributions for the selected consequence parameters. The study was formulated jointly and was limited to the current code models and to physical quantities that could be measured in experiments. To validate the distributions generated for the wet deposition input variables, samples were taken from these distributions and propagated through the wet deposition code model along with the Gaussian plume model (GPM) implemented in the MACCS and COSYMA codes. Resulting distributions closely replicated the aggregated elicited wet deposition distributions. Project teams from the NRC and CEC cooperated successfully to develop and implement a unified process for the elaboration of uncertainty distributions on consequence code input parameters. Formal expert judgment elicitation proved valuable for synthesizing the best available information. Distributions on measurable atmospheric dispersion and deposition parameters were successfully elicited from experts involved in the many phenomenological areas of consequence analysis. This volume is the second of a three-volume document describing the project and contains two appendices describing the rationales for the dispersion and deposition data along with short biographies of the 16 experts who participated in the project.

Psychological Intervention for Post-traumatic Stress Disorder among Witnesses of a Fatal Industrial Accident in a Workers' Health Center

Directory of Open Access Journals (Sweden)

Dong-Mug Kang

2017-12-01

Full Text Available Post-traumatic stress disorder (PTSD is a serious problem not only among workers who experience industrial accidents but also among workers who witness such accidents. Early intervention is needed to prevent prolonged psychological problems. There has been no study conducted regarding the psychological problems of and interventions for bystander workers in Korea. This study introduces the experience of intervention on psychological problems at the Busan Workers' Health Center workers who witnessed their colleagues' death. An investigation and an intervention were conducted according to the Korean Occupational Safety and Health Agency (KOSHA Guide. In total, 21 individuals including indirect observers showed statistical differences on scores of the Impact Event Scale Revised and the Patient Health Questionnaire 9 after the intervention. Future interventions and research involving a larger sample size over a longer period are needed. The KOSHA Guide could be a useful tool for urgent psychological intervention in the event of major workplace disasters. Keywords: industrial accident, post-traumatic stress disorder, witness, workers' health center

Preliminary neutronic assessment for ATF (Accident Tolerant Fuel) based on iron alloy

International Nuclear Information System (INIS)

Abe, Alfredo; Carluccio, Thiago; Piovezan, Pamela; Giovedi, Claudia; Martins, Marcelo R.

2015-01-01

After Fukushima Daiichi nuclear accident in 2011, the nuclear fuel performance under accident condition became a very important issue and currently different research and development program are in progress toward to reliability and withstand under accident condition. These initiatives are known as ATF (Accident Tolerant Fuel) R and D program, which many countries with different research institutes, fuel vendors and others are nowadays involved. Accident Tolerant Fuel (ATF) can be defined as enhanced fuel which can tolerate loss of active cooling system capability for a considerably longer time period and the fuel/cladding system can be maintained without significant degradation and can also improve the fuel performance during normal operations and transients, as well as design-basis accident (DBA) and beyond design-basis (BDBA) accident. Different materials have being proposed as fuel cladding candidates considering thermo-mechanical properties and lower reaction kinetic with steam and slower hydrogen production. The aim of this work is to perform a neutronic assessment for several cladding candidates based on iron alloy considering a standard PWR fuel rod (fuel pellet and dimension). The purpose of the assessment is to address different parameters that might contribute for possible neutronic reactivity gain in order to overcome the penalty due to increase of neutron absorption in the cladding materials. All the neutronic assessment is performed using MCNP, Monte Carlo code. (author)

Review of severe accidents and the results of accident consequence assessment in different energy systems (Contract research)

International Nuclear Information System (INIS)

Matsuki, Yoshio; Muramatsu, Ken

2008-05-01

The cases of severe accidents and the consequence assessments in different energy systems, Coal, Oil, Gas, Hydro and Nuclear, were collected, and then they were further analyzed. In this report, the information on the accidents in various energy systems were collected from the sources of the Paul Scherrer Institute (hereinafter, 'PSI') and the International Atomic Energy Agency (hereinafter, 'IAEA'). The information on the severe accidents of nuclear power plants were collected from the report of the US Presidential Commission on Catastrophic Nuclear Accidents and several relevant reports issued in the countries of the European Union, together with the reports of the PSI and the IAEA. To analyze the collected information, several parameters, which are numbers of fatalities, injuries, evacuees and the costs of the damages, were chosen to characterize those accidents in different energy systems. And then, upon the comparison of these characteristics of different accidents, the impacts of the accidents in nuclear and other energy systems were compared. Upon the results of the analysis, it is pointed out that the cost caused by the Chernobyl Accident, the severe accident in nuclear energy, tends to be higher than in the other energy systems. On the other hand, from the aspects of fatalities and injuries, it is not confirmed that the damages of the Chernobyl Accident are larger than in the other energy systems. However, it is also recognized, as the specific characteristics of the severe nuclear accident, that the impacts of the accident spread in a wider area, and stay for a longer period, in comparison with the ones in the other energy systems. (author)

Complex method to calculate objective assessments of information systems protection to improve expert assessments reliability

Science.gov (United States)

Abdenov, A. Zh; Trushin, V. A.; Abdenova, G. A.

2018-01-01

The paper considers the questions of filling the relevant SIEM nodes based on calculations of objective assessments in order to improve the reliability of subjective expert assessments. The proposed methodology is necessary for the most accurate security risk assessment of information systems. This technique is also intended for the purpose of establishing real-time operational information protection in the enterprise information systems. Risk calculations are based on objective estimates of the adverse events implementation probabilities, predictions of the damage magnitude from information security violations. Calculations of objective assessments are necessary to increase the reliability of the proposed expert assessments.

A simple assessment scheme for severe accident consequences using release parameters

Energy Technology Data Exchange (ETDEWEB)

Silva, Kampanart, E-mail: kampanarts@tint.or.th [Thailand Institute of Nuclear Technology, 16 Vibhavadi-Rangsit Rd., Latyao, Chatuchak, 10900 (Thailand); Okamoto, Koji [The University of Tokyo, 7-3-1 Hongo, Bunkyo, Tokyo 113-8654 (Japan)

2016-08-15

Highlights: • Nuclear accident consequence index can assess overall consequences of an accident. • Correlations between the index and release parameters are developed. • Relation between the index and release amount follows power function. • The exponent of the power function is the key to the relation. - Abstract: Nuclear accident consequence index (NACI) which can assess the overall consequences of a severe accident on people and the environment is developed based on findings from previous studies. It consists of three indices: radiation effect index, relocation index and decontamination index. Though the NACI can cover large range of consequences, its assessment requires extensive resources. The authors then attempt to simplify the assessment, by investigating the relations between the release parameters and the NACI, in order to use the release parameters for severe accident consequence assessment instead of the NACI. NACI and its components increase significantly when the release amount is increased, while the influences of the release period and the release starting time on the NACI are nearly negligible. Relations between the release amount and the NACI and its components follow simple power functions (y = ax{sup b}). The exponent of the power functions seems to be the key to the relations. The exponent of the relation between the release amount and the NACI was around 0.8–1.0 when the release amount is smaller than 100 TBq, and it increased to around 1.3–1.4 when the release amount is equal to or larger than 100 TBq.

Assessment of Equipment Capability to Perform Reliably under Severe Accident Conditions

International Nuclear Information System (INIS)

2017-07-01

The experience from the last 40 years has shown that severe accidents can subject electrical and instrumentation and control (I&C) equipment to environmental conditions exceeding the equipment’s original design basis assumptions. Severe accident conditions can then cause rapid degradation or damage to various degrees up to complete failure of such equipment. This publication provides the technical basis to consider when assessing the capability of electrical and I&C equipment to perform reliably during a severe accident. It provides examples of calculation tools to determine the environmental parameters as well as examples and methods that Member States can apply to assess equipment reliability.

Investigations of Human and Organizational Factors in hazardous vapor accidents

International Nuclear Information System (INIS)

Wang Yanfu; Faghih Roohi, Shahrzad; Hu Xiuming; Xie Min

2011-01-01

Highlights: → HFACS provides a systematic guideline in accident investigations. The hierarchal structure of HFACS forces investigators to seek out latent HOFs. → Bayesian Network enhances the ability of the HFACS by allowing experts to quantify the degree of relationships among the HOFs. → The fuzzy AHP helps to reduce the subjective biases by avoiding the need to give explicit probability values for the variables' states. - Abstract: This paper presents a model to assess the contribution of Human and Organizational Factor (HOF) to accidents. The proposed model is made up of two phases. The first phase is the qualitative analysis of HOF responsible for accidents, which utilizes Human Factors Analysis and Classification System (HFACS) to seek out latent HOFs. The hierarchy of HOFs identified in the first phase provides inputs for the analysis in the second phase, which is a quantitative analysis using Bayesian Network (BN). BN enhances the ability of HFACS by allowing investigators or domain experts to measure the degree of relationships among the HOFs. In order to estimate the conditional probabilities of BN, fuzzy analytical hierarchy process and decomposition method are applied in the model. Case studies show that the model is capable of seeking out critical latent human and organizational errors and carrying out quantitative analysis of accidents. Thereafter, corresponding safety prevention measures are derived.

Health-related quality of life and post-traumatic stress disorder symptoms in accident and emergency attenders suffering from psychosocial crises: a longitudinal study

Science.gov (United States)

Senneseth, Mette; Alsaker, Kjersti; Natvig, Gerd Karin

2012-01-01

Aims This paper is a report of a study of health-related quality of life and post-traumatic stress disorder symptoms in patients attending an Accident and Emergency department because of psychosocial crises. Background Psychosocial crises are commonplace globally, but there is little knowledge about patients attending Accident and Emergency departments because of psychosocial crises. Methods Data were collected at an Accident and Emergency department in Norway from September 2008 to June 2009. A total of 99 adults participated in the baseline study and 41 of these participated at 2 months follow-up. The Short Form-36 Health Survey and the Post Traumatic Symptom Scale were used to obtain data. Findings Participants reported significantly lower scores in all health-related quality of life domains at baseline compared with the general Norwegian population. The mental health score was two standard deviations below the norm. Health-related quality of life scores were improved and post-traumatic stress disorder symptoms were reduced after 2 months. High levels of post-traumatic stress disorder symptoms were reported by 78% of the participants at baseline and 59% at follow-up. Participants with high levels of post-traumatic stress disorder symptoms at follow-up also reported low health-related quality of life scores. Conclusion This study suggests a need for an acute psychosocial intervention and an opportunity to receive follow-up support at Accident and Emergency departments. PMID:21740459

Assessment of WWER fuel condition in design basis accident

International Nuclear Information System (INIS)

Bibilashvili, Yu.; Sokolov, N.; Andreeva-Andrievskaya, L.; Vlasov, Yu.; Nechaeva, O.; Salatov, A.

1994-01-01

The fuel behaviour in design basis accidents is assessed by means of the verified code RAPTA-5. The code uses a set of high temperature physico-chemical properties of the fuel components as determined for commercially produced materials, fuel rod simulators and fuel rod bundles. The WWER fuel criteria available in Russia for design basis accidents do not generally differ from the similar criteria adopted for PWR's. 12 figs., 11 refs

Assessment of PASS Effectiveness under Severe Accidents in Nuclear Power Plants

International Nuclear Information System (INIS)

Choi, Yu Jung; Lee, Sung Bok; Kim, Hyeong Taek; Lee, Jin Yong

2008-01-01

Following the accident at Three Mile Island Unit 2 (TMI-2) on March 28, 1979, the USNRC formed a lessons-learned Task Force to identify and evaluate safety concerns originating with the TMI-2 accident. NUREG-0578 documented the results of the task force effort. One of the recommendations of the task force was for licensees to upgrade the capability to obtain samples from the reactor coolant system and containment atmosphere under high radioactivity conditions and to provide the capability for chemical and spectral analyses of high-level samples on site. NUREG-0737 contained the details of the TMI recommendations that were to be implemented by the licensees. Additional criteria for post accident sampling system(PASS) were issued by Regulatory Guide 1.97. As the results, PASS has been installed on nuclear power plants(NPPs) in Korea as well as United States. However, significant improvements have been achieved since the TMI-2 accident in the areas of understanding risks associated with nuclear plant operations and developing better strategies for managing the response to potential severe accidents at NPPs. Thus, the requirements for PASS have been re-evaluated in some reports. According to the reports, the samples and measurements from PASS do not contribute significantly to emergency management response to severe accidents due to the long analyzing time, 3 hours. Hence, this paper focused on the development of the quantitative analysis methodology to analyze the sequence of the severe accident in Yonggwang nuclear power plants (YGN) and presented the results of the analysis according to the developed methodology

77 FR 66649 - Proposed Revision to Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors

Science.gov (United States)

2012-11-06

... and Severe Accident Evaluation for New Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Assessment and Severe Accident Evaluation for New Reactors.'' The NRC is extending the public comment period... assessment (PRA) information and severe accident assessments for new reactors submitted to support design...

Risk assessment of maintenance operations: the analysis of performing task and accident mechanism.

Science.gov (United States)

Carrillo-Castrillo, Jesús A; Rubio-Romero, Juan Carlos; Guadix, Jose; Onieva, Luis

2015-01-01

Maintenance operations cover a great number of occupations. Most small and medium-sized enterprises lack the appropriate information to conduct risk assessments of maintenance operations. The objective of this research is to provide a method based on the concepts of task and accident mechanisms for an initial risk assessment by taking into consideration the prevalence and severity of the maintenance accidents reported. Data were gathered from 11,190 reported accidents in maintenance operations in the manufacturing sector of Andalusia from 2003 to 2012. By using a semi-quantitative methodology, likelihood and severity were evaluated based on the actual distribution of accident mechanisms in each of the tasks. Accident mechanisms and tasks were identified by using those variables included in the European Statistics of Accidents at Work methodology. As main results, the estimated risk of the most frequent accident mechanisms identified for each of the analysed tasks is low and the only accident mechanisms with medium risk are accidents when lifting or pushing with physical stress on the musculoskeletal system in tasks involving carrying, and impacts against objects after slipping or stumbling for tasks involving movements. The prioritisation of public preventive actions for the accident mechanisms with a higher estimated risk is highly recommended.

Severe accidents: in nuclear power plants

International Nuclear Information System (INIS)

1986-01-01

A ''severe'' nuclear accident refers to a reactor accident that could exceed reactor design specifications to such a degree as to prevent cooling of the reactor's core by normal means. This report summarizes the work of a NEA Senior Group of Experts who have studied the potential response of existing light-water reactors to severe accidents and have found that current designs of reactors are far more capable of coping with severe accidents than design specifications would suggest. The report emphasises the specific knowledge and means that can be used for diagnosing a severe accident and for managing its progression in order to prevent or mitigate its consequences

Chernobyl - what can natural scientists or physicians say to that accident?

International Nuclear Information System (INIS)

1986-01-01

The public discussion meeting was intended to offer to the general public a platform for discussion of questions evoked by the Chernobyl reactor accident, and scientific information on what has happened there. The brief lectures therefore deal with the accident scenario as far as assessable at the time, and with the consequences to be expected for the Federal Republic of Germany, with the fallout situation in the Mainz area, and the atmospheric dispersion and transfer of air masses from Chernobyl to the FRG. The medical experts presented information on the radiation exposure of the population and the possible genetic risk. (DG) [de

Assessment of WWER fuel condition in design basis accident

Energy Technology Data Exchange (ETDEWEB)

Bibilashvili, Yu; Sokolov, N; Andreeva-Andrievskaya, L; Vlasov, Yu; Nechaeva, O; Salatov, A [Vsesoyuznyj Nauchno-Issledovatel` skij Inst. Neorganicheskikh Materialov, Moscow (Russian Federation)

1994-12-31

The fuel behaviour in design basis accidents is assessed by means of the verified code RAPTA-5. The code uses a set of high temperature physico-chemical properties of the fuel components as determined for commercially produced materials, fuel rod simulators and fuel rod bundles. The WWER fuel criteria available in Russia for design basis accidents do not generally differ from the similar criteria adopted for PWR`s. 12 figs., 11 refs.

77 FR 60481 - Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Units of Post-Accident...

Science.gov (United States)

2012-10-03

... filtration and iodine adsorption units of ESF atmosphere cleanup systems in light-water-cooled nuclear power... Filtration and Adsorption Units of Post-Accident Engineered-Safety-Feature Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants AGENCY: Nuclear Regulatory Commission. ACTION: Regulatory guide...

Expert systems applied to two problems in nuclear power plants

International Nuclear Information System (INIS)

Kim, K.Y.

1988-01-01

This dissertation describes two prototype expert systems applied to two problems in nuclear power plants. One problem is spare parts inventory control, and the other one is radionuclide release from containment during severe accident. The expert system for spare parts inventory control can handle spare parts requirements not only in corrective, preventive, or predictive maintenance, but also when failure rates of components or parts are updated by new data. Costs and benefits of spare parts inventory acquisition are evaluated with qualitative attributes such as spare part availability to provide the inventory manager with an improved basis for decision making. The expert system is implemented with Intelligence/Compiler on an IBM-AT. The other expert system for radionuclide release from containment can estimate magnitude, type, location, and time of release of radioactive materials from containment during a severe accident nearly on line, based on the actual measured physical parameters such as temperature and pressure inside the containment. The expert system has a function to check the validation of sensor data. The expert system is implemented with KEE on a Symbolics LISP machine

LMFBR post accident heat removal testing needs and conceptual design of a test facility

International Nuclear Information System (INIS)

Kleefeldt, K.; Kuechle, M.; Royl, P.; Werle, H.; Boenisch, G.; Heinzel, V.; Mueller, R.A.; Schramm, K.; Smidt, D.

1977-03-01

A study has been carried out in which the needs and requirements for a test facility were derived, enabling detailed investigation of key phenomena anticipated during the post accident heat removal (PAHR) phase as a consequence of a postulated LMFBR whole core accident. Part I of the study concentrates on demonstrating the PAHR phenomena and related testing needs. Three types of experiments were identified which require in-pile testing, ranging from 10 to 70 cm test bed diameter and correspondingly, 30 to 5 W/g minimum power density in the test fuel. In part II a conceptual design for a test facility is presented, emphasizing the capability for accomodating large test beds. This is achieved by a below-reactor-vessel testing device, neutronically coupled to a 100 MWt sodium cooled fast reactor. (orig.) [de

IAEA Expert Team Completes Mission to Review Japan's Nuclear Power Plant Safety Assessment Process, 31 January 2012, Tokyo, Japan

International Nuclear Information System (INIS)

2012-01-01

Full text: A team of international nuclear safety experts today completed a review of Japan's two-stage process for assessing nuclear safety at the nation's nuclear power plants. The team began its work on 23 January and delivered a Preliminary Summary Report to Japanese officials today and plans to finish the final report by the end of February. National safety assessments and their peer review by the IAEA are a key component of the IAEA's Action Plan on Nuclear Safety, which was approved by the Agency's 152 Member States following last year's nuclear accident at Fukushima Daiichi Nuclear Power Station. At the request of the Government of Japan, the International Atomic Energy Agency (IAEA) organized a 10-person team to review the Japanese Nuclear and Industrial Safety Agency's (NISA) approach to the Comprehensive Assessments for the Safety of Existing Power Reactor Facilities and how NISA examines the results submitted by nuclear operators. The IAEA safety review mission consisted of five IAEA and three international nuclear safety experts. To help its review, the team held meetings in Tokyo with officials from NISA, the Japanese Nuclear Energy Safety (JNES) Organization, and the Kansai Electric Power Company (KEPCO), and the team visited the Ohi Nuclear Power Station to see an example of how Japan's Comprehensive Safety Assessment is being implemented by nuclear operators. 'We concluded that NISA's instructions to power plants and its review process for the Comprehensive Safety Assessments are generally consistent with IAEA Safety Standards', said team leader James Lyons, director of the IAEA's Nuclear Installation Safety Division. In its Preliminary Summary Report delivered today, the team highlighted a number of good practices and identified some improvements that would enhance the overall effectiveness of the Comprehensive Safety Assessment process. Good practices identified by the mission team include: Based on NISA instructions and commitments of the

Accident Diagnosis and Prognosis Aide (ADPA)

International Nuclear Information System (INIS)

Gunter, A.D.; Touchton, R.A.

1987-01-01

This presentation provides a demonstration of a prototypical expert system developed by Technology Applications, Inc. (TAI) under a contract with the Department of Energy as a part of their Small Business Innovation Research Program. The Accident Diagnosis and Prognosis Aide (ADPA) Demonstration Prototype is a working scale model of a real-time expert system which: Diagnoses an accident situation (as well as a number of underlying failures, events, and conditions deduced along the way). Calculates the change in the likelihood of core damage as a function of the events and failures diagnosed. Dynamically generates a recovery procedure tailored to the specific plant state at hand

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

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

Fukushima Daiichi. Post accident countermeasures and mid-and-long-term action plan

International Nuclear Information System (INIS)

Yamashita, Kazuhiko

2012-01-01

Since accident at Fukushima Daiichi nuclear power station, TEPCO had made strenuous efforts to maintain stable cooling of reactors and suppress discharge of radioactive materials under the cooperation of the government and others concerned. Various activities towards settlement of the accident had been done according to the roadmap proposed on April 17, 2011. This article described outlines of post accident countermeasures and current activities toward decommissioning based on mid-and-long-term action plan. Step 2 completion in roadmap towards settlement of the accident was confirmed on Dec. 16, 2011; (1) reactors: attainment of a condition equivalent to 'cold shutdown' with temperature of RPV bottom below 100degC, release of radioactive materials from PCV under control and public radiation exposure by additional release being significantly held down (evaluated as maximum 0.2 mSv/y at the site boundary), and mid-term safety of circulating water injection cooling system, (2) spent fuel pools: more stable cooling with circulating cooling system by installation of heat exchanger, (3) radioactive contaminated water: reduction of total amount with full-fledged processing facilities, desalination processing (reuse), storage and mitigation of contamination in the ocean, (4) commence construction of water shielding wall to prevent contamination of the ocean via the underground water, (5) installation of reactor building cover of unit 1 to restrain release of radioactive materials, and others. According to mid-to-long-term roadmap towards the decommissioning of Fukushima Nuclear Power Units 1-4, current activities were to (1) maintain stable reactor cooling and accumulated water processing and improve their credibility, (2) reduce the radiation impact due to additional emissions from the whole site and radioactive waste generated after the accident, (3) commence removal of fuel from the spent fuel pools (unit 4 within 2 years), (4) commence R and D and decontamination the

Guidelines for the review of accident management programmes in nuclear power plants. Reference document for the IAEA safety service missions on review of accident management programmes in nuclear power plants

International Nuclear Information System (INIS)

2003-01-01

Similarly as for other IAEA safety services, the objectives of accident management safety service are to assist the Member States in ensuring and enhancing the safety of NPPs. In particular, the objective is to assist at the utility and NPP (i.e. licensee) level in effective plant specific AMP preparation, development and implementation. However, assistance can also be provided to the regulatory body in its reviewing of AMPs. Objectives of the safety service can be summarized as follows: To explain to licensee personnel principles and possible approaches in effective implementation of AMP based on experience world-wide; To give opportunities to experts from the host plant to broaden their experience and knowledge in the field; To perform an objective assessment of the status in various phases of AMP implementation, compared with international experience and practices; To provide the licensee with suggestions and assistance for improvements in various stages of AMP implementation. The objective of the IAEA safety services is to offer two options to respond to individual requirements. These options include missions to review accident analysis needed for accident management and missions to review the whole AMP. Review of accident analysis for accident management (RAAAM): this review is intended to check completeness and quality of accident analysis covering BDBA and severe accidents. The review should be typically performed prior to use of accident analysis for development of AMP. It is considered that 2 experts and 1 IAEA team leader in one-week mission can perform the review. Detailed guidelines for review of analysis are provided in Section 2. Reference is also made to another IAEA Safety Report (Safety Standards Series No. NS-R-1) which is devoted to guidance for accident analysis of nuclear power plants (NPPs). Review of AMP (RAMP): this review of AMP, which is in particular appropriate prior to its implementation, is intended to check its quality, consistency

Post-traumatic stress disorders: Aftereffects of the Goiania radiological accident

International Nuclear Information System (INIS)

Bandeira de Carvalho, A.

2000-01-01

In September of 1987, a 137 Cs medical source was violated in the city of Goiania, Brazil and, as a result, four people died, 249 were internally and/or externally contaminated and 20 people were hospitalised at the Goiania General Hospital. The most serious cases were transferred to the Marcello Dias Naval Hospital in Rio de Janeiro. The objective of this paper is to describe the emotional and social impacts caused by this accident on the hospitalised victims. As these people belonged to a low social-economic level and did not understand the meaning of radioactivity, panic only started to be felt with the first deaths. The origin of their stress was, primarily, the radiation syndrome, but mostly the rejection suffered by them in the Goiania General Hospital, caused by fear of radioactivity; the difficulty in understanding the safety regulations; the always covered staffs; and the isolation due to immune-suppression. They feared having a limb amputated, were worried about their families, ambivalent about death: wishing and afraid of it. They felt guilty about the accident but at the same time they were seeking compensation from the State. Aggressive, due to the long confinement, they rioted inside the hospital. The need for psychological support was felt by the staff but few psychologists volunteered because of fear. Human beings present a number of psychological disorders after the trauma suffered in a catastrophe. These have a pattern, called post-traumatic stress disorders, which was noticed in this accident. It is a lesson learned in Goiania that psychological support is of vital importance to help the patient fight the consequences of the suffering. Psychologists should be trained to work without fear, during emergencies of this kind. (author)

Severe accident risks: An assessment for five US nuclear power plants

International Nuclear Information System (INIS)

1991-01-01

This report summarizes an assessment of the risks from severe accidents in five commercial nuclear power plants in the United State. These risks are measured in a number of ways, including: the estimated frequencies of core damage accidents from internally initiated accidents and externally initiated accidents for two of the plants; the performance of containment structures under severe accident loadings; the potential magnitude of radionuclide releases and offsite consequences of such accidents; and the overall risk (the product of accident frequencies and consequences). Supporting this summary report are a large number of reports written under contract to NRC that provide the detailed discussion of the methods used and results obtained in these risk studies. This report, Volume 3, contains two appendices. Appendix D summarizes comments received, and staff responses, on the first (February 1987) draft of NUREG-1150. Appendix E provides a similar summary of comments and responses, but for the second (June 1989) version of the report

Radiological accident in Panama - IAEA to send assistance team

International Nuclear Information System (INIS)

2001-01-01

Full text: The International Atomic Energy Agency (IAEA) is sending a team of six international experts to assist the authorities of Panama to deal with the aftermath of a radiological accident that occurred at Panama's National Oncology Institute. The Government of Panama informed the IAEA on 22 May about the accident, reported that 28 patients have been affected, and requested IAEA's assistance under the Convention on Assistance in the Case of a Nuclear Accident or Radiological Emergency, to which Panama is a party. The assistance to be provided by the expert mission will include: ensuring that the radiation source(s) involved in the accident is (are) in a safe and secure condition; evaluating the doses incurred by the affected patients, inter alia, by analysing the treatment records and physical measurements; undertaking a medical evaluation of the affected patients' prognosis and treatment, taking into account, inter alia, the autopsy findings for those who died; and identifying issues on which the IAEA could offer to provide and/or co-ordinate assistance to minimize the consequences of the accident. The team, which includes senior experts in radiology, radiotherapy, radiopathology, radiation dosimetry and radiation protection from France, USA and Japan, and the IAEA itself, will leave for Panama tomorrow, 26 May

The need to study of bounding accident in reprocessing plant

International Nuclear Information System (INIS)

Segawa, Satoshi; Fujita, Kunio

2013-01-01

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

Principles and techniques for post-accident assessment and recovery in a contaminated environment of a nuclear facility

International Nuclear Information System (INIS)

1989-01-01

To assist operators and public authorities alike in their advance preparation of emergency plans and in the establishment of emergency preparedness infrastructures, the IAEA has already issued several Safety Series publications dealing with these matters. This Safety Guide complements the technical guidance already published. It provides: a) Information and practical guidance relevant to assessing the off-site consequences during the late phase of a serious accident in a nuclear facility; b) Guidance on recovery operations off the site and the associated decision making process; and c) Proposals for consideration by national authorities regarding the organizational structure for the conduct of recovery operations. 52 refs, 8 figs, 4 tabs.

Evaluation of nuclear accidents consequences. Risk assessment methodologies, current status and applications

International Nuclear Information System (INIS)

Rodriguez, J.M.

1996-01-01

General description of the structure and process of the probabilistic methods of assessment the external consequences in the event of nuclear accidents is presented. attention is paid in the interface with Probabilistic Safety Analysis level 3 results (source term evaluation) Also are described key issues in accident consequence evaluation as: effects evaluated (early and late health effects and economic effects due to countermeasures), presentation of accident consequences results, computer codes. Briefly are presented some relevant areas for the applications of Accident Consequence Evaluation

Meeting of the High Committee - 13 December 2012: Report by the national commission of assessment of the financing of nuclear liabilities on a long term (CNEF); Elements of doctrine for the post-accidental management of a nuclear accident; Interventional radiology

International Nuclear Information System (INIS)

Lepine, M.; Gallay, Florence; Aubert, Bernard

2012-01-01

The first part of this document proposes a report made by the national commission of assessment of the financing of the nuclear liabilities on a long term (CNEF): presentation of the administrative authority, assessment of liabilities, assets, and remarks. The second part contains elements of doctrine for the post-accidental management of a nuclear accident: works made by the CODIRPA, definition of three critical objectives and six key issues for post-accidental management, post-accidental zoning, first post-accidental measures for the protection of the population, future objectives for CODIRPA work groups. The third part addresses interventional radiology: definition, specificities (uses, characteristics, continuous increase, number of sessions), related risks, clinic evidence and thresholds and delays of appearance of cutaneous effects, international concern, legal and regulatory framework (radiation protection measures), incidents known by the IRSN, results of a national survey on patient radiation protection in interventional cardiology centres, perspectives and expectations, worker exposure, recommendations regarding the regulatory framework, professional training, equipment and patient monitoring, information on potential undesired effects

Researches and Applications of ESR Dosimetry for Radiation Accident Dose Assessment

International Nuclear Information System (INIS)

Wu, K.; Guo, L.; Cong, J.B.; Sun, C.P.; Hu, J.M.; Zhou, Z.S.; Wang, S.; Zhang, Y.; Zhang, X.; Shi, Y.M.

1998-01-01

The aim of this work was to establish methods suitable for practical dose assessment of people involved in ionising radiation accidents. Some biological materials of the human body and materials possibly carried or worn by people were taken as detection samples. By using electron spin resonance (ESR) techniques, the basic dosimetric properties of selected materials were investigated in the range above the threshold dose of human acute haemopoietic radiation syndrome. The dosimetric properties involved included dose response properties of ESR signals, signal stabilities, distribution of background signals, the lowest detectable dose value, radiation conditions, environmental effects on the detecting process, etc. Several practical dose analytical indexes and detecting methods were set up. Some of them (bone, watch glass and tooth enamel) had also been successfully used in the dose assessment of people involved in three radiation accidents, including the Chernobyl reactor accident. This work further proves the important role of ESR techniques in radiation accident dose estimation. (author)

Evaluation of severe accident risks: Methodology for the containment, source term, consequence, and risk integration analyses. Volume 1, Revision 1

International Nuclear Information System (INIS)

Gorham, E.D.; Breeding, R.J.; Brown, T.D.; Harper, F.T.; Helton, J.C.; Murfin, W.B.; Hora, S.C.

1993-12-01

NUREG-1150 examines the risk to the public from five nuclear power plants. The NUREG-1150 plant studies are Level III probabilistic risk assessments (PRAs) and, as such, they consist of four analysis components: accident frequency analysis, accident progression analysis, source term analysis, and consequence analysis. This volume summarizes the methods utilized in performing the last three components and the assembly of these analyses into an overall risk assessment. The NUREG-1150 analysis approach is based on the following ideas: (1) general and relatively fast-running models for the individual analysis components, (2) well-defined interfaces between the individual analysis components, (3) use of Monte Carlo techniques together with an efficient sampling procedure to propagate uncertainties, (4) use of expert panels to develop distributions for important phenomenological issues, and (5) automation of the overall analysis. Many features of the new analysis procedures were adopted to facilitate a comprehensive treatment of uncertainty in the complete risk analysis. Uncertainties in the accident frequency, accident progression and source term analyses were included in the overall uncertainty assessment. The uncertainties in the consequence analysis were not included in this assessment. A large effort was devoted to the development of procedures for obtaining expert opinion and the execution of these procedures to quantify parameters and phenomena for which there is large uncertainty and divergent opinions in the reactor safety community

FN-curves: preliminary estimation of severe accident risks after Fukushima

International Nuclear Information System (INIS)

Vasconcelos, Vanderley de; Soares, Wellington Antonio; Costa, Antonio Carlos Lopes da

2015-01-01

Doubts of whether the risks related to severe accidents in nuclear reactors are indeed very low were raised after the nuclear accident at Fukushima Daiichi in 2011. Risk estimations of severe accidents in nuclear power plants involve both probability and consequence assessment of such events. Among the ways to display risks, risk curves are tools that express the frequency of exceeding a certain magnitude of consequence. Societal risk is often represented graphically in a FN-curve, a type of risk curve, which displays the probability of having N or more fatalities per year, as a function of N, on a double logarithmic scale. The FN-curve, originally introduced for the assessment of the risks in the nuclear industry through the U.S.NRC Reactor Safety Study WASH-1400 (1975), is used in various countries to express and limit risks of hazardous activities. This first study estimated an expected rate of core damage equal to 5x10 -5 by reactor-year and suggested an upper bound of 3x10 -4 by reactor-year. A more recent report issued by Electric Power Research Institute - EPRI (2008) estimates a figure of the order of 2x10 -5 by reactor-year. The Fukushima nuclear accident apparently implies that the observed core damage frequency is higher than that predicted by these probabilistic safety assessments. Therefore, this paper presents a preliminary analyses of the FN-curves related to severe nuclear reactor accidents, taking into account a combination of available data of past accidents, probability modelling to estimate frequencies, and expert judgments. (author)

FN-curves: preliminary estimation of severe accident risks after Fukushima

Energy Technology Data Exchange (ETDEWEB)

Vasconcelos, Vanderley de; Soares, Wellington Antonio; Costa, Antonio Carlos Lopes da, E-mail: vasconv@cdtn.br, E-mail: soaresw@cdtn.br, E-mail: aclc@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2015-07-01

Doubts of whether the risks related to severe accidents in nuclear reactors are indeed very low were raised after the nuclear accident at Fukushima Daiichi in 2011. Risk estimations of severe accidents in nuclear power plants involve both probability and consequence assessment of such events. Among the ways to display risks, risk curves are tools that express the frequency of exceeding a certain magnitude of consequence. Societal risk is often represented graphically in a FN-curve, a type of risk curve, which displays the probability of having N or more fatalities per year, as a function of N, on a double logarithmic scale. The FN-curve, originally introduced for the assessment of the risks in the nuclear industry through the U.S.NRC Reactor Safety Study WASH-1400 (1975), is used in various countries to express and limit risks of hazardous activities. This first study estimated an expected rate of core damage equal to 5x10{sup -5} by reactor-year and suggested an upper bound of 3x10{sup -4} by reactor-year. A more recent report issued by Electric Power Research Institute - EPRI (2008) estimates a figure of the order of 2x10{sup -5} by reactor-year. The Fukushima nuclear accident apparently implies that the observed core damage frequency is higher than that predicted by these probabilistic safety assessments. Therefore, this paper presents a preliminary analyses of the FN-curves related to severe nuclear reactor accidents, taking into account a combination of available data of past accidents, probability modelling to estimate frequencies, and expert judgments. (author)

Overview of domestic and foreign comments on Fukushima nuclear accident in the past half year

International Nuclear Information System (INIS)

Liu Yuanyuan; Chen Haiying; Zhang Chunming

2011-01-01

On March 11, 2011, Fukushima Dai-ichi nuclear power plant happened severity level nuclear accident, and caused great disasters. To be able to get more experience feedbacks for the safe development of nuclear technology from this accident, this paper investigated the domestic and foreign comments on Fukushima nuclear accident in the past half year, summarized and analyzed the attentions of experts a- bout this accident and improvement methods for the existing nuclear power technology. The results show that most domestic experts focus on improving nuclear power technology and management of emergency response and supervision, while foreign experts pay more attentions to how to deal with the environmental impact of accident on the biological side. All of these attentions and improvement methods will provide more useful information for developing safe nuclear power technology. (authors)

Probabilistic accident consequence uncertainty analysis -- Uncertainty assessment for deposited material and external doses. Volume 2: Appendices

Energy Technology Data Exchange (ETDEWEB)

Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Boardman, J. [AEA Technology (United Kingdom); Jones, J.A. [National Radiological Protection Board (United Kingdom); Harper, F.T.; Young, M.L. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States)

1997-12-01

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA deposited material and external dose models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the panel on deposited material and external doses, (4) short biographies of the experts, and (5) the aggregated results of their responses.

Fuel Accident Condition Simulator (FACS) Furnace for Post-Irradiation Heating Tests of VHTR Fuel Compacts

Energy Technology Data Exchange (ETDEWEB)

Paul A Demkowicz; Paul Demkowicz; David V Laug

2010-10-01

Abstract –Fuel irradiation testing and post-irradiation examination are currently in progress as part of the Next Generation Nuclear Plant Fuels Development and Qualification Program. The PIE campaign will include extensive accident testing of irradiated very high temperature reactor fuel compacts to verify fission product retention characteristics at high temperatures. This work will be carried out at both the Idaho National Laboratory (INL) and the Oak Ridge National Laboratory, beginning with accident tests on irradiated fuel from the AGR-1 experiment in 2010. A new furnace system has been designed, built, and tested at INL to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000°C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, Eu, and I) and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator (FACS) furnace system, as well as preliminary system calibration results.

Problems of probabilistic safety assessment after Fukushima Daiichi nuclear power plant accident

International Nuclear Information System (INIS)

Sugiyama, Naoki

2011-01-01

Probabilistic safety assessment (PSA) methodology to assure nuclear safety is had great expectations of lessons learned from Fukushima Daiichi nuclear power plant (NPP) accident and on the other hand this accident made actualized technical problems of PSA. Effectiveness of current PSA methodology for risk assessment was confirmed by comparing the accident development with accident scenario of PSA and equipment failure rate. From a viewpoint of nuclear safety objective and defense in depth approach of IAEA, technical problems of PSA were (1) extension of PSA for spent fuel pool and waste disposal system as well as level 3PSA for broader environmental contamination and (2) overlapping of accident scenario of plural unit site, balance of high quality plant management and preceding negation, treatment of uncertainty of external events, severe accident measure and human reliability analysis and reflection of disaster prevention capability to level 3PSA. In order to upgrade PSA technology, six proposals were described for nuclear safety and defense in depth, comprehensive evaluation scope and catch-up of latest technology, necessity of strategic preparation of PSA standard, human resources fostering and risk communication. (T. Tanaka)

Advanced sodium fast reactor accident source terms :

Energy Technology Data Exchange (ETDEWEB)

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

2010-09-01

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

Reactor accident diagnostic expert system: DISKET

International Nuclear Information System (INIS)

Yoshida, Kazuo; Yokobayashi, Masao

1989-11-01

A reactor accident diagnostic system DISKET has been developed to identify the cause and the type of an abnormal transient of a nuclear power plant. The system is based on the knowledge engineering and consists of an inference engine IERIAS and a knowledge base. The main features of DISKET are the following: Time-varying characteristics of transient can be treated and knowledge base can be divided into several knowledge units to handle a lot of rules effectively. This report has been provided for the convenience of DISKET's users and consists of three parts. The first part is the description of the whole system, the details of the knowledge base of DISKET are described in the second part, and how to use the DISKET system is explained in the third part. (author)

Radiation risk perception: a discrepancy between the experts and the general population

International Nuclear Information System (INIS)

Perko, Tanja

2014-01-01

Determining the differences in the perception of risks between experts who are regularly exposed to radiation, and lay people provides important insights into how potential hazards may be effectively communicated to the public. In the present study we examined lay people's (N = 1020) and experts' (N = 332) perception of five different radiological risks: nuclear waste, medical x-rays, natural radiation, an accident at a nuclear installation in general, and the Fukushima accident in particular. In order to link risk perception with risk communication, media reporting about radiation risks is analysed using quantitative and qualitative content analyses. The results showed that experts perceive radiological risks differently from the general public. Experts' perception of medical X-rays and natural radiation is significantly higher than in general population, while for nuclear waste and an accident at a nuclear installation, experts have lower risk perception than the general population. In-depth research is conducted for a group of workers that received an effective dose higher than 0.5 mSv in the year before the study; for this group we identify predictors of risk perception. The results clearly show that mass media don't use the same language as technical experts in addressing radiological risks. The study demonstrates that the discrepancy in risk perception and the communication gap between the experts and the general population presents a big challenge in understanding each other

Safety analysis of an expert reactor protection system in nuclear power plants

International Nuclear Information System (INIS)

El-Kafas, A.A.

1997-01-01

The purpose of the dissertation is to develop real time expert reactor protection system (ERPS) for operational safety of pressurized water reactor nuclear power plant. The system is developed to diagnose plant failures and for identification plant transients (with and without scram). For this erps, probabilistic safety analysis techniques are used to check the availability and priority of the recommended safety system in case of plant accidents. The real - time information during transients and accidents can be obtained to assess the operator in his decision - making. Also, the ERPS is able to give advice for the reactor operator to take the appropriate corrective action during abnormal situations. 5-15 figs., 42 refs

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

Data base of accident and agricultural statistics for transportation risk assessment

Energy Technology Data Exchange (ETDEWEB)

Saricks, C.L.; Williams, R.G.; Hopf, M.R.

1989-11-01

A state-level data base of accident and agricultural statistics has been developed to support risk assessment for transportation of spent nuclear fuels and high-level radioactive wastes. This data base will enhance the modeling capabilities for more route-specific analyses of potential risks associated with transportation of these wastes to a disposal site. The data base and methodology used to develop state-specific accident and agricultural data bases are described, and summaries of accident and agricultural statistics are provided. 27 refs., 9 tabs.

Data base of accident and agricultural statistics for transportation risk assessment

International Nuclear Information System (INIS)

Saricks, C.L.; Williams, R.G.; Hopf, M.R.

1989-11-01

A state-level data base of accident and agricultural statistics has been developed to support risk assessment for transportation of spent nuclear fuels and high-level radioactive wastes. This data base will enhance the modeling capabilities for more route-specific analyses of potential risks associated with transportation of these wastes to a disposal site. The data base and methodology used to develop state-specific accident and agricultural data bases are described, and summaries of accident and agricultural statistics are provided. 27 refs., 9 tabs

USING EXPERT OPINION IN HEALTH TECHNOLOGY ASSESSMENT: A GUIDELINE REVIEW.

Science.gov (United States)

Hunger, Theresa; Schnell-Inderst, Petra; Sahakyan, Narine; Siebert, Uwe

2016-01-01

External experts can be consulted at different stages of an HTA. When using vague information sources, it is particularly important to plan, analyze, and report the information processing in a standardized and transparent way. Our objective was to search and analyze recommendations regarding where and how to include expert data in HTA. We performed a systematic database search and screened the Internet pages of seventy-seven HTA organizations for guidelines, recommendations, and methods papers that address the inclusion of experts in HTA. Relevant documents were downloaded, and information was extracted in a standard form. Results were merged in tables and narrative evidence synthesis. From twenty-two HTA organizations, we included forty-two documents that consider the use of expert opinion in HTA. Nearly all documents mention experts in the step of preparation of the evidence report. Six documents address their role for priority setting of topics, fifteen for scoping, twelve for the appraisal of evidence and results, another twelve documents mention experts when considering the dissemination of HTA results. During the assessment step, experts are most often asked to amend the literature search or to provide expertise for special data analyses. Another issue for external experts is to appraise the HTA results and refer them back to a clinical and social context. Little is reported on methods of expert elicitation when their input substitutes study data. Despite existing recommendations on the use of expert opinion in HTA, common standards for elicitation are scarce in HTA guidelines.

[Occupational accidents due to exposure to biological material in the multidisciplinary team of the emergency service].

Science.gov (United States)

Oliveira, Adriana Cristina; Lopes, Aline Cristine Souza; Paiva, Maria Henriqueta Rocha Siqueira

2009-09-01

This transversal, survey-based research was carried out with a multiprofessional emergency care team in Belo Horizonte, between June and December 2006. The study aimed at estimating the incidence of occupational accidents by exposure to biological material, post-accidents conducts and demographic determinant factors. The study applied a structured questionnaire and descriptive analyses, as well as incidence calculations and logistic regression. The incidence of accidents with biological material reached 20.6%, being 40.8% by sharp materials and 49.0% by body fluids; 35.3% of the accidents took place among physicians and 24.0% among nurses. Post-accidents procedures: no medical assessment, 63.3%; under-notification, 81.6%; no conduct, 55.0%; and no serological follow-up, 61.2%. Factors associated with accidents: working time in the institution (Odds Ratio--OR, 2.84; Credible Interval--CI 95%-1.22-6.62); working in advanced support units (OR = 4.18; CI 95%--1.64-10.64); and interaction between working time in the institution and working in Basic Support Unit (OR 0.27; CI 95%--0.07-1.00). In order to reduce accidents, the implementation of post-accident protocols and follow-up, as well as under-notification norms, are suggested.

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

Technology, safety, and costs of decommissioning reference light-water reactors following postulated accidents. Appendices

Energy Technology Data Exchange (ETDEWEB)

Murphy, E S; Holter, G M

1982-11-01

Appendices contain information concerning the reference site description; reference PWR facility description; details of reference accident scenarios and resultant contamination levels; generic cleanup and decommissioning information; details of activities and manpower requirements for accident cleanup at a reference PWR; activities and manpower requirements for decommissioning at a reference PWR; costs of decommissioning at a reference PWR; cost estimating bases; safety assessment details; and details of post-accident cleanup and decommissioning at a reference BWR.

Navigation safety and risk assessment challenges in the High North

DEFF Research Database (Denmark)

Marchenko, N.A.; Borch, O.J.; Andreassen, N.

2017-01-01

marine accidents. Therefore, in this study a mostly qualitative analysis and expert judgement is the basis for the risk assessments. Implications for the emergency preparedness system of the region are discussed. The consequences of incidents depend on the incident type, scale and location,....... In this paper we look into the risks of accidents in the Atlantic Arctic based on previous ship accidents and the changes in maritime activity. The risk has to be assessed to ensure a proper level of response in emergency situations. As accidents are rare, there are limited statistics available for Arctic...

An application of probabilistic safety assessment methods to model aircraft systems and accidents

Energy Technology Data Exchange (ETDEWEB)

Martinez-Guridi, G.; Hall, R.E.; Fullwood, R.R.

1998-08-01

A case study modeling the thrust reverser system (TRS) in the context of the fatal accident of a Boeing 767 is presented to illustrate the application of Probabilistic Safety Assessment methods. A simplified risk model consisting of an event tree with supporting fault trees was developed to represent the progression of the accident, taking into account the interaction between the TRS and the operating crew during the accident, and the findings of the accident investigation. A feasible sequence of events leading to the fatal accident was identified. Several insights about the TRS and the accident were obtained by applying PSA methods. Changes proposed for the TRS also are discussed.

Users guide for NRC145-2 accident assessment computer code

International Nuclear Information System (INIS)

Pendergast, M.M.

1982-08-01

An accident assessment computer code has been developed for use at the Savannah River Plant. This computer code is based upon NRC Regulatory Guide 1.145 which provides guidence for accident assessements for power reactors. The code contains many options so that the user may utilize the code for many different assessments. For example the code can be used for non-nuclear assessments such as Sulpher Dioxide which may be required by the EPA. A discription of the code is contained in DP-1646. This document is a compilation of step-by-step instructions on how to use the code on the SRP IBM 3308 computer. This document consists of a number of tables which contain copies of computer listings. Some of the computer listings are copies of input; other listings give examples of computer output

Generic assessment procedures for determining protective actions during a reactor accident

International Nuclear Information System (INIS)

1997-08-01

This manual provides the tools, procedures and data needed to evaluate the consequences of a nuclear accident occurring at a nuclear power plant throughout all phases of the emergency before, during and after a release of radioactive material. It is intended for use by on-site and off-site groups responsible for evaluating the accident consequences and making recommendations for the protection of the plant personnel, the emergency workers and the public. The scope of this manual is restricted to the technical assessment of radiological consequences. It does not address the emergency response infrastructure requirements, nor does it cover the emergency management aspects of accident assessment (e.g. reporting, staff qualification, shift replacement, and procedure implementation). The procedures and methods in this manual were developed based on a number of assumptions concerning the design and operation of the nuclear power plant and national practices. Therefore, this manual must be reviewed as part of the planning process to match the potential accidents, local conditions, national criteria and other unique characteristics of an area or nuclear reactor where it may be used. Refs, figs, tabs

Assessment of the accident response of a light-water-moderated breeder-reactor system: AWBA development program

International Nuclear Information System (INIS)

High, H.M.

1983-05-01

The predicted accident response for a light water moderated, thorium/U-233 fueled, seed-blanket reactor concept was assessed. The first part of the assessment compared breeder accident response with that of a current commercial pressurized water reactor design for several different types of transients. Based on these comparisons and a review of the various parameter differences between the breeder and a U-235 fueled plant, the second part of the assessment studied the breeder accident behavior in more detail, particularly in areas of potential concern. Based on the two parts of the assessment, it was concluded that the breeder accident response would be very similar to that of present commercial pressurized water reactor plants. The large Doppler and moderator reactivity coefficients of the breeder would significantly reduce the severity of many of the accidents that must be considered. It is expected that the accident response of the breeder can be shown to meet regulatory criteria

Accident of the Fukushima-Daiichi nuclear power station. Situation two years after the event - IRSN file

International Nuclear Information System (INIS)

2013-03-01

Two years after the Fukushima accident, this report proposes a review of the situation in Japan, and of the European and international actions aimed at preventing the occurrence of another nuclear accident and its radiological consequences. It is based on information available at the end of January or February 2013. After a recall of the situation in Japan and Europe in 2011 (recall of the accident, of the different simulation, calculation and information actions undertaken by the IRSN, launching of a program of additional safety assessments and of European stress tests), the report addresses the situation in Japan two years after the accident: evolution of the nuclear risk management governance, status of the Fukushima-Daiichi power station, health and environmental impact and management of the post-accidental phase, actions undertaken by the IRSN (dose assessment, cooperation in the field of severe accidents, participation to the Fukushima Dialogue). The next part details the contribution of the IRSN to the strengthening of safety and radiation protection at the international level (in relationship with international organizations: IAEA, UNSCEAR and WHO). Additional technical information is provided in appendix, as well as a report on the environmental impact of the accident, and a report on the post-accidental management of the accident

Hazard Identification, Risk Assessment and Risk Control (HIRARC Accidents at Power Plant

Directory of Open Access Journals (Sweden)

Ahmad Asmalia Che

2016-01-01

Full Text Available Power plant had a reputation of being one of the most hazardous workplace environments. Workers in the power plant face many safety risks due to the nature of the job. Although power plants are safer nowadays since the industry has urged the employer to improve their employees’ safety, the employees still stumble upon many hazards thus accidents at workplace. The aim of the present study is to investigate work related accidents at power plants based on HIRARC (Hazard Identification, Risk Assessment and Risk Control process. The data were collected at two coal-fired power plant located in Malaysia. The finding of the study identified hazards and assess risk relate to accidents occurred at the power plants. The finding of the study suggested the possible control measures and corrective actions to reduce or eliminate the risk that can be used by power plant in preventing accidents from occurred

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

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

MELCOR assessment of sequential severe accident mitigation actions under SGTR accident

International Nuclear Information System (INIS)

Choi, Wonjun; Jeon, Joongoo; Kim, Nam Kyung; Kim, Sung Joong

2017-01-01

The representative example of the severe accident studies using the severe accident code is investigation of effectiveness of developed severe accident management (SAM) strategy considering the positive and adverse effects. In Korea, some numerical studies were performed to investigate the SAM strategy using various severe accident codes. Seo et.al performed validation of RCS depressurization strategy and investigated the effect of severe accident management guidance (SAMG) entry condition under small break loss of coolant accident (SBLOCA) without safety injection (SI), station blackout (SBO), and total loss of feed water (TLOFW) scenarios. The SGTR accident with the sequential mitigation actions according to the flow chart of SAMG was simulated by the MELCOR 1.8.6 code. Three scenariospreventing the RPV failure were investigated in terms of fission product release, hydrogen risk, and the containment pressure. Major conclusions can be summarized as follows: (1) According to the flow chart of SAMG, RPV failure can be prevented depending on the method of RCS depressurization. (2) To reduce the release of fission product during the injecting into SGs, a temporary opening of SDS before the injecting into SGs was suggested. These modified sequences of mitigation actions can reduce the release of fission product and the adverse effect of SDS.

Development on Dose Assessment Model of Northeast Asia Nuclear Accident Simulator

Energy Technology Data Exchange (ETDEWEB)

Kim, Ju Yub; Kim, Ju Youl; Kim, Suk Hoon; Lee, Seung Hee; Yoon, Tae Bin [FNC Techology, Yongin (Korea, Republic of)

2016-05-15

In order to support the emergency response system, the simulator for overseas nuclear accident is under development including source-term estimation, atmospheric dispersion modeling and dose assessment. The simulator is named NANAS (Northeast Asia Nuclear Accident Simulator). For the source-term estimation, design characteristics of each reactor type should be reflected into the model. Since there are a lot of reactor types in neighboring countries, the representative reactors of China, Japan and Taiwan have been selected and the source-term estimation models for each reactor have been developed, respectively. For the atmospheric dispersion modeling, Lagrangian particle model will be integrated into the simulator for the long range dispersion modeling in Northeast Asia region. In this study, the dose assessment model has been developed considering external and internal exposure. The dose assessment model has been developed as a part of the overseas nuclear accidents simulator which is named NANAS. It addresses external and internal pathways including cloudshine, groundshine and inhalation. Also, it uses the output of atmospheric dispersion model (i.e. the average concentrations of radionuclides in air and ground) and various coefficients (e.g. dose conversion factor and breathing rate) as an input. Effective dose and thyroid dose for each grid in the Korean Peninsula region are printed out as a format of map projection and chart. Verification and validation on the dose assessment model will be conducted in further study by benchmarking with the measured data of Fukushima Daiichi Nuclear Accident.

DELPHI expert panel evaluation of Hanford high level waste tank failure modes and release quantities

Energy Technology Data Exchange (ETDEWEB)

Dunford, G.L.; Han, F.C.

1996-09-30

The Failure Modes and Release Quantities of the Hanford High Level Waste Tanks due to postulated accident loads were established by a DELPHI Expert Panel consisting of both on-site and off-site experts in the field of Structure and Release. The Report presents the evaluation process, accident loads, tank structural failure conclusion reached by the panel during the two-day meeting.

Interactive Rapid Dose Assessment Model (IRDAM): reactor-accident assessment methods. Vol.2

International Nuclear Information System (INIS)

Poeton, R.W.; Moeller, M.P.; Laughlin, G.J.; Desrosiers, A.E.

1983-05-01

As part of the continuing emphasis on emergency preparedness, the US Nuclear Regulatory Commission (NRC) sponsored the development of a rapid dose assessment system by Pacific Northwest Laboratory (PNL). This system, the Interactive Rapid Dose Assessment Model (IRDAM) is a micro-computer based program for rapidly assessing the radiological impact of accidents at nuclear power plants. This document describes the technical bases for IRDAM including methods, models and assumptions used in calculations. IRDAM calculates whole body (5-cm depth) and infant thyroid doses at six fixed downwind distances between 500 and 20,000 meters. Radionuclides considered primarily consist of noble gases and radioiodines. In order to provide a rapid assessment capability consistent with the capacity of the Osborne-1 computer, certain simplifying approximations and assumptions are made. These are described, along with default values (assumptions used in the absence of specific input) in the text of this document. Two companion volumes to this one provide additional information on IRDAM. The user's Guide (NUREG/CR-3012, Volume 1) describes the setup and operation of equipment necessary to run IRDAM. Scenarios for Comparing Dose Assessment Models (NUREG/CR-3012, Volume 3) provides the results of calculations made by IRDAM and other models for specific accident scenarios

Underreporting of maritime accidents to vessel accident databases.

Science.gov (United States)

Hassel, Martin; Asbjørnslett, Bjørn Egil; Hole, Lars Petter

2011-11-01

Underreporting of maritime accidents is a problem not only for authorities trying to improve maritime safety through legislation, but also to risk management companies and other entities using maritime casualty statistics in risk and accident analysis. This study collected and compared casualty data from 01.01.2005 to 31.12.2009, from IHS Fairplay and the maritime authorities from a set of nations. The data was compared to find common records, and estimation of the true number of occurred accidents was performed using conditional probability given positive dependency between data sources, several variations of the capture-recapture method, calculation of best case scenario assuming perfect reporting, and scaling up a subset of casualty information from a marine insurance statistics database. The estimated upper limit reporting performance for the selected flag states ranged from 14% to 74%, while the corresponding estimated coverage of IHS Fairplay ranges from 4% to 62%. On average the study results document that the number of unreported accidents makes up roughly 50% of all occurred accidents. Even in a best case scenario, only a few flag states come close to perfect reporting (94%). The considerable scope of underreporting uncovered in the study, indicates that users of statistical vessel accident data should assume a certain degree of underreporting, and adjust their analyses accordingly. Whether to use correction factors, a safety margin, or rely on expert judgment, should be decided on a case by case basis. Copyright © 2011 Elsevier Ltd. All rights reserved.

'What on earth can this possibly mean'? French reentry courts and experts' risk assessment.

Science.gov (United States)

Herzog-Evans, Martine

2016-01-01

Against the backdrop of ten years of punitive criminal justice policies, the number of cases in which risk assessments by psychiatrist experts are mandatory has considerably increased in France. Because of complex and deeply ingrained cultural factors, most experts and academics oppose the use of actuarial or other structured judgement tools, which they assimilate to these policy changes. Parallel to this, the reentry judges in charge of making release and other community sentence decisions have maintained a strong rehabilitative and desistance-focused culture. Drawing on interviews with these judges and experts, the author wanted to assess the judges' expectations of experts' reports, their opinion on actuarial tools, and how they perceived experts and their aptitude to assess risk. The study showed that French reentry judges manage to keep experts' conclusions at bay when they do not fit with their desistance goals, as they can draw upon their own expertise and that of probation services. They do not have much faith in the professionalism and methodology of experts, and would like them to better demonstrate how they reach their conclusions. Moreover, criminogenic needs assessment would be much more useful to them than static risk assessment, which raises the issue as to why this is not the French probation services' role. Reentry judges who never encountered a report which uses a structured tool are influenced by the French ideological debate; those who have read such reports are unanimously in favour of such tools. It thus seems clear that they would like experts to be more strongly guided by science, but are not yet fully aware of what this entails. Copyright © 2015 Elsevier Ltd. All rights reserved.

Reflection of risk communication in 9 months after Fukushima nuclear accident

International Nuclear Information System (INIS)

Ogino, Haruyuki

2012-01-01

At the third symposium corresponding to the Fukushima accident held by the Japan Health Physics Society (JHPS) in December 2011, Youth Researcher's Association of JHPS presented 'risk communication in 9 months after Fukushima accident'. Focusing on the presentation, this article summarized social response of radiation risk in 9 months and activities of 'Q and A site that provides public with questions and answers related to radiation and daily life' by Society members as radiation protection experts after the Fukushima accident. The author raised some two problems related with (1) what and how should be 'judgments of best experts' reflecting the BSE case as 'trans-science' problem and (2) what's 'recommendation/standards document to be referred to' harmonizing with trend of international standards. Risk communication network should be not only between 'experts' and 'public' but also including international organizations. (T. Tanaka)

Factors associated with non-return to work in the severely injured victims 3 years after a road accident: A prospective study.

Science.gov (United States)

Pélissier, C; Fort, E; Fontana, L; Charbotel, B; Hours, M

2017-09-01

Road accidents may impact victims' physical and/or mental health and socio-occupational life, particularly the capacity to return to work. The purpose of our study is to assess modifiable medical and socio-occupational factors of non-return to work in the severely injured 3 years after a road accident. Among1,168 road accidents casualties in the Rhône administrative Département of France followed for five years, 141 of the 222 severely injured (Maximal Abbreviated Injury Scale ≥ 3) aged more than 16 years who were in work at the time of the accident, reported whether they had returned to work in the 3 years following the accident. The subgroups of those who had (n=113) and had not returned to work (n=28) were compared for socio-occupational (gender, age, educational level, marital status, socio-occupational group) accident-related medical factors (type of road user, type of journey, responsibility in the accident, initial care) and post-accident medical factors (pain intensity, post-traumatic stress disorder, physical sequelae, quality of life) by using standardized tools. Severity of initial head, face and lower-limb injury, intense persistent pain, post-traumatic stress disorder, poor self-assessed quality of life and health status at 3 years were associated with non-return to work on univariate analysis. On multivariate analysis, severity of initial head and lower-limb injury, intense persistent pain at 3 years and post-traumatic stress disorder were significantly associated with non-return to work 3 years following severe road-accident injury. Post-traumatic stress disorder and chronic pain were essential modifiable medical determinants of non-return to work in the severely injured after a road accident: early adapted management could promote return to work in the severely injured. Improve early adapted treatment of pain and PTSD in the rehabilitation team should help the severely injured return to work following a road accident. Copyright © 2017 Elsevier

Human Factors in Nuclear Reactor Accidents

International Nuclear Information System (INIS)

Mustafa, M.E.

2016-01-01

While many people would blame nature for the disaster of the “Fukushima Daiichi” accident, experts considered this accident to be also a human-induced disaster. This confirmed the importance of human errors which have been getting a growing interest in the nuclear field after the Three Mile Island accident. Personnel play an important role in design, operation, maintenance, planning, and management. The interface between machine and man is known as a human factor. In the present work, the human factors that have to be considered were discussed. The effect of the control room configuration and equipment design effect on the human behavior was also discussed. Precise reviewing of person’s qualifications and experience was focused. Insufficient training has been a major cause of human error in the nuclear field. The effective training issues were introduced. Avoiding complicated operational processes and non responsive management systems was stressed. Distinguishing between the procedures for normal and emergency operations was emphasised. It was stated that human error during maintenance and testing activities could cause a serious accident. This is because safety systems do not cover much more risk probabilities in the maintenance and testing activities like they do in the normal operation. In nuclear industry, the need for a classification and identification of human errors has been well recognised. As a result of this, human reliability must be assessed. These errors are analyzed by a probabilistic safety assessment which deals with errors in reading, listening and implementing procedures but not with cognitive errors. Much efforts must be accomplished to consider cognitive errors in the probabilistic safety assessment. The ways of collecting human factor data were surveyed. The methods for identifying safe designs, helping decision makers to predict how proposed or current policies will affect safety, and comprehensive understanding of the relationship

Assessment of ICARE/CATHARE V1 Severe Accident Code

International Nuclear Information System (INIS)

Chatelard, Patrick; Fleurot, Joelle; Marchand, Olivier; Drai, Patrick

2006-01-01

The ICARE/CATHARE code system has been developed by the French 'Institut de Radioprotection et de Surete Nucleaire' (IRSN) in the last decade for the detailed evaluation of Severe Accident (SA) consequences in a primary system. It is composed of the coupling of the core degradation IRSN code ICARE2 and of the thermal-hydraulics French code CATHARE2. It has been extensively used to support the level 2 Probabilistic Safety Assessment (PSA-2) of the 900 MWe PWR. This paper presents the synthesis of the ICARE/CATHARE V1 assessment which was conducted in the frame of the 'International ICARE/CATHARE Users' Club', under the management of IRSN. The ICARE/CATHARE V1 validation matrix is composed of more than 60 experiments, distributed in few thermal-hydraulics non-regression tests (to handle the front end phase of a severe accident), numerous Separate-Effect Tests, about 30 Integral Tests covering both the early and the late degradation phases, as well as a 'circuit' experiment including hydraulics loops. Finally, the simulation of the TMI-2 accident was also added to assess the code against real conditions. This validation task was aimed at assessing the ICARE/CATHARE V1 capabilities (including the stand-alone ICARE2 V3mod1 version) and also at proposing recommendations for an optimal use of this version ('Users' Guidelines'). Thus, with a correct account for the recommended guidelines, it appeared that the last ICARE/CATHARE V1 version could be reasonably used to perform best-estimate reactor studies up to a large corium slumping into the lower head. (authors)

Selected source term topics. Report to CSNI by an OECD/NEA Group of experts

International Nuclear Information System (INIS)

1987-04-01

CSNI Report 136 summarizes the results of the work performed by the Group of Experts on the Source Term and Environmental Consequences (PWG4) during the period extending from 1983 and 1986. This report is complementary to Part 1, 'Technical Status of the Source Term' of CSNI Report 135, 'Report to CSNI on Source Term Assessment, Containment atmosphere control systems, and accident consequences'; it considers in detail a number of very specific issues thought to be important in the source term area. It consists of: an executive summary (prepared by the Chairman of the Group), a section on conclusions and recommendations, and five technical chapters (fission product chemistry in the primary circuit of a LWR during severe accidents; resuspension/re-entrainment of aerosols in LWRs following a meltdown accident; iodine chemistry under severe accident conditions; effects of combustion, steam explosions and pressurized melt ejection on fission product behaviour; radionuclide removal by pool scrubbing), a technical annex and two appendices

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1988-02-15

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

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

Jules Horowitz reactor - Complementary safety assessment in the light of the Fukushima accident

International Nuclear Information System (INIS)

2011-01-01

This CSA (Complementary Safety Assessment) analyses the robustness of the Jules Horowitz reactor (RJH) to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. RJH is being built on the Cadarache CEA's site. Robustness is the ability for the facility to withstand events beyond the level for which the facility was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence (cliff edge effect). Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like RJH's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. This report is divided into 9 main chapters: 1) main features of the RJH facility, 2) identification of cliff edge risks and of equipment essential for safety, 3) earthquake risk, 4) flood risk, 5) risks due to other extreme natural disasters, 6) the loss of electrical power supplies and of cooling systems, 7) management of severe accidents, 8) subcontracting policy, 9) synthesis and list of improvements. This study shows a globally good robustness of the RJH for the considered risks. Nevertheless it can considered relevant to increase the robustness of the plant on a few points: -) to increase the seismic safety margins of some pieces of equipment, -) to increase the robustness of the internal electrical power supplies, -) to increase the fuel cooling capacity, and -) to improve the management of the post-accidental period. (A.C.)

The fuzzy set theory application to the analysis of accident progression event trees with phenomenological uncertainty issues

International Nuclear Information System (INIS)

Chun, Moon-Hyun; Ahn, Kwang-Il

1991-01-01

Fuzzy set theory provides a formal framework for dealing with the imprecision and vagueness inherent in the expert judgement, and therefore it can be used for more effective analysis of accident progression of PRA where experts opinion is a major means for quantifying some event probabilities and uncertainties. In this paper, an example application of the fuzzy set theory is first made to a simple portion of a given accident progression event tree with typical qualitative fuzzy input data, and thereby computational algorithms suitable for application of the fuzzy set theory to the accident progression event tree analysis are identified and illustrated with example applications. Then the procedure used in the simple example is extended to extremely complex accident progression event trees with a number of phenomenological uncertainty issues, i.e., a typical plant damage state 'SEC' of the Zion Nuclear Power Plant risk assessment. The results show that the fuzzy averages of the fuzzy outcomes are very close to the mean values obtained by current methods. The main purpose of this paper is to provide a formal procedure for application of the fuzzy set theory to accident progression event trees with imprecise and qualitative branch probabilities and/or with a number of phenomenological uncertainty issues. (author)

Instrumentation Performance during the TMI-2 Accident

International Nuclear Information System (INIS)

Rempe, Joy L.; Knudson, Darrell L.

2013-06-01

The accident at the Three Mile Island Unit 2 (TMI- 2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focused upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this paper. As noted within this paper, several techniques were invoked in the TMI-2 post-accident program to evaluate sensor survivability status and data qualification, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this paper provides recommendations related to sensor survivability and the data evaluation process that could be implemented in upcoming Fukushima Daiichi recovery efforts. (authors)

Elicitation and use of expert judgment in performance assessment for high-level radioactive waste repositories

International Nuclear Information System (INIS)

Bonano, E.J.; Hora, S.C.; Keeney, R.L.; von Winterfeldt, D.

1990-05-01

This report presents the concept of formalizing the elicitation and use of expert judgment in the performance assessment of high-level radioactive waste (HLW) repositories in deep geologic formations. The report begins with a discussion of characteristics (advantages and disadvantages) of formalizing expert judgment examples of previous uses of expert judgment in radioactive waste programs, criteria that can assist in deciding when to formalize expert judgment, and the relationship of formal use of expert judgment to data collection and modeling. The current state of the art with respect to the elicitation, use, and communication of formal expert judgment is presented. The report concludes with a discussion on potential applications of formal expert judgment in performance assessment of HLW repositories. 93 refs

Expert elicitation for deriving input data for probabilistic risk assessment of shipwrecks.

Science.gov (United States)

Landquist, H; Norrman, J; Lindhe, A; Norberg, T; Hassellöv, I-M; Lindgren, J F; Rosén, L

2017-12-15

The necessity of having a process in place for adequate risk assessment of shipwrecks that pose a threat to the marine environment is today internationally acknowledged. However, retrieving the desired data for such a risk assessment can prove challenging. One means of addressing this problem is to make use of experts' knowledge and experience. The purpose of this paper is therefore to present and analyse data for risk assessment of shipwrecks derived by expert elicitation. The main outcome is the experts' estimations of (i) the generic probability of an opening in a shipwreck due to the occurrence of a number of activities and (ii) estimations of the degree to which site-specific and wreck-specific indicators affect the probability of opening. Results show that the derived information is applicable in probabilistic shipwreck risk assessment and that the VRAKA framework now contains needed information for integrating generic and site-specific information using Bayesian updating. Copyright © 2017 Elsevier Ltd. All rights reserved.

Environmental Impact Assessment following a Nuclear Accident to a Candu NPP

International Nuclear Information System (INIS)

Margeanu, C.A.; Margeanu, S.; Olteanu, Gh.

2009-01-01

The paper presents calculations of nuclear accident consequences to public and environment, for a Candu NPP using advanced fuel in two hypothetical accident scenarios: (1) large LOCA followed by partial core melting with early containment failure; (2) late core disassembly and containment bypass through ECCS. During both accidents a release occurs, radioactive contaminants being dispersed into atmosphere. As reference, estimations for Candu standard UO 2 fuel were used. The radioactive core inventory was obtained by using ORIGEN-S computer code included in ORNL,SCALE 5 programs package. Radiological consequences assessment to public and environment was performed by means of PC COSYMA computer code

Uncertainty and sensitivity analysis in nuclear accident consequence assessment

International Nuclear Information System (INIS)

Karlberg, Olof.

1989-01-01

This report contains the results of a four year project in research contracts with the Nordic Cooperation in Nuclear Safety and the National Institute for Radiation Protection. An uncertainty/sensitivity analysis methodology consisting of Latin Hypercube sampling and regression analysis was applied to an accident consequence model. A number of input parameters were selected and the uncertainties related to these parameter were estimated within a Nordic group of experts. Individual doses, collective dose, health effects and their related uncertainties were then calculated for three release scenarios and for a representative sample of meteorological situations. From two of the scenarios the acute phase after an accident were simulated and from one the long time consequences. The most significant parameters were identified. The outer limits of the calculated uncertainty distributions are large and will grow to several order of magnitudes for the low probability consequences. The uncertainty in the expectation values are typical a factor 2-5 (1 Sigma). The variation in the model responses due to the variation of the weather parameters is fairly equal to the parameter uncertainty induced variation. The most important parameters showed out to be different for each pathway of exposure, which could be expected. However, the overall most important parameters are the wet deposition coefficient and the shielding factors. A general discussion of the usefulness of uncertainty analysis in consequence analysis is also given. (au)

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.

Expert assessment of the current state of the energy management system in the company

Science.gov (United States)

Minnullina, Anna; Abdrazakov, Rais

2017-10-01

The authors’ expert assessment of the current state of the energy management system in the company is proposed in the article. The experts are invited to assess the status of the energy management system in the following categories: energy policy, organizational structure, training, motivation, control, communication, investment, and energy consumption culture. For the purposes of interpretation of the results of the expert evaluation obtained, a gradation based on a possible range of values is proposed. The expert evaluation allows representing the status of the energy management system in general and at each of its individual levels, which makes it possible to identify the problem areas more accurately. To confirm the applied nature of the proposed methodology, the authors assessed the opinions of 8 experts, employed by the road construction company of the Tyumen Region and related in one way or another to the process of energy consumption in the company due to the nature of their activities.

Lessons learned from the Fukushima Dai-ichi accident and responses in NRA regulatory requirements

International Nuclear Information System (INIS)

Fuketa, Toyoshi

2014-01-01

The author would like to present significant lessons learned from the TEPCO’s Fukushima Dai-ichi accident and responses in regulatory requirements developed by the Nuclear Regulation Authority for power-producing light water reactors. The presentation will cover prevention of structures, systems and components failures, measures to prevent common cause failures, prevention of core damage, mitigation of severe accidents, emergency preparedness, continuous improvement of safety, use of probabilistic risk assessment, and post-accident regulation on the Fukushima Dai-ichi. (author)

Computer Based Road Accident Reconstruction Experiences

Directory of Open Access Journals (Sweden)

Milan Batista

2005-03-01

Full Text Available Since road accident analyses and reconstructions are increasinglybased on specific computer software for simulationof vehicle d1iving dynamics and collision dynamics, and forsimulation of a set of trial runs from which the model that bestdescribes a real event can be selected, the paper presents anoverview of some computer software and methods available toaccident reconstruction experts. Besides being time-saving,when properly used such computer software can provide moreauthentic and more trustworthy accident reconstruction, thereforepractical experiences while using computer software toolsfor road accident reconstruction obtained in the TransportSafety Laboratory at the Faculty for Maritime Studies andTransport of the University of Ljubljana are presented and discussed.This paper addresses also software technology for extractingmaximum information from the accident photo-documentationto support accident reconstruction based on the simulationsoftware, as well as the field work of reconstruction expertsor police on the road accident scene defined by this technology.

Application of fuzzy expert system on LILW performance assessment

International Nuclear Information System (INIS)

Lemos, F.L. de; Sullivan, T.

2002-01-01

A complete LILW repository performance assessment requires the involvement between several experts in many fields of science. Many sources of uncertainties arise due to complexity of interaction of environmental parameters, lack of data and ignorance, this makes predictive analysis and interpretation difficult. This difficulty in understanding the impact of the ambiguities is even higher when it comes to public and decision makers involvement. Traditional methods of data analysis, while having strong mathematical basis, many times are not adequate to deal with ambiguous data. These ambiguities can be an obstacle to make the results easier to understand and defensible. A methodology of decision making, based on fuzzy logic, can help the interaction between experts, decision makers and the public. This method is the basis of an expert system which can help the analysis of very complex and ambiguous processes. (author)

Cost-effectiveness analysis of countermeasures using accident consequence assessment models

International Nuclear Information System (INIS)

Alonso, A.; Gallego, E.

1987-01-01

In the event of a large release of radionuclides from a nuclear power plant, protective actions for the population potentially affected must be implemented. Cost-effectiveness analysis will be useful to define the countermeasures and the criteria needed to implement them. This paper shows the application of Accident Consequence Assessment (ACA) models to cost-effectiveness analysis of emergency and long-term countermeasures, making use of the different relationships between dose, contamination levels, affected areas and population distribution, included in such a model. The procedure is illustrated with the new Melcor Accident Consequence Code System (MACCS 1.3), developed at Sandia National Laboratories (USA), for a fixed accident scenario. Different alternative actions are evaluated with regard to their radiological and economical impact, searching for an 'optimum' strategy. (author)

Engaging the expert in public dialogue: Developing the role of performance assessment

International Nuclear Information System (INIS)

Littleboy, Anna; Hodgkinson, David

2001-01-01

Participatory approaches to decision-making call for two-way dialogue with a wide range of stakeholders. In the United Kingdom, probabilistic performance assessments of a potential deep geological disposal facility have typically been undertaken by experts to meet regulations set by other experts. However, there is potential for performance assessments to fulfil the additional role of enabling dialogue with other stakeholder groups. This paper suggests developments in the focus and presentation of performance assessments to foster such dialogue. It reflects issues relevant to performance assessment that arise as waste management organisations consider the consequences of involving wider stakeholder groups in decision making about repository development

Engaging the expert in public dialogue: Developing the role of performance assessment

Energy Technology Data Exchange (ETDEWEB)

Littleboy, Anna [United Kingdom Nirex Ltd., Harwell (United Kingdom); Hodgkinson, David [Quintessa Ltd., Henley-on-Thames (United Kingdom)

2001-07-01

Participatory approaches to decision-making call for two-way dialogue with a wide range of stakeholders. In the United Kingdom, probabilistic performance assessments of a potential deep geological disposal facility have typically been undertaken by experts to meet regulations set by other experts. However, there is potential for performance assessments to fulfil the additional role of enabling dialogue with other stakeholder groups. This paper suggests developments in the focus and presentation of performance assessments to foster such dialogue. It reflects issues relevant to performance assessment that arise as waste management organisations consider the consequences of involving wider stakeholder groups in decision making about repository development.

Assessment of the severity of asthma by an expert system. Description and evaluation.

Science.gov (United States)

Redier, H; Daures, J P; Michel, C; Proudhon, H; Vervloet, D; Charpin, D; Marsac, J; Dusser, D; Brambilla, C; Wallaert, B

1995-02-01

Asthmaexpert, an expert system (ES), was produced at the special request of several clinicians in order to better understand the medical decisions made clinical experts in managing an asthmatic patient. We describe and evaluate this knowledge base, focusing mainly on assessment of the severity of asthma. After compiling data from a patient, Asthmaexpert assesses the severity of the disease and identifies the trigger factors involved, suggests any further investigations that may be required, and offers a treatment strategy. Implemented with Nexpert and Hypercard, it runs on a MacIntosh personal computer. The validation stage involved eight clinical experts who provided 20 case report forms (CRF) with their conclusions about management of asthma. The CRF were then programmed into the ES, which provided its own conclusions about the same subjects. Afterward, all the experts evaluated the conclusions given by ES or by their colleagues in a double-blind manner. One hundred twenty-seven CRF were available. The reliability of the experts' opinions was good, with a substantial consensus between them when assessing severity scores (kappa = 0.27 to 0.54). There was no difference in concordance of opinions on severity scores either between the experts who designed the system and ES or between the other experts and ES (weighted kappa = 0.72 and 0.69, respectively). Experts judged that the severity scores given by ES were as good as those proposed by their colleagues, and that the overall conclusions given by ES were as good as or better than those given by their colleagues. The conclusions drawn by ES were given a good rating.(ABSTRACT TRUNCATED AT 250 WORDS)

Techniques and decision making in the assessment of off-site consequences of an accident in a nuclear facility

International Nuclear Information System (INIS)

1987-01-01

This Guide is intended to complement the IAEA's existing technical guidance on emergency planning and preparedness by providing information and practical guidance related to the assessment of off-site consequences of an accident in a nuclear or radioactive materials installation and to the decision making process in implementing protective measures. This Guide contains information on emergency response philosophy, fundamental factors affecting accident consequences, principles of accident assessment, data acquisition and handling, systems, techniques and decision making principles. Many of the accident assessment concepts presented are considerably more advanced than some of those that now pertain in most countries. They could, if properly interpreted, developed and applied, significantly improve emergency response in the early and intermediate phases of an accident. Furthermore, they are considered to be applicable to a broad range of serious nuclear accidents and radiological emergencies. The extent of their application is governed by both the scale of the accident and by the availability of preplanned resources for accident assessment and emergency response. 68 refs, 28 figs, 14 tabs

The importance of post-mortem computed tomography (PMCT) in confrontation with conventional forensic autopsy of victims of motorcycle accidents.

Science.gov (United States)

Moskała, Artur; Woźniak, Krzysztof; Kluza, Piotr; Romaszko, Karol; Lopatin, Oleksij

2016-01-01

Since traffic accidents are an important problem in forensic medicine, there is a constant search for new solutions to help with an investigation process in such cases. In recent years there was a rapid development of post-mortem imaging techniques, especially post-mortem computed tomography (PMCT). In our work we concentrated on a potential advantage of PMCT in cases of motorcycle accident fatalities. The results of forensic autopsy were compared with combined results of the autopsy and PMCT to check in which areas use of these two techniques gives statistically important increase in number of findings. The hypothesis was confirmed in case of pneumothorax and fractures of skull, spine, clavicle, scapula, lower leg bones. As for majority of other bone fractures locations and brain injures there were single cases with pathologies visible only in PMCT, but too few to reach expected level of p-value. In case of injuries of solid organs and soft tissues statistical analysis did not confirmed any advantage of unenhanced PMCT use. On the whole it has been shown that PMCT used as an adjunct to forensic autopsy can cause an increase in information about vitally important regions in case of motorcycle accident fatalities. Copyright © 2015 Elsevier Ireland Ltd. All rights reserved.

Improvement of the following accident dose assessment system

Energy Technology Data Exchange (ETDEWEB)

Kim, Enn Han; Han, Moon Hee; Suh, Kyung Suk; Hwang, Won Tae; Choi, Young Gil [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1999-12-15

The FADAS has been updates for calculating the real-time wind fields continuously at the nuclear sites in Korea. The system has been constructed to compute the wind fields using its own process for the dummy meteorological data, and dose not effect on the overall wind field module. If the radioactive materials are released into the atmosphere in real situation, the calculations of wind fields and exposure dose in the previous FADAS are performed in the case of the recognition of the above situation in the source term evaluation module. The current version of FADAS includes the program for evaluating the effect of the predicted accident and the assumed scenario together. The dose assessment module is separated into the real-time and the supposed accident respectively.

Chernobyl radiological data for accident consequence assessment

International Nuclear Information System (INIS)

Bottino, A.; Sacripanti, A.

1989-01-01

In this draft is presented the results of a first effort to summarize information related to the radionuclides behaviour in rural areas, in order to estimate pathway parameters to assess accident consequences. This topic encloses relevant aspects concerning contamination of rural environment, the most important being: 1) dry deposition velocities; 2) washout coefficient; 3) accumulation in lakes; 4) migration in soil; 5) winter conditions; 6) filtering effects of forests

Expert judgements in performance assessments. Report of an SKI/SSI seminar

International Nuclear Information System (INIS)

Wilmot, R.D.; Galson, D.A.; Hora, S.C.

2000-09-01

Expert judgements are an important element of all performance assessments and are made when alternative approaches to decision-making are not available or are not feasible. Decisions regarding the scope of a particular assessment or the type of modelling approach adopted must be made through judgements because there are no observations that can be made. Similarly, any assumptions concerning human activities in the far future are essentially speculative and must be based on expert judgement. Observations of spatial heterogeneity within a disposal system may, on the other hand, be theoretically possible but not be feasible because of excessive cost or because they would adversely affect the system they were intended to characterise. Because there is a wide range of judgements made within a performance assessment, there are several ways of making the judgements and of assessing how they have been made. Judgements may, for example, be made by individuals or by groups, and they may be formally elicited or made without formal elicitation. Dialogue with stake holders can be an important means of assessing judgements, as can peer review. Documentation is a key element throughout the process of making and reviewing judgements, and appropriate quality assurance procedures can also build confidence in judgements. In order to develop an understanding of the processes of making and assessing judgements, the Swedish Nuclear Power Inspectorate (SKI) and the Swedish Radiation Protection Institute (SSI) jointly sponsored a seminar entitled 'The Use of Expert Judgements in Performance Assessments'. The seminar was held in Norrtaelje, Sweden, on 17-19 January 2000. The seminar was organised by Galson Sciences Ltd (GSL) on behalf of SKI, and conducted jointly by GSL and Professor Steve Hora of the University of Hawaii. Participants at the seminar included SKI and SSI staff and independent experts. A key element of the seminar was an illustrative expert elicitation session, designed to

Expert judgements in performance assessments. Report of an SKI/SSI seminar

Energy Technology Data Exchange (ETDEWEB)

Wilmot, R.D.; Galson, D.A. [Galson Sciences Ltd, Oakham (United Kingdom); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States)

2000-09-01

Expert judgements are an important element of all performance assessments and are made when alternative approaches to decision-making are not available or are not feasible. Decisions regarding the scope of a particular assessment or the type of modelling approach adopted must be made through judgements because there are no observations that can be made. Similarly, any assumptions concerning human activities in the far future are essentially speculative and must be based on expert judgement. Observations of spatial heterogeneity within a disposal system may, on the other hand, be theoretically possible but not be feasible because of excessive cost or because they would adversely affect the system they were intended to characterise. Because there is a wide range of judgements made within a performance assessment, there are several ways of making the judgements and of assessing how they have been made. Judgements may, for example, be made by individuals or by groups, and they may be formally elicited or made without formal elicitation. Dialogue with stake holders can be an important means of assessing judgements, as can peer review. Documentation is a key element throughout the process of making and reviewing judgements, and appropriate quality assurance procedures can also build confidence in judgements. In order to develop an understanding of the processes of making and assessing judgements, the Swedish Nuclear Power Inspectorate (SKI) and the Swedish Radiation Protection Institute (SSI) jointly sponsored a seminar entitled 'The Use of Expert Judgements in Performance Assessments'. The seminar was held in Norrtaelje, Sweden, on 17-19 January 2000. The seminar was organised by Galson Sciences Ltd (GSL) on behalf of SKI, and conducted jointly by GSL and Professor Steve Hora of the University of Hawaii. Participants at the seminar included SKI and SSI staff and independent experts. A key element of the seminar was an illustrative expert elicitation session

Exploring barriers and facilitators to the clinical use of virtual reality for post-stroke unilateral spatial neglect assessment.

Science.gov (United States)

Ogourtsova, Tatiana; Archambault, Philippe S; Lamontagne, Anouk

2017-11-07

Hemineglect, defined as a failure to attend to the contralesional side of space, is a prevalent and disabling post-stroke deficit. Conventional hemineglect assessments lack sensitivity as they contain mainly non-functional tasks performed in near-extrapersonal space, using static, two-dimensional methods. This is of concern given that hemineglect is a strong predictor for functional deterioration, limited post-stroke recovery, and difficulty in community reintegration. With the emerging field of virtual reality, several virtual tools have been proposed and have reported better sensitivity in neglect-related deficits detection than conventional methods. However, these and future virtual reality-based tools are yet to be implemented in clinical practice. The present study aimed to explore the barriers/facilitators perceived by clinicians in the use of virtual reality for hemineglect assessment; and to identify features of an optimal virtual assessment. A qualitative descriptive process, in the form of focus groups, self-administered questionnaire and individual interviews was used. Two focus groups (n = 11 clinicians) were conducted and experts in the field (n = 3) were individually interviewed. Several barriers and facilitators, including personal, institutional, client suitability, and equipment factors, were identified. Clinicians and experts in the field reported numerous features for the virtual tool optimization. Factors identified through this study lay the foundation for the development of a knowledge translation initiative towards an implementation of a virtual assessment for hemineglect. Addressing the identified barriers/facilitators during implementation and incorporating the optimal features in the design of the virtual assessment could assist and promote its eventual adoption in clinical settings. Implications for rehabilitation A multimodal and active knowledge translation intervention built on the presently identified modifiable factors is

Cleanup and decommissioning of a nuclear reactor after a severe accident

International Nuclear Information System (INIS)

1992-01-01

Although the development of commercial nuclear power plants has in general been associated with an excellent record of nuclear safety, the possibility of a severe accident resulting in major fuel and core damage cannot be excluded and such accidents have in fact already occurred. For over a decade, IAEA publications have provided technical guidance and recommendations for post-accident planning to be considered by appropriate authorities. Guidance and recommendations have recently been published on the management of damaged nuclear fuel, sealing of the reactor building and related safety and performance assessment aspects. The present technical report on the cleanup and decommissioning of reactors which have undergone a severe accident represents a further publication in the series. Refs, figs and tabs.

Dose assessment around TR-2 reactor due to maximum credible accident

International Nuclear Information System (INIS)

Turgut, M. H.; Adalioglu, U.; Aytekin, A.

2001-01-01

The revision of safety analysis report of TR-2 research reactor had been initiated in 1995. The whole accident analysis and accepted scenario for maximum credible accident has been revised according to the new safety concepts and the impact to be given to the environment due to this scenario has been assessed. This paper comprises all results of these calculations. The accepted maximum credible accident scenario is the partial blockage of the whole reactor core which resulted in the release of 25% of the core inventory. The DOSER code which uses very conservative modelling of atmospheric distributions were modified for the assessment calculations. Pasquill conditions based on the local weather observations, topography, and building affects were considered. The thyroid and whole body doses for 16 sectors and up to 10 km of distance around CNAEM were obtained. Release models were puff and a prolonged one of two hours of duration. Release fractions for the active isotopes were chosen from literature which were realistic

The TEX-I real-time expert system applied to situation assessment for the SNR-300 reactor

International Nuclear Information System (INIS)

Schmal, N.; Doerbecker, K.; Leder, H.J.; Rueckert, M.; Schade, H.J.

1990-01-01

Within the German TEX-I Project, which was sponsored by the Federal Ministry for Research and Technology, several companies developed industrial applications of technical expert systems for data interpretation, diagnosis and process control. The purpose of the diagnosis expert system reported here is to support the operators of the LMFBR SNR-300 in assessing plant status in real-time, based on readings from a large number of sensors. By online connection to the process control computer, it can monitor all incoming signal values, check the consistency of data, continuously diagnose the current plant status, detect unusual trends prior to accidents, localize faulty components and recommend operators response in abnormal conditions. The systems architecture consists of two basic subsystems, an inference engine and an intelligent process interface, implemented in Lisp on a Symbolics-Workstation. The inference engine has been derived from BABYLON, a hybrid shell developed by the German computer research institute GMD. This shell includes rules, prologue, constraints and an object oriented frame processor. The extended version has a component description language and a top-down diagnosis scheme including a mechanism of attention focusing. Inference run as independent, quasi-parallel processes. These so-called inference tasks can interrupt or abort each other, if higher priority events must be processed. The complete system is modelled in an object oriented matter and is divided into several subsystems and each subsystem into its physical components. At present the expert system is connected to a real-time simulation of the reactor. The simulation is based on a thermohydraulic code for simulation of the transient behaviour of temperatures and flow rates in the reactor core, plena, pipes, pumps, valves, intermediate heat exchangers and cooling components. Additionally, the systems response to an asynchronous operator interaction can be simulated. Several types of anomalies

Accident investigation practices in Europe--main responses from a recent study of accidents in industry and transport.

Science.gov (United States)

Roed-Larsen, Sverre; Valvisto, T; Harms-Ringdahl, L; Kirchsteiger, C

2004-07-26

Europe has during recent years been shocked by disasters from natural events and technical breakdowns. The consequences have been comprehensive, measured by lost lives, injuries, and material and environmental damage. ESReDA wanted in 2000--by setting up a special expert group on accident investigation--to clarify the state of art of accident investigation practices and to map the use of thoroughly accident investigation in order to learn lessons from past disasters and prevent new ones. The scope was to cover three sectors in the society: transport, production processes and storage of hazardous materials, and energy production. The main method used was a questionnaire, which was sent in 2001 to about 150 organisations. About 50 replies were analysed. The replies showed great variations but also similarities, among others in definition of accident and incident, the objectives of the investigation team, criteria used to start an investigation, the status of the investigation organisation, the flow of information, the composition of the investigation team, and the use of internal or international procedures or rules. Several methods (in total 14 different methods were mentioned) were used for carrying out accident /incident investigations. Most of the respondents were willing to co-operate in one or another way with ESReDA. Although there are important biases in the material, the results from questionnaire are important inputs to the future work of ESReDA Expert group in this field. 3 safety approaches have been identified.

Expert assessment of preventive measures in the zone of guaranteed voluntary resettlement.

Science.gov (United States)

Ozerova, I I; Prilipko, V A; Myshkovska, A A

2013-01-01

An expert evaluation of preventive measures focused on population health maintenance in the zone of guaranteed voluntary resettlement taking into account the limitations of the provisions of certain articles of the Law of Ukraine on the clean-up of the Chornobyl disaster aftermath. The research survey was applied using a peer review method. The three-level (national, regional, district) expert workgroup was involved been employed within recovery of the Chornobyl disaster aftermath. Professional experience of minimum 10 years, corresponding profile and current position were the requirements for selection. A simple streamlining the values of paired or consistent comparison was applied on the primary empirical data. The arrangements made in recent years by the Cabinet of Ministers of Ukraine and ministries engaged in clean-up and recovery from the Chornobyl nuclear power plant accident are ineffective. Costs from the Fund for the implementation of measures to clean-up the consequences of the Chornobyl disaster and social protection are reduced. There is no consensus among experts regarding some constraint of an action of certain articles of the Law of Ukraine "On the status and social protection of citizens affected by the Chornobyl disaster" on the principles of social protection and healthcare of the population living in contaminated areas. In addressing the social and health consequences of the Chornobyl disaster the Verkhovna Rada of Ukraine just for a long time has passes no amendment to the legislation on the subject. At that there are several restrictions in force for a large number of articles of the existing laws that were adopted in 1991. The mentioned above exacerbates social strain and destroys the "citizen-state relationship". The extreme need has emerged in a comprehensive public review of existing laws with involvement of multidisciplinary group of experts to a "round table" to justify the draft laws. Ozerova Ju. Ju., Prylypko V. A., Myshkovska A. A

Health risk assessment for nanoparticles: A case for using expert judgment

International Nuclear Information System (INIS)

Kandlikar, Milind; Ramachandran, Gurumurthy; Maynard, Andrew; Murdock, Barbara; Toscano, William A.

2007-01-01

Uncertainties in conventional quantitative risk assessment typically relate to values of parameters in risk models. For many environmental contaminants, there is a lack of sufficient information about multiple components of the risk assessment framework. In such cases, the use of default assumptions and extrapolations to fill in the data gaps is a common practice. Nanoparticle risks, however, pose a new form of risk assessment challenge. Besides a lack of data, there is deep scientific uncertainty regarding every aspect of the risk assessment framework: (a) particle characteristics that may affect toxicity; (b) their fate and transport through the environment; (c) the routes of exposure and the metrics by which exposure ought to be measured; (d) the mechanisms of translocation to different parts of the body; and (e) the mechanisms of toxicity and disease. In each of these areas, there are multiple and competing models and hypotheses. These are not merely parametric uncertainties but uncertainties about the choice of the causal mechanisms themselves and the proper model variables to be used, i.e., structural uncertainties. While these uncertainties exist for PM2.5 as well, risk assessment for PM2.5 has avoided dealing with these issues because of a plethora of epidemiological studies. However, such studies don't exist for the case of nanoparticles. Even if such studies are done in the future, they will be very specific to a particular type of engineered nanoparticle and not generalizable to other nanoparticles. Therefore, risk assessment for nanoparticles will have to deal with the various uncertainties that were avoided in the case of PM2.5. Consequently, uncertainties in estimating risks due to nanoparticle exposures may be characterized as 'extreme'. This paper proposes a methodology by which risk analysts can cope with such extreme uncertainty. One way to make these problems analytically tractable is to use expert judgment approaches to study the degree of

Experiments for post accident hydrogen dispersion in F.M. vault using helium

International Nuclear Information System (INIS)

Bajaj, S.S.; Bhattacharyya, D.; Mishra, S.

1994-01-01

Under certain postulated accident scenarios involving a Loss of Coolant Accident (LOCA) simultaneous with impairment of Emergency Core Cooling (ECC), generation of hydrogen due to reaction between the zirconium clad and coolant is predicted in the coolant channel. The hydrogen generated in the coolant channels would eventually get released either in Fuelling Machine (FM) vault or in the pump room atmosphere depending on the location of the break. Analytical studies carried out so far to estimate the time dependent hydrogen concentration in the accident FM Vault consider the entire vault as a single volume. Tests were, therefore, planned to assess the mixing within the FM vault atmosphere with and without the availability of cooling fan units by releasing a known quantity of helium (instead of hydrogen) at selected locations and monitoring the relative concentration of helium in air at various locations. Test was conducted by releasing about 360 1 helium over a period of to 4 minutes at preselected locations and by measuring the relative concentration (leak rates indicated by helium leak detectors) at various locations in the FM vault. The results of cases with fans operating indicate repeatable and consistent trends of good mixing in the vault. For other cases (non turbulent, still condition) the results are sensitive to various factors including orientation of release. The former set of cases (turbulent. fans operating) are more relevant for postulated accident conditions. (author). 1 tab., 18 figs

Thermohydraulic accident behavior of reactors

International Nuclear Information System (INIS)

Horche, W.; Kirmse, R.; Reichenbach, D.; Weber, J.P.

1992-01-01

GRS, on behalf of the German Federal Ministry for the Environment, conducted an assessment of the technical safety of the Greifswald nuclear generating units of the Soviet WWER-440/W-230 and W-213 reactor lines, respectively. The evaluation of existing accident analyses and the execution of some first calculations by GRS added to the know-how of GRS. This is reflected in the increased participation by GRS in international expert bodies investigating safety problems of WWER-440 plants. The contributions made towards international WWER projects within the framework of IAEA missions or as a result of bilateral consultations strengthen international partnership in the field of reactor safety in Central and Eastern Europe. (orig.) [de

Assessment of off-site consequences of nuclear accidents (MARIA)

International Nuclear Information System (INIS)

Haywood, S.M.

1985-01-01

A brief report is given of a workshop held in Luxembourg in 1985 on methods for assessing the off-site radiological consequences of nuclear accidents (MARIA). The sessions included topics such as atmospheric dispersion; foodchain transfer; urban contamination; demographic and land use data; dosimetry, health effects, economic and countermeasures models; uncertainty analysis; and application of probabilistic risk assessment results as input to decision aids. (U.K.)

Colloquium / Preparation for nuclear post-accident. Citizens and local authorities involvement in major risks governance - proceedings

International Nuclear Information System (INIS)

Pascal, Michel; Villers, Anita; Sene, Monique; Godin, Francois; Quilichini, Jean-Michel; Rollinger, Francois; Delalonde, Jean-Claude; Mehl-Auget, Isabelle; Autret, Jean-Claude; Heriard Dubreuil, Gilles; Petitfrere, Michael; Lochard, Jacques; Demet, Michel; Boucherie, Jean-Claude; Calafat, Alexis; Sommade, Christian; Villain, Alain; Ragazzo, Romeo; Rengot, Marielle; Sename, Jean; Thellier, Yvette; Tremelet, David; Olszewski, Richard; Fournier, Nicolas

2009-12-01

During 40 years, France has chosen nuclear energy as main energy source for power generation. Today, nuclear energy covers 80% of the French electricity needs. For this reason, in France, each inhabitant lives at less than 200 km of a nuclear power plant. The September 11, 2001 terror attack has led to reconsider the nuclear risk in terms of security. In 2005, the French nuclear safety authority (ASN) has been entrusted with the preparation of a nuclear post-accident management study. This study has been mainly based on the experience feedback of the Chernobyl accident and on the dialogue with different actors of the French territory: the local authorities, the habitants, the associations and the health, environment and education actors. This colloquium represents the opening act of the joint elaboration of the management of a potential nuclear accident at the region scale. The colloquium was organized around 5 round tables: 1 - Did the nuclear safety transparency law change the situation?; 2 - Examples of pluralistic dialogues: the long-term management of a nuclear accident; 3 - How to encourage the necessary skills development of citizens? 4 - Regional development strategies in terms of nuclear risks management (incidents, effluents, wastes, transports..); 5 - New territory liabilities and citizens' legitimate aspiration in terms of health/environment: building something together. This document is the proceedings of this colloquium. It reports the exchanges between the participants

A methodology for the transfer of probabilities between accident severity categories

International Nuclear Information System (INIS)

Whitlow, J.D.; Neuhauser, K.S.

1992-01-01

Evaluation of the radiological risks of accidents involving vehicles transporting radioactive materials requires consideration of both accident probability and consequences. The probability that an accident will occur may be estimated from historical accident data for the given mode of transport. In addition to an overall accident rate, information regarding accident severity and the resulting package environments across the range of all credible accidents is needed to determine the potential for a release of radioactive material from the package or for an increase in direct radiation from the package caused by damage to packaging shielding. This information is usually obtained from a variety of sources such as historical data, experimental data, analyses of accident and package environments, and expert opinion. The consequences of an accident depend on a number of factors including the type, quantity, and physical form of radioactive material being transported; the response of the package to accident environments; the fraction of material released from the package; and the dispersion of any released material. One approach for the classification and treatment of transportation accidents in risk analysis divides the complete range of critical accident environments resulting from all credible accidents into some number of accident-severity categories. The types of accident environments that a package may be subjected to in transportation are often classified into the following five groups: impact, fire, crush, puncture, and immersion. A open-quotes criticalclose quotes accident environment is one of a type that could present a plausible threat to a package. Each severity category represents a portion of all credible accidents, and the total of all severity categories covers the complete range of critical accident environments. This approach is used in the risk assessment codes RADTRAN (Neuhauser and Kanipe 1992) and INTERTRAN (Ericsson and Elert 1983)

A critical assessment of energy accident studies

International Nuclear Information System (INIS)

Felder, Frank A.

2009-01-01

A comparison of two studies conducted ten years apart on energy accidents provides important insights into methodological issues and policy implications. Recommendations for further improvements in energy accident studies are developed including accounting for differences between average and incremental accident damages, testing for appropriate levels of aggregation of accidents, making references and databases publicly available, more precisely defining and reporting different types of economic damages, accounting for involuntary and voluntary risks, reporting normalized damages, raising broader public policy and planning implications and updating existing accident databases.

A critical assessment of energy accident studies

Energy Technology Data Exchange (ETDEWEB)

Felder, Frank A. [Edward J. Bloustein School of Planning and Public Policy, Rutgers, The State University of New Jersey, 33 Livingston Avenue, New Brunswick, NJ 08901 (United States)

2009-12-15

A comparison of two studies conducted ten years apart on energy accidents provides important insights into methodological issues and policy implications. Recommendations for further improvements in energy accident studies are developed including accounting for differences between average and incremental accident damages, testing for appropriate levels of aggregation of accidents, making references and databases publicly available, more precisely defining and reporting different types of economic damages, accounting for involuntary and voluntary risks, reporting normalized damages, raising broader public policy and planning implications and updating existing accident databases. (author)

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

International Nuclear Information System (INIS)

1986-11-01

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

Development of a Methodology for VHTR Accident Consequence Assessment

Energy Technology Data Exchange (ETDEWEB)

Lee, Joeun; Kim, Jintae; Jae, Moosung [Hanyang University, Seoul (Korea, Republic of)

2016-05-15

The substitution of the VHTR for burning fossil fuels conserves these hydrocarbon resources for other uses and eliminates the emissions of greenhouse. In Korea, for these reasons, constructing the VHTR plan for hydrogen production is in progress. In this study, the consequence analysis for the off-site releases of radioactive materials during severe accidents has been performed using the level 3 PRA technology. The offsite consequence analysis for a VHTR using the MACCS code has been performed. Since the passive system such as the RCCS(Reactor Cavity Cooling System) are equipped, the frequency of occurrence of accidents has been evaluated to be very low. For further study, the assessment for characteristic of VHTR safety system and precise quantification of its accident scenarios is expected to conduct more certain consequence analysis. This methodology shown in this study might contribute to enhancing the safety of VHTR design by utilizing the results having far lower effect on the environment than the LWRs.

Accident consequence assessment and siting criteria development

International Nuclear Information System (INIS)

Kollas, J.G.

1988-01-01

The methodology developed is based on assessing the average over a large spectrum of meteorological conditions whole body collective dose resulting from a severe reference accident. The assessment of this dose is performed by code CRAC.GAEC, the Greek A.E.C. version of code CRAC2. The collective dose, which is chosen as a measure of the social radiation risk, is compared to the dose corresponding to a level of social risk encountered historically in energy production as a whole. The outcome of the comparison can be the determination of one or more sectors of acceptable sites for a set of specific conditions considered, such as the reactor characteristics. The present approach was aimed to deal with the problems stemming from the demographic idiomorphy of Greece, where one third of the country's population is concentrated in Athens, with the rest of the country exhibiting small population densities. One of the applications of the methodology developed concerned the identification of acceptable sites near Athens. For these sites the risk from the reference severe accident of a standard reactor to the over three millions inhabitants of Athens is less tan the risk corresponding to the same population that is due to energy production

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.))

Hip arthroscopy protocol: expert opinions on post-operative weight bearing and return to sports guidelines.

Science.gov (United States)

Rath, Ehud; Sharfman, Zachary T; Paret, Matan; Amar, Eyal; Drexler, Michael; Bonin, Nicolas

2017-01-01

The objectives of this study are to survey the weight-bearing limitation practices and delay for returning to running and impact sports of high volume hip arthroscopy orthopedic surgeons. The study was designed in the form of expert survey questionnaire. Evidence-based data are scares regarding hip arthroscopy post-operative weight-bearing protocols. An international cross-sectional anonymous Internet survey of 26 high-volume hip arthroscopy specialized surgeons was conducted to report their weight-bearing limitations and rehabilitation protocols after various arthroscopic hip procedures. The International Society of Hip Arthroscopy invited this study. The results were examined in the context of supporting literature to inform the studies suggestions. Four surgeons always allow immediate weight bearing and five never offer immediate weight bearing. Seventeen surgeons provide weight bearing depending on the procedures performed: 17 surgeons allowed immediate weight bearing after labral resection, 10 after labral repair and 8 after labral reconstruction. Sixteen surgeons allow immediate weight bearing after psoas tenotomy. Twenty-one respondents restrict weight bearing after microfracture procedures for 3-8 weeks post-operatively. Return to running and impact sports were shorter for labral procedures and bony procedures and longer for cartilaginous and capsular procedures. Marked variability exists in the post-operative weight-bearing practices of hip arthroscopy surgeons. This study suggests that most surgeons allow immediate weight bearing as tolerated after labral resection, acetabular osteoplasty, chondroplasty and psoas tenotomy. For cartilage defect procedures, 6 weeks or more non-weight bearing is suggested depending on the area of the defect and lateral central edge angle. Delayed return to sports activities is suggested after microfracture procedures. The level of evidence was Level V expert opinions.

Chernobyl reactor accident

International Nuclear Information System (INIS)

1986-05-01

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

Gender-Sensitive Post-Disaster Assessments

OpenAIRE

World Bank

2012-01-01

This note on gender-sensitive post-disaster assessments is the eighth in a series of guidance notes on gender issues in disaster risk management (DRM) in the East Asia and the Pacific region. Targeting World Bank staff, clients and development partners, this note gives an overview of the main reasons for assessing gender impacts as part of a post-disaster needs assessment, identifies the k...

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

Evaluation of food chain transfer data for use in accident consequence assessment

International Nuclear Information System (INIS)

Coughtrey, P.J.; Kirton, J.A.; Mitchell, N.G.

1991-01-01

Input data for the food chain transport component of radiological assessment models are summarised in the context of the sources of information available prior to the Chernobyl accident and those derived after the accident. Particular attention is devoted to interception and retention soil-to-plant, and plant-to-animal transfer, and to the applicability of environmental data to both equilibrium and time-dependent models. It is argued that much of the current uncertainty in parameter values for use in radiological assessment models reflects lack of understanding of processes involved in the various stages of transfer of radionuclides to man. The Chernobyl accident highlighted this lack of fundamental knowledge and illustrated a number of areas where further research and model development is justified. These areas are identified and suggestions given for appropriate research to support model development

Radiological aspects of nuclear accident scenarios. Volume 2 the Rade-Aid system post-Chernobyl action

International Nuclear Information System (INIS)

Sinnaeve, J.

1991-01-01

In the event of a nuclear accident, there is a need for a rapid assessment of the resulting levels of environmental contamination in order to facilitate decisions on possible countermeasures. Volume 2 describes the RADE-AID project to develop a computer system which can be used to support the formulation of decisions on countermeasures following an accidental release of radionuclides. The system is intended as an aid following an actual accident and a tool for assistance in planning and training

Comparative Assessment Of Natural Gas Accident Risks

International Nuclear Information System (INIS)

Burgherr, P.; Hirschberg, S.

2005-01-01

The study utilizes a hierarchical approach including (1) comparative analyses of different energy chains, (2) specific evaluations for the natural gas chain, and (3) a detailed overview of the German situation, based on an extensive data set provided by Deutsche Vereinigung des Gas- und Wasserfaches (DVGW). According to SVGW-expertise DVGW-data can be regarded as fully representative for Swiss conditions due to very similar technologies, management, regulations and safety culture, but has a substantially stronger statistical basis because the German gas grid is about 30 times larger compared to Switzerland. Specifically, the following tasks were carried out by PSI to accomplish the objectives of this project: (1) Consolidation of existing ENSAD data, (2) identification and evaluation of additional sources, (3) comparative assessment of accident risks, and (4) detailed evaluations of specific issues and technical aspects for severe and smaller accidents in the natural gas chain that are relevant under Swiss conditions. (author)

Human reliability data, human error and accident models--illustration through the Three Mile Island accident analysis

International Nuclear Information System (INIS)

Le Bot, Pierre

2004-01-01

Our first objective is to provide a panorama of Human Reliability data used in EDF's Safety Probabilistic Studies, and then, since these concepts are at the heart of Human Reliability and its methods, to go over the notion of human error and the understanding of accidents. We are not sure today that it is actually possible to provide in this field a foolproof and productive theoretical framework. Consequently, the aim of this article is to suggest potential paths of action and to provide information on EDF's progress along those paths which enables us to produce the most potentially useful Human Reliability analyses while taking into account current knowledge in Human Sciences. The second part of this article illustrates our point of view as EDF researchers through the analysis of the most famous civil nuclear accident, the Three Mile Island unit accident in 1979. Analysis of this accident allowed us to validate our positions regarding the need to move, in the case of an accident, from the concept of human error to that of systemic failure in the operation of systems such as a nuclear power plant. These concepts rely heavily on the notion of distributed cognition and we will explain how we applied it. These concepts were implemented in the MERMOS Human Reliability Probabilistic Assessment methods used in the latest EDF Probabilistic Human Reliability Assessment. Besides the fact that it is not very productive to focus exclusively on individual psychological error, the design of the MERMOS method and its implementation have confirmed two things: the significance of qualitative data collection for Human Reliability, and the central role held by Human Reliability experts in building knowledge about emergency operation, which in effect consists of Human Reliability data collection. The latest conclusion derived from the implementation of MERMOS is that, considering the difficulty in building 'generic' Human Reliability data in the field we are involved in, the best

Managing Post-Disaster Needs Assessments (PDNA)

OpenAIRE

Jones, Brett

2010-01-01

This knowledge note provides an overview of the post-disaster assessment process, extracting lessons learned in the East Asia Pacific Region (EAP) and presenting best practices from recent assessments. The note explains the Post-Disaster Needs Assessment (PDNA) methodology, and outlines: (i) the assessment triggers, (ii) key steps in assessment planning, and (iii) dos and don'ts in assessm...

Radiological consequences of the Chernobyl reactor accident

International Nuclear Information System (INIS)

Hill, P.; Hille, R.

2003-01-01

The reactor accident at unit 4 of the Chernobyl nuclear power plant in Ukraine has deeply affected the living conditions of millions of people. Especially the health consequences have been of public concern up to the present and also been the subject of sometimes absurd claims. The current knowledge on the radiological consequences of the accident is reviewed. Though an increased hazard for some risk groups with high radiation exposure, e.g., liquidators, still cannot be totally excluded for the future, the majority of the population shows no statistically significant indication of radiation-induced illnesses. The contribution of the Research Center Juelich to the assessment of the post-accidental situation and psychological relief of the population is reported. The population groups still requiring special attention include, in particular, children growing up in highly contaminated regions and the liquidators of the years 1986 and 1987 deployed immediately after the accident. (author)

Development of a reliable, valid measure to assess parents' and teachers' understanding of postural care for children with physical disabilities: the (UKC PostCarD) questionnaire.

Science.gov (United States)

Hotham, S; Hutton, E; Hamilton-West, K E

2015-11-01

Previous research has highlighted lack of knowledge, understanding and confidence among parents and teachers responsible for the postural care of children with physical disability. Interventions designed to improve these qualities require a reliable and validated tool to assess pre- and post-intervention levels. Currently, however, no validated measure of postural care confidence (i.e. self-efficacy) exists. Hence, the aim of this research was to develop a reliable and valid questionnaire to assess parents' and teachers' confidence, alongside knowledge and understanding of postural care - the Understanding Knowledge and Confidence in providing POSTural CARe for children with Disabilities (UKC PostCarD) questionnaire. Items were developed by a multidisciplinary team and designed to map onto the content of 'An A-to-Z of Postural Care'. Parents, teachers and therapists assessed items for face validity. Scale reliability was then assessed using Cronbach's alpha and known-group validity was assessed by comparing scores of an 'expert' group (physiotherapists and occupational therapists) with those of a 'non-expert' group (with no formal training in postural care). The total scale and all three subscales (understanding and knowledge, confidence and concerns) demonstrated adequate reliability (α > 0.83) and subscale correlations formed a logical pattern (understanding and knowledge correlated positively with confidence and negatively with concerns). Experts' (n = 111) scores were higher than non-experts' (n = 79) for the total scale and all subscales (P children with disabilities. © 2015 John Wiley & Sons Ltd.

Consequences of radioactive releases into the sea resulting from the accident at the Fukushima Dai-ichi nuclear power plant - Evolution of expert investigation according to the data available

International Nuclear Information System (INIS)

Laguionie, P.; Bailly Du Bois, P.; Boust, D.; Fievet, B.; Connan, O.; Garreau, P.; Charmasson, S.; Arnaud, M.; Duffa, C.; Champion, D.

2012-01-01

The accident at the Fukushima Dai-ichi Nuclear Power Plant (FDNPP) in March 2011 led to an unprecedented direct input of artificial radioactivity into the marine environment. The Institute for Radioprotection and Nuclear Safety was requested by the French authorities to investigate the radioecological impact of this input, in particular the potential contamination of products of marine origin used for human consumption. This article describes the close link between the responses provided and the availability of the data, as well as their nature and ability to meet the requirements of expert investigation. These responses were needed: (i) to evaluate the inputs of radionuclides into the marine environment, (ii) to understand their dispersion in seawater, and (iii) to estimate their transfer to the biota and sediments. Three phases can be distinguished which characterise these processes during the accident and post-accident periods. The first phase corresponds to an emergency phase during which no measurements were available on samples from the marine environment. It involved the formulation of hypotheses based solely on the expertise of the Institute for Radioprotection and Nuclear Safety. The second phase started when the Japanese authorities provided measurements of the concentrations of radionuclides in seawater. Although these data were not yet adapted to addressing the problems of radioecology, the scenarios could then be refined and the estimates developed in more detail. During the third phase, the accumulation of data over the course of time made it possible to study the phenomena in an appropriate way. The chronology of the events shows that it is essential to have (i) significant measurements of concentration from samples collected in the various matrices of the marine environment, regularly updated and sufficiently well-documented, (ii) samples of seawater collected at the earliest opportunity as close as possible to the damaged site to characterise the

Synthesis of the IRSN report related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor. Referral of the Permanent Group of Experts for nuclear reactors (GPR), examination of probabilistic level-2 safety studies (EPS 2) and severe accidents (AG) of the Flamanville reactor nr 3. Opinion related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor (FA3). Electronuclear reactors - EDF - Flamanville 3 EPR reactor. Severe accidents and probabilistic level 2 studies

International Nuclear Information System (INIS)

2015-01-01

This document gathers several documents. The first one recalls the main arrangements implemented on the FA3 EPR reactor regarding accidents with core fusion, reports the analysis made by the IRSN about the sizing of these arrangements to reach a controlled status of the installation after a severe accident, regarding the probabilistic level-2 safety assessment, regarding the radiological impact of a severe accident on the population and on the environment, regarding those aimed at facing a total and long duration loss of electric power sources and cold sources, and about the situation of the reactor with respect to WENRA positions on severe accidents for new reactors. The second document is a letter sent by the ASN to the Permanent Group of Experts for nuclear reactors (GPR) to address probabilistic level-2 safety studies (EPS2) and severe accidents for the Flamanville 3 reactor. The third one reports the opinion of the GPR on these both issues and proposes a set of recommendations. The next document is a letter sent by the ASN to the Flamanville 3 project manager at EDF which recalls the related objectives, the ASN opinion on the implemented arrangements for severe accidents (de-pressurization of the primary circuit, management of hydrogen-related risks, corium recovery and cooling outside the vessel, limitation of vapour explosion risks outside the vessel, heat evacuation system, containment enclosure, management of the risk of a return to criticality), to face a total and long duration loss of electricity sources and cold sources, and other aspects addressed in the IRSN analysis. Requests and remarks formulated by the ASN are provided in an appendix to this last document

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)

Development Of Dynamic Probabilistic Safety Assessment: The Accident Dynamic Simulator (ADS) Tool

International Nuclear Information System (INIS)

Chang, Y.H.; Mosleh, A.; Dang, V.N.

2003-01-01

The development of a dynamic methodology for Probabilistic Safety Assessment (PSA) addresses the complex interactions between the behaviour of technical systems and personnel response in the evolution of accident scenarios. This paper introduces the discrete dynamic event tree, a framework for dynamic PSA, and its implementation in the Accident Dynamic Simulator (ADS) tool. Dynamic event tree tools generate and quantify accident scenarios through coupled simulation models of the plant physical processes, its automatic systems, the equipment reliability, and the human response. The current research on the framework, the ADS tool, and on Human Reliability Analysis issues within dynamic PSA, is discussed. (author)

Development Of Dynamic Probabilistic Safety Assessment: The Accident Dynamic Simulator (ADS) Tool

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.H.; Mosleh, A.; Dang, V.N

2003-03-01

The development of a dynamic methodology for Probabilistic Safety Assessment (PSA) addresses the complex interactions between the behaviour of technical systems and personnel response in the evolution of accident scenarios. This paper introduces the discrete dynamic event tree, a framework for dynamic PSA, and its implementation in the Accident Dynamic Simulator (ADS) tool. Dynamic event tree tools generate and quantify accident scenarios through coupled simulation models of the plant physical processes, its automatic systems, the equipment reliability, and the human response. The current research on the framework, the ADS tool, and on Human Reliability Analysis issues within dynamic PSA, is discussed. (author)

The risk of a major nuclear accident: calculation and perception of probabilities

International Nuclear Information System (INIS)

Leveque, Francois

2013-07-01

observations of the real world, the objective probability of an accident and the subjective assessment of risks? Did the experts err on the side of optimism? How should risk and its perception be measured? (author)

Proceedings of the Second OECD/NEA Organisation Meeting on Increased Accident Tolerance of Fuels for LWRs

International Nuclear Information System (INIS)

Massara, Simone; ); Bragg-Sitton, Shannon; Braase, Lori; Merrill, Brad; Teague, Melissa; Stanek, Chris R.; Montgomery, Robert H.; Ott, Larry J.; Robb, Kevin; Snead, Lance; Farmer, Mitch; Billone, Michael C.; Todosow, Michael; Brown, Nicholas; Brachet, J.C.; Le Flem, M.; Sauder, C.; Idarraga-Trujillo, I.; Michaux, A.; Lorrette, C.; Le Saux, M.; Blanpain, P.; Park, Jeong-Yong; Yang, Jae-Ho; Kim, Weon-Ju; Koo, Yang-Hyun; Liu, T.; Hallstadius, Lars; Lahoda, Ed; Waeckel, N.; Bonnet, J.M.; Vitanza, Carlo; Ohta, Hirokazu; Ogata, Takanari; Nakamura, Kinya; Dyck, Gary; Inozemtsev, Victor; )

2013-01-01

Under the guidance of the OECD-NEA Nuclear Science Committee, the expert group acts as a forum for scientific and technical information exchange on advanced light water reactor (LWR) fuels with enhanced accident tolerance. The expert group focusses on the fundamental properties and behaviour under normal operations and accident conditions for advanced core materials and components (fuels, cladding, control rods, etc.). The materials considered are applicable to Gen II and Gen III Light Water Reactors, as well as Gen III+ reactors under construction. The objective of the expert group is to define and coordinate a programme of work to help advance the scientific knowledge needed to provide the technical underpinning for the development of advanced LWR fuels with enhanced accident tolerance compared to currently used zircaloy/UO 2 fuel systems, as well as other non-fuel core components with important roles in LWR performance under accident conditions. This document brings together the available presentations (slides) given at the 2. Meeting on Increased Accident Tolerance of Fuels for LWRs. Content: 1 - Overview of the exchanges after the December-2012 Workshop through the discussion forum established at the OECD-NEA (S. Massara, NEA); 2 - Metrics Development for Enhanced Accident Tolerant LWR Fuels (S. Bragg-Sitton, INL); 3 - Candidate ATF Clad Technologies and Key Feasibility Issues (L. Snead, ORNL); 4 - CEA studies on nuclear fuel claddings for LWRs enhanced accident tolerant fuel. Some recent results, pending issues and prospects (J.C. Brachet, CEA); 5 - Current status on the accident tolerant fuel development in the Republic of Korea (J.Y. Park, J.H. Chang, KAERI); 6 - The current status of fuel R and D in the P.R. of China (T. Liu, CGN). Session 2: Key elements for a work programme on ATF: 7 - Beneficial characteristics of ATF (metrics) (L. Hallstadius, Westinghouse); 8 - Reactor types of interest (applicability) (L. Ott, ORNL); 9 - Impact on normal operations

To improve nuclear plant safety by learning from accident's experience

International Nuclear Information System (INIS)

Matsumoto, Hidezo; Kida, Masanori; Kato, Hiroyuki; Hara, Shin-ichi

1994-01-01

The ultimate goal of this study is to produce an expert system that enables the experience (records and information) gained from accidents to be put to use towards improving nuclear plant safety. A number of examples have been investigated, both domestic and overseas, in which experience gained from accidents was utilized by utilities in managing and operating their nuclear power stations to improve safety. The result of investigation has been used to create a general 'basic flow' to make the best use of experience. The ultimate goal is achieved by carrying out this 'basic flow' with artificial intelligence (AI). To do this, it is necessary (1) to apply language analysis to process the source information (primary data base; domestic and overseas accident's reports) into the secondary data base, and (2) to establish an expert system for selecting (screening) significant events from the secondary data base. In the processing described in item (1), a multi-lingual thesaurus for nuclear-related terms become necessary because the source information (primary data bases) itself is multi-lingual. In the work described in item (2), the utilization of probabilistic safety assessment (PSA), for example, is a candidate method for judging the significance of events. Achieving the goal thus requires developing various new techniques. As the first step of the above long-term study project, this report proposes the 'basic flow' and presents the concept of how the nuclear-related AI can be used to carry out this 'basic flow'. (author)

Method for Assessing Risk of Road Accidents in Transportation of School Children

Science.gov (United States)

Pogotovkina, N. S.; Volodkin, P. P.; Demakhina, E. S.

2017-11-01

The rationale behind the problem being investigated is explained by the remaining high level of the accident rates with the participation of vehicles carrying groups of children, including school buses, in the Russian Federation over the period of several years. The article is aimed at the identification of new approaches to improve the safety of transportation of schoolchildren in accordance with the Concept of children transportation by buses and the plan for its implementation. The leading approach to solve the problem under consideration is the prediction of accidents in the schoolchildren transportation. The article presents the results of the accident rate analysis with the participation of school buses in the Russian Federation for five years. Besides, a system to monitor the transportation of schoolchildren is proposed; the system will allow analyzing and forecasting traffic accidents which involve buses carrying groups of children, including school buses. In addition, the article presents a methodology for assessing the risk of road accidents during the transportation of schoolchildren.

Assessment of radiation risks as a result of the Chernobyl accident

International Nuclear Information System (INIS)

Ivanov, V.K.

1998-01-01

Full text of publication follows: the Government of the former USSR had made decision on establishing common registry of exposed persons in several months after the Chernobyl accident. The registry had served in Medical Radiological Research Centre of Russian Academy of Medical Sciences, Obninsk City till 1992 (the time of dissolution of the USSR). Individual medical and dosimetric information on 659292 persons, including 284907 emergency accident workers (liquidators) had been collected for the period between 1986 and 1991. As of 01.01.1998, National Chernobyl Registry of the Russian Federation has kept individual data on 508236 persons including 167726 liquidators. As it is known, long-term epidemiological study of Hiroshima and Nagasaki A-bomb survivors resulted in statistically significant assessments of radiation risks for induction of cancer at the dose level above 0.5 Gy. Radiation doses after the Chernobyl accident do not exceed 0.3-0.5 Gy. That is why assessment of radiation risks at low radiation doses is a problem of great importance. As a result of the epidemiological studies performed on the basis of the Russian Chernobyl registry we pioneered the assessment of statistically significant radiation risks for induction of cancer at low radiation dose. (author)

Learning Safety Assessment from Accidents in a University Environment

OpenAIRE

Jensen, Niels; Jørgensen, Sten Bay

2013-01-01

This contribution describes how a chemical engineering department started learning from accidents during experimental work and ended up implementing an industrially inspired system for risk assessment of new and existing experimental setups as well as a system for assessing potential risk from the chemicals used in the experimental work. These experiences have led to recent developments which focus increasingly on the a theoretical basis for modeling and reasoning on safety as well as operati...

Evaluation of severe accident risk in the Pickering a risk assessment

International Nuclear Information System (INIS)

Dinnie, K.S.; Raina, V.M.

1997-01-01

The nature of the design of commercial power plants is such that significant impacts on public health can only occur if a number of barriers fail. Rigorous design and licensing requirements ensure that the more likely accidents do not fail all these barriers and their contribution to risk is likely to be small. The task of estimating accident risk must, therefore, focus more towards those less likely but potentially more serious combinations of failures that are characterized by the following: a) a large release of fission products into the containment atmosphere, b) a breach in the containment envelope, and c) the existence of a driving force to expel the containment atmosphere to the outside environment. The likelihood of such conditions existing simultaneously during the course of an accident is expected to be small, such that experience and data regarding the behaviour of plant systems under such conditions is sparse or non-existent. The challenge of Probabilistic Safety Assessments (PSAs) is to examine the potential for severe accidents using approaches that are sufficiently detailed and realistic to provide valid information regarding plant risk and susceptibilities, while simple enough to keep the analysis manageable. This paper outlines the key features of the Pickering A Risk Assessment (PARA) (1) and the manner in which it addresses these issues, and provides some insights into the results and conclusions drawn from the study. (author)

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

Criticality accident:

International Nuclear Information System (INIS)

Canavese, Susana I.

2000-01-01

A criticality accident occurred at 10:35 on September 30, 1999. It occurred in a precipitation tank in a Conversion Test Building at the JCO Tokai Works site in Tokaimura (Tokai Village) in the Ibaraki Prefecture of Japan. STA provisionally rated this accident a 4 on the seven-level, logarithmic International Nuclear Event Scale (INES). The September 30, 1999 criticality accident at the JCO Tokai Works Site in Tokaimura, Japan in described in preliminary, technical detail. Information is based on preliminary presentations to technical groups by Japanese scientists and spokespersons, translations by technical and non-technical persons of technical web postings by various nuclear authorities, and English-language non-technical reports from various news media and nuclear-interest groups. (author)

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)

Site assessment after a pipeline accident at Moutnice

International Nuclear Information System (INIS)

Kult, L.; Sara, V.; Vavra, J.

1993-12-01

The current condition of land contaminated with crude oil due to the accident which occurred at the Druzhba (Friendship) pipeline in 1988, and of the vegetation growing on it is assessed. The contours of maximum pollution shortly after the accident can be easily found based on the observed nonpolar substance contents in the soil. The pH values are about 7.4. Analyses revealed no elevated heavy metal contents as compared with normal unpolluted soil. The above-ground parts of barley exhibit retarded growth corresponding to the degree of soil pollution. With one exception, the vanadium and nickel contents of plants grown in the polluted soil are lower than as encountered in clean soil. In the most affected areas the level of pollution is too high to enable the land to be used for farming. (J.B.). 6 tabs., 5 figs

Regulation Plans on Severe Accidents developed by KINS Severe Accident Regulation Preparation TFT

International Nuclear Information System (INIS)

Kim, Kyun Tae; Chung, Ku Young; Na, Han Bee

2016-01-01

Some nuclear power plants in Fukushima Daiichi site had lost their emergency reactor cooling function for long-time so the fuels inside the reactors were molten, and the integrity of containment was damaged. Therefore, large amount of radioactive material was released to environment. Because the social and economic effects of severe accidents are enormous, Korean Government already issued 'Severe Accident Policy' in 2001 which requires nuclear power plant operators to set up 'Quantitative Safety Goal', to do 'Probabilistic Safety Analysis', to install 'Severe Accident Countermeasures' and to make 'Severe Accident Management Plan'. After the Fukushima disaster, a Special Safety Inspection was performed for all operating nuclear power plants of Korea. The inspection team from industry, academia, and research institutes assessed Korean NPPs capabilities to cope with or respond to severe accidents and emergency situation caused by natural disasters such as a large earthquake or tsunami. As a result of the special inspection, about 50 action items were identified to increase the capability to cope with natural disaster and severe accidents. Nuclear Safety Act has been amended to require NPP operators to submit Accident Management Plant as part of operating license application. The KINS Severe Accident Regulation Preparation TFT had first investigated oversea severe accident regulation trend before and after the Fukushima accident. Then, the TFT has developed regulation draft for severe accidents such as Severe accident Management Plans, the required design features for new NPPs to prevent severe accident against multiple failures and beyond-design external events, countermeasures to mitigate severe accident and to keep the integrity of containment, and assessment methodology on safety assessment plan and probabilistic safety assessment

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

The role of nuclear reactor containment in severe accidents

International Nuclear Information System (INIS)

1989-04-01

The containment is a structural envelope which completely surrounds the nuclear reactor system and is designed to confine the radioactive releases in case of an accident. This report summarises the work of an NEA Senior Group of Experts who have studied the potential role of containment in accidents exceeding design specifications (so-called severe accidents). Some possibilities for enhancing the ability of plants to reduce the risk of significant off-site consequences by appropriate management of the acident have been examined

A methodology for assessing the effect of countermeasures against a nuclear accident using fuzzy set theory

International Nuclear Information System (INIS)

Han, M.H.; Hwang, W.T.; Kim, E.H.; Suh, K.S.; Choi, Y.G.

2000-01-01

A methodology for assessing the effectiveness of countermeasures against a nuclear accident has been designed by means of the concept of fuzzy set theory. In most of the existing countermeasure models in actions under radiological emergencies, the large variety of possible features is simplified by a number of rough assumptions. During this simplification procedure, a lot of information is lost which results in much uncertainty concerning the output of the countermeasure model. Furthermore, different assumptions should be used for different sites to consider the site specific conditions. In this study, the diversity of each variable related to protective action has been modelled by the linguistic variable. The effectiveness of sheltering and evacuation has been estimated using the proposed method. The potential advantage of the proposed method is in reducing the loss of information by incorporating the opinions of experts and by introducing the linguistic variables which represent the site specific conditions. (author)

Radiation risk models for all solid cancers other than those types of cancer requiring individual assessments after a nuclear accident

International Nuclear Information System (INIS)

Walsh, Linda; Zhang, Wei

2016-01-01

In the assessment of health risks after nuclear accidents, some health consequences require special attention. For example, in their 2013 report on health risk assessment after the Fukushima nuclear accident, the World Health Organisation (WHO) panel of experts considered risks of breast cancer, thyroid cancer and leukaemia. For these specific cancer types, use was made of already published excess relative risk (ERR) and excess absolute risk (EAR) models for radiation-related cancer incidence fitted to the epidemiological data from the Japanese A-bomb Life Span Study (LSS). However, it was also considered important to assess all other types of solid cancer together and the WHO, in their above-mentioned report, stated ''No model to calculate the risk for all other solid cancer excluding breast and thyroid cancer risks is available from the LSS data''. Applying the LSS models for all solid cancers along with the models for the specific sites means that some cancers have an overlap in the risk evaluations. Thus, calculating the total solid cancer risk plus the breast cancer risk plus the thyroid cancer risk can overestimate the total risk by several per cent. Therefore, the purpose of this paper was to publish the required models for all other solid cancers, i.e. all solid cancers other than those types of cancer requiring special attention after a nuclear accident. The new models presented here have been fitted to the same LSS data set from which the risks provided by the WHO were derived. Although it is known already that the EAR and ERR effect modifications by sex are statistically significant for the outcome ''all solid cancer'', it is shown here that sex modification is not statistically significant for the outcome ''all solid cancer other than thyroid and breast cancer''. It is also shown here that the sex-averaged solid cancer risks with and without the sex modification are very similar once breast and thyroid cancers are factored out. Some other notable model

Formation of decontamination cost calculation model for severe accident consequence assessment

International Nuclear Information System (INIS)

Silva, Kampanart; Promping, Jiraporn; Okamoto, Koji; Ishiwatari, Yuki

2014-01-01

In previous studies, the authors developed an index “cost per severe accident” to perform a severe accident consequence assessment that can cover various kinds of accident consequences, namely health effects, economic, social and environmental impacts. Though decontamination cost was identified as a major component, it was taken into account using simple and conservative assumptions, which make it difficult to have further discussions. The decontamination cost calculation model was therefore reconsidered. 99 parameters were selected to take into account all decontamination-related issues, and the decontamination cost calculation model was formed. The distributions of all parameters were determined. A sensitivity analysis using the Morris method was performed in order to identify important parameters that have large influence on the cost per severe accident and large extent of interactions with other parameters. We identified 25 important parameters, and fixed most negligible parameters to the median of their distributions to form a simplified decontamination cost calculation model. Calculations of cost per severe accident with the full model (all parameters distributed), and with the simplified model were performed and compared. The differences of the cost per severe accident and its components were not significant, which ensure the validity of the simplified model. The simplified model is used to perform a full scope calculation of the cost per severe accident and compared with the previous study. The decontamination cost increased its importance significantly. (author)

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)

Identification and Prioritization of “Black Spots” without Using Accident Information

Directory of Open Access Journals (Sweden)

Mahmoudreza Keymanesh

2017-01-01

Full Text Available It is attempted to identify and prioritize the accident prone points (black spots in “Iraanshahr-Sarbaaz-Chabahar” road located in Baluchistan, Iran, without no use of accident data but rather using Analytic Hierarchy Process (AHP, which is the enhanced procedure of road safety audit technique. First, by surveying the whole route, all factors that could influence accidents in this road were specified; then the route was divided into eight sections; this division was performed based on the uniformity and homogeny of each section in terms of geometric design and regional conditions. In each section, potentially hazardous locations were identified and some questionnaires were prepared, which were filled by 5 road traffic experts familiar with the route; then the collected data were analyzed by Analytical Hierarchy Process (AHP using Expert Choice Software and the black spots were identified and prioritized. Finally, these black spots were compared with the black spots that had been obtained by traffic police based on accident data.

10 years from the Chernobyl nuclear reactor accident: consequences and lesson learned

International Nuclear Information System (INIS)

1996-01-01

Published jointly by the Czech State Office for Nuclear Safety and the Czech National Radiation Protection Institute, the publication gives a succinct account of the cause of the Chernobyl accident and its impact on the former Soviet Union, and concentrates on the effects of the accident on the Czech Republic. The topics dealt with in this respect include, among others: radionuclide contents of foods with particular emphasis on milk products for babies, assessment of surface contamination of the Czech Republic due to the accident, internal contamination of the population as determined by whole-body measurements, assessment of the effective dose equivalents from external irradiation and effective dose equivalent commitments from internal irradiation, cesium radioisotopes in natural ecosystems, and the use of post-Chernobyl monitoring to test radionuclide migration models within the IAEA VAMP programme. (P.A.). 12 tabs., 30 figs., 64 refs

A Model for Traffic Accidents Prediction Based on Driver Personality Traits Assessment

Directory of Open Access Journals (Sweden)

Marjana Čubranić-Dobrodolac

2017-12-01

Full Text Available The model proposed in this paper uses four psychological instruments for assessing driver behaviour and personality traits aiming to find a relationship between the considered constructs and the occurrence of traffic accidents. A Barratt Impulsiveness Scale (BIS-11 was used for the assessment of impulsivity, Aggressive Driving Behaviour Questionnaire (ADBQ for assessing the aggressiveness while driving, Manchester Driver Attitude Questionnaire (DAQ and the Questionnaire for self-assessment of driving ability. Besides these instruments, the participants filled out an extensive demographic survey. Within the statistical analysis, in addition to the descriptive indicators, correlation coefficients were calculated and four hierarchical regression analyses were performed to determine the predictive power of personality traits on the occurrence of traffic accidents. Further, to confirm the results and to obtain additional information about the relationship between the considered variables, the structural equation modelling and binary logistic regression have been implemented. A sample of this research covered 305 drivers, of which there were 100 bus drivers and 102 truck drivers, as well as 103 drivers of privately owned vehicles. The results indicate that BIS-11 and ADBQ questionnaires show the best predictive power which means that impulsivity and aggressiveness as personality traits have the greatest influence on the occurrence of traffic accidents. This research could be useful in many fields, such as the design of selection procedures for professional drivers, development of programs for the prevention of traffic accidents and violations of law, rehabilitation of drivers who have been deprived of the driving license, etc.

Testing of an accident consequence assessment model using field data

International Nuclear Information System (INIS)

Homma, Toshimitsu; Matsubara, Takeshi; Tomita, Kenichi

2007-01-01

This paper presents the results obtained from the application of an accident consequence assessment model, OSCAAR to the Iput dose reconstruction scenario of BIOMASS and also to the Chernobyl 131 I fallout scenario of EMRAS, both organized by International Atomic Energy Agency. The Iput Scenario deals with 137 Cs contamination of the catchment basin and agricultural area in the Bryansk Region of Russia, which was heavily contaminated after the Chernobyl accident. This exercise was used to test the chronic exposure pathway models in OSCAAR with actual measurements and to identify the most important sources of uncertainty with respect to each part of the assessment. The OSCAAR chronic exposure pathway models had some limitations but the refined model, COLINA almost successfully reconstructed the whole 10-year time course of 137 Cs activity concentrations in most requested types of agricultural products and natural foodstuffs. The Plavsk scenario provides a good opportunity to test not only the food chain transfer model of 131 I but also the method of assessing 131 I thyroid burden. OSCAAR showed in general good capabilities for assessing the important 131 I exposure pathways. (author)

Some lessons on radiological protection learnt from the accident at the Fukushima Dai-ichi nuclear power plant

International Nuclear Information System (INIS)

Kai, M

2012-01-01

The accident at the Fukushima Dai-ichi nuclear power plant released a large quantity of radioactive iodine and caesium into the environment. In terms of radiological protection, the evacuation and food restrictions that were adopted in a timely manner by the authorities effectively reduced the dose received by people living in the affected area. Since late March, the transition from an emergency to an existing exposure situation has been in progress. In selecting the reference exposure levels in some areas under an existing exposure situation, the authorities tried to follow the situation-based approach recommended by the ICRP. However, a mixture of emergency and post-emergency approaches confused the people living in the contaminated areas because the reactor conditions continued to be not completely stable. In deriving the criteria in an existing exposure situation, the regulatory authority selected 20 mSv y −1 . The mothers in the affected area believed that a dose of 20 mSv y −1 was unacceptably high for children since 1 mSv y −1 is the dose limit for the public under normal conditions. Internet information accelerated concern about the internal exposure to children and the related health effects. From some experiences after the accident the following lessons could be learned. The selection of reference doses in existing exposure situations after an accident must be openly communicated with the public using a risk-informed approach. The detriment-adjusted nominal risk coefficient was misused for calculating the hypothetical number of cancer deaths by some non-radiation experts. It would not be possible to resolve this problem unless the ICRP addressed an alternative risk assessment to convey the meaning and associated uncertainty of the risk to an exposed population. A situation-based approach in addition to a risk-informed approach needs to be disseminated properly in order to select the level of protection that would be the best possible under the

Regulation Plans on Severe Accidents developed by KINS Severe Accident Regulation Preparation TFT

Energy Technology Data Exchange (ETDEWEB)

Kim, Kyun Tae; Chung, Ku Young; Na, Han Bee [KINS, Daejeon (Korea, Republic of)

2016-05-15

Some nuclear power plants in Fukushima Daiichi site had lost their emergency reactor cooling function for long-time so the fuels inside the reactors were molten, and the integrity of containment was damaged. Therefore, large amount of radioactive material was released to environment. Because the social and economic effects of severe accidents are enormous, Korean Government already issued 'Severe Accident Policy' in 2001 which requires nuclear power plant operators to set up 'Quantitative Safety Goal', to do 'Probabilistic Safety Analysis', to install 'Severe Accident Countermeasures' and to make 'Severe Accident Management Plan'. After the Fukushima disaster, a Special Safety Inspection was performed for all operating nuclear power plants of Korea. The inspection team from industry, academia, and research institutes assessed Korean NPPs capabilities to cope with or respond to severe accidents and emergency situation caused by natural disasters such as a large earthquake or tsunami. As a result of the special inspection, about 50 action items were identified to increase the capability to cope with natural disaster and severe accidents. Nuclear Safety Act has been amended to require NPP operators to submit Accident Management Plant as part of operating license application. The KINS Severe Accident Regulation Preparation TFT had first investigated oversea severe accident regulation trend before and after the Fukushima accident. Then, the TFT has developed regulation draft for severe accidents such as Severe accident Management Plans, the required design features for new NPPs to prevent severe accident against multiple failures and beyond-design external events, countermeasures to mitigate severe accident and to keep the integrity of containment, and assessment methodology on safety assessment plan and probabilistic safety assessment.

Utilization of dose assessment models to facilitate off-site recovery operations for accidents at nuclear facilities

International Nuclear Information System (INIS)

Dickerson, M.H.; Foster, K.T.

1989-09-01

One of the most important uses of dose assessment models in response to accidents at nuclear facilities is to help provide guidance to emergency response managers for identifying, and mitigating, the consequences of an accident once the accident has been terminated. By combining results from assessment models with radiological measurements, a qualitative methodology can be developed to aid emergency response managers in determining the total dose received by the population and to minimize future doses through the use of mitigation procedures. To illustrate the methodology, this discussion focuses on the use of models to estimate the dose delivered to the public both during and after a nuclear accident. 4 refs., 10 figs., 1 tab

Remarks on the assessment, representation, aggregation and utilization of expert opinion

International Nuclear Information System (INIS)

Fine, T.L.

1980-04-01

This report considers the relevance of recent ideas in the foundations of probability to the rational use of expert opinion in the design of a nuclear waste repository, and the assessment of its performance. The main probability concepts introduce are those of modal (probably A), comparative (A is at least as probable as B) and interval-valued (the lower probability of A is P(A) and the upper probability of A is P(anti A)) probabilities. We then outline an approach first using comparative probability to model the resuls of binary elicitation of an expert's opinions concerning repository uncertainties and then employing interval-valued probability to represent comparative probability in a computationally convenient form. We further consider the issue of aggregating or amalgamating the responses of several experts, and we emphasize the need to preserve some measure of the disagreements among the experts. The resulting aggregated interval-valued representation of the responses concerning the uncertainties surrounding the performance of a nuclear waste repository design can then be used to numerically assess this performance in a manner parallel to that of utility theory. Utility theory is the basis for statistical decision theory. Our recommendations can only be tentative, and research is recommended to gain some working experience with the results of the proposed decision-making process in the repostory design context

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)

Constructing the Indicators of Assessing Human Vulnerability to Industrial Chemical Accidents: A Consensus-based Fuzzy Delphi and Fuzzy AHP Approach.

Science.gov (United States)

Fatemi, Farin; Ardalan, Ali; Aguirre, Benigno; Mansouri, Nabiollah; Mohammadfam, Iraj

2017-04-10

Industrial chemical accidents have been increased in developing countries. Assessing the human vulnerability in the residents of industrial areas is necessary for reducing the injuries and causalities of chemical hazards. The aim of this study was to explore the key indicators for the assessment of human vulnerability in the residents living near chemical installations. The indicators were established in the present study based on the Fuzzy Delphi method (FDM) and Fuzzy Analytic Hierarchy Process (FAHP). The reliability of FDM and FAHP was calculated. The indicators of human vulnerability were explored in two sets of social and physical domains. Thirty-five relevant experts participated in this study during March-July 2015. According to experts, the top three indicators of human vulnerability according to the FDM and FAHP were vulnerable groups, population density, and awareness. Detailed sub-vulnerable groups and awareness were developed based on age, chronic or severe diseases, disability, first responders, and residents, respectively. Each indicator and sub-indicator was weighted and ranked and had an acceptable consistency ratio. The importance of social vulnerability indicators are about 7 times more than physical vulnerability indicators. Among the extracted indicators, vulnerable groups had the highest weight and the greatest impact on human vulnerability. however, further research is needed to investigate the applicability of established indicators and generalizability of the results to other studies. Fuzzy Delphi; Fuzzy AHP; Human vulnerability; Chemical hazards.

The risk of accident in nuclear power plants - Quotes and questions. National debate on energy transition. Taking the risk of nuclear accident into account. Note to the 'scenarios' subgroup of the group of experts

International Nuclear Information System (INIS)

Laponche, Bernard

2013-01-01

After an overview of the production of electricity from nuclear energy and of its risks, the author discusses the issue of nuclear safety by distinguishing the different points of view, by describing the different levels between a severe and a major accident, and by recalling the statements made by the ASN and the IRSN on this issue. He describes the various reasons and consequences of accidents: types of accidents which could result in a core fusion, the containment failure as the major accident. He discusses the questions and comments about major issues like: strength of reactor vessels, hydrogen explosion, water vapour explosion, corium crossing the concrete sill plate, corium recovery, MOX as an aggravating situation for safety. In the last part, the author discusses the global assessment of a risk of a nuclear accident: probabilities and occurrences, significant accidents on PWRs, premonitory analysis, demonstration of nuclear safety

Assessment of Technogenic Accident Risk of Industrial Building Structures

Science.gov (United States)

Baiburin, D. A.; Baiburin, A. Kh

2017-11-01

A methodology for assessing the risk of an industrial building accident was developed taking into account the damage caused by various localization of collapse. Before the beginning of the survey of a facility technical condition, groups including the same type of building structures are selected. Further, assessment is made for the reduction in their load-carrying capacity from the strength and stability conditions taking into account defects. The characteristics of the influence of defects and structural damage on a building safety is the degree of compliance with the standards expressed by the reliability level. Reliability levels assignment is carried out on the basis of calculations, operating experience and inspection of a particular type of structure according to the formalized rules. The risk of collapse according to a separate scenario is calculated for structures that are capable and incapable of causing a progressive ossification. The results of the technique application are based on the analysis of the accident risk at the welding shop “Vysota (Height) 239” of the Chelyabinsk Pipe Rolling Plant.

Cernavoda CANDU severe accident evaluation

International Nuclear Information System (INIS)

Negut, G.; Marin, A.

1997-01-01

The papers present the activities dedicated to Romania Cernavoda Nuclear Power Plant first CANDU Unit severe accident evaluation. This activity is part of more general PSA assessment activities. CANDU specific safety features are calandria moderator and calandria vault water capabilities to remove the residual heat in the case of severe accidents, when the conventional heat sinks are no more available. Severe accidents evaluation, that is a deterministic thermal hydraulic analysis, assesses the accidents progression and gives the milestones when important events take place. This kind of assessment is important to evaluate to recovery time for the reactor operators that can lead to the accident mitigation. The Cernavoda CANDU unit is modeled for the of all heat sinks accident and results compared with the AECL CANDU 600 assessment. (orig.)

A limited assessment of the ASEP human reliability analysis procedure using simulator examination results

International Nuclear Information System (INIS)