Accident sequence precursor events with age-related contributors

Energy Technology Data Exchange (ETDEWEB)

Murphy, G.A.; Kohn, W.E.

1995-12-31

The Accident Sequence Precursor (ASP) Program at ORNL analyzed about 14.000 Licensee Event Reports (LERs) filed by US nuclear power plants 1987--1993. There were 193 events identified as precursors to potential severe core accident sequences. These are reported in G/CR-4674. Volumes 7 through 20. Under the NRC Nuclear Plant Aging Research program, the authors evaluated these events to determine the extent to which component aging played a role. Events were selected that involved age-related equipment degradation that initiated an event or contributed to an event sequence. For the 7-year period, ORNL identified 36 events that involved aging degradation as a contributor to an ASP event. Except for 1992, the percentage of age-related events within the total number of ASP events over the 7-year period ({approximately}19%) appears fairly consistent up to 1991. No correlation between plant ape and number of precursor events was found. A summary list of the age-related events is presented in the report.

BWR severe accident sequence analyses at ORNL - some lessons learned

International Nuclear Information System (INIS)

Hodge, S.A.

1983-01-01

Boiling water reactor severe accident sequence studies are being carried out using Browns Ferry Unit 1 as the model plant. Four accident studies were completed, resulting in recommendations for improvements in system design, emergency procedures, and operator training. Computer code improvements were an important by-product

Human factors review for Severe Accident Sequence Analysis (SASA)

International Nuclear Information System (INIS)

Krois, P.A.; Haas, P.M.; Manning, J.J.; Bovell, C.R.

1984-01-01

The paper will discuss work being conducted during this human factors review including: (1) support of the Severe Accident Sequence Analysis (SASA) Program based on an assessment of operator actions, and (2) development of a descriptive model of operator severe accident management. Research by SASA analysts on the Browns Ferry Unit One (BF1) anticipated transient without scram (ATWS) was supported through a concurrent assessment of operator performance to demonstrate contributions to SASA analyses from human factors data and methods. A descriptive model was developed called the Function Oriented Accident Management (FOAM) model, which serves as a structure for bridging human factors, operations, and engineering expertise and which is useful for identifying needs/deficiencies in the area of accident management. The assessment of human factors issues related to ATWS required extensive coordination with SASA analysts. The analysis was consolidated primarily to six operator actions identified in the Emergency Procedure Guidelines (EPGs) as being the most critical to the accident sequence. These actions were assessed through simulator exercises, qualitative reviews, and quantitative human reliability analyses. The FOAM descriptive model assumes as a starting point that multiple operator/system failures exceed the scope of procedures and necessitates a knowledge-based emergency response by the operators. The FOAM model provides a functionally-oriented structure for assembling human factors, operations, and engineering data and expertise into operator guidance for unconventional emergency responses to mitigate severe accident progression and avoid/minimize core degradation. Operators must also respond to potential radiological release beyond plant protective barriers. Research needs in accident management and potential uses of the FOAM model are described. 11 references, 1 figure

Human factors review for nuclear power plant severe accident sequence analysis

International Nuclear Information System (INIS)

Krois, P.A.; Haas, P.M.

1985-01-01

The paper discusses work conducted to: (1) support the severe accident sequence analysis of a nuclear power plant transient based on an assessment of operator actions, and (2) develop a descriptive model of operator severe accident management. Operator actions during the transient are assessed using qualitative and quantitative methods. A function-oriented accident management model provides a structure for developing technical operator guidance on mitigating core damage preventing radiological release

Accident sequence precursor analysis level 2/3 model development

International Nuclear Information System (INIS)

Lui, C.H.; Galyean, W.J.; Brownson, D.A.

1997-01-01

The US Nuclear Regulatory Commission's Accident Sequence Precursor (ASP) program currently uses simple Level 1 models to assess the conditional core damage probability for operational events occurring in commercial nuclear power plants (NPP). Since not all accident sequences leading to core damage will result in the same radiological consequences, it is necessary to develop simple Level 2/3 models that can be used to analyze the response of the NPP containment structure in the context of a core damage accident, estimate the magnitude of the resulting radioactive releases to the environment, and calculate the consequences associated with these releases. The simple Level 2/3 model development work was initiated in 1995, and several prototype models have been completed. Once developed, these simple Level 2/3 models are linked to the simple Level 1 models to provide risk perspectives for operational events. This paper describes the methods implemented for the development of these simple Level 2/3 ASP models, and the linkage process to the existing Level 1 models

Accident Sequence Evaluation Program: Human reliability analysis procedure

International Nuclear Information System (INIS)

Swain, A.D.

1987-02-01

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

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)

Accident Sequence Evaluation Program: Human reliability analysis procedure

Energy Technology Data Exchange (ETDEWEB)

Swain, A.D.

1987-02-01

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

Event sequence quantification for a loss of shutdown cooling accident in the GCFR

International Nuclear Information System (INIS)

Frank, M.; Reilly, J.

1979-10-01

A summary is presented of the core-wide sequence of events of a postulated total loss of forced and natural convection decay heat removal in a shutdown Gas-Cooled Fast Reactor (GCFR). It outlines the analytical methods and results for the progression of the accident sequence. This hypothetical accident proceeds in the distinct phases of cladding melting, assembly wall melting and molten steel relocation into the interassembly spacing, and fuel relocation. It identifies the key phenomena of the event sequence and the concerns and mechanisms of both recriticality and recriticality prevention

Source terms associated with two severe accident sequences in a 900 MWe PWR

International Nuclear Information System (INIS)

Fermandjian, J.; Evrard, J.M.; Berthion, Y.; Lhiaubet, G.; Lucas, M.

1983-12-01

Hypothetical accidents taken into account in PWR risk assessment result in fission product release from the fuel, transfer through the primary circuit, transfer into the reactor containment building (RCB) and finally release to the environment. The objective of this paper is to define the characteristics of the source term (noble gases, particles and volatile iodine forms) released from the reactor containment building during two dominant core-melt accident sequences: S 2 CD and TLB according to the ''Reactor Safety Study'' terminology. The reactor chosen for this study is a French 900 MWe PWR unit. The reactor building is a prestressed concrete containment with an internal liner. The first core-melt accident sequence is a 2-break loss-of-coolant accident on the cold leg, with failure of both system and the containment spray system. The second one is a transient initiated by a loss of offsite and onsite power supply and auxiliary feedwater system. These two sequences have been chosen because they are representative of risk dominant scenarios. Source terms associated with hypothetical core-melt accidents S 2 CD and TLB in a French PWR -900 MWe- have been performed using French computer codes (in particular, JERICHO Code for containment response analysis and AEROSOLS/31 for aerosol behavior in the containment)

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)

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

Incorporation of advanced accident analysis methodology into safety analysis reports

International Nuclear Information System (INIS)

2003-05-01

The IAEA Safety Guide on Safety Assessment and Verification defines that the aim of the safety analysis should be by means of appropriate analytical tools to establish and confirm the design basis for the items important to safety, and to ensure that the overall plant design is capable of meeting the prescribed and acceptable limits for radiation doses and releases for each plant condition category. Practical guidance on how to perform accident analyses of nuclear power plants (NPPs) is provided by the IAEA Safety Report on Accident Analysis for Nuclear Power Plants. The safety analyses are performed both in the form of deterministic and probabilistic analyses for NPPs. It is customary to refer to deterministic safety analyses as accident analyses. This report discusses the aspects of using the advanced accident analysis methods to carry out accident analyses in order to introduce them into the Safety Analysis Reports (SARs). In relation to the SAR, purposes of deterministic safety analysis can be further specified as (1) to demonstrate compliance with specific regulatory acceptance criteria; (2) to complement other analyses and evaluations in defining a complete set of design and operating requirements; (3) to identify and quantify limiting safety system set points and limiting conditions for operation to be used in the NPP limits and conditions; (4) to justify appropriateness of the technical solutions employed in the fulfillment of predetermined safety requirements. The essential parts of accident analyses are performed by applying sophisticated computer code packages, which have been specifically developed for this purpose. These code packages include mainly thermal-hydraulic system codes and reactor dynamics codes meant for the transient and accident analyses. There are also specific codes such as those for the containment thermal-hydraulics, for the radiological consequences and for severe accident analyses. In some cases, codes of a more general nature such

A study on hydrogen deflagration for selected severe accident sequences in Ringhals 3

Energy Technology Data Exchange (ETDEWEB)

Gustavsson, V.; Moeller, E. [SwedPower AB (Sweden)

2002-01-01

In this report, we have investigated the most important severe accident sequences in Ringhals 3, a Westinghouse 3-loop PWR, concerning hydrogen generation and containment pressure at hydrogen deflagration. In order to analyze the accident sequences and to calculate the hydrogen production, the computer code MAAP (Modular Accident Analysis Program) was used. Six accident sequences were studied, where four were LOCA cases and two transients. MAAP gives the evolution of the accident and particularly the pressure in the containment and the production of hydrogen as a function of time. The pressure peaks at deflagration were calculated by the method AICC-Adiabatic Isochoric Complete Combustion. The results from these calculations are conservative for two reasons. Adiabatic combustion means that the heat losses to structures in the containment are neglected. The combustion is also assumed to occur once and all available hydrogen is burned. The maximum pressure in five analysed cases was compared with the failure pressure of the containment. In the LOCA case, 373 kg hydrogen was burned and the resulting peak pressure in the containment was 0,53 MPa. In the transient, where 720 kg hydrogen was burned, the peak pressure was 0,69 MPa. This is the same as the failure pressure of the containment. Finally, in the conservative case, 980 kg hydrogen was burned and the resulting peak pressure 0,96 MPa. However, it should be noted that these conclusions are conservative from two points of view. Firstly a more realistic (than AICC) calculation of the peak pressure would give a lower value than 0,69 MPa. Secondly, there is conservatism in the evaluation of the failure pressure. (au)

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.

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)

Detailed evaluation of RCS boundary rupture during high-pressure severe accident sequences

International Nuclear Information System (INIS)

Park, Rae-Joon; Hong, Seong-Wan

2011-01-01

A depressurization possibility of the reactor coolant system (RCS) before a reactor vessel rupture during a high-pressure severe accident sequence has been evaluated for the consideration of direct containment heating (DCH) and containment bypass. A total loss of feed water (TLOFW) and a station blackout (SBO) of the advanced power reactor 1400 (APR 1400) has been evaluated from an initiating event to a creep rupture of the RCS boundary by using the SCDAP/RELAP5 computer code. In addition, intentional depressurization of the RCS using power-operated safety relief valves (POSRVs) has been evaluated. The SCDAPRELAP5 results have shown that the pressurizer surge line broke before the reactor vessel rupture failure, but a containment bypass did not occur because steam generator U tubes did not break. The intentional depressurization of the RCS using POSRV was effective for the DCH prevention at a reactor vessel rupture. (author)

The Integrated Approach to the Accident Evaluation for Advanced LWRs

International Nuclear Information System (INIS)

Oriolo, F.; Paci, S.

1998-01-01

The present paper discusses some relevant phenomena occurring in advanced LWRs during postulated accident scenarios. In particular, the operation of ESF is the starting point for analysis of those phenomena that cause the mutual influence between PS and containment in these plants. As a consequence, it is highlighted as accident analyses which treat PS and containment phenomena completely separated may be not adequate when applied to innovate reactors. Exemplified thermal-hydraulic analysis are presented for AP600 and SBWR, using the FUMO integrated model, for highlight accident evolution taking into account these interactions. The architecture of this integrated code is presented highlighting the importance of an integrated approach to the safety analysis in innovative reactors. (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.

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)

The Accident Sequence Precursor program: Methods improvements and current results

International Nuclear Information System (INIS)

Minarick, J.W.; Manning, F.M.; Harris, J.D.

1987-01-01

Changes in the US NRC Accident Sequence Precursor program methods since the initial program evaluations of 1969-81 operational events are described, along with insights from the review of 1984-85 events. For 1984-85, the number of significant precursors was consistent with the number observed in 1980-81, dominant sequences associated with significant events were reasonably consistent with PRA estimates for BWRs, but lacked the contribution due to small-break LOCAs previously observed and predicted in PWRs, and the frequency of initiating events and non-recoverable system failures exhibited some reduction compared to 1980-81. Operational events which provide information concerning additional PRA modeling needs are also described

A methodology for analyzing precursors to earthquake-initiated and fire-initiated accident sequences

International Nuclear Information System (INIS)

Budnitz, R.J.; Lambert, H.E.; Apostolakis, G.

1998-04-01

This report covers work to develop a methodology for analyzing precursors to both earthquake-initiated and fire-initiated accidents at commercial nuclear power plants. Currently, the U.S. Nuclear Regulatory Commission sponsors a large ongoing project, the Accident Sequence Precursor project, to analyze the safety significance of other types of accident precursors, such as those arising from internally-initiated transients and pipe breaks, but earthquakes and fires are not within the current scope. The results of this project are that: (1) an overall step-by-step methodology has been developed for precursors to both fire-initiated and seismic-initiated potential accidents; (2) some stylized case-study examples are provided to demonstrate how the fully-developed methodology works in practice, and (3) a generic seismic-fragility date base for equipment is provided for use in seismic-precursors analyses. 44 refs., 23 figs., 16 tabs

Core damage frequency estimation using accident sequence precursor data: 1990-1993

International Nuclear Information System (INIS)

Martz, H.F.

1998-01-01

The Nuclear Regulatory Commission's (NRC's) ongoing Accident Sequence Precursor (ASP) program uses probabilistic risk assessment (PRA) techniques to assess the potential for severe core damage (henceforth referred to simply as core damage) based on operating events. The types of operating events considered include accident sequence initiators, safety equipment failures, and degradation of plant conditions that could increase the probability that various postulated accident sequences occur. Such operating events potentially reduce the margin of safety available for prevention of core damage an thus can be considered as precursors to core damage. The current process for identifying, analyzing, and documenting ASP events is described in detail in Vanden Heuval et al. The significance of a Licensee Event Report (LER) event (or events) is measured by means of the conditional probability that the event leads to core damage, the so-called conditional core damage probability or, simply, CCDP. When the first ASP study results were published in 1982, it covered the period 1969--1979. In addition to identification and ranking of precursors, the original study attempted to estimate core damage frequency (CDF) based on the precursor events. The purpose of this paper is to compare the average annual CDF estimates calculated using the CCDP sum, Cooke-Goossens, Bier, and Abramson estimators for various reactor classes using the combined ASP data for the four years, 1990--1993. An important outcome of this comparison is an answer to the persistent question regarding the degree and effect of the positive bias of the CCDP sum method in practice. Note that this paper only compares the estimators with each other. Because the true average CDF is unknown, the estimation error is also unknown. Therefore, any observations or characterizations of bias are based on purely theoretical considerations

Severe accident sequence assessment for boiling water reactors: program overview

International Nuclear Information System (INIS)

Fontana, M.H.

1980-10-01

The Severe Accident Sequence Assessment (SASA) Program was started at the Oak Ridge National Laboratory (ORNL) in June 1980. This report documents the initial planning, specification of objectives, potential uses of the results, plan of attack, and preliminary results. ORNL was assigned the Brown's Ferry Unit 1 Plant with the station blackout being the initial sequence set to be addressed. This set includes: (1) loss of offsite and onsite ac power with no coolant injection; and (2) loss of offsite and onsite ac power with high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) as long as dc power supply lasts. This report includes representative preliminary results for the former case

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

Accident sequences evaluation using SFATs for low power and shutdown operation of pressurized heavy water reactors

International Nuclear Information System (INIS)

Kim, Chansoo; Chung, Chang-Hyun; Yang, Huichang

2004-01-01

To maintain the level of defense-in-depth safety of Pressurized Heavy Water Reactor (PHWR) during LP/SD operation, the qualitative risk evaluation methods such as Safety Function Assessment Trees (SFATs) are required. Therefore SFATs are suggested to assess and manage the PHWR safety in LP/SD. Before this study, safety functions of PHWR were classified into 7 groups; Reactivity Control, Core Cooling, Secondary Heat Removal, Primary Heat Transport Inventory, Essential Electrical Power, Cooling Water, and Containment Integrity. The SFATs for PHWR LP/SD operations were developed along with the Plant Outage Status (POS) variation, and totally 38 SFATs were developed for Wolsung Unit 2. For the verification of SFATs logics developed, top 5 accident sequences those contribute the CDF of PHWR were selected, and plant safety status were evaluated for those accident sequences. Accident sequences such as DCC-4 (Dual Control Computer Failure), CL4-16 (Total Loss of Class IV Power), and FWPV-11 (Loss of Feedwater Supply to SG due to Failure of Pumps/Values) were included. In this research the evaluation of plant safety status by accident sequences using SFATs and the verification of the SFATs were performed. Through the verification of SFAT logics, the enhancements to the internal logics of the SFATs were made, and the dependencies between safety systems and support systems were considered. It is expected the defense-in-depth evaluation model of PHW just as SFATs can be utilized in the configuration risk management program (CRMP) development and improve technical specifications development for Korean PHWRs. (author)

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

International Nuclear Information System (INIS)

Passalacqua, R.

2004-01-01

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

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.

A severe accident analysis for the system-integrated modular advanced reactor

International Nuclear Information System (INIS)

Jung, Gunhyo; Jae, Moosung

2015-01-01

The System-Integrated Modular Advanced Reactor (SMART) that has been recently designed in KOREA and has acquired standard design certification from the nuclear power regulatory body (NSSC) is an integral type reactor with 330MW thermal power. It is a small sized reactor in which the core, steam generator, pressurizer, and reactor coolant pump that are in existing pressurized light water reactors are designed to be within a pressure vessel without any separate pipe connection. In addition, this reactor has much different design characteristics from existing pressurized light water reactors such as the adoption of a passive residual heat removal system and a cavity flooding system. Therefore, the safety of the SMART against severe accidents should be checked through severe accident analysis reflecting the design characteristics of the SMART. For severe accident analysis, an analysis model has been developed reflecting the design information presented in the standard design safety analysis report. The severe accident analysis model has been developed using the MELCOR code that is widely used to evaluate pressurized LWR severe accidents. The steady state accident analysis model for the SMART has been simulated. According to the analysis results, the developed model reflecting the design of the SMART is found to be appropriate. Severe accident analysis has been performed for the representative accident scenarios that lead to core damage to check the appropriateness of the severe accident management plan for the SMART. The SMART has been shown to be safe enough to prevent severe accidents by utilizing severe accident management systems such as a containment spray system, a passive hydrogen recombiner, and a cavity flooding system. In addition, the SMART is judged to have been technically improved remarkably compared to existing PWRs. The SMART has been designed to have a larger reactor coolant inventory compared to its core's thermal power, a large surface area in

Current understanding of the sequence of events. Overview of current understanding of accident progression at Fukushima Dai-ichi

International Nuclear Information System (INIS)

Gulliford, Jim

2013-01-01

An overview of the main sequence of events, particularly the evolution of the cores in Units 1-3 was given. The presentation is based on information provided by Dr Okajima of JAEA to the June 2012 Nuclear Science Committee meeting. During the accident, conditions at the plant were such that operators were initially unable to obtain instruments readouts from the control panel and hence could not know what condition the reactors were in. (Reactor Power, Pressure, Temperature, Water height and flow rate, etc.). Subsequently, as electrical power supplies were gradually restored more data became available. In addition to the reactor data, other information from off-site measurements and from measuring stations inside the site boundary is now available, particularly for radiation dose rates in air. These types of information, combined with detailed knowledge of the plant design and operations history up to the time of the accident are being used to construct detailed computer models which simulate the behaviour of the reactor core, pressure vessel and containment during the accident sequence. This combination of detailed design/operating data, limited measured data during the accident and computer modelling allows us to construct a fairly clear picture of the accident progression. The main sequence of events (common to Units 1, 2 and 3) is summarised. The OECD/NEA is currently coordinating an international benchmark study of the accident at Fukushima Daiichi known as the BSAF Project. The objectives of this activity are to analyse and evaluate the accident progression and improve severe accident (SA) analysis methods and models. The project provides valuable additional (and corrected) data from plant measurements as well as an improved understanding of the role played by the fuel and cladding design. Based on (limited) plant data and extensive modelling analysis, we have a detailed qualitative description of the Fukushima-Daiichi accident. Further analyses of the type

Postulated accidents

International Nuclear Information System (INIS)

Ullrich, W.

1980-01-01

This lecture on 'Postulated Accidents' is the first of a series of lectures on the dynamic and transient behaviour of nuclear power plants, especially pressurized water reactors. The main points covered will be: Reactivity Accidents, Transients (Intact Loop) and Loss of Cooland Accidents (LOCA) including small leak. This lecture will discuss the accident analysis in general, the definition of the various operational phases, the accident classification, and, as an example, an accident sequence analysis on the basis of 'Postulated Accidents'. (orig./RW)

Modeling framework for crew decisions during accident sequences

International Nuclear Information System (INIS)

Lukic, Y.D.; Worledge, D.H.; Hannaman, G.W.; Spurgin, A.J.

1986-01-01

The ability to model the average behavior of operating crews in the course of accident sequences is vital in learning on how to prevent damage to power plants and to maintain safety. This paper summarizes the work carried out in support of a Human Reliability Model framework. This work develops the mathematical framework of the model and identifies the parameters which could be measured in some way, e.g., through simulator experience and/or small scale tests. Selected illustrative examples are presented, of the numerical experiments carried out in order to understand the model sensitivity to parameter variation. These examples ar discussed with the objective of deriving insights of general nature regarding operating of the model which may lead to enhanced understanding of man/machine interactions

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

International Nuclear Information System (INIS)

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

1989-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1989-11-01

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

Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications

International Nuclear Information System (INIS)

2013-12-01

Requests for severe accident investigations and assurance of mitigation measures have increased for operating nuclear power plants and the design of advanced nuclear power plants. Severe accident analysis investigations necessitate the analysis of the very complex physical phenomena that occur sequentially during various stages of accident progression. Computer codes are essential tools for understanding how the reactor and its containment might respond under severe accident conditions. The IAEA organizes coordinated research projects (CRPs) to facilitate technology development through international collaboration among Member States. The CRP on Benchmarking Severe Accident Computer Codes for HWR Applications was planned on the advice and with the support of the IAEA Nuclear Energy Department's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). This publication summarizes the results from the CRP participants. The CRP promoted international collaboration among Member States to improve the phenomenological understanding of severe core damage accidents and the capability to analyse them. The CRP scope included the identification and selection of a severe accident sequence, selection of appropriate geometrical and boundary conditions, conduct of benchmark analyses, comparison of the results of all code outputs, evaluation of the capabilities of computer codes to predict important severe accident phenomena, and the proposal of necessary code improvements and/or new experiments to reduce uncertainties. Seven institutes from five countries with HWRs participated in this CRP

Advanced Approach to Consider Aleatory and Epistemic Uncertainties for Integral Accident Simulations

International Nuclear Information System (INIS)

Peschke, Joerg; Kloos, Martina

2013-01-01

The use of best-estimate codes together with realistic input data generally requires that all potentially important epistemic uncertainties which may affect the code prediction are considered in order to get an adequate quantification of the epistemic uncertainty of the prediction as an expression of the existing imprecise knowledge. To facilitate the performance of the required epistemic uncertainty analyses, methods and corresponding software tools are available like, for instance, the GRS-tool SUSA (Software for Uncertainty and Sensitivity Analysis). However, for risk-informed decision-making, the restriction on epistemic uncertainties alone is not enough. Transients and accident scenarios are also affected by aleatory uncertainties which are due to the unpredictable nature of phenomena. It is essential that aleatory uncertainties are taken into account as well, not only in a simplified and supposedly conservative way but as realistic as possible. The additional consideration of aleatory uncertainties, for instance, on the behavior of the technical system, the performance of plant operators, or on the behavior of the physical process provides a quantification of probabilistically significant accident sequences. Only if a safety analysis is able to account for both epistemic and aleatory uncertainties in a realistic manner, it can provide a well-founded risk-informed answer for decision-making. At GRS, an advanced probabilistic dynamics method was developed to address this problem and to provide a more realistic modeling and assessment of transients and accident scenarios. This method allows for an integral simulation of complex dynamic processes particularly taking into account interactions between the plant dynamics as simulated by a best-estimate code, the dynamics of operator actions and the influence of epistemic and aleatory uncertainties. In this paper, the GRS method MCDET (Monte Carlo Dynamic Event Tree) for probabilistic dynamics analysis is explained

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

International Nuclear Information System (INIS)

Sdouz, Gert

2006-01-01

The goal of this work is the investigation of the influence of different accident management strategies on the thermal-hydraulics in the containment during a Large Break Loss of Coolant Accident with a large containment leak from the beginning of the accident. The increasing relevance of terrorism suggests a closer look at this kind of severe accidents. Normally the course of severe accidents and their associated phenomena are investigated with the assumption of an intact containment from the beginning of the accident. This intact containment has the ability to retain a large part of the radioactive inventory. In these cases there is only a release via a very small leakage due to the un-tightness of the containment up to cavity bottom melt through. This paper represents the last part of a comprehensive study on the influence of accident management strategies on the source term of VVER-1000 reactors. Basically two different accident sequences were investigated: the 'Station Blackout'- sequence and the 'Large Break LOCA'. In a first step the source term calculations were performed assuming an intact containment from the beginning of the accident and no accident management action. In a further step the influence of different accident management strategies was studied. The last part of the project was a repetition of the calculations with the assumption of a damaged containment from the beginning of the accident. This paper concentrates on the last step in the case of a Large Break LOCA. To be able to compare the results with calculations performed years ago the calculations were performed using the Source Term Code Package (STCP), hydrogen explosions are not considered. In this study four different scenarios have been investigated. The main parameter was the switch on time of the spray systems. One of the results is the influence of different accident management strategies on the source term. In the comparison with the sequence with intact containment it was

Factors correlated with traffic accidents as a basis for evaluating Advanced Driver Assistance Systems.

Science.gov (United States)

Staubach, Maria

2009-09-01

This study aims to identify factors which influence and cause errors in traffic accidents and to use these as a basis for information to guide the application and design of driver assistance systems. A total of 474 accidents were examined in depth for this study by means of a psychological survey, data from accident reports, and technical reconstruction information. An error analysis was subsequently carried out, taking into account the driver, environment, and vehicle sub-systems. Results showed that all accidents were influenced by errors as a consequence of distraction and reduced activity. For crossroad accidents, there were further errors resulting from sight obstruction, masked stimuli, focus errors, and law infringements. Lane departure crashes were additionally caused by errors as a result of masked stimuli, law infringements, expectation errors as well as objective and action slips, while same direction accidents occurred additionally because of focus errors, expectation errors, and objective and action slips. Most accidents were influenced by multiple factors. There is a safety potential for Advanced Driver Assistance Systems (ADAS), which support the driver in information assimilation and help to avoid distraction and reduced activity. The design of the ADAS is dependent on the specific influencing factors of the accident type.

Recent advances in severe accident technology - direct containment heating in advanced light water reactors

International Nuclear Information System (INIS)

Fontana, M.H.

1993-01-01

The issues affecting high-pressure melt ejection (HPME) and the consequential containment pressurization from direct containment heating (DCH), as they affect advanced light water reactors (ALWRs), specifically advanced pressurized water reactors (APWRs), were reviewed by the U.S. Department of Energy Advanced Reactor Severe Accident Program (ARSAP). Recommendations from ARSAP regarding the design of APWRs to minimize DCH are embodied within the Electric Power Research Institute ALWR Utility Requirements Document, which specifies (a) a large, strong containment; (b) an in-containment refueling water storage tank; (c) a reactor cavity configuration that minimizes energy transport to the containment atmosphere; and (d) a reactor coolant system depressurization system. Experimental and analytical efforts, which have focused on current-generation plants, and analyses for APWRs were reviewed. Although DCH is a subject of continuous research and considerable uncertainties remain, it is the judgment of the ARSAP that reactors complying with the recommended design requirements would have a low probability of early containment failure due to HPME and DCH

Reactivity insertion accident analysis

International Nuclear Information System (INIS)

Moreira, J.M.L.; Nakata, H.; Yorihaz, H.

1990-04-01

The correct prediction of postulated accidents is the fundamental requirement for the reactor licensing procedures. Accident sequences and severity of their consequences depend upon the analysis which rely on analytical tools which must be validated against known experimental results. Present work presents a systematic approach to analyse and estimate the reactivity insertion accident sequences. The methodology is based on the CINETHICA code which solves the point-kinetics/thermohydraulic coupled equations with weighted temperature feedback. Comparison against SPERT experimental results shows good agreement for the step insertion accidents. (author) [pt

The next nuclear power station generation: Beyond-design accident concepts, methods, and action sequence

International Nuclear Information System (INIS)

Asmolov, V.G.; Khakh, O.Ya.; Shashkov, M.G.

1993-01-01

The problem of beyond-design accidents at nuclear stations will not be solved unless a safety culture becomes a basic characteristic of all lines of activity. Only then can the danger of accidents as an objective feature of nuclear stations be eliminated by purposive skilled and responsible activities of those implementing safety. Nuclear-station safety is provided by the following interacting and complementary lines of activity: (1) the design and construction of nuclear stations by properly qualified design and building organizations; (2) monitoring and supervision of safety by special state bodies; (3) control of the station by the exploiting organization; and (4) scientific examination of safety within the above framework and by independent organizations. The distribution of the responsibilities, powers, and right in these lines should be defined by a law on atomic energy, but there is not such law in Russian. The beyond-design accident problem is a key one in nuclear station safety, as it clear from the serious experience with accidents and numerous probabilistic studies. There are four features of the state of this topic in Russia that are of major significance for managing accidents: the lack of an atomic energy law, the inadequacy of the technical standards, the lack of a verified program package for nuclear-station designs in order to calculate the beyond-design accidents and analyze risks, and a lack of approach by designers to such accidents on the basis of international recommendations. This paper gives a brief description of three-forming points in the scientific activity: the general concept of nuclear-station safety, methods of analyzing and providing accident management, and the sequence of actions developed by specialists at this institute in recent years

A decision theoretic approach to an accident sequence: when feedwater and auxiliary feedwater fail in a nuclear power plant

International Nuclear Information System (INIS)

Svenson, Ola

1998-01-01

This study applies a decision theoretic perspective on a severe accident management sequence in a processing industry. The sequence contains loss of feedwater and auxiliary feedwater in a boiling water nuclear reactor (BWR), which necessitates manual depressurization of the reactor pressure vessel to enable low pressure cooling of the core. The sequence is fast and is a major contributor to core damage in probabilistic risk analyses (PRAs) of this kind of plant. The management of the sequence also includes important, difficult and fast human decision making. The decision theoretic perspective, which is applied to a Swedish ABB-type reactor, stresses the roles played by uncertainties about plant state, consequences of different actions and goals during the management of a severe accident sequence. Based on a theoretical analysis and empirical simulator data the human error probabilities in the PRA for the plant are considered to be too small. Recommendations for how to improve safety are given and they include full automation of the sequence, improved operator training, and/or actions to assist the operators' decision making through reduction of uncertainties, for example, concerning water/steam level for sufficient cooling, time remaining before insufficient cooling level in the tank is reached and organizational cost-benefit evaluations of the events following a false alarm depressurization as well as the events following a successful depressurization at different points in time. Finally, it is pointed out that the approach exemplified in this study is applicable to any accident scenario which includes difficult human decision making with conflicting goals, uncertain information and with very serious consequences

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.

Phenomenological uncertainty analysis of containment building pressure load caused by severe accident sequences

International Nuclear Information System (INIS)

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

2014-01-01

Highlights: • Phenomenological uncertainty analysis has been applied to level 2 PSA. • The methodology provides an alternative to simple deterministic analyses and sensitivity studies. • A realistic evaluation provides a more complete characterization of risks. • Uncertain parameters of MAAP code for the early containment failure were identified. - Abstract: This paper illustrates an application of a severe accident analysis code, MAAP, to the uncertainty evaluation of early containment failure scenarios employed in the containment event tree (CET) model of a reference plant. An uncertainty analysis of containment pressure behavior during severe accidents has been performed for an optimum assessment of an early containment failure model. The present application is mainly focused on determining an estimate of the containment building pressure load caused by severe accident sequences of a nuclear power plant. Key modeling parameters and phenomenological models employed for the present uncertainty analysis are closely related to the in-vessel hydrogen generation, direct containment heating, and gas combustion. The basic approach of this methodology is to (1) develop severe accident scenarios for which containment pressure loads should be performed based on a level 2 PSA, (2) identify severe accident phenomena relevant to an early containment failure, (3) identify the MAAP input parameters, sensitivity coefficients, and modeling options that describe or influence the early containment failure phenomena, (4) prescribe the likelihood descriptions of the potential range of these parameters, and (5) evaluate the code predictions using a number of random combinations of parameter inputs sampled from the likelihood distributions

Efficient method for simulation of BWR severe accident sequence events before core uncovery

International Nuclear Information System (INIS)

Harrington, R.M.

1984-01-01

BWR-LACP has been a versatile tool for the ORNL SASA program. The development effort was minimal, and the code is fast running and economical. Operator actions are easily simulated and the complete scope of both reactor vessel and primary containment are modeled. Valuable insights have been gained into accident sequences. A Fortran version is under development and it will be modified for application to Mark II plants

Heath Effects Sequence of Meet Halfa Radiological Accident After Twelve Years

International Nuclear Information System (INIS)

Shabon, M.H.

2013-01-01

The accident of Meet-Halfa developed consequent upon the loss of an industrial gamma radiography source. The source was found by a farmer resident of Meet-Halfa who took it to his house occupied by his family. The sequence of events developed over a period of seven weeks from the time the source was found on May 5, 2000, till the day of its retrieval from the house by the national authorities on June 26. The protracted exposure patterns of the family members during the period of source possession are not precisely known, however these exposures resulted in two fatalities, clinical forms of bone marrow depression, and several skin burns of different severities. The recent sequences of the accident is as follows:-The three survived siblings married and get good children. That mean there is no hereditary stochastic effects. The sister died at 2007 with 72 years old with senility and no specific disease. The youngest daughter amputate the left thumb and index fingers at 2001. The elder son amputate the terminal phalanx of the right thumb at 2009. The youngest daughter amputate the right index finger at 2009. The elder son graft the burn at the lower right quadrant of the abdomen for more than 20 times (3 of them were in the Mansheat Al-Bakry Millitary Hospital), but there is residual of burn untill now. Sever abdominal hernia in the elder son due to necroses in the right quadrant abdominal muscles. Grafting for these muscles occur but failed.

Methodology for time-dependent reliability analysis of accident sequences and complex reactor systems

International Nuclear Information System (INIS)

Paula, H.M.

1984-01-01

The work presented here is of direct use in probabilistic risk assessment (PRA) and is of value to utilities as well as the Nuclear Regulatory Commission (NRC). Specifically, this report presents a methodology and a computer program to calculate the expected number of occurrences for each accident sequence in an event tree. The methodology evaluates the time-dependent (instantaneous) and the average behavior of the accident sequence. The methodology accounts for standby safety system and component failures that occur (a) before they are demanded, (b) upon demand, and (c) during the mission (system operation). With respect to failures that occur during the mission, this methodology is unique in the sense that it models components that can be repaired during the mission. The expected number of system failures during the mission provides an upper bound for the probability of a system failure to run - the mission unreliability. The basic event modeling includes components that are continuously monitored, periodically tested, and those that are not tested or are otherwise nonrepairable. The computer program ASA allows practical applications of the method developed. This work represents a required extension of the presently available methodology and allows a more realistic PRA of nuclear power plants

Method for improving accident sequence recognition in nuclear power plant control rooms

International Nuclear Information System (INIS)

Heising, C.D.; Dinsmore, S.C.

1983-01-01

This work adapts fault trees from plant-specific probabilistic risk analyses (PRAs) to construct and quantitatively evaluate an alarm analysis system for the engineered safety features (ESFs). The purpose is to help improve reactor operator recognition and identification of potential accident sequences. The PRA system fault trees provide system failure mode information which can be used to construct alarm trees. These alarm trees provide a framework for assessing the plant indicators so that the plant conditions are made more readily apparent to plant personnel. In the alarm tree, possible states of each instrumented alarem are identified as true or false. In addition, a warning status is defined and integrated into an alarm analysis routine. The impact of this additional status conditioned on the Boolean laws used to evaluate the alarm trees is examined. An application is described for BWR high pressure coolant injection system (HPCI) that would be utilized during many severe reactor accidents

Advances in global development and deployment of small modular reactors and incorporating lessons learned from the Fukushima Daiichi accident into the designs of engineered safety features of advanced reactors

International Nuclear Information System (INIS)

Hadid Subki, M.; )

2014-01-01

The IAEA has been facilitating the Member States in incorporating the lessons-learned from the Fukushima Dai-ichi Accident into the designs of engineered safety features of advanced reactors, including small modular reactors. An extended assessment is required to address challenges for advancing reactor safety in the new evolving generation of SMR plants to preserve the historic lessons in safety, through: assuring the diversity in emergency core cooling systems following loss of onsite AC power; ensuring diversity in reactor depressurization following a transient or accident; confirming independence in reactor trip and safety systems for sensors, power supplies and actuation systems, and finally diversity in maintaining containment integrity following a severe accident

A PC Mathcad-based computational aid for severe accident analysis and its application to a BWR small LOCA sequence

International Nuclear Information System (INIS)

Wu, Laung-Kuang T.; Lee, S.J.

2004-01-01

A PC-based Mathcad program is used to develop a computational aid for analyzing severe accident phenomena. This computational aid uses simple engineering expressions and empirical correlations to estimate key quantities and timings at various stages of accident progressions. In this paper, the computational aid is applied to analyze an early phase of a BWR small LOCA sequence. The accident phenomena analyzed include: break flow rates, boiled-up water level in the core, core uncovery time, depressurization of the reactor pressure vessel, core heat-up, onset of clad oxidation, hydrogen generation, and onset of fuel relocation. The results are compared with those obtained running the MAAP 3.0B code. This PC-based computational aid can be used to train plant personnel in understanding severe accident phenomena and to assist them in managing severe accidents. (author)

The investigation of Passive Accident Mitigation Scheme for advanced PWR NPP

International Nuclear Information System (INIS)

Shi, Er-bing; Fang, Cheng-yue; Wang, Chang; Xia, Geng-lei; Zhao, Cui-na

2015-01-01

Highlights: • We put forward a new PAMS and analyze its operation characteristics under SBO. • We conduct comparative analysis between PAMS and Traditional Secondary Side PHRS. • The PAMS could cope with SBO accident and maintain the plant in safe conditions. • PAMS could decrease heat removal capacity of PHRS. • PAMS has advantage in reducing cooling rate and PCCT temperature rising amplitude. - Abstract: To enhance inherent safety features of nuclear power plant, the advanced pressurized water reactors implement a series of passive safety systems. This paper puts forward and designs a new Passive Accident Mitigation Scheme (PAMS) to remove residual heat, which consists of two parts: the first part is Passive Auxiliary Feedwater System (PAFS), and the other part is Passive Heat Removal System (PHRS). This paper takes the Westinghouse-designed Advanced Passive PWR (AP1000) as research object and analyzes the operation characteristics of PAMS to cope with the Station Blackout Accident (SBO) by using RELAP5 code. Moreover, the comparative analysis is also conducted between PAMS and Traditional Secondary Circuit PHRS to derive the advantages of PAMS. The results show that the designed scheme can remove core residual heat significantly and maintain the plant in safe conditions; the first part of PAMS would stop after 120 min and the second part has to come into use simultaneously; the low pressurizer (PZR) pressure signal would be generated 109 min later caused by coolant volume shrinkage, which would actuate the Passive Safety Injection System (PSIS) to recovery the water level of pressurizer; the flow instability phenomenon would occur and last 21 min after the PHRS start-up; according to the comparative analysis, the coolant average temperature gradient and the Passive Condensate Cooling Tank (PCCT) water temperature rising amplitude of PAMS are lower than those of Traditional Secondary Circuit PHRS

The tsunami probabilistic risk assessment (PRA). Example of accident sequence analysis of tsunami PRA according to the standard for procedure of tsunami PRA for nuclear power plants

International Nuclear Information System (INIS)

Ohara, Norihiro; Hasegawa, Keiko; Kuroiwa, Katsuya

2013-01-01

After the Fukushima Daiichi nuclear power plant (NPP) accident, standard for procedure of tsunami PRA for NPP had been established by the Standardization Committee of AESJ. Industry group had been conducting analysis of Tsunami PRA for PWR based on the standard under the cooperation with electric utilities. This article introduced overview of the standard and examples of accident sequence analysis of Tsunami PRA studied by the industry group according to the standard. The standard consisted of (1) investigation of NPP's composition, characteristics and site information, (2) selection of relevant components for Tsunami PRA and initiating events and identification of accident sequence, (3) evaluation of Tsunami hazards, (4) fragility evaluation of building and components and (5) evaluation of accident sequence. Based on the evaluation, countermeasures for further improvement of safety against Tsunami could be identified by the sensitivity analysis. (T. Tanaka)

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

International Nuclear Information System (INIS)

Bava, G.; Zambardi, F.

1992-01-01

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

Seismically induced accident sequence analysis of the advanced test reactor

International Nuclear Information System (INIS)

Khericha, S.T.; Henry, D.M.; Ravindra, M.K.; Hashimoto, P.S.; Griffin, M.J.; Tong, W.H.; Nafday, A.M.

1991-01-01

A seismic probabilistic risk assessment (PRA) was performed for the Department of Energy (DOE) Advanced Test Reactor (ATR) as part of the external events analysis. The risk from seismic events to the fuel in the core and in the fuel storage canal was evaluated. The key elements of this paper are the integration of seismically induced internal flood and internal fire, and the modeling of human error rates as a function of the magnitude of earthquake. The systems analysis was performed by EG ampersand G Idaho, Inc. and the fragility analysis and quantification were performed by EQE International, Inc. (EQE)

Loss-of-coolant accident analyses of the Advanced Neutron Source Reactor

International Nuclear Information System (INIS)

Chen, N.C.J.; Yoder, G.L.; Wendel, M.W.

1991-01-01

Currently in the conceptual design stage, the Advanced Neutron Source Reactor (ANSR) will operate at a high heat flux, a high mass flux, an a high degree of coolant subcooling. Loss-of-coolant accident (LOCA) analyses using RELAP5 have been performed as part of an early evaluation of ANSR safety issues. This paper discusses the RELAP5 ANSR conceptual design system model and preliminary LOCA simulation results. Some previous studies were conducted for the preconceptual design. 12 refs., 7 figs

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

International Nuclear Information System (INIS)

1998-06-01

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

Methodology to classify accident sequences of an Individual Plant Examination according to the severe releases for BWR type reactors

International Nuclear Information System (INIS)

Sandoval V, S.

2001-01-01

The Light Water Reactor (LWR) operation regulations require to every operating plant to perform of an Individual Plant Examination study (Ipe). One of the main purposes of an Ipe is t o gain a more quantitative understanding of the overall probabilities of core damage and fission product releases . Probabilistic Safety Analysis (PSA) methodologies and Severe Accident Analysis are used to perform Ipe studies. PSA methodologies are used to identify and analyse the set of event sequences that might originate the fission product release from a nuclear power plant; these methodologies are combinatorial in nature and generate thousands of sequences. Among other uses within an Ipe, severe accident simulations are used to determine the characteristics of the fission product release for the identified sequences and in this way, the releases can be understood and characterized. A vast amount of resources is required to simulate and analyse every Ipe sequence. This effort is unnecessary if similar sequences are grouped. The grouping scheme must achieve an efficient trade off between problem reduction and accuracy. The methodology presented in this work enables an accurate characterization and analysis of the Ipe fission product releases by using a reduced problem. The methodology encourages the use of specific plant simulations. (Author)

Preliminary Analysis of Aircraft Loss of Control Accidents: Worst Case Precursor Combinations and Temporal Sequencing

Science.gov (United States)

Belcastro, Christine M.; Groff, Loren; Newman, Richard L.; Foster, John V.; Crider, Dennis H.; Klyde, David H.; Huston, A. McCall

2014-01-01

Aircraft loss of control (LOC) is a leading cause of fatal accidents across all transport airplane and operational classes, and can result from a wide spectrum of hazards, often occurring in combination. Technologies developed for LOC prevention and recovery must therefore be effective under a wide variety of conditions and uncertainties, including multiple hazards, and their validation must provide a means of assessing system effectiveness and coverage of these hazards. This requires the definition of a comprehensive set of LOC test scenarios based on accident and incident data as well as future risks. This paper defines a comprehensive set of accidents and incidents over a recent 15 year period, and presents preliminary analysis results to identify worst-case combinations of causal and contributing factors (i.e., accident precursors) and how they sequence in time. Such analyses can provide insight in developing effective solutions for LOC, and form the basis for developing test scenarios that can be used in evaluating them. Preliminary findings based on the results of this paper indicate that system failures or malfunctions, crew actions or inactions, vehicle impairment conditions, and vehicle upsets contributed the most to accidents and fatalities, followed by inclement weather or atmospheric disturbances and poor visibility. Follow-on research will include finalizing the analysis through a team consensus process, defining future risks, and developing a comprehensive set of test scenarios with correlation to the accidents, incidents, and future risks. Since enhanced engineering simulations are required for batch and piloted evaluations under realistic LOC precursor conditions, these test scenarios can also serve as a high-level requirement for defining the engineering simulation enhancements needed for generating them.

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.

Advanced surrogate model and sensitivity analysis methods for sodium fast reactor accident assessment

International Nuclear Information System (INIS)

Marrel, A.; Marie, N.; De Lozzo, M.

2015-01-01

Within the framework of the generation IV Sodium Fast Reactors, the safety in case of severe accidents is assessed. From this statement, CEA has developed a new physical tool to model the accident initiated by the Total Instantaneous Blockage (TIB) of a sub-assembly. This TIB simulator depends on many uncertain input parameters. This paper aims at proposing a global methodology combining several advanced statistical techniques in order to perform a global sensitivity analysis of this TIB simulator. The objective is to identify the most influential uncertain inputs for the various TIB outputs involved in the safety analysis. The proposed statistical methodology combining several advanced statistical techniques enables to take into account the constraints on the TIB simulator outputs (positivity constraints) and to deal simultaneously with various outputs. To do this, a space-filling design is used and the corresponding TIB model simulations are performed. Based on this learning sample, an efficient constrained Gaussian process metamodel is fitted on each TIB model outputs. Then, using the metamodels, classical sensitivity analyses are made for each TIB output. Multivariate global sensitivity analyses based on aggregated indices are also performed, providing additional valuable information. Main conclusions on the influence of each uncertain input are derived. - Highlights: • Physical-statistical tool for Sodium Fast Reactors TIB accident. • 27 uncertain parameters (core state, lack of physical knowledge) are highlighted. • Constrained Gaussian process efficiently predicts TIB outputs (safety criteria). • Multivariate sensitivity analyses reveal that three inputs are mainly influential. • The type of corium propagation (thermal or hydrodynamic) is the most influential

Uncertainties and severe-accident management

International Nuclear Information System (INIS)

Kastenberg, W.E.

1991-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

None

1977-03-01

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

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)

The study of steam explosions in nuclear systems. Advanced Reactor Severe Accident Program

International Nuclear Information System (INIS)

Theofanous, T.G.; Yuen, W.W.; Angelini, S.; Chen, X.

1995-01-01

This report presents an overview of the steam explosion issue in nuclear reactor safety and our approach to assessing it. Key physics, models, and computational tools are described, and illustrative results are presented for ex-vessel steam explosions in an open pool geometry. An extensive set of appendices facilitate access to previously reported work that is an integral part of this effort. These appendices include key developments in our approach, key advances in our understanding from physical and numerical experiments, and details of the most advanced computational results presented in this report. Of major significance are the following features: A consistent two-dimensional treatment for both premixing and propagation which in practical settings are ostensibly at least two-dimensional phenomena; experimental demonstration of voiding and microinteractions which represent key behaviors in premixing and propagation respectively; demonstration of the explosion venting phenomena in open pool geometries which, therefore, can be counted on as a very important mitigative feature; and introduction of the idea of penetration cutoff as a key mechanism prohibiting large-scale premixing in usual ex-vessel situations involving high pour velocities and subcooled pools. This report is intended as an overview and is to be followed by code manuals for PM-ALPHA and ESPROSE.m, respective verification reports, and application documents for reactor-specific applications. The applications will employ the Risk Oriented Accident Analysis Methodology (ROAAM) to address the safety importance of potential steam explosions phenomena in evaluated severe accidents for passive Advanced Light Water Reactors (ALWRs)

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

Stress in accident and post-accident management at Chernobyl

International Nuclear Information System (INIS)

Girard, P.; Dubreuil, G.H.

1996-01-01

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

Benchmarking severe accident computer codes for heavy water reactor applications

Energy Technology Data Exchange (ETDEWEB)

Choi, J.H. [International Atomic Energy Agency, Vienna (Austria)

2010-07-01

Consideration of severe accidents at a nuclear power plant (NPP) is an essential component of the defence in depth approach used in nuclear safety. Severe accident analysis involves very complex physical phenomena that occur sequentially during various stages of accident progression. Computer codes are essential tools for understanding how the reactor and its containment might respond under severe accident conditions. International cooperative research programmes are established by the IAEA in areas that are of common interest to a number of Member States. These co-operative efforts are carried out through coordinated research projects (CRPs), typically 3 to 6 years in duration, and often involving experimental activities. Such CRPs allow a sharing of efforts on an international basis, foster team-building and benefit from the experience and expertise of researchers from all participating institutes. The IAEA is organizing a CRP on benchmarking severe accident computer codes for heavy water reactor (HWR) applications. The CRP scope includes defining the severe accident sequence and conducting benchmark analyses for HWRs, evaluating the capabilities of existing computer codes to predict important severe accident phenomena, and suggesting necessary code improvements and/or new experiments to reduce uncertainties. The CRP has been planned on the advice and with the support of the IAEA Nuclear Energy Department's Technical Working Groups on Advanced Technologies for HWRs. (author)

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

Current status of low power/shutdown PSA and accident sequence analysis for loss of RHR during mid-loop operation

International Nuclear Information System (INIS)

Park, Chang Kyu; Choi, Young; Kim, Tae Woon; Jin, Young Ho

1994-07-01

Probabilistic safety assessment (PSA) has been applied to only full-power operation of nuclear power plant (NPP), but some events which were recently occurred could reach severe plant damage state. Thus, various countries around the world have focused their interests on the evaluation for low power/shutdown (LP/S) operation. This report covers the main stream of LP/S PSA methodology, current status of LP/S PSA practices and results, and accident sequence analysis for loss of RHR during mid-loop operation. Therefore this report would be helpful for us to practice LP/S PSA for YGN 5,6 NPP which will be built in the near future. Also the results of accident sequence analysis show that operator's mis-diagnosis and failure of recovery action would initiate core damage during LP/S operation. In summary, overall environmental improvements (equipments, procedures, Tech Spec, etc, ...) and operating support system will be very useful to reduce risk during LP/S operation. (Author) 5 figs., 9 tabs

Simulation with the MELCOR code of two severe accident sequences, Station Blackout and Small Break LOCA, for the Atucha I nuclear power plant

International Nuclear Information System (INIS)

Valle Cepero, Reinaldo

2004-01-01

The results of the PSA-I applied to the Atucha I nuclear power plant (CNA I) determine the accidental sequences with the most influence related to the probability of the core reactor damage. Among those sequences are include, the Station Blackout and lost of primary coolant, combine with the failure of the emergency injection systems by pipe breaks of diameters between DN100 - DN25 or equivalent areas, Small LOCA. This paper has the objective to model and analyze the behavior of the primary circuit and the pressure vessel during the evolution of those two accidental sequences. It presented a detailed analysis of the main phenomena that occur from the initial moment of the accident to the failure moment of the pressure vessel and the melt material fall to the reactor cavity. Two sequences were taken into account, considering the main phenomena (core uncover, heating, fuel element oxidation, hydrogen generation, degradation and relocation of the melt material, failure of the support structures, etc.) and the time of occurrence, of those events will be different, if it is considered that both sequences will be developed in different scenarios. One case is an accident with the primary circuit to a high pressure (Station Blackout scenario) and the other with a early primary circuit depressurization due to the lost of primary coolant. For this work the MELCOR 1.8.5 code was used and it allows within a unified framework to modeling an extensive spectrum of phenomenology associated with the severe accidents. (author)

Consideration of severe accidents in design of advanced WWER reactors

International Nuclear Information System (INIS)

Fedorov, V.G.; Rogov, M.F.; Podshibyakin, A.K.; Fil, N.S.; Volkov, B.E.; Semishkin, V.P.

1998-01-01

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

Parry-Romberg syndrome: findings in advanced magnetic resonance imaging sequences - case report

Energy Technology Data Exchange (ETDEWEB)

Paula, Rafael Alfenas de; Ribeiro, Bruno Niemeyer de Freitas, E-mail: alfenas85@gmail.com [Universidade Federal do Rio de Janeiro (UFRJ), Rio de Janeiro, RJ (Brazil). Hospital Universitario Clementino Fraga Filho; Bahia, Paulo Roberto Valle [Universidade Federal do Rio de Janeiro (UFRJ), Rio de Janeiro, RJ (Brazil). Dept. de radiologia; Ribeiro, Renato Niemeyer de Freitas [Hospital de Clinica de Jacarepagua, Rio de Janeiro, RJ (Brazil); Carvalho, Lais Balbi de [Universidade Presidente Antonio Carlos (Unipac), Juiz de Fora, MG (Brazil)

2014-05-15

Parry-Romberg syndrome is a rare disease characterized by progressive hemifacial atrophy associated with other systemic changes, including neurological symptoms. Currently, there are few studies exploring the utilization of advanced magnetic resonance sequences in the investigation of this disease. The authors report the case of a 45-year-old patient and describe the findings at structural magnetic resonance imaging and at advanced sequences, correlating them with pathophysiological data. (author)

Application of dynamic probabilistic safety assessment approach for accident sequence precursor analysis: Case study for steam generator tube rupture

Energy Technology Data Exchange (ETDEWEB)

Lee, Han Sul; Heo, Gyun Young [Kyung Hee University, Yongin (Korea, Republic of); Kim, Tae Wan [Incheon National University, Incheon (Korea, Republic of)

2017-03-15

The purpose of this research is to introduce the technical standard of accident sequence precursor (ASP) analysis, and to propose a case study using the dynamic-probabilistic safety assessment (D-PSA) approach. The D-PSA approach can aid in the determination of high-risk/low-frequency accident scenarios from all potential scenarios. It can also be used to investigate the dynamic interaction between the physical state and the actions of the operator in an accident situation for risk quantification. This approach lends significant potential for safety analysis. Furthermore, the D-PSA approach provides a more realistic risk assessment by minimizing assumptions used in the conventional PSA model so-called the static-PSA model, which are relatively static in comparison. We performed risk quantification of a steam generator tube rupture (SGTR) accident using the dynamic event tree (DET) methodology, which is the most widely used methodology in D-PSA. The risk quantification results of D-PSA and S-PSA are compared and evaluated. Suggestions and recommendations for using D-PSA are described in order to provide a technical perspective.

Analysis of radionuclide behavior in a BWR Mark-II containment under severe accident management condition in low pressure sequence

International Nuclear Information System (INIS)

Funayama, Kyoko; Kajimoto, Mitsuhiro; Nagayoshi, Takuji; Tanaka, Nobuo

1999-01-01

In the Level 2 PSA program at INS/NUPEC, MELCOR1.8.3 is extensively applied to analyze radionuclide behavior of dominant sequences. In addition, the revised source terms provided in the NUREG-1465 report have been also discussed to examine the potential of the radionuclides release to the environment in the conventional siting criteria. In the present study, characteristics of source terms to the environment were examined comparing with results by the Hypothetical Accident (LOCA), NUREG-1465 and MELCOR1.8.3. calculation for a typical BWR with a Mark-II containment in order to assure conservatives of the Hypothetical Accident in Japan. Release fractions of iodine to the environment for the Hypothetical Accident and NUREG-1465, which used engineering models for predicting radionuclide behaviors, were about 10 -4 and 10 -6 of core inventory, respectively, while the best estimate MELCOR1.8.3 code predicted 10 -9 of iodine to the environment. The present study showed that the engineering models in the Hypothetical Accident or NUREG-1465 have large conservatives to estimate source term of iodine to the environment. (author)

Application of dynamic probabilistic safety assessment approach for accident sequence precursor analysis: Case study for steam generator tube rupture

International Nuclear Information System (INIS)

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

2017-01-01

The purpose of this research is to introduce the technical standard of accident sequence precursor (ASP) analysis, and to propose a case study using the dynamic-probabilistic safety assessment (D-PSA) approach. The D-PSA approach can aid in the determination of high-risk/low-frequency accident scenarios from all potential scenarios. It can also be used to investigate the dynamic interaction between the physical state and the actions of the operator in an accident situation for risk quantification. This approach lends significant potential for safety analysis. Furthermore, the D-PSA approach provides a more realistic risk assessment by minimizing assumptions used in the conventional PSA model so-called the static-PSA model, which are relatively static in comparison. We performed risk quantification of a steam generator tube rupture (SGTR) accident using the dynamic event tree (DET) methodology, which is the most widely used methodology in D-PSA. The risk quantification results of D-PSA and S-PSA are compared and evaluated. Suggestions and recommendations for using D-PSA are described in order to provide a technical perspective

A Quantitative Accident Sequence Analysis for a VHTR

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

In Korea, the basic design features of VHTR are currently discussed in the various design concepts. Probabilistic risk assessment (PRA) offers a logical and structured method to assess risks of a large and complex engineered system, such as a nuclear power plant. It will be introduced at an early stage in the design, and will be upgraded at various design and licensing stages as the design matures and the design details are defined. Risk insights to be developed from the PRA are viewed as essential to developing a design that is optimized in meeting safety objectives and in interpreting the applicability of the existing demands to the safety design approach of the VHTR. In this study, initiating events which may occur in VHTRs were selected through MLD method. The initiating events were then grouped into four categories for the accident sequence analysis. Initiating events frequency and safety systems failure rate were calculated by using reliability data obtained from the available sources and fault tree analysis. After quantification, uncertainty analysis was conducted. The SR and LR frequency are calculated respectively 7.52E- 10/RY and 7.91E-16/RY, which are relatively less than the core damage frequency of LWRs.

Application of a Software tool for Evaluating Human Factors in Accident Sequences

International Nuclear Information System (INIS)

Queral, Cesar; Exposito, Antonio; Gonzalez, Isaac; Quiroga, Juan Antonio; Ibarra, Aitor; Hortal, Javier; Hulsund, John-Einar; Nilsen, Svein

2006-01-01

The Probabilistic Safety Assessment (PSA) includes the actions of the operator like elements in the set of the considered protection performances during accident sequences. Nevertheless, its impact throughout a sequence is not analyzed in a dynamic way. In this sense, it is convenient to make more detailed studies about its importance in the dynamics of the sequences, letting make studies of sensitivity respect to the human reliability and the response times. For this reason, the CSN is involved in several activities oriented to develop a new safety analysis methodology, the Integrated Safety Assessment (ISA), which must be able to incorporate operator actions in conventional thermo-hydraulic (TH) simulations. One of them is the collaboration project between CSN, HRP and the DSE-UPM that started in 2003. In the framework of this project, a software tool has been developed to incorporate operator actions in TH simulations. As a part of the ISA, this tool permits to quantify human error probabilities (HEP) and to evaluate its impact in the final state of the plant. Independently, it can be used for evaluating the impact of the execution by operators of procedures and guidelines in the final state of the plant and the evaluation of the allowable response times for the manual actions of the operator. The results obtained in the first pilot case are included in this paper. (authors)

Accident management approach in Armenia

International Nuclear Information System (INIS)

Ghazaryan, K.

1999-01-01

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

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

Probabilistic accident sequence recovery analysis

International Nuclear Information System (INIS)

Stutzke, Martin A.; Cooper, Susan E.

2004-01-01

Recovery analysis is a method that considers alternative strategies for preventing accidents in nuclear power plants during probabilistic risk assessment (PRA). Consideration of possible recovery actions in PRAs has been controversial, and there seems to be a widely held belief among PRA practitioners, utility staff, plant operators, and regulators that the results of recovery analysis should be skeptically viewed. This paper provides a framework for discussing recovery strategies, thus lending credibility to the process and enhancing regulatory acceptance of PRA results and conclusions. (author)

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

On the sequence and consequences of the Chernobyl reactor accident

Energy Technology Data Exchange (ETDEWEB)

Hennies, H H

1986-01-01

A serious reactor accident occurred on April 26, 1986 at Chernobyl near Kiev (Soviet Union) where, after melting of the core, there was a considerable release of radioactivity to the environment and to the atmosphere. The radioactivity release caused irradiation of the operating staff, which led to 24 deaths by June 1986. Hardly anything is known about the irradiation of the environment of the reactor plant, but the population within a radius of 30 km was evacuated. The radioactivity released into the atmosphere spread all over Europe, and Germany was affected a few days after the accident. The article gives a short description of the plant which suffered the accident, one tries to describe the course of the accident and to discuss the applicability to German plants.

Application of MELCOR Code to a French PWR 900 MWe Severe Accident Sequence and Evaluation of Models Performance Focusing on In-Vessel Thermal Hydraulic Results

International Nuclear Information System (INIS)

De Rosa, Felice

2006-01-01

In the ambit of the Severe Accident Network of Excellence Project (SARNET), funded by the European Union, 6. FISA (Fission Safety) Programme, one of the main tasks is the development and validation of the European Accident Source Term Evaluation Code (ASTEC Code). One of the reference codes used to compare ASTEC results, coming from experimental and Reactor Plant applications, is MELCOR. ENEA is a SARNET member and also an ASTEC and MELCOR user. During the first 18 months of this project, we performed a series of MELCOR and ASTEC calculations referring to a French PWR 900 MWe and to the accident sequence of 'Loss of Steam Generator (SG) Feedwater' (known as H2 sequence in the French classification). H2 is an accident sequence substantially equivalent to a Station Blackout scenario, like a TMLB accident, with the only difference that in H2 sequence the scram is forced to occur with a delay of 28 seconds. The main events during the accident sequence are a loss of normal and auxiliary SG feedwater (0 s), followed by a scram when the water level in SG is equal or less than 0.7 m (after 28 seconds). There is also a main coolant pumps trip when ΔTsat < 10 deg. C, a total opening of the three relief valves when Tric (core maximal outlet temperature) is above 603 K (330 deg. C) and accumulators isolation when primary pressure goes below 1.5 MPa (15 bar). Among many other points, it is worth noting that this was the first time that a MELCOR 1.8.5 input deck was available for a French PWR 900. The main ENEA effort in this period was devoted to prepare the MELCOR input deck using the code version v.1.8.5 (build QZ Oct 2000 with the latest patch 185003 Oct 2001). The input deck, completely new, was prepared taking into account structure, data and same conditions as those found inside ASTEC input decks. The main goal of the work presented in this paper is to put in evidence where and when MELCOR provides good enough results and why, in some cases mainly referring to its

A Preliminary Neutral Framework for the Accident Sequence Evaluation for a Hydrogen Conversion Reactor

International Nuclear Information System (INIS)

Han, Seok Jung; Yang, Joon Eon

2005-01-01

A framework for an early stage PSA for a hydrogen conversion reactor has been proposed in this paper. The approach is based on a functional and top-down approach. A main concerning point of this approach is to use a design neutral framework. A design neutral framework of PSA can provide a flexibility to apply to several candidate design concepts or options. This neutral-framework idea was borrowed from a proposed regulatory framework in US NRC. The feasibility of our proposed approach has been assessed to be applied in an accident sequence analysis for a hydrogen conversion reactor

Iodine chemistry effect on source term assessments. A MELCOR 186 YT study of a PWR severe accident sequence

International Nuclear Information System (INIS)

Herranz, Luis E.; Garcia, Monica; Otero, Bernadette

2009-01-01

Level-2 Probabilistic Safety Analysis has demonstrated to be a powerful tool to give insights into multiple aspects concerning severe accidents: phenomena with the greatest potential to lead to containment failure, safety systems performance and, even, to identify any additional accident management that could mitigate the consequences of such an even, etc. A major result of level-2 PSA is iodine content in Source Term since it is the main responsible for the radiological impact during the first few days after a hypothetical severe accident. Iodine chemistry is known to considerably affect iodine behavior and although understanding has improved substantially since the early 90's, a thorough understanding is still missing and most PSA studies do not address it when assessing severe accident scenarios. This paper emphasizes the quantitative and qualitative significance of considering iodine chemistry in level-2 PSA estimates. To do so a cold leg break, low pressure severe accident sequence of an actual pressurized water reactor has been analyzed with the MELCOR 1.8.6 YT code. Two sets of calculations, with and without chemistry, have been carried out and compared. The study shows that iodine chemistry could result in an iodine release to environment about twice higher, most of which would consist of around 60% of iodine in gaseous form. From these results it is concluded that exploratory studies on the potential effect of iodine chemistry on source term estimates should be carried out. (author)

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

International Nuclear Information System (INIS)

Silva, Kampanart; Ishiwatari, Yuki; Takahara, Shogo

2013-01-01

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

NIF: Impacts of chemical accidents and comparison of chemical/radiological accident approaches

International Nuclear Information System (INIS)

Lazaro, M.A.; Policastro, A.J.; Rhodes, M.

1996-01-01

The US Department of Energy (DOE) proposes to construct and operate the National Ignition Facility (NIF). The goals of the NIF are to (1) achieve fusion ignition in the laboratory for the first time by using inertial confinement fusion (ICF) technology based on an advanced-design neodymium glass solid-state laser, and (2) conduct high-energy-density experiments in support of national security and civilian applications. The primary focus of this paper is worker-public health and safety issues associated with postulated chemical accidents during the operation of NIF. The key findings from the accident analysis will be presented. Although NIF chemical accidents will be emphasized, the important differences between chemical and radiological accident analysis approaches and the metrics for reporting results will be highlighted. These differences are common EIS facility and transportation accident assessments

Evaluating advancements in accident investigations using a novel framework

NARCIS (Netherlands)

Karanikas, N.; Soltani, P.; de Boer, R.J.; Roelen, A.

2015-01-01

Safety is monitored by various proactive and reactive methods, including the investigation of adverse accidents and incidents, which are collectively known as safety investigations. In this study we demonstrate how accident and incident investigation reports can be useful to identify implicit safety

Development of the severe accident risk information database management system SARD

International Nuclear Information System (INIS)

Ahn, Kwang Il; Kim, Dong Ha

2003-01-01

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

Development of the severe accident risk information database management system SARD

Energy Technology Data Exchange (ETDEWEB)

Ahn, Kwang Il; Kim, Dong Ha

2003-01-01

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

Analysis of accident sequences and source terms at waste treatment and storage facilities for waste generated by U.S. Department of Energy Waste Management Operations, Volume 3: Appendixes C-H

International Nuclear Information System (INIS)

Mueller, C.; Nabelssi, B.; Roglans-Ribas, J.

1995-04-01

This report contains the Appendices for the Analysis of Accident Sequences and Source Terms at Waste Treatment and Storage Facilities for Waste Generated by the U.S. Department of Energy Waste Management Operations. The main report documents the methodology, computational framework, and results of facility accident analyses performed as a part of the U.S. Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies are assessed, and the resultant radiological and chemical source terms are evaluated. A personal computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for calculation of human health risk impacts. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also provide discussion of specific accident analysis data and guidance used or consulted in this report

Analysis of accident sequences and source terms at waste treatment and storage facilities for waste generated by U.S. Department of Energy Waste Management Operations, Volume 3: Appendixes C-H

Energy Technology Data Exchange (ETDEWEB)

Mueller, C.; Nabelssi, B.; Roglans-Ribas, J. [and others

1995-04-01

This report contains the Appendices for the Analysis of Accident Sequences and Source Terms at Waste Treatment and Storage Facilities for Waste Generated by the U.S. Department of Energy Waste Management Operations. The main report documents the methodology, computational framework, and results of facility accident analyses performed as a part of the U.S. Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies are assessed, and the resultant radiological and chemical source terms are evaluated. A personal computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for calculation of human health risk impacts. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also provide discussion of specific accident analysis data and guidance used or consulted in this report.

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

Recent advances in nanopore-based nucleic acid analysis and sequencing

International Nuclear Information System (INIS)

Shi, Jidong; Fang, Ying; Hou, Junfeng

2016-01-01

Nanopore-based sequencing platforms are transforming the field of genomic science. This review (containing 116 references) highlights some recent progress on nanopore-based nucleic acid analysis and sequencing. These studies are classified into three categories, biological, solid-state, and hybrid nanopores, according to their nanoporous materials. We begin with a brief description of the translocation-based detection mechanism of nanopores. Next, specific examples are given in nanopore-based nucleic acid analysis and sequencing, with an emphasis on identifying strategies that can improve the resolution of nanopores. This review concludes with a discussion of future research directions that will advance the practical applications of nanopore technology. (author)

MELCOR assessment of sequential severe accident mitigation actions under SGTR accident

International Nuclear Information System (INIS)

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

2017-01-01

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

The United States Department of Energy (DOE) Computerized Accident/Incident Reporting System (CAIRS)

International Nuclear Information System (INIS)

Briscoe, G.J.

1993-01-01

The Department of Energy's (DOE) Computerized Accident/Incident Reporting System (CAIRS) is a comprehensive data base containing more than 50,000 investigation reports of injury/illness, property damage and vehicle accident cases representing safety data from 1975 to the present for more than 150 DOE contractor organizations. A special feature is that the text of each accident report is translated using a controlled dictionary and rigid sentence structure called Factor Relationship and Sequence of Events (FRASE) that enhances the ability to retrieve specific types of information and to perform detailed analyses. DOE summary and individual contractor reports are prepared quarterly and annually. In addition, ''Safety Performance Profile'' reports for individual organizations are prepared to provide advance information to appraisal teams, and special topical reports are prepared for areas of concern such as an increase in the number of security injuries or environmental releases. The data base is open to all DOE and Contractor registered users with no access restrictions other than that required by the Privacy Act

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

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

International Nuclear Information System (INIS)

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

2003-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2003-10-01

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

Accident and emergency management

International Nuclear Information System (INIS)

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

1988-06-01

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

The influence of the technologically advanced evacuation models on the risk analyses during accidents in LNG terminal

Energy Technology Data Exchange (ETDEWEB)

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

2014-07-01

The evacuation of people located in different safety zones of an LNG terminal is a complex problem considering that the accidents involving LNG are very hazardous and post the biggest threat to the safety of the people located near the LNG leakage. The safety risk criteria define the parameters which one LNG terminal should meet in terms of safety. Those criteria also contain an evacuation as an evasive action with the objective to mitigate the influence of the LNG accident on the people at risk. Till date, not a lot of attention has been paid to technologically advanced evacuations intended for LNG terminals. Creating the technologically advanced evacuation influences directly on the decrease of the probability of fatalities P{sub f,i}, thus influencing the calculation of the individual risk as well as the societal risk which results in the positioning of the F-N curve in the acceptable part of the ALARP zone. With this paper, we aim to present the difference between the safety analyses in cases when conservative data for P{sub f,i} is being used while calculating the risk, and in cases when real data for P{sub f,i} is been used. (Author)

Severe accident considerations for modern KWU-PWR plants

International Nuclear Information System (INIS)

Eyink, J.

1987-01-01

In assumption of severe accident on modern KWU-PWR plants the author discusses on the: selection of core meltdown sequences, course of the accident, containment behaviour and source terms for fission products release to the environment

A DNA sequence element that advances replication origin activation time in Saccharomyces cerevisiae.

Science.gov (United States)

Pohl, Thomas J; Kolor, Katherine; Fangman, Walton L; Brewer, Bonita J; Raghuraman, M K

2013-11-06

Eukaryotic origins of DNA replication undergo activation at various times in S-phase, allowing the genome to be duplicated in a temporally staggered fashion. In the budding yeast Saccharomyces cerevisiae, the activation times of individual origins are not intrinsic to those origins but are instead governed by surrounding sequences. Currently, there are two examples of DNA sequences that are known to advance origin activation time, centromeres and forkhead transcription factor binding sites. By combining deletion and linker scanning mutational analysis with two-dimensional gel electrophoresis to measure fork direction in the context of a two-origin plasmid, we have identified and characterized a 19- to 23-bp and a larger 584-bp DNA sequence that are capable of advancing origin activation time.

Source terms derived from analyses of hypothetical accidents, 1950-1986

International Nuclear Information System (INIS)

Stratton, W.R.

1987-01-01

This paper reviews the history of reactor accident source term assumptions. After the Three Mile Island accident, a number of theoretical and experimental studies re-examined possible accident sequences and source terms. Some of these results are summarized in this paper

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

SWR-1000 concept on control of severe accidents

International Nuclear Information System (INIS)

Meyer, P.J.

1998-01-01

It is essential for the SWR-1000 probabilistic safety concept to consider the results from experiments and reliability system failure within the probabilistic safety analyses for passive systems. Active and passive safety features together reduce the probability of the occurrence of beyond design basis accidents in order to limit their consequences in accordance with the German law. As a reference case we analyzed the most probable core melt accident sequence with a very conservative assumption. An initial event, stuck open of safety and relief valves without the probability of active and passive feeding systems of the pressure vessel, was considered. Other sequences of the loss of coolant accidents lead to lower probability

Advances in safety countermeasures at the Tomari NPP of Hokkaido Electric Power on the basis of Fukushima Daiichi NPP accident. Fire protection and other advances

International Nuclear Information System (INIS)

Shibata, Taku; Dasai, Katsumi

2014-01-01

Fire protections for the nuclear power plants have been based on the fire laws and the conventional guide. After Fukushima Daiichi NPP accident, many safety countermeasures - also about Fire Protection - have been discussed in the Japanese authorities. This paper shows our present activities in the Tomari NPP about the fire protections from the view points of Fire Prevention, Fire Detection/Suppression Systems and Fire Protection, and other advances. (author)

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.

Fukushima. The accident sequence and important causes. Pt. 2/3; Fukushima. Unfallablauf und wesentliche Ursachen. T. 2/3

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

In this part on the accident sequence in the NPP Fukushima Daiichi on March 11, 2011 the important safety systems of a nuclear power plant are described, including the design of a nuclear boiling water reactor with Mark-II type containment, the high-pressure injection system and the systems for afterheat removal. The chronology of the accident progress in the NPP units 1-3 is described. The units 4-6 were shutdown due to revision work. Due to the earthquake an electric power transformation station close to the NPP site and the power poles were destroyed, the redundant power supply of the neighboring electricity supplier Tohoku did not work. All emergency diesel generators were flooded and destroyed resulting in the so-called station blackout. Firefighting trucks and materials for radiation protection and the infrastructure at the NPP site were destroyed. The release of radioactivity induced a severe contamination of the reactor site.

Simulation of severe accident using March-3 computer code

International Nuclear Information System (INIS)

Fernandes, A.; Nakata, H.

1991-01-01

The severe accident sensitivity analysis utilizing the March-3 approximate modelization options has been performed. The reference results against which the present results have been compared were obtained from the best published results for the most representative accident sequences: TMLU, S sub(2)DC sub(r) and S sub(2)DCF sub(r) for the Zion-1 reactor. The results of the present sensitivity analysis revealed the presence of very crude modelizations, in the March-3 program, to represent the critical phenomenologies involved in the severe accident sequences considered, even though large uncertainties must still be taken into account due primarily to the scarcity of the integral benchmark data. (author)

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)

Loss-of-coolant accident mitigation for the Advanced Neutron Source Reactor

International Nuclear Information System (INIS)

Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr.

1994-01-01

A RELAP5 Advanced Neutron Source Reactor system model has been developed for the conceptual design safety analysis. Three major regions modeled are the core, the heat exchanger loops, and letdown/pressurizing system. The model has been used to examine design alternatives for mitigation of loss-of-coolant accident (LOCA) transients. The safety margins to the flow excursion limit and critical heat flux are presented. The results show that the core can survive an instantaneous double-ended guillotine of the core outlet piping break (610 mm-diameter) provided a cavitating venturi is employed. RELAP5 calculations were also used to determine the effects of using a non-instantaneous break opening times. Both break opening time and break formation characteristics were included in these parametric calculations. Accumulator optimization studies were also performed which suggest that an optimum accumulator bubble size exists which improves system performance under some break scenarios

Comparative analysis of a hypothetical 0.1 $/SEC transient overpower accident in an irradiated LMFBR core using different computer models

International Nuclear Information System (INIS)

Cacciabue, P.C.; Fremont, R. de; Renard, A.

1982-01-01

The Report gives the results of comparative calculations performed by the Whole Core Accident Codes Group which is a subgroup of the Safety Working Group of the Fast Reactor Coordinating Committee for a hypothetical transient overpower accident in an irradiated LMFBR core. Different computer codes from members of the European Community and the United States were used. The calculations are based on a Benchmark problem, using commonly agreed input data for the most important phenomena, such as the fuel pin failure threshold, FCl parameters, etc. Beside this, results with alternative assumptions for theoretical modelling are presented with the scope to show in a parametric way the influence of more advanced modelling capabilities and/or better (so-called best estimate) input data for the most important phenomena on the accident sequences

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)

Analysis of severe accidents in pressurized heavy water reactors

International Nuclear Information System (INIS)

2008-06-01

Certain very low probability plant states that are beyond design basis accident conditions and which may arise owing to multiple failures of safety systems leading to significant core degradation may jeopardize the integrity of many or all the barriers to the release of radioactive material. Such event sequences are called severe accidents. It is required in the IAEA Safety Requirements publication on Safety of the Nuclear Power Plants: Design, that consideration be given to severe accident sequences, using a combination of engineering judgement and probabilistic methods, to determine those sequences for which reasonably practicable preventive or mitigatory measures can be identified. Acceptable measures need not involve the application of conservative engineering practices used in setting and evaluating design basis accidents, but rather should be based on realistic or best estimate assumptions, methods and analytical criteria. Recently, the IAEA developed a Safety Report on Approaches and Tools for Severe Accident Analysis. This publication provides a description of factors important to severe accident analysis, an overview of severe accident phenomena and the current status in their modelling, categorization of available computer codes, and differences in approaches for various applications of severe accident analysis. The report covers both the in- and ex-vessel phases of severe accidents. The publication is consistent with the IAEA Safety Report on Accident Analysis for Nuclear Power Plants and can be considered as a complementary report specifically devoted to the analysis of severe accidents. Although the report does not explicitly differentiate among various reactor types, it has been written essentially on the basis of available knowledge and databases developed for light water reactors. Therefore its application is mostly oriented towards PWRs and BWRs and, to a more limited extent, they can be only used as preliminary guidance for other types of reactors

The DOE technology development programme on severe accident management

International Nuclear Information System (INIS)

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

1998-01-01

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

Severe Accident Sequence Analysis Program: Anticipated transient without scram simulations for Browns Ferry Nuclear Plant Unit 1

International Nuclear Information System (INIS)

Dallman, R.J.; Gottula, R.C.; Holcomb, E.E.; Jouse, W.C.; Wagoner, S.R.; Wheatley, P.D.

1987-05-01

An analysis of five anticipated transients without scram (ATWS) was conducted at the Idaho National Engineering Laboratory (INEL). The five detailed deterministic simulations of postulated ATWS sequences were initiated from a main steamline isolation valve (MSIV) closure. The subject of the analysis was the Browns Ferry Nuclear Plant Unit 1, a boiling water reactor (BWR) of the BWR/4 product line with a Mark I containment. The simulations yielded insights to the possible consequences resulting from a MSIV closure ATWS. An evaluation of the effects of plant safety systems and operator actions on accident progression and mitigation is presented

Computerized accident management support system: development for severe accident management

International Nuclear Information System (INIS)

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

1998-01-01

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

The significance of domino effect in chemical accidents

OpenAIRE

Hemmatian, Behrouz; Abdolhamidzadeh, B; Darbra Roman, Rosa Maria; Casal Fàbrega, Joaquim

2014-01-01

A historical survey was performed on 330 accidents involving domino effect, occurred in process/storage plants and in the transportation of hazardous materials; only accidents occurred after 1st-January-1961 have been considered. The main features – geographical location, type of accident, materials involved, origin and causes, consequences, domino sequences – were analyzed, with special consideration to the situation in the developing countries and compared to those from other previous surve...

ALWR severe accident issue resolution in support of updated emergency planning

International Nuclear Information System (INIS)

Additon, Stephen L.; Leaver, David E.; Sorrell, Steven W.; Theofanous, Theo G.

2004-01-01

The Advanced Light Water Reactor (ALWR) Program in the U.S. is a cooperative, cost-sharing undertaking between the U.S. government, industry, and a number of international participants, with the objective of developing the next generation of nuclear power plants. The ALWR designs emphasize improvements in safety and operational reliability through simplification, improved safety margins, innovative passive safety systems, enhanced man-machine interfaces, and incorporation of the lessons learned from the operation of existing LWR plants. An important component of the improved safety characteristics of ALWRs is the consideration of severe accidents in the plant design. The U.S. Department of Energy (DOE) initiated the Advanced Reactor Severe Accident Program (ARSAP) to assist in the transfer of severe accident technology from the U.S. national laboratories to the industry to implement this approach. The basic design requirements for this new generation of nuclear power plants were developed, under the management of the Electric Power Research Institute (EPRI) by the utilities and documented in the Utility Requirements Document (URD). The URD safety policy is based on the traditional 'defense-in-depth' approach, which emphasizes prevention through safety systems which prevent accidents from progressing to core damage, and mitigation to ensure that accidents are mitigated and contained. In a major departure from previous practice, severe accidents, including postulated core melt events, are specifically included in the defense-in-depth design considerations for ALWRs. As a result of this approach, the emergency planning assumptions and criteria warrant a review and reevaluation for ALWR designs. ALWRs present a risk profile that is significantly different than that which served as the basis for the emergency planning requirements for operating plants. The determination of this profile necessarily requires the characterization of the severe accident response of ALWRs

Supplemental analysis of accident sequences and source terms for waste treatment and storage operations and related facilities for the US Department of Energy waste management programmatic environmental impact statement

International Nuclear Information System (INIS)

Folga, S.; Mueller, C.; Nabelssi, B.; Kohout, E.; Mishima, J.

1996-12-01

This report presents supplemental information for the document Analysis of Accident Sequences and Source Terms at Waste Treatment, Storage, and Disposal Facilities for Waste Generated by US Department of Energy Waste Management Operations. Additional technical support information is supplied concerning treatment of transuranic waste by incineration and considering the Alternative Organic Treatment option for low-level mixed waste. The latest respirable airborne release fraction values published by the US Department of Energy for use in accident analysis have been used and are included as Appendix D, where respirable airborne release fraction is defined as the fraction of material exposed to accident stresses that could become airborne as a result of the accident. A set of dominant waste treatment processes and accident scenarios was selected for a screening-process analysis. A subset of results (release source terms) from this analysis is presented

Severe Accidents: French Regulatory Practice for Nuclear Power Plants

International Nuclear Information System (INIS)

Colin, M.

1997-01-01

In the framework of a continuous and iterative process, the French Safety Authority asks the utility EDF to implement equipment and procedure modifications on the operating reactors, in order to cope with the most likely Severe Accident sequences. As a result of Probabilistic Safety Assessments published in 1990, important equipment and procedure modifications are being implemented on the French PWRs to improve the safety in shutdown states. The implementation of another set of modifications against some reactivity accident sequences is also in progress. More recently, the Safety Authority expressed specific Severe Accident requirements in terms of instrumentation, equipment qualification, high pressure core melt accidents and hydrogen risk prevention. In that respect, EDF was asked to implement hydrogen recombiners on its reactors. On the other hand, the French Safety authority is involved with its German counterpart in the assessment process of the European Pressurized Water Reactor Project. In consistency with the common recommendations of the Safety Authorities involved, Severe Accident provisions for this reactor are being taken into account at the design stage

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)

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)

Thermohydraulic and safety analysis on China advanced research reactor under station blackout accident

International Nuclear Information System (INIS)

Tian Wenxi; Qiu Suizheng; Su Guanghui; Jia Dounan; Liu Xingmin; Zhang Jianwei

2007-01-01

A thermohydraulic and safety analysis code-TSACC has been developed using Fortran90 language to evaluate the transient thermohydraulic behavior of the China advanced research reactor (CARR) under station blackout accident (SBA). For the development of TSACC, a series of corresponding mathematical and physical models were applied. Point reactor neutron kinetics model was adopted for solving the reactor power. All possible flow and heat transfer conditions under station blackout accident were considered and the optional correlations were supplied. The usual finite difference method was abandoned and the integral technique was adopted to evaluate the temperature field of the plate type fuel elements. A new simple and convenient equation was proposed for the resolution of the transient behaviors of the main pump instead of the complicated four-quadrant model. Gear method and Adams method were adopted alternately for a better solution to the stiff differential equations describing the dynamic behavior of the CARR. The computational result of TSACC showed the adequacy of the safety margin of CARR under SBA. For the purpose of Verification and Validation (V and V), the simulated results of TSACC were compared with those of RELAP5/MOD3 and a good agreement was obtained. The adoption of modular programming techniques enables TASCC to be applied to other reactors by easily modifying the corresponding function modules

Occurrence and countermeasures of urban power grid accident

Science.gov (United States)

Wei, Wang; Tao, Zhang

2018-03-01

With the advance of technology, the development of network communication and the extensive use of power grids, people can get to know power grid accidents around the world through the network timely. Power grid accidents occur frequently. Large-scale power system blackout and casualty accidents caused by electric shock are also fairly commonplace. All of those accidents have seriously endangered the property and personal safety of the country and people, and the development of society and economy is severely affected by power grid accidents. Through the researches on several typical cases of power grid accidents at home and abroad in recent years and taking these accident cases as the research object, this paper will analyze the three major factors that cause power grid accidents at present. At the same time, combining with various factors and impacts caused by power grid accidents, the paper will put forward corresponding solutions and suggestions to prevent the occurrence of the accident and lower the impact of the accident.

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.

Analysis of accident sequences and source terms at treatment and storage facilities for waste generated by US Department of Energy waste management operations

Energy Technology Data Exchange (ETDEWEB)

Mueller, C.; Nabelssi, B.; Roglans-Ribas, J.; Folga, S.; Policastro, A.; Freeman, W.; Jackson, R.; Mishima, J.; Turner, S.

1996-12-01

This report documents the methodology, computational framework, and results of facility accident analyses performed for the US Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies assessed, and the resultant radiological and chemical source terms evaluated. A personal-computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for the calculation of human health risk impacts. The WM PEIS addresses management of five waste streams in the DOE complex: low-level waste (LLW), hazardous waste (HW), high-level waste (HLW), low-level mixed waste (LLMW), and transuranic waste (TRUW). Currently projected waste generation rates, storage inventories, and treatment process throughputs have been calculated for each of the waste streams. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated, and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. Key assumptions in the development of the source terms are identified. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also discuss specific accident analysis data and guidance used or consulted in this report.

Advanced fuels safety comparisons

International Nuclear Information System (INIS)

Grolmes, M.A.

1977-01-01

The safety considerations of advanced fuels are described relative to the present understanding of the safety of oxide fueled Liquid Metal Fast Breeder Reactors (LMFBR). Safety considerations important for the successful implementation of advanced fueled reactors must early on focus on the accident energetics issues of fuel coolant interactions and recriticality associated with core disruptive accidents. It is in these areas where the thermal physical property differences of the advanced fuel have the greatest significance

Airborne concentrations of radioactive materials in severe accidents

International Nuclear Information System (INIS)

Ross, D.F. Jr.; Denning, R.S.

1989-01-01

Radioactive materials would be released to the containment building of a commercial nuclear reactor during each of the stages of a severe accident. Results of analyses of two accident sequences are used to illustrate the magnitudes of these sources of radioactive materials, the resulting airborne mass concentrations, the characteristics of the airborne aerosols, the potential for vapor forms of radioactive materials, the effectiveness of engineered safety features in reducing airborne concentrations, and the release of radioactive materials to the environment. Ability to predict transport and deposition of radioactive materials is important to assessing the performance of containment safety features in severe accidents and in the development of accident management procedures to reduce the consequences of severe accidents

An estimation of the accident sequence of the LOCA groups for the PSA model of the KSNP

International Nuclear Information System (INIS)

Han, Seok Jung; Yang, Joon Eon

2004-01-01

A new trend of the probabilistic safety assessment (PSA) technology is to improve and enhance the current PSA model to be adequate for risk-informed applications (RIA). Requirements of a PSA model for the RIA are summarized as (1) reduction of the conservatism in the model utilizing all available information and (2) consideration of the specific features of a plant as designed, as operated. This is because the PSA based on conservatism and insufficient consideration of the plant-specific features resulted in a shadow effect on the assessment results. When a PSA model is used in a risk-informed application, more precise risk-information is more helpful to decision making process, so the reduction of the conservatism and the consideration of the plant-specific features in a PSA model are the most essential elements. Recently, an effort has been performed to modify the current PSA model for the Korea Standard Nuclear Power plant (KSNP) to be used in risk-informed applications. A re-estimation of the accident sequence of the loss of coolant accident (LOCA) groups for the PSA model of the KSNP has been performed

Methodology for the Assessment of Confidence in Safety Margin for Small Break Loss of Coolant Accident Sequences

Energy Technology Data Exchange (ETDEWEB)

Nagrale, D. B.; Prasad, M.; Rao, R. S.; Gaikwad, A.J., E-mail: avinashg@aerb.gov.in [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai (India)

2014-10-15

Deterministic Safety Analysis and Probabilistic Safety Assessment (PSA) analyses are used concurrently to assess the Nuclear Power Plant (NPP) safety. The conventional deterministic analysis is conservative. The best estimate plus uncertainty analysis is increasingly being used for deterministic calculation in NPPs. The PSA methodology aims to be as realistic as possible while integrating information about accident phenomena, plant design, operating practices, component reliability and human behaviour. The peak clad temperature (PCT) distribution provides an insight into the confidence in safety margin for an initiating event. The paper deals with the concept of calculating the peak clad temperature with 95 percent confidence and 95 percent probability (PCT{sub 95/95}) in small break loss of coolant accident (SBLOCA) and methodologies for assessing safety margin. Five input parameters mainly, nominal power level, decay power, fuel clad gap conductivity, fuel thermal conductivity and discharge coefficient, were selected. A Uniform probability density function was assigned to the uncertain parameters and these uncertainties are propagated using Latin Hypercube Sampling (LHS) technique. The sampled data for 5 parameters were randomly mixed by LHS to obtain 25 input sets. A non-core damage accident sequence was selected from the SBLOCA event tree of a typical VVER study to estimate the PCTs and safety margin. A Kolmogorov– Smirnov goodness-of-fit test was carried out for PCTs. The smallest value of safety margin would indicate the robustness of the system with 95% confidence and 95% probability. Regression analysis was also carried out using 1000 sample size for the estimating PCTs. Mean, variance and finally safety margin were analysed. (author)

Severe accident tests and development of domestic severe accident system codes

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

According to lessons learned from Fukushima-Daiichi NPS accidents, the safety evaluation will be started based on the NRA's New Safety Standards. In parallel with this movement, reinforcement of Severe Accident (SA) Measures and Accident Managements (AMs) has been undertaken and establishments of relevant regulations and standards are recognized as urgent subjects. Strengthening responses against nuclear plant hazards, as well as realistic protection measures and their standardization is also recognized as urgent subjects. Furthermore, decommissioning of Fukushima-Daiichi Unit1 through Unit4 is promoted diligently. Taking into account JNES's mission with regard to these SA Measures, AMs and decommissioning, movement of improving SA evaluation methodologies inside and outside Japan, and prioritization of subjects based on analyses of sequences of Fukushima-Daiichi NPS accidents, three viewpoints was extracted. These viewpoints were substantiated as the following three groups of R and D subjects: (1) Obtaining near term experimental subjects: Containment venting, Seawater injection, Iodine behaviors. (2) Obtaining mid and long experimental subjects: Fuel damage behavior at early phase of core degradation, Core melting and debris formation. (3) Development of a macroscopic level SA code for plant system behaviors and a mechanistic level code for core melting and debris formation. (author)

Severe accident tests and development of domestic severe accident system codes

International Nuclear Information System (INIS)

2013-01-01

According to lessons learned from Fukushima-Daiichi NPS accidents, the safety evaluation will be started based on the NRA's New Safety Standards. In parallel with this movement, reinforcement of Severe Accident (SA) Measures and Accident Managements (AMs) has been undertaken and establishments of relevant regulations and standards are recognized as urgent subjects. Strengthening responses against nuclear plant hazards, as well as realistic protection measures and their standardization is also recognized as urgent subjects. Furthermore, decommissioning of Fukushima-Daiichi Unit1 through Unit4 is promoted diligently. Taking into account JNES's mission with regard to these SA Measures, AMs and decommissioning, movement of improving SA evaluation methodologies inside and outside Japan, and prioritization of subjects based on analyses of sequences of Fukushima-Daiichi NPS accidents, three viewpoints was extracted. These viewpoints were substantiated as the following three groups of R and D subjects: (1) Obtaining near term experimental subjects: Containment venting, Seawater injection, Iodine behaviors. (2) Obtaining mid and long experimental subjects: Fuel damage behavior at early phase of core degradation, Core melting and debris formation. (3) Development of a macroscopic level SA code for plant system behaviors and a mechanistic level code for core melting and debris formation. (author)

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.)

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Thermalydraulic processes in the reactor coolant system of a BWR under severe accident conditions

International Nuclear Information System (INIS)

Hodge, S.A.

1990-01-01

Boiling water reactors (BWRs) incorporate many unique structural features that make their expected response under severe accident conditions very different from that predicted in the case of pressurized water reactor accident sequences. Automatic main steam isolation valve (MIV) closure as the vessel water level approaches the top of the core would cause reactor vessel isolation while automatic recirculation pump trip would limit the in-vessel flows to those characteristic of natural circulation (as disturbed by vessel relief valve actuation). This paper provides a discussion of the BWR control blade, channel box, core plate, control rod guide tube, and reactor vessel safety relief valve (SRV) configuration and the effects of these structural components upon thermal hydraulic processes within the reactor vessel under severe accident conditions. The dominant BWR severe accident sequences as determined by probabilistic risk assessment are described and the expected timing of events for the unmitigated short-term station blackout severe accident sequence at the Peach Bottom atomic power station is presented

Analysis of accident sequences and source terms at waste treatment and storage facilities for waste generated by U.S. Department of Energy Waste Management Operations, Volume 1: Sections 1-9

International Nuclear Information System (INIS)

Mueller, C.; Nabelssi, B.; Roglans-Ribas, J.

1995-04-01

This report documents the methodology, computational framework, and results of facility accident analyses performed for the U.S. Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies are assessed, and the resultant radiological and chemical source terms are evaluated. A personal computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for calculation of human health risk impacts. The methodology is in compliance with the most recent guidance from DOE. It considers the spectrum of accident sequences that could occur in activities covered by the WM PEIS and uses a graded approach emphasizing the risk-dominant scenarios to facilitate discrimination among the various WM PEIS alternatives. Although it allows reasonable estimates of the risk impacts associated with each alternative, the main goal of the accident analysis methodology is to allow reliable estimates of the relative risks among the alternatives. The WM PEIS addresses management of five waste streams in the DOE complex: low-level waste (LLW), hazardous waste (HW), high-level waste (HLW), low-level mixed waste (LLMW), and transuranic waste (TRUW). Currently projected waste generation rates, storage inventories, and treatment process throughputs have been calculated for each of the waste streams. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also provide discussion of specific accident analysis data and guidance used or consulted in this report

Serum homocysteine levels in cerebrovascular accidents.

Science.gov (United States)

Zongte, Zolianthanga; Shaini, L; Debbarma, Asis; Singh, Th Bhimo; Devi, S Bilasini; Singh, W Gyaneshwar

2008-04-01

Hyperhomocysteinemia has been considered an independent risk factor in the development of stroke. The present study was undertaken to evaluate serum homocysteine levels in patients with cerebrovascular accidents among the Manipuri population and to compare with the normal cases. Ninety-three cerebrovascular accident cases admitted in the hospital were enrolled for the study and twenty-seven age and sex matched individuals free from cerebrovascular diseases were taken as control group. Serum homocysteine levels were estimated by ELISA method using Axis homocysteine EIA kit manufactured by Ranbaxy Diagnostic Ltd. India. The finding suggests that hyperhomocysteinemia is associated with cerebrovascular accident with male preponderance, which increases with advancing age. However, whether hyperhomocysteinemia is the cause or the result of cerebrovascular accidents needs further investigations.

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Small break LOCA [loss of coolant accident] mitigation for Bellefonte

International Nuclear Information System (INIS)

Bayless, P.D.; Dobbe, C.A.

1986-01-01

Several 5-cm (2-in.) diameter cold leg break loss coolant accidents for the Bellefonte nuclear plant were analyzed as part of the Severe Accident Sequence Analysis Program. The transients assumed various system failures, and included the S 2 D sequence. Operator actions to mitigate the S 2 D transient were also investigated. The transients were analyzed until either core damage began or long-term decay heat removal was established. The S 2 D sequence was analyzed into the core damage phase of the transient. The analyses showed that the flow from one high pressure injection pump was necessary and sufficient to prevent core damage in the absence of operator actions. Operator actions were also able to prevent core damage for the S 2 D sequence

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

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.)

Method of assessing severe accident management strategies

International Nuclear Information System (INIS)

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

1991-01-01

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

Studies of potential severe accidents in Finnish nuclear power plants. Quarterly report 3. quarter 1987

International Nuclear Information System (INIS)

Aro, Ilari.

1989-07-01

This thesis is based on six publications dealing with severe accident studies in Finnish nuclear power plants. Main emphasis has been put on general technical bases and methodologies applied in severe accident evaluation in Finland. As an example of the use of the analysis and evaluation methods, the analysis of one representative accident sequence, t otal loss of AC power , has been presented for both Finnish power plant types. This accident sequence is required to be analyzed in the Finnish safety guide YVL 2.2 which deals with transient and accident analyses as a basis of technical solutions at nuclear powr plants. Two different analysis methods, MAAP 3.0 and MARCH 3/STCP have been used for receiving as complete a picture as possible of the flow of events and for verifying the models to some extent. Besides the use of the two different models, the method of sensitivity analysis has been used for evaluating the effects of some important technical parameters on the accident flow. Finally, conclusions of the applicability of the two methods for analyzing severe accident sequences in Finnish plants have been discussed

Reactor accidents and the environment

International Nuclear Information System (INIS)

Beattie, J.R.; Griffiths, R.F.; Kaiser, G.D.; Kinchin, G.H.

1978-01-01

This is a condensed version of a paper, entitled 'The Environmental Impact of Radioactive Releases from Accidents in Nuclear Power Reactors', by the authors, presented to the Nuclear Energy Panel of the International Atomic Energy Agency/United Nations Environmental Programme. Headings include - Effects of ionising radiation on man; number of deaths expected from leukaemia and other cancers; risk estimates for incidence of benign nodules and thyroid cancer; maximum permissible levels and emergency levels of radiation and radioactivity; ICRP recommended dose limits for members of the general public; atmospheric dispersion and modelling; ICRP emergency reference levels for 1 131 , Cs 137 , Ru 106 and Sr 90 ; environmental consequences of accidental releases from nuclear power reactors; environmental impact of accidents to Magnox gas-cooled reactors; environmental impact of accidents to advanced gas-cooled reactors; environmental impact of accidents to fast reactors; and nature of risks. consequences are examined in terms of early and late biological effects on man, and contamination of land areas. Serious accidents are of low probability of occurrence, and the risk of accidents to nuclear power reactors is estimated to be very small. 43 references. (U.K.)

Thermal and hydraulic behaviour of CANDU cores under severe accident conditions - final report. Vol. 1

International Nuclear Information System (INIS)

Rogers, J.T.

1984-06-01

This report gives the results of a study of the thermo-hydraulic aspects of severe accident sequences in CANDU reactors. The accident sequences considered are the loss of the moderator cooling system and the loss of the moderator heat sink, each following a large loss-of-coolant accident accompanied by loss of emergency coolant injection. Factors considered include expulsion and boil-off of the moderator, uncovery, overheating and disintegration of the fuel channels, quenching of channel debris, re-heating of channel debris following complete moderator expulsion, formation and possible boiling of a molten pool of core debris and the effectiveness of the cooling of the calandria wall by the shield tank water during the accident sequences. The effects of these accident sequences on the reactor containment are also considered. Results show that there would be no gross melting of fuel during moderator expulsion from the calandria, and for a considerable time thereafter, as quenched core debris re-heats. Core melting would not begin until about 135 minutes after accident initiation in a loss of the moderator cooling system and until about 30 minutes in a loss of the moderator heat sink. Eventually, a pool of molten material would form in the bottom of the calandria, which may or may not boil, depending on property values. In all cases, the molten core would be contained within the calandria, as long as the shield tank water cooling system remains operational. Finally, in the period from 8 to 50 hours after the initiation of the accident, the molten core would re-solidify within the calandria. There would be no consequent damage to containment resulting from these accident sequences, nor would there be a significant increase in fission product releases from containment above those that would otherwise occur in a dual failure LOCA plus LOECI

Risk evaluation method for faults by engineering approach. (2) Application concept of margin analysis utilizing accident sequences

International Nuclear Information System (INIS)

Kamiya, Masanobu; Kanaida, Syuuji; Kamiya, Kouichi; Sato, Kunihiko; Kuroiwa, Katsuya

2016-01-01

The influence of the fault displacement on the facility should to be evaluated not only by the activity of the fault but also by obtaining risk information by considering scenarios including such as the frequency and the degree of the hazard, which should be an appropriate approach for nuclear safety. An applicable concept of margin analysis utilizing accident sequences for evaluating the influence of the fault displacement is proposed. By use of this analysis, we can evaluate of the safety functions and margin for core damage, verify the efficiency of equipment of portable type and make a decision to take additional measures to reduce the risk by using obtained risk information. (author)

Modelling Accident Tolerant Fuel Concepts

Energy Technology Data Exchange (ETDEWEB)

Hales, Jason Dean [Idaho National Laboratory; Gamble, Kyle Allan Lawrence [Idaho National Laboratory

2016-05-01

The catastrophic events that occurred at the Fukushima-Daiichi nuclear power plant in 2011 have led to widespread interest in research of alternative fuels and claddings that are proposed to be accident tolerant. The United States Department of Energy (DOE) through its Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has funded an Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The ATF HIP is a three-year project to perform research on two accident tolerant concepts. The final outcome of the ATF HIP will be an in-depth report to the DOE Advanced Fuels Campaign (AFC) giving a recommendation on whether either of the two concepts should be included in their lead test assembly scheduled for placement into a commercial reactor in 2022. The two ATF concepts under investigation in the HIP are uranium silicide fuel and iron-chromium-aluminum (FeCrAl) alloy cladding. Utilizing the expertise of three national laboratory participants (Idaho National Laboratory, Los Alamos National Laboratory, and Argonne National Laboratory), a comprehensive multiscale approach to modeling is being used that includes atomistic modeling, molecular dynamics, rate theory, phase-field, and fuel performance simulations. Model development and fuel performance analysis are critical since a full suite of experimental studies will not be complete before AFC must prioritize concepts for focused development. In this paper, we present simulations of the two proposed accident tolerance fuel systems: U3Si2 fuel with Zircaloy-4 cladding, and UO2 fuel with FeCrAl cladding. Sensitivity analyses are completed using Sandia National Laboratories’ Dakota software to determine which input parameters (e.g., fuel specific heat) have the greatest influence on the output metrics of interest (e.g., fuel centerline temperature). We also outline the multiscale modelling approach being employed. Considerable additional work is required prior to preparing the recommendation report for the Advanced

Atucha-I source terms for sequences initiated by transients

International Nuclear Information System (INIS)

Baron, J.; Bastianelli, B.

1997-01-01

The present work is part of an expected source terms study in the Atucha I nuclear power plant during severe accidents. From the accident sequences with a significant probability to produce core damage, those initiated by operational transients have been identified as the most relevant. These sequences have some common characteristics, in the sense that all of them resume in the opening of the primary system safety valves, and leave this path open for the coolant loss. In the case these sequences continue as severe accidents, the same path will be used for the release of the radionuclides, from the core, through the primary system and to the containment. Later in the severe accident sequence, the failure of the pressure vessel will occur, and the corium will fall inside the reactor cavity, interacting with the concrete. During these processes, more radioactive products will be released inside the containment. In the present work the severe accident simulation initiated by a blackout is performed, from the point of view of the phenomenology of the behavior of the radioactive products, as they are transported in the piping, during the core-concrete interactions, and inside the containment buildings until it failure. The final result is the source term into the atmosphere. (author) [es

An overview of selected severe accident research and applications

International Nuclear Information System (INIS)

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

2004-01-01

Severe accident research is being conducted world wide by industry organizations, utilities, and regulatory agencies. As this research is disseminated, it is being applied by utilities when they perform their Individual Plant Examinations (IPEs) and consider the preparation of Accident Management programs. The research is associated with phenomenological assessments of containment challenges and associated uncertainties, severe accident codes and analysis tools, systematic evaluation processes, and accident management planning. The continued advancement of this research and its applications will significantly contribute to the enhanced safety and operation of nuclear power plants. (author)

Release of fission products during controlled loss-of-coolant accidents and hypothetical core meltdown accidents

International Nuclear Information System (INIS)

Albrecht, H.; Malinauskas, A.P.

1978-01-01

A few years ago the Projekt Nukleare Sicherheit joined the United States Nuclear Regulatory Commission in the development of a research program which was designed to investigate fission product release from light water reactor fuel under conditions ranging from spent fuel shipping cask accidents to core meltdown accidents. Three laboratories have been involved in this cooperative effort. At Argonne National Laboratory (ANL), the research effort has focused on noble gas fission product release, whereas at Oak Ridge National Laboratory (ORNL) and at Kernforschungszentrum Karlsruhe (KfK), the studies have emphasized the release of species other than the noble gases. In addition, the ORNL program has been directed toward the development of fission product source terms applicable to analyses of spent fuel shipping cask accidents and controlled loss-of-coolant accidents, and the KfK program has been aimed at providing similar source terms which are characteristic of core meltdown accidents. The ORNL results are presented for fission product release from defected fuel rods into a steam atmosphere over the temperature range 500 to 1200 0 C, and the KfK results for release during core meltdown sequences

Severe accident analysis using MARCH 1.0 code

International Nuclear Information System (INIS)

Guimaraes, A.C.F.

1987-09-01

The description and utilization of the MARCH 1.0 computer code, which aim to analyse physical phenomena associated with core meltdown accidents in PWR type reactors, are presented. The primary system is modeled as a single volume which is partitioned into a gas (steam and hydrogen) region and a water region. March predicts blowdown from the primary system in single phase. Based on results of the probabilistic safety analysis for the Zion and Indian Point Nuclear Power Plants, the S 2 HFX sequence accident for Angra-1 reactor is studied. The S 2 HFX sequence means that the loss of coolant accident occurs through small break in primary system with bot total failures of the reactor safety system and containment in yours recirculation modes, leading the core melt and the containment failure due to overpressurization. The obtained results were considered reasonable if compared with the results obtained for the Zion and Indian Point nuclear power plants. (Author) [pt

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

ATHLET validation using accident management experiments

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

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

Probabilistic risk assessment for the Los Alamos Meson Physics Facility worst-case design-basis accident

International Nuclear Information System (INIS)

Sharirli, M.; Butner, J.M.; Rand, J.L.; Macek, R.J.; McKinney, S.J.; Roush, M.L.

1992-01-01

This paper presents results from a Los Alamos National Laboratory Engineering and Safety Analysis Group assessment of the worse-case design-basis accident associated with the Clinton P. Anderson Meson Physics Facility (LAMPF)/Weapons Neutron Research (WNR) Facility. The primary goal of the analysis was to quantify the accident sequences that result in personnel radiation exposure in the WNR Experimental Hall following the worst-case design-basis accident, a complete spill of the LAMPF accelerator 1L beam. This study also provides information regarding the roles of hardware systems and operators in these sequences, and insights regarding the areas where improvements can increase facility-operation safety. Results also include confidence ranges to incorporate combined effects of uncertainties in probability estimates and importance measures to determine how variations in individual events affect the frequencies in accident sequences

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

Structural aspects of the Chernobyl accident

International Nuclear Information System (INIS)

Murray, R.C.; Cummings, G.E.

1988-01-01

On April 26, 1986 the world's worst nuclear power plant accident occurred at the Unit 4 of the Chernobyl Nuclear Power Station in the USSR. This paper presents a discussion of the design of the Chernobyl Power Plant, the sequence of events that led to the accident and the damage caused by the resulting explosion. The structural design features that contributed to the accident and resulting damage will be highlighted. Photographs and sketches obtained from various worldwide news agencies will be shown to try and gain a perspective of the extent of the damage. The aftermath, clean-up, and current situation will be discussed and the important lessons learned for the structural engineer will be presented. 15 refs., 10 figs

Severe Accident Test Station Design Document

Energy Technology Data Exchange (ETDEWEB)

Snead, Mary A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yan, Yong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howell, Michael [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Keiser, James R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-09-01

The purpose of the ORNL severe accident test station (SATS) is to provide a platform for evaluation of advanced fuels under projected beyond design basis accident (BDBA) conditions. The SATS delivers the capability to map the behavior of advanced fuels concepts under accident scenarios across various temperature and pressure profiles, steam and steam-hydrogen gas mixtures, and thermal shock. The overall facility will include parallel capabilities for examination of fuels and irradiated materials (in-cell) and non-irradiated materials (out-of-cell) at BDBA conditions as well as design basis accident (DBA) or loss of coolant accident (LOCA) conditions. Also, a supporting analytical infrastructure to provide the data-needs for the fuel-modeling components of the Fuel Cycle Research and Development (FCRD) program will be put in place in a parallel manner. This design report contains the information for the first, second and third phases of design and construction of the SATS. The first phase consisted of the design and construction of an out-of-cell BDBA module intended for examination of non-irradiated materials. The second phase of this work was to construct the BDBA in-cell module to test irradiated fuels and materials as well as the module for DBA (i.e. LOCA) testing out-of-cell, The third phase was to build the in-cell DBA module. The details of the design constraints and requirements for the in-cell facility have been closely captured during the deployment of the out-of-cell SATS modules to ensure effective future implementation of the in-cell modules.

Severe Accident Test Station Design Document

International Nuclear Information System (INIS)

Snead, Mary A.; Yan, Yong; Howell, Michael; Keiser, James R.; Terrani, Kurt A.

2015-01-01

The purpose of the ORNL severe accident test station (SATS) is to provide a platform for evaluation of advanced fuels under projected beyond design basis accident (BDBA) conditions. The SATS delivers the capability to map the behavior of advanced fuels concepts under accident scenarios across various temperature and pressure profiles, steam and steam-hydrogen gas mixtures, and thermal shock. The overall facility will include parallel capabilities for examination of fuels and irradiated materials (in-cell) and non-irradiated materials (out-of-cell) at BDBA conditions as well as design basis accident (DBA) or loss of coolant accident (LOCA) conditions. Also, a supporting analytical infrastructure to provide the data-needs for the fuel-modeling components of the Fuel Cycle Research and Development (FCRD) program will be put in place in a parallel manner. This design report contains the information for the first, second and third phases of design and construction of the SATS. The first phase consisted of the design and construction of an out-of-cell BDBA module intended for examination of non-irradiated materials. The second phase of this work was to construct the BDBA in-cell module to test irradiated fuels and materials as well as the module for DBA (i.e. LOCA) testing out-of-cell, The third phase was to build the in-cell DBA module. The details of the design constraints and requirements for the in-cell facility have been closely captured during the deployment of the out-of-cell SATS modules to ensure effective future implementation of the in-cell modules.

Chernobylsk accident (Causes and Consequences)- Part 2

International Nuclear Information System (INIS)

Esteves, D.

1986-09-01

The causes and consequences of the nuclear accident at Chernobylsk-4 reactor are shortly described. The informations were provided by Russian during the specialist meeting, carried out at seat of IAEA. The Russian nuclear panorama; the site, nuclear power plant characteristics and sequence of events; the immediate measurements after accident; monitoring/radioactive releases; environmental contamination and ecological consequences; measurements of emergency; recommendations to increase the nuclear safety; and recommendations of work groups, are presented. (M.C.K.) [pt

Analysis and research status of severe core damage accidents

International Nuclear Information System (INIS)

1984-03-01

The Severe Core Damage Research and Analysis Task Force was established in Nuclear Safety Research Center, Tokai Research Establishment, JAERI, in May, 1982 to make a quantitative analysis on the issues related with the severe core damage accident and also to survey the present status of the research and provide the required research subjects on the severe core damage accident. This report summarizes the results of the works performed by the Task Force during last one and half years. The main subjects investigated are as follows; (1) Discussion on the purposes and necessities of severe core damage accident research, (2) proposal of phenomenological research subjects required in Japan, (3) analysis of severe core damage accidents and identification of risk dominant accident sequences, (4) investigation of significant physical phenomena in severe core damage accidents, and (5) survey of the research status. (author)

Discussion of the concept of safety indicators from the point of view of TfUX2 accident sequence for Forsmark 3

International Nuclear Information System (INIS)

Bujor, A.

1991-01-01

This paper contains general considerations on the safety indicators, with details at the system level and for the operator actions. For the system analysis, a modular analysis at a low detailed level is proposed (Module System Approach) in order to emphasize the safety related aspects at the subsystem (module) level. The operator actions are divided in ''active actions'' (actions in the control room during incident/accident situations) and ''passive actions'' (actions during tests, maintenance, repairs, etc.) and are analysed separately. In the second part, a discussion of a possible way to apply some SI to the TfUX2 accident sequence for FORSMARK-3, is done. For the analysis of the Auxiliary Feedwater Systems (AFWS) an equation is proposed to derive target values for the failure probability on demand at the train level, given the target value at the system level, including the common cause failures between the redundant trains. (author) 6 tabs., 18 refs

Chernobyl reactor accident

International Nuclear Information System (INIS)

Malinauskas, A.P.; Buchanan, J.R.; Lorenz, R.A.; Yamashita, T.

1986-01-01

On April 26, 1986, an explosion occurred at the newest of four operating nuclear reactors at the Chernobyl site in the USSR. The accident initiated an international technical exchange of almost unprecedented magnitude; this exchange was climaxed with a meeting at the International Atomic Energy Agency in Vienna during the week of August 25, 1986. The meeting was attended by more than 540 official representatives from 51 countries and 20 international organizations. Information gleaned from that technical exchange is presented in this report. A description of the Chernobyl reactor, which differs significantly from commercial US reactors, is presented, the accident scenario advanced by the Russian delegation is discussed, and observations that have been made concerning fission product release are described

Dose calculations for severe LWR accident scenarios

International Nuclear Information System (INIS)

Margulies, T.S.; Martin, J.A. Jr.

1984-05-01

This report presents a set of precalculated doses based on a set of postulated accident releases and intended for use in emergency planning and emergency response. Doses were calculated for the PWR (Pressurized Water Reactor) accident categories of the Reactor Safety Study (WASH-1400) using the CRAC (Calculations of Reactor Accident Consequences) code. Whole body and thyroid doses are presented for a selected set of weather cases. For each weather case these calculations were performed for various times and distances including three different dose pathways - cloud (plume) shine, ground shine and inhalation. During an emergency this information can be useful since it is immediately available for projecting offsite radiological doses based on reactor accident sequence information in the absence of plant measurements of emission rates (source terms). It can be used for emergency drill scenario development as well

Preliminary steps towards assessing aerosol retention in the break stage of a dry steam generator during severe accident SGTR sequences

International Nuclear Information System (INIS)

Herranz, L.E.; Lopez del Pra, C.; Sanchez Velasco, F.J.

2006-01-01

Severe accidents SGTR sequences are identified as major contributors to risk of PWRs. Their relevance lies in the potential radioactive release from reactor coolant system to the environment. Lack of knowledge on the source term attenuation capability of the steam generator has avoided its consideration in probabilistic safety studies and severe accident management guidelines. This paper describes a research program presently under way on the aerosol retention in the nearby of the tube breach within the secondary side of the steam generation in the absence of water. Its development has been internationally framed within the EU-SGTR and the ARTIST program. Experimental activities are focused on setting up a reliable database in which the influence of gas mass flow rate, breach configuration and particle nature in the aerosol retention are properly considered. Theoretical activities are aimed at developing a predictive tool (ARISG) capable of assessing source term attenuation in the scenario with reasonable accuracy. Given the major importance of jet aerodynamics, 3D CFD analyses are being conducted to assist both test interpretation and model development. (author)

Accident analysis for PRC-II reactor

International Nuclear Information System (INIS)

Wei Yongren; Tang Gang; Wu Qing; Lu Yili; Liu Zhifeng

1997-12-01

The computer codes, calculation models, transient results, sensitivity research, design improvement, and safety evaluation used in accident analysis for PRC-II Reactor (The Second Pulsed Reactor in China) are introduced. PRC-II Reactor is built in big populous city, so the public pay close attention to reactor safety. Consequently, Some hypothetical accidents are analyzed. They include an uncontrolled control rod withdrawal at rated power, a pulse rod ejection at rated power, and loss of coolant accident. Calculation model which completely depict the principle and process for each accident is established and the relevant analysis code is developed. This work also includes comprehensive computing and analyzing transients for each accident of PRC-II Reactor; the influences in the reactor safety of all kind of sensitive parameters; evaluating the function of engineered safety feature. The measures to alleviate the consequence of accident are suggested and taken in the construction design of PRC-II Reactor. The properties of reactor safety are comprehensively evaluated. A new advanced calculation model (True Core Uncovered Model) of LOCA of PRC-II Reactor and the relevant code (MCRLOCA) are first put forward

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.

A framework for the assessment of severe accident management strategies

International Nuclear Information System (INIS)

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

1993-09-01

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

A framework for the assessment of severe accident management strategies

Energy Technology Data Exchange (ETDEWEB)

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

1993-09-01

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

Radiological consequence analyses of loss of coolant accidents of various break sizes of Pressurized Heavy Water Reactor

International Nuclear Information System (INIS)

Sanyasi Rao, V.V.S.; Hari Prasad, M.; Ghosh, A.K.

2010-01-01

For any advanced technology, it is essential to ensure that the consequences associated with the accident sequences arising, if any, from the operation of the plant are as low as possible and certainly below the guidelines/limits set by the regulatory bodies. Nuclear power is no exception to this. In this paper consequences of the events arising from Loss of Coolant Accident (LOCA) sequences in Pressurized Heavy Water Reactor (PHWR), are analysed. The sequences correspond to different break sizes of LOCA followed by the operation or otherwise of Emergency Core Cooling System (ECCS). Operation or otherwise of the containment safety systems has also been considered. It has been found that there are no releases to the environment when ECCS is available. The releases, when ECCS is not available, arise from the slack and the ground. The radionuclides considered include noble gases, iodine, and cesium. The hourly meteorological parameters (wind speed, wind direction, precipitation and stability category), considered for this study, correspond to those of Kakrapar site. The consequences evaluated are the thyroid dose and the bone marrow dose received by a person located at various distances from the release point. Isodose curves are generated. From these evaluations, it has been found that the doses are very low. The complementary cumulative frequency distributions (CCFD) for thyroid and bone marrow doses have also been presented for the cases analysed. (author)

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)

CNE (central nuclear en Embalse): probabilistic safety study. Loss-of-coolant accidents. Analysis through events sequence

International Nuclear Information System (INIS)

Layral, S.I.

1987-01-01

The aim of this study was to perform for the Embalse nuclear power plant, a probabilistic evaluation of loss-of-coolant accidents (LOCA) to identify the risks associated with them and to determine their acceptability in accordance with norms. This study includes all ruptures in the primary system that produce the automatic activation of 'emergency core cooling system'. Three starting events were selected for the probabilistic evaluation: 100% rupture of an input collector; 5% rupture of an input collector; 1.2% rupture of an input collector. At this stage the evaluation is focussed on the identification and quantization of the main failure sequences that follow a LOCA and lead to an uncontrolled reactor state or 'core meltdown'. The most important contribution to the core meltdown due to LOCA is the failure of supplies that are required for the emergency core cooling system. (Author)

Advanced sodium fast reactor accident source terms :

Energy Technology Data Exchange (ETDEWEB)

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

2010-09-01

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

Simulation of LOF accidents with directly electrical heated UO2 pins

International Nuclear Information System (INIS)

Alexas, A.

1976-01-01

The behavior of directly electrical heated UO 2 pins has been investigated under loss of coolant conditions. Two types of hypothetical accidents have been simulated, first, a LOF accident without power excursion (LOF accident) and second, a LOF accident with subsequent power excursion (LOF-TOP accident). A high-speed film shows the sequence of events for two characteristic experiments. In consequence of the high-speed film analysis as well as the metallographical evaluation statements are given in respect to the cladding meltdown process, the fuel melt fraction and the energy input from the beginning of a power transient to the beginning of the molten fuel ejections

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

Advances in operational safety and severe accident research

Energy Technology Data Exchange (ETDEWEB)

Simola, K. [VTT Automation (Finland)

2002-02-01

A project on reactor safety was carried out as a part of the NKS programme during 1999-2001. The objective of the project was to obtain a shared Nordic view of certain key safety issues related to the operating nuclear power plants in Finland and Sweden. The focus of the project was on selected central aspects of nuclear reactor safety that are of common interest for the Nordic nuclear authorities, utilities and research bodies. The project consisted of three sub-projects. One of them concentrated on the problems related to risk-informed deci- sion making, especially on the uncertainties and incompleteness of probabilistic safety assessments and their impact on the possibilities to use the PSA results in decision making. Another sub-project dealt with questions related to maintenance, such as human and organisational factors in maintenance and maintenance management. The focus of the third sub-project was on severe accidents. This sub-project concentrated on phenomenological studies of hydrogen combustion, formation of organic iodine, and core re-criticality due to molten core coolant interaction in the lower head of reactor vessel. Moreover, the current status of severe accident research and management was reviewed. (au)

Conceptual design loss-of-coolant accident analysis for the Advanced Neutron Source reactor

International Nuclear Information System (INIS)

Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr.

1994-01-01

A RELAP5 system model for the Advanced Neutron Source Reactor has been developed for performing conceptual safety analysis report calculations. To better represent thermal-hydraulic behavior of the core, three specific changes in the RELAP5 computer code were implemented: a turbulent forced-convection heat transfer correlation, a critical heat flux (CHF) correlation, and an interfacial drag correlation. The model consists of the core region, the heat exchanger loop region, and the pressurizing/letdown system region. Results for three loss-of-coolant accident analyses are presented: (1) an instantaneous double-ended guillotine (DEG) core outlet break with a cavitating venturi installed downstream of the core, (b) a core pressure boundary tube outer wall rupture, and (c) a DEG core inlet break with a finite break-formation time. The results show that the core can survive without exceeding the flow excursion of CHF thermal limits at a 95% probability level if the proper mitigation options are provided

Medical aid in the initial period of radiation accident

International Nuclear Information System (INIS)

Selidovkin, G.D.

1995-01-01

The main tasks of medical arrangements on the initial stage of rendering aid after radiation accident are the prime medical classification of the injured persons among the personnel of the plant and population, and realization of measures to avoid the increase of doses. The volume of medical aid depends on the type of accident, on the after-accident radiation situation, on the influence of hazardous factors, on the number of people involved in accident situation and the spectrum of sanitary losses, etc., which is to be predicted in advance and to be taken into consideration when rendering aid. The proper and sufficient aid on the initial stage will build the foundation of the ultimate efficiency of medical aid after radiation accident. 14 refs

Injury protection and accident causation parameters for vulnerable road users based on German In-Depth Accident Study GIDAS.

Science.gov (United States)

Otte, Dietmar; Jänsch, Michael; Haasper, Carl

2012-01-01

Within a study of accident data from GIDAS (German In-Depth Accident Study), vulnerable road users are investigated regarding injury risk in traffic accidents. GIDAS is the largest in-depth accident study in Germany. Due to a well-defined sampling plan, representativeness with respect to the federal statistics is also guaranteed. A hierarchical system ACASS (Accident Causation Analysis with Seven Steps) was developed in GIDAS, describing the human causation factors in a chronological sequence. The accordingly classified causation factors - derived from the systematic of the analysis of human accident causes ("7 steps") - can be used to describe the influence of accident causes on the injury outcome. The bases of the study are accident documentations over ten years from 1999 to 2008 with 8204 vulnerable road users (VRU), of which 3 different groups were selected as pedestrians n=2041, motorcyclists n=2199 and bicyclists n=3964, and analyzed on collisions with cars and trucks as well as vulnerable road users alone. The paper will give a description of the injury pattern and injury mechanisms of accidents. The injury frequencies and severities are pointed out considering different types of VRU and protective measures of helmet and clothes of the human body. The impact points are demonstrated on the car, following to conclusion of protective measures on the vehicle. Existing standards of protection devices as well as interdisciplinary research, including accident and injury statistics, are described. With this paper, a summarization of the existing possibilities on protective measures for pedestrians, bicyclists and motorcyclists is given and discussed by comparison of all three groups of vulnerable road users. Also the relevance of special impact situations and accident causes mainly responsible for severe injuries are pointed out, given the new orientation of research for the avoidance and reduction of accident patterns. 2010 Elsevier Ltd. All rights reserved.

Protein Science by DNA Sequencing: How Advances in Molecular Biology Are Accelerating Biochemistry.

Science.gov (United States)

Higgins, Sean A; Savage, David F

2018-01-09

A fundamental goal of protein biochemistry is to determine the sequence-function relationship, but the vastness of sequence space makes comprehensive evaluation of this landscape difficult. However, advances in DNA synthesis and sequencing now allow researchers to assess the functional impact of every single mutation in many proteins, but challenges remain in library construction and the development of general assays applicable to a diverse range of protein functions. This Perspective briefly outlines the technical innovations in DNA manipulation that allow massively parallel protein biochemistry and then summarizes the methods currently available for library construction and the functional assays of protein variants. Areas in need of future innovation are highlighted with a particular focus on assay development and the use of computational analysis with machine learning to effectively traverse the sequence-function landscape. Finally, applications in the fundamentals of protein biochemistry, disease prediction, and protein engineering are presented.

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.

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)

Lessons from the Fukushima nuclear power accident

International Nuclear Information System (INIS)

Hatamura, Yotaro

2013-01-01

Through the investigation of the Fukushima Nuclear Power Accident as the chairman of the related Government's Committee, many things had been considered. Essence of the accident could be not only what occurred in the Fukushima nuclear power station, but also dispersed radioactive materials forced many residents to move and not to be returned. Such events as indication errors of water level meter occurring in severe accident could no be thought and remote mechanical operation of valves under high radiation environment were not prepared. Contamination by radioactive clouds caused the evacuation of residents for a long period. Lessons learned from the accident were described such as; (1) the verification of the road to failure connecting selected accident sequence and road to success with another supposed choice, (2) considering what might occur and then what should be needed on the contrary, (3) nuclear power, if should be continued, should be used with the premise of its hazards, and (4) advise to nuclear engineer for adequate information dissemination and technical explanation to the public and keeping nuclear technologies alive. (T. Tanaka)

On preparation for accident management in LWR power stations

International Nuclear Information System (INIS)

1996-01-01

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

Advanced computational methods for the assessment of reactor core behaviour during reactivity initiated accidents. Final report; Fortschrittliche Rechenmethoden zum Kernverhalten bei Reaktivitaetsstoerfaellen. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Pautz, A.; Perin, Y.; Pasichnyk, I.; Velkov, K.; Zwermann, W.; Seubert, A.; Klein, M.; Gallner, L.; Krzycacz-Hausmann, B.

2012-05-15

The document at hand serves as the final report for the reactor safety research project RS1183 ''Advanced Computational Methods for the Assessment of Reactor Core Behavior During Reactivity-Initiated Accidents''. The work performed in the framework of this project was dedicated to the development, validation and application of advanced computational methods for the simulation of transients and accidents of nuclear installations. These simulation tools describe in particular the behavior of the reactor core (with respect to neutronics, thermal-hydraulics and thermal mechanics) at a very high level of detail. The overall goal of this project was the deployment of a modern nuclear computational chain which provides, besides advanced 3D tools for coupled neutronics/ thermal-hydraulics full core calculations, also appropriate tools for the generation of multi-group cross sections and Monte Carlo models for the verification of the individual calculational steps. This computational chain shall primarily be deployed for light water reactors (LWR), but should beyond that also be applicable for innovative reactor concepts. Thus, validation on computational benchmarks and critical experiments was of paramount importance. Finally, appropriate methods for uncertainty and sensitivity analysis were to be integrated into the computational framework, in order to assess and quantify the uncertainties due to insufficient knowledge of data, as well as due to methodological aspects.

Fuel behaviour in the case of severe accidents and potential ATF designs. Fuel Behavior in Severe Accidents and Potential Accident Tolerance Fuel Designs

International Nuclear Information System (INIS)

Cheng, Bo

2013-01-01

This presentation reviews the conditions of fuel rods under severe loss of coolant conditions, approaches that may increase coping time for plant operators to recover, requirements of advanced fuel cladding to increase tolerance in accident conditions, potential candidate alloys for accident-tolerant fuel cladding and a novel design of molybdenum (Mo) -based fuel cladding. The current Zr-alloy fuel cladding will lose all its mechanical strength at 750-800 deg. C, and will react rapidly with high-pressure steam, producing significant hydrogen and exothermic heat at 700-1000 deg. C. The metallurgical properties of Zr make it unlikely that modifications of the Zr-alloy will improve the behaviour of Zr-alloys at temperatures relevant to severe accidents. The Mo-based fuel cladding is designed to (1) maintain fuel rod integrity, and reduce the release rate of hydrogen and exothermic heat in accident conditions at 1200-1500 deg. C. The EPRI research has thus far completed the design concepts, demonstration of feasibility of producing very thin wall (0.2 mm) Mo tubes. The feasibility of depositing a protective coating using various techniques has also been demonstrated. Demonstration of forming composite Mo-based cladding via mechanical reduction has been planned

Accidents - Chernobyl accident

International Nuclear Information System (INIS)

2004-01-01

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

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)

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.

Analysis of effects of calandria tube uncovery under severe accident conditions in CANDU reactors

International Nuclear Information System (INIS)

Rogers, J.T.; Currie, T.C.; Atkinson, J.C.; Dick, R.

1983-01-01

A study is being undertaken for the Atomic Energy Control Board to assess the thermal and hydraulic behaviour of CANDU reactor cores under accident conditions more severe than those normally considered in the licensing process. In this paper, we consider the effects on a coolant channel of the uncovery of a calandria tube by moderator boil-off following a LOCA in a Bruce reactor unit in which emergency cooling is ineffective and the moderator heat sink is impaired by the failure of the moderator cooling system. Calandria tube uncovery and its immediate consequences, as described here, constitute only one part of the entire accident sequence. Other aspects of this sequence as well as results of the analysis of the other accident sequences studied will be described in the final report on the project and in later papers

MELCOR Severe Accident Analysis on the SMART Reactor

International Nuclear Information System (INIS)

Kim, Tae Woon; Jin, Young Ho; Kim, Young In; Kim, Keung Koo; Wang, Ziao; Revankar, Shripad

2014-01-01

A severe accident is analyzed for Korea SMR reactor, SMART. Core melt down sequences are analyzed for SMART reactor core using MELCOR version 1.8.5. MELCOR is developed by Sandia National Laboratory for US NRC for the simulation of severe accidents in nuclear power plants. Two cases are simulated here and compared between them; one is the case for core having 3 concentric rings and the other is the case for core having 5 concentric rings. One inch break LOCA scenario is simulated and compared between these two core models. Time sequences for the thermal hydraulic behaviors of RPV and thermal heatup behaviors of reactor core are explained in graphically. Thermal hydraulic behavior such as the change of pressure, level, mass, and temperature of RPV is explained. Thermal heatup behavior of reactor core such as oxidation of cladding, hydrogen generation, core slumping down to lower plenum, and finally creep rupture of PRV lower head is explained. Engineered safety features such as safety injection systems (SIS), and Passive residual heat removal systems (PHRS), etc. are assumed to be not working. One inch break of severe accident is simulated on Korean SMR (SMART) Integral PWR with MELCOR code version 1.8.5. Core melt progression and lower head failure time is very slow compared to other commercial reactors. Simulation on 3 and 5 radial rings core models gives very similar pattern in core cell failure timings. Other various accident scenarios (for example, SBO in Fukushima) will be tried further. Containment behaviors and source term behaviors in severe accident conditions will be analyzed in future

Thirty years after the Chernobyl accident: What lessons have we learnt?

International Nuclear Information System (INIS)

Beresford, N.A.; Fesenko, S.; Konoplev, A.; Skuterud, L.; Smith, J.T.; Voigt, G.

2016-01-01

April 2016 sees the 30 th anniversary of the accident at the Chernobyl nuclear power plant. As a consequence of the accident populations were relocated in Belarus, Russia and Ukraine and remedial measures were put in place to reduce the entry of contaminants (primarily 134+137 Cs) into the human food chain in a number of countries throughout Europe. Remedial measures are still today in place in a number of countries, and areas of the former Soviet Union remain abandoned. The Chernobyl accident led to a large resurgence in radioecological studies both to aid remediation and to be able to make future predictions on the post-accident situation, but, also in recognition that more knowledge was required to cope with future accidents. In this paper we discuss, what in the authors' opinions, were the advances made in radioecology as a consequence of the Chernobyl accident. The areas we identified as being significantly advanced following Chernobyl were: the importance of semi-natural ecosystems in human dose formation; the characterisation and environmental behaviour of ‘hot particles'; the development and application of countermeasures; the “fixation” and long term bioavailability of radiocaesium and; the effects of radiation on plants and animals. - Highlights: • A review of 30 years of radioecological studies following the 1986 Chernobyl accident. • Key contributions to radioecology from post-Chernobyl research are discussed.

Bus accident severity and passenger injury: evidence from Denmark

DEFF Research Database (Denmark)

Prato, Carlo Giacomo; Kaplan, Sigal

2014-01-01

Purpose Bus safety is a concern not only in developing countries, but also in the U.S. and Europe. In Denmark, disentangling risk factors that are positively or negatively related to bus accident severity and injury occurrence to bus passengers can contribute to promote safety as an essential...... principle of sustainable transit and advance the vision “every accident is one too many”. Methods Bus accident data were retrieved from the national accident database for the period 2002–2011. A generalized ordered logit model allows analyzing bus accident severity and a logistic regression enables...... examining occurrence of injury to bus passengers. Results Bus accident severity is positively related to (i) the involvement of vulnerable road users, (ii) high speed limits, (iii) night hours, (iv) elderly drivers of the third party involved, and (v) bus drivers and other drivers crossing in yellow or red...

Development of instrumentation systems for severe accidents. 4. New accident tolerant in-containment pressure transducer for containment pressure monitoring system

International Nuclear Information System (INIS)

Oba, Masato; Teruya, Kuniyuki; Yoshitsugu, Makoto; Ikeuchi, Takeshi

2015-01-01

The accident at Tokyo Electric Power Company's Fukushima Dai-ichi Nuclear Power Plant (TF-1 accident) caused severe situations and resulted in a difficulty in measuring important parameters for monitoring plant conditions. Therefore, we have studied the TF-1 accident to select the important parameters that should be monitored at the severe accident and are developing the Severe Accident Instrumentations and Monitoring Systems that could measure the parameters in severe accident conditions. Mitsubishi Heavy Industries, LTD (MHI) developed a new accident tolerant containment pressure monitoring system and demonstrated that the monitoring system could endure extremely harsh environmental conditions that envelop severe accident environmental conditions inside a containment such as maximum operating temperature of up to 300degC and total integrated dose (TID) of 1 MGy gamma. The new containment pressure monitoring system comprises of a strain gage type pressure transducer and a mineral insulated (MI) cable with ceramic connectors, which are located in the containment, and a strain measuring amplifier located outside the containment. Less thermal and radiation degradation is achieved because of minimizing use of organic materials for in-containment equipment such as the transducer and connectors. Several tests were performed to demonstrate the performance and capability of the in-containment equipment under severe accident environmental conditions and the major steps in this testing were run in the following test sequences: (1) the baseline functional tests (e.g., repeatability, non-linearity, hysteresis, and so on) under normal conditions, (2) accident radiation testing, (3) seismic testing, and (4) steam/temperature test exposed to simulated severe accident environmental conditions. The test results demonstrate that the new pressure transducer can endure the simulated severe accident conditions. (author)

Thirty years after the Chernobyl accident: What lessons have we learnt?

Science.gov (United States)

Beresford, N A; Fesenko, S; Konoplev, A; Skuterud, L; Smith, J T; Voigt, G

2016-06-01

April 2016 sees the 30(th) anniversary of the accident at the Chernobyl nuclear power plant. As a consequence of the accident populations were relocated in Belarus, Russia and Ukraine and remedial measures were put in place to reduce the entry of contaminants (primarily (134+137)Cs) into the human food chain in a number of countries throughout Europe. Remedial measures are still today in place in a number of countries, and areas of the former Soviet Union remain abandoned. The Chernobyl accident led to a large resurgence in radioecological studies both to aid remediation and to be able to make future predictions on the post-accident situation, but, also in recognition that more knowledge was required to cope with future accidents. In this paper we discuss, what in the authors' opinions, were the advances made in radioecology as a consequence of the Chernobyl accident. The areas we identified as being significantly advanced following Chernobyl were: the importance of semi-natural ecosystems in human dose formation; the characterisation and environmental behaviour of 'hot particles'; the development and application of countermeasures; the "fixation" and long term bioavailability of radiocaesium and; the effects of radiation on plants and animals. Copyright © 2016 The Authors. Published by Elsevier Ltd.. All rights reserved.

Core fusion accidents in nuclear power reactors. Knowledge review

International Nuclear Information System (INIS)

Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Cenerino, Gerard; Jacquemain, Didier; Raimond, Emmanuel; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno

2013-01-01

This reference document proposes a large and detailed review of severe core fusion accidents occurring in nuclear power reactors. It aims at presenting the scientific aspects of these accidents, a review of knowledge and research perspectives on this issue. After having recalled design and operation principles and safety principles for reactors operating in France, and the main studied and envisaged accident scenarios for the management of severe accidents in French PWRs, the authors describe the physical phenomena occurring during a core fusion accident, in the reactor vessel and in the containment building, their sequence and means to mitigate their effects: development of the accident within the reactor vessel, phenomena able to result in an early failure of the containment building, phenomena able to result in a delayed failure with the corium-concrete interaction, corium retention and cooling in and out of the vessel, release of fission products. They address the behaviour of containment buildings during such an accident (sizing situations, mechanical behaviour, bypasses). They review and discuss lessons learned from accidents (Three Mile Island and Chernobyl) and simulation tests (Phebus-PF). A last chapter gives an overview of software and approaches for the numerical simulation of a core fusion accident

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

A new approach to incorporate operator actions in the simulation of accident sequences

International Nuclear Information System (INIS)

Antonio Exposito; Juan Antonio Quiroga; Javier Hortal; John-Einar Hulsund

2006-01-01

Full text of publication follows: Nowadays, simulation-based human reliability analysis (HRA) methods seem to provide a new direction for the development of advanced methodologies to study operator actions effect during accident sequences. Due to this, the Spanish Nuclear Safety Council (CSN) started a working group which has, among other objectives, to develop such simulation-based HRA methodology. As a result of its activities, a new methodology, named Integrated Safety Assessment (ISA), has been developed and is currently being incorporated into licensing activities at CSN. One of the key aspects of this approach is the incorporation of the capability to simulate operator actions, expanding the ISA methodology scopes to make HRA studies. For this reason, CSN is involved in several activities oriented to develop a new tool, which must be able to incorporate operator actions in conventional thermohydraulic (TH) simulations. One of them is the collaboration project between CSN, Halden Reactor Project (HRP) and the Department of Energy Systems (DSE) of the Polytechnic University of Madrid that started in 2003. The basic aim of the project is to develop a software tool that consists of a closed-loop plant/operator simulator, a thermal hydraulic (TH) code for simulating the plant transient and the procedures processor to give the information related with operator actions to the TH code, both coupled by a data communication system which allows the information exchange. For the plant simulation we have a plant transient simulator code (TRETA/TIZONA for PWR/BWR NPPs respectively), developed by the CSN, with PWR/BWR full scope models. The functionality of these thermalhydraulic codes has been expanded, allowing control the overall information flow between coupled codes, simulating the TH transient and determining when the operator actions must be considered. In the other hand, we have the COPMA-III code, a computerized procedure system able to manage XML operational

Recent insights from severe accident research and implications for containment criteria for advanced LWRs

International Nuclear Information System (INIS)

Speis, T.P.; King, T.L.; Eltawila, F.

1992-01-01

The Severe Accident Research Program (SARP) was begun after the TMI-2 accident in March, 1979. The rule for dealing with the generation of large quantity of hydrogen in BWRs and Ice Condenser PWRs was promulgated by the Nuclear Regulatory Commission (NRC). The NRC issued severe Accident Policy Statement in 1985, and the revised SARP in 1989. In this paper, the current understanding of the more important phenomena and the associated mechanical and thermal loads to the containment is described, and the on-going works are summarized. The containment loadings in severe accidents are listed, and direct containment heating and the liner failure in BWR Mark I are added. A great deal of informations obtained on the early phase of melt progression are shown. The current understanding of the severe accident phenomena related to the containment and the on-going related research efforts are discussed more in detail. Fuel-coolant interaction including alpha-mode containment failure, direct containment heating, hydrogen deflagration and detonation, core-concrete interaction and debris coolability are described. (K.I.)

Vaporization of structural materials in severe accidents

International Nuclear Information System (INIS)

Lorenz, R.A.

1982-01-01

Vaporized structural materials form the bulk of aerosol particles that can transport fission products in severe LWR accidents. As part of the Severe Accident Sequence Analysis (SASA) program at Oak Ridge National Laboratory, a model has been developed based on a mass transport coefficient to describe the transport of materials from the surface of a molten pool. In many accident scenarios, the coefficient can be calculated from existing correlations for mass transfer by natural convection. Data from SASCHA fuel melting tests (Karlsruhe, Germany) show that the partial pressures of many of the melt components (Fe, Cr, Co, Mn, Sn) required for the model can be calculated from the vapor pressures of the pure species and Raoult's law. These calculations indicate much lower aerosol concentrations than reported in previous studies

Severe accidents and terrorist threats at nuclear reactors

International Nuclear Information System (INIS)

Pollack, G.L.

1987-01-01

Some of the key areas of uncertainty are the nature of the physical and chemical interactions of released fission products and of the interactions between a molten core and concrete, the completeness and validity of the computer codes used to predict accidents, and the behavior of the containment. Because of these and other uncertainties, it is not yet possible to reliably predict the consequences of reactor accidents. It is known that for many accident scenarios, especially less severe ones or where the containment is not seriously compromised, the amount of radioactive material expected to escape the reactor is less, even much less, than was previously calculated. For such accidents, the predictions are easier and more reliable. With severe accidents, however, there is considerable uncertainty as to the predicted results. For accidents of the type that terrorists might cause - for example, where the sequence of failure would be unexpected or where redundant safety features are caused to fail together - the uncertainties are still larger. The conclusion, then, is that there are potential dangers to the public from terrorist actions at a nuclear reactor; however, because of the variety of potential terrorist threats and the incompleteness of the knowledge about the behavior of reactor components and fission products during accidents, the consequences cannot yet be assessed quantitatively

Next-generation sequencing for endocrine cancers: Recent advances and challenges.

Science.gov (United States)

Suresh, Padmanaban S; Venkatesh, Thejaswini; Tsutsumi, Rie; Shetty, Abhishek

2017-05-01

Contemporary molecular biology research tools have enriched numerous areas of biomedical research that address challenging diseases, including endocrine cancers (pituitary, thyroid, parathyroid, adrenal, testicular, ovarian, and neuroendocrine cancers). These tools have placed several intriguing clues before the scientific community. Endocrine cancers pose a major challenge in health care and research despite considerable attempts by researchers to understand their etiology. Microarray analyses have provided gene signatures from many cells, tissues, and organs that can differentiate healthy states from diseased ones, and even show patterns that correlate with stages of a disease. Microarray data can also elucidate the responses of endocrine tumors to therapeutic treatments. The rapid progress in next-generation sequencing methods has overcome many of the initial challenges of these technologies, and their advantages over microarray techniques have enabled them to emerge as valuable aids for clinical research applications (prognosis, identification of drug targets, etc.). A comprehensive review describing the recent advances in next-generation sequencing methods and their application in the evaluation of endocrine and endocrine-related cancers is lacking. The main purpose of this review is to illustrate the concepts that collectively constitute our current view of the possibilities offered by next-generation sequencing technological platforms, challenges to relevant applications, and perspectives on the future of clinical genetic testing of patients with endocrine tumors. We focus on recent discoveries in the use of next-generation sequencing methods for clinical diagnosis of endocrine tumors in patients and conclude with a discussion on persisting challenges and future objectives.

The cost of nuclear accidents

International Nuclear Information System (INIS)

2015-01-01

Proposed by a technical section of the SFEN, and based on a meeting with representatives of different organisations (OECD-NEA, IRSN, EDF, and European Nuclear Energy Forum), this publication addresses the economic consequences of a severe accident (level 6 or 7) within an electricity producing nuclear power plant. Such an assessment essentially relies on three pillars: release of radio-elements outside the reactor, the scenario of induced consequences, and the method of economic quantification. After a recall and a comment of safety arrangements, and of the generally admitted probability of such an accident, this document notices that several actors are concerned by nuclear energy and are trying to assess accident costs. The issue of how to assess a cost (or costs) of a nuclear accident is discussed: there are in fact several types of costs and consequences. Thus, some costs can be rather precisely quantified when some others can be difficult to assess or with uncertainty. The relevance of some cost categories appears to be a matter of discussion and one must not forget that consequences can occur on a long term. The need for methodological advances is outlined and three categories of technical objectives are identified for the assessment (efficiency of safety measures to be put forward to mitigate the risk via a better accident management, compensation of victims and nuclear civil responsibility, and comparison of electricity production sectors and assessment of externalisation to guide public choices). It is outlined that the impact of accidents depend on several factors, that the most efficient mean to limit consequences of accidents is of course to limit radioactive emissions

Analysis of Three Mile Island - Unit 2 accident

International Nuclear Information System (INIS)

1979-07-01

The Nuclear Safety Analysis Center (NSAC) of the Electric Power Research Institute is analyzing the Three Mile Island-2 accident. An early result of this analysis was a brief narrative summary, issued in mid May 1979. The present report contains a revised version of that narrative summary, a highly detailed sequence of events, a standard reference list, a list of abbreviations and acronyms, and several appendices. The appendices serve either to describe plant features which are pertinent to the understanding of the sequence of events, or indicate how certain inferences and conclusions in the report were reached. Supplementing the appendices contained herein, additional appendices are in preparation; these will be issued when available (e.g., the appendices Hydrogen Phenomena and Operator Actions during Initial Transient will follow later). Also in preparation is a matrix of equipment and systems actions during the accident. This report together with future supplements and a separate Core Damage Assessment report, will embody the principal results of that phase of NSAC work which is devoted to learning and understanding what happened during the accident. Subsequent phases will concentrate on causes, lessons learned and generic remedial or preventive measures which may be appropriate

Analysis of Three Mile Island-Unit 2 accident

International Nuclear Information System (INIS)

1979-07-01

The Nuclear Safety Analysis Center (NSAC) of the Electic Power Research Institute is analyzing the Three Mile Island-2 accident. An early result of this analysis was a brief narrative summary, issued in mid-May 1979. The present report contains a revised version of that narrative summary, a highly detailed sequence of events, a standard reference list, a list of abbreviations and acronyms, and several appendices. The appendices serve either to describe plant features which are pertinent to the understanding of the sequence of events, or indicate how certain inferences and conclusions in the report were reached. Supplementing the appendices contained herein, additional appendices are in preparation; these will be issued when available (e.g., the appendices Hydrogen Phenomena and Operator Actions duing Initial Transient will follow later). Also in preparation is a matrix of equipment and systems actions during the accident. This report together with future supplements and a separate Core Damage Assessment report, will embody the principal results of that phase of NSAC's work which is devoted to learning and understanding what happened during the accident. Subsequent phases will concentrate on causes, lessons learned and generic remedial or preventive measures which may be appropriate

Passive depressurization accident management strategy for boiling water reactors

International Nuclear Information System (INIS)

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

2015-01-01

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

Passive depressurization accident management strategy for boiling water reactors

Energy Technology Data Exchange (ETDEWEB)

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

2015-04-01

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

Analysis of accident sequences and source terms at treatment and storage facilities for waste generated by US Department of Energy waste management operations. Volume 1: Sections 1-9

International Nuclear Information System (INIS)

Mueller, C.; Nabelssi, B.; Roglans-Ribas, J.; Folga, S.; Policastro, A.; Freeman, W.; Jackson, R.; Mishima, J.; Turner, S.

1996-12-01

This report documents the methodology, computational framework, and results of facility accident analyses performed for the US Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies assessed, and the resultant radiological and chemical source terms evaluated. A personal-computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for the calculation of human health risk impacts. The WM PEIS addresses management of five waste streams in the DOE complex: low-level waste (LLW), hazardous waste (HW), high-level waste (HLW), low-level mixed waste (LLMW), and transuranic waste (TRUW). Currently projected waste generation rates, storage inventories, and treatment process throughputs have been calculated for each of the waste streams. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated, and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. Key assumptions in the development of the source terms are identified. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also discuss specific accident analysis data and guidance used or consulted in this report

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Jerky driving--An indicator of accident proneness?

Science.gov (United States)

Bagdadi, Omar; Várhelyi, András

2011-07-01

This study uses continuously logged driving data from 166 private cars to derive the level of jerks caused by the drivers during everyday driving. The number of critical jerks found in the data is analysed and compared with the self-reported accident involvement of the drivers. The results show that the expected number of accidents for a driver increases with the number of critical jerks caused by the driver. Jerk analyses make it possible to identify safety critical driving behaviour or "accident prone" drivers. They also facilitate the development of safety measures such as active safety systems or advanced driver assistance systems, ADAS, which could be adapted for specific groups of drivers or specific risky driving behaviour. Copyright © 2011 Elsevier Ltd. All rights reserved.

Overview of severe accident research at KAERI

International Nuclear Information System (INIS)

Kim, H.D.; Kim, S.B.; Hong, S.W.; Kim, D.H.

2000-01-01

The severe accident research program at Korea Atomic Energy Research Institute, within the framework of governmental 10 year long-term nuclear R and D program, aims at the development of assessment techniques and accident management strategies for the prevention and mitigation of potential risk. The research program includes experimental efforts, development of phenomena specific models and development of an integrated computer code. The results of research program is intended to be utilized for the design of the advanced light water reactor and development of accident management strategies for the operating reactors. The main focused areas of recent investigation at KAERI are experiments on in-vessel core debris retention (SONATA-IV) and fuel coolant interaction (TROI) along with the development of models and integrated computer code (MIDAS). (author)

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.

Radionuclide release calculations for selected severe accident scenarios

International Nuclear Information System (INIS)

Denning, R.S.; Leonard, M.T.; Cybulskis, P.; Lee, K.W.; Kelly, R.F.; Jordan, H.; Schumacher, P.M.; Curtis, L.A.

1990-08-01

This report provides the results of source term calculations that were performed in support of the NUREG-1150 study. ''Severe Accident Risks: An Assessment for Five US Nuclear Power Plants.'' This is the sixth volume of a series of reports. It supplements results presented in the earlier volumes. Analyses were performed for three of the NUREG-1150 plants: Peach Bottom, a Mark I, boiling water reactor; Surry, a subatmospheric containment, pressurized water reactor; and Sequoyah, an ice condenser containment, pressurized water reactor. Complete source term results are presented for the following sequences: short term station blackout with failure of the ADS system in the Peach Bottom plant; station blackout with a pump seal LOCA for the Surry plant; station blackout with a pump seal LOCA in the Sequoyah plant; and a very small break with loss of ECC and spray recirculation in the Sequoyah plant. In addition, some partial analyses were performed which did not require running all of the modules of the Source Term Code Package. A series of MARCH3 analyses were performed for the Surry and Sequoyah plants to evaluate the effects of alternative emergency operating procedures involving primary and secondary depressurization on the progress of the accident. Only thermal-hydraulic results are provided for these analyses. In addition, three accident sequences were analyzed for the Surry plant for accident-induced failure of steam generator tubes. In these analyses, only the transport of radionuclides within the primary system and failed steam generator were examined. The release of radionuclides to the environment is presented for the phase of the accident preceding vessel meltthrough. 17 refs., 176 figs., 113 tabs

Instrumentation availability during severe accidents for a boiling water reactor with a Mark I containment

International Nuclear Information System (INIS)

Arcieri, W.C.; Hanson, D.J.

1992-02-01

In support of the US Nuclear Regulatory Commission Accident Management Research Program, the availability of instruments to supply accident management information during a broad range of severe accidents is evaluated for a Boiling Water Reactor with a Mark I containment. Results from this evaluation include: (1) the identification of plant conditions that would impact instrument performance and information needs during severe accidents; (2) the definition of envelopes of parameters that would be important in assessing the performance of plant instrumentation for a broad range of severe accident sequences; and (3) assessment of the availability of plant instrumentation during severe accidents

Planning for the Handling of Radiation Accidents

International Nuclear Information System (INIS)

1969-01-01

The developing atomic energy programmes and the widespread use of radiation sources in medicine, agriculture, industry and research have had admirable safety records. Throughout the world the number of known accidents in which persons have been exposed to harmful am ounts of ionizing radiation is relatively small, and only a few deaths have occurred. Meticulous precautions are being taken to maintain this good record in all work with radiation sources and to keep the exposure of persons as low as practicable. In spite of all the precautions that are taken, accidents may occur and they may be accompanied by the injury or death of persons and damage to property. It is only prudent to take those steps that are practicable to prevent accidents and to plan in advance the emergency action that would limit the injuries and damage caused by those accidents that do occur. Emergency plans should be sufficiently broad to cover unforeseen or very improbable accidents as well as those that are considered credible. Some accidents may involve only the workers in an establishment, those working directly with the source and possibly their colleagues. Other accidents may have consequences, notably in the form of radioactive contamination of the environment, that affect the general public, possibly far from the site of the accident. The preparation of plans for dealing with radiation accidents is therefore obligatory both for the various authorities that are responsible for protecting the health and the food and water supplies of the public, and for the operator of an installation containing radiation sources.

Smart Sensing of the Aux. Feed-water Pump Performance in NPP Severe Accidents Using Advanced GMDH Method

Energy Technology Data Exchange (ETDEWEB)

No, Young Gyu; Seong, Poong Hyun [KAIST, Daejeon (Korea, Republic of)

2016-05-15

In order to develop and verify the models, a number of data obtained by simulating station black out (SBO) scenario for the optimized power reactor 1000 (OPR1000) using MARS code were used. Most of monitoring systems for component have been suggested by using the directly measured data. However, it is very difficult to acquire data related to safety-critical component' status. Therefore, it is necessary to develop the new method that combines the data-based equipped with learning system and data miming technique. Many data-based modeling methods have been applied successfully to nuclear engineering area, such as signal validation, plant diagnostics and event identification. Also, the data miming is the process of analyzing data from different perspectives and summarizing it into useful information. In this study, the smart sensing technique was developed using advanced group method of data handing (GMDH) model. The original GMDH is an inductive self organizing algebraic model. The advanced GMDH model is equipped with a fuzzy concept. The proposed advanced GMDH model enhances the original GMDH model by reducing the effect of outliers and noise. The advanced GMDH uses different weightings according to their importance which is specified by the fuzzy membership grade. The developed model was verified using SBO accident simulation data for the OPR1000 nuclear power plant acquired with MARS code. Also, the advanced GMDH model was trained using the simulated development data and verified with simulated test data. The development and test data sets were independent. The simulation results show that the performance of the developed advanced GMDH model was very satisfactory, as shown in Table 1. Therefore, if the developed model can be optimized using diverse and specific data, it will be possible to predict the performance of Aux. feed water pump accurately.

Group unified accident reporting database (GUARD)

International Nuclear Information System (INIS)

Koene, W.; Waterfall, K.W.

1991-01-01

Significant advances have been made in recent years in enhancing the standard of safety within Shell Companies, such that safety has now been raised to a status equal to other primary business objectives. It is widely accepted that accident prevention is part of good business practice, and that a safe operation is normally an efficient operation. Safety programmes are being widely implemented which involve all employees from top management right down to the workforce including the contract staff, and the benefits are being realized. The effectiveness of any safety programme, however, must be continuously monitored, and in this respect injury and accident statistics play an important role as a prime indicator of safety performance. Statistics form part of the safety management process indicating the success of the safety programmes being implemented, and highlighting areas of weakness. Statistical information relating to the number and frequency of accidents, significant as it is, tells us little about how the accidents occur, or about how to improve the intrinsic safety of the operations. More detailed information on accident causes and lessons derived from the investigation of non-injurious accidents and near-misses is required for this, and for the setting of appropriate remedial actions. This paper concentrates on the feedback from accidents which have already occurred. This feedback plays a vital role as an indicator of safety performance upon which to judge the effectiveness of safety programmes, and also to provide important information relating to the immediate and underlying causes of accidents. To meet these requirements, however, a system for recording analyzing and communicating safety data is essential

Investigations of postulated accident sequences for the Fort St. Vrain HTGR

International Nuclear Information System (INIS)

Ball, S.J.; Cleveland, J.C.; Conklin, J.C.; Hatta, M.; Sanders, J.P.

1978-01-01

The systems analysis capability of the ORNL HTGR Safety analysis research program includes a family of computer codes: an overall plant NSSS simulation (ORTAP), and detailed component codes for investigating core neutronic accidents (CORTAP), shutdown emergency-cooling accidents via a 3-dimensional core model (ORECA), and once-through steam generator transients (BLAST). The component codes can either be run independently or in the overall NSSS code. Verification efforts have consisted primarily of using existing Fort St. Vrain reactor dynamics data to compare against code predictions. Comparisons of core thermal conditions made for reactor scrams from power levels between 30 and 50% showed good agreement. An optimization program was used to rationalize the difference between the predicted and measured refueling region outlet temperatures, and, in general, excellent agreement was attained by adjustment of models and parameters within their uncertainty ranges. However, more work is required to establish a unique and valid set of models

Development of Integrated Evaluation System for Severe Accident Management

Energy Technology Data Exchange (ETDEWEB)

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

2007-06-15

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

Development of Integrated Evaluation System for Severe Accident Management

International Nuclear Information System (INIS)

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

2007-06-01

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

Severe accident considerations in Canadian nuclear power reactors

International Nuclear Information System (INIS)

Omar, A.M.; Measures, M.P.; Scott, C.K.; Lewis, M.J.

1990-08-01

This paper describes a current study on severe accidents being sponsored by the Atomic Energy Control Board (AECB) and provides background on other related Canadian work. Scoping calculations are performed in Phase I of the AECB study to establish the relative consequences of several permutations resulting from six postulated initiating events, nine containment states, and a selection of meteorological conditions and health effects mitigating criteria. In Phase II of the study, selected accidents sequences would be analyzed in detail using models suitable for the design features of the Canadian nuclear power reactors

Analysis of severe core damage accident progression for the heavy water reactor

International Nuclear Information System (INIS)

Tong Lili; Yuan Kai; Yuan Jingtian; Cao Xuewu

2010-01-01

In this study, the severe accident progression analysis of generic Canadian deuterium uranium reactor 6 was preliminarily provided using an integrated severe accident analysis code. The selected accident sequences were multiple steam generator tube rupture and large break loss-of-coolant accidents because these led to severe core damage with an assumed unavailability for several critical safety systems. The progressions of severe accident included a set of failed safety systems normally operated at full power, and initiative events led to primary heat transport system inventory blow-down or boil off. The core heat-up and melting, steam generator response,fuel channel and calandria vessel failure were analyzed. The results showed that the progression of a severe core damage accident induced by steam generator tube rupture or large break loss-of-coolant accidents in a CANDU reactor was slow due to heat sinks in the calandria vessel and vault. (authors)

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

Road work zone accident studies : Advanced Research On Road Work Zone Safety Standard in Europe ARROWS Task 2.2 internal report. On behalf of the European Union, Directorate-General for Transport DG VII-E3, Transport RTD Programme.

NARCIS (Netherlands)

Gundy, C.M.

1998-01-01

The primary objective of this study, part of the ARROWS (Advanced Research on Road Work Zone Safety Standards in Europe) project, is to draw conclusions about the nature and extent of work zone traffic accidents. To that end, existing empirical studies concerning work zone traffic accidents have

The analysis of pressurizer safety valve stuck open accident for low power and shutdown PSA

Energy Technology Data Exchange (ETDEWEB)

Lim, Ho Gon; Park, Jin Hee; Jang, Seong Chul; Kim, Tae Woon

2005-01-01

The PSV (Pressurizer Safety Valve) popping test carried out practically in the early phase of a refueling outage has a little possibility of triggering a test-induced LOCA due to a PSV not fully closed or stuck open. According to a KSNP (Korea Standard Nuclear Power Plant) low power and shutdown PSA (Probabilistic Safety Assessment), the failure of a HPSI (High Pressure Safety Injection) following a PSV stuck open was identified as a dominant accident sequence with a significant contribution to low power and shutdown risks. In this study, we aim to investigate the consequences of the NPP for the various accident sequences following the PSV stuck open as an initiating event through the thermal-hydraulic system code calculations. Also, we search the accident mitigation method for the sequence of HPSI failure, then, the applicability of the method is verified by the simulations using T/H system code.

A new NEA expert group on accident-tolerant fuels

International Nuclear Information System (INIS)

Massara, Simone

2014-01-01

After the events at the Fukushima Daiichi nuclear power plant in March 2011, enhancing the accident tolerance of light water reactors (LWRs) became a topic of serious discussion. One outcome of those discussions has been to promote research into the development of advanced fuels and more robust reactor system technologies with improved performance, reliability and safety characteristics during normal operations and under accident conditions. The Fukushima Daiichi accident has highlighted in particular the importance of reducing hydrogen production rates and increasing fission product retention during extended loss of cooling accidents. In this context, the NEA organised two international workshops to share information and discuss technical and safety issues associated with the development of accident-tolerant fuels (ATFs) for LWRs. Presentations were given by experts from various organisations, industry and regulatory bodies of NEA member countries, as well as from representatives of international bodies. The presentations focused on lessons learnt from the Fukushima Daiichi accident, the desired characteristics of ATFs, potential design options and candidate materials, as well as the current state of the art in related modelling and simulation methods. During discussions following these workshop presentations, delegates agreed to establish a collaborative framework on ATFs within the NEA. Reporting to the Nuclear Science Committee, the Expert Group on Accident-tolerant Fuels for Light Water Reactors (EGATFL) will define and carry out a programme of work to help advance the scientific knowledge needed to provide the technical underpinning for the development of advanced LWR fuels with more enhanced accident tolerance compared to currently used zircaloy/UO 2 fuels. The group will foster information exchange on material properties and relevant phenomenological experiments, carry out state-of-the-art reviews, organise benchmark studies and foster international

Accident management for severe accidents

International Nuclear Information System (INIS)

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

1988-01-01

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

The Fukushima accident

International Nuclear Information System (INIS)

Maqua, M.; Stueck, R.

2012-01-01

On 11 March 2011, the Tohoku earthquake and the subsequent tsunami hit the Japanese east coast, causing more than 15,000 fatalities. To this date, 3,000 people are still missing. The Fukushima Dai-ichi NPP was the nuclear installation that was most affected by the tsunami. The earthquake cut off the NPP from the national grid. About 45 minutes later, the tsunami flooded units 1-4 and led to core meltdown events with large releases for units 1, 2 and 3. Unit 4 had been in refuelling outage at that time and lost the cooling of the spent fuel pool for several days. Considerable hydrogen explosions occurred in units 1, 3 and 4. Shortly after the accident, TEPCO started to mitigate the consequences of the accident by providing external cooling to the reactors and by removing the radioactive debris from the site. Great emphasis was laid on effective radiation protection measures for the clean-up workers. Thus, up to now there has been no fatality due to the radiation caused by the Fukushima accident. The main steps of the accident sequences are described, taking into account the latest findings of investigations performed by TEPCO or on behalf of the regulatory body. The presentation focuses on the description of the status of the Fukushima Dai-ichi nuclear power plant and the future steps for cleaning-up the site. In the presentation, the major phases of the roadmap that TEPCO has developed for the clean-up are highlighted. The risks associated with the current plant status and the clean-up phases are described. Abstract the content of the manuscript in a few lines.

Interface requirements to couple thermal hydraulics codes to severe accident codes: ICARE/CATHARE

Energy Technology Data Exchange (ETDEWEB)

Camous, F.; Jacq, F.; Chatelard, P. [IPSN/DRS/SEMAR CE-Cadarache, St Paul Lez Durance (France)] [and others

1997-07-01

In order to describe with the same code the whole sequence of severe LWR accidents, up to the vessel failure, the Institute of Protection and Nuclear Safety has performed a coupling of the severe accident code ICARE2 to the thermalhydraulics code CATHARE2. The resulting code, ICARE/CATHARE, is designed to be as pertinent as possible in all the phases of the accident. This paper is mainly devoted to the description of the ICARE2-CATHARE2 coupling.

Application of advanced statistical methods in assessment of the late phase of a nuclear accident

International Nuclear Information System (INIS)

Hofman, R.

2008-01-01

The paper presents a new methodology for improving of estimates of radiological situation on terrain in the late phase of a nuclear accident. Methods of Bayesian filtering are applied to the problem. The estimates are based on combination of modeled and measured data provided by responsible authorities. Exploiting information on uncertainty of both the data sources, we are able to produce improved estimate of the true situation on terrain. We also attempt to account for model error, which is unknown and plays crucial role in accuracy of the estimates. The main contribution of this paper is application of an approach based on advanced statistical methods, which allows for estimating of model error covariance structure upon measurements. Model error is estimated on basis of measured-minus-observed residuals evaluated upon measured and modeled values. The methodology is demonstrated on a sample scenario with simulated measurements. (authors)

Application of advanced statistical methods in assessment of the late phase of a nuclear accident

International Nuclear Information System (INIS)

Hofman, R.

2009-01-01

The paper presents a new methodology for improving of estimates of radiological situation on terrain in the late phase of a nuclear accident. Methods of Bayesian filtering are applied to the problem. The estimates are based on combination of modeled and measured data provided by responsible authorities. Exploiting information on uncertainty of both the data sources, we are able to produce improved estimate of the true situation on terrain. We also attempt to account for model error, which is unknown and plays crucial role in accuracy of the estimates. The main contribution of this paper is application of an approach based on advanced statistical methods, which allows for estimating of model error covariance structure upon measurements. Model error is estimated on basis of measured-minus-observed residuals evaluated upon measured and modeled values. The methodology is demonstrated on a sample scenario with simulated measurements. (authors)

Instrumentation availability for a pressurized water reactor with a large dry containment during severe accidents

International Nuclear Information System (INIS)

Arcieri, W.C.; Hanson, D.J.

1991-03-01

In support of the US Nuclear Regulatory Commission (NRC) Accident Management Research Program, the availability of instruments to supply accident management information during a broad range of severe accidents is evaluated for a pressurized water reactor with a large dry containment. Results from this evaluation include the following: (a) identification of plant conditions that would impact instrument performance and information needs during severe accidents, (b) definition of envelopes of parameters that would be important in assessing the performance of plant instrumentation for a broad range of severe accident sequences, and (c) assessment of the availability of plant instrumentation during severe accidents. 16 refs., 3 figs., 4 tabs

Severe accident research activities at Helmholtz-Zentrum Dresden-Rossendorf (HZDR)

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

In the frame of the nuclear safety research program of the Helmholtz Association HZDR performs fundamental and applied research to assess and to reduce the risks related to the nuclear fuel cycle and the production of electricity in nuclear power plants. One of the research topics focuses on the safety aspects of current and future reactor designs. This includes the development and application of methods for analyses of transients and postulated accidents, covering the whole spectrum from normal operation till severe accident sequences including core degradation. This paper gives an overview of the severe accident research activities at the Reactor Safety Division at the Institute of Resource Ecology.

LWR and HTGR coolant dynamics: the containment of severe accidents

International Nuclear Information System (INIS)

Theofanous, T.G.; Gherson, P.; Nourbakhsh, H.P.; Hu, K.; Iyer, K.; Viskanta, R.; Lommers, L.

1983-07-01

This is the final report of a project containing three major tasks. Task I deals with the fundamental aspects of energetic fuel/coolant interactions (steam explosions) as they pertain to LWR core melt accidents. Task II deals with the applied aspects of LWR core melt accident sequences and mechanisms important to containment response, and includes consideration of energetic fuel/coolant interaction events, as well as non-explosive ones, corium material disposition and eventual coolability, and containment pressurization phenomena. Finally, Task III is concerned with HTGR loss of forced circulation accidents. This report is organized into three major parts corresponding to these three tasks respectively

Proceedings of the international topical meeting on advanced reactors safety: Volume 2

International Nuclear Information System (INIS)

1997-01-01

In this volume, 89 papers are grouped under the following headings: advances in research/test reactor safety; advanced reactor accident management and emergency actions; advanced reactors instrumentation/controls/human factors; probabilistic risk/safety and reliability assessments; steam explosion research and issues; advanced reactor severe accident issues and research (analysis and assessments); advanced reactor thermal hydraulics; accelerator-driven source safety; liquid-metal reactor safety; structural assessments and issues; late papers

Analyses of containment loading by hydrogen burning during hypothetical core meltdown accidents

International Nuclear Information System (INIS)

Bracht, K.; Tiltmann, M.

1983-01-01

The possibility of occurance of violent hydrogen burning during a LWR meltdown accident and its consequences to containment atmosphere conditions are discussed. Two accident sequences with low and high system pressure during the in-vessel-melt phase of a meltdown accident are considered. In both sequences only deflagration, but no detonation may become possible, presuming homogeneity of the containment atmospheres. In a low pressure szenario the pressure increase due to deflagration will not reach the failure pressure of the containment, if combustion takes place when the flammability limit is reached. For the special situation of a rapid release of steam and hydrogen after a high-pressure failure of a reactor pressure vessel, calculations with a multicompartment code show that the possibility for hydrogen burning does not exist. Thus, an additional augmentation of the steam spike as a consequence of the failure of the pressure vessel cannot occur. (orig.)

Consequences of potential accidents in heavy water plants

International Nuclear Information System (INIS)

Croitoru, C.; Lazar, R.E.; Preda, I.A.; Dumitrescu, M.

2002-01-01

Heavy water plants achieve the primary isotopic concentration by H 2 O-H 2 S chemical exchange. In these plants are stored large quantities of hydrogen sulphide (high toxic, corrosive, flammable and explosive) maintained in process at relative high temperatures and pressures. It is required an assessment of risks associated with the potential accidents. The paper presents adopted model for quantitative consequences assessment in heavy water plants. Following five basic steps are used to identify the risks involved in plants operation: hazard identification, accident sequences development, H 2 S emissions calculus, dispersion analyses and consequences determination. A brief description of each step and some information from risk assessment for our heavy water pilot plant are provided. Accident magnitude, atmospheric conditions and population density in studied area were accounted for consequences calculus. (author)

Post-accident monitoring systems in Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

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

2018-01-01

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

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

Potential for containment leak paths through electrical penetration assemblies under severe accident conditions

International Nuclear Information System (INIS)

Sebrell, W.

1983-07-01

The leakage behavior of containments beyond design conditions and knowledge of failure modes is required for evaluation of mitigation strategies for severe accidents, risk studies, emergency preparedness planning, and siting. These studies are directed towards assessing the risk and consequences of severe accidents. An accident sequence analysis conducted on a Boiling Water Reactor (BWR), Mark I (MK I), indicated very high temperatures in the dry-well region, which is the location of the majority of electrical penetration assemblies. Because of the high temperatures, it was postulated in the ORNL study that the sealants would fail and all the electrical penetration assemblies would leak before structural failure would occur. Since other containments had similar electrical penetration assemblies, it was concluded that all containments would experience the same type of failure. The results of this study, however, show that this conclusion does not hold for PWRs because in the worst accident sequence, the long time containment gases stabilize to 350 0 F. BWRs, on the other hand, do experience high dry-well temperatures and have a higher potential for leakage

Studies of severe accidents in light-water reactors

International Nuclear Information System (INIS)

1987-01-01

From 10 to 12 November 1986 some 80 delegates met under the auspices of the CEC working group on the safety of light-water reactors. The participants from EC Member States were joined by colleagues from Sweden, Finland and the USA and met to discuss the subject of severe accidents in LWRs. Although this seminar had been planned well before Chernobyl, the ''severe-accident-that-really-happened'' made its mark on the seminar. The four main seminar topics were: (i) high source-term accident sequences identified in PSAs, (ii) containment performance, (iii) mitigation of core melt consequences, (iv) severe accident management in LWRs. In addition to the final panel discussion there was also a separate panel discussion on lessons learned from the Chernobyl accident. These proceedings include the papers presented during the seminar and they are arranged following the seminar programme outline. The presentations and discussions of the two panels are not included in the proceedings. The general conclusions and directions following from these two panels were, however, considered in a seminar review paper which was published in the March 1987 issue of Nuclear Engineering International

A framework for assessing severe accident management strategies

International Nuclear Information System (INIS)

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

1991-01-01

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

The yellow cake accident at the Ezeiza Airport

International Nuclear Information System (INIS)

Rodriguez, C.E.; Puntarulo, L.J.; Canibano, J.A.

1989-01-01

In January 1987 several drums containing yellow cake fell from about six meters during the loading operation of a Boeing 747 T-100 cargo aircraft. As a result of the accident, about 50% of the 38 drums involved lost their lids and a fraction of the radioactive content was released on an area of about 200 meters squared. Small amounts of yellow cake were dispersed down wind until about 100 meters from the accident place. The shipment was prepared for transport in standard 200 liter steel drums fulfilling the applicable Transport Regulations and the accident was the consequence of an erroneous operation during the cargo associated with a mechanical failure of the cargo lift. In order to avoid human contamination, immediate action was taken by the airport emergency team and in the meantime, the specialized groups of the National Atomic Energy Commission and the Federal Fire Brigades, were convened to take care of the decontamination and radiological evaluation problems. This paper describes the accidental sequences, the accident scenery, the countermeasures taken, the recovery and decontamination actions, and finally, as a conclusion, a brief description of the toxic and radiological aspects of the accident's mode

The effectiveness of using pictures in teaching young children about burn injury accidents.

Science.gov (United States)

Liu, Hsueh-Fen; Lin, Fang-Suey; Chang, Chien-Ju

2015-11-01

This study utilized the "story grammar" approach (Stein and Glenn, 1979) to analyze the within-corpus differences in recounting of sixty 6- and 7-year-old children, specifically whether illustrations (5-factor accident sequence) were or were not resorted to as a means to assist their narration of a home accident in which a child received a burn injury from hot soup. Our investigation revealed that the message presentation strategy "combining oral and pictures" better helped young children to memorize the story content (sequence of events leading to the burn injury) than "oral only." Specifically, the content of "the dangerous objects that caused the injury", "the unsafe actions that people involved took", and "how the people involved felt about the severity of the accident" differed significantly between the two groups. Copyright © 2015 Elsevier Ltd and The Ergonomics Society. All rights reserved.

Comparative analysis of station blackout accident progression in typical PWR, BWR, and PHWR

International Nuclear Information System (INIS)

Park, Soo Young; Ahn, Kwang Il

2012-01-01

Since the crisis at the Fukushima plants, severe accident progression during a station blackout accident in nuclear power plants is recognized as a very important area for accident management and emergency planning. The purpose of this study is to investigate the comparative characteristics of anticipated severe accident progression among the three typical types of nuclear reactors. A station blackout scenario, where all off-site power is lost and the diesel generators fail, is simulated as an initiating event of a severe accident sequence. In this study a comparative analysis was performed for typical pressurized water reactor (PWR), boiling water reactor (BWR), and pressurized heavy water reactor (PHWR). The study includes the summarization of design differences that would impact severe accident progressions, thermal hydraulic/severe accident phenomenological analysis during a station blackout initiated-severe accident; and an investigation of the core damage process, both within the reactor vessel before it fails and in the containment afterwards, and the resultant impact on the containment.

Accident analysis of HANARO fuel test loop

Energy Technology Data Exchange (ETDEWEB)

Kim, J. Y.; Chi, D. Y

1998-03-01

Steady state fuel test loop will be equipped in HANARO to obtain the development and betterment of advanced fuel and materials through the irradiation tests. The HANARO fuel test loop was designed to match the CANDU and PWR fuel operating conditions. The accident analysis was performed by RELAP5/MOD3 code based on FTL system designs and determined the detail engineering specification of in-pile test section and out-pile systems. The accident analysis results of FTL system could be used for the fuel and materials designer to plan the irradiation testing programs. (author). 23 refs., 20 tabs., 178 figs.

Study of fission products (Cs, Ba, Mo, Ru) behaviour in irradiated and simulated nuclear fuels during severe accidents using X-ray absorption Spectroscopy, SIMS and EPMA

International Nuclear Information System (INIS)

Geiger, Ernesto

2016-01-01

The identification of Fission Products (FP) release mechanism from irradiated nuclear fuels during a severe accident is of main importance for the development of codes for the estimation of the source-term (nature and quantity of radionuclides released into the environment). among the many FP Ba, Cs, Mo and Ru present a particular interest, since they may interact with each other or other elements and thus affect their release. In the framework of this thesis, two work axes have been set up in order to identify, firstly, the chemical phases initially present before the accident and, secondly, their evolution during the accident itself. The experimental approach consisted in reproducing nuclear severe accidents conditions at laboratory scale using both irradiated fuels and model materials (natural UO_2 doped with 12 FP). The advantage of these latter is the possibility of using characterization methods such as X-ray absorption Spectroscopy which are not available for irradiated fuels. Three irradiated fuel samples have been studied, representative to an initial state (before the accident), to an intermediate stage (1773 K) and to an advanced stage (2873 K) of a nuclear severe accident. Regarding to model materials, many accident sequences have been carried out, from 573 to 1973 K. Experimental results have allowed to establish a new release mechanism, considering both reducing and oxidizing conditions during an accident. These results have also demonstrated the importance of model materials as a complement to irradiated nuclear fuels in the study of nuclear severe accidents. (author) [fr

OVERVIEW OF MODULAR HTGR SAFETY CHARACTERIZATION AND POSTULATED ACCIDENT BEHAVIOR LICENSING STRATEGY

Energy Technology Data Exchange (ETDEWEB)

Ball, Sydney J [ORNL

2014-06-01

This report provides an update on modular high-temperature gas-cooled reactor (HTGR) accident analyses and risk assessments. One objective of this report is to improve the characterization of the safety case to better meet current regulatory practice, which is commonly geared to address features of today s light water reactors (LWRs). The approach makes use of surrogates for accident prevention and mitigation to make comparisons with LWRs. The safety related design features of modular HTGRs are described, along with the means for rigorously characterizing accident selection and progression methodologies. Approaches commonly used in the United States and elsewhere are described, along with detailed descriptions and comments on design basis (and beyond) postulated accident sequences.

Traffic Accidents Involving Cyclists Identifying Causal Factors Using Questionnaire Survey, Traffic Accident Data, and Real-World Observation.

Science.gov (United States)

Oikawa, Shoko; Hirose, Toshiya; Aomura, Shigeru; Matsui, Yasuhiro

2016-11-01

The purpose of this study is to clarify the mechanism of traffic accidents involving cyclists. The focus is on the characteristics of cyclist accidents and scenarios, because the number of traffic accidents involving cyclists in Tokyo is the highest in Japan. First, dangerous situations in traffic incidents were investigated by collecting data from 304 cyclists in one city in Tokyo using a questionnaire survey. The survey indicated that cyclists used their bicycles generally while commuting to work or school in the morning. Second, the study investigated the characteristics of 250 accident situations involving cyclists that happened in the city using real-world bicycle accident data. The results revealed that the traffic accidents occurred at intersections of local streets, where cyclists collided most often with vehicles during commute time in the morning. Third, cyclists' behavior was observed at a local street intersection in the morning in the city using video pictures. In one hour during the morning commute period, 250 bicycles passed through the intersection. The results indicated that one of the reasons for traffic accidents involving cyclists might be the combined effect of low visibility, caused by the presence of box-like building structures close to the intersections, and the cyclists' behavior in terms of their velocity and no confirming safety. It was observed that, on average, bicycle velocity was 3.1 m/s at the initial line of an intersection. The findings from this study could be useful in developing new technologies to improve cyclist safety, such as alert devices for cyclists and vehicle drivers, wireless communication systems between cyclists and vehicle drivers, or advanced vehicles with bicycle detection and collision mitigation systems.

Development on quantitative safety analysis method of accident scenario. The automatic scenario generator development for event sequence construction of accident

International Nuclear Information System (INIS)

Kojima, Shigeo; Onoue, Akira; Kawai, Katsunori

1998-01-01

This study intends to develop a more sophisticated tool that will advance the current event tree method used in all PSA, and to focus on non-catastrophic events, specifically a non-core melt sequence scenario not included in an ordinary PSA. In the non-catastrophic event PSA, it is necessary to consider various end states and failure combinations for the purpose of multiple scenario construction. Therefore it is anticipated that an analysis work should be reduced and automated method and tool is required. A scenario generator that can automatically handle scenario construction logic and generate the enormous size of sequences logically identified by state-of-the-art methodology was developed. To fulfill the scenario generation as a technical tool, a simulation model associated with AI technique and graphical interface, was introduced. The AI simulation model in this study was verified for the feasibility of its capability to evaluate actual systems. In this feasibility study, a spurious SI signal was selected to test the model's applicability. As a result, the basic capability of the scenario generator could be demonstrated and important scenarios were generated. The human interface with a system and its operation, as well as time dependent factors and their quantification in scenario modeling, was added utilizing human scenario generator concept. Then the feasibility of an improved scenario generator was tested for actual use. Automatic scenario generation with a certain level of credibility, was achieved by this study. (author)

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

Comparison of Severe Accident Results Among SCDAP/RELAP5, MAAP, and MELCOR Codes

International Nuclear Information System (INIS)

Wang, T.-C.; Wang, S.-J.; Teng, J.-T.

2005-01-01

This paper demonstrates a large-break loss-of-coolant accident (LOCA) sequence of the Kuosheng nuclear power plant (NPP) and station blackout sequence of the Maanshan NPP with the SCDAP/RELAP5 (SR5), Modular Accident Analysis Program (MAAP), and MELCOR codes. The large-break sequence initiated with double-ended rupture of a recirculation loop. The main steam isolation valves (MSIVs) closed, the feedwater pump tripped, the reactor scrammed, and the assumed high-pressure and low-pressure spray systems of the emergency core cooling system (ECCS) were not functional. Therefore, all coolant systems to quench the core were lost. MAAP predicts a longer vessel failure time, and MELCOR predicts a shorter vessel failure time for the large-break LOCA sequence. The station blackout sequence initiated with a loss of all alternating-current (ac) power. The MSIVs closed, the feedwater pump tripped, and the reactor scrammed. The motor-driven auxiliary feedwater system and the high-pressure and low-pressure injection systems of the ECCS were lost because of the loss of all ac power. It was also assumed that the turbine-driven auxiliary feedwater pump was not functional. Therefore, the coolant system to quench the core was also lost. MAAP predicts a longer time of steam generator dryout, time interval between top of active fuel and bottom of active fuel, and vessel failure time than those of the SR5 and MELCOR predictions for the station blackout sequence. The three codes give similar results for important phenomena during the accidents, including SG dryout, core uncovery, cladding oxidation, cladding failure, molten pool formulation, debris relocation to the lower plenum, and vessel head failure. This paper successfully demonstrates the large-break LOCA sequence of the Kuosheng NPP and the station blackout sequence of the Maanshan NPP

System 80+ design features for severe accident prevention and mitigation

International Nuclear Information System (INIS)

Jacob, M.C.; Schneider, R.E.; Finnicum, D.J.

1993-01-01

ABB-CE, in cooperation with the US Department of Energy, is working to develop and certify the System 80+ design, which is ABB-CE's standardized evolutionary Advanced Light Water Reactor (ALWR) design. It incorporates design enhancements based on Probabilistic Risk Assessment (PRA) insights, guidance from the EPRI's Utility Requirements Document, and US NRC's Severe Accident Policy. Major severe accident prevention and mitigation design features of the system is discussed along with its conformance to EPRI URD guidance, as applicable. Computer simulation of a best estimate severe accident scenario is presented to illustrate the acceptable containment performance of the design. It is concluded that by considering severe accident prevention and mitigation early in the design process, the System 80+ design represents a robust plant design that has low core damage frequencies, low containment conditional failure probabilities, and acceptable deterministic containment performance under severe accident conditions

Severe accident testing of electrical penetration assemblies

International Nuclear Information System (INIS)

Clauss, D.B.

1989-11-01

This report describes the results of tests conducted on three different designs of full-size electrical penetration assemblies (EPAs) that are used in the containment buildings of nuclear power plants. The objective of the tests was to evaluate the behavior of the EPAs under simulated severe accident conditions using steam at elevated temperature and pressure. Leakage, temperature, and cable insulation resistance were monitored throughout the tests. Nuclear-qualified EPAs were produced from D. G. O'Brien, Westinghouse, and Conax. Severe-accident-sequence analysis was used to generate the severe accident conditions (SAC) for a large dry pressurized-water reactor (PWR), a boiling-water reactor (BWR) Mark I drywell, and a BWR Mark III wetwell. Based on a survey conducted by Sandia, each EPA was matched with the severe accident conditions for a specific reactor type. This included the type of containment that a particular EPA design was used in most frequently. Thus, the D. G. O'Brien EPA was chosen for the PWR SAC test, the Westinghouse was chosen for the Mark III test, and the Conax was chosen for the Mark I test. The EPAs were radiation and thermal aged to simulate the effects of a 40-year service life and loss-of-coolant accident (LOCA) before the SAC tests were conducted. The design, test preparations, conduct of the severe accident test, experimental results, posttest observations, and conclusions about the integrity and electrical performance of each EPA tested in this program are described in this report. In general, the leak integrity of the EPAs tested in this program was not compromised by severe accident loads. However, there was significant degradation in the insulation resistance of the cables, which could affect the electrical performance of equipment and devices inside containment at some point during the progression of a severe accident. 10 refs., 165 figs., 16 tabs

TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis

International Nuclear Information System (INIS)

Liles, D.R.; Mahaffy, J.H.

1984-02-01

The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field; flow-regime-dependent constitutive equation treatment; optional reflood tracking capability for both bottom flood and falling-film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. This report describes the thermal-hydraulic models and the numerical solution methods used in the code. Detailed programming and user information also are provided

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

Safety and risk questions following the nuclear incidents and accidents in Japan. Summary final report

International Nuclear Information System (INIS)

Mildenberger, Oliver

2015-03-01

After the nuclear accidents in Japan, GRS has carried out in-depth investigations of the events. On the one hand, the accident sequences in the affected units have been analysed from various viewpoints. On the other hand, the transferability of the findings to German plants has been examined to possibly make recommendations for safety improvements. The accident sequences at Fukushima Daiichi have been traced with as much detail as possible based on all available information. Additional insights have been drawn from thermohydraulic analyses with the GRS code system ATHLET-CD/COCOSYS focusing on the events in units 2 and 3, e.g. with regard to core damage and the state of the containments in the first days of the accident sequence. In-depth investigations have also been carried out on topics such as natural external hazards, electrical power supply or organizational measures. In addition, methodological studies on further topics related with the accidents have been performed. Through a detailed analysis of the relevant data from the events in Japan, the basis for an in-depth examination of the transferability to German plants was created. It was found that an implementation of most of the insights gained from the investigations had already been initiated as part of the GRS information notice 2012/02. Further findings have been communicated to the federal government and introduced into other relevant bodies, e.g. the Nuclear Safety Standards Committee (KTA) or the Reactor Safety Commission (RSK).

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

International Nuclear Information System (INIS)

Asmolov, V.G.

1997-01-01

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

Severe Accident Mitigation by using Core Catcher applicable for Korea standard nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Park, Hae Kyun; Kim, Sang Nyung [Kyung Hee Univ., Yongin (Korea, Republic of)

2013-10-15

Nuclear power plants have been designed and operated in order to prevent severe accident because of their risk that contains tremendous radioactive materials that are potentially hazardous. Moreover, the government requested the nuclear industry to implement a severe accident management strategy for existing reactors to mitigate the risk of potential severe accidents. However, Korea standard nuclear power plant(APR-1400 and OPR-1000) are much more vulnerable for severe accident management than that of developed countries. Due to the design feature of reactor cavity in Korea standard nuclear power plant, inequable and serious Molten Core-Concrete Interaction(MCCI) may cause considerable safety problem to the reactor containment liner. At worst, it brings the release of radioactive materials to the environment. This accident applies to the fourth level of defense in depth(IAEA 1996), 'severe accident'. This study proposes and designs the 'slope' to secure reactor containment liner integrity when the corium spreads out from the destroyed reactor vessel to the reactor cavity due to the core melting accident. For this, make the initial corium distribution evenly exploit the 'slope' on the basis of the study of Ex-vessel corium behavior to prevent inequable and serious MCCI, in order to mitigate severe accident. The viscosity has a dominant position in the calculation. According to the result, the spread out distance on the slope is 10.7146841m, considering the rough surface of the concrete(slope) and margin of reactor cavity end(under 11m). Easy to design, production and economic feasibility are the advantage of the designed slope in this study. However, the slope design may unsuitable when the sequences of the accidents did not satisfy the assumptions as mentioned. Despite of those disadvantages, the slope will show a great performance to mitigate the severe accident. As mentioned in assumption, the corium releasing time property was

Severe Accident Mitigation by using Core Catcher applicable for Korea standard nuclear power plant

International Nuclear Information System (INIS)

Park, Hae Kyun; Kim, Sang Nyung

2013-01-01

Nuclear power plants have been designed and operated in order to prevent severe accident because of their risk that contains tremendous radioactive materials that are potentially hazardous. Moreover, the government requested the nuclear industry to implement a severe accident management strategy for existing reactors to mitigate the risk of potential severe accidents. However, Korea standard nuclear power plant(APR-1400 and OPR-1000) are much more vulnerable for severe accident management than that of developed countries. Due to the design feature of reactor cavity in Korea standard nuclear power plant, inequable and serious Molten Core-Concrete Interaction(MCCI) may cause considerable safety problem to the reactor containment liner. At worst, it brings the release of radioactive materials to the environment. This accident applies to the fourth level of defense in depth(IAEA 1996), 'severe accident'. This study proposes and designs the 'slope' to secure reactor containment liner integrity when the corium spreads out from the destroyed reactor vessel to the reactor cavity due to the core melting accident. For this, make the initial corium distribution evenly exploit the 'slope' on the basis of the study of Ex-vessel corium behavior to prevent inequable and serious MCCI, in order to mitigate severe accident. The viscosity has a dominant position in the calculation. According to the result, the spread out distance on the slope is 10.7146841m, considering the rough surface of the concrete(slope) and margin of reactor cavity end(under 11m). Easy to design, production and economic feasibility are the advantage of the designed slope in this study. However, the slope design may unsuitable when the sequences of the accidents did not satisfy the assumptions as mentioned. Despite of those disadvantages, the slope will show a great performance to mitigate the severe accident. As mentioned in assumption, the corium releasing time property was conservatively calculated

Causal factors in accidents of high-speed craft and conventional ocean-going vessels

International Nuclear Information System (INIS)

Antao, Pedro; Guedes Soares, C.

2008-01-01

An analysis of 40 ocean-going commercial vessel accidents is compared with the study of a similar number of high-speed crafts (HSCs) accidents, using in both cases a methodology that highlights the sequence of events leading to the accident and identifies the associated latent or causal factors. The main objective of this study was to identify and understand the difference in the pattern of causal factors associated with HSC accidents, as compared with the more traditional ocean-going ships. From the analysis one can see that the HSC accidents are mainly related to bridge personnel and operations, where the human element is the key factor identified as being responsible for the majority of the accidents. When compared with ocean-going commercial vessels, it is clear that navigational equipment and procedures have a larger preponderance in terms of the occurrence of accidents of HSC and particular attention should be given to these issues

Hot Cell Installation and Demonstration of the Severe Accident Test Station

Energy Technology Data Exchange (ETDEWEB)

Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Burns, Zachary M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yan, Yong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-08-01

A Severe Accident Test Station (SATS) capable of examining the oxidation kinetics and accident response of irradiated fuel and cladding materials for design basis accident (DBA) and beyond design basis accident (BDBA) scenarios has been successfully installed and demonstrated in the Irradiated Fuels Examination Laboratory (IFEL), a hot cell facility at Oak Ridge National Laboratory. The two test station modules provide various temperature profiles, steam, and the thermal shock conditions necessary for integral loss of coolant accident (LOCA) testing, defueled oxidation quench testing and high temperature BDBA testing. The installation of the SATS system restores the domestic capability to examine postulated and extended LOCA conditions on spent fuel and cladding and provides a platform for evaluation of advanced fuel and accident tolerant fuel (ATF) cladding concepts. This document reports on the successful in-cell demonstration testing of unirradiated Zircaloy-4. It also contains descriptions of the integral test facility capabilities, installation activities, and out-of-cell benchmark testing to calibrate and optimize the system.

Method for consequence calculations for severe accidents

International Nuclear Information System (INIS)

Nielsen, F.

1988-07-01

This report was commissioned by the Swedish State Power Board. The report contains a calculation of radiation doses in the surroundings caused by a theoretical core meltdown accident at Forsmark reactor No 3. The accident sequence chosen for the calculating was a release caused by total power failure. The calculations were made by means of the PLUCON4 code. Meteorological data for two years from the Forsmark meteorological tower were analysed to find representative weather situations. As typical weather, Pasquill D was chosen with a wind speed of 5 m/s, and as extreme weather, Pasquill F with a wind speed of 2 m/s. 23 tabs., 37 ills., 20 refs. (author)

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

International Nuclear Information System (INIS)

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

1988-01-01

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

Station blackout transient at the Browns Ferry Unit 1 Plant: a severe accident sequence analysis (SASA) program study

International Nuclear Information System (INIS)

Schultz, R.R.

1982-01-01

Operating plant transients are of great interest for many reasons, not the least of which is the potential for a mild transient to degenerate to a severe transient yielding core damage. Using the Browns Ferry (BF) Unit-1 plant as a basis of study, the station blackout sequence was investigated by the Severe Accident Sequence Analysis (SASA) Program in support of the Nuclear Regulatory Commission's Unresolved Safety Issue A-44: Station Blackout. A station blackout transient occurs when the plant's AC power from a comemrcial power grid is lost and cannot be restored by the diesel generators. Under normal operating conditions, f a loss of offsite power (LOSP) occurs [i.e., a complete severance of the BF plants from the Tennessee Valley Authority (TVA) power grid], the eight diesel generators at the three BF units would quickly start and power the emergency AC buses. Of the eight diesel generators, only six are needed to safely shut down all three units. Examination of BF-specific data show that LOSP frequency is low at Unit 1. The station blackout frequency is even lower (5.7 x 10 - 4 events per year) and hinges on whether the diesel generators start. The frequency of diesel generator failure is dictated in large measure by the emergency equipment cooling water (EECW) system that cools the diesel generators

Psychophysiological and other factors affecting human performance in accident prevention and investigation

International Nuclear Information System (INIS)

Klinestiver, L.R.

1980-01-01

Psychophysiological factors are not uncommon terms in the aviation incident/accident investigation sequence where human error is involved. It is highly suspect that the same psychophysiological factors may also exist in the industrial arena where operator personnel function; but, there is little evidence in literature indicating how management and subordinates cope with these factors to prevent or reduce accidents. It is apparent that human factors psychophysological training is quite evident in the aviation industry. However, while the industrial arena appears to analyze psychophysiological factors in accident investigations, there is little evidence that established training programs exist for supervisors and operator personnel

Advanced neutron source reactor conceptual safety analysis report, three-element-core design: Chapter 15, accident analysis

International Nuclear Information System (INIS)

Chen, N.C.J.; Wendel, M.W.; Yoder, G.L.; Harrington, R.M.

1996-02-01

In order to utilize reduced enrichment fuel, the three-element-core design for the Advanced Neutron Source has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. To assess the impact of changes in the core region configuration and the thermal-hydraulic steady-state conditions, the safety analysis has been updated. This report gives the safety margins for the loss-of-off-site power and pressure-boundary fault accidents based on the RELAP5 results. AU margins are greater for the three-element-core simulations than those calculated for the two-element core

Small break loss of coolant accident analysis of advanced PWR plant designs utilizing DVI line venturis

International Nuclear Information System (INIS)

Kemper, Robert M.; Gagnon, Andre F.; McNamee, Kevin; Cheung, Augustine C.

1995-01-01

The Westinghouse Advanced Passive and evolutionary Pressurizer Water Reactors (i.e. AP600 and APWR) incorporate direct vessel injection (DVI) of emergency core coolant as a means of minimizing the potential spilling of emergency core cooling water during a loss of coolant accident (LOCA). As a result, the most limiting small break LOCA (SBLOCA) event for these designs, with respect core inventory makeup capability, is a postulated double ended rupture of one of the DVI lines. This paper presents the results of a design optimization study that examines the installation of a venturi in the DVI line as a means of limiting the reactor coolant lost from the reactor vessel. The comparison results demonstrate that by incorporating a properly sized venturi in the DVI line, core uncovery concerns as a result of a DVI line break can be eliminated for both the AP600 and APWR plants. (author)

Accident termination by element dropout in the GCFR

International Nuclear Information System (INIS)

Torri, A.; Tomkins, J.L.

1976-01-01

Severe loss-of-flow accidents are being investigated for the GCFR in order to assess the risk from those low-probability accidents which lead to a loss of coolable core geometry. Accident mitigating phenomena unique to the GCFR have been identified for the loss of decay heat removal accident. Circumferential assembly duct melting is calculated to occur at the core mid-plane before the fuel within the assembly melts. The GCFR core assemblies are top-mounted and there is clearance between assemblies to accommodate swelling and thermal distortions without interference. No lateral core clamping system is employed and there are no structures in the plenum below the core. Thus it is possible for the lower portion of the individual assemblies, including most of the fuel, to drop to the cavity floor unless interference or bonding between assemblies develops during the accident. Due to the delay in duct corner melting the melt front at the duct mid-flat progresses over about one-half of the core height. The possibility of inter-element bonding by molten duct steel dislocated into the gap between assemblies has been recognized and a test program to verify the duct melting sequence and to investigate the duct dropout is being planned at the Los Alamos Scientific Laboratory

Environmental Impact Assessment following a Nuclear Accident to a Candu NPP

International Nuclear Information System (INIS)

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

2009-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-07-01

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

Advance of Hazardous Operation Robot and its Application in Special Equipment Accident Rescue

Science.gov (United States)

Zeng, Qin-Da; Zhou, Wei; Zheng, Geng-Feng

A survey of hazardous operation robot is given out in this article. Firstly, the latest researches such as nuclear industry robot, fire-fighting robot and explosive-handling robot are shown. Secondly, existing key technologies and their shortcomings are summarized, including moving mechanism, control system, perceptive technology and power technology. Thirdly, the trend of hazardous operation robot is predicted according to current situation. Finally, characteristics and hazards of special equipment accident, as well as feasibility of hazardous operation robot in the area of special equipment accident rescue are analyzed.

Assessment of two BWR accident management strategies

International Nuclear Information System (INIS)

Hodge, S.A.; Petek, M.

1991-01-01

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

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.

Limitations of systemic accident analysis methods

Directory of Open Access Journals (Sweden)

Casandra Venera BALAN

2016-12-01

Full Text Available In terms of system theory, the description of complex accidents is not limited to the analysis of the sequence of events / individual conditions, but highlights nonlinear functional characteristics and frames human or technical performance in relation to normal functioning of the system, in safety conditions. Thus, the research of the system entities as a whole is no longer an abstraction of a concrete situation, but an exceeding of the theoretical limits set by analysis based on linear methods. Despite the issues outlined above, the hypothesis that there isn’t a complete method for accident analysis is supported by the nonlinearity of the considered function or restrictions, imposing a broad vision of the elements introduced in the analysis, so it can identify elements corresponding to nominal parameters or trigger factors.

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

22 CFR 102.11 - Arranging for the payment of expenses attendant upon an accident.

Science.gov (United States)

2010-04-01

... FUNCTIONS CIVIL AVIATION United States Aircraft Accidents Abroad § 102.11 Arranging for the payment of... authorizes in advance the expenditure of such funds on a reimbursable basis. In the absence of such advance...

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Development of advanced claddings for suppressing the hydrogen emission in accident conditions. Development of advanced claddings for suppressing the hydrogen emission in the accident condition

International Nuclear Information System (INIS)

Park, Jeong-Yong; KIM, Hyun-Gil; JUNG, Yang-Il; PARK, Dong-Jun; KOO, Yang-Hyun

2013-01-01

The development of accident-tolerant fuels can be a breakthrough to help solve the challenge facing nuclear fuels. One of the goals to be reached with accident-tolerant fuels is to reduce the hydrogen emission in the accident condition by improving the high-temperature oxidation resistance of claddings. KAERI launched a new project to develop the accident-tolerant fuel claddings with the primary objective to suppress the hydrogen emission even in severe accident conditions. Two concepts are now being considered as hydrogen-suppressed cladding. In concept 1, the surface modification technique was used to improve the oxidation resistance of Zr claddings. Like in concept 2, the metal-ceramic hybrid cladding which has a ceramic composite layer between the Zr inner layer and the outer surface coating is being developed. The high-temperature steam oxidation behaviour was investigated for several candidate materials for the surface modification of Zr claddings. From the oxidation tests carried out in 1 200 deg. C steam, it was found that the high-temperature steam oxidation resistance of Cr and Si was much higher than that of zircaloy-4. Al 3 Ti-based alloys also showed extremely low-oxidation rate compared to zircaloy-4. One important part in the surface modification is to develop the surface coating technology where the optimum process needs to be established depending on the surface layer materials. Several candidate materials were coated on the Zr alloy specimens by a laser beam scanning (LBS), a plasma spray (PS) and a PS followed by LBS and subject to the high-temperature steam oxidation test. It was found that Cr and Si coating layers were effective in protecting Zr-alloys from the oxidation. The corrosion behaviour of the candidate materials in normal reactor operation condition such as 360 deg. C water will be investigated after the screening test in the high-temperature steam. The metal-ceramic hybrid cladding consisted of three major parts; a Zr liner, a

TITAN: a computer program for accident occurrence frequency analyses by component Monte Carlo simulation

Energy Technology Data Exchange (ETDEWEB)

Nomura, Yasushi [Department of Fuel Cycle Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Tamaki, Hitoshi [Department of Safety Research Technical Support, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Kanai, Shigeru [Fuji Research Institute Corporation, Tokyo (Japan)

2000-04-01

In a plant system consisting of complex equipments and components for a reprocessing facility, there might be grace time between an initiating event and a resultant serious accident, allowing operating personnel to take remedial actions, thus, terminating the ongoing accident sequence. A component Monte Carlo simulation computer program TITAN has been developed to analyze such a complex reliability model including the grace time without any difficulty to obtain an accident occurrence frequency. Firstly, basic methods for the component Monte Carlo simulation is introduced to obtain an accident occurrence frequency, and then, the basic performance such as precision, convergence, and parallelization of calculation, is shown through calculation of a prototype accident sequence model. As an example to illustrate applicability to a real scale plant model, a red oil explosion in a German reprocessing plant model is simulated to show that TITAN can give an accident occurrence frequency with relatively good accuracy. Moreover, results of uncertainty analyses by TITAN are rendered to show another performance, and a proposal is made for introducing of a new input-data format to adapt the component Monte Carlo simulation. The present paper describes the calculational method, performance, applicability to a real scale, and new proposal for the TITAN code. In the Appendixes, a conventional analytical method is shown to avoid complex and laborious calculation to obtain a strict solution of accident occurrence frequency, compared with Monte Carlo method. The user's manual and the list/structure of program are also contained in the Appendixes to facilitate TITAN computer program usage. (author)

TITAN: a computer program for accident occurrence frequency analyses by component Monte Carlo simulation

International Nuclear Information System (INIS)

Nomura, Yasushi; Tamaki, Hitoshi; Kanai, Shigeru

2000-04-01

In a plant system consisting of complex equipments and components for a reprocessing facility, there might be grace time between an initiating event and a resultant serious accident, allowing operating personnel to take remedial actions, thus, terminating the ongoing accident sequence. A component Monte Carlo simulation computer program TITAN has been developed to analyze such a complex reliability model including the grace time without any difficulty to obtain an accident occurrence frequency. Firstly, basic methods for the component Monte Carlo simulation is introduced to obtain an accident occurrence frequency, and then, the basic performance such as precision, convergence, and parallelization of calculation, is shown through calculation of a prototype accident sequence model. As an example to illustrate applicability to a real scale plant model, a red oil explosion in a German reprocessing plant model is simulated to show that TITAN can give an accident occurrence frequency with relatively good accuracy. Moreover, results of uncertainty analyses by TITAN are rendered to show another performance, and a proposal is made for introducing of a new input-data format to adapt the component Monte Carlo simulation. The present paper describes the calculational method, performance, applicability to a real scale, and new proposal for the TITAN code. In the Appendixes, a conventional analytical method is shown to avoid complex and laborious calculation to obtain a strict solution of accident occurrence frequency, compared with Monte Carlo method. The user's manual and the list/structure of program are also contained in the Appendixes to facilitate TITAN computer program usage. (author)

A physical tool for severe accident mitigation studies

Energy Technology Data Exchange (ETDEWEB)

Marie, N., E-mail: nathalie.marie@cea.fr [CEA, DEN, DER, F-13108 Saint Paul Lez Durance (France); Bachrata, A. [CEA, DEN, DER, F-13108 Saint Paul Lez Durance (France); Seiler, J.M. [CEA, DEN, DTN, F-38054 Grenoble (France); Barjot, F. [EDF R& D, SINETICS, F-93141 Clamart (France); Marrel, A. [CEA, DEN, DER, F-13108 Saint Paul Lez Durance (France); Gossé, S. [CEA, DEN, DPC, F-91191 Gif Sur Yvette (France); Bertrand, F. [CEA, DEN, DER, F-13108 Saint Paul Lez Durance (France)

2016-12-01

Highlights: • Physical tool for mitigation studies devoted to SFR safety. • Physical models to describe the material discharge from core. • Comparison to SIMMER III results. • Studies for ASTRID safety assessment and support to core design. - Abstract: Within the framework of the Generation IV Sodium-cooled Fast Reactors (SFR) R&D program of CEA, the core behavior in case of severe accidents is being assessed. Such transients are usually simulated with mechanistic codes (such as SIMMER-III). As a complement to this code, which gives reference accidental transient, a physico-statistical approach is currently followed; its final objective being to derive the variability of the main results of interest for the safety. This approach involves a fast-running simulation of extended accident sequences coupling low-dimensional physical models to advanced statistical analysis techniques. In this context, this paper presents such a low-dimensional physical tool (models and simulation results) dedicated to molten core materials discharge. This 0D tool handles heat transfers from molten (possibly boiling) pools, fuel crust evolution, phase separation/mixing of fuel/steel pools, radial thermal erosion of mitigation tubes, discharge of core materials and associated axial thermal erosion of mitigation tubes. All modules are coupled with a global neutronic evolution model of the degraded core. This physical tool is used to study and to define mitigation features (function of tubes devoted to mitigation inside the core, impact of absorbers falling into the degraded core…) to avoid energetic core recriticality during a secondary phase of a potential severe accident. In the future, this physical tool, associated to statistical treatments of the effect of uncertainties would enable sensitivity analysis studies. This physical tool is described before presenting its comparison against SIMMER-III code results, including a space-and energy-dependent neutron transport kinetic

Containment loading during severe core damage accidents

International Nuclear Information System (INIS)

Fermandjian, J.; Evrard, J.M.; Cenerino, C.; Berthion, Y.; Carvallo, G.

1984-11-01

The objective of the article is to study the influence of the state of the reactor cavity (dry or flooded) and of the corium coolability on the thermal-hydraulics in the containment in the case of an accident sequence involving core melting and subsequent containment basemat erosion, in a 900 MWe PWR unit. Calculations are performed by using the JERICHO thermal hydraulics code

Reference accident (Core disruption accident - safety analysis detailed report no. 11)

Energy Technology Data Exchange (ETDEWEB)

1988-01-15

The PEC safety analysis led to the conclusion that all credible sequences (incident sequences characterized by a frequency of occurrence above 10/sup minus 7/ events per year) are limited to the design basis conditions of components of the plant protection systems, and that none of them leads to a release of mechanical energy or to an extensive damage of the core and primary containment structures event in the case of failure to scram. Nevertheless, as is done in other countries for similar reactors, some events beyond the limits of credibility were considered for the PEC reactor. These were defined on a absolutely hypothetical basis that involves severe core disruption and dynamic loading of primary containment boundary. A series of containments, each having a different role, was designed to mitigate the radiological effects of a postulated core disruptive accident. The final aim was to demonstrate that residual heat can be removed and that the release of radioactivity to the environment is within acceptable limits.

Examination of some assumed severe reactor accidents at the Olkiluoto nuclear power plant

International Nuclear Information System (INIS)

Pekkarinen, E.; Rossi, J.

1989-02-01

Knowledge and analysis methods of severe accidents at nuclear power plants and of subsequent response of primary system and containment have been developed in last few years to the extent that realistic source tems of the specified accident sequences can be calculated for the Finnish nuclear power plants. The objective of this investigation was to calculate the source terms of off-site consequences brought about by some selected severe accident sequences initiated by the total loss of on-site and off-site AC power at the Olkiluoto nuclear power plant. The results describing the estimated off-site health risks are expressed as conditional assuming that the accident has taken place, because the probabilities of the occurence of the accident sequences considered have not been analysed in this study. The range and probabilities of occurence of health detriments are considered by calculating consequences in different weeather conditions and taking into account the annual frequency of each weather condition and statistical population distribution. The calculational results indicate that the reactor building provides and additional holdup and deposition of radioactive substance (except coble gases) released from the containment. Furthermore, the release fractions of the core inventory to the environment of volatile fission products such as iodine, cesium and tellurium remain under 0.03. No early health effects are predicted for the surrounding population in case the assumed short-tem countermeasures are performed effectively. Acute health effects are extremely improbable even without any active countermeasure. By reducing the long-term exposure from contaminated agricultural products, the collective dose from natural long-term background radiation, for instance in the sector of 30 degrees towards the southern Finland up to the distance of 300 kilometers, would be expected to increase with 2-20 percent depending on the release considered

The use of influence diagrams for evaluating severe accident management strategies

International Nuclear Information System (INIS)

Jae, M.; Apostolakis, G.E.

1992-01-01

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

A risk-based evaluation of the impact of key uncertainties on the prediction of severe accident source terms - STU

International Nuclear Information System (INIS)

Ang, M.L.; Grindon, E.; Dutton, L.M.C.; Garcia-Sedano, P.; Santamaria, C.S.; Centner, B.; Auglaire, M.; Routamo, T.; Outa, S.; Jokiniemi, J.; Gustavsson, V.; Wennerstrom, H.; Spanier, L.; Gren, M.; Boschiero, M-H; Droulas, J-L; Friederichs, H-G; Sonnenkalb, M.

2001-01-01

The purpose of this project is to address the key uncertainties associated with a number of fission product release and transport phenomena in a wider context and to assess their relevance to key severe accident sequences. This project is a wide-based analysis involving eight reactor designs that are representative of the reactors currently operating in the European Union (EU). In total, 20 accident sequences covering a wide range of conditions have been chosen to provide the basis for sensitivity studies. The appraisal is achieved through a systematic risk-based framework developed within this project. Specifically, this is a quantitative interpretation of the sensitivity calculations on the basis of 'significance indicators', applied above defined threshold values. These threshold values represent a good surrogate for 'large release', which is defined in a number of EU countries. In addition, the results are placed in the context of in-containment source term limits, for advanced light water reactor designs, as defined by international guidelines. Overall, despite the phenomenological uncertainties, the predicted source terms (both into the containment, and subsequently, into the environment) do not display a high degree of sensitivity to the individual fission product issues addressed in this project. This is due, mainly, to the substantial capacity for the attenuation of airborne fission products by the designed safety provisions and the natural fission product retention mechanisms within the containment

Potential for containment leak paths through electrical penetration assemblies under severe accident conditions. [PWR; BWR

Energy Technology Data Exchange (ETDEWEB)

Sebrell, W.

1983-07-01

The leakage behavior of containments beyond design conditions and knowledge of failure modes is required for evaluation of mitigation strategies for severe accidents, risk studies, emergency preparedness planning, and siting. These studies are directed towards assessing the risk and consequences of severe accidents. An accident sequence analysis conducted on a Boiling Water Reactor (BWR), Mark I (MK I), indicated very high temperatures in the dry-well region, which is the location of the majority of electrical penetration assemblies. Because of the high temperatures, it was postulated in the ORNL study that the sealants would fail and all the electrical penetration assemblies would leak before structural failure would occur. Since other containments had similar electrical penetration assemblies, it was concluded that all containments would experience the same type of failure. The results of this study, however, show that this conclusion does not hold for PWRs because in the worst accident sequence, the long time containment gases stabilize to 350/sup 0/F. BWRs, on the other hand, do experience high dry-well temperatures and have a higher potential for leakage.

Method for consequence calculations for severe accidents

International Nuclear Information System (INIS)

Nielsen, F.

1988-01-01

This report was commissioned by the Swedish State Power Board. The report contains a calculation of radiation doses in the surroundings caused by a theoretical core meltdown accident at Ringhals reactor No 3/4. The accident sequence chosen for the calcualtions was a release caused by total power failure. The calculations were made by means of the PLUCON4 code. A decontamination factor of 500 is used to account for the scrubber effect. Meteorological data for two years from the Ringhals meteorological tower were analysed to find representative weather situations. As typical weather, Pasquill D, was chosen with a wind speed of 10 m/s, and as extreme weather, Pasquill E, with a wind speed of 2 m/s. 19 refs. (author)

Severe accident analysis and management in nuclear fuel cycle facilities

International Nuclear Information System (INIS)

Golshan, Mina

2013-01-01

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

Development Status of Accident Tolerant Fuels for Light Water Reactors in Korea

Energy Technology Data Exchange (ETDEWEB)

Yang, Jae Ho; Kim, Hyun Gil; In, Wang Kee; Kim, Weon Ju; Koo, Yang Hyum [KAERI, Daejeon (Korea, Republic of); Lee, Seung Jae [KEPCONF, Daejeon (Korea, Republic of)

2016-05-15

Research on accident tolerant fuels (ATFs) is aimed at developing innovative fuels, which can mitigate or prevent the consequences of accidents. In Korea, innovative concepts are being developed to improve fuel safety and reliability of LWRs during accident events and normal operations. ATF technologies will be developed and commercialized through a sequence of long-lead and extensive activities. The interim milestone for new fuel program is that we would be ready for an irradiation test in commercial reactor by 2021. This presentation deals with the status of ATF development in KOREA and plan to implement new fuel technology successfully in commercial nuclear power plants.

Increase the Safety of Road Traffic Accidents by Applying Clustering

Directory of Open Access Journals (Sweden)

Kos Goran

2013-12-01

Full Text Available In terms of continual increase of number of traffic accidents and alarming trend of increasing number of traffic accidents with catastrophic consequences for human life and health, it is necessary to actively research and develop methods to combat these trends. One of the measures is the implementation of advanced information systems in existing traffic environment. Accidents clusters, as databases of traffic accidents, introduce a new dimension in traffic systems in the form of experience, providing information on current accidents and the ones that have previously occurred in a given period. This paper proposes a new approach to predictive management of traffic processes, based on the collection of data in real time and is based on accidents clusters. The modern traffic information services collects road traffic status data from a wide variety of traffic sensing systems using modern ICT technologies, creating the most accurate road traffic situation awareness achieved so far. Road traffic situation awareness enhanced by accident clusters' data can be visualized and distributed in various ways (including the forms of dynamic heat maps and on various information platforms, suiting the requirements of the end-users. Accent is placed on their significant features that are based on additional knowledge about existing traffic processes and distribution of important traffic information in order to prevent and reduce traffic accidents.

Comparison of the phenomenology of SBO sequences with and without seals LOCA Westinghouse PWRs

International Nuclear Information System (INIS)

Mena Rosell, L.; Queral, C.; Jimenez Varas, G.

2013-01-01

SBO sequences have gained notoriety after the accident at Fukushima. Within this type of sequence the appearance or not of seals of the RCP LOCA determines the evolution of the accident. This work has been applied the methodology of integrated safety analysis (ISA), developed by the CSN, sequences of SBO. The objective is to compare the evolution of SBO sequences in a wide spectrum of conditions and recovery times of AC and DC loss. The simulations have been performed with the SCAIS tool coupled to MAAP. The set of simulations carried out, of the order of 2,000 sequences, clearly show the differences in the evolution of sequences with and without seals crazy. This type of analysis allows you to verify which would be the most appropriate management of sequence depending on the appearance or not of the MADWOMAN of seals.

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)

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

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)

Validation of severe accident management guidance for the wolsong plants

International Nuclear Information System (INIS)

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

2006-01-01

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

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

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)

Generic implications of the Chernobyl accident

International Nuclear Information System (INIS)

Sege, G.

1989-01-01

The US Nuclear Regulatory Commission (NRC) staff's assessment of the generic implications of the Chernobyl accident led to the conclusion that no immediate changes in the NRC's regulations regarding design or operation of US commercial reactors are needed. However, further consideration of certain issues was recommended. This paper discusses those issues and the studies being addressed to them. Although 24 tasks relating to light water reactor issues are identified in the Chernobyl follow-up research program, only four are new initiatives originating from Chernobyl implications. The remainder are limited modifications of ongoing programs designed to ensure that those programs duly reflect any lessons that may be drawn from the Chernobyl experience. The four new study tasks discussed include a study of reactivity transients, to reconfirm or bring into question the adequacy of potential reactivity accident sequences hitherto selected as a basis for design approvals; analysis of risk at low power and shutdown; a study of procedure violations; and a review of current NRC testing requirements for balance of benefits and risks. Also discussed, briefly, are adjustments to ongoing studies in the areas of operational controls, design, containment, emergency planning, and severe accident phenomena

Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

Directory of Open Access Journals (Sweden)

Andrej Prošek

2013-01-01

Full Text Available Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO. Long-term SBO in a pressurized water reactor (PWR leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pumps (RCPs leaks assumed to delay the time before core uncovers and significantly heats up. The most important strategies assumed were primary side depressurization and additional makeup water to reactor coolant system (RCS. For simulations of long term SBO scenarios, including early stages of severe accident sequences, the best estimate RELAP5/MOD3.3 and the verified input model of Krško two-loop PWR were used. The results suggest that for the expected magnitude of RCPs seal leak, the core uncovery during the first seven days could be prevented by using the turbine-driven auxiliary feedwater pump and manually depressurizing the RCS through the secondary side. For larger RCPs seal leaks, in general this is not the case. Nevertheless, the core uncovery can be significantly delayed by increasing RCS depressurization.

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

International Nuclear Information System (INIS)

Wang Yuexing

1999-01-01

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

Consequence of potential accidents in heavy water plants

International Nuclear Information System (INIS)

Croitoru, C.; Lazar, R.E.; Preda, I.A.; Dumitrescu, M.

1998-01-01

Heavy water plants realize the primary isotopic concentrations of water using H 2 O-H 2 S chemical exchange and they are chemical plants. As these plants are handling and spreading large quantities of hydrogen sulphide (high toxic, corrosive, flammable and explosive as) maintained in the process at relative high temperatures and pressures, it is required an assessing of risks associated with the potential accidents. The H 2 S released in atmosphere as a result of an accident will have negative consequences to property, population and environment. This paper presents a model of consequences quantitative assessment and its outcome for the most dangerous accident in heavy water plants. Several states of the art risk based methods were modified and linked together to form a proper model for this analyse. Five basic steps to identify the risks involved in operating the plants are followed: hazard identification, accident sequence development, H 2 S emissions calculus, dispersion analyses and consequences determination. A brief description of each step and some information of analysis results are provided. The accident proportions, the atmospheric conditions and the population density in the respective area were accounted for consequences calculus. The specific results of the consequences analysis allow to develop the plant's operating safety requirements so that the risk remain at an acceptable level. (authors)

Heat transport and afterheat removal for gas cooled reactors under accident conditions

International Nuclear Information System (INIS)

2001-01-01

The Co-ordinated Research Project (CRP) on Heat Transport and Afterheat Removal for Gas Cooled Reactors Under Accident Conditions was organized within the framework of the International Working Group on Gas Cooled Reactors (IWGGCR). This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs) and supports the conduct of these activities. Advanced GCR designs currently being developed are predicted to achieve a high degree of safety through reliance on inherent safety features. Such design features should permit the technical demonstration of exceptional public protection with significantly reduced emergency planning requirements. For advanced GCRs, this predicted high degree of safety largely derives from the ability of the ceramic coated fuel particles to retain the fission products under normal and accident conditions, the safe neutron physics behaviour of the core, the chemical stability of the core and the ability of the design to dissipate decay heat by natural heat transport mechanisms without reaching excessive temperatures. Prior to licensing and commercial deployment of advanced GCRs, these features must first be demonstrated under experimental conditions representing realistic reactor conditions, and the methods used to predict the performance of the fuel and reactor must be validated against these experimental data. Within this CRP, the participants addressed the inherent mechanisms for removal of decay heat from GCRs under accident conditions. The objective of this CRP was to establish sufficient experimental data at realistic conditions and validated analytical tools to confirm the predicted safe thermal response of advance gas cooled reactors during accidents. The scope includes experimental and analytical investigations of heat transport by natural convection conduction and thermal

Application of probabilistic methods to accident analysis at waste management facilities

International Nuclear Information System (INIS)

Banz, I.

1986-01-01

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

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

International Nuclear Information System (INIS)

Anon.

1989-01-01

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

Analysis of two different types of hydrogen combustion during severe accidents in a typical pressurized water reactor

International Nuclear Information System (INIS)

Ko Yuchih; Lee Min

2005-01-01

Hydrogen combustion is an important phenomenon that may occur during severe accidents of Nuclear Power Plants (NPPs). Depending on the specific plant design, the initiating events, and mitigation actions executed, hydrogen combustion may have distinct characteristics and may damage the plant in various degrees. The worst scenario will be the catastrophic failure of containment. In this study two specific types of hydrogen combustion are analyzed to evaluate their impact on the containment integrity. In this paper, Station Blackout (SBO) and Loss of Coolant Accidents (LOCAs) sequences are analyzed using MAAP4 (Modular Accident Analysis Program) code. The former sequence is used to represent hydrogen combustion phenomenon under the condition that the reactor pressure vessel (RPV) breaches at high pressure and the latter sequence represents the phenomenon that RPV fails at low pressure. Two types of hydrogen combustion are observed in the simulation. The Type I hydrogen combustion represents global and instantaneous hydrogen combustion. Large pressure spike is created during the combustion and represents a threat to containment integrity. Type II hydrogen combustion is localized burn and burn continuously over a time period. There is hardly any impact of this type hydrogen burn on the containment pressurization rate. Both types of hydrogen combustion can occur in the severe accidents without any human intervention. From the accident mitigation point of view, operators should try to bring the containment into conditions that favor the Type II hydrogen combustion. (authors)

Accident management insights after the Fukushima Daiichi NPP accident

International Nuclear Information System (INIS)

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

2014-01-01

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

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)

The need to study of bounding accident in reprocessing plant

International Nuclear Information System (INIS)

Segawa, Satoshi; Fujita, Kunio

2013-01-01

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

A framework for the assessment of severe accident management strategies

International Nuclear Information System (INIS)

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

1992-01-01

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

Safety Assessment of Advanced Imaging Sequences I: Measurements

DEFF Research Database (Denmark)

Jensen, Jørgen Arendt; Rasmussen, Morten Fischer; Pihl, Michael Johannes

2016-01-01

intensity measurement program. The approach can measure and store data for a full imaging sequence in 3.8 to 8.2 s per spatial position. Based on Ispta, MI, and probe surface temperature, the method gives the ability to determine whether a sequence is within US FDA limits, or alternatively indicate how......A method for rapid measurement of intensities (Ispta), mechanical index (MI), and probe surface temperature for any ultrasound scanning sequence is presented. It uses the scanner’s sampling capability to give an accurate measurement of the whole imaging sequence for all emissions to yield the true...... measurement system (Onda Corporation, Sunnyvale, CA, USA). Four different sequences have been measured: a fixed focus emission, a duplex sequence containing B-mode and flow emissions, a vector flow sequence with B-mode and flow emissions in 17 directions, and finally a synthetic aperture (SA) duplex flow...

Unavoidable Accident

OpenAIRE

Grady, Mark F.

2009-01-01

In negligence law, "unavoidable accident" is the risk that remains when an actor has used due care. The counterpart of unavoidable accident is "negligent harm." Negligence law makes parties immune for unavoidable accident even when they have used less than due care. Courts have developed a number of methods by which they "sort" accidents to unavoidable accident or to negligent harm, holding parties liable only for the latter. These sorting techniques are interesting in their own right and als...

Modification of MELCOR for severe accident analysis of candidate accident tolerant cladding materials

Energy Technology Data Exchange (ETDEWEB)

Merrill, Brad J., E-mail: brad.merrill@inl.gov; Bragg-Sitton, Shannon M., E-mail: shannon.bragg-sitton@inl.gov; Humrickhouse, Paul W., E-mail: paul.humrickhouse@inl.gov

2017-04-15

Highlights: • Accident tolerant fuels (ATF) systems are currently under development for LWRs. • Many performance analysis tools are specifically developed for UO{sub 2}–Zr alloy fuel. • Modifications were made to the MELCOR code for candidate ATF cladding. • Preliminary analysis results for SiC and FeCrAl cladding concepts are presented. - Abstract: A number of materials are currently under development as candidate accident tolerant fuel and cladding for application in the current fleet of commercial light water reactors (LWRs). The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Enhancing the accident tolerance of light water reactors became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal for the development of accident tolerant fuel (ATF) systems for LWRs is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. Designed for use in the current fleet of commercial LWRs, or in reactor concepts with design certifications (GEN-III+), to achieve their goal enhanced ATF must endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system, while maintaining or improving performance during normal operation. Many available nuclear fuel performance analysis tools are specifically developed for the current UO{sub 2}–Zirconium alloy fuel system. The MELCOR severe-accident analysis code, under development at the Sandia National Laboratory in New Mexico (SNL-NM) for the US Nuclear Regulatory Commission (NRC), is one of these tools. This paper describes modifications

System 80+TM PRA insights on severe accident prevention and mitigation

International Nuclear Information System (INIS)

Finnicum, D.J.; Jacob, M.C.; Schneider, R.E.; Weston, R.A.

2004-01-01

The System 80 + design is ABB-CE's standardized evolutionary Advanced Light Water Reactor (ALWR) design. It incorporates design enhancements based on Probabilistic Risk Assessment (PRA) insights, guidance from the ALWR Utility Requirements Document (URD), and US NRC's Severe Accident Policy. Major severe accident prevention and mitigation design features of the System 80 + design are described. The results of the System 80 + PRA are presented and the insights gained from the PRA sensitivity analyses are discussed. ABB-CE considered defense-in-depth for accident prevention and mitigation early in the design process and used robust design features to ensure that the System 80 + design achieved a low core damage frequency, low containment conditional failure probability, and excellent deterministic containment performance under severe accident conditions and to ensure that the risk was properly allocated among design features and between prevention and mitigation. (author)

Implementation of severe accident management measures - Summary and conclusions

International Nuclear Information System (INIS)

2002-01-01

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

Road Traffic Accident Analysis of Ajmer City Using Remote Sensing and GIS Technology

Science.gov (United States)

Bhalla, P.; Tripathi, S.; Palria, S.

2014-12-01

With advancement in technology, new and sophisticated models of vehicle are available and their numbers are increasing day by day. A traffic accident has multi-facet characteristics associated with it. In India 93% of crashes occur due to Human induced factor (wholly or partly). For proper traffic accident analysis use of GIS technology has become an inevitable tool. The traditional accident database is a summary spreadsheet format using codes and mileposts to denote location, type and severity of accidents. Geo-referenced accident database is location-referenced. It incorporates a GIS graphical interface with the accident information to allow for query searches on various accident attributes. Ajmer city, headquarter of Ajmer district, Rajasthan has been selected as the study area. According to Police records, 1531 accidents occur during 2009-2013. Maximum accident occurs in 2009 and the maximum death in 2013. Cars, jeeps, auto, pickup and tempo are mostly responsible for accidents and that the occurrence of accidents is mostly concentrated between 4PM to 10PM. GIS has proved to be a good tool for analyzing multifaceted nature of accidents. While road safety is a critical issue, yet it is handled in an adhoc manner. This Study is a demonstration of application of GIS for developing an efficient database on road accidents taking Ajmer City as a study. If such type of database is developed for other cities, a proper analysis of accidents can be undertaken and suitable management strategies for traffic regulation can be successfully proposed.

Radiation dose assessment of ACP hot cell in accident

International Nuclear Information System (INIS)

Kook, D. H.; Jeong, W. M.; Koo, J. H.; Jeo, I. J.; Lee, E. P.; Ryu, K. S.

2003-01-01

The Advanced spent fuel Condition in Process(ACP) is under development for the effective management of spent fuel which had been generated in nuclear plants. The ACP needs a hot cell where most operations will be performed. To give priority to the environments safety, radiation doses evaluations for the radioactive nuclides in accident cases were preliminarily performed with the meteorological data around facility site. Fire accident prevails over several accidnets. Internal Dose and External Dose evaluation according to short dispersion data for that case show a safe margin for regulation limits and SAR limit of IMEF where this facility will be constructed

North Wales Group report on the effects of the Chernobyl accident

International Nuclear Information System (INIS)

1987-11-01

A report is presented by the North Wales Group concerning the sequence of events affecting North Wales and the identification of the residual problems following contamination from the Chernobyl accident. The first part of the report attempts to establish a time scale for radiation restrictions applicable in North Wales and the size of the areas which are involved. Part two deals with national arrangements to handle incidents like Chernobyl and examines the wider field of international arrangements. A review is given of events as seen by the affected community following the Chernobyl accident. (U.K.)

Dynamic Sequence Assignment.

Science.gov (United States)

1983-12-01

D-136 548 DYNAMIIC SEQUENCE ASSIGNMENT(U) ADVANCED INFORMATION AND 1/2 DECISION SYSTEMS MOUNTAIN YIELW CA C A 0 REILLY ET AL. UNCLSSIIED DEC 83 AI/DS...I ADVANCED INFORMATION & DECISION SYSTEMS Mountain View. CA 94040 84 u ,53 V,..’. Unclassified _____ SCURITY CLASSIFICATION OF THIS PAGE REPORT...reviews some important heuristic algorithms developed for fas- ter solution of the sequence assignment problem. 3.1. DINAMIC MOGRAMUNIG FORMULATION FOR

Analysis of SBO accident and natural circulation of 49-2 swimming pool reactor

International Nuclear Information System (INIS)

Wu Yuanyuan; Liu Tiancai; Sun Wei

2012-01-01

The transient thermal hydraulic characteristics of 49-2 Swimming Pool Reactor (SPR) were analyzed by RELAP5/MOD3.3 code to verify the capability of natural circulation and minus reactivity feedback for accident mitigation under the condition of station blackout (SBO). Then, the effects on accident consequence and sequence for core channels and primary pumps were briefly discussed. The calculation results show that the reactor can be shutdown by the effect of minus reactivity feedback, and the residual heat can be removed through the stable natural circulation. Therefore, it demonstrates that the 49-2 SPR is safe during the accident of SBO. (authors)

Core loss during a severe accident (COLOSS)

International Nuclear Information System (INIS)

Adroguer, B.; Bertrand, F.; Chatelard, P.; Cocuaud, N.; Van Dorsselaere, J.P.; Bellenfant, L.; Knocke, D.; Bottomley, D.; Vrtilkova, V.; Belovsky, L.; Mueller, K.; Hering, W.; Homann, C.; Krauss, W.; Miassoedov, A.; Schanz, G.; Steinbrueck, M.; Stuckert, J.; Hozer, Z.; Bandini, G.; Birchley, J.; Berlepsch, T. von; Kleinhietpass, I.; Buck, M.; Benitez, J.A.F.; Virtanen, E.; Marguet, S.; Azarian, G.; Caillaux, A.; Plank, H.; Boldyrev, A.; Veshchunov, M.; Kobzar, V.; Zvonarev, Y.; Goryachev, A.

2005-01-01

The COLOSS project was a 3-year shared-cost action, which started in February 2000. The work-programme performed by 19 partners was shaped around complementary activities aimed at improving severe accident codes. Unresolved risk-relevant issues regarding H 2 production, melt generation and the source term were studied through a large number of experiments such as (a) dissolution of fresh and high burn-up UO 2 and MOX by molten Zircaloy (b) simultaneous dissolution of UO 2 and ZrO 2 (c) oxidation of U-O-Zr mixtures (d) degradation-oxidation of B 4 C control rods. Corresponding models were developed and implemented in severe accident computer codes. Upgraded codes were then used to apply results in plant calculations and evaluate their consequences on key severe accident sequences in different plants involving B 4 C control rods and in the TMI-2 accident. Significant results have been produced from separate-effects, semi-global and large-scale tests on COLOSS topics enabling the development and validation of models and the improvement of some severe accident codes. Breakthroughs were achieved on some issues for which more data are needed for consolidation of the modelling in particular on burn-up effects on UO 2 and MOX dissolution and oxidation of U-O-Zr and B 4 C-metal mixtures. There was experimental evidence that the oxidation of these mixtures can contribute significantly to the large H 2 production observed during the reflooding of degraded cores under severe accident conditions. The plant calculation activity enabled (a) the assessment of codes to calculate core degradation with the identification of main uncertainties and needs for short-term developments and (b) the identification of safety implications of new results. Main results and recommendations for future R and D activities are summarized in this paper

An assessment the severe accident equipment survivability for the Korean Next Generation Reactor

International Nuclear Information System (INIS)

Lee, B. C.; Moon, Y. T.; Park, J. W.; Kho, H. J.; Lee, S. W.

1999-01-01

One of the prominent design approaches to cope with the severe accident challenges in the Korean Next Generation Reactor is an assessment of equipment survivability in the severe accident environment at early design stage. In compliance with 10CFR50.34(f) and SECY-93-087, this work addresses that a reasonable level of assurance be provided to demonstrate that sufficient instrumentation and equipment will survive the consequences of a severe accident and will be available so that the operator may recover from and trend severe core damage sequences, including those scenarios which result in 100 percent oxidation of the active fuel cladding. An analytical and systematic approach was used to identify the equipment and instrumentation of safety-function and define severe accident environments including temperature, pressure, humidity, and radiation before and after the reactor vessel breach. As a result, it was concluded that with minor exceptions, existing design basis equipment qualification methods are sufficient to provide a reasonable level of assurance that this equipment will function during a severe accident. Furthermore, supplemental severe accident equipment and instrument procurement requirements were identified. (author)

Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario

Energy Technology Data Exchange (ETDEWEB)

Rempe, J. L. [Rempe and Associates, LLC, Idaho Falls, ID (United States); Knudson, D. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lutz, R. J. [Lutz Nuclear Safety Consultant, LLC, Asheville, NC (United States)

2015-09-01

The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 nuclear power plants demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data. While progress in these areas has been made since TMI-2, the events at Fukushima suggests that there may still be a potential need to ensure that critical plant information is available to plant operators. Recognizing the significant technical and economic challenges associated with plant modifications, it is important to focus on instrumentation that can address these information critical needs. As part of a program initiated by the Department of Energy, Office of Nuclear Energy (DOE-NE), a scoping effort was initiated to assess critical information needs identified for severe accident management and mitigation in commercial Light Water Reactors (LWRs), to quantify the environment instruments monitoring this data would have to survive, and to identify gaps where predicted environments exceed instrumentation qualification envelop (QE) limits. Results from the Pressurized Water Reactor (PWR) scoping evaluations are documented in this report. The PWR evaluations were limited in this scoping evaluation to quantifying the environmental conditions for an unmitigated Short-Term Station BlackOut (STSBO) sequence in one unit at the Surry nuclear power station. Results were obtained using the MELCOR models developed for the US Nuclear Regulatory Commission (NRC)-sponsored State of the Art Consequence Assessment (SOARCA) program project. Results from this scoping evaluation indicate that some instrumentation identified to provide critical information would be exposed to conditions that

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

Energy Technology Data Exchange (ETDEWEB)

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

1988-01-01

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

Effect of alternative aging and accident simulations on polymer properties

International Nuclear Information System (INIS)

Bustard, L.D.; Chenion, J.; Carlin, F.; Alba, C.; Gaussens, G.; LeMeur, M.

1985-05-01

The influence of accident irradiation, steam, and chemical spray exposures on the behavior of twenty-three age-preconditioned polymer sample sets (twenty-one different materials) has been investigated. The test program varied the following conditions: (1) Accident simulations of irradiation and thermodynamic (steam and chemical spray) conditions were performed both sequentially and simultaneously. (2) Accident thermodynamic (steam and chemical spray) exposures were performed both with and without air present during the exposures. (3) Sequential accident irradiations were performed both at 28 0 C and 70 0 C. (4) Age preconditioning was performed both sequentially and simultaneously. (5) Sequential aging irradiations were performed both at 27 0 C and 70 0 C. (6) Sequential aging exposures were performed using two sequences: (1) thermal followed by irradiation and (2) irradiation followed by thermal. We report both general trends applicable to a majority of the tested materials as well as specific results for each polymer. Our data base consists of ultimate tensile properties at the completion of the accident exposure for three XLPO and XLPE, five EPR and EPDM, two CSPE (HYPALON), one CPE, one VAMAC, one polydiallylphtalate, and one PPS material. We also report bend test results at completion of the accident exposures for two TEFZEL materials and permanent set after compression results for three EPR, one VAMAC, one BUNA N, one SILICONE, and one VITON material

Damage of reactor buildings occurred at the Fukushima Daiichi accident. Focusing on sequence leading to hydrogen explosions

International Nuclear Information System (INIS)

Naito, Masanori

2011-01-01

Fukushima Daiichi accident discharged enormous radioactive materials confined inside into the environment due to hydrogen explosions occurred at reactor buildings and forced many people to live the refugee life. This article described overview of Great East Japan Earthquake, specifications of Fukushima Daiichi nuclear power plants, sequence of plant status after earthquake occurrence and computerized simulation of plant behavior of Unit 1 leading to core melt and hydrogen explosion. Simulation results with estimated and assumed conditions showed water level decreased to bottom of reactor core after 4 hrs and 15 minutes passed, core melt started after 6 hrs and 49 minutes passed, failure of core support plate after 7 hrs and 18 minutes passed and through failure of penetration at bottom of pressure vessel after 7 hrs and 25 minutes passed. Hydrogen concentration at operating floor of reactor building of Unit 1 would be 15% accumulated and the pressure would amount to about 5 bars after hydrogen explosion if reactor building did not rupture with leak-tight structure. Since reactor building was not pressure-proof structure, walls of operating floor would rupture before 5 bars attained. (T. Tanaka)

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

Organisational factors of occupational accidents with movement disturbance (OAMD) and prevention.

Science.gov (United States)

Leclercq, Sylvie

2014-01-01

Workplace design and upkeep, or human factors, are frequently advanced for explaining so-called Occupational Slip, Trip and Fall Accidents (OSTFAs). Despite scientific progress, these accidents, and more broadly Occupational Accidents with Movement Disturbance (OAMDs), are also commonly considered to be "simple". This paper aims to stimulate changes in such perceptions by focusing on organisational factors that often combine with other accident factors to cause movement disturbance and injury in work situations. These factors frequently lead to arbitration between production and safety, which involves implementation of controls by workers. These controls can lead to greater worker exposure to OAMD risk. We propose a model that focuses on such controls to account specifically for the need to confront production and safety logics within a company and to enhance the potential for appropriate prevention action. These are then integrated into the set of controls highlighted by work organisation model developed by the NIOSH.

Advanced Instrumentation for Transient Reactor Testing

Energy Technology Data Exchange (ETDEWEB)

Corradini, Michael L.; Anderson, Mark; Imel, George; Blue, Tom; Roberts, Jeremy; Davis, Kurt

2018-01-31

Transient testing involves placing fuel or material into the core of specialized materials test reactors that are capable of simulating a range of design basis accidents, including reactivity insertion accidents, that require the reactor produce short bursts of intense highpower neutron flux and gamma radiation. Testing fuel behavior in a prototypic neutron environment under high-power, accident-simulation conditions is a key step in licensing nuclear fuels for use in existing and future nuclear power plants. Transient testing of nuclear fuels is needed to develop and prove the safety basis for advanced reactors and fuels. In addition, modern fuel development and design increasingly relies on modeling and simulation efforts that must be informed and validated using specially designed material performance separate effects studies. These studies will require experimental facilities that are able to support variable scale, highly instrumented tests providing data that have appropriate spatial and temporal resolution. Finally, there are efforts now underway to develop advanced light water reactor (LWR) fuels with enhanced performance and accident tolerance. These advanced reactor designs will also require new fuel types. These new fuels need to be tested in a controlled environment in order to learn how they respond to accident conditions. For these applications, transient reactor testing is needed to help design fuels with improved performance. In order to maximize the value of transient testing, there is a need for in-situ transient realtime imaging technology (e.g., the neutron detection and imaging system like the hodoscope) to see fuel motion during rapid transient excursions with a higher degree of spatial and temporal resolution and accuracy. There also exists a need for new small, compact local sensors and instrumentation that are capable of collecting data during transients (e.g., local displacements, temperatures, thermal conductivity, neutron flux, etc.).

Self-reported accidents

DEFF Research Database (Denmark)

Møller, Katrine Meltofte; Andersen, Camilla Sloth

2016-01-01

The main idea behind the self-reporting of accidents is to ask people about their traffic accidents and gain knowledge on these accidents without relying on the official records kept by police and/or hospitals.......The main idea behind the self-reporting of accidents is to ask people about their traffic accidents and gain knowledge on these accidents without relying on the official records kept by police and/or hospitals....

Biological effects of ionizing radiations. Radiological accident from Goiania, GO, Brazil

International Nuclear Information System (INIS)

Okuno, Emico

2013-01-01

This article presents the fundaments of radiation physics, the natural and artificial sources, biological effects, radiation protection. We also examine the sequence of events that resulted in Goiania accident with a source of caesium-137 from abandoned radiotherapy equipment and its terrible consequences. (author)

Assessment of severe accident prevention and mitigation features: PWR, large dry containment design

International Nuclear Information System (INIS)

Perkins, K.R.; Hsu, C.J.; Lehner, J.R.; Luckas, W.J.; Cho, N.; Fitzpatrick, R.G.; Pratt, W.T.; Eltawila, F.; Maly, J.A.

1988-07-01

Plant features and operator actions which have been found to be important in either preventing or mitigating severe accidents in PWRs with large dry containments have been identified. These features and actions were developed from insights derived from reviews of risk assessments performed specifically for the Zion plant and from assessments of other relevant studies. Accident sequences that dominate the core-damage frequency and those accident sequences that are of potentially high consequence were identified. Vulnerabilities of the large dry containment to severe accident containment loads were also identified. In addition, those features of a PWR with a large dry containment, which are important for preventing core damage and are available for mitigating fission-product release to the environment were identified. The report is issued to provide focus to the analyst examining an individual plant. The report calls attention to plant features and operator actions and provides a list of deterministic tributes for assessing those features and actions found to be helpful in reducing the overall risk for Zion and other PWRs with large dry containments. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, plant features and operator actions will be of value in reducing overall plant risk. This report is intended to serve solely as guidance

Assessment of severe accident prevention and mitigation features: PWR, ice-condenser containment design

International Nuclear Information System (INIS)

Hsu, C.J.; Perkins, K.R.; Luckas, W.J.; Fitzpatrick, R.G.; Cho, N.; Lehner, J.R.; Pratt, W.T.; Eltawila, F.; Maly, J.A.

1988-07-01

Plant features and operator actions which have been found to be important in either preventing and mitigating severe accidents in PWRs with ice-condenser containments have been identified. Thus features and actions were developed from insights derived from reviews of risk assessments performed specifically for the Sequoyah plant and from assessments of other relevant studies. Accident sequences that dominate the core-damage frequency and those accident sequences that are of potentially high consequence were identified. Vulnerabilities of the ice-condenser containment to sever accident containment loads were also identified. In addition, those features of a PWR with an ice-condenser containment, which are important for preventing core damage and are available for mitigating fission-product release to the environment were identified. This report is issued to provide focus to an analyst examining an individual plant. The report calls attention to plant features and operator actions and provides a list of deterministic attributes for assessing those features and actions found to be helpful in reducing the overall risk for Sequoyah and other PWRs with ice-condenser containments. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, plant features and operator actions will be of value in reducing overall plant risk. This report is intended to serve solely as guidance. 14 tabs

Assessment of severe accident prevention and mitigation features: BWR, Mark II containment design

International Nuclear Information System (INIS)

Lehner, J.R.; Hsu, C.J.; Eltawila, F.; Perkins, K.R.; Luckas, W.J.; Fitzpatrick, R.G.; Pratt, W.T.

1988-07-01

Plant features and operator actions, which have been found to be important in either preventing or mitigating severe accidents in BWRs with Mark II containments (BWR Mark II's) have been identified. These features and actions were developed from insights derived from reviews of in-depth risk assessments performed specifically for the Limerick and Shoreham plants and from other relevant studies. Accident sequences that dominate the core-damage frequency and those accident sequences that are of potentially high consequence were identified. Vulnerabilities of the BWR Mark II to severe-accident containment loads were also noted. In addition, those features of a BWR Mark II, which are important for preventing core damage and are available for mitigating fission-product release to the environment were also identified. This report is issued to provide focus to an analyst examining an individual plant. This report calls attention to plant features and operator actions and provides a list of deterministic attributes for assessing those features and actions found to be helpful in reducing the overall risk for Mark II plants. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, plant features and operator actions will be of value in reducing overall plant risk. This report is intended to serve solely as guidance

The accident at the Three Mile Island nuclear power plant

International Nuclear Information System (INIS)

Butragueno, J.L.

1980-01-01

The sequence of events in the Three Mile Island, Unit 2, accident on the March 28, 1979 is analyzed. In this plant a loss of feed-water transient became a small LOCA that caused a serious core damage. A general emergency situation was declared after uncontrolled radioactive releases were detectec. (author)

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

International Nuclear Information System (INIS)

1986-11-01

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

Main safety issues related to IPSN severe accident research

International Nuclear Information System (INIS)

LeComte, C.

1991-01-01

The work performed at IPSN concerning accident studies on nuclear installations is focused on the characterization of accidental sequences with three major aims: prevention, mitigation, and organization of counter-measures. As criteria to optimize all efforts made to improve nuclear safety, the radioactive dispersal in the environment must be quantified as function of internal and external radioactive products transfers. During the short-term phase of the accident, potential radioactive releases can be evaluated by the realistic code system ESCADRE. This system is validated by numerous analytical studies related to containment and fission product behavior. It will be further qualified by the results of the global experiments performed in the PHEBUS FP facility at IPSN

Study on mitigation of in-vessel release of fission products in severe accidents of PWR

International Nuclear Information System (INIS)

Huang, G.F.; Tong, L.L.; Li, J.X.; Cao, X.W.

2010-01-01

Research highlights: → In-vessel release of fission products in severe accidents for 600 MW PWR is analyzed. → Mitigation effect of primary feed-and-bleed on in-vessel release is investigated. → Mitigation effect of secondary feed-and-bleed on in-vessel release is studied. → Mitigation effect of ex-vessel cooling on in-vessel release is evaluated. - Abstract: During the severe accidents in a nuclear power plant, large amounts of fission products release with accident progression, including in-vessel and ex-vessel release. Mitigation of fission products release is demanded for alleviating radiological consequence in severe accidents. Mitigation countermeasures to in-vessel release are studied for Chinese 600 MW pressurized water reactor (PWR), including feed-and-bleed in primary circuit, feed-and-bleed in secondary circuit and ex-vessel cooling. SBO, LOFW, SBLOCA and LBLOCA are selected as typical severe accident sequences. Based on the evaluation of in-vessel release with different startup time of countermeasure, and the coupling relationship between thermohydraulics and in-vessel release of fission products, some results are achieved. Feed-and-bleed in primary circuit is an effective countermeasure to mitigate in-vessel release of fission products, and earlier startup time of countermeasure is more feasible. Feed-and-bleed in secondary circuit is also an effective countermeasure to mitigate in-vessel release for most severe accident sequences that can cease core melt progression, e.g. SBO, LOFW and SBLOCA. Ex-vessel cooling has no mitigation effect on in-vessel release owing to inevitable core melt and relocation.

Relation between workplace accidents and the levels of carboxyhemoglobin in motorcycle taxi drivers.

Science.gov (United States)

da Silva, Luiz Almeida; Robazzi, Maria Lúcia do Carmo Cruz; Terra, Fábio de Souza

2013-01-01

To investigate the relation between workplace accidents and the levels of carboxyhemoglobin found in motorcycle taxi drivers. Correlational, quantitative study involving 111 workers and data obtained in July 2012 through a questionnaire to characterize the participants and blood collection to measure carboxyhemoglobin levels. 28.8% had suffered workplace accidents; 27.6% had fractured the lower limbs and significant symptoms of carbon monoxide exposure were verified in smokers. The carboxyhemoglobin levels were higher among smokers and victims of workplace accidents. Motorcycle taxi drivers had increased levels of carboxyhemoglobin, possibly due to the exposure to carbon monoxide; these levels are also increased among smokers and victims of workplace accidents. The study provides advances in the knowledge about occupational health and environmental science, and also shows that carboxyhemoglobin can be an indicator of exposure to environmental pollutants for those working outdoors, which can be related to workplace accidents.

Regulatory analyses for severe accident issues: an example

International Nuclear Information System (INIS)

Burke, R.P.; Strip, D.R.; Aldrich, D.C.

1984-09-01

This report presents the results of an effort to develop a regulatory analysis methodology and presentation format to provide information for regulatory decision-making related to severe accident issues. Insights and conclusions gained from an example analysis are presented. The example analysis draws upon information generated in several previous and current NRC research programs (the Severe Accident Risk Reduction Program (SARRP), Accident Sequence Evaluation Program (ASEP), Value-Impact Handbook, Economic Risk Analyses, and studies of Vented Containment Systems and Alternative Decay Heat Removal Systems) to perform preliminary value-impact analyses on the installation of either a vented containment system or an alternative decay heat removal system at the Peach Bottom No. 2 plant. The results presented in this report are first-cut estimates, and are presented only for illustrative purposes in the context of this document. This study should serve to focus discussion on issues relating to the type of information, the appropriate level of detail, and the presentation format which would make a regulatory analysis most useful in the decisionmaking process

Fuel relocation modeling in the SAS4A accident analysis code system

International Nuclear Information System (INIS)

Tentner, A.M.; Miles, K.J.

1985-01-01

SAS4A is a new code system which has been designed for analyzing the initial phase of Hypothetical Core Disruptive Accidents (HCDAs) up to gross melting or failure of the subassembly walls. During such postulated accident scenarios as the Loss-of-Flow (LOF) and Transient-Overpower (TOP) events, the relocation of the fuel plays a key role in determining the sequence of events and the amount of energy produced before neutronic shutdown. This paper discusses the general strategy used in modeling the various phenomena which lead to fuel relocation and presents the key fuel relocation models used in SAS4A. The implications of these models for the whole-core accident analysis as well as recent results of fuel motion experiment analyses are also presented

Fuel relocation modeling in the SAS4A accident analysis code system

International Nuclear Information System (INIS)

Tentner, A.M.; Miles, K.J.; Kalimullah; Hill, D.J.

1986-01-01

The SAS4A code system has been designed for the analysis of the initial phase of Hypothetical Core Disruptive Accidents (HCDAs) up to gross melting or failure of the subassembly walls. During such postulated accident scenarios as the Loss-of-Flow (LOF) and Transient-Overpower (TOP) events, the relocation of the fuel plays a key role in determining the sequence of events and the amount of energy produced before neutronic shutdown. This paper discusses the general strategy used in modelong the various phenomena which lead to fuel relocation and presents the key fuel relocation models used in SAS4A. The implications of these models for the whole-core accident analysis as well as recent results of fuel relocation are emphasized. 12 refs

APR1400 CEA Withdrawal at Power Accident Analysis using KNAP

International Nuclear Information System (INIS)

Lee, Dong-Hyuk; Yang, Chang-Keun; Kim, Yo-Han; Sung, Chang-Kyung

2006-01-01

KEPRI (Korea Electric Power Research Institute) has been developing safety analysis methodology for non- LOCA (Loss Of Coolant Accident) analysis of OPR1000 (Optimized Power Reactor 1000, formerly KSNP). The new methodology, named KNAP (Korea Non-LOCA Analysis Package), uses RETRAN as the main system analysis code. RETRAN code is a non- LOCA safety analysis code developed by EPRI. The new methodology will replace existing CE (Combustion Engineering) supplied codes and methodologies currently used in non-LOCA analysis of OPR1000. In this paper, we apply KNAP methodology to APR1400 (Advanced Power Reactor 1400). The CEA (Control Element Assembly) withdrawal at power accident is one of the 'reactivity and power distribution anomalies' events and the results are typically described in the chapter 15.4.2 of SAR (Safety Analysis Report). The APR1400 has been designed to generate 1,400MWe of electricity with advanced features for greatly enhanced safety and economic goals. The CEA withdrawal at power analysis in APR1400 SSAR (Standard Safety Analysis Report) is analyzed with CESEC-III computer code. In this study, to confirm the applicability of the KNAP methodology and code system to APR1400, CEA withdrawal at power accident is analyzed using RETRAN code and it is compared with results from APR1400 SSAR

Accident source terms for pressurized water reactors with high-burnup cores calculated using MELCOR 1.8.5.

Energy Technology Data Exchange (ETDEWEB)

Gauntt, Randall O.; Powers, Dana Auburn; Ashbaugh, Scott G.; Leonard, Mark Thomas; Longmire, Pamela

2010-04-01

In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs2MoO4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1988-02-15

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

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.

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.

Analysis of Three Mile Island-Unit 2 accident

International Nuclear Information System (INIS)

1980-03-01

The Nuclear Safety Analysis Center (NSAC) of the Electric Power Research Institute has analyzed the Three Mile Island-2 accident. Early results of this analysis were a brief narrative summary, issued in mid-May 1979 and an initial version of this report issued later in 1979 as noted in the Foreword. The present report is a revised version of the 1979 report, containing summaries, a highly detailed sequence of events, a comparison of that sequence of events with those from other sources, 25 appendices, references and a list of abbreviations and acronyms. A matrix of equipment and system actions is included as a folded insert

Minimization of the occupational doses during the liquidation of the radiation accident consequences

International Nuclear Information System (INIS)

Kuryndina, Lidia; Stroganov, Anatoly; Kuryndin, Anton

2008-01-01

Full text: As known the accident on the Chernobylskaya npp is the heaviest one in the nuclear energy history. It showed how considerable can be radiation levels on the breakdown nuclear facility. Nevertheless Russian specialists on radiation protection worked out and successfully realized a conception of the working in such conditions during the liquidation of the accident consequences. The conception based out on using ALARA principle, included the methods of radiation fields structure analysis and allowed to minimize of the occupational doses at operations of the accident consequences liquidation. The main idea of the conception is in strongly dependence between the radiation dose of the personnel performing the liquidation operations and concrete sequence of these operations. Also it is necessary from time to time to receive the experimental information about radiation situation dynamics on the breakdown facility and to make variant calculations for optimizing for the successful implementation of such approach. The structure of these calculations includes variable fraction for the actual state of the facility before the accident and after one and not variable fraction depend on the geometric and protection characteristics of the facility. And the second part is more complicated and bigger. Therefore the most part of these calculations required for the any successful liquidation of the accident consequences can be made on the facility projecting stage. If it will be made the following tasks can be solved in case of the accident: 1) To estimate a distribution of the contamination source using the radiation control system indications; 2) To determine a contribution from each source to the dose rate for any contaminated area; 3) To estimate the radiation doses of the personnel participated in the accident consequences liquidation; 4) To select and to realize the sequence of the liquidation operations giving the minimal doses. The paper will overview the description

Accident information needs

International Nuclear Information System (INIS)

Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

1992-01-01

A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information

Accident information needs

Energy Technology Data Exchange (ETDEWEB)

Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

1992-12-31

A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information.

Accident information needs

Energy Technology Data Exchange (ETDEWEB)

Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

1992-01-01

A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information.

Development of a prototype graphic simulation program for severe accident training

International Nuclear Information System (INIS)

Kim, Ko Ryu; Jeong, Kwang Sub; Ha, Jae Joo

2000-05-01

This is a report of the development process and related technologies of severe accident graphic simulators, required in industrial severe accident management and training. Here, we say 'a severe accident graphic simulator' as a graphics add-in system to existing calculation codes, which can show the severe accident phenomena dynamically on computer screens and therefore which can supplement one of main defects of existing calculation codes. With graphic simulators it is fairly easy to see the total behavior of nuclear power plants, where it was very difficult to see only from partial variable numerical information. Moreover, the fast processing and control feature of a graphic simulator can give some opportunities of predicting the severe accident advancement among several possibilities, to one who is not an expert. Utilizing graphic simulators' we expect operators' and TSC members' physical phenomena understanding enhancement from the realistic dynamic behavior of plants. We also expect that severe accident training course can gain better training effects using graphic simulator's control functions and predicting capabilities, and therefore we expect that graphic simulators will be effective decision-aids tools both in sever accident training course and in real severe accident situations. With these in mind, we have developed a prototype graphic simulator having surveyed related technologies, and from this development experiences we have inspected the possibility to build a severe accident graphic simulator. The prototype graphic simulator is developed under IBM PC WinNT environments and is suited to Uljin 3and4 nuclear power plant. When supplied with adequate severe accident scenario as an input, the prototype can provide graphical simulations of plant safety systems' dynamic behaviors. The prototype is composed of several different modules, which are phenomena display module, MELCOR data interface module and graphic database interface module. Main functions of

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

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

International Nuclear Information System (INIS)

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

1998-01-01

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

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

Identification of failure sequences sensitive to human error

International Nuclear Information System (INIS)

1987-06-01

This report prepared by the participants of the technical committee meeting on ''Identification of Failure Sequences Sensitive to Human Error'' addresses the subjects discussed during the meeting and the conclusions reached by the committee. Chapter 1 reviews the INSAG recommendations and the main elements of the IAEA Programme in the area of human element. In Chapter 2 the role of human actions in nuclear power plants safety from insights of operational experience is reviewed. Chapter 3 is concerned with the relationship between probabilistic safety assessment and human performance associated with severe accident sequences. Chapter 4 addresses the role of simulators in view of training for accident conditions. Chapter 5 presents the conclusions and future trends. The seven papers presented by members of this technical committee are also included in this technical document. A separate abstract was prepared for each of these papers

Thermodynamic correlations for the accident analysis of HTR's

International Nuclear Information System (INIS)

Rehm, W.; Jahn, W.; Finken, R.

1976-12-01

The thermal properties of Helium and for the case of a depressurized primary circuit, various mixtures of primary cooling gas were taken into consideration. The temperature dependence of the correlations for the thermal properties of the graphite components in the core and for the structural materials in the primary circuit are extrapolated about normal operation conditions. Furthermore the correlations for the effective thermal conductivity, the heat transfer and pressure drop are described for pebble bed HTR's. In addition some important heat transfer data of the steam generator are included. With these correlations, for example accident sequences with failure of the afterheat removal systems are discussed for pebble bed HTR's. It is concluded that the transient temperature behaviour demonstrates the inherent safety features of the HTR in extreme accidents. (orig.) [de

SWR 1000 severe accident control through in-vessel melt retention by external RPV cooling

Energy Technology Data Exchange (ETDEWEB)

Kolev, N.I. [Framatome Advanced Nuclear Power, NDSI, Erlangen (Germany)

2001-07-01

Framatome Advanced Nuclear Power is being designing a new generation NPP with boiling water reactor SWR1000. Besides of various of modern passive and active safety features the system is also designed for controlling of a postulated severe accident with extreme low probability of occurrence. This work presents the rationales behind the decision to select the external cooling as a safety management strategy during severe accident. Bounding scenery are analyzed regarding the core melting, melt-water interaction during relocation of the melt from the core region into the lower head and the external coolability of the lower head. The conclusion is reached that the external cooling for the SWR1000 is a valuable strategy for accident management during postulated severe accidents. (authors)

SWR 1000 severe accident control through in-vessel melt retention by external RPV cooling

International Nuclear Information System (INIS)

Kolev, N.I.

2001-01-01

Framatome Advanced Nuclear Power is being designing a new generation NPP with boiling water reactor SWR1000. Besides of various of modern passive and active safety features the system is also designed for controlling of a postulated severe accident with extreme low probability of occurrence. This work presents the rationales behind the decision to select the external cooling as a safety management strategy during severe accident. Bounding scenery are analyzed regarding the core melting, melt-water interaction during relocation of the melt from the core region into the lower head and the external coolability of the lower head. The conclusion is reached that the external cooling for the SWR1000 is a valuable strategy for accident management during postulated severe accidents. (authors)

MCCI study for Pressurized Heavy Water Reactor under hypothetical accident condition

International Nuclear Information System (INIS)

Verma, Vishnu; Mukhopadhyay, Deb; Chatterjee, B.; Singh, R.K.; Vaze, K.K.

2011-01-01

In case of severe core damage accident in Pressurized Heavy Water Reactor (PHWR), large amount of molten corium is expected to come out into the calandria vault due to failure of calandria vessel. Molten corium at high temperature is sufficient to decompose and ablate concrete. Such attack could fail CV by basement penetration. Since containment is ultimate barrier for activity release. The Molten Core Concrete Interaction (MCCI) of the resulting pool of debris with the concrete has been identified as an important part of the accident sequence. MCCI Analysis has been carried out for PHWR for a hypothetical accident condition where total core material is considered to be relocated in calandria vault. Concrete ablation rate in vertical and radial direction is evaluated for rectangular geometry using MEDICIS module of ASTEC Code. Amount of gases released during MCCI is also evaluated. (author)

Relation between workplace accidents and the levels of carboxyhemoglobin in motorcycle taxi drivers

Directory of Open Access Journals (Sweden)

Luiz Almeida da Silva

2013-09-01

Full Text Available OBJECTIVE: to investigate the relation between workplace accidents and the levels of carboxyhemoglobin found in motorcycle taxi drivers. METHOD: correlational, quantitative study involving 111 workers and data obtained in July 2012 through a questionnaire to characterize the participants and blood collection to measure carboxyhemoglobin levels. RESULT: 28.8% had suffered workplace accidents; 27.6% had fractured the lower limbs and significant symptoms of carbon monoxide exposure were verified in smokers. The carboxyhemoglobin levels were higher among smokers and victims of workplace accidents. CONCLUSION: motorcycle taxi drivers had increased levels of carboxyhemoglobin, possibly due to the exposure to carbon monoxide; these levels are also increased among smokers and victims of workplace accidents. The study provides advances in the knowledge about occupational health and environmental science, and also shows that carboxyhemoglobin can be an indicator of exposure to environmental pollutants for those working outdoors, which can be related to workplace accidents.

Theories of radiation effects and reactor accident analysis

International Nuclear Information System (INIS)

Williams, P.M.; Ball, S.J.

1996-01-01

Muckerheide's paper was a public breakthrough on how one might assess the public health effects of low-level radiation. By the organization of a wealth of data, including the consequences of Hiroshima and Nagasaki but not including Chernobyl, he was able to conclude that present radioactive waste disposal and cleanup efforts need to be much less arduous than forecast by the U.S. Department of Energy, which, together with regulators, uses the linear hypothesis of radiation damage to humans. While the linear hypothesis is strongly defended and even recommended for extension to noncarcinogenic pollutants, exploration of a conservative threshold for very low level exposures could save billions of dollars in disposing of radioactive waste, enhance the understanding of reactor accident consequences, and assist in the development of design and operating criteria pertaining to severe accidents. In this context, the authors discuss the major differences between design-basis and severe accidents. The authors propose that what should ultimately be done is to develop a regulatory formula for severe-accident analysis that relates the public health effects to the amount and type of radionuclides released and distributed by the Chernobyl accident. Answers to the following important questions should provide the basis of this study: (1) What should be the criteria for distinguishing between design-basis and severe accidents, and what should be the basis for these criteria? (2) How do, and should, these criteria differ for older plants, newer operating plants, type of plant (i.e., gas cooled, water cooled, and liquid metal), advanced designs, and plants of the former Soviet Union? (3) How safe is safe enough?

Approaches to accident analysis in recent US Department of Energy environmental impact statements

International Nuclear Information System (INIS)

Mueller, C.; Folga, S.; Nabelssi, B.

1996-01-01

A review of accident analyses in recent US Department of Energy (DOE) Environmental Impact Statements (EISs) was conducted to evaluate the consistency among approaches and to compare these approaches with existing DOE guidance. The review considered several components of an accident analysis: the overall scope, which in turn should reflect the scope of the EIS; the spectrum of accidents considered; the methods and assumptions used to determine frequencies or frequency ranges for the accident sequences; and the assumption and technical bases for developing radiological and chemical atmospheric source terms and for calculating the consequences of airborne releases. The review also considered the range of results generated with respect to impacts on various worker and general populations. In this paper, the findings of these reviews are presented and methods recommended for improving consistency among EISs and bringing them more into line with existing DOE guidance

Analysis of loss of decay heat removal sequences at Browns Ferry Unit One: Chapter 17

International Nuclear Information System (INIS)

Harrington, R.M.

1983-01-01

This paper summarizes the Oak Ridge National Laboratory (ORNL) report ''Loss of DHR Sequences at Browns Ferry Unit One - Accident Sequence Analysis'' (NUREG/CR-2973). The Loss of DHR investigation is the third in a series of accident studies concerning the BWR 4 - MK I containment plant design. These studies, sponsored by the Nuclear Regulatory Commission Severe Accident Sequence Analysis (SASA) program, have been conducted at ORNL with the full cooperation of the Tennessee Valley Authority (TVA), using Unit One of the Browns Ferry Nuclear Plant as the model design. Each unit of this three-unit plant has a maximum authorized power of 3293 MW(t) or 1067 net MW(e). The primary containments are of the Mark I pressure suppression pool type and the three units share a secondary containment of the controlled leakage, elevated release design. Each unit occupies a separate reactor building located in one structure underneath the common refueling floor

MAAP - modular program for analyses of severe accidents

International Nuclear Information System (INIS)

Henry, R.E.; Lutz, R.J.

1990-01-01

The MAAP computer code was developed by Westinghouse as a fast, user-friendly, integrated analytical tool for evaluations of the sequences and consequences of severe accidents. The code allows a fully integrated treatment of thermohydraulic behavior and of the fission products in the primary system, the containment, and the ancillary buildings. This ensures interactive inclusion of all thermohydraulic events and of fission product behavior. All important phenomena which may occur in a major accident are contained in the modular code. In addition, many of the important parameters affecting the multitude of different phenomena can be defined by the user. In this way, it is possible to study the accuracy of the predicted course and of the consequences of a series of major accident phenomena. The MAAP code was subjected to extensive benchmarking with respect to the results of the experimental and theoretical programs, the findings obtained in other safety analyses using computers and data from accidents and transients in plants actually in operation. With the expected connection of the validation and test programs, the computer code attains a quality standard meeting the most stringent requirements in safety analyses. The code will be enlarged further in order to expand the number of benchmarks and the resolution of individual comparisons, and to ensure that future MAAP models will be in better agreement with the experiments and experiences of industry. (orig.) [de

Comparison of HRA methods based on WWER-1000 NPP real and simulated accident scenarios

International Nuclear Information System (INIS)

Petkov, Gueorgui

2010-01-01

Full text: Adequate treatment of human interactions in probabilistic safety analysis (PSA) studies is a key to the understanding of accident sequences and their relative importance in overall risk. Human interactions with machines have long been recognized as important contributors to the safe operation of nuclear power plants (NPP). Human interactions affect the ordering of dominant accident sequences and hence have a significant effect on the risk of NPP. By virtue of the ability to combine the treatment of both human and hardware reliability in real accidents, NPP fullscope, multifunctional and computer-based simulators provide a unique way of developing an understanding of the importance of specific human actions for overall plant safety. Context dependent human reliability assessment (HRA) models, such as the holistic decision tree (HDT) and performance evaluation of teamwork (PET) methods, are the so-called second generation HRA techniques. The HDT model has been used for a number of PSA studies. The PET method reflects promising prospects for dealing with dynamic aspects of human performance. The paper presents a comparison of the two HRA techniques for calculation of post-accident human error probability in the PSA. The real and simulated event training scenario 'turbine's stop after loss of feedwater' based on standard PSA model assumptions is designed for WWER-1000 computer simulator and their detailed boundary conditions are described and analyzed. The error probability of post-accident individual actions will be calculated by means of each investigated technique based on student's computer simulator training archives

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

Transport and release of fission products during nuclear reactor accident

International Nuclear Information System (INIS)

Lee, K.W.; Kuhlman, M.R.; Gieseke, J.A.

1984-01-01

This study represents the identification and formulation of a systematic, mechanistic approach to estimating source terms and the implementation of this approach through calculations of fission products release to the environment for a large PWR reactor under a selected set of accident conditions. The development and improvement of calculational procedures is an evolutionary process and in the long term must be verified through experimental studies. It is anticipated that as additional information is obtained the accuracy of predictions can be improved and uncertainties reduced. Transport and deposition of radionuclides were found to be quite dependent on the accident sequences and the corresponding thremal hydraulic conditions. Reduced temperatures led to increased deposition of vapor species, and reduced flow rates to increased aerosol deposition. It is to be recognized that the estimates of release fractions are subject to uncertainties in the data and computer models employed in the calculations and are expected to have been influenced by assumptions regarding plant geometry, thermal hydraulics, deposition mechanisms, and sequence events. The effects of these assumptions will be investigated as this study continues. (Author)

JERICHO computer code: PWR containment response during severe accidents description and sensitivity analysis

International Nuclear Information System (INIS)

Fermandjian, J.; Evrard, J.M.

1983-12-01

The JERICHO code has been developed in order to study the thermodynamic behaviour inside the reactor containment building for the complete spectrum of accident sequences likely to occur in such a reactor, including models for the various mass and energy transfer phenomena, for water spray, for hydrogen and carbon monoxide flammability limits and combustion, as well as for containment venting. Sensitivity analyses have been performed on a severe accident sequence, (namely, small LOCA with failure of the emergency core cooling and containment spray systems), involving core melting and subsequent concrete containment basemat erosion. The effect of various models, such as mass and energy transfer to the structures, has been studied. The influence of the concrete composition, of the fission product deposition and of the thermal degradation of the reactor cavity concrete walls on long term thermodynamic behaviour has also been investigated

Reactor safety study. An assessment of accident risks in U.S. commercial nuclear power plants. Appendix XI. Analysis of comments on the draft WASH-1400 report

International Nuclear Information System (INIS)

1975-10-01

Information is presented concerning comments on reactor safety by governmental agencies and civilian organizations; reactor safety study methodology; consequence model; probability of accident sequences; and various accident conditions

Treatment of whole-body radiation accident victims

International Nuclear Information System (INIS)

Drum, D.E.; Rappeport, J.M.

1990-01-01

This paper discusses how whole-body radiation exposure incidents present a number of unique challenges. The acute, nonstochastic effects of high doses of radiation over 25 rads (0.25 Gy) delivered to humans is generally manifest in rather categorical fashion; depending on the dose, either the patient is largely unharmed functionally or he is seriously injured. Radiation initiates microchemical changes within a nanosecond time frame; there exists no specific therapy to stop or reverse the sequence of events that follow. Thus, the range for effective therapeutic intervention is rather small, between 150 to 1500 rad (1.5 to 15 Gy) for humans. Nevertheless, it is likely that a large uncomplicated exposure to as much as 750 rad (7.5 Gy) might be survivable without dramatic measures such as bone marrow transplantation. Review of the available information about past accidents shows that the majority of radiation accidents are mixed injuries

SCENARIO OF AN ACCIDENT OF SOIL DAMS IN CASE OF WATER SPILL OVER A DAM CREST BY USING FAULT TREE ANALYSIS

OpenAIRE

Kuznetsov Dmitriy Viktorovich

2016-01-01

The scenario of a hydrodynamic accident of water flow over a crest of a soil dam is considered by the method of fault tree analysis, for which the basic reasons and controlled diagnostic indicators of an accident have been defined. Logical operators “AND”/”OR” were used for creation of a sequence of logically connected events, leading to an undesired event in the scenario of accident. The scenario of the accident was plotted in case of three basic reasons - an excessive settling of a dam cres...

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

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)

Large LOCA accident analysis for AP1000 under earthquake

International Nuclear Information System (INIS)

Yu, Yu; Lv, Xuefeng; Niu, Fenglei

2015-01-01

Highlights: • Seismic failure event probability is induced by uncertainties in PGA and in Am. • Uncertainty in PGA is shared by all the components at the same place. • Relativity induced by sharing PGA value can be analyzed explicitly by MC method. • Multi components failures and accident sequences will occur under high PGA value. - Abstract: Seismic probabilistic safety assessment (PSA) is developed to give the insight of nuclear power plant risk under earthquake and the main contributors to the risk. However, component failure probability including the initial event frequency is the function of peak ground acceleration (PGA), and all the components especially the different kinds of components at same place will share the common ground shaking, which is one of the important factors to influence the result. In this paper, we propose an analysis method based on Monte Carlo (MC) simulation in which the effect of all components sharing the same PGA level can be expressed by explicit pattern. The Large LOCA accident in AP1000 is analyzed as an example, based on the seismic hazard curve used in this paper, the core damage frequency is almost equal to the initial event frequency, moreover the frequency of each accident sequence is close to and even equal to the initial event frequency, while the main contributors are seismic events since multi components and systems failures will happen simultaneously when a high value of PGA is sampled. The component failure probability is determined by uncertainties in PGA and in component seismic capacity, and the former is the crucial element to influence the result

Potential applications of advanced information technology in emergency management

International Nuclear Information System (INIS)

Andersson, H.; Holmstrom, C.

1987-01-01

Within nuclear-, offshore- and petrochemical industries there is always a potential risk for severe incidents and accidents. It is a commonly shared belief that timely and correct decisions in these situations could either prevent an incident to develop into a severe accident or to mitigate the negative consequences of an accident. It is also a common belief that in those cases where poor decisions have been taken it has been because of insufficient access to information and expert knowledge when the decisions were taken. These are the background experiences for the joint Nordic research program on the use of advanced information technology in emergency preparedness organizations. Important initial research tasks in the program have been to identify and specify the needs for advanced information technology applications in emergency preparedness organizations. So far a couple of studies aiming for a needs assessment of advanced information technologies in nuclear power emergency preparedness organizations in Sweden and Finland have been completed. The conclusions from these studies are presented in this paper

Severe accident phenomena

International Nuclear Information System (INIS)

Jokiniemi, J.; Kilpi, K.; Lindholm, I.; Maekynen, J.; Pekkarinen, E.; Sairanen, R.; Silde, A.

1995-02-01

Severe accidents are nuclear reactor accidents in which the reactor core is substantially damaged. The report describes severe reactor accident phenomena and their significance for the safety of nuclear power plants. A comprehensive set of phenomena ranging from accident initiation to containment behaviour and containment integrity questions are covered. The report is based on expertise gained in the severe accident assessment projects conducted at the Technical Research Centre of Finland (VTT). (49 refs., 32 figs., 12 tabs.)

Criticality accident of nuclear fuel facility. Think back on JCO criticality accident

International Nuclear Information System (INIS)

Naito, Keiji

2003-09-01

This book is written in order to understand the fundamental knowledge of criticality safety or criticality accident of nuclear fuel facility by the citizens. It consists of four chapters such as critical conditions and criticality accident of nuclear facility, risk of criticality accident, prevention of criticality accident and a measure at an occurrence of criticality accident. A definition of criticality, control of critical conditions, an aspect of accident, a rate of incident, damage, three sufferers, safety control method of criticality, engineering and administrative control, safety design of criticality, investigation of failure of safety control of JCO criticality accident, safety culture are explained. JCO criticality accident was caused with intention of disregarding regulation. It is important that we recognize the correct risk of criticality accident of nuclear fuel facility and prevent disasters. On the basis of them, we should establish safety culture. (S.Y.)

Professional experience and traffic accidents/near-miss accidents among truck drivers.

Science.gov (United States)

Girotto, Edmarlon; Andrade, Selma Maffei de; González, Alberto Durán; Mesas, Arthur Eumann

2016-10-01

To investigate the relationship between the time working as a truck driver and the report of involvement in traffic accidents or near-miss accidents. A cross-sectional study was performed with truck drivers transporting products from the Brazilian grain harvest to the Port of Paranaguá, Paraná, Brazil. The drivers were interviewed regarding sociodemographic characteristics, working conditions, behavior in traffic and involvement in accidents or near-miss accidents in the previous 12 months. Subsequently, the participants answered a self-applied questionnaire on substance use. The time of professional experience as drivers was categorized in tertiles. Statistical analyses were performed through the construction of models adjusted by multinomial regression to assess the relationship between the length of experience as a truck driver and the involvement in accidents or near-miss accidents. This study included 665 male drivers with an average age of 42.2 (±11.1) years. Among them, 7.2% and 41.7% of the drivers reported involvement in accidents and near-miss accidents, respectively. In fully adjusted analysis, the 3rd tertile of professional experience (>22years) was shown to be inversely associated with involvement in accidents (odds ratio [OR] 0.29; 95% confidence interval [CI] 0.16-0.52) and near-miss accidents (OR 0.17; 95% CI 0.05-0.53). The 2nd tertile of professional experience (11-22 years) was inversely associated with involvement in accidents (OR 0.63; 95% CI 0.40-0.98). An evident relationship was observed between longer professional experience and a reduction in reporting involvement in accidents and near-miss accidents, regardless of age, substance use, working conditions and behavior in traffic. Copyright © 2016 Elsevier Ltd. All rights reserved.

CEA studies on advanced nuclear fuel claddings for enhanced accident tolerant LWRs fuel (LOCA and beyond LOCA conditions)

International Nuclear Information System (INIS)

Brachet, J.C.; Lorrette, C.; Michaux, A.; Sauder, C.; Idarraga-Trujillo, I.; Le Saux, M.; Le Flem, M.; Schuster, F.; Billard, A.; Monsifrot, E.; Torres, E.; Rebillat, F.; Bischoff, J.; Ambard, A.

2015-01-01

This paper gives an overview of CEA studies on advanced nuclear fuel claddings for enhanced Accident Tolerant LWR Fuel in collaboration with industrial partners AREVA and EDF. Two potential solutions were investigated: chromium coated zirconium based claddings and SiC/SiC composite claddings with a metallic liner. Concerning the first solution, the optimization of chromium coatings on Zircaloy-4 substrate has been performed. Thus, it has been demonstrated that, due in particular to their slower oxidation rate, a significant additional 'grace period( can be obtained on high temperature oxidized coated claddings in comparison to the conventional uncoated ones, regarding their residual PQ (Post-Quench) ductility and their ability to survive to the final water quenching in LOCA and, to some extent, beyond LOCA conditions. Concerning the second solution, the innovative 'sandwich' SiC/SiC cladding concept is introduced. Initially designed for the next generation of nuclear reactors, it can be adapted to obtain high safety performance for LWRs in LOCA conditions. The key findings of this work highlight the low sensitivity of SiC/SiC composites under the explored steam oxidation conditions. No signification degradation of the mechanical properties of CVI-HNI SiC/SiC specimen is particularly acknowledged for relatively long duration (beyond 100 h at 1200 Celsius degrees). Despite these very positive preliminary results, significant studies and developments are still necessary to close the technology gap. Qualification for nuclear application requires substantial irradiation testing, additional characterization and the definition of design rules applicable to such a structure. The use of a SiC-based fuel cladding shows promise for the highest temperature accident conditions but remains a long term perspective

Preliminary Design of Optimized Reactor Insulator for Severe Accident Mitigation of APR1400

International Nuclear Information System (INIS)

Heo, Sun; Lee, Jae-Gon; Kang, Yong-Chul

2007-01-01

APR1400, a Korean evolutionary advance light water reactor, has many advanced safety feature to prevent and mitigate of design basis accident (DBA) and severe accident. When reactor cooling system (RCS) fails to cooling its core, the core melted down and the molten core gathers together on bottom of reactor vessel. The molten core hurts reactor vessel and is released to containment, which raises the release of radioactive isotopes and the heating of the containment atmosphere. Finally, the corium is accumulated in the bottom of reactor cavity and it also raises the Molten Core and Concrete Interaction (MCCI) and the heating of containment atmosphere. There are two strategies to cooling molten core. Those are in-vessel retention and ex-vessel cooling. At the early stage of APR1400 design, only ex-vessel cooling which is cooling of the molten core outside the vessel after vessel failure is considered based on EPRI Utility Requirement Document (URD) for Evolutionary LWR. However, a need has been arisen to reflect current research findings on severe accident phenomena and mitigation technologies to Korean URD and IVRERVC (In-Vessel corium Retention using Ex-Reactor Vessel Cooling) was adopted APR1400. The ERVC is not considered as a licensing design basis but based on the defense-in-depth principle and safety margin basis, which is the top-tier requirement of the severe accident mitigation design as stated in the KURD. The Severe Accident Management strategy for APR1400 is intended to aid the plant operating staff to secure reactor vessel integrity in the early stage of the severe accident. As a part of a design implementation of IVR-ERVC for APR1400, we developed the preliminary design requirement, design specification and conceptual design

Study of event sequence database for a nuclear power domain

International Nuclear Information System (INIS)

Kusumi, Yoshiaki

1998-01-01

A retrieval engine developed to extract event sequences from an accident information database using a time series retrieval formula expressed with ordered retrieval terms is explored. This engine outputs not only a sequence which completely matches with a time series retrieval formula, but also sequence which approximately matches the formula (fuzzy retrieval). An event sequence database in which records consist of three ordered parameters, namely the causal event, the process and result. Then the database is used to assess the feasibility of this engine and favorable results were obtained. (author)

OSSA. A second generation of severe accident management

International Nuclear Information System (INIS)

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

2009-01-01

diagnostic directs the users to a series of system guidelines. The system approach allows a quick and general overview of the plant status. Following the course of an accident scenario evolving to a severe accident, the reader will be taken through the advanced concepts that are developed within the frame of the OSSA. Step by step, in place of the operators or the technical support group, he will experiment the OSSA. The selected severe accident scenario presented is a station blackout during which the late recovery will lead to severe accident. (author)

Assessment of PASS Effectiveness under Severe Accidents in Nuclear Power Plants

International Nuclear Information System (INIS)

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

2008-01-01

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

Analysis of media coverage and KINS communication activities on Fukushima accident

International Nuclear Information System (INIS)

Lee, Ki Hyung; Hwang, Sun Chul; Yun, Yuen Wha; Lee, Gye Hwi; Jeong, Jin A; Song, Hye Rim; Yang, Cho Hee

2012-01-01

The people and mass media of Korea, the closest country to Japan, showed great interest in Fukushima nuclear power plant accident. The Korean government and KINS (Korea Institute of Nuclear Safety) attempted to provide accurate information to the press through various communication actions. In this study, we conducted an in-depth analysis of the tendencies of the press according to the accident sequence and tracked the diffusion of this issue. The purpose of this study is to determine the properties of the crisis and essence of the issue. We also carry out a general evaluation and draw implications through an analysis of the communication actions of KINS

Radiological impact of a loss of coolant accident at Angra 2 reactor

International Nuclear Information System (INIS)

Dias, W.

1992-01-01

A loss of coolant accident is analyzed which comprises a double ended rupture of a main primary system line. The accident sequence is described and the main assumptions as to the activity release are presented. On the basis of site specific meteorological data, the atmospheric dispersion factors are calculated using the Gaussian plume diffusion model and the doses are then determined at the boundary of the low population zone. The resulting values for the effective dose equivalent are more than one order of magnitude below that due to the average background radiation received in one year. (author)

Prevention of pedestrian accidents.

OpenAIRE

Kendrick, D

1993-01-01

Child pedestrian accidents are the most common road traffic accident resulting in injury. Much of the existing work on road traffic accidents is based on analysing clusters of accidents despite evidence that child pedestrian accidents tend to be more dispersed than this. This paper analyses pedestrian accidents in 573 children aged 0-11 years by a locally derived deprivation score for the years 1988-90. The analysis shows a significantly higher accident rate in deprived areas and a dose respo...

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

International Nuclear Information System (INIS)

Jae, M.; Moon, J.H.

2002-01-01

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

Development of a prototype graphic simulation program for severe accident training

Energy Technology Data Exchange (ETDEWEB)

Kim, Ko Ryu; Jeong, Kwang Sub; Ha, Jae Joo

2000-05-01

This is a report of the development process and related technologies of severe accident graphic simulators, required in industrial severe accident management and training. Here, we say 'a severe accident graphic simulator' as a graphics add-in system to existing calculation codes, which can show the severe accident phenomena dynamically on computer screens and therefore which can supplement one of main defects of existing calculation codes. With graphic simulators it is fairly easy to see the total behavior of nuclear power plants, where it was very difficult to see only from partial variable numerical information. Moreover, the fast processing and control feature of a graphic simulator can give some opportunities of predicting the severe accident advancement among several possibilities, to one who is not an expert. Utilizing graphic simulators' we expect operators' and TSC members' physical phenomena understanding enhancement from the realistic dynamic behavior of plants. We also expect that severe accident training course can gain better training effects using graphic simulator's control functions and predicting capabilities, and therefore we expect that graphic simulators will be effective decision-aids tools both in sever accident training course and in real severe accident situations. With these in mind, we have developed a prototype graphic simulator having surveyed related technologies, and from this development experiences we have inspected the possibility to build a severe accident graphic simulator. The prototype graphic simulator is developed under IBM PC WinNT environments and is suited to Uljin 3and4 nuclear power plant. When supplied with adequate severe accident scenario as an input, the prototype can provide graphical simulations of plant safety systems' dynamic behaviors. The prototype is composed of several different modules, which are phenomena display module, MELCOR data interface module and graphic database