Review of the severe accident risk reduction program (SARRP) containment event trees

International Nuclear Information System (INIS)

1986-05-01

A part of the Severe Accident Risk Reduction Program, researchers at Sandia National Laboratories have constructed a group of containment event trees to be used in the analysis of key accident sequences for light water reactors (LWR) during postulated severe accidents. The ultimate goal of the program is to provide to the NRC staff a current assessment of the risk from severe reactor accidents for a group of five light water reactors. This review specifically focuses on the development and construction of the containment event trees and the results for containment failure probability, modes and timing. The report first gives the background on the program, the review criteria, and a summary of the observations, findings and recommendations. secondly, the individual reviews of each committee member on the event trees is presented. Finally, a review is provided on the computer model used to construct and evaluate the event trees

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

Analysis on the nitrogen drilling accident of Well Qionglai 1 (I: Major inducement events of the accident

Directory of Open Access Journals (Sweden)

Yingfeng Meng

2015-12-01

Full Text Available Nitrogen drilling in poor tight gas sandstone should be safe because of very low gas production. But a serious accident of fire blowout occurred during nitrogen drilling of Well Qionglai 1. This is the first nitrogen drilling accident in China, which was beyond people's knowledge about the safety of nitrogen drilling and brought negative effects on the development of gas drilling technology still in start-up phase and resulted in dramatic reduction in application of gas drilling. In order to form a correct understanding, the accident was systematically analyzed, the major events resulting in this accident were inferred. It is discovered for the first time that violent ejection of rock clasts and natural gas occurred due to the sudden burst of downhole rock when the fractured tight gas zone was penetrated during nitrogen drilling, which has been named as “rock burst and blowout by gas bomb”, short for “rock burst”. Then all the induced events related to the rock burst are as following: upthrust force on drilling string from rock burst, bridging-off formed and destructed repeatedly at bit and centralizer, and so on. However, the most direct important event of the accident turns out to be the blockage in the blooie pipe from rock burst clasts and the resulted high pressure at the wellhead. The high pressure at the wellhead causes the blooie pipe to crack and trigged blowout and deflagration of natural gas, which is the direct presentation of the accident.

Containment severe accident thermohydraulic phenomena

International Nuclear Information System (INIS)

Frid, W.

1991-08-01

This report describes and discusses the containment accident progression and the important severe accident containment thermohydraulic phenomena. The overall objective of the report is to provide a rather detailed presentation of the present status of phenomenological knowledge, including an account of relevant experimental investigations and to discuss, to some extent, the modelling approach used in the MAAP 3.0 computer code. The MAAP code has been used in Sweden as the main tool in the analysis of severe accidents. The dependence of the containment accident progression and containment phenomena on the initial conditions, which in turn are heavily dependent on the in-vessel accident progression and phenomena as well as associated uncertainties, is emphasized. The report is in three parts dealing with: * Swedish reactor containments, the severe accident mitigation programme in Sweden and containment accident progression in Swedish PWRs and BWRs as predicted by the MAAP 3.0 code. * Key non-energetic ex-vessel phenomena (melt fragmentation in water, melt quenching and coolability, core-concrete interaction and high temperature in containment). * Early containment threats due to energetic events (hydrogen combustion, high pressure melt ejection and direct containment heating, and ex-vessel steam explosions). The report concludes that our understanding of the containment severe accident progression and phenomena has improved very significantly over the parts ten years and, thereby, our ability to assess containment threats, to quantify uncertainties, and to interpret the results of experiments and computer code calculations have also increased. (au)

Prediction of accident sequence probabilities in a nuclear power plant due to earthquake events

International Nuclear Information System (INIS)

Hudson, J.M.; Collins, J.D.

1980-01-01

This paper presents a methodology to predict accident probabilities in nuclear power plants subject to earthquakes. The resulting computer program accesses response data to compute component failure probabilities using fragility functions. Using logical failure definitions for systems, and the calculated component failure probabilities, initiating event and safety system failure probabilities are synthesized. The incorporation of accident sequence expressions allows the calculation of terminal event probabilities. Accident sequences, with their occurrence probabilities, are finally coupled to a specific release category. A unique aspect of the methodology is an analytical procedure for calculating top event probabilities based on the correlated failure of primary events

The management of severe accidents in modern pressure tube reactors

International Nuclear Information System (INIS)

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

2007-01-01

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

Event tree analysis of accidents during transport of radioactive materials in Japan

International Nuclear Information System (INIS)

Watabe, N.; Shirai, K.; Noguchi, K.; Suzuki, H.; Kinehara, Y.

1993-01-01

The Event Tree Method is one of the Probabilistic Safety Assessment Method. It introduces the accident scenario and the results of countermeasures. Therefore, it is effective in determining latent accident scenarios in the transfer. In this report the Event Tree Method is used to study the tunnel fire and its effects are evaluated. And this is the first trail of our Probabilistic Safety Assessment. The Event Tree for determining the early conditions when a car engine catches fire in a tunnel is examined. There are fire extinguishers, tunnel equipments for fire-fighting, fire stations and the heat-resisting property of the container for protecting from the fire. The protection level against the over 800degC-30min. fire accident is 88.3 %. (J.P.N.)

Severe Accident Progression and Consequence Assessment Methodology Upgrades in ISAAC for Wolsong CANDU6

International Nuclear Information System (INIS)

Song, Y.M.; Kim, D.H.; Nijhawan, Sunil

2015-01-01

Amongst the applications of integrated severe accident analysis codes like ISAAC, the principal are to a) help develop an understanding of the severe accident progression and its consequences; b) support the design of mitigation measures by providing for them the state of the reactor following an accident; and c) to provide a training platform for accident management actions. After Fukushima accident there is an increased awareness of the need to implement effective and appropriate mitigation measures and empower the operators with training and understanding about severe accident progression and control opportunities. An updated code with reduced uncertainties can better serve these needs of the utility making decisions about mitigation measures and corrective actions. Optimal deployment of systems such as PARS and filtered containment venting require information on reactor transients for a number of critical parameters. Thus there is a greater consensus now for a demonstrated ability to perform accident progression and consequence assessment analyses with reduced uncertainties. Analyses must now provide source term transients that represent the best in available understanding and so meaningfully support mitigation measures. This requires removal of known simplifications and inclusion of all quantifiable and risk significant phenomena. Advances in understanding of CANDU6 severe accident progression reflected in the severe accident integrated code ROSHNI are being incorporated into ISAAC using CANDU specific component and system models developed and verified for Wolsong CANDU 6 reactors. A significant and comprehensive upgrade of core behavior models is being implemented in ISAAC to properly reflect the large variability amongst fuel channels in feeder geometry, fuel thermal powers and burnup. The paper summarizes the models that have been added and provides some results to illustrate code capabilities. ISAAC is being updated to meet the current requirements and

Severe Accident Progression and Consequence Assessment Methodology Upgrades in ISAAC for Wolsong CANDU6

Energy Technology Data Exchange (ETDEWEB)

Song, Y.M.; Kim, D.H. [KAERI, Daejeon (Korea, Republic of); Nijhawan, Sunil [Prolet Inc. 98 Burbank Drive, Toronto (Canada)

2015-05-15

Amongst the applications of integrated severe accident analysis codes like ISAAC, the principal are to a) help develop an understanding of the severe accident progression and its consequences; b) support the design of mitigation measures by providing for them the state of the reactor following an accident; and c) to provide a training platform for accident management actions. After Fukushima accident there is an increased awareness of the need to implement effective and appropriate mitigation measures and empower the operators with training and understanding about severe accident progression and control opportunities. An updated code with reduced uncertainties can better serve these needs of the utility making decisions about mitigation measures and corrective actions. Optimal deployment of systems such as PARS and filtered containment venting require information on reactor transients for a number of critical parameters. Thus there is a greater consensus now for a demonstrated ability to perform accident progression and consequence assessment analyses with reduced uncertainties. Analyses must now provide source term transients that represent the best in available understanding and so meaningfully support mitigation measures. This requires removal of known simplifications and inclusion of all quantifiable and risk significant phenomena. Advances in understanding of CANDU6 severe accident progression reflected in the severe accident integrated code ROSHNI are being incorporated into ISAAC using CANDU specific component and system models developed and verified for Wolsong CANDU 6 reactors. A significant and comprehensive upgrade of core behavior models is being implemented in ISAAC to properly reflect the large variability amongst fuel channels in feeder geometry, fuel thermal powers and burnup. The paper summarizes the models that have been added and provides some results to illustrate code capabilities. ISAAC is being updated to meet the current requirements and

Monitoring Severe Accidents Using AI Techniques

International Nuclear Information System (INIS)

No, Young Gyu; Kim, Ju Hyun; Na, Man Gyun; Ahn, Kwang Il

2011-01-01

It is very difficult for nuclear power plant operators to monitor and identify the major severe accident scenarios following an initiating event by staring at temporal trends of important parameters. The objective of this study is to develop and verify the monitoring for severe accidents using artificial intelligence (AI) techniques such as support vector classification (SVC), probabilistic neural network (PNN), group method of data handling (GMDH) and fuzzy neural network (FNN). The SVC and PNN are used for event classification among the severe accidents. Also, GMDH and FNN are used to monitor for severe accidents. The inputs to AI techniques are initial time-integrated values obtained by integrating measurement signals during a short time interval after reactor scram. In this study, 3 types of initiating events such as the hot-leg LOCA, the cold-leg LOCA and SGTR are considered and it is verified how well the proposed scenario identification algorithm using the GMDH and FNN models identifies the timings when the reactor core will be uncovered, when CET will exceed 1200 .deg. F and when the reactor vessel will fail. In cases that an initiating event develops into a severe accident, the proposed algorithm showed accurate classification of initiating events. Also, it well predicted timings for important occurrences during severe accident progression scenarios, which is very helpful for operators to perform severe accident management

Probabilistic Accident Progression Analysis with application to a LMFBR design

International Nuclear Information System (INIS)

Jamali, K.M.

1982-01-01

A method for probabilistic analysis of accident sequences in nuclear power plant systems referred to as ''Probabilistic Accident Progression Analysis'' (PAPA) is described. Distinctive features of PAPA include: (1) definition and analysis of initiator-dependent accident sequences on the component level; (2) a new fault-tree simplification technique; (3) a new technique for assessment of the effect of uncertainties in the failure probabilities in the probabilistic ranking of accident sequences; (4) techniques for quantification of dependent failures of similar components, including an iterative technique for high-population components. The methodology is applied to the Shutdown Heat Removal System (SHRS) of the Clinch River Breeder Reactor Plant during its short-term (0 -2 . Major contributors to this probability are the initiators loss of main feedwater system, loss of offsite power, and normal shutdown

Some implications of an event-based definition of exposure to the risk of road accident.

Science.gov (United States)

Elvik, Rune

2015-03-01

This paper proposes a new definition of exposure to the risk of road accident as any event, limited in space and time, representing a potential for an accident to occur by bringing road users close to each other in time or space of by requiring a road user to take action to avoid leaving the roadway. A typology of events representing a potential for an accident is proposed. Each event can be interpreted as a trial as defined in probability theory. Risk is the proportion of events that result in an accident. Defining exposure as events demanding the attention of road users implies that road users will learn from repeated exposure to these events, which in turn implies that there will normally be a negative relationship between exposure and risk. Four hypotheses regarding the relationship between exposure and risk are proposed. Preliminary tests support these hypotheses. Advantages and disadvantages of defining exposure as specific events are discussed. It is argued that developments in vehicle technology are likely to make events both observable and countable, thus ensuring that exposure is an operational concept. Copyright © 2014 Elsevier Ltd. All rights reserved.

Statistical evaluation of design-error related nuclear reactor accidents

International Nuclear Information System (INIS)

Ott, K.O.; Marchaterre, J.F.

1981-01-01

In this paper, general methodology for the statistical evaluation of design-error related accidents is proposed that can be applied to a variety of systems that evolves during the development of large-scale technologies. The evaluation aims at an estimate of the combined ''residual'' frequency of yet unknown types of accidents ''lurking'' in a certain technological system. A special categorization in incidents and accidents is introduced to define the events that should be jointly analyzed. The resulting formalism is applied to the development of U.S. nuclear power reactor technology, considering serious accidents (category 2 events) that involved, in the accident progression, a particular design inadequacy. 9 refs

LMFBR accident delineation study: approach and preliminary results

International Nuclear Information System (INIS)

Williams, D.C.; Sholtis, J.A.; Rios, M.; Worledge, D.H.; Conrad, P.W.; Varela, D.W.; Pickard, P.S.

1979-01-01

Event trees have been constructed for all phases of LMFBR accidents. The trees proved useful for identifying meaningful initiating accident categories and containment responses. In these areas, quantification appears feasible, given an adequate data base. Event trees were also used to represent in-core phenomenological questions governing accident progression and energetics, but here quantification appears impracticable because pervasive phenomenological uncertainties exist. Infrequent accident initiation is the dominant factor in assuring low risk. Nevertheless, containment promises an additional measure of risk reduction provided severe energetics are highly unlikely. The delineation served to systematize LMFBR safety issues and should aid in evaluating LMFBR R and D priorities

Accident and Off-Normal Response and Recovery from Multi-Canister Overpack (MCO) Processing Events

International Nuclear Information System (INIS)

ALDERMAN, C.A.

2000-01-01

In the process of removing spent nuclear fuel (SNF) from the K Basins through its subsequent packaging, drymg, transportation and storage steps, the SNF Project must be able to respond to all anticipated or foreseeable off-normal and accident events that may occur. Response procedures and recovery plans need to be in place, personnel training established and implemented to ensure the project will be capable of appropriate actions. To establish suitable project planning, these events must first be identified and analyzed for their expected impact to the project. This document assesses all off-normal and accident events for their potential cross-facility or Multi-Canister Overpack (MCO) process reversal impact. Table 1 provides the methodology for establishing the event planning level and these events are provided in Table 2 along with the general response and recovery planning. Accidents and off-normal events of the SNF Project have been evaluated and are identified in the appropriate facility Safety Analysis Report (SAR) or in the transportation Safety Analysis Report for Packaging (SARP). Hazards and accidents are summarized from these safety analyses and listed in separate tables for each facility and the transportation system in Appendix A, along with identified off-normal events. The tables identify the general response time required to ensure a stable state after the event, governing response documents, and the events with potential cross-facility or SNF process reversal impacts. The event closure is predicated on stable state response time, impact to operations and the mitigated annual occurrence frequency of the event as developed in the hazard analysis process

Below Regulatory Conern Owners Group: Radiologic impact of accidents and unexpected events from disposal of BRC waste

International Nuclear Information System (INIS)

Waite, D.A.; Dolan, M.M.; Rish, W.R.; Rossi, A.J.; McCourt, J.E.

1989-07-01

This report determines the radiological impact of accidents and unexpected events in the disposal of Below Regulatory Concern (BRC) waste. The accident analysis considers the transportation, incineration, and disposal of BRC waste as municipal solid waste. The potential greatest radiological impact for each type of accident is identified through the use of event trees. These accident events are described in terms of the generic waste property(ies) (e.g., flammability, dispersibility, leachability, and solubility) that cause the greatest radiological impact. 7 refs., 32 figs., 12 tabs

Three Mile Island accident: a case study of life event appraisal

International Nuclear Information System (INIS)

Goldsteen, R.L.

1983-01-01

This research investigates community reactions to the accident at the Three Mile Island (TMI) nuclear powered electric generating plant in March, 1979. The investigation is placed in the context of life event research and chooses an appraisal orientation. Three innovations are argued: 1) perceived consequences of the event best predict reactions to it, 2) the attitudes of significant others toward the event influence reactions to the accident under certain circumstances, and 3) sense of well-being is a good outcome measure for a general population. The hypotheses posit that the attitudes of others will affect sense of well-being only when individual attitudes concerning the consequences of the accident are moderate; when individual attitudes are extreme, the attitudes of others will have no demonstrable effect on outcomes. The findings did not support all the prediction of the hypotheses. However, they indicate that perceived consequences are the best predictors of sense of well-being and that an individual's attitudinal position, his strength of attitude, and the nature of the stimulus are highly related to whether or not an individual will be influenced by the views of others

Some implications of an event-based definition of exposure to the risk of road accident

DEFF Research Database (Denmark)

Elvik, Rune

2015-01-01

This paper proposes a new definition of exposure to the risk of road accident as any event, limited in space and time, representing a potential for an accident to occur by bringing road users close to each other in time or space of by requiring a road user to take action to avoid leaving the road......This paper proposes a new definition of exposure to the risk of road accident as any event, limited in space and time, representing a potential for an accident to occur by bringing road users close to each other in time or space of by requiring a road user to take action to avoid leaving...

Life Change Events as a Predictor of Accident Incidence in a College Population.

Science.gov (United States)

Furney, Steven R.

1983-01-01

To test the relationship between stressful life-change events and accident incidence, researchers administered the College Schedule of Recent Experience to male students at a large midwestern university. The study's implications for identifying high-risk persons and for accident prevention are discussed. (PP)

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.)

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

International Nuclear Information System (INIS)

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

1986-01-01

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

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.

Probabilistic Dynamics for Integrated Analysis of Accident Sequences considering Uncertain Events

Directory of Open Access Journals (Sweden)

Robertas Alzbutas

2015-01-01

Full Text Available The analytical/deterministic modelling and simulation/probabilistic methods are used separately as a rule in order to analyse the physical processes and random or uncertain events. However, in the currently used probabilistic safety assessment this is an issue. The lack of treatment of dynamic interactions between the physical processes on one hand and random events on the other hand causes the limited assessment. In general, there are a lot of mathematical modelling theories, which can be used separately or integrated in order to extend possibilities of modelling and analysis. The Theory of Probabilistic Dynamics (TPD and its augmented version based on the concept of stimulus and delay are introduced for the dynamic reliability modelling and the simulation of accidents in hybrid (continuous-discrete systems considering uncertain events. An approach of non-Markovian simulation and uncertainty analysis is discussed in order to adapt the Stimulus-Driven TPD for practical applications. The developed approach and related methods are used as a basis for a test case simulation in view of various methods applications for severe accident scenario simulation and uncertainty analysis. For this and for wider analysis of accident sequences the initial test case specification is then extended and discussed. Finally, it is concluded that enhancing the modelling of stimulated dynamics with uncertainty and sensitivity analysis allows the detailed simulation of complex system characteristics and representation of their uncertainty. The developed approach of accident modelling and analysis can be efficiently used to estimate the reliability of hybrid systems and at the same time to analyze and possibly decrease the uncertainty of this estimate.

Indemnification of damage in the event of a nuclear accident

International Nuclear Information System (INIS)

2003-01-01

The Workshop on the Indemnification of Damage in the Event of a Nuclear Accident, organised by the OECD Nuclear Energy Agency in close co-operation with the French authorities, was held in Paris from 26 to 28 November 2001. This event was an integral part of the International Nuclear Emergency Exercise INEX 2000. It attracted wide participation from national nuclear authorities, regulators, operators of nuclear installations, nuclear insurers and international organisations. The objective was to test the capacity of the existing nuclear liability and compensation mechanisms in the 29 countries represented at the workshop to manage the consequences of a nuclear emergency. This workshop was based upon the scenario used for the INEX 2000 Exercise, i.e. an accident simulated at the Gravelines nuclear power plant in the north of France in May 2001. These proceedings contain a comparative analysis of legislative and regulatory provisions governing emergency response and nuclear third party liability, based upon country replies to a questionnaire. This publication also includes the full responses provided to that questionnaire, as well as the texts of presentations made by special guests from Germany and Japan describing the manner in which the public authorities in their respective countries responded to two nuclear accidents of a very different nature and scale. (authors)

Statistical evaluation of design-error related accidents

International Nuclear Information System (INIS)

Ott, K.O.; Marchaterre, J.F.

1980-01-01

In a recently published paper (Campbell and Ott, 1979), a general methodology was proposed for the statistical evaluation of design-error related accidents. The evaluation aims at an estimate of the combined residual frequency of yet unknown types of accidents lurking in a certain technological system. Here, the original methodology is extended, as to apply to a variety of systems that evolves during the development of large-scale technologies. A special categorization of incidents and accidents is introduced to define the events that should be jointly analyzed. The resulting formalism is applied to the development of the nuclear power reactor technology, considering serious accidents that involve in the accident-progression a particular design inadequacy

Pilot study of dynamic Bayesian networks approach for fault diagnostics and accident progression prediction in HTR-PM

Energy Technology Data Exchange (ETDEWEB)

Zhao, Yunfei; Tong, Jiejuan; Zhang, Liguo, E-mail: lgzhang@tsinghua.edu.cn; Zhang, Qin

2015-09-15

Highlights: • Dynamic Bayesian network is used to diagnose and predict accident progress in HTR-PM. • Dynamic Bayesian network model of HTR-PM is built based on detailed system analysis. • LOCA Simulations validate the above model even if part monitors are lost or false. - Abstract: The first high-temperature-reactor pebble-bed demonstration module (HTR-PM) is under construction currently in China. At the same time, development of a system that is used to support nuclear emergency response is in progress. The supporting system is expected to complete two tasks. The first one is diagnostics of the fault in the reactor based on abnormal sensor measurements obtained. The second one is prognostic of the accident progression based on sensor measurements obtained and operator actions. Both tasks will provide valuable guidance for emergency staff to take appropriate protective actions. Traditional method for the two tasks relies heavily on expert judgment, and has been proven to be inappropriate in some cases, such as Three Mile Island accident. To better perform the two tasks, dynamic Bayesian networks (DBN) is introduced in this paper and a pilot study based on the approach is carried out. DBN is advantageous in representing complex dynamic systems and taking full consideration of evidences obtained to perform diagnostics and prognostics. Pearl's loopy belief propagation (LBP) algorithm is recommended for diagnostics and prognostics in DBN. The DBN model of HTR-PM is created based on detailed system analysis and accident progression analysis. A small break loss of coolant accident (SBLOCA) is selected to illustrate the application of the DBN model of HTR-PM in fault diagnostics (FD) and accident progression prognostics (APP). Several advantages of DBN approach compared with other techniques are discussed. The pilot study lays the foundation for developing the nuclear emergency response supporting system (NERSS) for HTR-PM.

Restructuring of an Event Tree for a Loss of Coolant Accident in a PSA model

International Nuclear Information System (INIS)

Lim, Ho-Gon; Han, Sang-Hoon; Park, Jin-Hee; Jang, Seong-Chul

2015-01-01

Conventional risk model using PSA (probabilistic Safety Assessment) for a NPP considers two types of accident initiators for internal events, LOCA (Loss of Coolant Accident) and transient event such as Loss of electric power, Loss of cooling, and so on. Traditionally, a LOCA is divided into three initiating event (IE) categories depending on the break size, small, medium, and large LOCA. In each IE group, safety functions or systems modeled in the accident sequences are considered to be applicable regardless of the break size. However, since the safety system or functions are not designed based on a break size, there exist lots of mismatch between safety system/function and an IE, which may make the risk model conservative or in some case optimistic. Present paper proposes new methodology for accident sequence analysis for LOCA. We suggest an integrated single ET construction for LOCA by incorporating a safety system/function and its applicable break spectrum into the ET. Integrated accident sequence analysis in terms of ET for LOCA was proposed in the present paper. Safety function/system can be properly assigned if its applicable range is given by break set point. Also, using simple Boolean algebra with the subset of the break spectrum, final accident sequences are expressed properly in terms of the Boolean multiplication, the occurrence frequency and the success/failure of safety system. The accident sequence results show that the accident sequence is described more detailed compared with the conventional results. Unfortunately, the quantitative results in terms of MCS (minimal Cut-Set) was not given because system fault tree was not constructed for this analysis and the break set points for all 7 point were not given as a specified numerical quantity. Further study may be needed to fix the break set point and to develop system fault tree

Restructuring of an Event Tree for a Loss of Coolant Accident in a PSA model

Energy Technology Data Exchange (ETDEWEB)

Lim, Ho-Gon; Han, Sang-Hoon; Park, Jin-Hee; Jang, Seong-Chul [KAERI, Daejeon (Korea, Republic of)

2015-05-15

Conventional risk model using PSA (probabilistic Safety Assessment) for a NPP considers two types of accident initiators for internal events, LOCA (Loss of Coolant Accident) and transient event such as Loss of electric power, Loss of cooling, and so on. Traditionally, a LOCA is divided into three initiating event (IE) categories depending on the break size, small, medium, and large LOCA. In each IE group, safety functions or systems modeled in the accident sequences are considered to be applicable regardless of the break size. However, since the safety system or functions are not designed based on a break size, there exist lots of mismatch between safety system/function and an IE, which may make the risk model conservative or in some case optimistic. Present paper proposes new methodology for accident sequence analysis for LOCA. We suggest an integrated single ET construction for LOCA by incorporating a safety system/function and its applicable break spectrum into the ET. Integrated accident sequence analysis in terms of ET for LOCA was proposed in the present paper. Safety function/system can be properly assigned if its applicable range is given by break set point. Also, using simple Boolean algebra with the subset of the break spectrum, final accident sequences are expressed properly in terms of the Boolean multiplication, the occurrence frequency and the success/failure of safety system. The accident sequence results show that the accident sequence is described more detailed compared with the conventional results. Unfortunately, the quantitative results in terms of MCS (minimal Cut-Set) was not given because system fault tree was not constructed for this analysis and the break set points for all 7 point were not given as a specified numerical quantity. Further study may be needed to fix the break set point and to develop system fault tree.

Regulatory approach to enhanced human performance during accidents

International Nuclear Information System (INIS)

Palla, R.L. Jr.

1990-01-01

It has become increasingly clear in recent years that the risk associated with nuclear power is driven by human performance. Although human errors have contributed heavily to the two core-melt events that have occurred at power reactors, effective performance during an event can also prevent a degraded situation from progressing to a more serious accident, as in the loss-of-feedwater event at Davis-Besse. Sensitivity studies in which human error rates for various categories of errors in a probabilistic risk assessment (PRA) were varied confirm the importance of human performance. Moreover, these studies suggest that actions taken during an accident are at least as important as errors that occur prior to an initiating event. A program that will lead to enhanced accident management capabilities in the nuclear industry is being developed by the US Nuclear Regulatory Commission (NRC) and industry and is a key element in NRC's integration plan for closure of severe-accident issues. The focus of the accident management (AM) program is on human performance during accidents, with emphasis on in-plant response. The AM program extends the defense-in-depth principle to plant operating staff. The goal is to take advantage of existing plant equipment and operator skills and creativity to find ways to terminate accidents that are beyond the design basis. The purpose of this paper is to describe the NRC's objectives and approach in AM as well as to discuss several human performance issues that are central to AM

Use of simulators in severe accident management

International Nuclear Information System (INIS)

Evans, R.C.

1994-01-01

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

Fukushima Daiichi Unit 1 Accident Progression Uncertainty Analysis and Implications for Decommissioning of Fukushima Reactors - Volume I.

Energy Technology Data Exchange (ETDEWEB)

Gauntt, Randall O. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Mattie, Patrick D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2016-01-01

Sandia National Laboratories (SNL) has conducted an uncertainty analysis (UA) on the Fukushima Daiichi unit (1F1) accident progression with the MELCOR code. The model used was developed for a previous accident reconstruction investigation jointly sponsored by the US Department of Energy (DOE) and Nuclear Regulatory Commission (NRC). That study focused on reconstructing the accident progressions, as postulated by the limited plant data. This work was focused evaluation of uncertainty in core damage progression behavior and its effect on key figures-of-merit (e.g., hydrogen production, reactor damage state, fraction of intact fuel, vessel lower head failure). The primary intent of this study was to characterize the range of predicted damage states in the 1F1 reactor considering state of knowledge uncertainties associated with MELCOR modeling of core damage progression and to generate information that may be useful in informing the decommissioning activities that will be employed to defuel the damaged reactors at the Fukushima Daiichi Nuclear Power Plant. Additionally, core damage progression variability inherent in MELCOR modeling numerics is investigated.

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

International Nuclear Information System (INIS)

Huerta B, A.; Aguilar T, O.; Nunez C, A.; Lopez M, R.

1994-01-01

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

Insights from Severe Accident Analyses for Verification of VVER SAMG

Energy Technology Data Exchange (ETDEWEB)

Gaikwad, A. J.; Rao, R. S.; Gupta, A.; Obaidurrahaman, K., E-mail: avinashg@aerb.gov.in [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai (India)

2014-10-15

The severe accident analyses of simultaneous rupture of all four steam lines (case-a), simultaneous occurrence of LOCA with SBO (case-b) and Station blackout (case-c) were performed with the computer code ASTEC V2r2 for a typical VVER-1000. The results obtained will be used for verification of sever accident provisions and Severe Accident Management Guidelines (SAMG). Auxiliary feed water and emergency core cooling systems are modelled as boundary conditions. The ICARE module is used to simulate the reactor core, which is divided into five radial regions by grouping similarly powered fuel assemblies together. Initially, CESAR module computes thermal hydraulics in primary and secondary circuits. As soon as core uncovery begins, the ICARE module is actuated based on certain parameters, and after this, ICARE module computes the thermal hydraulics in the core, bypass, downcomer and the lower plenum. CESAR handles the remaining components in the primary and secondary loops. CPA module is used to simulate the containment and to predict the thermal-hydraulic and hydrogen behaviour in the containment. The accident sequences were selected in such a way that they cover low/high pressure and slow/fast core damage progression events. Events simulated included slow progression events with high pressure and fast accident progression with low primary pressure. Analysis was also carried out for the case of SBO with the opening of the PORVs when core exit temperature exceeds certain value as part of SAMG. Time step sensitivity study was carried out for LOCA with SBO. In general the trends and magnitude of the parameters are as expected. The key results of the above analyses are presented in this paper. (author)

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Accident analyses in nuclear power plants following external initiating events and in the shutdown state. Final report

International Nuclear Information System (INIS)

Loeffler, Horst; Kowalik, Michael; Mildenberger, Oliver; Hage, Michael

2016-06-01

The work which is documented here provides the methodological basis for improvement of the state of knowledge for accident sequences after plant external initiating events and for accident sequences which begin in the shutdown state. The analyses have been done for a PWR and for a BWR reference plant. The work has been supported by the German federal ministry BMUB under the label 3612R01361. Top objectives of the work are: - Identify relevant event sequences in order to define characteristic initial and boundary conditions - Perform accident analysis of selected sequences - Evaluate the relevance of accident sequences in a qualitative way The accident analysis is performed with the code MELCOR 1.8.6. The applied input data set has been significantly improved compared to previous analyses. The event tree method which is established in PSA level 2 has been applied for creating a structure for a unified summarization and evaluation of the results from the accident analyses. The computer code EVNTRE has been applied for this purpose. In contrast to a PSA level 2, the branching probabilities of the event tree have not been determined with the usual accuracy, but they are given in an approximate way only. For the PWR, the analyses show a considerable protective effect of the containment also in the case of beyond design events. For the BWR, there is a rather high probability for containment failure under core melt impact, but nevertheless the release of radionuclides into the environment is very limited because of plant internal retention mechanisms. This report concludes with remarks about existing knowledge gaps and with regard to core melt sequences, and about possible improvements of the plant safety.

Development of severe accident evaluation technology (level 2 PSA) for sodium-cooled fast reactors. (3) Identification of dominant factors in transition phase of unprotected events

International Nuclear Information System (INIS)

Tobita, Yoshiharu; Yamano, Hidemasa; Sato, Ikken

2009-01-01

The event progression of the transition phase in the unprotected loss of flow accident of the JSFR design concept was analyzed using the SIMMER-III code reflecting the knowledge obtained from the EAGLE experimental program. It was clarified through the parametric calculations that the fuel discharge behavior through the paths such as the inner duct of modified-FAIDUS and control-rod guide tube is playing a very important role. Effective fuel discharge through these paths prevents possibility of severe recriticality events. Important factors dominating the transition phase were identified through these parametric calculations. (author)

Outline of Fukushima nuclear accident and future action. Lessons learned from accident and countermeasure plan

International Nuclear Information System (INIS)

Fukuda, Toshihiko

2012-01-01

Fukushima nuclear accident was caused by loss of all AC power sources (SBO) and loss of ultimate heat sink (LUHS) at Fukushima Daiichi Nuclear Power Plants (NPPs) hit by the Great East Japan Earthquake. This article reviewed outline of Fukushima nuclear accident progression when on year had passed since and referred to lessons learned from accident and countermeasure plan to prevent severe accident in SBO and LUHS events by earthquake and tsunami as future action. This countermeasure would be taken to (1) prevent serious flooding in case a tsunami overwhelms the breakwater, with improving water tightness of rooms for emergency diesel generator, batteries and power centers, (2) enhance emergency power supply and cooling function with mobile electricity generator, high pressure fire pump car and alternate water supply source, (3) mitigate environmental effects caused by core damage with installing containment filtered venting, and (4) enforce emergency preparedness in case of severe accident. Definite countermeasure plan for Kashiwazaki-Kariwa NPPs was enumerated. (T. Tanaka)

Selection of events at Ukrainian NPPs using the algorithm based on accident precursor method

International Nuclear Information System (INIS)

Vorontsov, D.V.; Lyigots'kij, O.Yi.; Serafin, R.Yi.; Tkachova, L.M.

2012-01-01

The paper describes a general approach to the first stage of research and development on analysis of Ukrainian NPP operation events from 1 January 2000 to 31 December 2010 using the accident precursor approach. Groups of potentially important events formed after their selection and classification are provided

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

Upgrading the electrical system of the IEA-R1 reactor to avoid triggering event of accidents

International Nuclear Information System (INIS)

Mello, Jose Roberto de; Madi Filho, Tufic

2015-01-01

The IEA-R1 research reactor at the Institute of Energy and Nuclear Research (IPEN) is a research reactor open pool type, built and designed by the American firm 'Babcox and Wilcox', having as coolant and moderator demineralized light water and Beryllium and graphite, as reflectors. The power supply system is designed to meet the electricity demand required by the loads of the reactor (Security systems and systems not related to security) in different situations the plant can meet, such as during startup, normal operation at power, shutdown, maintenance, exchange of fuel elements and accident situations. Studies have been done on possible accident initiating events and deterministic techniques were applied to assess the consequences of such incidents. Thus, the methods used to identify and select the accident initiating events, the methods of analysis of accidents, including sequence of events, transient analysis and radiological consequences, have been described. Finally, acceptance criteria of radiological doses are described. Only a brief summary of the item concerning loss of electrical power will be presented. The loss of normal electrical power at the IEA-R1 reactor is very common. In the case of Electric External Power Loss, at the IEA-R1 reactor building, there may be different sequences of events, as described below. When the supply of external energy in the IEA-R1 facility fails, the Electrical Distribution Vital System, consisting of 4 (four) generators type 'UPS', starts operation, immediately and it will continue supplying power to the reactor control table, core cooling system and other security systems. To contribute to security, in the electric power failure, starts to operate the Emergency Cooling System (SRE). SRE has the function of removing residual heat from the core to prevent the melting of fuel elements in the event of loss of refrigerant to the core. Adding to the generators with batteries group system, new auxiliary

Preliminary Analysis of Severe Accident Progression Initiated from Small Break LOCA of a SMART Reactor

International Nuclear Information System (INIS)

Jin, Young Ho; Park, Jong Hwa; Kim, Dong Ha; Cho, Seong Won

2010-01-01

SMART (System integrated Modular Advanced ReacTor), is under the development at Korea Atomic Energy Research Institute (KAERI). SMART is an integral type pressurized water reactor which contains a pressurizer, 4 reactor coolant pumps (RCPs), and 8 steam generator cassettes(S/Gs) in a single reactor vessel. This reactor has substantially enhanced its safety with an integral layout of its major components, 4 trains of safety injection systems (SISs), and an adoption of 4 trains of passive residual heat removal systems (PRHRS) instead of an active auxiliary feedwater system . The thermal power is 330 MWth. During the conceptual design stage, a preliminary PSA was performed. PSA results identified that a small break loss of coolant accident (SLOCA) with all safety injections unavailable is one of important severe core damage sequences. Clear understanding of this sequence helps in the developing accident mitigation strategies. MIDAS/SMR computer code is used to simulate the severe accident progression initiated from a small break LOCA in SMART reactor. This code has capability to model a helical steam generator which is adopted in SMART reactor. The important accident progression results for SMART reactor are then compared with the typical pressurized water reactor (PWR) result

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

events and accidents at NPPs, and what additional measures should be considered as an accident progresses to the severe accident stage. Insights are provided on the experiences and practices existing or being proposed in the NEA member states, as well as new findings from post-Fukushima studies. Emphasis is placed on identifying commendable practices that support enhanced and integrated on-site accident management response and decision-making by NPP operators. The report provides information (including commendable practices) useful for regulatory authorities to consider as they implement enhancements to their regulatory framework in the area of integrated accident management building on the lessons learnt from the Fukushima Daiichi NPP accident. The report's insights also should be useful to regulatory authorities, operating organisations and others in the nuclear safety community for addressing accident management issues such as procedures and guidelines, equipment, infrastructure and instrumentation, and human and organisational resources. Factors such as accidents involving spent fuel pools, multi-unit aspects of accident management, the interface between onsite and off-site organisations and resources, and degradation of the surrounding infrastructure are also discussed. (authors)

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.)

'It was a freak accident': an analysis of the labelling of injury events in the US press.

Science.gov (United States)

Smith, Katherine C; Girasek, Deborah C; Baker, Susan P; Manganello, Jennifer A; Bowman, Stephen M; Samuels, Alicia; Gielen, Andrea C

2012-02-01

Given that the news media shape our understanding of health issues, a study was undertaken to examine the use by the US media of the expression 'freak accident' in relation to injury events. This analysis is intended to contribute to the ongoing consideration of lay conceptualisation of injuries as 'accidents'. LexisNexis Academic was used to search three purposively selected US news sources (Associated Press, New York Times and Philadelphia Inquirer) for the expression 'freak accident' over 5 years (2005-9). Textual analysis included both structured and open coding. Coding included measures for who used the expression within the story, the nature of the injury event and the injured person(s) being reported upon, incorporation of prevention information within the story and finally a phenomenological consideration of the uses and meanings of the expression within the story context. Results The search yielded a dataset of 250 human injury stories incorporating the term 'freak accident'. Injuries sustained by professional athletes dominated coverage (61%). Fewer than 10% of stories provided a clear and explicit injury prevention message. Stories in which journalists employed the expression 'freak accident' were less likely to include prevention information than stories in which the expression was used by people quoted in the story. Journalists who frame injury events as freak accidents may be an appropriate focus for advocacy efforts. Effective prevention messages should be developed and disseminated to accompany injury reporting in order to educate and protect the public.

Heat transfer phenomena revelant to severe accidents

International Nuclear Information System (INIS)

Dallman, R.J.; Duffey, R.B.

1990-01-01

A number of aspects of severe accidents have been reviewed, particularly in relation to the heat transfer characteristics and the important phenomena. It is shown that natural circulation, forced convection, and entrainment phenomena are important for both the reactor system and ex-vessel events. It is also shown that the phenomena related to two component enhanced heat transfer is important in the pool of molten core debris, in relation to the potential for attack of the liner structure and the concrete. These mechanisms are discussed within the general context of severe accident progression

Heat transfer phenomena relevant to severe accidents

International Nuclear Information System (INIS)

Dallman, R.J.; Duffey, R.B.

1990-01-01

A number of aspects of severe accidents have been reviewed, particularly in relation to the heat transfer characteristics and the important phenomena. It is shown that natural circulation, forced convection, and entrainment phenomena are important for both the reactor system and ex-vessel events. It is also shown that the phenomena related to two component enhanced heat transfer is important in the pool of molten core debris, in relation to the potential for attack of the liner structure and the concrete. These mechanisms are discussed within the general context of severe accident progression. 26 refs

Severe accidents and operator training - discussion of potential issues

International Nuclear Information System (INIS)

Vidard, Michel

1997-01-01

R and D programs developed throughout the world allowed significant progress in the understanding of physical phenomena and Severe Accident Management (SAM) programs started in many OECD countries. Basically, the common denominator to all these SAM programs was to provide utility operators with procedures or guidelines allowing to deal with complex situations not formally considered in the Design Basis, including accidents where a significant portion of the core had molten. These SAM procedures or guidelines complement the traditional accident management procedures (event, symptom or physical-state oriented) and should allow operators to deal with a reasonably bounding set of situations. Dealing with operator or crisis team training, it was recognized that training would be beneficial but that training programs were lagging, i.e. though training sessions were either organized or contemplated after implementation of SAM programs, they seemed to be somewhat different from more traditional training sessions on Accident Management. After some explanations on the differences between Design Basis Accidents (DBAs) and Beyond Design Basis Accidents (BDBAs), this paper underlines some potential difficulties for training operators and discuss problems to be addressed by organisms contemplating SAM training sessions consistent with similar activities for less complex events

On-site habitability in the event of an accident at a nuclear facility

International Nuclear Information System (INIS)

1989-01-01

This publication is intended to provide technical guidance and a methodology for regulatory bodies, designers, constructors and operators of nuclear facilities to assist them in assessing the current situation as regards on-site habitability for their specific nuclear facilities. Initially, the aim will be to ensure that the ''vital areas'' of the facility which are necessary for the safe operation and shutdown of the facility will remain habitable, in some cases continuously and in others transiently, in the event of an accident inside or outside the installation. The assessment procedure can be used not only for potential radiation accidents but also to consider the effects on habitability of those probable non-radiological events which, if not correctly and effectively countered, could lead to the development of potentially unsafe conditions in the facility itself. 30 refs, 4 figs, 8 tabs

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

International Nuclear Information System (INIS)

Saleh, Joseph H.; Saltmarsh, Elizabeth A.; Favarò, Francesca M.; Brevault, Loïc

2013-01-01

An important consideration in safety analysis and accident prevention is the identification of and response to accident precursors. These off-nominal events are opportunities to recognize potential accident pathogens, identify overlooked accident sequences, and make technical and organizational decisions to address them before further escalation can occur. When handled properly, the identification of precursors provides an opportunity to interrupt an accident sequence from unfolding; when ignored or missed, precursors may only provide tragic proof after the fact that an accident was preventable. In this work, we first provide a critical review of the concept of precursor, and we highlight important features that ought to be distinguished whenever accident precursors are discussed. We address for example the notion of ex-ante and ex-post precursors, identified for postulated and instantiated (occurred) accident sequences respectively, and we discuss the feature of transferability of precursors. We then develop a formal (mathematical) definition of accident precursors as truncated accident sequences within the modeling framework of Discrete Event Systems. Additionally, we examine the related notions of “accident pathogens” as static or lurking adverse conditions that can contribute to or aggravate an accident, as well as “near misses”, “warning signs” and the novel concept of “accident pathway”. While these terms are within the same linguistic neighborhood as “accident precursors”, we argue that there are subtle but important differences between them and recommend that they not be used interchangeably for the sake of accuracy and clarity of communication within the risk and safety community. We also propose venues for developing quantitative importance measures for accident precursors, similar to component importance measures in reliability engineering. Our objective is to establish a common understanding and clear delineation of these terms, and

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Development of accident event trees and evaluation of safety system failure modes for the nuclear ultra large crude carrier

International Nuclear Information System (INIS)

Lewe, C.K.; Coffey, R.S.; Goodwin, E.F.; Maltese, J.G.; Pyatt, D.W.

1978-01-01

A method of applying the probabilistic accident event tree methodology to safety assessments of a nuclear powered Ultra Large Crude Carrier is presented. Also presented are the procedures by which an external accident initiating event, such as a ship collision, may be correlated with the probabilities of damage to the ship's safety systems and to their ultimate availabilities to perform required safety functions

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-08-01

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

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

International Nuclear Information System (INIS)

1996-08-01

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

Key Characteristics of Combined Accident including TLOFW accident for PSA Modeling

Energy Technology Data Exchange (ETDEWEB)

Kim, Bo Gyung; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of); Yoon, Ho Joon [Khalifa University of Science, Technology and Research, Abu Dhabi (United Arab Emirates)

2015-05-15

The conventional PSA techniques cannot adequately evaluate all events. The conventional PSA models usually focus on single internal events such as DBAs, the external hazards such as fire, seismic. However, the Fukushima accident of Japan in 2011 reveals that very rare event is necessary to be considered in the PSA model to prevent the radioactive release to environment caused by poor treatment based on lack of the information, and to improve the emergency operation procedure. Especially, the results from PSA can be used to decision making for regulators. Moreover, designers can consider the weakness of plant safety based on the quantified results and understand accident sequence based on human actions and system availability. This study is for PSA modeling of combined accidents including total loss of feedwater (TLOFW) accident. The TLOFW accident is a representative accident involving the failure of cooling through secondary side. If the amount of heat transfer is not enough due to the failure of secondary side, the heat will be accumulated to the primary side by continuous core decay heat. Transients with loss of feedwater include total loss of feedwater accident, loss of condenser vacuum accident, and closure of all MSIVs. When residual heat removal by the secondary side is terminated, the safety injection into the RCS with direct primary depressurization would provide alternative heat removal. This operation is called feed and bleed (F and B) operation. Combined accidents including TLOFW accident are very rare event and partially considered in conventional PSA model. Since the necessity of F and B operation is related to plant conditions, the PSA modeling for combined accidents including TLOFW accident is necessary to identify the design and operational vulnerabilities.The PSA is significant to assess the risk of NPPs, and to identify the design and operational vulnerabilities. Even though the combined accident is very rare event, the consequence of combined

Severe accident progression perspectives for Mark I containments based on the IPE results

International Nuclear Information System (INIS)

Lin, C.C.; Lehner, J.R.; Pratt, W.T.; Drouin, M.

1995-01-01

Based on level 2 analyses in IPE (Individual Plant Examination) submittals accident progression, perspectives were obtained for all containment types. These perspectives consisted of insights on containment failure modes, releases therein, and factors responsible for the results. To illustrate the types of perspectives acquired on severe accident progresssion, insights obtained for (BWR) Mark I containments are discussed here. Mark I containments have high strength but small volumes and rely on pressure suppression pools to condense steam released from the reactor coolant system during an accident. Accidents causing structural failure of the drywell shortly after the core debris melts through the reactor vessel were found to be dominant contributors to risk. Importance of individual containment failure mechanisms depends on plant features and in some cases on modeling assumptions; however the following mechanisms were found important: drywell shell melt-through caused by direct contact with core debris and drywell failure caused by rapid pressure/temperature pulses at time of vessel melt-through. Drywell failure caused by gradual pressure/temperature buildup due to gases and steam released during core/concrete interactions is important in some IPEs. In other IPEs vent was an important contributor. However, accidents that bypass containment (eg interfacing systems LOCA)or involve containment isolation failure were not important contributors to the CDF in any of the IPEs for Mark I plants. These accidents are also not important to risk (even though they can involve large fission product release) because their frequencies of occurrence are so much lower than frequencies of early structural failure caused by other accidents that dominate the CDF

Procedural and submittal guidance for the individual plant examination of external events (IPEEE) for severe accident vulnerabilities

International Nuclear Information System (INIS)

Chen, J.T.; Chokshi, N.C.; Kenneally, R.M.; Kelly, G.B.; Beckner, W.D.; McCracken, C.; Murphy, A.J.; Reiter, L.; Jeng, D.

1991-06-01

Based on a Policy statement on Severe Accidents, the licensee of each nuclear power plant is requested to perform an individual plant examination. The plant examination systematically looks for vulnerabilities to severe accidents and cost-effective safety improvements that reduce or eliminate the important vulnerabilities. This document presents guidance for performing and reporting the results of the individual plant examination of external events (IPEEE). The guidance for reporting the results of the individual plant examination of internal events (IPE) is presented in NUREG-1335. 53 refs., 1 figs., 2 tabs

Thermal-hydraulic uncertainties affecting severe accident progression

International Nuclear Information System (INIS)

Haskin, F.E.; Behr, V.L.

1984-01-01

To provide the proper technical bases for decisions regarding severe accidents, the US Nuclear Regulatory Commission (NRC) is sponsoring the following activities: (a) a variety of severe accident research programs, combined under the Severe Accident Research Plan; (b) nationwide task forces on containment loading, containment response, and fission product source terms; (c) a review by the American Physical Society of state-of-the-art methods for calculating radiological source terms; and (d) technical exchange meetings with the Industry Degraded Core (IDCOR) program. One of the means for integrating this developing array of technical information is the Severe Accident Risk Reduction Program (SARRP). One of the current SARRP objectives is to utilize insights gained from the activities listed above to characterize the relative likelihoods of competing containment failure modes for core-melt accidents

WHEN MODEL MEETS REALITY – A REVIEW OF SPAR LEVEL 2 MODEL AGAINST FUKUSHIMA ACCIDENT

Energy Technology Data Exchange (ETDEWEB)

Zhegang Ma

2013-09-01

The Standardized Plant Analysis Risk (SPAR) models are a set of probabilistic risk assessment (PRA) models used by the Nuclear Regulatory Commission (NRC) to evaluate the risk of operations at U.S. nuclear power plants and provide inputs to risk informed regulatory process. A small number of SPAR Level 2 models have been developed mostly for feasibility study purpose. They extend the Level 1 models to include containment systems, group plant damage states, and model containment phenomenology and accident progression in containment event trees. A severe earthquake and tsunami hit the eastern coast of Japan in March 2011 and caused significant damages on the reactors in Fukushima Daiichi site. Station blackout (SBO), core damage, containment damage, hydrogen explosion, and intensive radioactivity release, which have been previous analyzed and assumed as postulated accident progression in PRA models, now occurred with various degrees in the multi-units Fukushima Daiichi site. This paper reviews and compares a typical BWR SPAR Level 2 model with the “real” accident progressions and sequences occurred in Fukushima Daiichi Units 1, 2, and 3. It shows that the SPAR Level 2 model is a robust PRA model that could very reasonably describe the accident progression for a real and complicated nuclear accident in the world. On the other hand, the comparison shows that the SPAR model could be enhanced by incorporating some accident characteristics for better representation of severe accident progression.

Accidents in nuclear ships

Energy Technology Data Exchange (ETDEWEB)

Oelgaard, P L [Risoe National Lab., Roskilde (Denmark); [Technical Univ. of Denmark, Lyngby (Denmark)

1996-12-01

This report starts with a discussion of the types of nuclear vessels accidents, in particular accidents which involve the nuclear propulsion systems. Next available information on 61 reported nuclear ship events in considered. Of these 6 deals with U.S. ships, 54 with USSR ships and 1 with a French ship. The ships are in almost all cases nuclear submarines. Only events that involve the sinking of vessels, the nuclear propulsion plants, radiation exposures, fires/explosions, sea-water leaks into the submarines and sinking of vessels are considered. For each event a summary of available information is presented, and comments are added. In some cases the available information is not credible, and these events are neglected. This reduces the number of events to 5 U.S. events, 35 USSR/Russian events and 1 French event. A comparison is made between the reported Soviet accidents and information available on dumped and damaged Soviet naval reactors. It seems possible to obtain good correlation between the two types of events. An analysis is made of the accident and estimates are made of the accident probabilities which are found to be of the order of 10{sup -3} per ship reactor years. It if finally pointed out that the consequences of nuclear ship accidents are fairly local and does in no way not approach the magnitude of the Chernobyl accident. It is emphasized that some of the information on which this report is based, may not be correct. Consequently some of the results of the assessments made may not be correct. (au).

Accidents in nuclear ships

International Nuclear Information System (INIS)

Oelgaard, P.L.

1996-12-01

This report starts with a discussion of the types of nuclear vessels accidents, in particular accidents which involve the nuclear propulsion systems. Next available information on 61 reported nuclear ship events in considered. Of these 6 deals with U.S. ships, 54 with USSR ships and 1 with a French ship. The ships are in almost all cases nuclear submarines. Only events that involve the sinking of vessels, the nuclear propulsion plants, radiation exposures, fires/explosions, sea-water leaks into the submarines and sinking of vessels are considered. For each event a summary of available information is presented, and comments are added. In some cases the available information is not credible, and these events are neglected. This reduces the number of events to 5 U.S. events, 35 USSR/Russian events and 1 French event. A comparison is made between the reported Soviet accidents and information available on dumped and damaged Soviet naval reactors. It seems possible to obtain good correlation between the two types of events. An analysis is made of the accident and estimates are made of the accident probabilities which are found to be of the order of 10 -3 per ship reactor years. It if finally pointed out that the consequences of nuclear ship accidents are fairly local and does in no way not approach the magnitude of the Chernobyl accident. It is emphasized that some of the information on which this report is based, may not be correct. Consequently some of the results of the assessments made may not be correct. (au)

Severe accident management at South Africa's Koeberg plant

International Nuclear Information System (INIS)

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

1997-01-01

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

Fukushima. The accident sequence and important causes. Pt. 1/3

International Nuclear Information System (INIS)

Pistner, Christoph

2013-01-01

On March 11, 2011 a strong earthquake at the east coast of Japan and a subsequent tsunami caused severe damage at the NPP site of Fukushima Daiichi. The article covers the fundamental safety aspects of the accident progress according to the state of knowledge. The principles of nuclear technology and reactor safety are summarized in order to allow the understanding of the accidental sequence. Even two years after the disaster many questions on the sequence of accident events are still open.

Containment event analysis for postulated severe accidents: Peach Bottom Atomic Power Station, Unit 2. Draft report for comment

Energy Technology Data Exchange (ETDEWEB)

Amos, C N [Technadyne Engineering Consultants, Inc., Albuquerque, NM (United States); Griesmeyer, J M [Sandia National Laboratories, Albuquerque, NM (United States); Kolaczkowski, A M [Science Applications International Corporation, Albuquerque, NM (United States)

1987-05-01

A study has been performed as part of the Severe Accident Risk Reduction Program (SARRP) to investigate the response of a particular boiling water reactor with a Mark I containment (Peach Bottom Unit 2) to postulated severe accidents. A detailed containment event tree for the Peach Bottom plant has been developed to describe the various possible accident pathways that can lead to radioactive releases from containment. Data and analyses from a large number of NRC and industry-sponsored programs have been reviewed and used as a basis for quantifying the event tree, i.e., determining the likelihood of the pathways at each branch point for a variety of accident sequence initiators. A generalized containment event tree code, called EVNTRE, has been developed to facilitate the quantification. The uncertainty in the results has been examined by performing the quantification three times, using a different set of input each time to represent the variation of opinion in the reactor safety community. In the so-called 'central' estimate, the likelihood of early containment failure (occurring before or within a short time after reactor vessel breach) was found to be significant because of the possible occurrence of the following phenomena that can threaten containment integrity: (1) meltthrough of the drywell shell caused by thermal attack from core debris, and (2) drywell overpressurization caused by rapid depressurization of the reactor vessel in combination with other events such as direct heating. However, uncertainties surrounding these issues could cause the early failure likelihood to be significantly lower than in the central estimate. This work supports NRC's assessment of severe accident risks to be published in NUREG-1150. (author)

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.

Work safety in the light of technological progress in mining. [Poland, accident and technology assessment for period 1950 to 1972

Energy Technology Data Exchange (ETDEWEB)

Stecko, R.

1976-01-01

This paper discusses positive and negative aspects of technological progress in black coal mining in Poland in the period from 1950 to 1972 on the basis of statistical data supplied by various institutions. Twenty three indexes characterizing both work safety and technological progress in underground coal mines are used. Equations of linear and non-linear even regression and equations of linear and non-linear multiple regression are derived on the basis of statistical records of 23 years. The equations are used to estimate indexes of work safety on the basis of technical progress indexes. It is suggested that from among 8 indexes characterizing work safety 4 are most useful and important: danger index, absenteeism index, frequency of underground fire index, and frequency of rock burst index. Analysis of the regression equation shows that technological progress in underground coal mines in Poland has caused a decrease in indexes describing: frequency of work accidents, frequency of underground fires, frequency of rock burst, and danger of underground work; and an increase in indexes characterizing: seriousness of work accidents, and absenteeism caused by work accidents. (84 refs.)

International policy on intervention in the event of a nuclear accident

International Nuclear Information System (INIS)

Jensen, P.H.; Crick, M.J.; Gonzalez, A.J.

1996-01-01

Criteria for taking particular protective actions with the aim of preventing or reducing radiation exposures to the population or to workers in the event of a nuclear accident or radiological emergency can be established on the basis of radiological protection principles for intervention situations. It is of utmost importance that pre-established intervention levels for different protective measures form an integral part of an emergency response plan. Generic optimized intervention levels and their derived operational quantities based on the principles given in this paper are judged to provide protection that would be justified and reasonable optimized for a wide range of accident situations although they can only be used as guidelines. Any specific optimization would lead to intervention levels that might be either higher or lower than those emerging from a generic optimization. (author). 9 refs

International policy on intervention in the event of a nuclear accident

Energy Technology Data Exchange (ETDEWEB)

Jensen, P H [Risoe National Lab., Roskilde (Denmark); Crick, M J; Gonzalez, A J [International Atomic Energy Agency, Vienna (Austria)

1996-08-01

Criteria for taking particular protective actions with the aim of preventing or reducing radiation exposures to the population or to workers in the event of a nuclear accident or radiological emergency can be established on the basis of radiological protection principles for intervention situations. It is of utmost importance that pre-established intervention levels for different protective measures form an integral part of an emergency response plan. Generic optimized intervention levels and their derived operational quantities based on the principles given in this paper are judged to provide protection that would be justified and reasonable optimized for a wide range of accident situations although they can only be used as guidelines. Any specific optimization would lead to intervention levels that might be either higher or lower than those emerging from a generic optimization. (author). 9 refs.

Identification and evaluation of accident sequences in nuclear power reactors

International Nuclear Information System (INIS)

Amendola, A.; Capobianchi, S.; Mancini, G.; Olivi, L.; Volta, G.; Reina, G.

1981-01-01

Probabilistic analysis techniques are being more and more used for the evaluation of accident progression in nuclear power plants, especially after the issue of the Reactor Safety Study (Report WASH-1400). This study and subsequent discussions have indicated the necessity of better investigating some major items, namely: adequate data base for the probabilistic evaluations; completeness of the analysis with respect both to accident initiation and behaviour; adequate treatment of uncertainties on the physical and operational parameters governing the accident behaviour. Furthermore, recent occurrences have stressed the importance of the operational aspects of reactor safety, such as plant-specific identification of possible occurrences, their prompt recognition, on-line prediction of subsequent developments and actions to be taken. The paper reviews the contributions in progress at JRC-Ispra to all these aspects, and specifically reports on the following: (1) The set-up of a European Reliability Data System for the acquisition and organisation of operational data of LWRs in the European Community. (2) The development of more complete and realistic models of systems. This work includes multistate static models of components and systems with a view to automatic fault-tree construction and dynamic models for accident sequence identification. The dynamic modelling approach ESCS (Event Sequence and Consequences Spectrum), shown in detail with an example, represents a step forward with respect to event-tree technique and opens new possibilities in dealing with human factors and on-line diagnosis problems. (3) The development of RSM (Response Surface Methodology) for the analysis of uncertainty propagations in consequence and in probability of accident chains. (author)

A digest of the Nuclear Safety Division report on the Fukushima Dai-ichi accident seminar (4). Issues identified by the accident

International Nuclear Information System (INIS)

Moriyama, Kumiaki; Abe, Kiyoharu

2013-01-01

AESJ Nuclear Safety Division published 'Report on the Fukushima Dai-ichi Accident Seminar - what was wrong and what should been down in future-' which would be published as five special articles of the AESJ journal. The Fukushima Dai-ichi accident identified issues of several activities directly related with nuclear safety in the areas of safety design, severe accident management and safety regulations. PRA, operational experiences and safety research could not always contribute safety assurance of nuclear power plant so much. This article (4) summarized technical issues based on related facts of the accident as much as possible and discussed' what was wrong and what should be down in future'. Important issues were identified from defense-in-depth philosophy and lessons learned on safety design were obtained from accident progression analysis. Activities against external events and continuous improvements of safety standards based on latest knowledge were most indispensable. Strong cooperation among experts in different areas was also needed. (T. Tanaka)

Investigation on accident management measures for VVER-1000 reactors

International Nuclear Information System (INIS)

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

2009-01-01

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

Domino effect in chemical accidents: main features and accident sequences.

Science.gov (United States)

Darbra, R M; Palacios, Adriana; Casal, Joaquim

2010-11-15

The main features of domino accidents in process/storage plants and in the transportation of hazardous materials were studied through an analysis of 225 accidents involving this effect. Data on these accidents, which occurred after 1961, were taken from several sources. Aspects analyzed included the accident scenario, the type of accident, the materials involved, the causes and consequences and the most common accident sequences. The analysis showed that the most frequent causes are external events (31%) and mechanical failure (29%). Storage areas (35%) and process plants (28%) are by far the most common settings for domino accidents. Eighty-nine per cent of the accidents involved flammable materials, the most frequent of which was LPG. The domino effect sequences were analyzed using relative probability event trees. The most frequent sequences were explosion→fire (27.6%), fire→explosion (27.5%) and fire→fire (17.8%). Copyright © 2010 Elsevier B.V. All rights reserved.

Visualization of Traffic Accidents

Science.gov (United States)

Wang, Jie; Shen, Yuzhong; Khattak, Asad

2010-01-01

Traffic accidents have tremendous impact on society. Annually approximately 6.4 million vehicle accidents are reported by police in the US and nearly half of them result in catastrophic injuries. Visualizations of traffic accidents using geographic information systems (GIS) greatly facilitate handling and analysis of traffic accidents in many aspects. Environmental Systems Research Institute (ESRI), Inc. is the world leader in GIS research and development. ArcGIS, a software package developed by ESRI, has the capabilities to display events associated with a road network, such as accident locations, and pavement quality. But when event locations related to a road network are processed, the existing algorithm used by ArcGIS does not utilize all the information related to the routes of the road network and produces erroneous visualization results of event locations. This software bug causes serious problems for applications in which accurate location information is critical for emergency responses, such as traffic accidents. This paper aims to address this problem and proposes an improved method that utilizes all relevant information of traffic accidents, namely, route number, direction, and mile post, and extracts correct event locations for accurate traffic accident visualization and analysis. The proposed method generates a new shape file for traffic accidents and displays them on top of the existing road network in ArcGIS. Visualization of traffic accidents along Hampton Roads Bridge Tunnel is included to demonstrate the effectiveness of the proposed method.

Development of severe accident evaluation technology (level 2 PSA) for sodium-cooled fast reactors. (5) Identification of dominant factors in ex-vessel accident sequences

International Nuclear Information System (INIS)

Ohno, Shuji; Seino, Hiroshi; Miyahara, Shinya

2009-01-01

The evaluation of accident progression outside of a reactor vessel (ex-vessel) and subsequent transfer behavior of radioactive materials is of great importance from the viewpoint of Level 2 PSA. Hence typical ex-vessel accident sequences in the JAEA Sodium-cooled Fast Reactor are qualitatively discussed in this paper and dominant behaviors or factors in the sequences are investigated through parametric calculations using the CONTAIN/LMR code. Scenarios to be focused on are, 1) sodium vapor leakage from the reactor vessel and 2) sodium-concrete reaction, which are both to be considered in the accident category of LOHRS (loss of heat removal system) and might be followed by an early containment failure due to the thermal effect of sodium combustion and hydrogen burning respectively. The calculated results clarify that the sodium vapor leak rate and the scale of sodium-concrete reaction are the important factors to dominate the ex-vessel accident progression. In addition to the understandings of the dominant factors, the analyzed results also provide the specific information such as pressure loading value to the containment and the timing of pressurization, which is indispensable as technical base in Level 2 PSA for developing event trees and for quantifying the accident consequences. (author)

Current status and issues of external event PSA for extreme natural hazards after Fukushima accident

International Nuclear Information System (INIS)

Choi, In-Kil; Hahm, Daegi; Kim, Min Kyu

2014-01-01

Extreme external events is emerged as significant risk contributor to the nuclear power plants after Fukushima Daiichi accident due to the catastrophic earthquake followed by great tsunami greater than a design basis. This accident shows that the extreme external events have the potential to simultaneously affect redundant and diverse safety systems and thereby induce common cause failure or common cause initiators. The probabilistic risk assessment methodology has been used for the risk assessment and safety improvement against the extreme natural hazards. The earthquake and tsunami hazard is an important issue for the nuclear industry in Korea. In this paper, the role and application of probabilistic safety assessment for the post Fukushima action will be introduced. For the evaluation of the extreme natural hazard, probabilistic seismic and tsunami hazard analysis is being performed for the safety enhancement. The research activity on the external event PSA and its interim results will be introduced with the issues to be solved in the future for the reliability enhancement of the risk analysis results. (authors)

30 years of the Goiania Accident: a comparative study with other radioactivity dispersion events

International Nuclear Information System (INIS)

Smith, Ricardo Bastos; Vicente, Roberto

2017-01-01

The year 2017 marks 30 years since the radioactive accident that occurred in the city of Goiania, capital of the state of Goias. It was the largest radiological accident in Brazil, and one of the largest in the world occurring outside nuclear facilities. Regarding the accidents at nuclear power plants, two of the biggest were Chernobyl in Ukraine, a year and a half before Goiania, and the Fukushima accident in Japan, in 2011. Different amounts of radioactive material were dispersed in the environment in each of these events. However, each one’s main pathway of dispersion was different: the accident of Goiania was terrestrial, Chernobyl was at the atmosphere, and Fukushima was mainly in the ocean. This work aims to study these different amounts, comparing such activities. In addition, it proposes to compare the sea dispersion of Fukushima with the amount of radioactive waste dumped in the oceans, when the release of radioactive waste at sea was permitted. It also proposes to compare the Chernobyl aerial dispersion with the radioactive material dissipated in the atmosphere, resulting from the more than 500 atmospheric nuclear tests conducted between 1945 and 1962 by the United States, the former Soviet Union, England, France and China. (author)

30 years of the Goiania Accident: a comparative study with other radioactivity dispersion events

Energy Technology Data Exchange (ETDEWEB)

Smith, Ricardo Bastos; Vicente, Roberto, E-mail: rbsmith@ipen.br, E-mail: rvicente@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

2017-07-01

The year 2017 marks 30 years since the radioactive accident that occurred in the city of Goiania, capital of the state of Goias. It was the largest radiological accident in Brazil, and one of the largest in the world occurring outside nuclear facilities. Regarding the accidents at nuclear power plants, two of the biggest were Chernobyl in Ukraine, a year and a half before Goiania, and the Fukushima accident in Japan, in 2011. Different amounts of radioactive material were dispersed in the environment in each of these events. However, each one’s main pathway of dispersion was different: the accident of Goiania was terrestrial, Chernobyl was at the atmosphere, and Fukushima was mainly in the ocean. This work aims to study these different amounts, comparing such activities. In addition, it proposes to compare the sea dispersion of Fukushima with the amount of radioactive waste dumped in the oceans, when the release of radioactive waste at sea was permitted. It also proposes to compare the Chernobyl aerial dispersion with the radioactive material dissipated in the atmosphere, resulting from the more than 500 atmospheric nuclear tests conducted between 1945 and 1962 by the United States, the former Soviet Union, England, France and China. (author)

Retrieval system for emplaced spent unreprocessed fuel (SURF) in salt bed depository: accident event analysis and mechanical failure probabilities. Final report

International Nuclear Information System (INIS)

Bhaskaran, G.; McCleery, J.E.

1979-10-01

This report provides support in developing an accident prediction event tree diagram, with an analysis of the baseline design concept for the retrieval of emplaced spent unreprocessed fuel (SURF) contained in a degraded Canister. The report contains an evaluation check list, accident logic diagrams, accident event tables, fault trees/event trees and discussions of failure probabilities for the following subsystems as potential contributors to a failure: (a) Canister extraction, including the core and ram units; (b) Canister transfer at the hoist area; and (c) Canister hoisting. This report is the second volume of a series. It continues and expands upon the report Retrieval System for Emplaced Spent Unreprocessed Fuel (SURF) in Salt Bed Depository: Baseline Concept Criteria Specifications and Mechanical Failure Probabilities. This report draws upon the baseline conceptual specifications contained in the first report

Accident sequence quantification with KIRAP

Energy Technology Data Exchange (ETDEWEB)

Kim, Tae Un; Han, Sang Hoon; Kim, Kil You; Yang, Jun Eon; Jeong, Won Dae; Chang, Seung Cheol; Sung, Tae Yong; Kang, Dae Il; Park, Jin Hee; Lee, Yoon Hwan; Hwang, Mi Jeong

1997-01-01

The tasks of probabilistic safety assessment(PSA) consists of the identification of initiating events, the construction of event tree for each initiating event, construction of fault trees for event tree logics, the analysis of reliability data and finally the accident sequence quantification. In the PSA, the accident sequence quantification is to calculate the core damage frequency, importance analysis and uncertainty analysis. Accident sequence quantification requires to understand the whole model of the PSA because it has to combine all event tree and fault tree models, and requires the excellent computer code because it takes long computation time. Advanced Research Group of Korea Atomic Energy Research Institute(KAERI) has developed PSA workstation KIRAP(Korea Integrated Reliability Analysis Code Package) for the PSA work. This report describes the procedures to perform accident sequence quantification, the method to use KIRAP`s cut set generator, and method to perform the accident sequence quantification with KIRAP. (author). 6 refs.

Accident sequence quantification with KIRAP

International Nuclear Information System (INIS)

Kim, Tae Un; Han, Sang Hoon; Kim, Kil You; Yang, Jun Eon; Jeong, Won Dae; Chang, Seung Cheol; Sung, Tae Yong; Kang, Dae Il; Park, Jin Hee; Lee, Yoon Hwan; Hwang, Mi Jeong.

1997-01-01

The tasks of probabilistic safety assessment(PSA) consists of the identification of initiating events, the construction of event tree for each initiating event, construction of fault trees for event tree logics, the analysis of reliability data and finally the accident sequence quantification. In the PSA, the accident sequence quantification is to calculate the core damage frequency, importance analysis and uncertainty analysis. Accident sequence quantification requires to understand the whole model of the PSA because it has to combine all event tree and fault tree models, and requires the excellent computer code because it takes long computation time. Advanced Research Group of Korea Atomic Energy Research Institute(KAERI) has developed PSA workstation KIRAP(Korea Integrated Reliability Analysis Code Package) for the PSA work. This report describes the procedures to perform accident sequence quantification, the method to use KIRAP's cut set generator, and method to perform the accident sequence quantification with KIRAP. (author). 6 refs

Outline of the Desktop Severe Accident Graphic Simulator Module for OPR-1000

International Nuclear Information System (INIS)

Park, S. Y.; Ahn, K. I.

2015-01-01

This paper introduce the desktop severe accident graphic simulator module (VMAAP) which is a window-based severe accident simulator using MAAP as its engine. The VMAAP is one of the submodules in SAMEX system (Severe Accident Management Support Expert System) which is a decision support system for use in a severe accident management following an incident at a nuclear power plant. The SAMEX system consists of four major modules as sub-systems: (a) Severe accident risk data base module (SARDB): stores the data of integrated severe accident analysis code results like MAAP and MELCOR for hundreds of high frequency scenarios for the reference plant; (b) Risk-informed severe accident risk data base management module (RI-SARD): provides a platform to identify the initiating event, determine plant status and equipment availability, diagnoses the status of the reactor core, reactor vessel and containment building, and predicts the plant behaviors; (c) Severe accident management simulator module (VMAAP): runs the MAAP4 code with user friendly graphic interface for input deck and output display; (d) On-line severe accident management guidance module (On-line SAMG); provides available accident management strategies with an electronic format. The role of VMAAP in SAMEX can be described as followings. SARDB contains the most of high frequency scenarios based on a level 2 probabilistic safety analysis. Therefore, there is good chance that a real accident sequence is similar to one of the data base cases. In such a case, RI-SARD can predict an accident progression by a scenario-base or symptom-base search depends on the available plant parameter information. Nevertheless, there still may be deviations or variations between the actual scenario and the data base scenario. The deviations can be decreased by using a real-time graphic accident simulator, VMAAP.. VMAAP is a MAAP4-based severe accident simulation model for OPR-1000 plant. It can simulate spectrum of physical processes

Outline of the Desktop Severe Accident Graphic Simulator Module for OPR-1000

Energy Technology Data Exchange (ETDEWEB)

Park, S. Y.; Ahn, K. I. [KAERI, Daejeon (Korea, Republic of)

2015-05-15

This paper introduce the desktop severe accident graphic simulator module (VMAAP) which is a window-based severe accident simulator using MAAP as its engine. The VMAAP is one of the submodules in SAMEX system (Severe Accident Management Support Expert System) which is a decision support system for use in a severe accident management following an incident at a nuclear power plant. The SAMEX system consists of four major modules as sub-systems: (a) Severe accident risk data base module (SARDB): stores the data of integrated severe accident analysis code results like MAAP and MELCOR for hundreds of high frequency scenarios for the reference plant; (b) Risk-informed severe accident risk data base management module (RI-SARD): provides a platform to identify the initiating event, determine plant status and equipment availability, diagnoses the status of the reactor core, reactor vessel and containment building, and predicts the plant behaviors; (c) Severe accident management simulator module (VMAAP): runs the MAAP4 code with user friendly graphic interface for input deck and output display; (d) On-line severe accident management guidance module (On-line SAMG); provides available accident management strategies with an electronic format. The role of VMAAP in SAMEX can be described as followings. SARDB contains the most of high frequency scenarios based on a level 2 probabilistic safety analysis. Therefore, there is good chance that a real accident sequence is similar to one of the data base cases. In such a case, RI-SARD can predict an accident progression by a scenario-base or symptom-base search depends on the available plant parameter information. Nevertheless, there still may be deviations or variations between the actual scenario and the data base scenario. The deviations can be decreased by using a real-time graphic accident simulator, VMAAP.. VMAAP is a MAAP4-based severe accident simulation model for OPR-1000 plant. It can simulate spectrum of physical processes

Protection of the Population in the event of a Nuclear accident. A Basis for Intervention

International Nuclear Information System (INIS)

1990-01-01

During the years following the Chernobyl accident in 1986, the NEA actively participated in the international effort towards the improvement and better harmonization of the international and national criteria for the protection of the public in the event of a nuclear accident. A first report on this matter, titled Nuclear Accidents: Intervention Levels for Protection of the Public was published by the NEA in 1989. Subsequently, the NEA Committee on Radiation Protection and Public Health set up a small Task Group to provide additional guidance, and to take into account recent developments in other international organizations. The report outlines the status of relevant international activities in the period following the preparation of the 1989 report, discusses the intervention principles and describes both the proposed accident management system and a general scheme for its application. It is to be noted that the principles and criteria for intervention discussed in this report, although developed with specific reference to reactor accidents, apply equally well to activities and possible accidents at other nuclear facilities. The report briefly describes the transition from an accident management situation back to a normal situation and the related problem of changing criteria for the protection of the public. In addition to the traditional exposure pathways -inhalation from the cloud, external irradiation from the cloud and the ground and ingestion of food - the report acknowledges the existence of special pathways, proposing criteria for protecting workers and the public and some examples of their application

Ultimate Electrical Means for Severe Accident and Multi Unit Event Management

International Nuclear Information System (INIS)

Guisez, Xavier

2015-01-01

Following the Multi Unit Severe Accident that occurred at Fukushima as a result of the tsunami on 11 March 2011, the European Council decided to submit its Nuclear Power Plants to a Stress Test. In Belgium, this Stress Test, named BEST (Belgian Stress Test), was successfully concluded at the end of 2011. Nevertheless, Electrabel decided, in agreement with the Authorities, to start a beyond design basis action plan, with the goal to mitigate the consequences of a Beyond Design Basis Accident and a Multi Unit Event. Consequently, this has led to an improvement of the robustness of its Nuclear Power Plants. Considering the importance of electrical power supply to a nuclear power plant, a significant part of this action plan consisted of setting up a mobile, 'plug and play' method for the electrical power supply to some major safety systems. In order to install this ultimate power supply, three factors were retained as essential. First, important reactor monitoring instrumentation is preserved. Second, core cooling is provided at all times. Finally, systems are easily made operational within a very short delay of time. During normal operation and Design Basis Events, core cooling is provided by High Voltage equipment. However, during high stress circumstances, it is too complex to realize connections on this equipment. Therefore, analysis was performed to realize core cooling with, easier to handle, Low Voltage equipment. These systems are powered by several GenSets, especially designed and manufactured for this application. The outcome of this project are easy to use, ultimate means, that supply electric power to important safety systems in order to drastically reduce the risk of core damage, during a beyond design basis event. Additionally, for all ultimate means, procedures and training modules were developed for the operators. (authors)

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

International Nuclear Information System (INIS)

Vrbanic, I.; Spiler, J.; Mikulicic, V.; Simic, Z.

2002-01-01

Treatment of events that represent systems' successes in accident sequences is well known issue associated primarily with those PSA models that employ event tree / fault tree (ET / FT) linking technique. Even theoretically clear, practical implementation and usage creates for certain PSA models a number of difficulties regarding result correctness. Strict treatment of success-events would require consistent applying of de Morgan laws. However, there are several problems related to it. First, Boolean resolution of the overall model, such as the one representing occurrence of reactor core damage, becomes very challenging task if De Morgan rules are applied consistently at all levels. Even PSA tools of the newest generation have some problems with performing such a task in a reasonable time frame. The second potential issue is related to the presence of negated basic events in minimal cutsets. If all the basic events that result from strict applying of De Morgan rules are retained in presentation of minimal cutsets, their readability and interpretability may be impaired severely. It is also worth noting that the concept of a minimal cutset is tied to equipment failures, rather than to successes. For reasons like these, various simplifications are employed in PSA models and tools, when it comes to the treatment of success-events in the sequences. This paper provides a discussion of major concerns associated with the treatment of success-events in accident sequences of a typical PSA model. (author)

Simulation of severe accidents in COTELS experiments

International Nuclear Information System (INIS)

Vasilev, Yu.S.; Zhdanov, V.S.; Kolodeshnikov, A.A.; Kadyrov, Kh. G.; Turkebaev, T.E.; Tsaj, K.V.; Suslov, E.E.

1999-01-01

At present, the issue of atomic reactor operation safety is of a great attention. It is evident that the accident accompanied with a core materials melting is an improbable event. To fully assess a hazard of a reactor use and enhance its safety, it is necessary to predict a possible accident progress and specify possible consequences of severe accidents and eliminating measures. In COTELS experiments, aimed at investigation of interaction of corium with concrete and water, the corium s imulator m elt is discharged on the concrete. The concrete erosion parameters, composition and rate of aerosol and gas escaping are recorded. The solidified melt and concrete fragments structure is studied after the testing, using the X-ray diffractometer DRON-3. This paper gives consideration to possible mechanisms of formation of uranium-containing and other phases of products of interaction of the corium melt with concrete and water

Indemnification of Damage in the Event of a Nuclear Accident

International Nuclear Information System (INIS)

2006-01-01

The Second International Workshop on the Indemnification of Nuclear Damage was held in Bratislava, Slovak Republic, from 18 to 20 May 2005. The workshop was co-organised by the OECD Nuclear Energy Agency and the Nuclear Regulatory Authority of the Slovak Republic. It attracted wide participation from national nuclear authorities, regulators, operators of nuclear installations, nuclear insurers and international organisations. The purpose of the workshop was to assess the third party liability and compensation mechanisms that would be implemented by participating countries in the event of a nuclear accident taking place within or near their borders. To accommodate this objective, two fictitious accident scenarios were developed: one involving a fire in a nuclear installation located in the Slovak Republic and resulting in the release of significant amounts of radioactive materials off-site, and the other a fire on board a ship transporting enriched uranium hexafluoride along the Danube River. The first scenario was designed to involve the greatest possible number of countries, with the second being restricted to countries with a geographical proximity to the Danube. These proceedings contain the papers presented at the workshop, as well as reports on the discussion sessions held. (author)

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

International Nuclear Information System (INIS)

Leonard, M.T.; Ashbaugh, S.G.; Cole, R.K.; Bergeron, K.D.; Nagashima, K.

1996-01-01

This paper presents a comparison of calculations of severe accident progression for several postulated accident sequences for representative Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) nuclear power plants performed with the MELCOR 1.8.3 and the MAAP4 computer codes. The PWR system examined in this study is a 1100 MWe system similar in design to a Westinghouse 3-loop plant with a large dry containment; the BWR is a 1100 MWe system similar in design to General Electric BWR/4 with a Mark I containment. A total of nine accident sequences were studied with both codes. Results of these calculations are compared to identify major differences in the timing of key events in the calculated accident progression or other important aspects of severe accident behavior, and to identify specific sources of the observed differences

Review of progress on enhanced accident tolerant fuel

International Nuclear Information System (INIS)

McCoy, K.; Dunn, B.; Kochendarfer, R.

2015-01-01

The accident at Fukushima has resulted in renewed interest in understanding the performance of nuclear power plants under accident conditions. Part of that interest is directed toward determining how to improve the performance of fuel during an accident that involves long exposures of the fuel to high temperatures. This paper describes the method being used by AREVA to select and evaluate approaches for improving the accident tolerance of nuclear fuel. The method involves starting with a large number of approaches that might enhance accident tolerance, and reviewing how well each approach satisfies a set of engineering requirements and goals. Among the approaches investigated we have the development of fuel pellets that contain a second phase to improve thermal conductivity, the use of molybdenum alloy tubing as fuel cladding, the use of oxidation-resistant coatings to zirconium cladding, and the use of nanoparticles in the coolant to improve heat transfer

International exchange of radiological information in the event of a nuclear accident - future perspectives

International Nuclear Information System (INIS)

De-Cort, M.; De-Vries, G.; Breitenbach, L.; Leeb, H.; Weiss, W.

1996-01-01

Immediately after the Chernobyl accident most European countries established or enhanced their national radioactivity monitoring and information systems. The large transboundary effect of the radioactive release also triggered the need for bilateral and international agreements on the exchange of radiological information in case of a nuclear accident. Based on the experiences gained from existing bi- and multilateral data exchange the Commission of the European Communities has made provision for and is developing technical systems to exchange information of common interest. Firstly the existing national systems and systems based on bilateral agreements are summarized. The objectives and technical realizations of the EC international information exchange systems ECURIE and EURDEP, are described. The experiences gained over the past few years and the concepts for the future, in which central and eastern European countries will be included, are discussed. The benefits that would result from improving the international exchange of radiological information in the event of a future nuclear accident are further being described

A methodology for analysing human errors of commission in accident scenarios for risk assessment

International Nuclear Information System (INIS)

Kim, J. H.; Jung, W. D.; Park, J. K

2003-01-01

As the concern on the impact of the operator's inappropriate interventions, so-called Errors Of Commissions(EOCs), on the plant safety has been raised, the interest in the identification and analysis of EOC events from the risk assessment perspective becomes increasing accordingly. To this purpose, we propose a new methodology for identifying and analysing human errors of commission that might be caused from the failures in situation assessment and decision making during accident progressions given an initiating event. The proposed methodology was applied to the accident scenarios of YGN 3 and 4 NPPs, which resulted in about 10 EOC situations that need careful attention

Initiating events of accidents in the practice of oil well logging in Cuba

International Nuclear Information System (INIS)

Alles Leal, A.; Perez Reyes, Y.; Dumenigo Gonzalez, C.

2013-01-01

The oil well logging is an extremely important activity within the oil industry, but in turn, brings risks that occasionally result in damage to health, the environment and economic losses. In this context, risk analysis has become an important tool to control them through their prediction and the study of the factors that determine them, enabling substantiated decisions to, first, foresee accidents and, secondly, to minimize their consequences. This paper proposes the elaboration of a list of initiating events of accidents in the practice of oil well logging which is one of the most important aspects for further evaluation of radiation safety of this practice. For its determination the technique employed to identify risks was 'Failure Modes and Effects Analysis (FMEA)' by applying it to the different stages and processes of practice. (Author)

Validation of the metal fuel version of the SAS4A accident analysis code

International Nuclear Information System (INIS)

Tentner, A.M.

1991-01-01

This paper describes recent work directed towards the validation of the metal fuel version of the SAS4A accident analysis code. The SAS4A code system has been developed at Argonne National Laboratory for the simulation of hypothetical severe accidents in Liquid Metal-Cooled Reactors (LMR), designed to operate in a fast neutron spectrum. SAS4A was initially developed for the analysis of oxide-fueled liquid metal-cooled reactors and has played an important role in the simulation and assessment of the energetics potential for postulated severe accidents in these reactors. Due to the current interest in the metal-fueled liquid metal-cooled reactors, a metal fuel version of the SAS4A accident analysis code is being developed in the Integral Fast Reactor program at Argonne. During such postulated accident scenarios as the unprotected (i.e. without scram) loss-of-flow and transient overpower events, a large number of interrelated physical phenomena occur during a relatively short time. These phenomena include transient heat transfer and hydrodynamic events, coolant boiling, and fuel and cladding melting and relocation. Due to strong neutronic feedbacks these events can significantly influence the reactor power history in the accident progression. The paper presents the results of a recent SAS4A simulation of the M7 TREAT experiment. 6 refs., 5 figs

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.)

Severe Accident Research Program plan update

International Nuclear Information System (INIS)

1992-12-01

In August 1989, the staff published NUREG-1365, ''Revised Severe Accident Research Program Plan.'' Since 1989, significant progress has been made in severe accident research to warrant an update to NUREG-1365. The staff has prepared this SARP Plan Update to: (1) Identify those issues that have been closed or are near completion, (2) Describe the progress in our understanding of important severe accident phenomena, (3) Define the long-term research that is directed at improving our understanding of severe accident phenomena and developing improved methods for assessing core melt progression, direct containment heating, and fuel-coolant interactions, and (4) Reflect the growing emphasis in two additional areas--advanced light water reactors, and support for the assessment of criteria for containment performance during severe accidents. The report describes recent major accomplishments in understanding the underlying phenomena that can occur during a severe accident. These include Mark I liner failure, severe accident scaling methodology, source term issues, core-concrete interactions, hydrogen transport and combustion, TMI-2 Vessel Investigation Project, and direct containment heating. The report also describes the major planned activities under the SARP over the next several years. These activities will focus on two phenomenological issues (core melt progression, and fuel-coolant interactions and debris coolability) that have significant uncertainties that impact our understanding and ability to predict severe accident phenomena and their effect on containment performance SARP will also focus on severe accident code development, assessment and validation. As the staff completes the research on severe accident issues that relate to current generation reactors, continued research will focus on efforts to independently evaluate the capability of new advanced light water reactor designs to withstand severe accidents

Comparison of event tree, fault tree and Markov methods for probabilistic safety assessment and application to accident mitigation

International Nuclear Information System (INIS)

James, H.; Harris, M.J.; Hall, S.F.

1992-01-01

Probabilistic safety assessment (PSA) is used extensively in the nuclear industry. The main stages of PSA and the traditional event tree method are described. Focussing on hydrogen explosions, an event tree model is compared to a novel Markov model and a fault tree, and unexpected implication for accident mitigation is revealed. (author)

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

One-year progression of moderate asymptomatic carotid stenosis predicts the risk of vascular events.

Science.gov (United States)

Balestrini, Simona; Lupidi, Francesca; Balucani, Clotilde; Altamura, Claudia; Vernieri, Fabrizio; Provinciali, Leandro; Silvestrini, Mauro

2013-03-01

This study aimed at evaluating whether ultrasound monitoring of moderate asymptomatic carotid stenosis may help in identifying subjects at high risk for vascular events. We included 523 subjects with unilateral asymptomatic carotid stenosis of 50% to 69%. Follow-up carotid ultrasound was performed within 12 months from inclusion to detect the frequency and degree of stenosis progression. Subjects were prospectively evaluated for a median period of 42 months (interquartile range, 38-45) after a second ultrasound evaluation. Outcome measures were any stroke and transient ischemic attack, myocardial infarction, and death. Carotid stenosis progression was associated with the occurrence of vascular events (hazard ratio, 21.57; 95% confidence interval, 11.81-39.39; P<0.001). During follow-up, 96.7% of subjects without progressive carotid stenosis remained free from vascular events. Among patients with progressive stenosis, 53.7% experienced a vascular event and 27.1% experienced an ipsilateral stroke. One-year moderate asymptomatic carotid stenosis progression is related to higher risk of vascular events, including ipsilateral stroke.

The handling of radiation accidents

International Nuclear Information System (INIS)

1977-01-01

The symposium was attended by 204 participants from 39 countries and 5 international organizations. Forty-two papers were presented in 8 sessions. The purpose of the meeting was to foster an exchange of experiences gained in establishing and exercising plans for mitigating the effects of radiation accidents and in the handling of actual accident situations. Only a small number of accidents were reported at the symposium, and this reflects the very high standards of safety that has been achieved by the nuclear industry. No accidents of radiological significance were reported to have occurred at commercial nuclear power plants. Of the accidents reported, industrial radiography continues to be the area in which most of the radiation accidents occur. The experience gained in the reported accident situations served to confirm the crucial importance of the prompt availability of medical and radiological services, particularly in the case of uptake of radioactive material, and emphasized the importance of detailed investigation into the causes of the accident in order to improve preventative measures. One of the principal themes of the symposium involved emergency procedures related to nuclear power plant accidents, and several papers defining the scope, progression and consequences of design base accidents for both thermal and fast reactor systems were presented. These were complemented by papers defining the resultant protection requirements that should be satisfied in the establishment of plans designed to mitigate the effects of the postulated accident situations. Several papers were presented describing existing emergency organizational arrangements relating both to specific nuclear power plants and to comprehensive national schemes, and a particularly informative session was devoted to the topic of training of personnel in the practical conduct of emergency arrangements. The general feeling of the participants was one of studied confidence in the competence and

Radioactive Reversal? The Fukushima Accident as a Focusing Event for Comparative Policy Change on Nuclear Energy

Science.gov (United States)

Sanchez, Victoria Justine

This dissertation project examines the 2011 Fukushima nuclear accident as a focusing event for policy change on nuclear energy. For example, following the accident, Germany (and much of Europe) experienced a reversal of policy on nuclear energy. Conversely, many others such as China, Russia, and France, did not exhibit such a retraction against nuclear power, albeit with public debate about the risks and consequences of accidents. Why has there been dramatic policy change in some cases but not others? The political and literal fallout of Fukushima has provoked a wave of policy change towards nuclear energy at the national level. Through qualitative and quantitative measures, we can view Fukushima as an impetus for comparing the dynamics of nuclear policy change. Quantitatively, this project employs logistic regression to explore variables such as regime type, energy security, trade supply and demand, climate change concerns, and public acceptance are related to policy outcomes and change on nuclear energy in the post-Fukushima context of 49 different countries. Qualitatively, country cases (Russia, Germany, and Canada) are assessed into three categories based on the outcome of policy decisions on nuclear energy following Fukushima for a richer analysis. Beyond the Fukushima example, we can hope to better understand how political focusing events can gain influence in an international context.

LOSP-initiated event tree analysis for BWR

International Nuclear Information System (INIS)

Watanabe, Norio; Kondo, Masaaki; Uno, Kiyotaka; Chigusa, Takeshi; Harami, Taikan

1989-03-01

As a preliminary study of 'Japanese Model Plant PSA', a LOSP (loss of off-site power)-initiated Event Tree Analysis for a Japanese typical BWR was carried out solely based on the open documents such as 'Safety Analysis Report'. The objectives of this analysis are as follows; - to delineate core-melt accident sequences initiated by LOSP, - to evaluate the importance of core-melt accident sequences in terms of occurrence frequency, and - to develop a foundation of plant information and analytical procedures for efficiently performing further 'Japanese Model Plant PSA'. This report describes the procedure and results of the LOSP-initiated Event Tree Analysis. In this analysis, two types of event trees, Functional Event Tree and Systemic Event Tree, were developed to delineate core-melt accident sequences and to quantify their frequencies. Front-line System Event Tree was prepared as well to provide core-melt sequence delineation for accident progression analysis of Level 2 PSA which will be followed in a future. Applying U.S. operational experience data such as component failure rates and a LOSP frequency, we obtained the following results; - The total frequency of core-melt accident sequences initiated by LOSP is estimated at 5 x 10 -4 per reactor-year. - The dominant sequences are 'Loss of Decay Heat Removal' and 'Loss of Emergency Electric Power Supply', which account for more than 90% of the total core-melt frequency. In this analysis, a higher value of 0.13/R·Y was used for the LOSP frequency than experiences in Japan and any recovery action was not considered. In fact, however, there has been no experience of LOSP event in Japanese nuclear power plants so far and it is also expected that offsite power and/or PCS would be recovered before core melt. Considering Japanese operating experience and recovery factors will reduce the total core-melt frequency to less than 10 -6 per reactor-year. (J.P.N.)

Introduction of operator actions in the event trees

International Nuclear Information System (INIS)

Bars, G.; Lanore, J.M.; Villeroux, C.

1984-11-01

In the PRA in progress in France for a 900 MW PWR plant, an effort is done for introducing operator actions during accident sequences. A first approach of this complex problem relies on an extensive use of existing methods an knowledge in diverse fields. Identification of actions is based on the operating procedures, and in particular on the existence of special emergency procedures which define the optimal actions during severe accidents. This approach implies the introduction in the event trees of the notion of procedure failure. Quantification of the corresponding probabilities leads to several problems including physics of the sequences, systems availability and human behaviour for decision making and actions. This treatment is illustrated by the example of the small break event tree

Accident management

International Nuclear Information System (INIS)

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

1989-01-01

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

Risk analysis of releases from accidents during mid-loop operation at Surry

International Nuclear Information System (INIS)

Jo, J.; Lin, C.C.; Nimnual, S.; Mubayi, V.; Neymotin, L.

1992-11-01

Studies and operating experience suggest that the risk of severe accidents during low power operation and/or shutdown (LP/S) conditions could be a significant fraction of the risk at full power operation. Two studies have begun at the Nuclear Regulatory Commission (NRC) to evaluate the severe accident progression from a risk perspective during these conditions: One at the Brookhaven National Laboratory for the Surry plant, a pressurized water reactor (PWR), and the other at the Sandia National Laboratories for the Grand Gulf plant, a boiling water reactor (BWR). Each of the studies consists of three linked, but distinct, components: a Level I probabilistic risk analysis (PRA) of the initiating events, systems analysis, and accident sequences leading to core damage; a Level 2/3 analysis of accident progression, fuel damage, releases, containment performance, source term and consequences-off-site and on-site; and a detailed Human Reliability Analysis (HRA) of actions relevant to plant conditions during LP/S operations. This paper summarizes the approach taken for the Level 2/3 analysis at Surry and provides preliminary results on the risk of releases and consequences for one plant operating state, mid-loop operation, during shutdown

Evaluation of severe accident risks, Grand Gulf, Unit 1: Appendices

International Nuclear Information System (INIS)

Brown, T.D.; Breeding, R.J.; Jow, H.N.; Higgins, S.J.; Shiver, A.W.; Helton, J.C.; Amos, C.N.

1990-12-01

In support of the Nuclear Regulatory Commission's (NRC's) assessment of the risk from severe accidents at commercial nuclear power plants in the US report in NUREG-1150, the Severe Accident Risk Reduction Program (SARRP) has completed a revised calculation of the risk to the general public from severe accidents at the Grand Gulf Nuclear Station, Unit 1. This power plant, located in Port Gibson, Mississippi, is operated by the System Energy Resources, Inc. (SERI). The emphasis in this risk analysis was not on determining a ''so-called'' point estimate of risk. Rather, it was to determine the distribution of risk, and to discover the uncertainties that account for the breadth of this distribution. Off-site risk initiated by events internal to the power plant was assessed. This document provides Appendices A through E for this report. Topics included are, respectively: supporting information for the accident progression analysis; supporting information for the source term analysis; supporting information for the consequence analysis; risk results; and sampling information

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)

Mitigation of Severe Accident Consequences Using Inherent Safety Principles

International Nuclear Information System (INIS)

Wigeland, R.A.; Cahalan, J.E.

2009-01-01

Sodium-cooled fast reactors are designed to have a high level of safety. Events of high probability of occurrence are typically handled without consequence through reliable engineering systems and good design practices. For accidents of lower probability, the initiating events are characterized by larger and more numerous challenges to the reactor system, such as failure of one or more major engineered systems and can also include a failure to scram the reactor in response. As the initiating conditions become more severe, they have the potential for creating serious consequences of potential safety significance, including fuel melting, fuel pin disruption and recriticality. If the progression of such accidents is not mitigated by design features of the reactor, energetic events and dispersal of radioactive materials may result. For severe accidents, there are several approaches that can be used to mitigate the consequences of such severe accident initiators, which typically include fuel pin failures and core disruption. One approach is to increase the reliability of the reactor protection system so that the probability of an ATWS event is reduced to less than 1 x 10-6 per reactor year, where larger accident consequences are allowed, meeting the U.S. NRC goal of relegating such accident consequences as core disruption to these extremely low probabilities. The main difficulty with this approach is to convincingly test and guarantee such increased reliability. Another approach is to increase the redundancy of the reactor scram system, which can also reduce the probability of an ATWS event to a frequency of less than 1 x 10-6 per reactor year or lower. The issues with this approach are more related to reactor core design, with the need for a greater number of control rod positions in the reactor core and the associated increase in complexity of the reactor protection system. A third approach is to use the inherent reactivity feedback that occurs in a fast reactor to

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

Recent Developments in Level 2 PSA and Severe Accident Management

International Nuclear Information System (INIS)

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

2007-01-01

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

Event course analysis of core disruptive accidents

International Nuclear Information System (INIS)

Hering, W.; Homann, C.; Sengpiel, W.; Struwe, D.; Messainguiral, C.

1995-01-01

The theortical studies of the behavior of a PWR core in a meltdown accident are focused on hydrogen release, materials redistribution in the core area including forming of an oxide melt pool, quantity of melt and its composition, and temperatures attained by the RPV internals (esp. in the upper plenum) during the accident up to the time of melt relocation into the lower plenum. The calculations are done by the SCDAP/RELAP5 code. For its validation selected CORA results and Phebus FPTO results have been used. (orig.)

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

International Nuclear Information System (INIS)

Choi, Young; Park, Soo Yong; Ahn, Kwang-Il; Kim, D.H.

2006-01-01

Although a Level 2 PSA was performed for the Korean Standard Power Plants (KSNPs), and it considered the necessary sequences for an assessment of the containment integrity and source term analysis. In terms of an accident management, however, more cases causing severe core damage need to be analyzed and arranged systematically for an easy access to the results. At present, KAERI is calculating the severe accident sequences intensively for various initiating events and generating a database for the accident progression including thermal hydraulic and source term behaviours. The developed Database (DB) system includes a graphical display for a plant and equipment status, previous research results by knowledge-base technique, and the expected plant behaviour. The plant model used in this paper is oriented to the case of LOCAs related severe accident phenomena and thus can simulate the plant behaviours for a severe accident. Therefore the developed system may play a central role as an information source for decision-making for a severe accident management, and will be used as a training simulator for a severe accident management. (author)

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

Historical aspects of radiation accidents

International Nuclear Information System (INIS)

Mettler, F.A. Jr.; Ricks, R.C.

1990-01-01

Radiation accidents are extremely rare events; however, the last two years have witnessed the largest radiation accidents in both the eastern and western hemispheres. It is the purpose of this chapter to review how radiation accidents are categorized, examine the temporal changes in frequency and severity, give illustrative examples of several types of radiation accidents, and finally, to describe the various registries for radiation accidents

Major Accidents (Gray Swans) Likelihood Modeling Using Accident Precursors and Approximate Reasoning.

Science.gov (United States)

Khakzad, Nima; Khan, Faisal; Amyotte, Paul

2015-07-01

Compared to the remarkable progress in risk analysis of normal accidents, the risk analysis of major accidents has not been so well-established, partly due to the complexity of such accidents and partly due to low probabilities involved. The issue of low probabilities normally arises from the scarcity of major accidents' relevant data since such accidents are few and far between. In this work, knowing that major accidents are frequently preceded by accident precursors, a novel precursor-based methodology has been developed for likelihood modeling of major accidents in critical infrastructures based on a unique combination of accident precursor data, information theory, and approximate reasoning. For this purpose, we have introduced an innovative application of information analysis to identify the most informative near accident of a major accident. The observed data of the near accident were then used to establish predictive scenarios to foresee the occurrence of the major accident. We verified the methodology using offshore blowouts in the Gulf of Mexico, and then demonstrated its application to dam breaches in the United Sates. © 2015 Society for Risk Analysis.

Technical basis document for external events

International Nuclear Information System (INIS)

OBERG, B.D.

2003-01-01

This document supports the Tank Farms Documented Safety Analysis and presents the technical basis for the FR-equencies of externally initiated accidents. The consequences of externally initiated events are discussed in other documents that correspond to the accident that was caused by the external event. The external events include aircraft crash, vehicle accident, range fire, and rail accident

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

International Nuclear Information System (INIS)

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

2001-01-01

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

The Chernobyl accident consequences; Consequences de l'accident de Tchernobyl

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-04-01

Five teen years later, Tchernobyl remains the symbol of the greater industrial nuclear accident. To take stock on this accident, this paper proposes a chronology of the events and presents the opinion of many international and national organizations. It provides also web sites references concerning the environmental and sanitary consequences of the Tchernobyl accident, the economic actions and propositions for the nuclear safety improvement in the East Europe. (A.L.B.)

Expert software for accident identification

International Nuclear Information System (INIS)

Dobnikar, M.; Nemec, T.; Muehleisen, A.

2003-01-01

Each type of an accident in a Nuclear Power Plant (NPP) causes immediately after the start of the accident variations of physical parameters that are typical for that type of the accident thus enabling its identification. Examples of these parameter are: decrease of reactor coolant system pressure, increase of radiation level in the containment, increase of pressure in the containment. An expert software enabling a fast preliminary identification of the type of the accident in Krsko NPP has been developed. As input data selected typical parameters from Emergency Response Data System (ERDS) of the Krsko NPP are used. Based on these parameters the expert software identifies the type of the accident and also provides the user with appropriate references (past analyses and other documentation of such an accident). The expert software is to be used as a support tool by an expert team that forms in case of an emergency at Slovenian Nuclear Safety Administration (SNSA) with the task to determine the cause of the accident, its most probable scenario and the source term. The expert software should provide initial identification of the event, while the final one is still to be made after appropriate assessment of the event by the expert group considering possibility of non-typical events, multiple causes, initial conditions, influences of operators' actions etc. The expert software can be also used as an educational/training tool and even as a simple database of available accident analyses. (author)

Monitoring severe accidents using AI techniques

Energy Technology Data Exchange (ETDEWEB)

No, Young Gyu; Ahn, Kwang Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Ju Hyun; Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of); Lim, Dong Hyuk [Korea Institute of Nuclear Nonproliferation and Control, Daejon (Korea, Republic of)

2012-05-15

After the Fukushima nuclear accident in 2011, there has been increasing concern regarding severe accidents in nuclear facilities. Severe accident scenarios are difficult for operators to monitor and identify. Therefore, accurate prediction of a severe accident is important in order to manage it appropriately in the unfavorable conditions. In this study, artificial intelligence (AI) techniques, such as support vector classification (SVC), probabilistic neural network (PNN), group method of data handling (GMDH), and fuzzy neural network (FNN), were used to monitor the major transient scenarios of a severe accident caused by three different initiating events, the hot-leg loss of coolant accident (LOCA), the cold-leg LOCA, and the steam generator tube rupture in pressurized water reactors (PWRs). The SVC and PNN models were used for the event classification. The GMDH and FNN models were employed to accurately predict the important timing representing severe accident scenarios. In addition, in order to verify the proposed algorithm, data from a number of numerical simulations were required in order to train the AI techniques due to the shortage of real LOCA data. The data was acquired by performing simulations using the MAAP4 code. The prediction accuracy of the three types of initiating events was sufficiently high to predict severe accident scenarios. Therefore, the AI techniques can be applied successfully in the identification and monitoring of severe accident scenarios in real PWRs.

Monitoring severe accidents using AI techniques

International Nuclear Information System (INIS)

No, Young Gyu; Ahn, Kwang Il; Kim, Ju Hyun; Na, Man Gyun; Lim, Dong Hyuk

2012-01-01

After the Fukushima nuclear accident in 2011, there has been increasing concern regarding severe accidents in nuclear facilities. Severe accident scenarios are difficult for operators to monitor and identify. Therefore, accurate prediction of a severe accident is important in order to manage it appropriately in the unfavorable conditions. In this study, artificial intelligence (AI) techniques, such as support vector classification (SVC), probabilistic neural network (PNN), group method of data handling (GMDH), and fuzzy neural network (FNN), were used to monitor the major transient scenarios of a severe accident caused by three different initiating events, the hot-leg loss of coolant accident (LOCA), the cold-leg LOCA, and the steam generator tube rupture in pressurized water reactors (PWRs). The SVC and PNN models were used for the event classification. The GMDH and FNN models were employed to accurately predict the important timing representing severe accident scenarios. In addition, in order to verify the proposed algorithm, data from a number of numerical simulations were required in order to train the AI techniques due to the shortage of real LOCA data. The data was acquired by performing simulations using the MAAP4 code. The prediction accuracy of the three types of initiating events was sufficiently high to predict severe accident scenarios. Therefore, the AI techniques can be applied successfully in the identification and monitoring of severe accident scenarios in real PWRs.

Probabilistic approach in treatment of deterministic analyses results of severe accidents

International Nuclear Information System (INIS)

Krajnc, B.; Mavko, B.

1996-01-01

Severe accidents sequences resulting in loss of the core geometric integrity have been found to have small probability of the occurrence. Because of their potential consequences to public health and safety, an evaluation of the core degradation progression and the resulting effects on the containment is necessary to determine the probability of a significant release of radioactive materials. This requires assessment of many interrelated phenomena including: steel and zircaloy oxidation, steam spikes, in-vessel debris cooling, potential vessel failure mechanisms, release of core material to the containment, containment pressurization from steam generation, or generation of non-condensable gases or hydrogen burn, and ultimately coolability of degraded core material. To asses the answer from the containment event trees in the sense of weather certain phenomenological event would happen or not the plant specific deterministic analyses should be performed. Due to the fact that there is a large uncertainty in the prediction of severe accidents phenomena in Level 2 analyses (containment event trees) the combination of probabilistic and deterministic approach should be used. In fact the result of the deterministic analyses of severe accidents are treated in probabilistic manner due to large uncertainty of results as a consequence of a lack of detailed knowledge. This paper discusses approach used in many IPEs, and which assures that the assigned probability for certain question in the event tree represent the probability that the event will or will not happen and that this probability also includes its uncertainty, which is mainly result of lack of knowledge. (author)

First Responders and Criticality Accidents

Energy Technology Data Exchange (ETDEWEB)

Valerie L. Putman; Douglas M. Minnema

2005-11-01

Nuclear criticality accident descriptions typically include, but do not focus on, information useful to first responders. We studied these accidents, noting characteristics to help (1) first responders prepare for such an event and (2) emergency drill planners develop appropriate simulations for training. We also provide recommendations to help people prepare for such events in the future.

Action to be taken in the event of a radiological accident

International Nuclear Information System (INIS)

Bresson, G.; Nenot, J.C.

1977-01-01

In the event of a radiological accident affecting people, the measures that have to be taken are the responsibility of a large number of persons whose original disciplines differ widely. In the interest of efficiency, it is obviously essential that these measures should be co-ordinated; this implies smoothly functioning liaison between the persons responsible for action at different levels. These levels of action are numerous and differ very considerably; they include, in the first place, the links between the nuclear facility and the medical authorities, either directly within the hospital framework or through the intermediary of an industrial medicine service; then the links between the hospital sector and the large number of experts concerned with the highly specialized aspects of the diagnostic, therapeutic and prognostic problems of irradiation or contamination by radioisotopes; lastly, the links between these various specialists. In view of the wide variety of the parameters involved in accidents, the organization of the action to be taken cannot be encompassed within a rigid framework, especially as it should be possible to apply this organization at both the national and the international level, taking into account the diversity of the medico-legal aspects. The efficiency of the means applied is therefore governed by the flexibility of the procedure; however, the relative scarcity of accidents, i.e. the absence of any involvement of persons and equipment on a true scale, makes it imperative that a high degree of precision be applied in preparing emergency plans, since the omission of one step or one link may have serious or irreparable consequences which cannot always be offset by improvization. The outline of such an operational organization is presented and discussed in the light of past experience. (author)

Do failures in non-technical skills contribute to fatal medical accidents in Japan? A review of the 2010-2013 national accident reports.

Science.gov (United States)

Uramatsu, Masashi; Fujisawa, Yoshikazu; Mizuno, Shinya; Souma, Takahiro; Komatsubara, Akinori; Miki, Tamotsu

2017-02-16

We sought to clarify how large a proportion of fatal medical accidents can be considered to be caused by poor non-technical skills, and to support development of a policy to reduce number of such accidents by making recommendations about possible training requirements. Summaries of reports of fatal medical accidents, published by the Japan Medical Safety Research Organization, were reviewed individually. Three experienced clinicians and one patient safety expert conducted the reviews to determine the cause of death. Views of the patient safety expert were given additional weight in the overall determination. A total of 73 summary reports of fatal medical accidents were reviewed. These reports had been submitted by healthcare organisations across Japan to the Japan Medical Safety Research Organization between April 2010 and March 2013. The cause of death in fatal medical accidents, categorised into technical skills, non-technical skills and inevitable progress of disease were evaluated. Non-technical skills were further subdivided into situation awareness, decision making, communication, team working, leadership, managing stress and coping with fatigue. Overall, the cause of death was identified as non-technical skills in 34 cases (46.6%), disease progression in 33 cases (45.2%) and technical skills in two cases (5.5%). In two cases, no consensual determination could be achieved. Further categorisation of cases of non-technical skills were identified as 14 cases (41.2%) of problems with situation awareness, eight (23.5%) with team working and three (8.8%) with decision making. These three subcategories, or combinations of them, were identified as the cause of death in 33 cases (97.1%). Poor non-technical skills were considered to be a significant cause of adverse events in nearly half of the fatal medical accidents examined. Improving non-technical skills may be effective for reducing accidents, and training in particular subcategories of non-technical skills may be

The Human Aspect of the Fukushima Daiichi Accident

International Nuclear Information System (INIS)

Anegawa, T.; Kawano, A.

2016-01-01

situation where a large number of aftershocks were observed and associated tsunami cautions were announced from time to time. Due to the occurrence of the SBO (Station Black Out), first responders had to engage in field works in the complete darkness while the field were scattered with damaged equipment, vehicles and other debris caused by the tsunami and explosions. Eventual loss of effective communication tools such as paging and PHS also hampered communication between the field, main control rooms and the ERC. In spite of the loss of effective communication tool and other equipment prepared for emergency response, the ERC personnel and shift crew members had to deal with concurrent event progress at six units at the same time; where sometimes the accident progress at one unit (e.g., explosion of the reactor building) also inversely affected the accident response at the adjacent units. Communication within the ERC and between the site and the Headquarters as well as outside the company (e.g., Cabinet, regulatory authority) became more and more complicated and caused further confusion as the progress of accident at 6 units in Fukushima Daiichi and 4 units in Fukushima Daini NPSs. The presentation will describe actions and decisions being taken in such extreme circumstances, to highlight the key lessons learned; such as importance of establishing strong command and control functions, data sharing system etc. Learning from the accident, TEPCO has introduced new command and control system and staff are being trained with the new system. Also, reflecting the lessons from the accident response by shift crew at main control rooms and the field, training program for shift workers and first responders has been revised and more extensive and frequent emergency drills are conducted. In the presentation, such activities currently performed by TEPCO will be addressed. (author)

Severe accidents in nuclear reactors

International Nuclear Information System (INIS)

Ohai, Dumitru; Dumitrescu, Iulia; Tunaru, Mariana

2004-01-01

The likelihood of accidents leading to core meltdown in nuclear reactors is low. The consequences of such an event are but so severe that developing and implementing of adequate measures for preventing or diminishing the consequences of such events are of paramount importance. The analysis of major accidents requires sophisticated computation codes but necessary are also relevant experiments for checking the accuracy of the predictions and capability of these codes. In this paper an overview of the severe accidents worldwide with definitions, computation codes and relating experiments is presented. The experimental research activity of severe accidents was conducted in INR Pitesti since 2003, when the Institute jointed the SARNET Excellence Network. The INR activity within SARNET consists in studying scenarios of severe accidents by means of ASTEC and RELAP/SCDAP codes and conducting bench-scale experiments

A trend analysis of human error events for proactive prevention of accidents. Methodology development and effective utilization

International Nuclear Information System (INIS)

Hirotsu, Yuko; Ebisu, Mitsuhiro; Aikawa, Takeshi; Matsubara, Katsuyuki

2006-01-01

This paper described methods for analyzing human error events that has been accumulated in the individual plant and for utilizing the result to prevent accidents proactively. Firstly, a categorization framework of trigger action and causal factors of human error events were reexamined, and the procedure to analyze human error events was reviewed based on the framework. Secondly, a method for identifying the common characteristics of trigger action data and of causal factor data accumulated by analyzing human error events was clarified. In addition, to utilize the results of trend analysis effectively, methods to develop teaching material for safety education, to develop the checkpoints for the error prevention and to introduce an error management process for strategic error prevention were proposed. (author)

A methodology for the quantitative risk assessment of major accidents triggered by seismic events

International Nuclear Information System (INIS)

Antonioni, Giacomo; Spadoni, Gigliola; Cozzani, Valerio

2007-01-01

A procedure for the quantitative risk assessment of accidents triggered by seismic events in industrial facilities was developed. The starting point of the procedure was the use of available historical data to assess the expected frequencies and the severity of seismic events. Available equipment-dependant failure probability models (vulnerability or fragility curves) were used to assess the damage probability of equipment items due to a seismic event. An analytic procedure was subsequently developed to identify, evaluate the credibility and finally assess the expected consequences of all the possible scenarios that may follow the seismic events. The procedure was implemented in a GIS-based software tool in order to manage the high number of event sequences that are likely to be generated in large industrial facilities. The developed methodology requires a limited amount of additional data with respect to those used in a conventional QRA, and yields with a limited effort a preliminary quantitative assessment of the contribution of the scenarios triggered by earthquakes to the individual and societal risk indexes. The application of the methodology to several case-studies evidenced that the scenarios initiated by seismic events may have a relevant influence on industrial risk, both raising the overall expected frequency of single scenarios and causing specific severe scenarios simultaneously involving several plant units

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit-1: Analysis of core damage frequency from internal events during mid-loop operations. Appendices F-H, Volume 2, Part 4

International Nuclear Information System (INIS)

Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J.; Bley, D.; Johnson, D.; Holmes, B.

1994-06-01

Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analyses that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program includes that of a level-3 PRA. In phase 2, mid-loop operation was selected as the plant configuration to be analyzed based on the results of the phase 1 study. The objective of the phase 2 study is to perform a detailed analysis of the potential accident scenarios that may occur during mid-loop operation, and compare the results with those of NUREG-1150. The scope of the level-1 study includes plant damage state analysis, and uncertainty analysis. Volume 1 summarizes the results of the study. Internal events analysis is documented in Volume 2. It also contains an appendix that documents the part of the phase 1 study that has to do with POSs other than mid-loop operation. Internal fire and internal flood analyses are documented in Volumes 3 and 4. A separate study on seismic analysis, documented in Volume 5, was performed for the NRC by Future Resources Associates, Inc. Volume 6 documents the accident progression, source terms, and consequence analysis

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

Energy Technology Data Exchange (ETDEWEB)

Pistner, Christoph [Oeko-Institut e.V., Darmstadt (Germany). Bereich Nukleartechnik und Anlagensicherheit

2013-07-01

On March 11, 2011 a strong earthquake at the east coast of Japan and a subsequent tsunami caused severe damage at the NPP site of Fukushima Daiichi. The article covers the fundamental safety aspects of the accident progress according to the state of knowledge. The principles of nuclear technology and reactor safety are summarized in order to allow the understanding of the accidental sequence. Even two years after the disaster many questions on the sequence of accident events are still open.

Continued efforts to improve the robustness of the French Gen II PWRs with respect to the risk of severe accidents. Safety assessment and research activities

International Nuclear Information System (INIS)

Raimond, E.; Bonnet, J.M.; Generino, G.; Dubreuil, M.; Pichereau, F.; Van Dorsselaere, J.P.

2012-01-01

In the context of post Fukushima accident, the paper presents the continuous efforts performed in France to upgrade progressively the French Gen II pressurised water reactors safety features in order to face the risks of any severe accident. It reminds some decisions taken after the TMI2 and the Chernobyl accidents and describes the situation in France before the Fukushima accident: -) progress done on severe accident consequences analysis thanks to recent research activities, -) improvement of Gen II PWRs safety features, in relation with the periodic safety review process, -) definition of higher safety levels requirement directly linked to the protection of population in the framework of Gen II PWRs long term operation. The last part of the paper comments carefully how the Fukushima accident will interfere on all these previous efforts to increase the Gen II PWRs robustness. The Fukushima accident clearly highlights a need of additional efforts to identify possible cliff edge effect in case of beyond design events (especially external events). The definition of additional accident management procedures and means to secure a reactor (or a site) whatever the conditions will be a major consequence for the French NPPs. In a second step, some complements on the existing defense-in-depth approach are now expected: additional requirements to define line of defense against adverse consequences of beyond design situations. The need for specific additional research activities after the Fukushima accident seems to be limited to some specific issues (for example spent fuel pool behaviour in case of long term loss of cooling). This paper is followed by the slides of the presentation

Do failures in non-technical skills contribute to fatal medical accidents in Japan? A review of the 2010–2013 national accident reports

Science.gov (United States)

Uramatsu, Masashi; Fujisawa, Yoshikazu; Mizuno, Shinya; Souma, Takahiro; Komatsubara, Akinori; Miki, Tamotsu

2017-01-01

Objectives We sought to clarify how large a proportion of fatal medical accidents can be considered to be caused by poor non-technical skills, and to support development of a policy to reduce number of such accidents by making recommendations about possible training requirements. Design Summaries of reports of fatal medical accidents, published by the Japan Medical Safety Research Organization, were reviewed individually. Three experienced clinicians and one patient safety expert conducted the reviews to determine the cause of death. Views of the patient safety expert were given additional weight in the overall determination. Setting A total of 73 summary reports of fatal medical accidents were reviewed. These reports had been submitted by healthcare organisations across Japan to the Japan Medical Safety Research Organization between April 2010 and March 2013. Primary and secondary outcome measures The cause of death in fatal medical accidents, categorised into technical skills, non-technical skills and inevitable progress of disease were evaluated. Non-technical skills were further subdivided into situation awareness, decision making, communication, team working, leadership, managing stress and coping with fatigue. Results Overall, the cause of death was identified as non-technical skills in 34 cases (46.6%), disease progression in 33 cases (45.2%) and technical skills in two cases (5.5%). In two cases, no consensual determination could be achieved. Further categorisation of cases of non-technical skills were identified as 14 cases (41.2%) of problems with situation awareness, eight (23.5%) with team working and three (8.8%) with decision making. These three subcategories, or combinations of them, were identified as the cause of death in 33 cases (97.1%). Conclusions Poor non-technical skills were considered to be a significant cause of adverse events in nearly half of the fatal medical accidents examined. Improving non-technical skills may be effective for

22 CFR 102.8 - Reporting accidents.

Science.gov (United States)

2010-04-01

... 22 Foreign Relations 1 2010-04-01 2010-04-01 false Reporting accidents. 102.8 Section 102.8... Accidents Abroad § 102.8 Reporting accidents. (a) To airline and Civil Aeronautics Administration... probably be the first to be informed of the accident, in which event he will be expected to report the...

Progress and Updates of Regulatory Challenges and Safety Issues in Korea during Three Years after Fukushima Accident

International Nuclear Information System (INIS)

Lee, Young Eal; Kim, Kyun Tae

2014-01-01

Before the public fear on radiation risk caused by neighboring country's severe accident disappeared, a series of nuclear safety issues last 3 years made a few reactors shut down and the public trust much lower than before. Because of these scandals such as cover-ups, forged certificated items, corruption of manager of licensee and so on, many efforts made during three year after Fukushima accident on improving the nuclear safety were invalidated and even regulators as well as operators have been sharply criticized for its responsibility and transparency. This paper shares information on the progress and uprates achieved in Korea so far in connection with the safety issues caused during last 3 years and actions taken by the regulatory body. Before the public fear on radiation risk caused by neighboring country's severe accident disappeared, a series of nuclear safety issues last 3 years made a few reactors shut down and the public trust much lower than before. Because of these scandals such as cover-ups, forged certificated items, corruption of manager of licensee and so on, many efforts made during three year after Fukushima accident on improving the nuclear safety were invalidated and even regulators as well as operators have been sharply criticized for its responsibility and transparency

Progress and Updates of Regulatory Challenges and Safety Issues in Korea during Three Years after Fukushima Accident

Energy Technology Data Exchange (ETDEWEB)

Lee, Young Eal; Kim, Kyun Tae [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2014-10-15

Before the public fear on radiation risk caused by neighboring country's severe accident disappeared, a series of nuclear safety issues last 3 years made a few reactors shut down and the public trust much lower than before. Because of these scandals such as cover-ups, forged certificated items, corruption of manager of licensee and so on, many efforts made during three year after Fukushima accident on improving the nuclear safety were invalidated and even regulators as well as operators have been sharply criticized for its responsibility and transparency. This paper shares information on the progress and uprates achieved in Korea so far in connection with the safety issues caused during last 3 years and actions taken by the regulatory body. Before the public fear on radiation risk caused by neighboring country's severe accident disappeared, a series of nuclear safety issues last 3 years made a few reactors shut down and the public trust much lower than before. Because of these scandals such as cover-ups, forged certificated items, corruption of manager of licensee and so on, many efforts made during three year after Fukushima accident on improving the nuclear safety were invalidated and even regulators as well as operators have been sharply criticized for its responsibility and transparency.

Accident analyses in nuclear power plants following external initiating events and in the shutdown state. Final report; Unfallanalysen in Kernkraftwerken nach anlagenexternen ausloesenden Ereignissen und im Nichtleistungsbetrieb. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Loeffler, Horst; Kowalik, Michael; Mildenberger, Oliver; Hage, Michael

2016-06-15

The work which is documented here provides the methodological basis for improvement of the state of knowledge for accident sequences after plant external initiating events and for accident sequences which begin in the shutdown state. The analyses have been done for a PWR and for a BWR reference plant. The work has been supported by the German federal ministry BMUB under the label 3612R01361. Top objectives of the work are: - Identify relevant event sequences in order to define characteristic initial and boundary conditions - Perform accident analysis of selected sequences - Evaluate the relevance of accident sequences in a qualitative way The accident analysis is performed with the code MELCOR 1.8.6. The applied input data set has been significantly improved compared to previous analyses. The event tree method which is established in PSA level 2 has been applied for creating a structure for a unified summarization and evaluation of the results from the accident analyses. The computer code EVNTRE has been applied for this purpose. In contrast to a PSA level 2, the branching probabilities of the event tree have not been determined with the usual accuracy, but they are given in an approximate way only. For the PWR, the analyses show a considerable protective effect of the containment also in the case of beyond design events. For the BWR, there is a rather high probability for containment failure under core melt impact, but nevertheless the release of radionuclides into the environment is very limited because of plant internal retention mechanisms. This report concludes with remarks about existing knowledge gaps and with regard to core melt sequences, and about possible improvements of the plant safety.

Thermal Hydraulic design parameters study for severe accidents using neural networks

Energy Technology Data Exchange (ETDEWEB)

Roh, Chang Hyun; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Chang, Keun Sun [Sunmoon University, Asan (Korea, Republic of)

1998-12-31

To provide the information on severe accident progression is very important for advanced or new type of nuclear power plant (NPP) design. A parametric study, therefore, was performed to investigate the effect of thermal hydraulic design parameters on severe accident progression of pressurized water reactors (PWRs). Nine parameters, which are considered important in NPP design or severe accident progression, were selected among the various thermal hydraulic design parameters. The backpropagation neural network (BPN) was used to determine parameters, which might more strongly affect the severe accident progression, among nine parameters. For training, different input patterns were generated by the latin hypercube sampling (LHS) technique and then different target patterns that contain core uncovery time and vessel failure time were obtained for Young Gwang Nuclear (YGN) Units 3 and 4 using modular accident analysis program (MAAP) 3.0B code. Three different severe accident scenarios, such as two loss of coolant accidents (LOCAs) and station blackout (SBO), were considered in this analysis. Results indicated that design parameters related to refueling water storage tank (RWST), accumulator and steam generator (S/G) have more dominant effects on the progression of severe accidents investigated, compared to the other six parameters. 9 refs., 5 tabs. (Author)

Thermal Hydraulic design parameters study for severe accidents using neural networks

Energy Technology Data Exchange (ETDEWEB)

Roh, Chang Hyun; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Chang, Keun Sun [Sunmoon University, Asan (Korea, Republic of)

1997-12-31

To provide the information on severe accident progression is very important for advanced or new type of nuclear power plant (NPP) design. A parametric study, therefore, was performed to investigate the effect of thermal hydraulic design parameters on severe accident progression of pressurized water reactors (PWRs). Nine parameters, which are considered important in NPP design or severe accident progression, were selected among the various thermal hydraulic design parameters. The backpropagation neural network (BPN) was used to determine parameters, which might more strongly affect the severe accident progression, among nine parameters. For training, different input patterns were generated by the latin hypercube sampling (LHS) technique and then different target patterns that contain core uncovery time and vessel failure time were obtained for Young Gwang Nuclear (YGN) Units 3 and 4 using modular accident analysis program (MAAP) 3.0B code. Three different severe accident scenarios, such as two loss of coolant accidents (LOCAs) and station blackout (SBO), were considered in this analysis. Results indicated that design parameters related to refueling water storage tank (RWST), accumulator and steam generator (S/G) have more dominant effects on the progression of severe accidents investigated, compared to the other six parameters. 9 refs., 5 tabs. (Author)

Application of RASCAL code for multiunit accident in domestic nuclear sites

Energy Technology Data Exchange (ETDEWEB)

Park, Sang Hyun; Jeong, Seung Young [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2014-10-15

All of domestic nuclear power plant sites are multiunit site (at least 5 - 6 reactors are operating), so this capability has to be quickly secured for nuclear licensee and institutes responsible for nuclear emergency response. In this study, source term and offsite dose from multiunit event were assessed using a computer code, RASCAL. An emergency exercise scenario was chosen to verify applicability of the codes to domestic nuclear site accident. Employing tools and new features of the code, such as merging more than two individual source terms and source term estimate for long term progression accident, main parameters and information in the scenario, release estimates and dose projections were performed. Radiological releases and offsite doses from multiunit accident were calculated using RASCAL.. A scenario, in which three reactors were damaged coincidently by a great natural disaster, was considered. Surrogate plants were chosen for the code calculation. Source terms of each damaged unit were calculated individually first, and then total source term and integrated offsite dose assessment data was acquired using a source term merge function in the code. Also comparison between LTSBO and LOCA source term estimate options was performed. Differences in offsite doses were caused by release characteristics. From LTSBO option, iodines were released much higher than LOCA. Also LTSBO source term release was delayed and the duration was longer than LOCA. This option would be useful to accidents which progress with much longer time frame than LOCA. RASCAL can be useful tool for radiological consequence assessment in domestic nuclear site accidents.

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

International Nuclear Information System (INIS)

O'Kula, K.R.; Taylor, R.P. Jr.; Ashbaugh, S.G.

1995-01-01

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

APRI-7 Accident Phenomena of Risk Importance. A progress report on research in the field of severe accidents in 2009-2011

International Nuclear Information System (INIS)

Garis, Ninos; Agrell, Maria; Glaenneskog, Henrik

2012-01-01

Knowledge of the phenomena that may occur during severe accidents in a nuclear power plant is an important prerequisite for being able to predict the plant behavior, in order to formulate procedures and instructions for incident handling, for contingency planning, and to get good quality at the accident analysis and risk studies. Since the early 80's nuclear power companies and authorities in Sweden has collaborated in research on severe reactor accidents. Cooperation in the beginning was mostly linked to strengthening the protection against environmental impacts after a severe reactor accident, in particular to develop systems for filtered depressurization of the reactor containment. Since the early 90's the cooperation has partially changed and shifted to the phenomenological questions of risk dominance. During the years 2009-2011, cooperation continued in the research-program APRI-7. The aim was to show whether the solutions adopted in the Swedish strategy for accident management provides reasonable protection for the environment. This was done by gaining detailed knowledge of both important phenomena in the hearth melting behavior, and the amount of radioactivity that can be discharged to the surroundings during a severe accident. To achieve this aim, the research program has included a follow-up of international research in severe accidents and evaluation of results, and continued to support research at KTH and Chalmers Univ. of severe accidents. The follow-up of international research has promoted the exchange of knowledge and experience and has provided access to a wealth of information about various phenomena relevant to the events at severe accidents. This was important to obtain a good basis for assessment of abatement measures in the Swedish nuclear reactors. Continuing support to the Royal Inst. of Technology has provided increased knowledge about the ability to cool the molten core of the reactor vessel and the processes associated with cooling the

Radiation, accidents, society

International Nuclear Information System (INIS)

1988-01-01

This book is meant to be used as a reference book for information officers at the event of a nuclear accident. The main part is edited in alphabetical order to facilitate use under stress. The book gives a short review of the health risks of radiation, and descriptions of accidents that have occured. The index words that have been chosen for the main part of the book have been selected due to experiences in connection with incidents and accidents. (L.E.)

Use of PSA to support accident management at NPPs

International Nuclear Information System (INIS)

Gomez Cobo, A.

1997-01-01

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

Criticality accident in Argentina

International Nuclear Information System (INIS)

Oliveira, A.R. de.

1984-01-01

A recent criticality type accident, ocurred in Argetina, is commented. Considerations about the nature of the facility where this accident took place, its genesis, type of operation carried out on the day of the event, and the medical aspects involved are done. (Author) [pt

Desktop Severe Accident Graphic Simulator Module for CANDU6 : PSAIS

International Nuclear Information System (INIS)

Park, S. Y.; Song, Y. M.

2015-01-01

The ISAAC ((Integrated Severe Accident Analysis Code for CANDU Plant) code is a system level computer code capable of performing integral analyses of potential severe accident progressions in nuclear power plants, whose main purpose is to support a Level 2 probabilistic safety assessment or severe accident management strategy developments. The code has the capability to predict a severe accident progression by modeling the CANDU6- specific systems and the expected physical phenomena based on the current understanding of the unique accident progressions. The code models the sequence of accident progressions from a core heatup, pressure tube/calandria tube rupture after an uncovery from inside and outside, a relocation of the damaged fuel to the bottom of the calandria, debris behavior in the calandria, corium quenching after a debris relocation from the calandria to the calandria vault and an erosion of the calandria vault concrete floor, a hydrogen burn, and a reactor building failure. Along with the thermal hydraulics, the fission product behavior is also considered in the primary system as well as in the reactor building

The Chernobyl accident consequences

International Nuclear Information System (INIS)

2001-04-01

Five teen years later, Tchernobyl remains the symbol of the greater industrial nuclear accident. To take stock on this accident, this paper proposes a chronology of the events and presents the opinion of many international and national organizations. It provides also web sites references concerning the environmental and sanitary consequences of the Tchernobyl accident, the economic actions and propositions for the nuclear safety improvement in the East Europe. (A.L.B.)

Industrial accidents triggered by lightning.

Science.gov (United States)

Renni, Elisabetta; Krausmann, Elisabeth; Cozzani, Valerio

2010-12-15

Natural disasters can cause major accidents in chemical facilities where they can lead to the release of hazardous materials which in turn can result in fires, explosions or toxic dispersion. Lightning strikes are the most frequent cause of major accidents triggered by natural events. In order to contribute towards the development of a quantitative approach for assessing lightning risk at industrial facilities, lightning-triggered accident case histories were retrieved from the major industrial accident databases and analysed to extract information on types of vulnerable equipment, failure dynamics and damage states, as well as on the final consequences of the event. The most vulnerable category of equipment is storage tanks. Lightning damage is incurred by immediate ignition, electrical and electronic systems failure or structural damage with subsequent release. Toxic releases and tank fires tend to be the most common scenarios associated with lightning strikes. Oil, diesel and gasoline are the substances most frequently released during lightning-triggered Natech accidents. Copyright © 2010 Elsevier B.V. All rights reserved.

Radiation accidents

International Nuclear Information System (INIS)

Nenot, J.C.

1996-01-01

Analysis of radiation accidents over a 50 year period shows that simple cases, where the initiating events were immediately recognised, the source identified and under control, the medical input confined to current handling, were exceptional. In many cases, the accidents were only diagnosed when some injuries presented by the victims suggested the radiological nature of the cause. After large-scale accidents, the situation becomes more complicated, either because of management or medical problems, or both. The review of selected accidents which resulted in severe consequences shows that most of them could have been avoided; lack of regulations, contempt for rules, human failure and insufficient training have been identified as frequent initiating parameters. In addition, the situation was worsened because of unpreparedness, insufficient planning, unadapted resources, and underestimation of psychosociological aspects. (author)

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...

Addressing severe accidents in the CANDU 9 design

International Nuclear Information System (INIS)

Nijhawan, S.M.; Wight, A.L.; Snell, V.G.

1998-01-01

CANDU 9 is a single-unit evolutionary heavy-water reactor based on the Bruce/Darlington plants. Severe accident issues are being systematically addressed in CANDU 9, which includes a number of unique features for prevention and mitigation of severe accidents. A comprehensive severe accident program has been formulated with feedback from potential clients and the Canadian regulatory agency. Preliminary Probabilistic Safety Analyses have identified the sequences and frequency of system and human failures that may potentially lead to initial conditions indicating onset of severe core damage. Severe accident consequence analyses have used these sequences as a guide to assess passive heat sinks for the core, and containment performance. Estimates of the containment response to mass and energy injections typical of postulated severe accidents have been made and the results are presented. We find that inherent CANDU severe accident mitigation features, such as the presence of large water volumes near the fuel (moderator and shield tank), permit a relatively slow severe accident progression under most plant damage states, facilitate debris coolability and allow ample time for the operator to arrest the progression within, progressively, the fuel channels, calandria vessel or shield tank. The large-volume CANDU 9 containment design complements these features because of the long times to reach failure

Tchernobyl accident

International Nuclear Information System (INIS)

1986-06-01

First, R.M.B.K type reactors are described. Then, safety problems are dealt with reactor control, behavior during transients, normal loss of power and behavior of the reactor in case of leak. A possible scenario of the accident of Tchernobyl is proposed: events before the explosion, possible initiators, possible scenario and events subsequent to the core meltdown (corium-concrete interaction, interaction with the groundwater table). An estimation of the source term is proposed first from the installation characteristics and the supposed scenario of the accident, and from the measurements in Europe; radiological consequences are also estimated. Radioactivity measurements (Europe, Scandinavia, Western Europe, France) are given in tables (meteorological maps and fallouts in Europe). Finally, a description of the site is given [fr

Implementation of accident management programmes in nuclear power plants

International Nuclear Information System (INIS)

2004-01-01

good practices and developments in Member States and is intended as reference material for NPPs, as well as an information source for other organizations such as regulatory bodies. It is a follow-up to the IAEA report on Accident Management Programmes in Nuclear Power Plants, published in 1994, and reflects the considerable progress made since that time. The objective of this report is to provide a description of the elements to be addressed by the team responsible for developing and implementing a plant specific AMP at an NPP. Although it is intended primarily for use by NPP operators, utilities and their technical support organizations, it can also facilitate preparation of the relevant national regulatory requirements. Important event sequences that may lead to severe accidents shall be identified using a combination of probabilistic methods, deterministic methods and sound engineering judgement. These event sequences shall then be reviewed against a set of criteria aimed at determining which severe accidents should be addressed in the design. Potential design or procedural changes that could either reduce the likelihood of these selected events, or mitigate their consequences, should these selected events occur, shall be evaluated, and shall be implemented if reasonably practicable. Consideration shall be given to the plant full design capabilities, including the possible use of some systems (i.e. safety and non-safety systems) beyond their originally intended function and anticipated operating conditions, and the use of additional temporary systems to return the plant to a controlled state and/or to mitigate the consequences of a severe accident, provided that it can be shown that the systems are able to function in the environmental conditions to be expected. For multiunit plants, consideration shall be given to the use of available means and/or support from other units, provided that the safe operation of the other units is not compromised. Accident management

Development of Accident Scenarios and Quantification Methodology for RAON Accelerator

International Nuclear Information System (INIS)

Lee, Yongjin; Jae, Moosung

2014-01-01

The RIsp (Rare Isotope Science Project) plans to provide neutron-rich isotopes (RIs) and stable heavy ion beams. The accelerator is defined as radiation production system according to Nuclear Safety Law. Therefore, it needs strict operate procedures and safety assurance to prevent radiation exposure. In order to satisfy this condition, there is a need for evaluating potential risk of accelerator from the design stage itself. Though some of PSA researches have been conducted for accelerator, most of them focus on not general accident sequence but simple explanation of accident. In this paper, general accident scenarios are developed by Event Tree and deduce new quantification methodology of Event Tree. In this study, some initial events, which may occur in the accelerator, are selected. Using selected initial events, the accident scenarios of accelerator facility are developed with Event Tree. These results can be used as basic data of the accelerator for future risk assessments. After analyzing the probability of each heading, it is possible to conduct quantification and evaluate the significance of the accident result. If there is a development of the accident scenario for external events, risk assessment of entire accelerator facility will be completed. To reduce the uncertainty of the Event Tree, it is possible to produce a reliable data via the presented quantification techniques

Overview of core disruptive accidents

International Nuclear Information System (INIS)

Marchaterre, J.F.

1977-01-01

An overview of the analysis of core-disruptive accidents is given. These analyses are for the purpose of understanding and predicting fast reactor behavior in severe low probability accident conditions, to establish the consequences of such conditions and to provide a basis for evaluating consequence limiting design features. The methods are used to analyze core-disruptive accidents from initiating event to complete core disruption, the effects of the accident on reactor structures and the resulting radiological consequences are described

Progress in Addressing DNFSB Recommendation 2002-1 Issues: Improving Accident Analysis Software Applications

International Nuclear Information System (INIS)

VINCENT, ANDREW

2005-01-01

Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2002-1 (''Quality Assurance for Safety-Related Software'') identified a number of quality assurance issues on the use of software in Department of Energy (DOE) facilities for analyzing hazards, and designing and operating controls to prevent or mitigate potential accidents. Over the last year, DOE has begun several processes and programs as part of the Implementation Plan commitments, and in particular, has made significant progress in addressing several sets of issues particularly important in the application of software for performing hazard and accident analysis. The work discussed here demonstrates that through these actions, Software Quality Assurance (SQA) guidance and software tools are available that can be used to improve resulting safety analysis. Specifically, five of the primary actions corresponding to the commitments made in the Implementation Plan to Recommendation 2002-1 are identified and discussed in this paper. Included are the web-based DOE SQA Knowledge Portal and the Central Registry, guidance and gap analysis reports, electronic bulletin board and discussion forum, and a DOE safety software guide. These SQA products can benefit DOE safety contractors in the development of hazard and accident analysis by precluding inappropriate software applications and utilizing best practices when incorporating software results to safety basis documentation. The improvement actions discussed here mark a beginning to establishing stronger, standard-compliant programs, practices, and processes in SQA among safety software users, managers, and reviewers throughout the DOE Complex. Additional effort is needed, however, particularly in: (1) processes to add new software applications to the DOE Safety Software Toolbox; (2) improving the effectiveness of software issue communication; and (3) promoting a safety software quality assurance culture

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

International Nuclear Information System (INIS)

Chang, Hyun Sop

1996-02-01

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

50 CFR 25.72 - Reporting of accidents.

Science.gov (United States)

2010-10-01

... 50 Wildlife and Fisheries 6 2010-10-01 2010-10-01 false Reporting of accidents. 25.72 Section 25... Reporting of accidents. Accidents involving damage to property, injury to the public or injury to wildlife..., but in no event later than 24 hours after the accident, by the persons involved, to the refuge manager...

TIARA: treatment initiatives after radiological accidents

International Nuclear Information System (INIS)

Menetrier, F.; Berard, Ph.; Joussineau, S.; Stradling, N.; Hodgson, A.; List, V.; Morcillo, M.A.; Paile, W.; Holt, D.C.B.; Eriksson, T.

2007-01-01

This paper describes the objectives, and reviews the progress, of the European project 'Treatment Initiatives After Radiological Accidents' (TIARA). TIARA forms part of the 'Preparatory Action for Security Research' (PASR) launched by the European Commission in 2004. The Preparatory Action is intended to reach preliminary conclusions on the needs for the security of EU citizens. It prepared a comprehensive Security Research Programme as part of the Commission's Seventh Framework Programme proposal, which was adopted in 2006 and launched in 2007. The principal purpose of TIARA is to constitute a European network that will participate in facilitating the management of a crisis in the event of the malevolent dispersal of radionuclides into the public environment. (authors)

Radiological Emergency Preparedness after the Early Phase of an Accident : Focusing on an Air Contamination Event

International Nuclear Information System (INIS)

Jeong, Hyo Joon; Hwang, Won Tae; Kim, Eun Han; Han, Moon Hee

2010-01-01

Toxic materials in an urban area can be caused by a variety of events, such as accidental releases on industrial complexes, accidents during the transportation of hazardous materials and intentional explosions. Most governments around the world and their citizens have become increasingly worried about intentional accidents in urban area after the 911 terrorist attack in the United States of America. Even though there have been only a few attempted uses of Radiological Dispersal Devices (RDDs), accidental releases have occurred many times at commercial nuclear power plants and nuclear waste disposal sites. When an intentional release of radioactive materials occurs in an urban area, air quality for radioactive materials in the environment is of great importance to take action for countermeasures and environmental risk assessments. Atmospheric modeling is part of the decision making tasks and that it is particularly important for emergency managers as they often need to take actions quickly on very inadequate information(1). A simple model such as HOTSPOT required wind direction and source term would be enough to support the decision making in the early phase of an accident, but more sophisticated atmospheric modeling is required to adjust decontamination area and relocation etc after the early phase of an accidental event. In this study, we assume an explosion of 137 Cs using RDDs in the metropolitan area of Soul, South Korea. California Puff Model (CALPUFF) is used to calculate an atmospheric dispersion and transport for 137 Cs. Atmospheric dispersion and quantitative radiological risk analysis for 137 Cs were performed assuming an intentional explosion in the metropolitan area of Soul, South Korea after the early phase of emergency. These kinds of atmospheric modeling and risk analysis could provide a means for decision makers to take action on important issues such as the cleanup of the contaminated area and countermeasures to protect the public caused by

Guidance on accidents involving radioactivity

International Nuclear Information System (INIS)

1989-01-01

This annex contains advice to Health Authorities on their response to accidents involving radioactivity. The guidance is in six parts:-(1) planning the response required to nuclear accidents overseas, (2) planning the response required to UK nuclear accidents a) emergency plans for nuclear installations b) nuclear powered satellites, (3) the handling of casualties contaminated with radioactive substances, (4) background information for dealing with queries from the public in the event of an accident, (5) the national arrangements for incident involving radioactivity (NAIR), (6) administrative arrangements. (author)

Radiological accident 'The Citadel' medical aspects

International Nuclear Information System (INIS)

Cardenas Herrera, Juan; Fernandez, Isis M.; Lopez, Gladys; Garcia, Omar; Lamadrid, Ana I.; Ramos, Enma O.; Villa, Rosario; Giron, Carmen M.; Escobar, Myrian; Zerpa, Miguel; Romero, Argenis H.; Medina, Julio; Laurenti, Zenia; Oliva, Maria T.; Sierra, Nitza; Lorenzo, Alexis

2008-01-01

The work exposes the medical actions carried out in the mitigation of the consequences of the accident and its main results. In a facility of storage of radioactive waste in Caracas, Venezuela, it was happened a radiological accident. This event caused radioactive contamination of the environment, as well as the irradiation and radioactive contamination of at least 10 people involved in the fact, in its majority children. Cuban institutions participated in response to the accident. Among the decisions adopted by the team of combined work Cuban-Venezuelan, we find the one of transferring affected people to Cuba, for their dosimetric and medical evaluation. Being designed a work strategy to develop the investigations to people affected by the radiological accident, in correspondence with the circumstances, magnitude and consequences of the accident. The obtained main results are: 100% presented affectations in its health, not associate directly to the accident, although the accident influenced in its psychological state. In 3 of studied people they were detected radioactive contamination with Cesium -137 with dose among 2.01 X 10-4 Sv up to 2.78 X 10-4 Sv. This accident demonstrated the necessity to have technical capacities to face these events and the importance of the international solidarity. (author)

Nuclear ship accidents

International Nuclear Information System (INIS)

Oelgaard, P.L.

1993-05-01

In this report available information on 28 nuclear ship accident and incidents is considered. Of these 5 deals with U.S. ships and 23 with USSR ships. The ships are in almost all cases nuclear submarines. Only events that involve the nuclear propulsion plants, radiation exposures, fires/explosions and sea water leaks into the submarines are considered. Comments are made on each of the events, and at the end of the report an attempt is made to point out the weaknesses of the submarine designs which have resulted in the accidents. It is emphasized that much of the available information is of a rather dubious nature. consequently some of the assessments made may not be correct. (au)

Progress in the U.S. nuclear utility industry 1979-1989

Energy Technology Data Exchange (ETDEWEB)

NONE

1989-07-01

March 28, 1979 changed the course of the commercial U.S. nuclear utility industry. An accident at Three Mile Island Nuclear Station Unit 2 in Middletown, Pennsylvania damaged the reactor's fuel core, as well as the industry's reputation and confidence. In the months after the event, the president of the United States, the nuclear industry, the public, government regulators and the media sought answers to many questions. Among the most important were, how and why did the accident happen? The President's Commission on the Accident at TMI, the Kemeny Commission, was formed to address these questions. Four major causes were identified by the commission: - inadequate or inappropriate operator training; - mechanical problems and faulty instrumentation; - poor control room design; - communication failures at the facility and in information exchange within the industry. Of even greater importance was another question: How could another such accident be prevented? A look at the industry's progress in the 10 years since the TMI accident shows this question has been vigorously addressed and that corrective actions have been taken.

Progress in the U.S. nuclear utility industry 1979-1989

International Nuclear Information System (INIS)

1989-01-01

March 28, 1979 changed the course of the commercial U.S. nuclear utility industry. An accident at Three Mile Island Nuclear Station Unit 2 in Middletown, Pennsylvania damaged the reactor's fuel core, as well as the industry's reputation and confidence. In the months after the event, the president of the United States, the nuclear industry, the public, government regulators and the media sought answers to many questions. Among the most important were, how and why did the accident happen? The President's Commission on the Accident at TMI, the Kemeny Commission, was formed to address these questions. Four major causes were identified by the commission: - inadequate or inappropriate operator training; - mechanical problems and faulty instrumentation; - poor control room design; - communication failures at the facility and in information exchange within the industry. Of even greater importance was another question: How could another such accident be prevented? A look at the industry's progress in the 10 years since the TMI accident shows this question has been vigorously addressed and that corrective actions have been taken

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

International Nuclear Information System (INIS)

Saleh, Joseph H.; Pendley, Cynthia C.

2012-01-01

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

The protection of on-site personnel in the event of a radiological accident

International Nuclear Information System (INIS)

Morrey, M.; Simister, D.N.

2003-01-01

The National Radiological Protection Board (NPRB) is responsible in the UK for advising Government and other responsible bodies on the principles for responding to radiological emergencies. NRPB has published appropriate advice on the off-site protection of the public and on the protection of workers involved in taking mitigating actions to reduce the exposure of others. This paper puts forward a suggested framework for the protection of on-site personnel in the event of a radiological emergency which might include a criticality accident. This framework both dovetails with existing planning for the protection of members of the public off-site, and also takes account of specific differences between the situations on and off-site. (author)

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)

Organizational forms of medical care in the event of radiation accidents in the German Democratic Republic

International Nuclear Information System (INIS)

Nack, P.; Arndt, D.; Schuettmann, W.

1977-01-01

Medical care of radiation casualties in the German Democratic Republic (GDR) is organized on two levels. On the level of users the responsible Medical Officers guarantee both the routine control of persons occupationally exposed to radiation and first aid in the event of accidents. On the second level medical treatment is given either in the Clinical Department of the National Board of Nuclear Safety and Radiation Protection or in specialized national health system clinics having facilities for intensive medical care. A decision on hospitalization is made according to the conditions of the accident and the necessary diagnostic and therapeutic measures as a rule are based on consultations between the responsible Medical Officer and the departments of the Board (Emergency Assistance Service, Clinical Department, Consultative Committee). For serious cases where haematological complications can be expected, a central medical clinic with facilities for bone-marrow transplants is available. The casualties are treated in local clinics which are provided with continuous support and advice by the Board. This support consists in: (i) immediate activity by a consultative committee of the Board's physicians and scientists experienced and trained in radiation protection and the treatment of radiation accidents; (ii) the requirement of compulsory examination methods and take-over of specialized laboratory investigations; and (iii) the use of a mobile emergency measuring system in cases of additional incorporation. It is the main principle of medical care in case of radiation accidents to consult, as early as possible, a medical consultative committee of the Board in the field of radiation protection at each step of medical care. (author)

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

46 CFR 78.33-5 - Accidents to machinery.

Science.gov (United States)

2010-10-01

... 46 Shipping 3 2010-10-01 2010-10-01 false Accidents to machinery. 78.33-5 Section 78.33-5 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) PASSENGER VESSELS OPERATIONS Reports of Accidents, Repairs, and Unsafe Equipment § 78.33-5 Accidents to machinery. (a) In the event of an accident...

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

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations. Appendix E (Sections E.9-E.16), Volume 2, Part 3B

International Nuclear Information System (INIS)

Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J.; Wong, S.M.; Bley, D.; Johnson, D.

1994-06-01

Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analyses that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program includes that of a level-3 PRA. In phase 2, mid-loop operation was selected as the plant configuration to be analyzed based on the results of the phase 1 study. The objective of the phase 2 study is to perform a detailed analysis of the potential accident scenarios that may occur during mid-loop operation, and compare the results with those of NUREG-1150. The scope of the level-1 study includes plant damage state analysis, and uncertainty analysis. Volume 1 summarizes the results of the study. Internal events analysis is documented in Volume 2. It also contains an appendix that documents the part of the phase 1 study that has to do with POSs other than mid-loop operation. Internal fire and internal flood analyses are documented in Volumes 3 and 4. A separate study on seismic analysis, documented in Volume 5, was performed for the NRC by Future Resources Associates, Inc. Volume 6 documents the accident progression, source terms, and consequence analysis

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations. Appendix E (Sections E.9-E.16), Volume 2, Part 3B

Energy Technology Data Exchange (ETDEWEB)

Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J.; Wong, S.M. [Brookhaven National Lab., Upton, NY (United States); Bley, D.; Johnson, D. [PLG Inc., Newport Beach, CA (United States)] [and others

1994-06-01

Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analyses that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program includes that of a level-3 PRA. In phase 2, mid-loop operation was selected as the plant configuration to be analyzed based on the results of the phase 1 study. The objective of the phase 2 study is to perform a detailed analysis of the potential accident scenarios that may occur during mid-loop operation, and compare the results with those of NUREG-1150. The scope of the level-1 study includes plant damage state analysis, and uncertainty analysis. Volume 1 summarizes the results of the study. Internal events analysis is documented in Volume 2. It also contains an appendix that documents the part of the phase 1 study that has to do with POSs other than mid-loop operation. Internal fire and internal flood analyses are documented in Volumes 3 and 4. A separate study on seismic analysis, documented in Volume 5, was performed for the NRC by Future Resources Associates, Inc. Volume 6 documents the accident progression, source terms, and consequence analysis.

Synthesis of public authorities organisation in case of emergency and in a post-event situation (following a nuclear accident or a radiological attack) in France and abroad

International Nuclear Information System (INIS)

Kayser, O.

2010-01-01

After having briefly recalled how an emergency situation (notably in case of nuclear accident or radiological attack) is taken into account in the organisation of public authorities through specific plans (PPI or plans particuliers d'intervention, intervention specific plans), this report also describes how the situation is handled by these authorities after the end of the emergency situation (i.e. when the risk of new radioactive releases is over). This post-event stage is split into two phases: a transition phase which lasts several weeks or months, and a long term consequence management phase (over months or years). The author first describes the specificities of a nuclear or radiological event (accident or attack). He recalls the global public organisation and the involved actors. For the post-event period, he indicates the various actions, describes the interdepartmental coordination and the various aspects of the program designed to manage accident consequences on the long term. He also describes the roles of permanent bodies, agencies and institutes (ASN, ASND, MSNR, IRSN, INVS, ADEME, AFSSA, Meteo France, CEA, ANDRA, AREVA, EDF, ministries). The last part describes the action of public authorities in case of a nuclear accident occurring abroad. This includes relationship with European and international bodies

Accident considerations in LMFBR design

International Nuclear Information System (INIS)

Simpson, D.E.; Alter, H.; Fauske, H.K.; Hikido, K.; Keaten, R.W.; Stevenson, M.G.; Strawbridge, L.

1975-12-01

LMFBR safety design criteria are discussed from the standpoints of accident severity classification and damage criteria, and the following design events are considered: fuel failure propagation, reactivity addition faults, heat transport system events, steam generator faults, sodium spills, fuel handling and storage faults, and external events

Technical organization of safety authorities for the event of an accident at a nuclear installation

International Nuclear Information System (INIS)

Scherrer, J.; Evrard, J.M.; Ney, J.

1986-01-01

Within the general context of nuclear safety, the Central Nuclear Installation Safety Service of the French Ministry for Industry and its technical backup, the Institute for Radiation Protection and Nuclear Safety of the CEA (Atomic Energy Commission), have established a special organization designed to provide real-time forecasts of the evolution of a nuclear accident situation with sufficient forewarning for the local representative of the Government (the Commissaire de la Republique in the Departement affected) to implement, as required, effective countermeasures to protect the population - for example, confinement indoors or evacuation. Descriptions are given of the principles of this organization and the particular precautions taken to confront the problems of mobilizing experts and of dealing with the saturation of normal telecommunications channels to be expected in the event of a nuclear accident. The organization set up for the installations belonging to Electricite de France is given as a detailed example. Particular stress is placed on the organizational arrangements of the Institute for Radiation Protection and Nuclear Safety designed to provide the emergency teams with the evaluation and forecasting tools they require to carry out their tasks. The procedures are on the whole well developed for atmospheric radioactivity transport, for which operational models already exist. Computer-backed methods with improved performance are at present being developed. A method of forecasting the behaviour of the releases resulting from nuclear accidents is set out for pressurized water reactors, based on evaluating the physical state of the installation, confinement integrity, availability of safety and backup systems, support systems and feed sources and on forecasting how this state will develop on the basis of measured and inferred physical values transmitted from the affected power station through a national network. The experience acquired during accident

Chernobyl accident and Denmark

International Nuclear Information System (INIS)

1986-12-01

The report describes the Chernobyl accident and its consequences for Denmark in particular. It was commissioned by The Secretary of State for the Environment. The event at the accident site, the release and dispersal of radioactive substances into the atmosphere and over Europe, is described. A discussion of the Danish organisation for nuclear emergencies, how it was activated and adapted to the actual situation, is given. A comprehensive description of the radiological contamination in Denmark following the accident and the estimated health effects, is presented. The situation in other European countries is mentioned. (author)

Potential applicability of fuzzy set theory to analyses of containment response and uncertainty for postulated severe accidents

International Nuclear Information System (INIS)

Chun, M.H.; Ahn, K.I.

1991-01-01

An important issue faced by contemporary risk analysts of nuclear power plants is how to deal with uncertainties that arise in each phase of probabilistic risk assessments. The major uncertainty addressed in this paper is the one that arises in the accident-progression event trees (APETs), which treat the physical processes affecting the core after an initiating event occurs. Recent advances in the theory of fuzzy sets make it possible to analyze the uncertainty related to complex physical phenomena that may occur during a severe accident of nuclear power plants by means of fuzzy set or possibility concept. The main purpose of this paper is to prevent the results of assessment of the potential applicability of the fuzzy set theory to the uncertainty analysis of APETs as a possible alternative procedure to that used in the most recent risk assessment

International aspects of nuclear accidents

International Nuclear Information System (INIS)

Uematsu, K.

1989-09-01

The accident at Chernobyl revealed that there were shortcomings and gaps in the existing international mechanisms and brought home to governments the need for stronger measures to provide better protection against the risks of severe accidents. The main thrust of international co-operation with regard to nuclear safety issues is aimed at achieving a uniformly high level of safety in nuclear power plants through continuous exchanges of research findings and feedback from reactor operating experience. The second type of problem posed in the event of an accident resulting in radioactive contamination of several countries relates to the obligation to notify details of the circumstances and nature of the accident speedily so that the countries affected can take appropriate protective measures and, if necessary, organize mutual assistance. Giving the public accurate information is also an important aspect of managing an emergency situation arising from a severe accident. Finally, the confusion resulting from the unwarranted variety of protective measures implemented after the Chernobyl accident has highlighted the need for international harmonization of the principles and scientific criteria applicable to the protection of the public in the event of an accident and for a more consistent approach to emergency plans. The international conventions on third party liability in the nuclear energy sector (Paris/Brussels Conventions and the Vienna Convention) provide for compensation for damage caused by nuclear accidents in accordance with the rules and jurisdiction that they lay down. These provisions impose obligations on the operator responsible for an accident, and the State where the nuclear facility is located, towards the victims of damage caused in another country

Causal Analysis to a Subway Accident: A Comparison of STAMP and RAIB

Directory of Open Access Journals (Sweden)

Zhou Yao

2018-01-01

Full Text Available Accident investigation and analysis after the accident, vital to prevent the occurrence of similar accident and improve the safety of the system. Different methods led to a different understanding of the accident. In this paper, a subway accident was analysed with a systemic accident analysis model – STAMP (System-Theoretic Accident Modelling and Processes. The hierarchical safety control structure was obtained, and the system-level safety constraints were obtained, controllers of the physical layer were analysed one by one, and put forward the relevant safety requirements and constraints, the dynamic analysis of the structure of the safety control is carried out, and the targeted recommendations are pointed out. In comparison with the analysis results obtained by the Rail Accident Investigation Branch (RAIB. Some useful findings have been concluded. STAMP treats safety as a control problem and reduces or eliminates causes of the accident from the controlling perspective. Whereas RAIB obtains causes of the accident by analysing the sequence of events related to the accident and reasons of these events, then chooses one(or moreevent(s as the immediate cause and some of the key events as causal factors. RAIB analysis is based on the sequential event models, but STAMP analysis provides us with a holistic, dynamic way to control system to maintain safety.

46 CFR 97.30-5 - Accidents to machinery.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Accidents to machinery. 97.30-5 Section 97.30-5 Shipping... Reports of Accidents, Repairs, and Unsafe Equipment § 97.30-5 Accidents to machinery. (a) In the event of an accident to a boiler, unfired pressure vessel, or machinery tending to render the further use of...

46 CFR 196.30-5 - Accidents to machinery.

Science.gov (United States)

2010-10-01

... 46 Shipping 7 2010-10-01 2010-10-01 false Accidents to machinery. 196.30-5 Section 196.30-5... Reports of Accidents, Repairs, and Unsafe Equipment § 196.30-5 Accidents to machinery. (a) In the event of an accident to a boiler, unfired pressure vessel, or machinery tending to render the further use of...

The Chernobyl accident

International Nuclear Information System (INIS)

Berg, J.O.; Christensen, G.; Lingjaerde, R.; Smidt Olsen, H.; Wethe, P.I.

1986-10-01

In connection with the Chernobyl accident the report gives a description of the technical features of importance to the accident, the course of events, and the estimated health hazards in the local environment. Dissimilarities in western and Sovjet reactor safety philosophy are dealt with, as well as conceivable concequences in relation to technology and research in western nuclear power programmes. Results of activity level measurements of air and foodstuff, made in Norway by Institute for Energy Technology, are given

Analysis on the nitrogen drilling accident of Well Qionglai 1 (II: Restoration of the accident process and lessons learned

Directory of Open Access Journals (Sweden)

Yingfeng Meng

2015-12-01

Full Text Available All the important events of the accident of nitrogen drilling of Well Qionglai 1 have been speculated and analyzed in the paper I. In this paper II, based on the investigating information, the well log data and some calculating and simulating results, according to the analysis method of the fault tree of safe engineering, the every possible compositions, their possibilities and time schedule of the events of the accident of Well Qionglai 1 have been analyzed, the implications of the logging data have been revealed, the process of the accident of Well Qionglai 1 has been restored. Some important understandings have been obtained: the objective causes of the accident is the rock burst and the induced events form rock burst, the subjective cause of the accident is that the blooie pipe could not bear the flow burden of the clasts from rock burst and was blocked by the clasts. The blocking of blooie pipe caused high pressure in wellhead, the high pressure made the blooie pipe burst, natural gas came out and flared fire. This paper also thinks that the rock burst in gas drilling in fractured tight sandstone gas zone is objective and not avoidable, but the accidents induced from rock burst can be avoidable by improving the performance of the blooie pipe, wellhead assemblies and drilling tool accessories aiming at the downhole rock burst.

A defense in depth approach for nuclear power plant accident management

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

Accident management information needs

International Nuclear Information System (INIS)

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

1990-04-01

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

Accident management information needs

Energy Technology Data Exchange (ETDEWEB)

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

1990-04-01

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

Principles for establishing intervention levels for the protection of the public in the event of a nuclear accident or radiological emergency

International Nuclear Information System (INIS)

1985-01-01

This Safety Guide is based on the report of an Advisory Group which met in Vienna in October 1984 in order to develop guidance on the radiation protection principles concerning emergency response planning and the establishment of intervention levels to be applied for the protection of the public in the event of a nuclear accident or radiological emergency. It considers the relationship between emergency response planning and various accident sequences, examines the pathways for radiation exposure and the sources of advice to decision makers during each of the three main accident phases, and specifies the dosimetric quantities that apply. The relevant pathological effects that must be protected against are summarized and the measures that may need to be implemented to provide protection with respect to each of the exposure pathways are discussed. It sets out the principles which underline decisions on intervention planning for each of the accident phases, gives guidance on dose values for the introduction of relevant protective measures and considers the application of cost-benefit analysis and the determination of the optimum dose level at which to withdraw protective measures

The Chernobyl accidents: Causes and Consequences

International Nuclear Information System (INIS)

Chihab-Eddine, A.

1988-01-01

The objective of this communication is to discuss the causes and the consequences of the Chernobyl accident. To facilitate the understanding of the events that led to the accident, the author gave a simplified introduction to the important physics that goes on in a nuclear reactor and he presented a brief description and features of chernobyl reactor. The accident scenario and consequences have been presented. The common contribution factors that led to both Three Mile Island and Chernobyl accidents have been pointed out.(author)

Persistence of airline accidents.

Science.gov (United States)

Barros, Carlos Pestana; Faria, Joao Ricardo; Gil-Alana, Luis Alberiko

2010-10-01

This paper expands on air travel accident research by examining the relationship between air travel accidents and airline traffic or volume in the period from 1927-2006. The theoretical model is based on a representative airline company that aims to maximise its profits, and it utilises a fractional integration approach in order to determine whether there is a persistent pattern over time with respect to air accidents and air traffic. Furthermore, the paper analyses how airline accidents are related to traffic using a fractional cointegration approach. It finds that airline accidents are persistent and that a (non-stationary) fractional cointegration relationship exists between total airline accidents and airline passengers, airline miles and airline revenues, with shocks that affect the long-run equilibrium disappearing in the very long term. Moreover, this relation is negative, which might be due to the fact that air travel is becoming safer and there is greater competition in the airline industry. Policy implications are derived for countering accident events, based on competition and regulation. © 2010 The Author(s). Journal compilation © Overseas Development Institute, 2010.

ANS severe accident program overview & planning document

Energy Technology Data Exchange (ETDEWEB)

Taleyarkhan, R.P.

1995-09-01

The Advanced Neutron Source (ANS) severe accident document was developed to provide a concise and coherent mechanism for presenting the ANS SAP goals, a strategy satisfying these goals, a succinct summary of the work done to date, and what needs to be done in the future to ensure timely licensability. Guidance was received from various bodies [viz., panel members of the ANS severe accident workshop and safety review committee, Department of Energy (DOE) orders, Nuclear Regulatory Commission (NRC) requirements for ALWRs and advanced reactors, ACRS comments, world-wide trends] were utilized to set up the ANS-relevant SAS goals and strategy. An in-containment worker protection goal was also set up to account for the routine experimenters and other workers within containment. The strategy for achieving the goals is centered upon closing the severe accident issues that have the potential for becoming certification issues when assessed against realistic bounding events. Realistic bounding events are defined as events with an occurrency frequency greater than 10{sup {minus}6}/y. Currently, based upon the level-1 probabilistic risk assessment studies, the realistic bounding events for application for issue closure are flow blockage of fuel element coolant channels, and rapid depressurization-related accidents.

Internal Accident Report: fill it out!

CERN Multimedia

2012-01-01

It is important to report all accidents, near-misses and dangerous situations so that they can be avoided in the future.   Reporting these events allows the relevant services to take appropriate action and implement corrective and preventive measures. It should be noted that the routing of the internal accident report was recently changed to make sure that the people who need to know are informed. Without information, corrective action is not possible. Without corrective action, there is a risk that the events will recur. As soon as you experience or see something amiss, fill out an internal accident report! If you have any questions the HSE Unit will be happy to answer them. Contact us at safety-general@cern.ch. The HSE Unit

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

Science.gov (United States)

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

2015-06-01

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

The nature of reactor accidents

International Nuclear Information System (INIS)

Domaratzki, Z.; Campbell, F.R.; Atchison, R.J.

1981-01-01

Reactor accidents are events which result in the release of radioactive material from a nuclear power plant due to the failure of one or more critical components of that plant. The failures, depending on their number and type, can result in releases whose consequences range from negligible to catastrophic. By way of examples, this paper describes four specific accidents which cover this range of consequence: failure of a reactor control system, loss of coolant, loss of coolant with impaired containment, and reactor core meltdown. For each a possible sequence of events and an estimate of the expected frequency are presented

Risk-based ranking of dominant contributors to maritime pollution events

International Nuclear Information System (INIS)

Wheeler, T.A.

1993-01-01

This report describes a conceptual approach for identifying dominant contributors to risk from maritime shipping of hazardous materials. Maritime transportation accidents are relatively common occurrences compared to more frequently analyzed contributors to public risk. Yet research on maritime safety and pollution incidents has not been guided by a systematic, risk-based approach. Maritime shipping accidents can be analyzed using event trees to group the accidents into 'bins,' or groups, of similar characteristics such as type of cargo, location of accident (e.g., harbor, inland waterway), type of accident (e.g., fire, collision, grounding), and size of release. The importance of specific types of events to each accident bin can be quantified. Then the overall importance of accident events to risk can be estimated by weighting the events' individual bin importance measures by the risk associated with each accident bin. 4 refs., 3 figs., 6 tabs

Millstone Unit 1 plant vulnerabilities during postulated severe nuclear accidents

International Nuclear Information System (INIS)

Khalil, Y.F.

1993-01-01

Generic Letter 88-20, Supplement No. 1 (Ref. 1), issued by the Nuclear Regulatory Commission (NRC) requested all licensees holding operating licenses and construction permits for nuclear power reactor facilities to perform Individual Plant Examinations (IPE) of their plant(s) for severe accident vulnerabilities and to submit the results to the Commission. This paper summarizes the major Front-End (Level-1 PRA) and Back-End (Level-2 PRA) insights gained from the Millstone Unit 1 (MP-1) IPE study. No major plant vulnerabilities have been identified from a Front-End perspective. The Back-End analysis, however, has identified two potential containment vulnerabilities during postulated events that progress beyond the Design Basis Accidents (DBAs), namely, (1) MP-1 is dominated by early source term releases that would occur within a six-hour time frame from time of accident initiation, or reactor trip, and (2) MP-1 containment is somewhat vulnerable to leak-type failure through the drywell head. As a result of the second finding, a recommendation currently under evaluation, has been made to increase the drywell head bolt's preload from 54 Kips to resist the containment design pressure value (62 psig)

Enhanced Accident Tolerant Fuels for LWRS - A Preliminary Systems Analysis

Energy Technology Data Exchange (ETDEWEB)

Gilles Youinou; R. Sonat Sen

2013-09-01

The severe accident at Fukushima Daiichi nuclear plants illustrates the need for continuous improvements through developing and implementing technologies that contribute to safe, reliable and cost-effective operation of the nuclear fleet. Development of enhanced accident tolerant fuel contributes to this effort. These fuels, in comparison with the standard zircaloy – UO2 system currently used by the LWR industry, should be designed such that they tolerate loss of active cooling in the core for a longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, and design-basis events. This report presents a preliminary systems analysis related to most of these concepts. The potential impacts of these innovative LWR fuels on the front-end of the fuel cycle, on the reactor operation and on the back-end of the fuel cycle are succinctly described without having the pretension of being exhaustive. Since the design of these various concepts is still a work in progress, this analysis can only be preliminary and could be updated as the designs converge on their respective final version.

The nuclear accidents: Causes and consequences

International Nuclear Information System (INIS)

Rochd, M.

1988-01-01

The author discussed and compared the real causes of T.M.I. and Chernobyl accidents and cited their consequences. To better understand how these accidents occurred, a brief description of PWR type (reactor type of T.M.I.) and of RBMK type (reactor type of Chernobyl) has been presented. The author has also set out briefly the safety analysis objectives and the three barriers established to protect the public against the radiological consequences. To distinguish failures that cause severe accidents and to analyze them in details, it is necessary to classify the accidents. There are many ways to do it according to their initiator event, or to their frequency, or to their degree of gravity. The safety criteria adopted by nuclear industry have been explained. These criteria specify the limits of certain physical parameters that should not be exceeded in case of incidents or accidents. To compare the real causes of T.M.I. and Chernobyl accidents, the events that led to both have been presented. As observed the main common contributing factors in both cases are that the operators did not pay attention to warnings and signals that were available to them and that they were not trained to handle these accident sequences. The essential conclusions derived from these severe accidents are: -The improvement of operators competence contribute to reduce the accident risks; -The rapid and correct diagnosis of real conditions at each point of the accidents permits an appropriate behavior that would bring the plant to a stable state; -Competent technical teams have to intervene and to assist the operators in case of emergency; -Emergency plans and an international collaboration are necessary to limit the accident risks. 11 figs. (author)

Analysis of reactivity accidents in PWR'S

International Nuclear Information System (INIS)

Camous, F.; Chesnel, A.

1989-12-01

This note describes the French strategy which has consisted, firstly, in examining all the accidents presented in the PWR unit safety reports in order to determine for each parameter the impact on accident consequences of varying the parameter considered, secondly in analyzing the provisions taken into account to restrict variation of this parameter to within an acceptable range and thirdly, in checking that the reliability of these provisions is compatible with the potential consequences of transgression of the authorized limits. Taking into consideration violations of technical operating specifications and/or non-observance of operating procedures, equipment failures, and partial or total unavailability of safety systems, these studies have shown that fuel mechanical strength limits can be reached but that the probability of occurrence of the corresponding events places them in the residual risk field and that it must, in fact, be remembered that there is a wide margin between the design basis accidents and accidents resulting in fuel destruction. However, during the coming year, we still have to analyze scenarios dealing with cumulated events or incidents leading to a reactivity accident. This program will be mainly concerned with the impact of the cases examined relating to dilution incidents under normal operating conditions or accident operating conditions

Analysis of accident progression in the TEPCO Fukushima Daiichi Nuclear Power Station

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

One of the objectives of this study is to investigate the early stage of the TEPCO Fukushima Daiichi accident and to check the validity of the countermeasures against the accident. Last year the early stage of the accident was analyzed with use of RELAP5 code, and the longer term analysis was done by MELCOR code. This year, the simulation of reactor water level instrumentation behavior by MELCOR code was performed. Another objective of this study is to analyze of the long term cooling after the Fukushima Daiichi accident by TRACE5 code. In order to simulate the cooling conditions in Fukushima plants after the accident, the parametric calculations were done on the assumption of the existence of steam/liquid leak in Reactor Pressure Vessel (RPV) and Pressure Containment Vessel (PCV) and the variety of debris distribution in RPV and PCV. As a result, the debris distribution in RPV and PCV was estimated by referring plant parameter such as reactor pressure and temperature. (author)

Safety in New Zealand's adventure tourism industry: the client accident experience of adventure tourism operators.

Science.gov (United States)

Bentley , T A; Page, S J; Laird, I S

2000-01-01

Injuries and fatalities among participants of adventure tourism activities have the potential to seriously impact on New Zealand's tourism industry. However, the absence of statistics for tourist accidents in New Zealand, and the lack of detailed academic research into adventure tourism safety, means the extent of the problem is unknown. The aims of the present study were to determine the incidence of client injuries across a range of adventure tourism activity sectors, and to identify common accident events and contributory risk factors. A postal questionnaire survey of New Zealand adventure tourism operators was used. Operators were asked to provide information related to their business; the number of recorded client injuries during the preceding 12 month period, January to December 1998; common accident and injury events associated with their activity; and perceived risk factors for accidents in their sector of the adventure tourism industry. The survey was responded to by 142 New Zealand adventure tourism operators. The operators' reported client injury experience suggests the incidence of serious client injuries is very low. Highest client injury incidence rates were found for activities that involved the risk of falling from a moving vehicle or animal (e.g., cycle tours, quad biking, horse riding, and white-water rafting). Slips, trips, and falls on the level were common accident events across most sectors of the industry. Perceived accident/incident causes were most commonly related to the client, and in particular, failure to attend to and follow instructions. The prevalence of client injuries in activity sectors not presently covered by government regulation, suggests policy makers should look again at extending codes of practice to a wider range of adventure tourism activities. Further research considering adventure tourism involvement in overseas visitor hospitalized injuries in New Zealand, is currently in progress. This will provide supporting evidence

Numerical methods operational at the French Meteorologie Nationale for nuclear accident situation

International Nuclear Information System (INIS)

Marais, C.; Musson-Genon, L.

1990-01-01

Since the Chernobyl accident, the Meteorologie Nationale has developed new numerical simulation methods to assist predictions provided as part of the meteorological support to the public authorities in the event of a nuclear accident. The present paper describes these new tools now operational at the Meteorologie Nationale. In the event of an accident, the first task of the forecaster is to anticipate the evolution of meteorological conditions at the site concerned. A fine scale, numerical forecasting model, PERIDOT, is used covering Western Europe with a resolution of 35 x 35 km. A comparison between PERIDOT wind forecasts and measurements at French NPS sites is presented which shows these forecasts to be of good overall quality, except for Chooz and Gravelines NPSs where the orographic complexity and the proximity of the sea require statistical corrections to be introduced. In all cases PERIDOT forecasts are clearly superior to those based on wind persistence. For accidents of any significance, the transport and dispersion of the atmopsheric polluants need to be evaluated as a matter of urgency. Again the forecaster has a vital role to play using numerical forecasting resources: in particular trajectory forecasts available by FAX within one hour of the meteorological Service Central d'Exploitation being alerted, and subsequently the Eulerian transport and diffusion code MEDIA which can be interfaced with either PERIDOT or EMERAUDE, a model operating on global meteorological conditions with a resolution of 150 x 150 km. This latter model has been tested against the Chernobyl accident with good results, the output is available in 4 to 5 hours after the alert and work is in hand to reduce the response time. Further studies are now in progress to provide a much finer regional resolution (5-10 km) and improved representation of wet and dry disposition at this resolution within MEDIA

Cooperation in the Event of Nuclear Accidents; Cooperation en Matiere d'Accidents Nucleaires

Energy Technology Data Exchange (ETDEWEB)

Bresson, G. [CEA, Centre d' etudes nucleaires, de Fontenay-aux-Roses (France)

1969-10-15

This paper is concerned only with the action to be taken in respect of an individual directly affected by an accident and not with the more general measures relating to the population as a whole. Keeping the same sequence of ideas, the paper deals with nuclear establishments and cites criteria for classifying them; hence only the relationship between the establishment and the hospital, and between the radiation protection experts and medical personnel, is discussed. The complex organization of emergency measures, reception of the victim of the accident, and the treatment possibly required should be based on standard practice and published material, both national and international, allowance being made for the characteristics of each sector. A ''flexible'' plan of co-ordination is given as an illustration. Action must be taken in such cases at the site of the accident, inside and outside the establishment, and above all at the hospital. All categories of persons are involved in the process, i.e. fellow-workers, management, specialized services, and medical personnel, each with their own part to play. The manpower and equipment brought into service therefore vary, and depend upon the internal and external relations maintained by the establishment. The measures envisaged should provide for the transport, reception and treatment of those involved in the accident. An existing organization of this kind is described as an illustration. Finally, no action can be of value without full knowledge of the facts and thorough training of the personnel. Some clearly defined ideas on the.subject are considered under this heading. (author) [French] Le memoire ne traite que de la conduite a tenir envers un accidente et non du probleme, plus general, des mesures relatives a une population. Dans le meme ordre d'idees, l'etude porte sur les etablissements nucleaires et leurs criteres de classement; il ne s'agit donc que des liaisons entre retablissement et l'hopital et entre les

Use of NUREG-1150 and IPEs in accident management

International Nuclear Information System (INIS)

Mauersberger

1992-01-01

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

ANS severe accident program overview ampersand planning document

International Nuclear Information System (INIS)

Taleyarkhan, R.P.

1995-09-01

The Advanced Neutron Source (ANS) severe accident document was developed to provide a concise and coherent mechanism for presenting the ANS SAP goals, a strategy satisfying these goals, a succinct summary of the work done to date, and what needs to be done in the future to ensure timely licensability. Guidance was received from various bodies [viz., panel members of the ANS severe accident workshop and safety review committee, Department of Energy (DOE) orders, Nuclear Regulatory Commission (NRC) requirements for ALWRs and advanced reactors, ACRS comments, world-wide trends] were utilized to set up the ANS-relevant SAS goals and strategy. An in-containment worker protection goal was also set up to account for the routine experimenters and other workers within containment. The strategy for achieving the goals is centered upon closing the severe accident issues that have the potential for becoming certification issues when assessed against realistic bounding events. Realistic bounding events are defined as events with an occurrency frequency greater than 10 -6 /y. Currently, based upon the level-1 probabilistic risk assessment studies, the realistic bounding events for application for issue closure are flow blockage of fuel element coolant channels, and rapid depressurization-related accidents

Organization of the French emergency teams in the event of a radiological accident

Energy Technology Data Exchange (ETDEWEB)

Dumon, F. [Faculte de Pharmacie, 13 - Marseille (France); Pizzocaro, Y. [CSP, Risques Technologiques, 83 - Toulon (France)

2001-07-01

Nowadays, the intervention in ionizing environment is increasingly probable. It is still rare, but with the development of the nuclear programme of electricity production which was held in the french past and the significant rise in the use of the radioelements, as well in the medical field as industrial, the radioactive risk cannot be neglected. Technical and human resources, brought by mobile emergency teams called CMIR, were thus implemented to ensure either the safety of only hard-working exposed to the ionizing radiations, but also that of the population. In France, the organization of the public authorities in the event of nuclear accident, fixed by Directives of the Prime Minister which relate to nuclear safety, protection against radiation, the law and order and the civil safety, is described in Particular Plan for Intervention (PPI). (author)

Organization of the French emergency teams in the event of a radiological accident

International Nuclear Information System (INIS)

Dumon, F.; Pizzocaro, Y.

2001-01-01

Nowadays, the intervention in ionizing environment is increasingly probable. It is still rare, but with the development of the nuclear programme of electricity production which was held in the french past and the significant rise in the use of the radioelements, as well in the medical field as industrial, the radioactive risk cannot be neglected. Technical and human resources, brought by mobile emergency teams called CMIR, were thus implemented to ensure either the safety of only hard-working exposed to the ionizing radiations, but also that of the population. In France, the organization of the public authorities in the event of nuclear accident, fixed by Directives of the Prime Minister which relate to nuclear safety, protection against radiation, the law and order and the civil safety, is described in Particular Plan for Intervention (PPI). (author)

Reactor accidents of four decades

International Nuclear Information System (INIS)

Szabo, Z.

1982-11-01

The report covers the period between 1942 and June 30, 1982. A detailed description and a comparative analysis of reactor accidents and chemical-processing-plant excursions are presented. The analysis takes into account the following points: causes (design, maintenance, operation); events (initiating event and sequence of events); consequences (environmental impacts, personnel effects and equipment damages). (author)

HTGR accident initiation and progression analysis status report. Volume 1. Introduction and summary

International Nuclear Information System (INIS)

Raabe, P.H.; Houghton, W.J.; Joksimovic, V.

1976-01-01

Probabilistic risk assessment techniques have been applied to obtain guidance in choosing nuclear safety research and development that is most worthwhile for high-temperature gas-cooled reactor (HTGR) nuclear power plants. The probabilistic techniques used are similar to those employed in the Reactor Safety Study for light water reactors (LWRs), WASH-1400, directed by Dr. N. C. Rasmussen. The recommendations for research include studies related to core heatup even though this event poses a very low risk to the public. In fact, it was found that under the many conditions covered by the study to date, even very infrequent accidents in HTGRs (say, once in ten million years) will not produce fatalities. Potential cost reduction areas have been found where alternate design options protect the public and meet regulatory safety criteria

Severe Accident Research Network (SARNET). Level 2 PSA work package: comparison of partners methods for uncertainties assessment

International Nuclear Information System (INIS)

Chaumont, B.; Haesendonck, M.; Vidal, S.; Eyink, J.; Loeffler, H.; Radu, G.; Kopustinskas, V.; Ming, A.; Guntay, S.; Gustavsson, V.; Ivanov, I.; Dienstbier, J.; Bareith, A.; Hollo, E.; Lajtha, G.

2007-01-01

The PSA2 work package (PSA2 WP) is a part of the Joined Programme Activity of the European Severe Accident Network (SARNET) related to level 2 PSA methodologies. The general objectives of this work package is to provide a comparison of the different methodologies used or under development for level 2 PSA application by the partners involved in the work package and to promote their harmonization. The PSA2 WP is organized into three main topics: methodologies in general, methodologies for uncertainties assessment, and dynamic reliability methods. The different tasks initially defined for these three topics are shortly described and the partners involved identified. Attention is then paid on the methodologies used so far by the different partners to assess the uncertainties in their level 2 PSA. A review of partners approaches to assess - as far as possible - the different sources of possible uncertainties is done for the different following topics: - uncertainties propagated from the level 1 PSA, - uncertainties (in sense of approximation) due to the binning of the level 1 sequences in Plant Damage, - uncertainties related to the structure of the Accident Progression Event Tree, - uncertainties related to the probabilities of stochastic events (system failure or recovery, human actions, some physical phenomena such as ignition of hydrogen combustion or triggering of steam explosion), - uncertainties elated to the modelling of the different physical phenomena, - uncertainties related to the cut-off frequency used in the probabilistic quantification of the Accident Progression Event Tree; - uncertainties related to the binning of level 2 sequences in Release Categories (variables not considered, values of eventual continuous variables). First conclusions of the comparison are given in terms of improvement needs and then of perspectives of the work for the following period of work. (authors)

The 1986 Chernobyl accident; Der Unfall von Tschernobyl 1986

Energy Technology Data Exchange (ETDEWEB)

Kerner, Alexander; Stueck, Reinhard; Weiss, Frank-Peter [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Garching bei Muenchen, Koeln (Germany). Bereich Reaktorsicherheitsanalysen; Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Koeln (Germany)

2011-02-15

April 26, 2011 marks the 25th anniversary of the Chernobyl reactor accident, the worst incident in the history of the peaceful utilization of nuclear power. While investigations of the course of events and the causes of the accident largely present a uniform picture, descriptions still vary widely when it comes to the impact on the population and the environment. This treatment of the Chernobyl accident constitutes a summary of facts about the initiation of the accident and the sequence of events that followed. In addition, measures are described which were taken to exclude any repetition of a disaster of this kind. The health consequences and the socio-economic impact of the accident are not discussed in any detail. The first section contains an introduction and an overview of the Soviet RBMK (Chernobyl) reactor line. In section 2, fundamental characteristics of this special type of reactor, which was exclusively built in the former Soviet Union, are discussed. This information is necessary to understand the sequence of accident events and provides an answer to the frequent question whether that accident could be transferred to reactors in this country. The third section outlines the history of the accident caused ultimately by a commissioning test never performed before. The section is completed by a brief description of radiological releases and the state of the plant after the accident when entombed in the ''sarcophagus.'' The different causes are then summarized and the modifications afterwards made to RBMK reactors are outlined. (orig.)

Report from the Special Committee on Fukushima Nuclear Accident

International Nuclear Information System (INIS)

Ozawa, Mamoru

2012-01-01

The Special Committee on Fukushima Nuclear Accident was established in April 2011 under the Heat Transfer Society of Japan (HTSJ) and discussed (1) how had evolved heat transfer research in progress of nuclear technology, (2) role of expert group in the area of heat transfer academy and technology and (3) energy prospect in Japan after the Fukushima nuclear accident. This report was described by the chairman of the special committee summarizing one year discussions as (1) background of heat transfer research progress, (2) progression of Fukushima Daiichi Nuclear Power Plant accident, (3) energy problem in Japan after the Fukushima accident and (4) social role of the HTSJ. This HTSJ was a unique, nonprofit association in Japan of the people engaged in heat transfers research or in various engineering aspects related to heat transfer, which meant interdisciplinary or common platform of heat transfer as elementary technologies. Such actual complex problems could be discussed in the HTSJ from an overlooking viewpoint in order for the HTSJ to play a social role. (T. Tanaka)

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-07-01

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

Identification of NPP accidents using support vector classification

Energy Technology Data Exchange (ETDEWEB)

Back, Ju Hyun; Yoo, Kwae Hwan; Na, Man Gyun [Chosun University, Gwangju (Korea, Republic of)

2016-10-15

In case of the accidents that happens in a nuclear power plants (NPPs), it is very important to identify its accidents for the operator. Therefore, in order to effectively manage the accidents, the initial short time trends of major parameters have to be observed and NPP accidents have to accurately be identified to provide its information to operators and technicians. In this regard, the objective of this study is to identify the accidents when the accidents happen in NPPs. In this study, we applied the support vector classification (SVC) model to classify the initiating events of critical accidents such as loss of coolant accidents (LOCA), total loss of feedwater (TLOFW), station blackout (SBO), and steam generator tube rupture (SGTR). Input variables were used as the initial integral value of the signal measured in the reactor coolant system (RCS), steam generator, and containment vessel after reactor trip. The proposed SVC model is verified by using the simulation data of the modular accident analysis program (MAAP4) code. In this study, the proposed SVC model is verified by using the simulation data of the modular accident analysis program (MAAP4) code. We used an initial integral value of the simulated sensor signals to identify the NPP accidents. The training data was used to train the SVC model. And, the trained model was confirmed using the test data. As a result, it was known that it can accurately classify five events.

Accomplishments and challenges of the severe accident research

International Nuclear Information System (INIS)

Sehgal, B.R.

2001-01-01

This paper briefly describes the progress of the severe accident research since 1980, in terms of the accomplishments made so far and the challenges that remain. Much has been accomplished: many important safety issues have been resolved and consensus is near on some others. However, some of the previously identified safety issues remain as challenges, while some new ones have arisen due to the shift in focus from containment to vessel integrity. New reactor designs have also created some new challenges. In general, the regulatory demands for new reactor designs are stricter, thereby requiring much greater attention to the safety issues concerned with the containment design of the new large reactors, and to the accident management procedures for mitigating the consequences of a severe accident. We apologize for not providing references to many fine investigations that contributed to the great progress made so far in the severe accident research

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

International Nuclear Information System (INIS)

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

1992-01-01

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

Radiological accidents: education for prevention and confrontation

International Nuclear Information System (INIS)

Cardenas Herrera, Juan; Fernandez Gomez, Isis Maria

2008-01-01

The purpose of this work is to train and inform on radiological accidents as a preventive measure to improve the people life quality. Radiological accidents are part of the events of technological origin which are composed of nuclear and radiological accidents. As a notable figure is determined that there have been 423 radiological accidents from 1944 to 2005 and among the causes prevail industrial accidents, by irradiations, medical accidents and of laboratories, among others. Latin American countries such as Argentina, Brazil, Mexico and Peru are some where most accidents have occurred by radioactivity. The radiological accidents can have sociological, environmental, economic, social and political consequences. In addition, there are scenarios of potential nuclear accidents and in them the potential human consequences. Also, the importance of the organization and planning in a nuclear emergency is highlighted. Finally, the experience that Cuba has lived on the subject of radiological accidents is described [es

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

Note on the stock market's reaction to the accident at Three Mile Island

International Nuclear Information System (INIS)

Spudeck, R.E.; Moyer, C.R.

1989-01-01

This note provides new information regarding the market reaction toward electric utility stocks that resulted both from the accident at Three Mile Island, and the events predating and postdating the accident. The results suggest that some of the market reaction heretofore ascribed to the accident resulted instead from regulatory activity occurring before the accident. We also provide results suggesting that regulatory activity by the Pennsylvania Public Utilities Commission in the wake of the accident served to offset a majority of the increased systematic risk resulting from the accident. Our results imply that previously reported lingering effects of the accident at Three Mile Island may be regulatory effects from events predating the accident

Event course analysis of core disruptive accidents; Ereignisablaufanalyse kernzerstoerender Unfaelle

Energy Technology Data Exchange (ETDEWEB)

Hering, W.; Homann, C.; Sengpiel, W.; Struwe, D.; Messainguiral, C.

1995-08-01

The theortical studies of the behavior of a PWR core in a meltdown accident are focused on hydrogen release, materials redistribution in the core area including forming of an oxide melt pool, quantity of melt and its composition, and temperatures attained by the RPV internals (esp. in the upper plenum) during the accident up to the time of melt relocation into the lower plenum. The calculations are done by the SCDAP/RELAP5 code. For its validation selected CORA results and Phebus FPTO results have been used. (orig.)

A study on the operator's errors of commission (EOC) in accident scenarios of nuclear power plants: methodology development and application

International Nuclear Information System (INIS)

Kim, Jae Whan; Jung, Won Dea; Park, Jin Kyun; Kang, Da Il

2003-04-01

As the concern on the operator's inappropriate interventions, the so-called Errors Of Commission (EOCs), that can exacerbate the plant safety has been raised, much of interest in the identification and analysis of EOC events from the risk assessment perspective has been increased. Also, one of the items in need of improvement for the conventional PSA and HRA that consider only the system-demanding human actions is the inclusion of the operator's EOC events into the PSA model. In this study, we propose a methodology for identifying and analysing human errors of commission that might be occurring from the failures in situation assessment and decision making during accident progressions given an initiating event. In order to achieve this goal, the following research items have been performed: Firstly, we analysed the error causes or situations contributed to the occurrence of EOCs in several incidents/accidents of nuclear power plants. Secondly, limitations of the advanced HRAs in treating EOCs were reviewed, and a requirement for a new methodology for analysing EOCs was established. Thirdly, based on these accomplishments a methodology for identifying and analysing EOC events inducible from the failures in situation assessment and decision making was proposed and applied to all the accident sequences of YGN 3 and 4 NPP which resulted in the identification of about 10 EOC situations

Analysis of unmitigated large break loss of coolant accidents using MELCOR code

Science.gov (United States)

Pescarini, M.; Mascari, F.; Mostacci, D.; De Rosa, F.; Lombardo, C.; Giannetti, F.

2017-11-01

In the framework of severe accident research activity developed by ENEA, a MELCOR nodalization of a generic Pressurized Water Reactor of 900 MWe has been developed. The aim of this paper is to present the analysis of MELCOR code calculations concerning two independent unmitigated large break loss of coolant accident transients, occurring in the cited type of reactor. In particular, the analysis and comparison between the transients initiated by an unmitigated double-ended cold leg rupture and an unmitigated double-ended hot leg rupture in the loop 1 of the primary cooling system is presented herein. This activity has been performed focusing specifically on the in-vessel phenomenology that characterizes this kind of accidents. The analysis of the thermal-hydraulic transient phenomena and the core degradation phenomena is therefore here presented. The analysis of the calculated data shows the capability of the code to reproduce the phenomena typical of these transients and permits their phenomenological study. A first sequence of main events is here presented and shows that the cold leg break transient results faster than the hot leg break transient because of the position of the break. Further analyses are in progress to quantitatively assess the results of the code nodalization for accident management strategy definition and fission product source term evaluation.

Cerebrovascular Accidents Associated with Sorafenib in Hepatocellular Carcinoma

OpenAIRE

Saif, Muhammad W.; Isufi, Iris; Peccerillo, Jennifer; Syrigos, Kostas N.

2011-01-01

Sorafenib is an oral angiogenetic multikinase inhibitor approved in the treatment of renal and hepatocellular carcinoma. Bleeding and venous thrombotic events have been described with angiogenetic agents but cerebrovascular accidents are rarely reported. We report two cases of patients with hepatocellular carcinoma who developed a cerebrovascular accident while on sorafenib. Neither patient had any risk factors for the cerebrovascular events apart from gender and age in the second patient. La...

Derivation of working levels for animal feedstuffs for use in the event of a future nuclear accident

International Nuclear Information System (INIS)

Nisbet, A.; Woodman, R.; Brown, J.

1998-04-01

In the event of a future nuclear accident, European Council Food Intervention Levels (CFILs) would be legally binding for foodstuffs marketed in the UK. Practical guidance has been developed on the activity concentrations of radiocaesium and radiostrontium in animal feedstuffs that would give rise to concentrations equivalent to the relevant CFIL in the final animal product. The animals considered were dairy and beef cattle, lambs, pigs, broiler chickens and laying hens. Typical diets have been derived for each animal. The NRPB foodchain model FARMLAND has been used to predict activity concentrations in different feedstuffs for accidents occurring at different times of the year. The predicted concentrations were combined with the data on dietary composition, information on feed-to-product transfer and the relevant CFIL to estimate the corresponding Working levels in Animal Feedstuffs (WAFs). The calculations were carried out using a dedicated software system called SILAFOD. This flexible system can be used to carry out more specific assessments. A handbook that accompanies this report contains detailed information on animal diets, contributions from various feedstuffs to intakes of activity and the corresponding WAFs. The early phase after an accident and the longer-term phase are both considered. The work received partial financial support from the Ministry of Agriculture, Fisheries and Food, Radiological Safety and Nutrition Division. (author)

Analysis of accidents at the LPR (Radiochemical Processes Laboratory)

International Nuclear Information System (INIS)

Kaufmann, F.; Boutet, L.I.

1987-01-01

Accidents are defined as not planned events that may result in the emission of significative quantities of radioactive materials to the environment. The pilot plant has been specifically designed to prevent this type of accidents but there still exists the possibility that one or more accidents can be produced during the plant life. In a first phase, the emission of radionuclides to the environment were evaluated for 13 credible accidents. In a second phase, by means of the calculation program SEDA, specially adapted to this purpose, the critical doses of critical group were calculated for each accident. Due to the small capacity of the pilot plant and the long cooling period of treated fuel, it is concluded that the radiological consequences for the external environment are of very small magnitude. In this way, without need of developing complex fault- or event-trees, it is shown that any of the accidents falls into the non acceptable zone of Farmer diagram. (Author)

Study on severe fuel damage and in-vessel melt progression

International Nuclear Information System (INIS)

Kim, Hee Dong; Kim, Sang Baik; Lee, Gyu Jung

1992-06-01

In-vessel core melt progression describes the progression of the state of a reactor core from core uncovery up to reactor vessel melt through in uncovered accidents or through temperature stabilization in accidents recovered by core reflooding. Melt progression can be thought as two parts; early melt progression and late melt progression. Early phase of core melt progression includes the progression of core material melting and relocation, which mostly consist of metallic materials. On the other hand, the late phase of core melt progression involves ceramic material melt and relocation to the lower plenum and heat-up the reactor vessel lower head. A large number of information are available for the early melt progression through experiments such as SFD, DF, FLHT test and utilized in the severe accident analysis codes. However, understanding of the late phase melt progression phenomenology is based primary on TMI-2 core examinations and not much experimental information is available. Especilally, the great uncertainties exist in vessel failure mode, melt composition, mass, and temperature. Further research is planned to perform to reduce the uncertainties in understanding of core melt down accidents as parts of long term melt progression research program. A study on the core melt progression at KAERI has been being performed through the Severe Accident Research Program with USNRC. KAERI staff had participated in the PBF SFD experiments at INEL and analyses of experiments were performed using SCDAP code. Experiments of core melt program have not been carried out at KAERI yet. It is planned that further research on core melt down accidents will be performed, which is related to design of future generations of nuclear reactors as parts of long-term project for improvement of nuclear reactor safety. (Author)

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

International Nuclear Information System (INIS)

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

2010-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-03-01

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

Deterministic analyses of severe accident issues

International Nuclear Information System (INIS)

Dua, S.S.; Moody, F.J.; Muralidharan, R.; Claassen, L.B.

2004-01-01

Severe accidents in light water reactors involve complex physical phenomena. In the past there has been a heavy reliance on simple assumptions regarding physical phenomena alongside of probability methods to evaluate risks associated with severe accidents. Recently GE has developed realistic methodologies that permit deterministic evaluations of severe accident progression and of some of the associated phenomena in the case of Boiling Water Reactors (BWRs). These deterministic analyses indicate that with appropriate system modifications, and operator actions, core damage can be prevented in most cases. Furthermore, in cases where core-melt is postulated, containment failure can either be prevented or significantly delayed to allow sufficient time for recovery actions to mitigate severe accidents

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

International Nuclear Information System (INIS)

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

1989-01-01

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

Iodine prophylaxis following nuclear accidents - a concept how to distribute potassium-iodide tablets out of the central stocks in the event of an accident

International Nuclear Information System (INIS)

Portius, U.

2007-01-01

With its recommendation ''Iodine prophylaxis following nuclear accidents'' (1996) and its reports of 1997 and 2001 the German Commission on Radiological Protection (SSK) followed the recommendations of the WHO ''Guidelines for iodine prophylaxis following nuclear accidents'' of 1989. The intervention levels were lowered (50 mSv for children/adolescents (up to the age of 18 years) and pregnant women, 250 mSv for adults), the iodine prophylaxis was restricted to persons up to the age of 45 years and the recommended dosage of stable iodine was changed. Due to the lowered reference levels the radius of 25 km around a nuclear power plant that had been the planning radius for the distribution of iodine tablets so far was extended to 100 km. Based on these recommendations the German authorities began to set up new strategies for the provision and distribution of potassium-iodide tablets (iodine tablets). Since 2004, within the radius of 25 km the iodine tablets are pre-distributed to households and/or stored at several points in the municipality for persons up to the age of 45 years. For the new planning radius of 25-100 km iodine tablets are stored in 8 central stocks in Germany for children/adolescents (up to the age of 18 years) and pregnant women. A working group with representatives from federal and Laender authorities has developed a distribution strategy for the distribution out of these central stocks in the event of an accident. It describes a possibility of organising and implementing the distribution of the iodine tablets within the radius of 25-100 km in a nationwide standardised way. (orig.)

Tchernobyl: a severe accident and its image

International Nuclear Information System (INIS)

Strazzulla, J.

1996-01-01

This paper gives a strong criticism about the false informations that were disseminated by the mass media immediately after the Tchernobyl accident. This accident is taken as an example to illustrate a common attitude in journalistic comments of geopolitical events. (J.S.). 1 photo

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)

ANALYSIS OF LABOUR ACCIDENTS DUE TO ROCK FALL EVENTS IN CUTTING FACE OF TUNNEL AND STUDY OF THE COUNTERMEASURES FOR SAFETY

Science.gov (United States)

Kikkawa, Naotaka; Itoh, Kazuya; Hori, Tomohito; Tamate, Satoshi; Toyosawa, Yasuo

In this paper, we analysed the labour accidents which had casualties due to rock fall events in the headings of tunnel and cleared the condition of the occurrence. It was clearly revealed that the accidents mostly happened when the workers mounted the explosive and the steel arch in the headings of the mountain tunnel. In addition, the dimension of the rocks fallen were averagely 0.6m diameter, it was not so much large. Therefore, the countermeasures based on both soft and hard faces would be useful and effective, such as the displacement measurement of a cutting face of tunnel, securing the sufficient lights to observe the cutting face, boring for drainage and shotcreting in a heading of tunnel.

Analysis of system and of course of events

International Nuclear Information System (INIS)

Hoertner, H.; Kersting, E.J.; Puetter, B.M.

1986-01-01

The analysis of the system and of the course of events is used to determine the frequency of core melt-out accidents and to describe the safety-related boundary conditions of appropriate accidents. The lecture is concerned with the effect of system changes in the reference plant and the effect of triggering events not assessed in detail or not sufficiently assessed in detail in phase A of the German Risk Study on the frequency of core melt-out accidents, the minimum requirements for system functions for controlling triggering events, i.e. to prevent core melt-out accidents, the reliability data important for reliability investigations and frequency assessments. (orig./DG) [de

Safety improvements at Canadian nuclear power plants in the aftermath of Fukushima accident

International Nuclear Information System (INIS)

Rzentkowski, G.; Khouaja, H.

2014-01-01

This paper describes the safety review of operating nuclear power plants undertaken by the Canadian Nuclear Safety Commission in light of the March 11, 2011 accident at the Fukushima Daiichi Nuclear Power Plants (NPPs). The review confirmed that the Canadian NPPs are robust and have a strong design relying on multiple layers of defence to protect the public from credible external events. Nevertheless, in the spirit of continuous safety improvements, the review identified a number of recommendations to further strengthen reactor defence-in-depth in preventing and mitigating the consequences of beyond design basis accidents, enhance onsite and offsite emergency response, and improve the CNSC regulatory framework. Progress achieved to date, in implementing these measures, is described in this paper along with a summary of safety benefits for each level of the reactor defence-in-depth. (author)

Safety improvements at Canadian nuclear power plants in the aftermath of Fukushima accident

Energy Technology Data Exchange (ETDEWEB)

Rzentkowski, G.; Khouaja, H. [Canadian Nuclear Safety Commission, Ottawa, ON (Canada)

2014-07-01

This paper describes the safety review of operating nuclear power plants undertaken by the Canadian Nuclear Safety Commission in light of the March 11, 2011 accident at the Fukushima Daiichi Nuclear Power Plants (NPPs). The review confirmed that the Canadian NPPs are robust and have a strong design relying on multiple layers of defence to protect the public from credible external events. Nevertheless, in the spirit of continuous safety improvements, the review identified a number of recommendations to further strengthen reactor defence-in-depth in preventing and mitigating the consequences of beyond design basis accidents, enhance onsite and offsite emergency response, and improve the CNSC regulatory framework. Progress achieved to date, in implementing these measures, is described in this paper along with a summary of safety benefits for each level of the reactor defence-in-depth. (author)

An analysis of the Three Mile Island accident

International Nuclear Information System (INIS)

Brooks, G.L.; Siddal, E.

1980-09-01

Starting with a systematic analysis of the chain of events that took place during the Three Mile Island accident, the authors assess the significance of the four distinct phases of the accident. Inferences that can be drawn with respect to the safety of CANDU reactors are discussed. A rational reaction to the accident is suggested, and several factors are shown not to have played an important part, contrary to public impressions. The authors point out that over-reaction to the accident could detract from public safety. The Canadian response to the accident is discussed. (auth)

Accident analysis for nuclear power plants

International Nuclear Information System (INIS)

2002-01-01

Deterministic safety analysis (frequently referred to as accident analysis) is an important tool for confirming the adequacy and efficiency of provisions within the defence in depth concept for the safety of nuclear power plants (NPPs). Owing to the close interrelation between accident analysis and safety, an analysis that lacks consistency, is incomplete or is of poor quality is considered a safety issue for a given NPP. Developing IAEA guidance documents for accident analysis is thus an important step towards resolving this issue. Requirements and guidelines pertaining to the scope and content of accident analysis have, in the past, been partially described in various IAEA documents. Several guidelines relevant to WWER and RBMK type reactors have been developed within the IAEA Extrabudgetary Programme on the Safety of WWER and RBMK NPPs. To a certain extent, accident analysis is also covered in several documents of the revised NUSS series, for example, in the Safety Requirements on Safety of Nuclear Power Plants: Design (NS-R-1) and in the Safety Guide on Safety Assessment and Verification for Nuclear Power Plants (NS-G-1.2). Consistent with these documents, the IAEA has developed the present Safety Report on Accident Analysis for Nuclear Power Plants. Many experts have contributed to the development of this Safety Report. Besides several consultants meetings, comments were collected from more than fifty selected organizations. The report was also reviewed at the IAEA Technical Committee Meeting on Accident Analysis held in Vienna from 30 August to 3 September 1999. The present IAEA Safety Report is aimed at providing practical guidance for performing accident analyses. The guidance is based on present good practice worldwide. The report covers all the steps required to perform accident analyses, i.e. selection of initiating events and acceptance criteria, selection of computer codes and modelling assumptions, preparation of input data and presentation of the

Perspectives on the economic risks of LWR accidents

International Nuclear Information System (INIS)

Ritchie, L.T.; Burke, R.P.

1986-01-01

Models which can be used for the analysis of the economic risks from events which may occur during LWR operation have been developed. The models include capabilities to estimate both onsite and offsite costs of LWR events ranging from routine plant forced outages to severe core-melt accidents resulting in large releases of radioactive material to the environment. The economic consequence models have been applied in studies of the economic risks from the operation of US LWR plants. The results of the analyses provide some important perspectives regarding the economic risks of LWR accidents. The analyses indicate that economic risks, in contrast to public health risks, are dominated by the onsite costs of relatively high-frequency forced outage events. Even for severe (e.g., core-melt) accidents, expected offsite costs are less than expected onsite costs for a typical US plant

Accident scenarios triggered by lightning strike on atmospheric storage tanks

International Nuclear Information System (INIS)

Necci, Amos; Argenti, Francesca; Landucci, Gabriele; Cozzani, Valerio

2014-01-01

Severe Natech accidents may be triggered by lightning strike affecting storage tanks containing relevant inventories of hazardous materials. The present study focused on the identification of event sequences and accident scenarios following lightning impact on atmospheric tanks. Reference event trees, validated using past accident analysis, are provided to describe the specific accident chains identified, accounting for reference protection and mitigation safety barriers usually adopted in current industrial practice. An overall methodology was outlined to allow the calculation of the expected frequencies of final scenarios following lightning impact on atmospheric storage tanks, taking into account the expected performance of available safety barriers. The methodology was applied to a case study in order to better understand the data that may be obtained and their importance in the framework of quantitative risk assessment (QRA) and of the risk management of industrial facilities with respect to external hazards due to natural events. - Highlights: • Event sequences following lightning impact on atmospheric tanks were identified. • Reference event trees including standard safety barriers were obtained. • Safety barriers applied in industrial practice were assessed to quantify event trees. • Frequencies of final scenarios following lightning impact on tanks were calculated. • Natech scenarios caused by lightning have an important influence on risk profiles

Extending the application range of a fuel performance code from normal operating to design basis accident conditions

International Nuclear Information System (INIS)

Van Uffelen, P.; Gyori, C.; Schubert, A.; Laar, J. van de; Hozer, Z.; Spykman, G.

2008-01-01

Two types of fuel performance codes are generally being applied, corresponding to the normal operating conditions and the design basis accident conditions, respectively. In order to simplify the code management and the interface between the codes, and to take advantage of the hardware progress it is favourable to generate a code that can cope with both conditions. In the first part of the present paper, we discuss the needs for creating such a code. The second part of the paper describes an example of model developments carried out by various members of the TRANSURANUS user group for coping with a loss of coolant accident (LOCA). In the third part, the validation of the extended fuel performance code is presented for LOCA conditions, whereas the last section summarises the present status and indicates needs for further developments to enable the code to deal with reactivity initiated accident (RIA) events

Situation at TMI unit 2 after the accident of March 28, 1979. Radiation protection aspects

International Nuclear Information System (INIS)

Chevalier, C.

1983-01-01

After a brief flashback on the chronology of the accident that occurred on TMI unit 2, the author, who took part in the first operations of diagnosis on the spot, gives an outlook of the situation at the end of 1981. The damages are considerable, and, in spite of the progression of decontamination works, the obstacles, mainly administrative and financial, do not allow, four years after the event, to give a serious forecast on the repairing of the damaged reactor [fr

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

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

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.

Safety climate and accidents at work

DEFF Research Database (Denmark)

Ajslev, Jeppe; Dastjerdi, Efat Lali; Dyreborg, Johnny

2017-01-01

Aim: Occupational safety climate is utilized as a way to measure the risk of accidents and injuries at work. This study investigates which factors are associated with safety climate and accidents at work. Methods: In the 2012 round of the Danish Work Environment and Health Study, 15,144 workers...... from the general working population of Denmark replied to questions about safety climate and accidents at work. Mutually adjusted logistic regression analyses determined the association between variables. Results: Within the last year, 5.7% had experienced an accident resulting in sickness absence....... The number of safety climate problems was progressively associated with the odds ratio (OR) for accidents. For one safety climate problem the OR for accidents was 2.01 (95% CI 1.67–2.42), for four or more safety climate problems the OR was 4.57 (95% CI 3.64–5.74). Young workers (18–24 years) had higher odds...

Evaluation of major polluting accidents in China-Results and perspectives

International Nuclear Information System (INIS)

Hou Yu; Zhang Tianzhu

2009-01-01

Lessons learnt from accidents are essential sources for updating state-of-the-art requirements in pollution accident prevention. To improve this input in the People's Republic of China in a systematic way, a database for collecting and evaluating major pollution accidents is being established. This is being done in co-operation with Chinese Society for Environment Sciences and other national Institutions. At the time of writing over 80 major events from 2002-2006 have been collected. In this paper, a summary evaluation on the major polluting events in China from 2002 to 2006 is presented and some basic lessons drawn shown. There is no a systematic pollution accident notification system currently in China. The results from root cause analysis underline the importance of emergency measures, maintenance, human factor issues and the role of safety organization. Chronic pollution, especially water pollution and air pollution should be paid the same attention as the sudden pollution. It is important to keep in mind that collecting information from major accidents represents a small percentage of the actual number of events taking place.

Safety and Health Standard 110: Incident/accident reporting and investigation

Energy Technology Data Exchange (ETDEWEB)

Sones, K. [West Kootenay Power, BC (Canada)

1999-10-01

Incident/accident reporting requirements in effect at West Kootenay Power are discussed. Details provided include definitions of low risk, high risk, and critical events, the incidents to be reported, the nature of the reports, the timelines, the investigation to be undertaken for each type of incident/accident, counselling services available to employees involved in serious incidents, and the procedures to be followed in accidents involving serious injury to non-employees. The emphasis is on the `critical five` high risk events and the procedures relating to them.

Report on Fukushima Daiichi NPP precursor events

International Nuclear Information System (INIS)

2014-01-01

The main questions to be answered by this report were: The Fukushima Daiichi NPP accident, could it have been prevented? If there is a next severe accident, may it be prevented? To answer the first question, the report addressed several aspects. First, the report investigated whether precursors to the Fukushima Daiichi NPP accident existed in the operating experience; second, the reasons why these precursors did not evolve into a severe accident. Third, whether lessons learned from these precursor events were adequately considered by member countries; and finally, if the operating experience feedback system needs to be improved, based on the previous analysis. To address the second question which is much more challenging, the report considered precursor events identified through a search and analysis of the IRS database and also precursors events based on risk significance. Both methods can point out areas where further work may be needed, even if it depends heavily on design and site-specific factors. From the operating experience side, more efforts are needed to ensure timely and full implementation of lessons learnt from precursor events. Concerning risk considerations, a combined use of risk precursors and operating experience may drive to effective changes to plants to reduce risk. The report also contains a short description and evaluation of selected precursors that are related to the course of the Fukushima Daiichi NPP accident. The report addresses the question whether operating experience feedback can be effectively used to identify plant vulnerabilities and minimize potential for severe core damage accidents. Based on several of the precursor events national or international in-depth evaluations were started. The vulnerability of NPPs due to external and internal flooding has clearly been addressed. In addition to the IRS based investigation, the WGRISK was asked to identify important precursor events based on risk significance. These precursors have

Preliminary safety analysis of the PWR with accident-tolerant fuels during severe accident conditions

International Nuclear Information System (INIS)

Wu, Xiaoli; Li, Wei; Wang, Yang; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng; Liu, Tong; Deng, Yongjun; Huang, Heng

2015-01-01

Highlights: • Analysis of severe accident scenarios for a PWR fueled with ATF system is performed. • A large-break LOCA without ECCS is analyzed for the PWR fueled with ATF system. • Extended SBO cases are discussed for the PWR fueled with ATF system. • The accident-tolerance of ATF system for application in PWR is illustrated. - Abstract: Experience gained in decades of nuclear safety research and previous nuclear accidents direct to the investigation of passive safety system design and accident-tolerant fuel (ATF) system which is now becoming a hot research point in the nuclear energy field. The ATF system is aimed at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the current UO 2 –Zr system when the reactor is in normal operation. By virtue of advanced materials with improved properties, the ATF system will obviously slow down the progression of accidents, allowing wider margin of time for the mitigation measures to work. Specifically, the simulation and analysis of a large break loss of coolant accident (LBLOCA) without ECCS and extended station blackout (SBO) severe accident are performed for a pressurized water reactor (PWR) loaded with ATF candidates, to reflect the accident-tolerance of ATF

The accidents during shutdown conditions Temelin NPP

International Nuclear Information System (INIS)

Sykora, M.; Mlady, O.

1996-01-01

Two parallel activities oriented for the accidents during shutdown conditions are performed at Temelin NPP: Development of symptom based emergency operating procedures (EOPs) applicable for the accidents which could occur during operational modes 1 through 4; independent evaluation of plant safety as part of the Temelin Shutdown probabilistic assessment to define the accidents which could occur during mode 5 and 6 for which the EOPs must be extended. Both these activities are in progress now because Temelin plant is still in the construction phase

Use of casual tree method for investigation of incidents and accidents involving radioactive materials

International Nuclear Information System (INIS)

Vasconcelos, Vanderley de; Senne Junior, Murillo; Marques, Raissa Oliveira

2013-01-01

There are many methodologies used for investigation of accidents to facilitate the search of the factors that cause these events in different areas of industry. These can be called proactive methods, if they are used before the occurrence of the events, or reactive methods that are applied after the occurrence of the incident or accident, and are used as a basis of information to prevent further events. One of these methods is the Causal Tree Method (CTM). The basic idea of this technique is that incidents and accidents result from variations in usual processes. These variations can be related to the individual, the task, the material or the environment. The tree starts with the end event (incident or accident) and works backwards. The facts relating to the end event are used in the construction of the causal tree. The end event is the starting point and only the facts that contributed to the incident or accident should be selected. The analyst has to identify and list the variations and then display them in the analytic tree, showing causal relations. The objective of this paper is to test the application of the CTM method in investigation of incidents and accidents involving radioactive materials, in order to evaluate its efficiency on finding the typical factors causing these events. (author)

Correction to: Progressive multifocal leukoencephalopathy in rituximab-treated rheumatic diseases: a rare event.

Science.gov (United States)

Berger, Joseph R; Malik, Vineeta; Lacey, Stuart; Brunetta, Paul; Lehane, Patricia B

2018-04-10

The article "Progressive multifocal leukoencephalopathy in rituximab-treated rheumatic diseases: a rare event," written by Joseph R. Berger, Vineeta Malik, Stuart Lacey, Paul Brunetta, and Patricia B. Lehane 3 , was originally published electronically on the publisher's internet portal (currently SpringerLink).

Application of forensic image analysis in accident investigations.

Science.gov (United States)

Verolme, Ellen; Mieremet, Arjan

2017-09-01

Forensic investigations are primarily meant to obtain objective answers that can be used for criminal prosecution. Accident analyses are usually performed to learn from incidents and to prevent similar events from occurring in the future. Although the primary goal may be different, the steps in which information is gathered, interpreted and weighed are similar in both types of investigations, implying that forensic techniques can be of use in accident investigations as well. The use in accident investigations usually means that more information can be obtained from the available information than when used in criminal investigations, since the latter require a higher evidence level. In this paper, we demonstrate the applicability of forensic techniques for accident investigations by presenting a number of cases from one specific field of expertise: image analysis. With the rapid spread of digital devices and new media, a wealth of image material and other digital information has become available for accident investigators. We show that much information can be distilled from footage by using forensic image analysis techniques. These applications show that image analysis provides information that is crucial for obtaining the sequence of events and the two- and three-dimensional geometry of an accident. Since accident investigation focuses primarily on learning from accidents and prevention of future accidents, and less on the blame that is crucial for criminal investigations, the field of application of these forensic tools may be broader than would be the case in purely legal sense. This is an important notion for future accident investigations. Copyright © 2017 Elsevier B.V. All rights reserved.

The accident in Fukushima. Preliminary report on the accident progress in the nuclear power plants as a consequence of the earth quake on 11th March 2011; Der Unfall in Fukushima. Zwischenbericht zu den Ablaeufen in den Kernkraftwerken nach dem Erdbeben vom 11. Maerz 2011

Energy Technology Data Exchange (ETDEWEB)

Borghoff, Stefan; Brueck, Benjamin; Kilian-Huelsmeyer, Yvonne; Maqua, Michael; Mildenberger, Oliver; Quester, Claudia; Stahl, Thorsten; Thuma, Gernot; Wetzel, Norbert; Wild, Volker

2011-08-15

The preliminary report on the accident progress in the nuclear power plants as a consequence of the earth quake on 11th March 2011 describes the chronologic sequence of the accident in the different units of the power plant. The measures for mitigation of the accident impact at the site of Fukushima Daiichi and Fukushima Daini included the efforts to reach and maintain stable plant conditions. The issue radiological situation includes an estimation of the air-borne radionuclide release, the contamination of the environment and the sea water, measures for protection of the public. The lessons learned following the NISA and IAEA fact finding missions and the open questions are summarized.

Consequence analysis of core damage states following severe accidents for the CANDU reactor design

International Nuclear Information System (INIS)

Wahba, N.N.; Kim, Y.T.; Lie, S.G.

1997-01-01

The analytical methodology used to evaluate severe accident sequences is described. The relevant thermal-mechanical phenomena and the mathematical approach used in calculating the timing of the accident progression and source term estimate are summarized. The postulated sever accidents analyzed, in general, mainly differ in the timing to reach and progress through each defined c ore damage state . This paper presents the methodology and results of the timing and steam discharge calculations as well as source term estimate out of containment for accident sequences classified as potentially leading to core disassembly following a small break loss-of-coolant accident (LOCA) scenario as a specific example. (author)

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

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

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

Return on experience on nuclear accidents

International Nuclear Information System (INIS)

Barre, Bertrand

2015-09-01

After a presentation of the International Nuclear and radiological Events Scale (INES scale), of its levels and criteria, this article proposes brief recalls of some nuclear accidents which occurred in nuclear reactors: Chalk River in Canada (1952), Windscale in England (1957), the universal Canadian reactor (NRU in 1958), the SL1 reactor of the Idaho National Laboratory in the USA (1961), the Swiss Lucens reactor (1969), Saint-Laurent des Eaux in France (1969 and 1980). More detailed descriptions are then given for the Three Mile Island accident in 1979, the Chernobyl accident in 1986, and the Fukushima accident in 2011. The main causes of these accidents are identified: loss of control of chain reaction, cooling defect on a stopped reactor, cooling defect on an operated reactor. Some lessons are drawn from these facts, and some characteristics of the EPR are outlined with respect with problems encountered in these accidents

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.

A study on the operator's errors of commission (EOC) in accident scenarios of nuclear power plants: methodology development and application

Energy Technology Data Exchange (ETDEWEB)

Kim, Jae Whan; Jung, Won Dea; Park, Jin Kyun; Kang, Da Il

2003-04-01

As the concern on the operator's inappropriate interventions, the so-called Errors Of Commission (EOCs), that can exacerbate the plant safety has been raised, much of interest in the identification and analysis of EOC events from the risk assessment perspective has been increased. Also, one of the items in need of improvement for the conventional PSA and HRA that consider only the system-demanding human actions is the inclusion of the operator's EOC events into the PSA model. In this study, we propose a methodology for identifying and analysing human errors of commission that might be occurring from the failures in situation assessment and decision making during accident progressions given an initiating event. In order to achieve this goal, the following research items have been performed: Firstly, we analysed the error causes or situations contributed to the occurrence of EOCs in several incidents/accidents of nuclear power plants. Secondly, limitations of the advanced HRAs in treating EOCs were reviewed, and a requirement for a new methodology for analysing EOCs was established. Thirdly, based on these accomplishments a methodology for identifying and analysing EOC events inducible from the failures in situation assessment and decision making was proposed and applied to all the accident sequences of YGN 3 and 4 NPP which resulted in the identification of about 10 EOC situations.

Calculation of spent fuel pool severe accident with MELCOR

International Nuclear Information System (INIS)

Deng Jian; Xiang Qing'an; Zhou Kefeng

2014-01-01

A calculation model was established for spent fuel pool (SFP) using MELCOR code to study the severe accident phenomena caused by the long term station black-out (SBO), including spent fuel heatup, zirconium cladding oxidation, and the injection into SFP to mitigate the severe accident. The results show that the severe accident progression is slow and relates directly with the initial water level in SFP. It is illustrated that the injection into SFP is one of the best mitigated measures for the SFP severe accident. (authors)

A Scoping Analysis Of The Impact Of SiC Cladding On Late-Phase Accident Progression Involving Core–Concrete Interaction

Energy Technology Data Exchange (ETDEWEB)

Farmer, M. T. [Argonne National Lab. (ANL), Argonne, IL (United States)

2015-11-01

The overall objective of the current work is to carry out a scoping analysis to determine the impact of ATF on late phase accident progression; in particular, the molten core-concrete interaction portion of the sequence that occurs after the core debris fails the reactor vessel and relocates into containment. This additional study augments previous work by including kinetic effects that govern chemical reaction rates during core-concrete interaction. The specific ATF considered as part of this study is SiC-clad UO₂.

Strategies for the prevention and mitigation of severe accidents

International Nuclear Information System (INIS)

Ader, C.; Heusener, G.; Snell, V.G.

1999-01-01

The currently operating nuclear power plants have, in general, achieved a high level of safety, as a result of design philosophies that have emphasized concepts such as defense-in-depth. This type of an approach has resulted in plants that have robust designs and strong containments. These designs were later found to have capabilities to protect the public from severe accidents (accidents more severe than traditional design basis in which substantial damage is done to the reactor core). In spite of this high level of safety, it has also been recognized that future plants need to be designed to achieve an enhanced level of safety, in particular with respect to severe accidents. This has led both regulatory authorities and utilities to develop guidance and/or requirements to guide plant designers in achieving improved severe accident performance through prevention and mitigation. The considerable research programs initiated after the TMI-2 accident have provided a large body of technical data, analytical methods, and the expertise necessary to provide for an understanding of a range of severe accident phenomena. This understanding of the ways severe accidents can progress and challenge containments, combined with the wide use of probabilistic safety assessments, have provided designers of evolutionary water cooled reactors opportunities to develop designs that minimize the challenges to the plant and to the public from severe accidents, including the development of accident management strategies intended to further reduce the risk of severe accidents. This paper describes some of the recent progress made in the understanding of severe accidents and related safety assessment methodology and how this knowledge has supported the incorporation of features into representative evolutionary designs that will prevent or mitigate many of the severe accident challenges present in current plants. (author)

Learning from nuclear accident experience

International Nuclear Information System (INIS)

Vaurio, J.K.

1984-01-01

Statistical procedures are developed to estimate accident occurrence rates from historical event records, to predict future rates and trends, and to estimate the accuracy of the rate estimates and predictions. Maximum likelihood estimation is applied to several learning models, and results are compared to earlier graphical and analytical estimates. The models are based on (1) the cumulative number of operating years, (2) the cumulative number of plants built, and (3) accidents (explicitly), with the accident rate distinctly different before and after an accident. The statistical accuracies of the parameters estimated are obtained in analytical form using the Fisher information matrix. Using data on core damage accidents in electricity producing plants, it is estimated that the probability for a plant to have a serious flaw has decreased from 0.1 to 0.01 during the developmental phase of the nuclear industry. At the same time the equivalent frequency of accidents has decreased from 0.04 per reactor year to 0.0004 per reactor year, partly due to the increasing population of plants. 10 references, 7 figures, 2 tables

Accident management-defence in depth in Indian PHWRS

International Nuclear Information System (INIS)

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

2015-01-01

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

Reactivity accident analysis in MTR cores

International Nuclear Information System (INIS)

Waldman, R.M.; Vertullo, A.C.

1987-01-01

The purpose of the present work is the analysis of reactivity transients in MTR cores with LEU and HEU fuels. The analysis includes the following aspects: the phenomenology of the principal events of the accident that takes place, when a reactivity of more than 1$ is inserted in a critical core in less than 1 second. The description of the accident that happened in the RA-2 critical facility in September 1983. The evaluation of the accident from different points of view: a) Theoretical and qualitative analysis; b) Paret Code calculations; c) Comparison with Spert I and Cabri experiments and with post-accident inspections. Differences between LEU and HEU RA-2 cores. (Author)

[Implementation of safety devices: biological accident prevention].

Science.gov (United States)

Catalán Gómez, M Teresa; Sol Vidiella, Josep; Castellà Castellà, Manel; Castells Bo, Carolina; Losada Pla, Nuria; Espuny, Javier Lluís

2010-04-01

Accidental exposures to blood and biological material were the most frequent and potentially serious accidents in healthcare workers, reported in the Prevention of Occupational Risks Unit within 2002. Evaluate the biological percutaneous accidents decrease after a progressive introduction of safety devices. Biological accidents produced between 2.002 and 2.006 were analyzed and reported by the injured healthcare workers to the Level 2b Hospital Prevention of Occupational Risk Unit with 238 beds and 750 employees. The key of the study was the safety devices (peripheral i.v. catheter, needleless i.v. access device and capillary blood collection lancet). Within 2002, 54 percutaneous biological accidents were registered and 19 in 2006, that represents a 64.8% decreased. There has been no safety devices accident reported involving these material. Accidents registered during the implantation period occurred because safety devices were not used at that time. Safety devices have proven to be effective in reducing needle stick percutaneous accidents, so that they are a good choice in the primary prevention of biological accidents contact.

Accidents in industrial radiography in Brazil from 2005 to 2010

International Nuclear Information System (INIS)

Lopes, Ricardo Tadeu

2011-01-01

Analysis of accidents occurring in industrial radiography in Brazil from 2005 until 2010 led to the study of the main characteristics of the events, their risks and dangers. This study outlines the main doubts on the subject, through a simplified analysis of the contents of high dose reports sent to CNEN by the companies that provide services for industrial radiography and from examining the growing number of radioactive sources for industrial radiography in Brazil, over this period. We classified the recorded events, as incidents, accidents, negligence, sabotage, and others, and studied their main consequences. We concluded that from 76 accidents that occurred during that period - 25 were real accidents, 13 minor accidents and 22 were inadvertent incidents. We found that the rate of growth in the number of sources is much greater than the rate of growth of accidents, with a ratio of 7.57 between them. The continuation of this study over some years, will allow the construction of a pyramid of accidents like the one developed by the Insurance Company of North America, specifically for industrial radiography to forecast the number of incidents and accidents that lead to serious or fatal injury. (author)

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.)

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

The Impact of Severe Nuclear Accidents on National Decision for Nuclear Decommissioning

International Nuclear Information System (INIS)

Suh, Young A; Hornibrook, Carol; Yim, Man Sung

2016-01-01

Many researchers have tried to identify the impact of severe nuclear accidents on a country's or international nuclear energy policy [2-3]. However, there is little research on the influence of nuclear accidents and historical events on a country's decision to permanently shutdown an NPP versus international nuclear decommissioning trends. To demonstrate the correlation between a nuclear severe accident and the impact on world nuclear decommissioning, this research reviewed case studies of individual historical events, such as the St. Lucens, TMI, Chernobyl, Fukushima accidents and the series of events leading up to the collapse of the Soviet Union. For validation of the results of these case studies, a statistical analysis was conducted using the R code. This will be useful in explaining how international and national decommissioning strategies are affected by shutdown reasons, i.e. world historical events. The number of permanently shutdown NPPs was selected as an indicator because any relationship between the number of permanently In conclusion, nuclear severe accidents and historical events have an impact on the number of international NPPs that shutdown permanently and cancelled NPP construction. This directly impacts international nuclear decommissioning policy and nuclear energy policy trends. The number of permanently shutdown NPPs was selected as an indicator because any relationship between the number of permanently

Safety related studies on the accident behaviour of the HTR-100

International Nuclear Information System (INIS)

Wolters, J.; Mertens, J.; Altes, J.; Bongartz, R.; Breitbach, G.; David, P.H.; Degen, G.; Ehrlich, H.G.; Escherich, K.H.; Frank, E.; Hennings, W.; Jahn, W.; Koschmieder, R.; Marx, J.; Meister, G.; Moormann, R.; Rehm, W.; Verfondern, K.

1991-10-01

The aim of investigations was to verify the safety concept of the plant for balance and to quantify the radiological risk to be expected in operating an HTR-100 double unit system. Moreover, aspects of the investment risk were considered. The spectrum of initiating events ranged from so-called transients to leaks in the primary circuit and steam generator and even included earthquakes. Some of the event trees derived were highly complex and extensive due to the situation of the steam generator above the core and with regard to the double unit plant concept with increased possibilities of accident control, but also with respect to potential accident propagation. Correspondingly sophisticated analyses were required to identify risk-relevant event sequences. Environmental exposure for all risk-relevant accidents is so low that accident consequence calculations do not reveal any lethal radiation doses and practically no stochastic fatal injuries. These calculations neither assumed acute protective measures nor long-term resettlement or decontamination. The radiological risk caused by an HTR-100 plant is therefore to be classified as very low. The initiating events selected as representative and the event sequences studied in detail cover the risk-relevant event spectrum well into the hypothetical range. (orig./HP) [de

Key regulatory and safety issues emerging NEA activities. Lessons Learned from Fukushima Dai-ichi NPS Accident - Key Regulatory and Safety Issues

International Nuclear Information System (INIS)

Nakoski, John

2013-01-01

A presentation was provided on the key safety and regulatory issues and an update of activities undertaken by the NEA and its members in response to the accident at the Fukushima Daiichi nuclear power stations (NPS) on 11 March 2011. An overview of the accident sequence and the consequences was provided that identified the safety functions that were lost (electrical power, core cooling, and primary containment) that lead to units 1, 2, and 3 being in severe accident conditions with large off-site releases. Key areas identified for which activities of the NEA and member countries are in progress include accident management; defence-in-depth; crisis communication; initiating events; operating experience; deterministic and probabilistic assessments; regulatory infrastructure; radiological protection and public health; and decontamination and recovery. For each of these areas, a brief description of the on-going and planned NEA activities was provided within the three standing technical committees of the NEA with safety and regulatory mandates (the Committee on Nuclear Regulatory Activities - CNRA, the Committee on the Safety of Nuclear Installations - CSNI, and the Committee on Radiation Protection and Public Health - CRPPH). On-going activities of CNRA include a review of enhancement being made to the regulatory aspects for the oversight of on-site accident management strategies and processes in light of the lessons learned from the accident; providing guidance to regulators on crisis communication; and supporting the peer review of the safety assessments of risk-significant research reactor facilities in light of the accident. Within the scope of the CSNI mandate, activities are being undertaken to better understand accident progression; characteristics of new fuel designs; and a benchmarking study of fast-running software for estimating source term under severe accident conditions to support protective measure recommendations. CSNI also has ongoing work in human

Emotional stress and traffic accidents: the impact of separation and divorce.

Science.gov (United States)

Lagarde, Emmanuel; Chastang, Jean-François; Gueguen, Alice; Coeuret-Pellicer, Mireille; Chiron, Mireille; Lafont, Sylviane

2004-11-01

Personal responses to stressful life events are suspected of increasing the risk of serious traffic accidents. We analyzed data from a French cohort study (the GAZEL cohort), including a retrospective driving behavior questionnaire, from 13,915 participants (10,542 men age 52-62 years and 3373 women age 47-62 years in 2001). Follow-up data covered 1993-2000. Hazard ratios for serious accidents (n = 713) were computed by Cox's proportional hazard regression with time-dependent covariates. Separate analyses were also performed to consider only at-fault accidents. Marital separation or divorce was associated with an increased risk of a serious accident (all serious accidents: hazard ratio 2.9, 95% confidence interval = 1.7-5.0; at-fault accidents: 4.4, 2.3-8.3). The impact of separation and divorce did not differ according to alcohol consumption levels. Other life events associated with increased risk of serious accident were a child leaving home (all accidents: 1.2, 0.97-1.6; at-fault accidents: 1.5, 1.1-2.1), an important purchase (all accidents: 1.4, 1.1-1.7; at-fault accidents: 1.6, 1.2-2.1), and hospitalization of the partner (all accidents: 1.4, 1.1-2.0). This study suggests that recent separation and divorce are associated with an increase in serious traffic accidents.

Overview of severe accident research at JAERI

International Nuclear Information System (INIS)

Sugimoto, Jun

1999-01-01

Severe accident research at JAERI aims at the confirmation of the safety margin, the quantification of the associated risk, and the evaluation of the effectiveness of the accident management measures of the nuclear power reactors, in accordance with the government five-year nuclear safety research program. JAERI has been conducting a wide range of severe accident research activities both in experiment and analysis, such as melt coolant interactions, fission product behaviors in coolant system, containment integrity and assessment of accident management measures. Molten core/coolant interaction and in-vessel molten coolability have been investigated in ALPHA Program. MUSE experiments in ALPHA Program has been conducted for the precise energy measurement due to steam explosion in melt jet and stratified geometries. In VEGA Program, which aims at FP release from irradiated fuels at high temperature and high pressure under various atmospheric conditions, the facility construction is almost completed. In WIND Program the revaporization of aerosols due to decay heating and also the integrity of the piping from this heat source are being investigated. Code development activities are in progress for an integrated source term analysis with THALES, fission product behaviors with ART, steam explosion with JASMINE, and in-vessel debris behaviors with CAMP. The experimental analyses and reactor application have made progress by participating international standard problem and code comparison exercises, along with the use of introduced codes, such as SCDAP/RELAP5 and MELCOR. The outcome of the severe accident research will be utilized for the evaluation of more reliable severe accident scenarios, detailed implementation of the accident management measures, and also for the future reactor development, basically through the sophisticated use of verified analytical tools. (author)

Generalization of Nuclear Safety and Course of Accident Events Research in the Ignalina NPP

International Nuclear Information System (INIS)

Kaliatka, A.; Uspuras, E.

2001-01-01

The safety analysis shown that after implementation of SAR recommendations Ignalina NPP is adequately protected against accidents which required fast initiation of automatic protections. In case of accidents with long-term loss of core cooling additional operator actions are required. Accident management in case long-term core cooling are analyzed in this paper. (author)

Keynote on lessons from major radiation accidents

International Nuclear Information System (INIS)

Ortiz, P.; Oresegun, M.; Wheatley, J.

2000-01-01

Generic lessons have been learned from a relatively large number of accidents in the most relevant practices (a set of analysis have been made on about 90 radiotherapy events, 43 industrial radiography and nine from industrial irradiations); more specific lessons have been drawn from in-depth investigations of individual accidents. The body of knowledge is grouped as follows: a) radiotherapy is very unique in that humans (patients) are purposely given very high radiation doses (20-75 Gy) by placing them in the radiation beam or by placing radioactive sources in contact with tissues. Intended deterministic effects are the essence of the normal radiotherapy practice and relatively small deviation from the intended doses, i.e,, slightly higher or lower than intended may cause increased rate of severe complication or reduce probability of cure. Consequences of major accidents have been devastating, affecting tens, even hundreds of patients and causing death (directly or indirectly) to a large number of them; b) accidents involving industrial radiography are the most frequent cause of overexposure to workers (radiographers); c) accidents with industrial irradiators have lower probability of occurrence, however, they are deemed to be fatal, especially when whole body exposure to panoramic gamma irradiators occur; partial body irradiation from industrial or research accelerator beams has led to amputation of hands and legs; d) when control of sources was relinquished ('orphan' sources) this has resulted in severe injuries, in some cases death and widespread contamination of the environment. A tool for further dissemination of lessons will be an international reporting system of unusual radiation events (RADEV), being introduced world-wide. Accidents were rarely due to a single human error or isolated equipment failure. In most cases there was a combination of elements such as: a) unawareness of the potential for an accident, b) poor education, which usually did not

Quantitative risk trends deriving from PSA-based event analyses. Analysis of results from U.S.NRC's accident sequence precursor program

International Nuclear Information System (INIS)

Watanabe, Norio

2004-01-01

The United States Nuclear Regulatory Commission (U.S.NRC) has been carrying out the Accident Sequence Precursor (ASP) Program to identify and categorize precursors to potential severe core damage accident sequences using the probabilistic safety assessment (PSA) technique. The ASP Program has identified a lot of risk significant events as precursors that occurred at U.S. nuclear power plants. Although the results from the ASP Program include valuable information that could be useful for obtaining and characterizing risk significant insights and for monitoring risk trends in nuclear power industry, there are only a few attempts to determine and develop the trends using the ASP results. The present study examines and discusses quantitative risk trends for the industry level, using two indicators, that is, the occurrence frequency of precursors and the annual core damage probability, deriving from the results of the ASP analysis. It is shown that the core damage risk at U.S. nuclear power plants has been lowered and the likelihood of risk significant events has been remarkably decreasing. As well, the present study demonstrates that two risk indicators used here can provide quantitative information useful for examining and monitoring the risk trends and/or risk characteristics in nuclear power industry. (author)

The study on development of emergency operating procedures based on symptom and risk for accident management

International Nuclear Information System (INIS)

Kang, K. S.; Jeong, H. J.

1998-01-01

The Advanced EOP(AEOP) has been developed by focusing on the importance of the operators role in emergency conditions. In the AEOP, to overcome the complexity of current EOPs and maintain the consistency of operators action according to plant emergency conditions, operator's task were allocated according to their duties. As an alternative, the Computerized Operator Aid System (COAS) has been developed to reduce operator's burden and provide detailed instructions of procedure. Probabilistic Safety Assessment (PSA) results were synthesized in the AEOP using the event tree to give the awareness and the prediction of accident progression in advance. In conclusion, the existing EOP with its inherent complexity should be simplified and consolidated using computerized operator support system and task allocation to prevent more severe accidents and to reduce operator cognitive overload in emergency conditions

Questions concerning safety and risk after the nuclear accidents in Japan. Deepened accident analysis for the Fukushima Daiichi power plant; Sicherheits- und Risikofragen im Nachgang zu den nuklearen Stoer- und Unfaellen in Japan. Vertiefte Ereignisanalyse zur Anlage Fukushima-Daini

Energy Technology Data Exchange (ETDEWEB)

Pistner, Christoph; Englert, Matthias [Oeko-Institut e.V. - Institut fuer Angewandte Oekologie, Darmstadt (Germany)

2015-02-25

The study questions concerning safety and risk in Japanese power plants following the disastrous nuclear accident covers the following issues: the nuclear facility Fukushima Daiichi, site characterization, important technical equipment, important electro-technical equipment, personal; description of the accident progression in the Fukushima nuclear power plant: impact of the earthquake, impact of the tsunami, short-term measures of the operating personnel, pressure and temperature situation in the containments, restoration of the after-heat cooling system in the units 1/2 and 4, fuel element storage pool, summarized parameters during the accident progress; comparative analysis of the accident progression at the Fukushima Daiichi site.

Kyshtym riddle: possible kind of the accident

International Nuclear Information System (INIS)

Ballereau, P.

1988-01-01

It can been postulated from varied rumors, soviet testimonies, analysis of radioecological russian publications that a serious radiological accident occurred in late 1957 - early 1958 in the Oural mountains. Isotopic ratio 90 Sr/ 137 Cs in the environment following the accident was abnormally high. Several types of accidents has been postulated; the more credible event is an explosion in a storage tank containing dried high activity wastes and NH 4 N0 3 , from which 137 Cs had been extracted [fr

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

Science.gov (United States)

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

2015-01-01

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

On high-temperature reactor accident topology

International Nuclear Information System (INIS)

Fassbender, J.; Kroeger, W.; Wolters, J.

1981-01-01

American and German risk studies for an HTGR and independent investigations of hypothetical accident sequences led to a fundamental understanding of the topology of HTGR accident sequences. The dominating importance of core heat-up accidents was confirmed and the initiating events were identified. Complications of core heat-up accidents by air or water ingress are of minor importance for the risk, whereas the long-term development of accidents during days and weeks plays an important role for the environmental impact. The risk caused by an HTGR at a German site cannot yet be determined exactly, because no modern German HTGR design has passed a licensing procedure. Cautious estimates show that risk will appear to be substantially smaller than the LWR risk. The main reasons are the considerably reduced release of fission procucts and the slow development of core heat-up accidents leaving much time for measures which reduce the risk. (orig.) [de

Chemical phenomena under severe accident conditions

International Nuclear Information System (INIS)

Powers, D.A.

1988-01-01

A severe nuclear reactor accident is expected to involve a vast number of chemical processes. The chemical processes of major safety significance begin with the production of hydrogen during steam oxidation of fuel cladding. Physico-chemical changes in the fuel and the vaporization of radionuclides during reactor accidents have captured much of the attention of the safety community in recent years. Protracted chemical interactions of core debris with structural concrete mark the conclusion of dynamic events in a severe accident. An overview of the current understanding of chemical processes in severe reactor accident is provided in this paper. It is shown that most of this understanding has come from application of findings from other fields though a few areas have in the past been subject to in-depth study of a fundamental nature. Challenges in the study of severe accident chemistry are delineated

Accident at the Three Mile Island Nuclear Power Plant

International Nuclear Information System (INIS)

Bajusz, J.; Vamos, G.

1979-01-01

A short description of the TMI power plant is given. The course of events leading to the reactor accident and that of the first two weeks is described. The effect on the environment is estimated. The reasons and consequences of the accident are analysed. The probability of such an accident at the Paks Nuclear Power Plant is estimated. (R.J.)

Accident at Harrisburg

International Nuclear Information System (INIS)

1979-05-01

The course of events during the accident on 28 March 1979 at Three Mile Island-2 Reactor at Harrisburg, Pennsylvania, is described in detail. The effects (in the environment and within the safety containment) are described. The following points are then discussed: the possibility of a comparable accident occurring in the nuclear power stations in the German Federal Republic; the possibility of any point having been overlooked in the design of nuclear power stations in the Federal Republic; whether previous risk analyses are still valid; and how near the Three Mile Island reactor was to a core meltdown. Some conclusions are drawn. (U.K.)

Risk evaluation for protection of the public in radiation accidents

International Nuclear Information System (INIS)

1967-01-01

Evaluation of the risk that would be involved in the exposure of the public in the event of a radiation accident requires information on the biological consequences expected of such an exposure. This report defines a range of reference doses of radiation and their corresponding risks to the public in the event of a radiation accident. The reference doses and the considerations on which they were based will be used for assessing the hazards of nuclear installations and for policy decisions by the authorities responsible for measures taken to safeguards the public in the case of a nuclear accident.

Radiological accidents/incidents with caesium-137 in Estonia

International Nuclear Information System (INIS)

Sinisoo, M.

1998-01-01

A report is provided of an accident and an incident involving radioactive sources in Estonia. In the 1994 occurrence, looters of a depository of radioactive waste manipulated a source containing 137 Cs and received dangerous doses of radiation. One of the persons involved died, others suffered minor burns. Another event, which occurred in early 1995, did not have a tragic outcome: an abandoned 137 Cs source was found in the vicinity of the highway linking Tallinn and Narva and was disposed of safely. Both these accidents draw attention to the potential dangers caused by the insufficient survey of the territory, radiation protection structures not yet fully operable, and the lack of equipment and know-how. The lessons to be drawn from these events are considered on the basis of the chronologies and factual data. The report contains concise descriptions of the accidents, a medical overview of the fate of the injured persons and the lessons learned from these accidents. (author)

Sequence Tree Modeling for Combined Accident and Feed-and-Bleed Operation

International Nuclear Information System (INIS)

Kim, Bo Gyung; Kang Hyun Gook; Yoon, Ho Joon

2016-01-01

In order to address this issue, this study suggests the sequence tree model to analyze accident sequence systematically. Using the sequence tree model, all possible scenarios which need a specific safety action to prevent the core damage can be identified and success conditions of safety action under complicated situation such as combined accident will be also identified. Sequence tree is branch model to divide plant condition considering the plant dynamics. Since sequence tree model can reflect the plant dynamics, arising from interaction of different accident timing and plant condition and from the interaction between the operator action, mitigation system, and the indicators for operation, sequence tree model can be used to develop the dynamic event tree model easily. Target safety action for this study is a feed-and-bleed (F and B) operation. A F and B operation directly cools down the reactor cooling system (RCS) using the primary cooling system when residual heat removal by the secondary cooling system is not available. In this study, a TLOFW accident and a TLOFW accident with LOCA were the target accidents. Based on the conventional PSA model and indicators, the sequence tree model for a TLOFW accident was developed. If sampling analysis is performed, practical accident sequences can be identified based on the sequence analysis. If a realistic distribution for the variables can be obtained for sampling analysis, much more realistic accident sequences can be described. Moreover, if the initiating event frequency under a combined accident can be quantified, the sequence tree model can translate into a dynamic event tree model based on the sampling analysis results

Sequence Tree Modeling for Combined Accident and Feed-and-Bleed Operation

Energy Technology Data Exchange (ETDEWEB)

Kim, Bo Gyung; Kang Hyun Gook [KAIST, Daejeon (Korea, Republic of); Yoon, Ho Joon [Khalifa University of Science, Abu Dhabi (United Arab Emirates)

2016-05-15

In order to address this issue, this study suggests the sequence tree model to analyze accident sequence systematically. Using the sequence tree model, all possible scenarios which need a specific safety action to prevent the core damage can be identified and success conditions of safety action under complicated situation such as combined accident will be also identified. Sequence tree is branch model to divide plant condition considering the plant dynamics. Since sequence tree model can reflect the plant dynamics, arising from interaction of different accident timing and plant condition and from the interaction between the operator action, mitigation system, and the indicators for operation, sequence tree model can be used to develop the dynamic event tree model easily. Target safety action for this study is a feed-and-bleed (F and B) operation. A F and B operation directly cools down the reactor cooling system (RCS) using the primary cooling system when residual heat removal by the secondary cooling system is not available. In this study, a TLOFW accident and a TLOFW accident with LOCA were the target accidents. Based on the conventional PSA model and indicators, the sequence tree model for a TLOFW accident was developed. If sampling analysis is performed, practical accident sequences can be identified based on the sequence analysis. If a realistic distribution for the variables can be obtained for sampling analysis, much more realistic accident sequences can be described. Moreover, if the initiating event frequency under a combined accident can be quantified, the sequence tree model can translate into a dynamic event tree model based on the sampling analysis results.

Study on severe accident induced by large break loss of coolant accident for pressureized water reactor

International Nuclear Information System (INIS)

Zhang Longfei; Zhang Dafa; Wang Shaoming

2007-01-01

Using the best estimate computer code SCDAP/RELAP5/MOD3.2 and taking US Westinghouse corporation Surry nuclear power plant as the reference object, a typical three-loop pressurized water reactor severe accident calculation model was established and 25 cm large break loss of coolant accident (LBLOCA) in cold and hot leg of primary loop induced core melt accident was analyzed. The calculated results show that core melt progression is fast and most of the core material melt and relocated to the lower plenum. The lower head of reactor pressure vessel failed at an early time and the cold leg break is more severe than the hot leg break in primary loop during LBLOCA. (authors)

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit-1: Analysis of core damage frequency from internal events during mid-loop operations. Appendix I, Volume 2, Part 5

Energy Technology Data Exchange (ETDEWEB)

Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J. [Brookhaven National Lab., Upton, NY (United States); Bley, D.; Johnson, D. [PLG Inc., Newport Beach, CA (United States); Holmes, B. [AEA Technology, Dorset (United Kingdom)] [and others

1994-06-01

Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analyses that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Lab. (BNL) and Sandia National Labs. (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this volume of the report is to document the approach utilized in the level-1 internal events PRA for the Surry plant, and discuss the results obtained. A phased approach was used in the level-1 program. In phase 1, which was completed in Fall 1991, a coarse screening analysis examining accidents initiated by internal events (including internal fire and flood) was performed for all plant operational states (POSs). The objective of the phase 1 study was to identify potential vulnerable plant configurations, to characterize (on a high, medium, or low basis) the potential core damage accident scenarios, and to provide a foundation for a detailed phase 2 analysis.

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit-1: Analysis of core damage frequency from internal events during mid-loop operations. Appendix I, Volume 2, Part 5

International Nuclear Information System (INIS)

Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J.; Bley, D.; Johnson, D.; Holmes, B.

1994-06-01

Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analyses that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Lab. (BNL) and Sandia National Labs. (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this volume of the report is to document the approach utilized in the level-1 internal events PRA for the Surry plant, and discuss the results obtained. A phased approach was used in the level-1 program. In phase 1, which was completed in Fall 1991, a coarse screening analysis examining accidents initiated by internal events (including internal fire and flood) was performed for all plant operational states (POSs). The objective of the phase 1 study was to identify potential vulnerable plant configurations, to characterize (on a high, medium, or low basis) the potential core damage accident scenarios, and to provide a foundation for a detailed phase 2 analysis

Impact of the accident at TMI-2 on new safety regulations

International Nuclear Information System (INIS)

Collins, J.T.

1981-01-01

The Nuclear Regulatory Commission (NRC) has been very busy, since the accident, looking into the causes surrounding the events that occurred on the morning of March 28, 1979. To date, the Commission has implemented the Short-Term Lessons Learned and has provided a schedule for implementing the Long-Term Lessons Learned. Some of these requirements have resulted in delays in licensing of new plants and the temporary shutdown of some operating plants. However, the NRC believes these new requirements are essential to increase the safety of nuclear power plants and to protect the health and safety of the public. Although the accident occurred almost 19 months ago, the cleanup of TMI-2 continues and will continue for the next 5 to 7 years. As the cleanup progresses and ultimately the fuel removed, the Commission will continue to learn from the information generated by this program. This information will be factored into the licensing process. If nuclear power is to remain a viable option as a source of electrical power in the United States, then NRC must continue to assure the general public that these plants can be operated safely from the lessons learned at TMI and that systems required to mitigate the consequences of accidents will indeed perform their intended functions

Chernobyl accident

International Nuclear Information System (INIS)

Bar'yakhtar, V.G.

1995-01-01

The monograph contains the catastrophe's events chronology, the efficiency assessed of those measures assumed for their localization as well as their environmental and socio-economic impact. Among materials of the monograph the results are presented of research on the radioactive contamination field forming as well as those concerning the investigation of biogeochemical properties of Chernobyl radionuclides and their migration process in the environment of the Ukraine. The data dealing with biological effects of the continued combined internal and external radioactive influence on plants, animals and human health under the circumstances of Chernobyl accident are of the special interest. In order to provide the scientific generalizing information on the medical aspects of Chernobyl catastrophe, the great part of the monograph is allotted to appraise those factors affecting the health of different population groups as well as to depict clinic aspects of Chernobyl events and medico-sanitarian help system. The National Programme of Ukraine for the accident consequences elimination and population social protection assuring for the years 1986-1993 and this Programme concept for the period up to the year 2000 with a special regard of the world community participation there

Study on risk factors of PWR accidents beyond design basis

International Nuclear Information System (INIS)

Ahn, Seung Hoon; Nah, W. J.; Bang, Y. S.; Oh, D. Y.; Oh, S. H.

2005-01-01

Development of the regulatory guidelines for Beyond Design Basis Accidents (BDBA) with high risk requires a detailed investigation of major factors contributing to the event risk. In this study, each event was classified by the level of risk, based on the probabilistic safety assessment results, so that BDBA with high risk could be selected, with consideration of foreign and domestic regulations, and operating experiences. The regulatory requirements and technical backgrounds for the selected accidents were investigated, and effective regulatory approaches for risk reduction of the accidents. The following conclusions were drawn from this study: - Selected high risk BDBA is station blackout, anticipated without scram, total loss of feedwater. - Major contributors to the risk of selected events were investigated, and appropriate assessment of them was recommended for development of the regulatory guidelines

Review of accident analyses performed at Mochovce NPP

International Nuclear Information System (INIS)

Siko, D.

2000-01-01

In this paper the review of accident analysis performed in NPP Mochovce V-1 is presented. The scope of these safety measures was defined and development in the T SSM for NPP Mochovce Nuclear Safety Improvements Report' issued in July 1995. The main objectives of these safety measures were the followings: (a) to establish the criteria for selection and classification of accidental events, as well as defining the list of initiating events to be analysed. Accident classification to the individual groups must be performed in accordance with RG 1.70 and IAEA recommendations 'Guidelines for Accidental Analysis of WWER NPP' (IAEA-EBR-WWER-01) to select boundary cases to be calculated from the scope of initiating events; (b ) to elaborate the accident analysis methodology that also includes acceptance criteria for their result evaluation, initial and boundary conditions, assumption related with the application of the single failure criteria, requirements on the analysis quality, used computer codes, as well as NPP models and input data for the accident analysis; (c) to perform the accident analysis for the Pre-operational Safety Report (POSAR); (d) to provide a synthetic report addressing the validity range of codes models and correlations, the assessment against relevant tests results, the evidence of the user qualification, the modernisation and nodding scheme for the plant and the justification of used computer codes. Analyses results showed that all acceptance criteria were met with satisfactory margin and design of the NPP Mochovce is accurate. (author)

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

International Nuclear Information System (INIS)

Nomura, Shigeo; Katoh, Kazuyuki; Okano, Kenta

2016-01-01

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

Accidents - Chernobyl accident; Accidents - accident de Tchernobyl

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

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

Experimental Investigation of Operation of VVER Steam Generator in Condensation Mode in the Event of the Severe Accident

Energy Technology Data Exchange (ETDEWEB)

Morozov, Andrey [Institute for Physics and Power Engineering by A.I. Leypunsky, 1 Bondarenko sq. Obninsk, 249033 (Russian Federation)

2008-07-01

For new Russian nuclear power plants with VVER-1200 reactor in the event of a beyond design basis accident, provision is made for the use of passive safety systems for necessary core cooling. These safety systems include the passive heat removal system (PHRS). In the case of leakage in the primary circuit this system assures the transition of steam generators (SG) to operation in the mode of condensation of the primary circuit steam. As a result, the condensate from SG arrives at the core providing its additional cooling. To investigate the condensation mode of VVER SG operation, a large scale HA2M-SG test facility was constructed. The rig incorporates: buffer tank, SG model with scale is 1:46, PHRS heat exchanger. Experiments at the test facility have been performed to investigate condensation mode of operation of SG model at the pressure 0.4 MPa, correspond to VVER reactor pressure at the last stage of the beyond design basis accident. The report presents the test procedure and the basic obtained test results. (authors)

Investigation into information flow during the accident at Three Mile Island

International Nuclear Information System (INIS)

1981-01-01

This report was prepared in response to a request from NRC Chairman Ahearne that directed the Office of Inspection and Enforcement to resume its investigation of information flow during the accident at Three Mile Island (TMI) that occurred on March 28, 1979. This investigation was resumed on March 21, 1980. The transfer of information among individuals, agencies, and personnel from Metropolitan Edison was analyzed to ascertain what knowledge was held by various individuals of the specific events, parameters, and systems during the accident at TMI. Maximum use was made of existing records, and additional interviews were conducted to clarify areas that had not been pursued during earlier investigations. Although the passage of time between the accident and post-accident interviews hampered precise recollections of events and circumstances, the investigation revealed that information was not intentionally withheld during the accident and that the system for effective transfer of information was inadequate during the accident

Accident sequences and causes analysis in a hydrogen production process

Energy Technology Data Exchange (ETDEWEB)

Jae, Moo Sung; Hwang, Seok Won; Kang, Kyong Min; Ryu, Jung Hyun; Kim, Min Soo; Cho, Nam Chul; Jeon, Ho Jun; Jung, Gun Hyo; Han, Kyu Min; Lee, Seng Woo [Hanyang Univ., Seoul (Korea, Republic of)

2006-03-15

Since hydrogen production facility using IS process requires high temperature of nuclear power plant, safety assessment should be performed to guarantee the safety of facility. First of all, accident cases of hydrogen production and utilization has been surveyed. Based on the results, risk factors which can be derived from hydrogen production facility were identified. Besides the correlation between risk factors are schematized using influence diagram. Also initiating events of hydrogen production facility were identified and accident scenario development and quantification were performed. PSA methodology was used for identification of initiating event and master logic diagram was used for selection method of initiating event. Event tree analysis was used for quantification of accident scenario. The sum of all the leakage frequencies is 1.22x10{sup -4} which is similar value (1.0x10{sup -4}) for core damage frequency that International Nuclear Safety Advisory Group of IAEA suggested as a criteria.

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

Science.gov (United States)

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

2015-01-01

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

The Impact of Severe Nuclear Accidents on National Decision for Nuclear Decommissioning

Energy Technology Data Exchange (ETDEWEB)

Suh, Young A; Hornibrook, Carol; Yim, Man Sung [KAIST, Daejeon (Korea, Republic of)

2016-05-15

Many researchers have tried to identify the impact of severe nuclear accidents on a country's or international nuclear energy policy [2-3]. However, there is little research on the influence of nuclear accidents and historical events on a country's decision to permanently shutdown an NPP versus international nuclear decommissioning trends. To demonstrate the correlation between a nuclear severe accident and the impact on world nuclear decommissioning, this research reviewed case studies of individual historical events, such as the St. Lucens, TMI, Chernobyl, Fukushima accidents and the series of events leading up to the collapse of the Soviet Union. For validation of the results of these case studies, a statistical analysis was conducted using the R code. This will be useful in explaining how international and national decommissioning strategies are affected by shutdown reasons, i.e. world historical events. The number of permanently shutdown NPPs was selected as an indicator because any relationship between the number of permanently In conclusion, nuclear severe accidents and historical events have an impact on the number of international NPPs that shutdown permanently and cancelled NPP construction. This directly impacts international nuclear decommissioning policy and nuclear energy policy trends. The number of permanently shutdown NPPs was selected as an indicator because any relationship between the number of permanently.

Analysis of steam generator tube rupture as a severe accident using MELCOR 1.8.4

International Nuclear Information System (INIS)

Yang Hongrun; Hidaka, Akihide; Sugimoto, Jun

1999-03-01

This report presents the results from the MELCOR 1.8.4 calculations for Steam Generator Tube Rupture (SGTR) with stuck open of all the safety valves in faulted SG as a severe accident. The calculations are based on Surry nuclear power plant. After performed using the once-through primary system model alone by 1.0x10 5 s, the calculations were conducted with both of the once-through and the hot leg countercurrent natural circulation models. The results, including event sequences, processes and progressions of core degradation, radionuclides release from core and reactor cavity, and source terms to the environment are described in detail. It is concluded that the availability of High Pressure Safety Injection (HPSI) can significantly delay the progression of core heat-up and approximately 7% of cesium iodide (CsI) can be released to the environment directly through the stuck open safety valve. Comparisons between the results from the two models are also given in this report. The present analyses also showed that during SGTR accident, the hot leg countercurrent natural circulation flow cannot be established well and therefore it has little effect on the mitigation of the core degradation. (author)

Safety regulations regarding to accident monitoring and accident sampling at Russian NPPs with VVER type reactors

International Nuclear Information System (INIS)

Sharafutdinov, Rachet; Lankin, Michail; Kharitonova, Nataliya

2014-01-01

The paper describes a tendency by development of regulatory document requirements related to accident monitoring and accident sampling at Russia's NPPs. Lessons learned from the Fukushima Daiichi accident pointed at the importance and necessary to carry out an additional safety check at Russia's nuclear power plants in the preparedness for management of severe accidents at NPPs. Planned measures for improvement of severe accidents management include development and implementation of the accident instrumentation systems, providing, monitoring, management and storage of information in a severe accident conditions. The draft of Safety Guidelines <accident monitoring system of nuclear power plants with VVER reactors' prepared by Scientific and Engineering Centre for Nuclear and Radiation Safety (SEC NRS) established the main criteria for accident monitoring instrumentation that can monitor relevant plant parameters in the reactor and inside containment during and after a severe accident in nuclear power plants. Development of these safety guidelines is in line with the recommendations of IAEA Action Plan on Nuclear Safety in response to the Fukushima Daiichi event and recommendations of the IAEA Nuclear Energy series Report <<Accident Monitoring Systems for Nuclear Power Plants' (Draft V 2.7). The paper presents the principles, which are used as the basis for selection of plant parameters for accident monitoring and for establishing of accident monitoring instrumentation. The recommendations to the accident sampling system capable to obtain the representative reactor coolant and containment air and fluid samples that support accurate analytical results for the parameters of interest are considered. The radiological and chemistry parameters to be monitored for primary coolant and sump and for containment air are specified. (author)

The psychological impact of the radiological accident in Goiania

International Nuclear Information System (INIS)

Carvalho, A.B. de.

1988-01-01

This work describes the psychological impact of an accident caused by the violation of a capsule containing Cesium 137 in the city of Goiania, Goias, Brazil, in September of 1987. Its object is to confirm the importance of having mental health teams working, not only with accident victims, but also side by side with the rescue teams in the event of radiation accidents. (author) [pt

Severe accident sequences simulated at the Grand Gulf Nuclear Station

International Nuclear Information System (INIS)

Carbajo, J.J.

1999-01-01

Different severe accident sequences employing the MELCOR code, version 1.8.4 QK, have been simulated at the Grand Gulf Nuclear Station (Grand Gulf). The postulated severe accidents simulated are two low-pressure, short-term, station blackouts; two unmitigated small-break (SB) loss-of-coolant accidents (LOCAs) (SBLOCAs); and one unmitigated large LOCA (LLOCA). The purpose of this study was to calculate best-estimate timings of events and source terms for a wide range of severe accidents and to compare the plant response to these accidents

REAC/TS radiation accident registry. Update of accidents in the United States

International Nuclear Information System (INIS)

Ricks, R.C.; Berger, M.E.; Holloway, E.C.; Goans, R.E.

2000-01-01

Serious injury due to ionizing radiation is a rare occurrence. From 1944 to the present, 243 US accidents meeting dose criteria for classification as serious are documented in the REAC/TS Registry. Thirty individuals have lost their lives in radiation accidents in the United States. The Registry is part of the overall REAC/TS program providing 24-hour direct or consultative assistance regarding medical and heath physics problems associated with radiation accidents in local, national, and international incidents. The REAC/TS Registry serves as a repository of medically important information documenting the consequences of these accidents. Registry data are gathered from various sources. These include reports from the World Heath Organization (WHO), International Atomic Energy Agency (IAEA), US Nuclear Regulatory Commission (US NRC), state radiological health departments, medical/health physics literature, personal communication, the Internet, and most frequently, from calls for medical assistance to REAC/TS, as part of our 24-hour medical assistance program. The REAC/TS Registry for documentation of radiation accidents serves several useful purposes: 1) weaknesses in design, safety practices, training or control can be identified, and trends noted; 2) information regarding the medical consequences of injuries and the efficacy of treatment protocols is available to the treating physician; and 3) Registry case studies serve as valuable teaching tools. This presentation will review and summarize data on the US radiation accidents including their classification by device, accident circumstances, and frequency by respective states. Data regarding accidents with fatal outcomes will be reviewed. The inclusion of Registry data in the IAEA's International Reporting System of Radiation Events (RADEV) will also be discussed. (author)

[Violence and accidents among older and younger adults: evidence from the Surveillance System for Violence and Accidents (VIVA), Brazil].

Science.gov (United States)

Luz, Tatiana Chama Borges; Malta, Deborah Carvalho; Sá, Naíza Nayla Bandeira de; Silva, Marta Maria Alves da; Lima-Costa, Maria Fernanda

2011-11-01

Data from the Brazilian Surveillance System for Violence and Accidents (VIVA) in 2009 were used to examine socio-demographic characteristics, outcomes, and types of accidents and violence treated at 74 sentinel emergency services in 23 Brazilian State capitals and the Federal District. The analysis included 25,201 individuals aged > 20 years (10.1% > 60 years); 89.3% were victims of accidents and 11.9% victims of violence. Hospitalization was the outcome in 11.1% of cases. Compared to the general population, there were more men and non-white individuals among victims of accidents, and especially among victims of violence. As compared to younger adults (20-59 years), accidents and violence against elderly victims showed less association with alcohol, a higher proportion of domestic incidents, more falls and pedestrian accidents, and aggression by family members. Policies for the prevention of accidents and violence should consider the characteristics of these events in the older population.

Accident sequence analysis of human-computer interface design

International Nuclear Information System (INIS)

Fan, C.-F.; Chen, W.-H.

2000-01-01

It is important to predict potential accident sequences of human-computer interaction in a safety-critical computing system so that vulnerable points can be disclosed and removed. We address this issue by proposing a Multi-Context human-computer interaction Model along with its analysis techniques, an Augmented Fault Tree Analysis, and a Concurrent Event Tree Analysis. The proposed augmented fault tree can identify the potential weak points in software design that may induce unintended software functions or erroneous human procedures. The concurrent event tree can enumerate possible accident sequences due to these weak points

Development Status of Accident Tolerant Fuel Cladding for LWRs

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyun-Gil; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jung-Hwan; Yang, Jae-Ho; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Hydrogen explosions and the release of radionuclides are caused by severe damage of current nuclear fuels, which are composed of fuel pellets and fuel cladding, during an accident. To reduce the damage to the public, the fuels have to enhance their integrity under an accident environment. Enhanced accident tolerance fuels (ATFs) can tolerate a loss of active cooling in the reactor core for a considerably longer time period during design-basis and beyond design-basis events while maintaining or improving the fuel performance during normal operations as well as operational transients, in comparison with the current UO{sub 2}-Zr alloy system used in the LWR. Surface modified Zr cladding as a new concept was suggested to apply an enhanced ATF cladding. The aim of the partial ODS treatment is to increase the high-temperature strength to suppress the ballooning/rupture behavior of fuel cladding during an accident event. The target of the surface coating is to increase the corrosion resistance during normal operation and increase the oxidation resistance during an accident event. The partial ODS treatment of Zircaloy-4 cladding can be produced using a laser beam scanning method with Y2O3 powder, and the surface Cr-alloy and Cr/FeCrAl coating on Zircaloy-4 cladding can be obtained after the development of 3D laser coating and arc ion plating technologies.

Questions about the reactor accident with Chernobyl-4

International Nuclear Information System (INIS)

Heijboer, R.J.

1986-01-01

The author presents an inventory of existing information about the Chernobyl-4 accident. Several possible scenarios are described and a comparison is drawn with the Three Mile Island-2 accident. The author concludes that the event is connected to an inherent instability of the RBMK-1000 reactor type. (G.J.P.)

Accident analysis of Fukushima Daiichi Nuclear Power Station unit 1

International Nuclear Information System (INIS)

Kobayashi, Masahide; Narabayashi, Tadashi; Tsuji, Masashi; Chiba, Go; Nagata, Yasunori; Shimoe, Tomohiro

2015-01-01

As a result of the Great East Japan Earthquake that occurred on 11 March 2011, all AC and DC power at the Fukushima Daiichi NPP units 1 to 3 were lost soon after the tsunami. The core cooling function was lost, and the cores of units 1 to 3 were damaged. The purpose of this work is to clarify the progress of the accident in unit 1, which was damaged the earliest among the 3 units. Therefore, an original severe accident analysis code was developed, and the progress of the accident was evaluated from the analysis results and the actual data. As a result, the leakage path from a pressure vessel was clarified, and some lessons and knowledge were gained. (author)

Development of an accident sequence precursor methodology and its application to significant accident precursors

Energy Technology Data Exchange (ETDEWEB)

Jang, Seung Hyun; Park, Sung Hyun; Jae, Moo Sung [Dept. of of Nuclear Engineering, Hanyang University, Seoul (Korea, Republic of)

2017-03-15

The systematic management of plant risk is crucial for enhancing the safety of nuclear power plants and for designing new nuclear power plants. Accident sequence precursor (ASP) analysis may be able to provide risk significance of operational experience by using probabilistic risk assessment to evaluate an operational event quantitatively in terms of its impact on core damage. In this study, an ASP methodology for two operation mode, full power and low power/shutdown operation, has been developed and applied to significant accident precursors that may occur during the operation of nuclear power plants. Two operational events, loss of feedwater and steam generator tube rupture, are identified as ASPs. Therefore, the ASP methodology developed in this study may contribute to identifying plant risk significance as well as to enhancing the safety of nuclear power plants by applying this methodology systematically.

Accidents involving Brazilian indigenous treated at urgent and emergency services of the Unified Health System.

Science.gov (United States)

Souza, Edinilsa Ramos de; Njaine, Kathie; Mascarenhas, Márcio Dênis Medeiros; Oliveira, Maria Conceição de

2016-12-01

Abstract We analyzed the accidents with Brazilian indigenous treated at urgent and emergency services of the Unified Health System (SUS). Data were obtained from the 2014 Viva Survey, which included 86 services from 24 capitals and the Federal District. The demographic profile of the indigenous, the event and the attendance were characterized. Most of the attended people were male in the 20-39 years age group. Falls and traffic accidents were the main reasons for attendance. Alcohol use was informed by 5.6% of the attended people, a figure that increases to 19.1% in traffic accidents, 26.1% among drivers and 22.8% among motorcyclists. There was a statistical difference between genders in relation to age, disability, place of occurrence of the event, work-related event and victim's condition in the traffic accident. We emphasize the importance of providing visibility to accidents with indigenous and engage them in the prevention of such events. Data reliability depends on the adequate completion in indigenous health information systems.

Accident Sequence Precursor Analysis for SGTR by Using Dynamic PSA Approach

International Nuclear Information System (INIS)

Lee, Han Sul; Heo, Gyun Young; Kim, Tae Wan

2016-01-01

In order to address this issue, this study suggests the sequence tree model to analyze accident sequence systematically. Using the sequence tree model, all possible scenarios which need a specific safety action to prevent the core damage can be identified and success conditions of safety action under complicated situation such as combined accident will be also identified. Sequence tree is branch model to divide plant condition considering the plant dynamics. Since sequence tree model can reflect the plant dynamics, arising from interaction of different accident timing and plant condition and from the interaction between the operator action, mitigation system, and the indicators for operation, sequence tree model can be used to develop the dynamic event tree model easily. Target safety action for this study is a feed-and-bleed (F and B) operation. A F and B operation directly cools down the reactor cooling system (RCS) using the primary cooling system when residual heat removal by the secondary cooling system is not available. In this study, a TLOFW accident and a TLOFW accident with LOCA were the target accidents. Based on the conventional PSA model and indicators, the sequence tree model for a TLOFW accident was developed. Based on the results of a sampling analysis and data from the conventional PSA model, the CDF caused by Sequence no. 26 can be realistically estimated. For a TLOFW accident with LOCA, second accident timings were categorized according to plant condition. Indicators were selected as branch point using the flow chart and tables, and a corresponding sequence tree model was developed. If sampling analysis is performed, practical accident sequences can be identified based on the sequence analysis. If a realistic distribution for the variables can be obtained for sampling analysis, much more realistic accident sequences can be described. Moreover, if the initiating event frequency under a combined accident can be quantified, the sequence tree model

Work accidents and self-esteem of nursing professional in hospital settings.

Science.gov (United States)

Santos, Sérgio Valverde Marques Dos; Macedo, Flávia Ribeiro Martins; Silva, Luiz Almeida da; Resck, Zelia Marilda Rodrigues; Nogueira, Denismar Alves; Terra, Fábio de Souza

2017-04-20

to analyze the occurrence of work accidents and the self-esteem of nurses in hospitals of a municipality of Minas Gerais. descriptive-analytical and cross-sectional study developed with 393 nursing professionals from three hospitals of a municipality in southern Minas Gerais. The Rosenberg Self-Esteem Scale and a questionnaire to characterize the population and work accidents were used for data collection. Data analysis was performed using Person's chi-squared test, Fisher's exact test, Cronbach's alpha, odds ratio and logistic regression. of the professionals studied, 15% had suffered an accident at work and 70.2% presented high self-esteem. Through the analysis, it was observed that smoking, religious belief and an outstanding event in the career were significantly associated with work accidents. In relation to self-esteem, family income, length of time working in the profession and an outstanding event in the career presented significant associations. factors such as smoking, religious belief, family income, length of time working in the profession and an outstanding event in the career can cause professionals to have accidents and/or cause changes in self-esteem, which can compromise their physical and mental health and their quality of life and work.

Stocks and energy shocks : the impact of energy accidents on stock market value

NARCIS (Netherlands)

Scholtens, B.; Boersen, A.

We investigate how financial market participants value energy accidents. We employ an event study to look into the response of stock markets to 209 accidents. These accidents were derived from Sovacool's (2008) database on major energy accidents from 1907 to 2007. It appears that the stock market in

Psychological and social aspects verified after the Goiania's radioactive accident

International Nuclear Information System (INIS)

Helou, Suzana

1995-01-01

Psychological and social aspects verified after the radioactive accident occurred in 1987 in Goiania - brazilian city - are discussed. With this goal was going presented a public opinion research in order to retract the Goiania's radioactive accident residual psychological effects. They were going consolidated data obtained in 1.126 interviews. Four involvement different levels groups with the accident are compared with regard to the event. The research allowed to conclude that the accident affected psychologically somehow all Goiania's population. Besides, the research allowed to analyze the professionals performance quality standard in terms of the accident

Self-similar risk characteristics of industrial accidents

International Nuclear Information System (INIS)

Puzanov, Y.V.

1994-01-01

At the present time there is no logically consistent theory of risk of industrial accidents, just as for the risk of other catastrophic phenomena (natural disasters, ecological castastrophes). Moreover, there is no unique interpretation of the term risk itself in application to catastrophic phenomena, and different authors employ the concept of risk arbitrarily, often proceeding from intuitively obvious ideas. The risk of an accident is most often identified with the probability of the accident itself (with a flux of accident events), the probability of loss of life or damage due to catastrophic phenomena. However, every such concept has its own independent meaning, and identifying these concepts with risk is fraught with confusion

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-07-01

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

United States position on severe accidents

International Nuclear Information System (INIS)

Ross, D.F.

1988-01-01

The United States policy on severe accidents was published in 1985 for both new plant applications and for existing plants. Implementation of this policy is in progress. This policy, aided by a related safety goal policy and by analysis capabilities emerging from improved understanding of accident phenomenology, is viewed as a logical development from the pioneering work in the WASH-1400 Reactor Safety Study published by the United States Nuclear Regulatory Commission (NRC) in 1975. This work provided an estimate of the probability and consequences of severe accidents which, prior to that time, had been mostly evaluated by somewhat arbitrary assumptions dating back 30 years. The early history of severe accident evaluation is briefly summarized for the period 1957-1979. Then, the galvanizing action of Three Mile Island Unit 2 (TMI-2) on severe accident analysis, experimentation and regulation is reviewed. Expressions of US policy in the form of rulemaking, severe accident policy, safety research, safety goal policy and court decisions (on adequacy of safety) are discussed. Finally, the NRC policy as of March 1988 is stated, along with a prospective look at the next few years. (author). 19 refs

The study on development of emergency operating procedures based on symptom and risk for accident management

Energy Technology Data Exchange (ETDEWEB)

Kang, K. S.; Jeong, H. J. [KOPEC, Taejon (Korea, Republic of)

1998-10-01

The Advanced EOP(AEOP) has been developed by focusing on the importance of the operators role in emergency conditions. In the AEOP, to overcome the complexity of current EOPs and maintain the consistency of operators action according to plant emergency conditions, operator's task were allocated according to their duties. As an alternative, the Computerized Operator Aid System (COAS) has been developed to reduce operator's burden and provide detailed instructions of procedure. Probabilistic Safety Assessment (PSA) results were synthesized in the AEOP using the event tree to give the awareness and the prediction of accident progression in advance. In conclusion, the existing EOP with its inherent complexity should be simplified and consolidated using computerized operator support system and task allocation to prevent more severe accidents and to reduce operator cognitive overload in emergency conditions.

Case for integral core-disruptive accident analysis

International Nuclear Information System (INIS)

Luck, L.B.; Bell, C.R.

1985-01-01

Integral analysis is an approach used at the Los Alamos National Laboratory to cope with the broad multiplicity of accident paths and complex phenomena that characterize the transition phase of core-disruptive accident progression in a liquid-metal-cooled fast breeder reactor. The approach is based on the combination of a reference calculation, which is intended to represent a band of similar accident paths, and associated system- and separate-effect studies, which are designed to determine the effect of uncertainties. Results are interpreted in the context of a probabilistic framework. The approach was applied successfully in two studies; illustrations from the Clinch River Breeder Reactor licensing assessment are included

Computer code calculations of the TMI-2 accident: initial and boundary conditions

International Nuclear Information System (INIS)

Behling, S.R.

1985-05-01

Initial and boundary conditions during the Three Mile Island Unit 2 (TMI-2) accident are described and detailed. A brief description of the TMI-2 plant configuration is given. Important contributions to the progression of the accident in the reactor coolant system are discussed. Sufficient information is provided to allow calculation of the TMI-2 accident with computer codes

Japanese Nuclear Accident and U.S. Response

International Nuclear Information System (INIS)

Douet, Randy

2011-01-01

U.S. Government response to the Fukushima accident: • Multi-agency task force (Nuclear Regulatory Commission, Department of Energy, Department of Defense) supporting Japan recovery efforts; • President Obama directed the NRC to perform a comprehensive review of U.S. reactors; • NRC established agency task force to develop lessons learned from Fukushima Daiichi accident to provide short-term and long-term analysis of the events

Safety assurance logic techniques for evaluation of accident prevention and mitigation

International Nuclear Information System (INIS)

McWethy, L.M.; Hagan, J.W.

1976-01-01

Safety assurance methods have been developed and applied in reactor safety assessments of FFTF. These methods promote visibility of the total safety provided by the plant, both in prevention of off-normal or accident conditions as well as provision of various features which terminate conditions within acceptable bounds if such conditions should occur. One of the primary techniques applied in safety assurance is the development of safety assurance diagrams. These diagrams explicitly identify the multiple lines of defense which prevent accident progression. The diagrams graphically demonstrate the defense-in-depth provided by the plant for each postulated occurrence. Lines of defense are shown against ever having an occurrence in the first place; thus giving appropriate emphasis on accident prevention, and visibility to the designer's role in promoting this level of safety. These diagrams, or accident process trees, also show graphically the various paths of postulated accident progression to their logical termination. Evaluation of the importance and strength of each line-of-defense assures fulfillment of the safety objectives of the overall plant system

The development of severe accident analysis technology

Energy Technology Data Exchange (ETDEWEB)

Kim, Heuy Dong; Cho, Sung Won; Kim, Sang Baek; Park, Jong Hwa; Lee, Kyu Jung; Park, Lae Joon; Hu, Hoh; Hong, Sung Wan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1993-07-01

The objective of the development of severe accident analysis technology is to understand the severe accident phenomena such as core melt progression and to provide a reliable analytical tool to assess severe accidents in a nuclear power plant. Furthermore, establishment of the accident management strategies for the prevention/mitigation of severe accidents is also the purpose of this research. The study may be categorized into three areas. For the first area, two specific issues were reviewed to identify the further research direction, that is the natural circulation in the reactor coolant system and the fuel-coolant interaction as an in-vessel and an ex-vessel phenomenological study. For the second area, the MELCOR and the CONTAIN codes have been upgraded, and a validation calculation of the MELCOR has been performed for the PHEBUS-B9+ experiment. Finally, the experimental program has been established for the in-vessel and the ex-vessel severe accident phenomena with the in-pile test loop in KMRR and the integral containment test facilities, respectively. (Author).

SAMEX: A severe accident management support expert

International Nuclear Information System (INIS)

Park, Soo-Yong; Ahn, Kwang-Il

2010-01-01

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

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

Accidents with sulfuric acid

Directory of Open Access Journals (Sweden)

Rajković Miloš B.

2006-01-01

Full Text Available Sulfuric acid is an important industrial and strategic raw material, the production of which is developing on all continents, in many factories in the world and with an annual production of over 160 million tons. On the other hand, the production, transport and usage are very dangerous and demand measures of precaution because the consequences could be catastrophic, and not only at the local level where the accident would happen. Accidents that have been publicly recorded during the last eighteen years (from 1988 till the beginning of 2006 are analyzed in this paper. It is very alarming data that, according to all the recorded accidents, over 1.6 million tons of sulfuric acid were exuded. Although water transport is the safest (only 16.38% of the total amount of accidents in that way 98.88% of the total amount of sulfuric acid was exuded into the environment. Human factor was the common factor in all the accidents, whether there was enough control of the production process, of reservoirs or transportation tanks or the transport was done by inadequate (old tanks, or the accidents arose from human factor (inadequate speed, lock of caution etc. The fact is that huge energy, sacrifice and courage were involved in the recovery from accidents where rescue teams and fire brigades showed great courage to prevent real environmental catastrophes and very often they lost their lives during the events. So, the phrase that sulfuric acid is a real "environmental bomb" has become clearer.

NPP Krsko Severe Accident Management Guidelines Upgrade

International Nuclear Information System (INIS)

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

2014-01-01

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

Proposal strategy and policy on nuclear safety for no-more severe accidents

International Nuclear Information System (INIS)

2013-01-01

Following the outspoken advice saying 'scientists and engineers concerning with nuclear power promotion and safety should be responsible for clarifying how preventable or what measures should be needed to prevent severe accidents occurring at Fukushima Daiichi nuclear power plants (NPPs)', committee on prevention of severe accidents at NPPs was established by relevant nuclear scientists and engineers involved so as to discuss basic issues to be solved from scientific and technical viewpoints. Based on the review of 'defense in depth' concept and accident analysis at Fukushima nuclear accident, four major proposals and six supplements to be established were identified such as: (1) finding mechanism of beyond imagination events for natural disaster, terrorism, and internal events, (2) reform of comprehensive safety standards and guidelines with performance basis easy to reflect latest knowledge and technology as 'back-fitting', (3) severe accidents measures, their validation, and drilling on accident management to advance procedures and develop human resources, and (4) risk communications and public disclosure of information. This article described backgrounds of committee's proposals on nuclear safety for no-more severe accidents. (T. Tanaka)

Analysis of the radiation accident in El Salvador

International Nuclear Information System (INIS)

Melara, N.E.

1998-01-01

On 5 February 1989 at 2 a.m. local time in a cobalt-60 industrial irradiation facility, a series of events started leading to one of the most serious radiation accidents in this type of installation. It took place in Soyapango, a city situated 5 km from San Salvador, the capital of the Republic of El Salvador. In this accident, three workers were involved in the first event and a further four in the second. When the accident took place, the activity level was approximately 0.66 PBq (18,000 Ci). The source became blocked when being lowered to its safe position, where upon the technician responsible for the irradiator entered the chamber in breach of the few inadequate safety procedures, accompanied by two colleagues from an adjacent department; the three workers suffered acute radiation exposure, with the result that one of them died six-and-a-half months later, the second had both his legs amputated at mid-thigh, while the third recovered completely. This article describes the irradiator, outlines the causes of the accident and analyses the economic and social repercussions, with the aim of helping teams responsible for radiation protection and safety in industrial irradiation facilities to identify potentially hazardous circumstances and avoid accidents. (author)

Real and mythical consequences of Chernobyl accident

International Nuclear Information System (INIS)

Osmachkin, V.S.

1999-01-01

This presentation describes the public Unacceptance of Nuclear Power as a consequence of Chernobyl Accident, an accident which was a severest event in the history of the nuclear industry. It was a shock for everybody, who has been involved in nuclear power programs. But nobody could expect that it was also the end romantic page in the nuclear story. The scale of the detriment was a great, and it could be compared with other big technological man-made catastrophes. But immediately after an accident mass media and news agencies started to transmit an information with a great exaggerations of the consequences of the event. In a report on the Seminar T he lessons of the Chernobyl - 1' in 1996 examples of such incorrect information, were cited. Particularly, in the mass media it was declared that consequences of the accident could be compared with a results of the second world war, the number of victims were more than hundred thousand people, more than million of children have the serious health detriments. Such and other cases of the misconstruction have been called as myths. The real consequences of Chernobyl disaster have been summed on the International Conference 'One decade after Chernobyl' - 2, in April 1996. A very important result of the Chernobyl accident was a dissemination of stable unacceptance of the everything connected with 'the atom'. A mystic horror from invisible mortal radiation has been inspired in the masses. And from such public attitude the Nuclear Power Programs in many countries have changed dramatically. A new more pragmatic and more careful atomic era started with a slogan: 'Kernkraftwerk ? Nein, danke'. No doubt, a Chernobyl accident was a serious technical catastrophe in atomic industry. The scale of detriment is connected with a number of involved peoples, not with a number of real victims. In comparison with Bhopal case, earthquakes, crashes of the airplanes, floods, traffic accidents and other risky events of our life - the Chernobyl is

Probabilistic risk assessment using event tables and the BNL [Brookhaven National Laboratory] event-tree analyzer

International Nuclear Information System (INIS)

Fullwood, R.R.; Shier, W.G.

1989-01-01

Probabilistic risk analysis (PRA) is being used to study design alternatives for the advanced neutron source research reactor being designed at Oak Ridge National Laboratory for operation in the 1990s. Major communication paths between the designers and the safety analysts are accident discussions supported by event tables, event-tree graphics, and accident sequence probabilities. The BETA code used in conjunction with a word processor provides this linkage. This paper describes the process, features of the BETA, how it works, and some examples of usage

The IAEA Accident Management Programme

Energy Technology Data Exchange (ETDEWEB)

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

1993-02-01

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

The IAEA Accident Management Programme

International Nuclear Information System (INIS)

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

1993-01-01

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

Severe accident analysis for level 2 PSA of SMART reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Jin Yong; Lee, Jeong Hun; Kim, Jong Uk; Yoo, Tae Geun; Chung, Soon Il; Kim, Min Gi [FNC Technology Co., Seoul (Korea, Republic of)

2010-12-15

The objectives of this study are to produce data for level 2 PSA and evaluation results of severe accident by analyzing severe accident sequence of transient events, producing fault tree of containment systems and evaluating direct containment heating of the SMART. In this project, severe accident analysis results were produced for general transient, loss of feedwater, station blackout, and steam line break events, and based on the results, design safety of SMART was verified. Also, direct containment heating phenomenon of the SMART was evaluated using TCE methodology. For level 2 PSA, fault tree of the containment isolation system, reactor cavity flooding system, plant chilled water system, and reactor containment building HVAC system was produced and analyzed

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

Directory of Open Access Journals (Sweden)

Marcos Rosa Costa

2013-06-01

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

The External Cost Evaluation of the Nuclear Severe Accident Using CVM

International Nuclear Information System (INIS)

Lee, Yong Suk; Lee, Byung Chul

2006-01-01

The external cost of energy can be defined as 'the cost not included in the energy market price', such as air pollution, noise, etc. Within the evaluation of the external cost of nuclear energy, the estimation of the external cost of severe accident is one of the major topics to be addressed. For the evaluation of the external cost of severe accident, the effect of risk aversion of the public against the severe accident must be addressed, because people are more concerned about low probability - high consequence events than about high probability - low consequence events having the same mean damage. It is generally recognized that there is a discrepancy between the social acceptability of the risk and the average monetary value which corresponds in principle to the compensation of the consequences for each individual of the population affected by the accident. In this paper, the CVM (Contingent Valuation Method) is used to integrate the risk aversion in the external costs of nuclear severe accidents in Korea

Accidents involving specialized aircraft in agriculture aerial spraying

Directory of Open Access Journals (Sweden)

Marcelo Boamorte Ravelli

Full Text Available ABSTRACT: The great challenge for the practice of agricultural aviation has been to avoid accidents. Although, there are technological progress and high resources for safety, accidents continue to occur. The objective of this research was to analyze the influence and occurrence of factors in agricultural aviation accidents in Brazil recently. Based on research and technical - scientific papers written by researchers and aviation authorities, recommendations directed towards reducing the risks associated with this aircraft modality are assessed. The main factors responsible for accidents are normally operational errors and maneuvers that cause flight collisions, engine failures and altitude loss. Professional awareness and qualification converge towards the success of the agricultural pilot in the detection of inherent dangers or occasional in the various systems involved.

System response of a DOE Defense Program package in a transportation accident environment

International Nuclear Information System (INIS)

Chen, T.F.; Hovingh, J.; Kimura, C.Y.

1992-01-01

The system response in a transportation accident environment is an element to be considered in an overall Transportation System Risk Assessment (TSRA) framework. The system response analysis uses the accident conditions and the subsequent accident progression analysis to develop the accident source term, which in turn, is used in the consequence analysis. This paper proposes a methodology for the preparation of the system response aspect of the TSRA

National practices in relation to severe accidents

International Nuclear Information System (INIS)

Soda, Kunihisa

1989-01-01

After the accidents at Three Mile Island and Chernobyl, many studies have been carried out on severe accidents by various organizations including IAEA and OECD/CSNI. In the present article, measures taken in different countries against severe accidents are outlined based on the results of these studies. In Sweden, policies for the management of a severe accident and reduction in the release of radioactive materials were established based on reports issued by the Atomic Energy Committee, which was set up after the Three Mile Island accident. The current policies require that filter vents be provided where necessary. France, following Sweden, adopted the use of filter vents. Operation procedures to be followed in the event of a severe accident have been established in the nation. The measures against severe accidents adopted in West Germany mainly focus on the weakening of the effects of accidents, and are not covered by the design standards. The use of filter vents are also required in Finland and Switzerland. In the U.S., a program for individual plant examination will be implemented over the three-year period beginning in 1989. Studies on measures against severe accidents seem to be performed also in the Soviet Union. (N.K.)

Goiania radiation accident: activities carried out and lessons learned based on personal experience

International Nuclear Information System (INIS)

Silva, F.C.A. da

2017-01-01

Goiânia Radiological Accident, on September 13, 1987, with a radioactive source of cesium-137 with 50.9 TBq, used in radiotherapy, is one of the most important accidents in the scientific area, representing a milestone for all workers in the areas of radiation protection and radiological emergency that worked during the event. A personal view of the Goiânia Radiological Accident is presented, showing some activities carried out in contaminated areas and lessons learned based on own experience during the event

Work accidents and self-esteem of nursing professional in hospital settings

Directory of Open Access Journals (Sweden)

Sérgio Valverde Marques dos Santos

Full Text Available Abstract Objective: to analyze the occurrence of work accidents and the self-esteem of nurses in hospitals of a municipality of Minas Gerais. Method: descriptive-analytical and cross-sectional study developed with 393 nursing professionals from three hospitals of a municipality in southern Minas Gerais. The Rosenberg Self-Esteem Scale and a questionnaire to characterize the population and work accidents were used for data collection. Data analysis was performed using Person's chi-squared test, Fisher's exact test, Cronbach's alpha, odds ratio and logistic regression. Results: of the professionals studied, 15% had suffered an accident at work and 70.2% presented high self-esteem. Through the analysis, it was observed that smoking, religious belief and an outstanding event in the career were significantly associated with work accidents. In relation to self-esteem, family income, length of time working in the profession and an outstanding event in the career presented significant associations. Conclusion: factors such as smoking, religious belief, family income, length of time working in the profession and an outstanding event in the career can cause professionals to have accidents and/or cause changes in self-esteem, which can compromise their physical and mental health and their quality of life and work.

Work accidents and self-esteem of nursing professional in hospital settings

Science.gov (United States)

dos Santos, Sérgio Valverde Marques; Macedo, Flávia Ribeiro Martins; da Silva, Luiz Almeida; Resck, Zelia Marilda Rodrigues; Nogueira, Denismar Alves; Terra, Fábio de Souza

2017-01-01

Abstract Objective: to analyze the occurrence of work accidents and the self-esteem of nurses in hospitals of a municipality of Minas Gerais. Method: descriptive-analytical and cross-sectional study developed with 393 nursing professionals from three hospitals of a municipality in southern Minas Gerais. The Rosenberg Self-Esteem Scale and a questionnaire to characterize the population and work accidents were used for data collection. Data analysis was performed using Person's chi-squared test, Fisher's exact test, Cronbach's alpha, odds ratio and logistic regression. Results: of the professionals studied, 15% had suffered an accident at work and 70.2% presented high self-esteem. Through the analysis, it was observed that smoking, religious belief and an outstanding event in the career were significantly associated with work accidents. In relation to self-esteem, family income, length of time working in the profession and an outstanding event in the career presented significant associations. Conclusion: factors such as smoking, religious belief, family income, length of time working in the profession and an outstanding event in the career can cause professionals to have accidents and/or cause changes in self-esteem, which can compromise their physical and mental health and their quality of life and work. PMID:28443993

Statistical analysis of the early phase of SBO accident for PWR

Energy Technology Data Exchange (ETDEWEB)

Kozmenkov, Yaroslav, E-mail: y.kozmenkov@hzdr.de; Jobst, Matthias, E-mail: m.jobst@hzdr.de; Kliem, Soeren, E-mail: s.kliem@hzdr.de; Schaefer, Frank, E-mail: f.schaefer@hzdr.de; Wilhelm, Polina, E-mail: p.wilhelm@hzdr.de

2017-04-01

Highlights: • Best estimate model of generic German PWR is used in ATHLET-CD simulations. • Uncertainty and sensitivity analysis of the early phase of SBO accident is presented. • Prediction intervals for occurrence of main events are evaluated. - Abstract: A statistical approach is used to analyse the early phase of station blackout accident for generic German PWR with the best estimate system code ATHLET-CD as a computation tool. The analysis is mainly focused on the timescale uncertainties of the accident events which can be detected at the plant. The developed input deck allows variations of all input uncertainty parameters relevant to the case. The list of identified and quantified input uncertainties includes 30 parameters related to the simulated physical phenomena/processes. Time uncertainties of main events as well as the major contributors to these uncertainties are defined. The uncertainty in decay heat has the highest contribution to the uncertainties of the analysed events. A linear regression analysis is used for predicting times of future events from detected times of occurred/past events. An accuracy of event predictions is estimated and verified. The presented statistical approach could be helpful for assessing and improving existing or elaborating additional emergency operating procedures aimed to prevent severe damage of reactor core.

Research investigation report on Fukushima Daiichi nuclear accident

International Nuclear Information System (INIS)

2012-03-01

This report was issued in February 2012 by Rebuild Japan Initiative Foundation's Independent Investigation Commission on the Fukushima Daiichi Nuclear Accident, which consisted of six members from the private sector in independent positions and with no direct interest in the business of promoting nuclear power. Commission aimed to determine the truth behind the accident by clarifying the various problems and reveal systematic problems behind these issues so as to create a new starting point by identifying clear lessons learned. Report composed of four chapters; (1) progression of Fukushima accident and resulting damage (accident management after Fukushima accident, and effects and countermeasure of radioactive materials discharged into the environment), (2) response against Fukushima accident (emergency response of cabinet office against nuclear disaster, risk communication and on-site response against nuclear disaster), (3) analysis of historical and structural factors (technical philosophy of nuclear safety, problems of nuclear safety regulation of Fukushima accident, safety regulatory governance and social background of 'Safety Myth'), (4) Global Context (implication in nuclear security, Japan in nuclear safety regime, U.S.-Japan relations for response against Fukushima accident, lessons learned from Fukushima accident - aiming at creation of resilience). Report could identify causes of Fukushima accident and factors related to resulting damages, show the realities behind failure to prevent the spread of damage, and analyze the overall structural and historical background behind the accidents. (T. Tanaka)

Development of intervention levels for the protection of the public in the event of a major nuclear accident. Past, present and future

International Nuclear Information System (INIS)

Emmerson, B.W.

1989-01-01

Since the mid-1950's nuclear energy has played an increasing role in meeting the world demand for electricity production. Although during this period incidents and accidents have occurred, in most cases their effect was confined to the plant. Three accidents, however, were sufficienty serious as to involve off-site consequences for the public. The experience from each contributed significantly in the development of current emergency response criteria and planning arrangements at the national and international level. This paper summarizes these contributions as they relate to the development of intervention levels for the protection of the public in the event of an accidental release of radioactive materials to the environment. It indicates the various measures taken by those countries that were affected by the release from the Chernobyl accident and reviews the subsequent actions by relevant international organizations to provide more comprehensive guidance on applying the principles of intervention and developing derived levels, particularly those aimed at controlling the consumption of contamined foodstuffs, or their movement in international trade. Finally, it considers the prospects for developing a more harmonized intervention approach based on the guidance now being completed at the international level [fr

Severe accident management. Optimized guidelines and strategies

International Nuclear Information System (INIS)

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

2014-01-01

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

Uranium storage bed accident hazards evaluation

International Nuclear Information System (INIS)

Longhurst, G.R.; Shmayda, W.T.

1989-01-01

To properly assess hazards and risks associated with the use of uranium beds as tritium storage devices in fusion reactor systems, it is necessary to understand the consequences occurring in the event of an accident. Accidents involving uranium beds are postulated, and the possible results are considered. A research program to more fully and accurately understand those results has been initiated involving the Idaho National Engineering Laboratory and Ontario Hydro. The plan and objectives of that program are presented. 11 refs., 1 tab

Uranium storage bed accident hazards evaluation

International Nuclear Information System (INIS)

Longhurst, G.R.; Shmayda, W.T.

1989-10-01

To properly assess hazards and risks associated with the use of uranium beds as tritium storage devices in fusion reactor systems, it is necessary to understand the consequences occurring in the event of an accident. Accidents involving uranium beds are postulated, and the possible results are considered. A research program to more fully and accurately understand those results has been initiated involving the Idaho National Engineering Laboratory and Ontario Hydro. The plan and objectives of that program are presented. 11 refs., 1 tab

Type A Accident Investigation Board report on the January 17, 1996, electrical accident with injury in Technical Area 21 Tritium Science and Fabrication Facility Los Alamos National Laboratory. Final report

International Nuclear Information System (INIS)

1996-04-01

An electrical accident was investigated in which a crafts person received serious injuries as a result of coming into contact with a 13.2 kilovolt (kV) electrical cable in the basement of Building 209 in Technical Area 21 (TA-21-209) in the Tritium Science and Fabrication Facility (TSFF) at Los Alamos National Laboratory (LANL). In conducting its investigation, the Accident Investigation Board used various analytical techniques, including events and causal factor analysis, barrier analysis, change analysis, fault tree analysis, materials analysis, and root cause analysis. The board inspected the accident site, reviewed events surrounding the accident, conducted extensive interviews and document reviews, and performed causation analyses to determine the factors that contributed to the accident, including any management system deficiencies. Relevant management systems and factors that could have contributed to the accident were evaluated in accordance with the guiding principles of safety management identified by the Secretary of Energy in an October 1994 letter to the Defense Nuclear Facilities Safety Board and subsequently to Congress

Type A Accident Investigation Board report on the January 17, 1996, electrical accident with injury in Technical Area 21 Tritium Science and Fabrication Facility Los Alamos National Laboratory. Final report

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-04-01

An electrical accident was investigated in which a crafts person received serious injuries as a result of coming into contact with a 13.2 kilovolt (kV) electrical cable in the basement of Building 209 in Technical Area 21 (TA-21-209) in the Tritium Science and Fabrication Facility (TSFF) at Los Alamos National Laboratory (LANL). In conducting its investigation, the Accident Investigation Board used various analytical techniques, including events and causal factor analysis, barrier analysis, change analysis, fault tree analysis, materials analysis, and root cause analysis. The board inspected the accident site, reviewed events surrounding the accident, conducted extensive interviews and document reviews, and performed causation analyses to determine the factors that contributed to the accident, including any management system deficiencies. Relevant management systems and factors that could have contributed to the accident were evaluated in accordance with the guiding principles of safety management identified by the Secretary of Energy in an October 1994 letter to the Defense Nuclear Facilities Safety Board and subsequently to Congress.

Prediction of road accidents: A Bayesian hierarchical approach

DEFF Research Database (Denmark)

Deublein, Markus; Schubert, Matthias; Adey, Bryan T.

2013-01-01

the expected number of accidents in which an injury has occurred and the expected number of light, severe and fatally injured road users. Additionally, the methodology is used for geo-referenced identification of road sections with increased occurrence probabilities of injury accident events on a road link......In this paper a novel methodology for the prediction of the occurrence of road accidents is presented. The methodology utilizes a combination of three statistical methods: (1) gamma-updating of the occurrence rates of injury accidents and injured road users, (2) hierarchical multivariate Poisson......-lognormal regression analysis taking into account correlations amongst multiple dependent model response variables and effects of discrete accident count data e.g. over-dispersion, and (3) Bayesian inference algorithms, which are applied by means of data mining techniques supported by Bayesian Probabilistic Networks...

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)

Alternative evacuation strategies for nuclear power accidents

International Nuclear Information System (INIS)

Hammond, Gregory D.; Bier, Vicki M.

2015-01-01

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

Formulating the Canadian regulatory position on severe accidents

International Nuclear Information System (INIS)

Viktorov, Alex

2006-01-01

In response to the increasing potential of new nuclear build in Canada, and as part of documentation harmonization effort, CNSC staff has initiated development of requirements for design of nuclear power plants. These requirements build both on the IAEA standards, most notably, NS-R-1, and the Canadian practices and experience. The three safety objectives, formulated by the IAEA, are adopted, and Safety Goals are proposed consistent with the international trend. This Canadian standard will require, for the first time, explicit consideration of severe accidents in design and safety assessments. Specific requirements are formulated for several plant systems that assure an effective fourth level of defence in depth. Available results from probabilistic safety assessments indicate that the risks posed by severe accidents are acceptably low. Nevertheless, such risks are not negligible. CNSC staff considers that severe accident management (SAM) represents the most practical way to achieve risk reduction with a moderate effort. Ultimately, SAM actions are aimed at bringing the reactor, and the plant in general, into a controlled and stable state. For the operating reactors, SAM provides an additional defense barrier against the consequences of those accidents that fall beyond the scope of events considered in the reactor design basis. The establishment of a SAM program ensures availability of the information, procedures, and resources necessary to take full advantage of existing plant capabilities to arrest core degradation, and prevent or mitigate large releases of radioactive material. To the extent practicable, a SAM program builds on the existing emergency operating procedures and makes use of the plant design capabilities. On this basis, the CNSC requested nuclear power reactor licensees to develop and implement SAM at all operating reactors. To be able to demonstrate compliance with requirements for plant design and severe accident management, it is necessary to

Analysis of hot leg natural circulation under station blackout severe accident

International Nuclear Information System (INIS)

Deng Jian; Cao Xuewu

2007-01-01

Under severe accidents, natural circulation flows are important to influence the accident progression and result in a pressurized water reactor (PWR). In a station blackout accident with no recovery of steam generator (SG) auxiliary feedwater (TMLB' severe accident scenario), the hot leg countercurrent natural circulation flow is analyzed by using a severe-accident code, to better understand its potential impacts on the creep-rupture timing among the surge line, the hot leg; and SG tubes. The results show that the natural circulation may delay the failure time of the hot leg. The recirculation ratio and the hot mixing factor are also calculated and discussed. (authors)

Device for bonding iodine in the event of nuclear reactor accidents

International Nuclear Information System (INIS)

Hladik, O.

1988-01-01

A device for bonding iodine, in particular radioiodine released during nuclear reactor accidents, is presented. Radioiodine is bonded, even at high temperatures, so that it is neither volatile nor soluble

Radiation protection experience in Yugoslavia from the Vinca accident to nowadays

International Nuclear Information System (INIS)

Ninkovic, M.M

2000-01-01

This Paper is the expression of the author opinion about development of radiation protection in Yugoslavia from its beginning forty years ago, which might affect its status in the foreseeable future at the first decades of the 21st century. It focuses on key events in this field starting from the Vinca Accident, which happened in the October 1958, to nowadays. Shortly reviewed some of key events are: Vinca Accident; Foundation of the Radiation Protection Laboratory in the Vinca Institute; International Vinca Dosimetry Experiment; First National Symposium and foundation of the Yugoslav Radiation Protection Association; International Intercomparison Experiment on Nuclear Accident Dosimetry and, International Summer Schools and Symposium on Radiation Protection organized in Yugoslavia. Finally, some specific experimental data obtained during and after Chernobyl Accident up to nowadays in radiation protection action in Yugoslavia are presented also. (author)

Analysis of Three Mile Island Unit 2 accident

International Nuclear Information System (INIS)

Anon.

1979-01-01

NSAC is conducting a detailed review of this accident and of the lessons to be learned. So far it has concentrated primarily on events during the sixteen hours following initiation of the accident. A sequence of events has been developed and is being verified and annotated by comparing oral and written statements with instrumentation records, data logs, operator logs, and inferences which can be made from these records. This report is being developed with the expectation that, while not completed or fully verified, it may be useful at this time. Supplements may be issued later as the analyses which are still under way are completed

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

International Nuclear Information System (INIS)

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

2004-01-01

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

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

Accidents and incidents with external and/or internal radiation-exposure

International Nuclear Information System (INIS)

Anon.

1990-01-01

An individual radiation exposure accident is an unexpected and unintended event which gives rise to an overexposure (external or internal). By overexposure one means all external or internal exposure which could lead to the exceeding of the regulatory norms. Going beyond these limits does not always produce pathological manifestations. The term radiological accident is, in practice, used only when there is an occurrence of some biological or clinical response, or when some therapeutic action is required. A radio-exposure accident can occur: within or from a nuclear power plant or from a center employing ionizing radiation. These are the most frequent; and during the transport of radioactive materials. These are exceptional events. The tanks and containers used in the transport of highly radioactive substances are exhaustively studied for their resistance to accidents, and the conditions of transport determined by very strict national and international regulations. The transport of substances of low radioactivity (labelled molecules, radiopharmaceuticals...) carries only minor risks

Jose Cabrera NPP severe accident management activities

International Nuclear Information System (INIS)

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

1998-01-01

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

In vessel core melt progression phenomena

International Nuclear Information System (INIS)

Courtaud, M.

1993-01-01

For all light water reactor (LWR) accidents, including the so called severe accidents where core melt down can occur, it is necessary to determine the amount and characteristics of fission products released to the environment. For existing reactors this knowledge is used to evaluate the consequences and eventual emergency plans. But for future reactors safety authorities demand decrease risks and reactors designed in such a way that fission products are retained inside the containment, the last protective barrier. This requires improved understanding and knowledge of all accident sequences. In particular it is necessary to be able to describe the very complex phenomena occurring during in vessel core melt progression because they will determine the thermal and mechanical loads on the primary circuit and the timing of its rupture as well as the fission product source term. On the other hand, in case of vessel failure, knowledge of the physical and chemical state of the core melt will provide the initial conditions for analysis of ex-vessel core melt progression and phenomena threatening the containment. Finally a good understanding of in vessel phenomena will help to improve accident management procedures like Emergency Core Cooling System water injection, blowdown and flooding of the vessel well, with their possible adverse effects. Research and Development work on this subject was initiated a long time ago and is still in progress but now it must be intensified in order to meet the safety requirements of the next generation of reactors. Experiments, limited in scale, analysis of the TMI 2 accident which is a unique source of global information and engineering judgment are used to establish and assess physical models that can be implemented in computer codes for reactor accident analysis

The influence of chemistry on severe accident phenomena in integral tests

International Nuclear Information System (INIS)

Hobbins, R.R.; Osetek, D.J.; Hagrman, D.L.

1988-01-01

The influence of chemical processes on severe accident phenomena in integral tests is reviewed and recommendations for areas of additional work are made. The results reviewed include those from tests conducted in the in-pile facilities at ACRR, PBF, and TREAT and the TMI-2 accident. Progress has been made in understanding the influence of chemistry on important severe accident phenomena such as core melt progression, hydrogen generation, aerosol generation and transport, and fission product release and transport (including revaporization). An example is the chemistry of volatile fission products, especially iodine and tellurium. Areas where understanding is inadequate are also apparent, such as chemical interactions between fission product vapors and aerosols. Influential chemical processes reviewed include oxidation by steam and interactions among control, structural, fuel, fission product, and aerosol materials

APRI-7 Accident Phenomena of Risk Importance. A progress report on research in the field of severe accidents in 2009-2011; APRI-7 Accident Phenomena of Risk Importance. En laegesrapport om forskningen inom omraadet svaara haverier under aaren 2009-2011

Energy Technology Data Exchange (ETDEWEB)

Garis, Ninos; Agrell, Maria [SSM, Stockholm (Sweden); Glaenneskog, Henrik [Vattenfall Research and Development AB, Aelvkarleby (Sweden)] [and others

2012-11-01

Knowledge of the phenomena that may occur during severe accidents in a nuclear power plant is an important prerequisite for being able to predict the plant behavior, in order to formulate procedures and instructions for incident handling, for contingency planning, and to get good quality at the accident analysis and risk studies. Since the early 80's nuclear power companies and authorities in Sweden has collaborated in research on severe reactor accidents. Cooperation in the beginning was mostly linked to strengthening the protection against environmental impacts after a severe reactor accident, in particular to develop systems for filtered depressurization of the reactor containment. Since the early 90's the cooperation has partially changed and shifted to the phenomenological questions of risk dominance. During the years 2009-2011, cooperation continued in the research-program APRI-7. The aim was to show whether the solutions adopted in the Swedish strategy for accident management provides reasonable protection for the environment. This was done by gaining detailed knowledge of both important phenomena in the hearth melting behavior, and the amount of radioactivity that can be discharged to the surroundings during a severe accident. To achieve this aim, the research program has included a follow-up of international research in severe accidents and evaluation of results, and continued to support research at KTH and Chalmers Univ. of severe accidents. The follow-up of international research has promoted the exchange of knowledge and experience and has provided access to a wealth of information about various phenomena relevant to the events at severe accidents. This was important to obtain a good basis for assessment of abatement measures in the Swedish nuclear reactors. Continuing support to the Royal Inst. of Technology has provided increased knowledge about the ability to cool the molten core of the reactor vessel and the processes associated with

Proposed classification scale for radiological incidents and accidents; Elaboration d'une echelle de classement des incidents et accidents radiologiques

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-04-15

The scale proposed in this report is intended to facilitate communication concerning the severity of incidents and accidents involving the exposure of human beings to ionising radiations. Like the INES, it comprises eight levels of severity and uses the same terms (accident, incident, anomaly, serious and major) for keeping the public and the media informed. In a radiological protection context, the severity of an event is considered as being directly proportional to the risk run by an individual (the probability of developing fatal or non-fatal health effects) following exposure to ionising radiation in an incident or accident situation. However for society, other factors have to be taken into account to determine severity. The severity scale proposed is therefore based on assessment of the individual radiological risk. A severity level corresponding to exposure of a member of the public in an incident or accident situation is determined on the basis of risk assessment concepts and methods derived from international consensus on dose/effect relationships for both stochastic and deterministic effects. The severity of all the possible exposure situations - worker exposure, collective exposure, potential exposure - is determined using a system of weighting in relation to situations involving members of the public. In the case of this scale, to indicate the severity of an event, it is proposed to make use of the most penalizing level of severity, comparing: - the severity associated with the probability of occurrence of deterministic effects and the severity associated with the probability of occurrence of stochastic effects, when the event gives rise to both types of risk; - the severity for members of the public and the severity for exposed workers, when both categories of individuals are involved; - the severity on the proposed radiological protection scale and that obtained using the INES, when radiological protection and nuclear safety aspects are associated with

A modular structure to accident identification using neural networks

International Nuclear Information System (INIS)

Duque Estrada, Cassius Rodrigo

2005-01-01

This work uses the accident identification method based on Artificial Neural Networks (ANN) as basic blocks of a modular structure, allowing the inclusion of new accidents to be identified without modifying the ANN already trained. This structure comprises several modules for accident identification and one module for analysis. Each identification module follows the structure of the basic block. The identification modules are responsible for the recognition of an accident belonging to the specific set of events for which it were trained. The analysis module processes the output from the identification module to determine the system response. In order to test this structure it was proposed a transient identification problem comprising fifty accidents distributed in five identification modules. The results have demonstrated that the accident identification method used as basic block of a modular structure allows the inclusion of new sets of accidents, or variations of a same accident, without modifying the ANN already trained. For this, it is enough to include into the system an specific module for this new set of accidents. (author)

NPP Krsko Severe Accident Management Guidelines Implementation

International Nuclear Information System (INIS)

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

2002-01-01

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

Serious reactor accidents reconsidered

International Nuclear Information System (INIS)

1987-12-01

The chance is determined for damage of the reactor core and that sequel events will cause excursion of radioactive materials into the environment. The gravity of such an accident is expressed by the source term. It appears that the chance for such an accident varies with the source term. In general it is valid that how larger the source term how smaller the chance is for it and vice versa. The chance for excursion is related to two complexes of events: serious damage (meltdown) of the reactor core, and the escape of the liberated radionuclides into the environment. The results are an order of magnitude consideration of the relation between the extent of the source term and the chance for it. From the spectrum of possible source terms three representative ones have been chosen: a large, a medium and a relative small source term. This choice is in accordance with international considerations. The hearth of this study is the estimation of the chance for occurrence of the three chosen source terms for new light-water reactors. refs.; figs.; tabs

HTGR accident initiation and progression analysis status report. Volume VIII. Responses to comments on AIPA status report

Energy Technology Data Exchange (ETDEWEB)

Raabe, P.H.

1977-01-01

The first seven volumes of the report series provide formal documentation of the status of the ERDA-sponsored Accident Initiation and Progression Analysis (AIPA) study as of the end of FY75. That portion of the report was given broad distribution to government agencies, industrial organizations, and academic institutions. Comments on the Status Report have been actively solicited from these and other organizations. The volume presented (the eighth in the AIPA Status Report) documents all of the formal written comments that have been received as of September 30, 1976, together with the responses to those comments. The comments as presented are direct quotations from the manuscripts as submitted by the reviewers; none have been paraphrased. The comments are presented in the same order as submitted by the reviewers and are generally addressed individually.

HTGR accident initiation and progression analysis status report. Volume VIII. Responses to comments on AIPA status report