A new approach to modeling aviation accidents

NASA Astrophysics Data System (ADS)

Rao, Arjun Harsha

General Aviation (GA) is a catchall term for all aircraft operations in the US that are not categorized as commercial operations or military flights. GA aircraft account for almost 97% of the US civil aviation fleet. Unfortunately, GA flights have a much higher fatal accident rate than commercial operations. Recent estimates by the Federal Aviation Administration (FAA) showed that the GA fatal accident rate has remained relatively unchanged between 2010 and 2015, with 1566 fatal accidents accounting for 2650 fatalities. Several research efforts have been directed towards betters understanding the causes of GA accidents. Many of these efforts use National Transportation Safety Board (NTSB) accident reports and data. Unfortunately, while these studies easily identify the top types of accidents (e.g., inflight loss of control (LOC)), they usually cannot identify why these accidents are happening. Most NTSB narrative reports for GA accidents are very short (many are only one paragraph long), and do not contain much information on the causes (likely because the causes were not fully identified). NTSB investigators also code each accident using an event-based coding system, which should facilitate identification of patterns and trends in causation, given the high number of GA accidents each year. However, this system is susceptible to investigator interpretation and error, meaning that two investigators may code the same accident differently, or omit applicable codes. To facilitate a potentially better understanding of GA accident causation, this research develops a state-based approach to check for logical gaps or omissions in NTSB accident records, and potentially fills-in the omissions. The state-based approach offers more flexibility as it moves away from the conventional event-based representation of accidents, which classifies events in accidents into several categories such as causes, contributing factors, findings, occurrences, and phase of flight. The method views aviation accidents as a set of hazardous states of a system (pilot and aircraft), and triggers that cause the system to move between hazardous states. I used the NTSB's accident coding manual (that contains nearly 4000 different codes) to develop a "dictionary" of hazardous states, triggers, and information codes. Then, I created the "grammar", or a set of rules, that: (1) orders the hazardous states in each accident; and, (2) links the hazardous states using the appropriate triggers. This approach: (1) provides a more correct count of the causes for accidents in the NTSB database; and, (2) checks for gaps or omissions in NTSB accident data, and fills in some of these gaps using logic-based rules. These rules also help identify and count causes for accidents that were not discernable from previous analyses of historical accident data. I apply the model to 6200 helicopter accidents that occurred in the US between 1982 and 2015. First, I identify the states and triggers that are most likely to be associated with fatal and non-fatal accidents. The results suggest that non-fatal accidents, which account for approximately 84% of the accidents, provide valuable opportunities to learn about the causes for accidents. Next, I investigate the causes of inflight loss of control using both a conventional approach and using the state-based approach. The conventional analysis provides little insight into the causal mechanism for LOC. For instance, the top cause of LOC is "aircraft control/directional control not maintained", which does not provide any insight. In contrast, the state-based analysis showed that pilots' tendency to clip objects frequently triggered LOC (16.7% of LOC accidents)--this finding was not directly discernable from conventional analyses. Finally, I investigate the causes for improper autorotations using both a conventional approach and the state-based approach. The conventional approach uses modifiers (e.g., "improper", "misjudged") associated with "24520: Autorotation" to identify improper autorotations in the pre-2008 system. In the psot-2008 system, the NTSB represents autorotation as a phase of flight, which has no modifier--making it impossible to determine if the autorotation was unsuccessful. In contrast, the state-based analysis identified 632 improper autorotation accidents, compared to 174 with a conventional analysis. Results from the state-based analysis show that not maintaining rotor RPM and improper flare were among the top reasons for improper autorotations. The presence of the "not possible" trigger in 11.6% of improper autorotations, suggests that it was impossible to make an autorotative landing. Improper use of collective is the sixth most frequent trigger for improper autorotation. Correct use of collective pitch control is crucial to maintain rotor RPM during an autorotation (considering that engines are generally not operational during autorotations).

A methodology for accident analysis of fusion breeder blankets and its application to helium-cooled lead–lithium blanket

DOE PAGES

Panayotov, Dobromir; Poitevin, Yves; Grief, Andrew; ...

2016-09-23

'Fusion for Energy' (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials,more » and phenomena while remaining consistent with the approach already applied to ITER accident analyses. Furthermore, the methodology phases are illustrated in the paper by its application to the EU HCLL TBS using both MELCOR and RELAP5 codes.« less

25 CFR 11.445 - Driving violations.

Code of Federal Regulations, 2010 CFR

2010-04-01

... violation of this section, or has been involved in a motor vehicle accident or collision resulting in property damage, personal injury, or death. (d) In the absence of an applicable tribal traffic code, the provisions of state traffic laws applicable in the state where a Court of Indian Offenses is located shall...

The kinetics of aerosol particle formation and removal in NPP severe accidents

NASA Astrophysics Data System (ADS)

Zatevakhin, Mikhail A.; Arefiev, Valentin K.; Semashko, Sergey E.; Dolganov, Rostislav A.

2016-06-01

Severe Nuclear Power Plant (NPP) accidents are accompanied by release of a massive amount of energy, radioactive products and hydrogen into the atmosphere of the NPP containment. A valid estimation of consequences of such accidents can only be carried out through the use of the integrated codes comprising a description of the basic processes which determine the consequences. A brief description of a coupled aerosol and thermal-hydraulic code to be used for the calculation of the aerosol kinetics within the NPP containment in case of a severe accident is given. The code comprises a KIN aerosol unit integrated into the KUPOL-M thermal-hydraulic code. Some features of aerosol behavior in severe NPP accidents are briefly described.

The kinetics of aerosol particle formation and removal in NPP severe accidents

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zatevakhin, Mikhail A.; Arefiev, Valentin K.; Semashko, Sergey E.

2016-06-08

Severe Nuclear Power Plant (NPP) accidents are accompanied by release of a massive amount of energy, radioactive products and hydrogen into the atmosphere of the NPP containment. A valid estimation of consequences of such accidents can only be carried out through the use of the integrated codes comprising a description of the basic processes which determine the consequences. A brief description of a coupled aerosol and thermal–hydraulic code to be used for the calculation of the aerosol kinetics within the NPP containment in case of a severe accident is given. The code comprises a KIN aerosol unit integrated into themore » KUPOL-M thermal–hydraulic code. Some features of aerosol behavior in severe NPP accidents are briefly described.« less

MELCOR computer code manuals

DOE Office of Scientific and Technical Information (OSTI.GOV)

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

Methodology, status and plans for development and assessment of the code ATHLET

DOE Office of Scientific and Technical Information (OSTI.GOV)

Teschendorff, V.; Austregesilo, H.; Lerchl, G.

1997-07-01

The thermal-hydraulic computer code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is being developed by the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) for the analysis of anticipated and abnormal plant transients, small and intermediate leaks as well as large breaks in light water reactors. The aim of the code development is to cover the whole spectrum of design basis and beyond design basis accidents (without core degradation) for PWRs and BWRs with only one code. The main code features are: advanced thermal-hydraulics; modular code architecture; separation between physical models and numerical methods; pre- and post-processing tools; portability. The codemore » has features that are of special interest for applications to small leaks and transients with accident management, e.g. initialization by a steady-state calculation, full-range drift-flux model, 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 code development is accompained by a systematic and comprehensive validation program. A large number of integral experiments and separate effect tests, including the major International Standard Problems, have been calculated by GRS and by independent organizations. The ATHLET validation matrix is a well balanced set of integral and separate effects tests derived from the CSNI proposal emphasizing, however, the German combined ECC injection system which was investigated in the UPTF, PKL and LOBI test facilities.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

White, J.E.; Roussin, R.W.; Gilpin, H.

A version of the CRAC2 computer code applicable for use in analyses of consequences and risks of reactor accidents in case work for environmental statements has been implemented for use on the Nuclear Regulatory Commission Data General MV/8000 computer system. Input preparation is facilitated through the use of an interactive computer program which operates on an IBM personal computer. The resulting CRAC2 input deck is transmitted to the MV/8000 by using an error-free file transfer mechanism. To facilitate the use of CRAC2 at NRC, relevant background material on input requirements and model descriptions has been extracted from four reports -more » ''Calculations of Reactor Accident Consequences,'' Version 2, NUREG/CR-2326 (SAND81-1994) and ''CRAC2 Model Descriptions,'' NUREG/CR-2552 (SAND82-0342), ''CRAC Calculations for Accident Sections of Environmental Statements, '' NUREG/CR-2901 (SAND82-1693), and ''Sensitivity and Uncertainty Studies of the CRAC2 Computer Code,'' NUREG/CR-4038 (ORNL-6114). When this background information is combined with instructions on the input processor, this report provides a self-contained guide for preparing CRAC2 input data with a specific orientation toward applications on the MV/8000. 8 refs., 11 figs., 10 tabs.« less

Computers in Traffic Education.

ERIC Educational Resources Information Center

Alexander, O. P.

1983-01-01

Traffic education covers basic road skills, legal/insurance aspects, highway code, accident causation/prevention, and vehicle maintenance. Microcomputer applications to traffic education are outlined, followed by a selected example of programs currently available (focusing on drill/practice, simulation, problem-solving, data manipulation, games,…

Development and application of the dynamic system doctor to nuclear reactor probabilistic risk assessments.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kunsman, David Marvin; Aldemir, Tunc; Rutt, Benjamin

2008-05-01

This LDRD project has produced a tool that makes probabilistic risk assessments (PRAs) of nuclear reactors - analyses which are very resource intensive - more efficient. PRAs of nuclear reactors are being increasingly relied on by the United States Nuclear Regulatory Commission (U.S.N.R.C.) for licensing decisions for current and advanced reactors. Yet, PRAs are produced much as they were 20 years ago. The work here applied a modern systems analysis technique to the accident progression analysis portion of the PRA; the technique was a system-independent multi-task computer driver routine. Initially, the objective of the work was to fuse the accidentmore » progression event tree (APET) portion of a PRA to the dynamic system doctor (DSD) created by Ohio State University. Instead, during the initial efforts, it was found that the DSD could be linked directly to a detailed accident progression phenomenological simulation code - the type on which APET construction and analysis relies, albeit indirectly - and thereby directly create and analyze the APET. The expanded DSD computational architecture and infrastructure that was created during this effort is called ADAPT (Analysis of Dynamic Accident Progression Trees). ADAPT is a system software infrastructure that supports execution and analysis of multiple dynamic event-tree simulations on distributed environments. A simulator abstraction layer was developed, and a generic driver was implemented for executing simulators on a distributed environment. As a demonstration of the use of the methodological tool, ADAPT was applied to quantify the likelihood of competing accident progression pathways occurring for a particular accident scenario in a particular reactor type using MELCOR, an integrated severe accident analysis code developed at Sandia. (ADAPT was intentionally created with flexibility, however, and is not limited to interacting with only one code. With minor coding changes to input files, ADAPT can be linked to other such codes.) The results of this demonstration indicate that the approach can significantly reduce the resources required for Level 2 PRAs. From the phenomenological viewpoint, ADAPT can also treat the associated epistemic and aleatory uncertainties. This methodology can also be used for analyses of other complex systems. Any complex system can be analyzed using ADAPT if the workings of that system can be displayed as an event tree, there is a computer code that simulates how those events could progress, and that simulator code has switches to turn on and off system events, phenomena, etc. Using and applying ADAPT to particular problems is not human independent. While the human resources for the creation and analysis of the accident progression are significantly decreased, knowledgeable analysts are still necessary for a given project to apply ADAPT successfully. This research and development effort has met its original goals and then exceeded them.« less

Insights Gained from Forensic Analysis with MELCOR of the Fukushima-Daiichi Accidents.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andrews, Nathan C.; Gauntt, Randall O.

Since the accidents at Fukushima-Daiichi, Sandia National Laboratories has been modeling these accident scenarios using the severe accident analysis code, MELCOR. MELCOR is a widely used computer code developed at Sandia National Laboratories since ~1982 for the U.S. Nuclear Regulatory Commission. Insights from the modeling of these accidents is being used to better inform future code development and potentially improved accident management. To date, our necessity to better capture in-vessel thermal-hydraulic and ex-vessel melt coolability and concrete interactions has led to the implementation of new models. The most recent analyses, presented in this paper, have been in support of themore » of the Organization for Economic Cooperation and Development Nuclear Energy Agency’s (OECD/NEA) Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF) Project. The goal of this project is to accurately capture the source term from all three releases and then model the atmospheric dispersion. In order to do this, a forensic approach is being used in which available plant data and release timings is being used to inform the modeled MELCOR accident scenario. For example, containment failures, core slumping events and lower head failure timings are all enforced parameters in these analyses. This approach is fundamentally different from a blind code assessment analysis often used in standard problem exercises. The timings of these events are informed by representative spikes or decreases in plant data. The combination of improvements to the MELCOR source code resulting from analysis previous accident analysis and this forensic approach has allowed Sandia to generate representative and plausible source terms for all three accidents at Fukushima Daiichi out to three weeks after the accident to capture both early and late releases. In particular, using the source terms developed by MELCOR, the MACCS software code, which models atmospheric dispersion and deposition, we are able to reasonably capture the deposition of radionuclides to the northwest of the reactor site.« less

Air oxidation of Zircaloy-4 in the 600-1000 °C temperature range: Modeling for ASTEC code application

NASA Astrophysics Data System (ADS)

Coindreau, O.; Duriez, C.; Ederli, S.

2010-10-01

Progress in the treatment of air oxidation of zirconium in severe accident (SA) codes are required for a reliable analysis of severe accidents involving air ingress. Air oxidation of zirconium can actually lead to accelerated core degradation and increased fission product release, especially for the highly-radiotoxic ruthenium. This paper presents a model to simulate air oxidation kinetics of Zircaloy-4 in the 600-1000 °C temperature range. It is based on available experimental data, including separate-effect experiments performed at IRSN and at Forschungszentrum Karlsruhe. The kinetic transition, named "breakaway", from a diffusion-controlled regime to an accelerated oxidation is taken into account in the modeling via a critical mass gain parameter. The progressive propagation of the locally initiated breakaway is modeled by a linear increase in oxidation rate with time. Finally, when breakaway propagation is completed, the oxidation rate stabilizes and the kinetics is modeled by a linear law. This new modeling is integrated in the severe accident code ASTEC, jointly developed by IRSN and GRS. Model predictions and experimental data from thermogravimetric results show good agreement for different air flow rates and for slow temperature transient conditions.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

2002-07-01

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 enablemore » 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)« less

MELCOR computer code manuals: Primer and user`s guides, Version 1.8.3 September 1994. Volume 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the US Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users` Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

Methods for nuclear air-cleaning-system accident-consequence assessment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andrae, R.W.; Bolstad, J.W.; Gregory, W.S.

1982-01-01

This paper describes a multilaboratory research program that is directed toward addressing many questions that analysts face when performing air cleaning accident consequence assessments. The program involves developing analytical tools and supportive experimental data that will be useful in making more realistic assessments of accident source terms within and up to the atmospheric boundaries of nuclear fuel cycle facilities. The types of accidents considered in this study includes fires, explosions, spills, tornadoes, criticalities, and equipment failures. The main focus of the program is developing an accident analysis handbook (AAH). We will describe the contents of the AAH, which include descriptionsmore » of selected nuclear fuel cycle facilities, process unit operations, source-term development, and accident consequence analyses. Three computer codes designed to predict gas and material propagation through facility air cleaning systems are described. These computer codes address accidents involving fires (FIRAC), explosions (EXPAC), and tornadoes (TORAC). The handbook relies on many illustrative examples to show the analyst how to approach accident consequence assessments. We will use the FIRAC code and a hypothetical fire scenario to illustrate the accident analysis capability.« less

Nuclear Fuels & Materials Spotlight Volume 5

DOE Office of Scientific and Technical Information (OSTI.GOV)

Petti, David Andrew

2016-10-01

As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • Evaluation and modeling of light water reactor accident tolerant fuel concepts • Status and results of recent TRISO-coated particle fuel irradiations, post-irradiation examinations, high-temperature safety testing to demonstrate the accident performance of this fuel system, and advanced microscopy to improve the understanding of fission product transport in this fuel system.more » • Improvements in and applications of meso and engineering scale modeling of light water reactor fuel behavior under a range of operating conditions and postulated accidents (e.g., power ramping, loss of coolant accident, and reactivity initiated accidents) using the MARMOT and BISON codes. • Novel measurements of the properties of nuclear (actinide) materials under extreme conditions, (e.g. high pressure, low/high temperatures, high magnetic field) to improve the scientific understanding of these materials. • Modeling reactor pressure vessel behavior using the GRIZZLY code. • New methods using sound to sense temperature inside a reactor core. • Improved experimental capabilities to study the response of fusion reactor materials to a tritium plasma. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at Idaho National Laboratory, and hope that you find this issue informative.« less

[Hospital information system performance for road traffic accidents analysis in a hospital recruitment based area].

PubMed

Jannot, A-S; Fauconnier, J

2013-06-01

Road traffic accidents in France are mainly analyzed through reports completed by the security forces (police and gendarmerie). But the hospital information systems can also identify road traffic accidents via specific documentary codes of the International Classification of Diseases (ICD-10). The aim of this study was therefore to determine whether hospital stays consecutive to road traffic accident were truly identified by these documentary codes in a facility that collects data routinely and to study the consistency of results from hospital information systems and from security forces during the 2002-2008 period. We retrieved all patients for whom a documentary code for road traffic accident was entered in 2002-2008. We manually checked the concordance of documentary code for road traffic accident and trauma origin in 350 patient files. The number of accidents in the Grenoble area was then inferred by combining with hospitalization regional data and compared to the number of persons injured by traffic accidents declared by the security force. These hospital information systems successfully report road traffic accidents with 96% sensitivity (95%CI: [92%, 100%]) and 97% specificity (95%CI: [95%, 99%]). The decrease in road traffic accidents observed was significantly less than that observed was significantly lower than that observed in the data from the security force (45% for security force data against 27% for hospital data). Overall, this study shows that hospital information systems are a powerful tool for studying road traffic accidents morbidity in hospital and are complementary to security force data. Copyright © 2013 Elsevier Masson SAS. All rights reserved.

Current and anticipated use of thermal-hydraulic codes for BWR transient and accident analyses in Japan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arai, Kenji; Ebata, Shigeo

1997-07-01

This paper summarizes the current and anticipated use of the thermal-hydraulic and neutronic codes for the BWR transient and accident analyses in Japan. The codes may be categorized into the licensing codes and the best estimate codes for the BWR transient and accident analyses. Most of the licensing codes have been originally developed by General Electric. Some codes have been updated based on the technical knowledge obtained in the thermal hydraulic study in Japan, and according to the BWR design changes. The best estimates codes have been used to support the licensing calculations and to obtain the phenomenological understanding ofmore » the thermal hydraulic phenomena during a BWR transient or accident. The best estimate codes can be also applied to a design study for a next generation BWR to which the current licensing model may not be directly applied. In order to rationalize the margin included in the current BWR design and develop a next generation reactor with appropriate design margin, it will be required to improve the accuracy of the thermal-hydraulic and neutronic model. In addition, regarding the current best estimate codes, the improvement in the user interface and the numerics will be needed.« less

Modeling of local steam condensation on walls in presence of non-condensable gases. Application to a loca calculation in reactor containment using the multidimensional geyser/tonus code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benet, L.V.; Caroli, C.; Cornet, P.

1995-09-01

This paper reports part of a study of possible severe pressurized water reactor (PWR) accidents. The need for containment modeling, and in particular for a hydrogen risk study, was reinforced in France after 1990, with the requirement that severe accidents must be taken into account in the design of future plants. This new need of assessing the transient local hydrogen concentration led to the development, in the Mechanical Engineering and Technology Department of the French Atomic Energy Commission (CEA/DMT), of the multidimensional code GEYSER/TONUS for containment analysis. A detailed example of the use of this code is presented. The mixturemore » consisted of noncondensable gases (air or air plus hydrogen) and water vapor and liquid water. This is described by a compressible homogeneous two-phase flow model and wall condensation is based on the Chilton-Colburn formula and the analogy between heat and mass transfer. Results are given for a transient two-dimensional axially-symmetric computation for the first hour of a simplified accident sequence. In this there was an initial injection of a large amount of water vapor followed by a smaller amount and by hydrogen injection.« less

The Initial Atmospheric Transport (IAT) Code: Description and Validation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Morrow, Charles W.; Bartel, Timothy James

The Initial Atmospheric Transport (IAT) computer code was developed at Sandia National Laboratories as part of their nuclear launch accident consequences analysis suite of computer codes. The purpose of IAT is to predict the initial puff/plume rise resulting from either a solid rocket propellant or liquid rocket fuel fire. The code generates initial conditions for subsequent atmospheric transport calculations. The Initial Atmospheric Transfer (IAT) code has been compared to two data sets which are appropriate to the design space of space launch accident analyses. The primary model uncertainties are the entrainment coefficients for the extended Taylor model. The Titan 34Dmore » accident (1986) was used to calibrate these entrainment settings for a prototypic liquid propellant accident while the recent Johns Hopkins University Applied Physics Laboratory (JHU/APL, or simply APL) large propellant block tests (2012) were used to calibrate the entrainment settings for prototypic solid propellant accidents. North American Meteorology (NAM )formatted weather data profiles are used by IAT to determine the local buoyancy force balance. The IAT comparisons for the APL solid propellant tests illustrate the sensitivity of the plume elevation to the weather profiles; that is, the weather profile is a dominant factor in determining the plume elevation. The IAT code performed remarkably well and is considered validated for neutral weather conditions.« less

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

NASA Astrophysics Data System (ADS)

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.

Insight from Fukushima Daiichi Unit 3 Investigations using MELCOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robb, Kevin R.; Francis, Matthew W.; Ott, Larry J.

During the emergency response period of the accidents that took place at Fukushima Daiichi in March of 2011, researchers at Oak Ridge National Laboratory (ORNL) conducted a number of studies using the MELCOR code to help understand what was occurring and what had occurred. During the post-accident period, the Department of Energy (DOE) and the US Nuclear Regulatory Commission (NRC) jointly sponsored a study of the Fukushima Daiichi accident with collaboration among Oak Ridge, Sandia, and Idaho national laboratories. The purpose of the study was to compile relevant data, reconstruct the accident progression using computer codes, assess the codes predictivemore » capabilities, and identify future data needs. The current paper summarizes some of the early MELCOR simulations and analyses conducted at ORNL of the Fukushima Daiichi Unit 3 accident. Extended analysis and discussion of the Unit 3 accident is also presented taking into account new knowledge and modeling refinements made since the joint DOE/NRC study.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Panayotov, Dobromir; Poitevin, Yves; Grief, Andrew

'Fusion for Energy' (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials,more » and phenomena while remaining consistent with the approach already applied to ITER accident analyses. Furthermore, the methodology phases are illustrated in the paper by its application to the EU HCLL TBS using both MELCOR and RELAP5 codes.« less

System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

NASA Astrophysics Data System (ADS)

Porter, Ian Edward

A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several additional fuels will also be analyzed, including uranium nitride (UN), uranium carbide (UC) and uranium silicide (U3Si2). Focusing on the system response in an accident scenario, an emphasis is placed on the fracture mechanics of the ceramic cladding by design the fuel rods to eliminate pellet cladding mechanical interaction (PCMI). The time to failure and how much of the fuel in the reactor fails with an advanced fuel design will be analyzed and compared to the current UO2/Zircaloy design using a full scale reactor model.

SOPHAEROS code development and its application to falcon tests

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lajtha, G.; Missirlian, M.; Kissane, M.

1996-12-31

One of the key issues in source-term evaluation in nuclear reactor severe accidents is determination of the transport behavior of fission products released from the degrading core. The SOPHAEROS computer code is being developed to predict fission product transport in a mechanistic way in light water reactor circuits. These applications of the SOPHAEROS code to the Falcon experiments, among others not presented here, indicate that the numerical scheme of the code is robust, and no convergence problems are encountered. The calculation is also very fast being three times longer on a Sun SPARC 5 workstation than real time and typicallymore » {approx} 10 times faster than an identical calculation with the VICTORIA code. The study demonstrates that the SOPHAEROS 1.3 code is a suitable tool for prediction of the vapor chemistry and fission product transport with a reasonable level of accuracy. Furthermore, the fexibility of the code material data bank allows improvement of understanding of fission product transport and deposition in the circuit. Performing sensitivity studies with different chemical species or with different properties (saturation pressure, chemical equilibrium constants) is very straightforward.« less

PACER -- A fast running computer code for the calculation of short-term containment/confinement loads following coolant boundary failure. Volume 2: User information

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sienicki, J.J.

A fast running and simple computer code has been developed to calculate pressure loadings inside light water reactor containments/confinements under loss-of-coolant accident conditions. PACER was originally developed to calculate containment/confinement pressure and temperature time histories for loss-of-coolant accidents in Soviet-designed VVER reactors and is relevant to the activities of the US International Nuclear Safety Center. The code employs a multicompartment representation of the containment volume and is focused upon application to early time containment phenomena during and immediately following blowdown. PACER has been developed for FORTRAN 77 and earlier versions of FORTRAN. The code has been successfully compiled and executedmore » on SUN SPARC and Hewlett-Packard HP-735 workstations provided that appropriate compiler options are specified. The code incorporates both capabilities built around a hardwired default generic VVER-440 Model V230 design as well as fairly general user-defined input. However, array dimensions are hardwired and must be changed by modifying the source code if the number of compartments/cells differs from the default number of nine. Detailed input instructions are provided as well as a description of outputs. Input files and selected output are presented for two sample problems run on both HP-735 and SUN SPARC workstations.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

BRISC is a developmental prototype for a nextgeneration “systems-level” integrated performance and safety code (IPSC) for nuclear reactors. Its development served to demonstrate how a lightweight multi-physics coupling approach can be used to tightly couple the physics models in several different physics codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled “burner” nuclear reactor. For example, the RIO Fluid Flow and Heat transfer code developed at Sandia (SNL: Chris Moen, Dept. 08005) is used in BRISC to model fluid flow and heat transfer, as well as conduction heat transfermore » in solids. Because BRISC is a prototype, its most practical application is as a foundation or starting point for developing a true production code. The sub-codes and the associated models and correlations currently employed within BRISC were chosen to cover the required application space and demonstrate feasibility, but were not optimized or validated against experimental data within the context of their use in BRISC.« less

Police accident report forms: safety device coding and enacted laws.

PubMed

Brock, K; Lapidus, G

2008-12-01

Safety device coding on state police accident report (PAR) forms was compared with provisions in state traffic safety laws. PAR forms were obtained from all 50 states and the District of Columbia (states/DC). For seat belts, 22 states/DC had a primary seat belt enforcement law vs 50 with a PAR code. For car seats, all 51 states/DC had a law and a PAR code. For booster seats, 39 states/DC had a law vs nine with a PAR code. For motorcycle helmets, 21 states/DC had an all-age rider helmet law and another 26 a partial-age law vs 50 with a PAR code. For bicycle helmets, 21 states/DC had a partial-age rider helmet law vs 48 with a PAR code. Therefore gaps in the ability of states to fully record accident data reflective of existing state traffic safety laws are revealed. Revising the PAR forms in all states to include complete variables for safety devices should be an important priority, independent of the laws.

Circumstances of Trauma and Accidents in Children: A Thesaurus-based Survey

PubMed

Séjourné, Claire; Philbois, Olivier; Vercherin, Paul; Patural, Hugues

2016-11-25

Introduction : Injuries and accidents are major causes of morbidity and mortality in children in France. Identification and description of the mechanisms of accidents are essential to develop adapted prevention methods. For this purpose, a specific thesaurus of ICD-10 codes relating to the circumstances of trauma and accidents in children was created in the French Loire department. The objective of this study was to evaluate the relevance and acceptability of the thesaurus in the pediatric emergency unit of Saint-Etienne university hospital.Material and Methods : This study was conducted in two phases. The first, longitudinal phase was conducted over three periods between May and October 2014 to compare codings by emergency room physicians before using the thesaurus with those defined on the basis of the thesaurus. The second phase retrospectively compared coding in July and August 2014 before introduction of the thesaurus with thesaurus-based coding in July and August 2015.Results : The first phase showed a loss of more than half of the information without the thesaurus. The circumstances of trauma can be described by an appropriate code in more than 90% of cases. The second phase showed a 13% increase in coding of the circumstances of trauma, which nevertheless remains insufficient.Discussion : The thesaurus facilitates coding and generally meets the coding physician’s expectations and should be used in large-scale epidemiological surveys.

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

DOE PAGES

Brown, Cameron S.; Zhang, Hongbin; Kucukboyaci, Vefa; ...

2016-09-07

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by amore » system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.« less

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Cameron S.; Zhang, Hongbin; Kucukboyaci, Vefa

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by amore » system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.« less

The development and evaluation of accident predictive models

NASA Astrophysics Data System (ADS)

Maleck, T. L.

1980-12-01

A mathematical model that will predict the incremental change in the dependent variables (accident types) resulting from changes in the independent variables is developed. The end product is a tool for estimating the expected number and type of accidents for a given highway segment. The data segments (accidents) are separated in exclusive groups via a branching process and variance is further reduced using stepwise multiple regression. The standard error of the estimate is calculated for each model. The dependent variables are the frequency, density, and rate of 18 types of accidents among the independent variables are: district, county, highway geometry, land use, type of zone, speed limit, signal code, type of intersection, number of intersection legs, number of turn lanes, left-turn control, all-red interval, average daily traffic, and outlier code. Models for nonintersectional accidents did not fit nor validate as well as models for intersectional accidents.

Evaluation of severe accident risks: Quantification of major input parameters: MAACS (MELCOR Accident Consequence Code System) input

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sprung, J.L.; Jow, H-N; Rollstin, J.A.

1990-12-01

Estimation of offsite accident consequences is the customary final step in a probabilistic assessment of the risks of severe nuclear reactor accidents. Recently, the Nuclear Regulatory Commission reassessed the risks of severe accidents at five US power reactors (NUREG-1150). Offsite accident consequences for NUREG-1150 source terms were estimated using the MELCOR Accident Consequence Code System (MACCS). Before these calculations were performed, most MACCS input parameters were reviewed, and for each parameter reviewed, a best-estimate value was recommended. This report presents the results of these reviews. Specifically, recommended values and the basis for their selection are presented for MACCS atmospheric andmore » biospheric transport, emergency response, food pathway, and economic input parameters. Dose conversion factors and health effect parameters are not reviewed in this report. 134 refs., 15 figs., 110 tabs.« less

Extracting recurrent scenarios from narrative texts using a Bayesian network: application to serious occupational accidents with movement disturbance.

PubMed

Abdat, F; Leclercq, S; Cuny, X; Tissot, C

2014-09-01

A probabilistic approach has been developed to extract recurrent serious Occupational Accident with Movement Disturbance (OAMD) scenarios from narrative texts within a prevention framework. Relevant data extracted from 143 accounts was initially coded as logical combinations of generic accident factors. A Bayesian Network (BN)-based model was then built for OAMDs using these data and expert knowledge. A data clustering process was subsequently performed to group the OAMDs into similar classes from generic factor occurrence and pattern standpoints. Finally, the Most Probable Explanation (MPE) was evaluated and identified as the associated recurrent scenario for each class. Using this approach, 8 scenarios were extracted to describe 143 OAMDs in the construction and metallurgy sectors. Their recurrent nature is discussed. Probable generic factor combinations provide a fair representation of particularly serious OAMDs, as described in narrative texts. This work represents a real contribution to raising company awareness of the variety of circumstances, in which these accidents occur, to progressing in the prevention of such accidents and to developing an analysis framework dedicated to this kind of accident. Copyright © 2014 Elsevier Ltd. All rights reserved.

NOTE: Development of modified voxel phantoms for the numerical dosimetric reconstruction of radiological accidents involving external sources: implementation in SESAME tool

NASA Astrophysics Data System (ADS)

Courageot, Estelle; Sayah, Rima; Huet, Christelle

2010-05-01

Estimating the dose distribution in a victim's body is a relevant indicator in assessing biological damage from exposure in the event of a radiological accident caused by an external source. When the dose distribution is evaluated with a numerical anthropomorphic model, the posture and morphology of the victim have to be reproduced as realistically as possible. Several years ago, IRSN developed a specific software application, called the simulation of external source accident with medical images (SESAME), for the dosimetric reconstruction of radiological accidents by numerical simulation. This tool combines voxel geometry and the MCNP(X) Monte Carlo computer code for radiation-material interaction. This note presents a new functionality in this software that enables the modelling of a victim's posture and morphology based on non-uniform rational B-spline (NURBS) surfaces. The procedure for constructing the modified voxel phantoms is described, along with a numerical validation of this new functionality using a voxel phantom of the RANDO tissue-equivalent physical model.

Development of modified voxel phantoms for the numerical dosimetric reconstruction of radiological accidents involving external sources: implementation in SESAME tool.

PubMed

Courageot, Estelle; Sayah, Rima; Huet, Christelle

2010-05-07

Estimating the dose distribution in a victim's body is a relevant indicator in assessing biological damage from exposure in the event of a radiological accident caused by an external source. When the dose distribution is evaluated with a numerical anthropomorphic model, the posture and morphology of the victim have to be reproduced as realistically as possible. Several years ago, IRSN developed a specific software application, called the simulation of external source accident with medical images (SESAME), for the dosimetric reconstruction of radiological accidents by numerical simulation. This tool combines voxel geometry and the MCNP(X) Monte Carlo computer code for radiation-material interaction. This note presents a new functionality in this software that enables the modelling of a victim's posture and morphology based on non-uniform rational B-spline (NURBS) surfaces. The procedure for constructing the modified voxel phantoms is described, along with a numerical validation of this new functionality using a voxel phantom of the RANDO tissue-equivalent physical model.

MELCOR simulations of the severe accident at Fukushima Daiichi Unit 3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cardoni, Jeffrey; Gauntt, Randall; Kalinich, Donald

In response to the accident at the Fukushima Daiichi nuclear power station in Japan, the U.S. Nuclear Regulatory Commission and U.S. Department of Energy agreed to jointly sponsor an accident reconstruction study as a means of assessing the severe accident modeling capability of the MELCOR code. Objectives of the project included reconstruction of the accident progressions using computer models and accident data, and validation of the MELCOR code and the Fukushima models against plant data. A MELCOR 2.1 model of the Fukushima Daiichi Unit 3 reactor is developed using plant-specific information and accident-specific boundary conditions, which involve considerable uncertainty duemore » to the inherent nature of severe accidents. Publicly available thermal-hydraulic data and radioactivity release estimates have evolved significantly since the accidents. Such data are expected to continually change as the reactors are decommissioned and more measurements are performed. As a result, the MELCOR simulations in this work primarily use boundary conditions that are based on available plant data as of May 2012.« less

MELCOR simulations of the severe accident at Fukushima Daiichi Unit 3

DOE PAGES

Cardoni, Jeffrey; Gauntt, Randall; Kalinich, Donald; ...

2014-05-01

In response to the accident at the Fukushima Daiichi nuclear power station in Japan, the U.S. Nuclear Regulatory Commission and U.S. Department of Energy agreed to jointly sponsor an accident reconstruction study as a means of assessing the severe accident modeling capability of the MELCOR code. Objectives of the project included reconstruction of the accident progressions using computer models and accident data, and validation of the MELCOR code and the Fukushima models against plant data. A MELCOR 2.1 model of the Fukushima Daiichi Unit 3 reactor is developed using plant-specific information and accident-specific boundary conditions, which involve considerable uncertainty duemore » to the inherent nature of severe accidents. Publicly available thermal-hydraulic data and radioactivity release estimates have evolved significantly since the accidents. Such data are expected to continually change as the reactors are decommissioned and more measurements are performed. As a result, the MELCOR simulations in this work primarily use boundary conditions that are based on available plant data as of May 2012.« less

The Role of Spatial Disorientation in Fatal General Aviation Accidents

NASA Technical Reports Server (NTRS)

Scheuring, RIchard

2005-01-01

In-flight Spatial Disorientation (SD) in pilots is a serious threat to aviation safety. Indeed, SD may play a much larger role in aviation accidents than the approximate 6-8% reported by the National Transportation Safety Board (NTSB) each year, because some accidents coded by the NTSB as aircraft control-not maintained (ACNM) may actually result from SD. The purpose of this study is to determine whether SD is underestimated as a cause of fatal general aviation (GA) accidents in the NTSB database. Fatal GA airplane accidents occurring between January 1995 and December 1999 were reviewed from the NTSB aviation accident database. Cases coded as ACNM or SD as the probable cause were selected for review by a panel of aerospace medicine specialists. Using a rating scale, each rater was instructed to determine if SD was the probable cause of the accident. Agreement between the raters and agreement between the raters and the NTSB were evaluated by Kappa statistics. The raters agreed that 11 out of 20 (55%) accidents coded by the NTSB as ACNM were probably caused by SD (p less than 0.05). Agreement between the raters and the NTSB did not reach significance (p greater than 0.05). The 95% C.I. for the sampling population estimated that between 33-77% of cases that the NTSB identified as ACNM could be identified by aerospace medicine experts as SD. Aerospace medicine specialists agreed that some cases coded by the NTSB as ACNM were probably caused by SD. Consequently, a larger number of accidents may be caused by the pilot succumbing to SD than indicated in the NTSB database. This new information should encourage regulating agencies to insure that pilots receive SD recognition training, enabling them to take appropriate corrective actions during flight. This could lead to new training standards, ultimately saving lives among GA airplane pilots.

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

NASA Astrophysics Data System (ADS)

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

2017-11-01

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

MELCOR Analysis of OSU Multi-Application Small Light Water Reactor (MASLWR) Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yoon, Dhongik S.; Jo, HangJin; Fu, Wen

A multi-application small light water reactor (MASLWR) conceptual design was developed by Oregon State University (OSU) with emphasis on passive safety systems. The passive containment safety system employs condensation and natural circulation to achieve the necessary heat removal from the containment in case of postulated accidents. Containment condensation experiments at the MASLWR test facility at OSU are modeled and analyzed with MELCOR, a system-level reactor accident analysis computer code. The analysis assesses its ability to predict condensation heat transfer in the presence of noncondensable gas for accidents where high-energy steam is released into the containment. This work demonstrates MELCOR’s abilitymore » to predict the pressure-temperature response of the scaled containment. Our analysis indicates that the heat removal rates are underestimated in the experiment due to the limited locations of the thermocouples and applies corrections to these measurements by conducting integral energy analyses along with CFD simulation for confirmation. Furthermore, the corrected heat removal rate measurements and the MELCOR predictions on the heat removal rate from the containment show good agreement with the experimental data.« less

MELCOR Analysis of OSU Multi-Application Small Light Water Reactor (MASLWR) Experiment

DOE PAGES

Yoon, Dhongik S.; Jo, HangJin; Fu, Wen; ...

2017-05-23

A multi-application small light water reactor (MASLWR) conceptual design was developed by Oregon State University (OSU) with emphasis on passive safety systems. The passive containment safety system employs condensation and natural circulation to achieve the necessary heat removal from the containment in case of postulated accidents. Containment condensation experiments at the MASLWR test facility at OSU are modeled and analyzed with MELCOR, a system-level reactor accident analysis computer code. The analysis assesses its ability to predict condensation heat transfer in the presence of noncondensable gas for accidents where high-energy steam is released into the containment. This work demonstrates MELCOR’s abilitymore » to predict the pressure-temperature response of the scaled containment. Our analysis indicates that the heat removal rates are underestimated in the experiment due to the limited locations of the thermocouples and applies corrections to these measurements by conducting integral energy analyses along with CFD simulation for confirmation. Furthermore, the corrected heat removal rate measurements and the MELCOR predictions on the heat removal rate from the containment show good agreement with the experimental data.« less

Pedestrian injury causation study (pedestrian accident typing)

DOT National Transportation Integrated Search

1982-08-01

A new computerized pedestrian accident typing procedure was tested on 1,997 cases from the Pedestrian Injury Causation Study (PICS). Two coding procedures were used to determine the effects of quantity and quality of information on accident typing ac...

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

1992-01-01

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

A Study to Determine the Need for a Standard Limiting the Horsepower of Recreational Boats.

DTIC Science & Technology

1978-09-01

Acceptance Number of Number Fatal Accidents Non -Fatal Accidents - (Lost control ) 1 93 2 :No attempt to avoid collision) 1 19 72 fAttempted to avoic, not enough...base, and an explanation of the computer SModel designed to aid in organizing and analyzing the data are presented with the results of the analyses. An...Standard 75 S 3.2 Non -Powering Related Accident Sample 76 3.3 Coded Information and Coding Form 77 • - 3.4 Effectiveness Evaluation of the Current

Analysis of the influence of the heat transfer phenomena on the late phase of the ThAI Iod-12 test

NASA Astrophysics Data System (ADS)

Gonfiotti, B.; Paci, S.

2014-11-01

Iodine is one of the major contributors to the source term during a severe accident in a Nuclear Power Plant for its volatility and high radiological consequences. Therefore, large efforts have been made to describe the Iodine behaviour during an accident, especially in the containment system. Due to the lack of experimental data, in the last years many attempts were carried out to fill the gaps on the knowledge of Iodine behaviour. In this framework, two tests (ThAI Iod-11 and Iod-12) were carried out inside a multi-compartment steel vessel. A quite complex transient characterizes these two tests; therefore they are also suitable for thermal- hydraulic benchmarks. The two tests were originally released for a benchmark exercise during the SARNET2 EU Project. At the end of this benchmark a report covering the main findings was issued, stating that the common codes employed in SA studies were able to simulate the tests but with large discrepancies. The present work is then related to the application of the new versions of ASTEC and MELCOR codes with the aim of carry out a new code-to-code comparison vs. ThAI Iod-12 experimental data, focusing on the influence of the heat exchanges with the outer environment, which seems to be one of the most challenging issues to cope with.

Radioactive release during nuclear accidents in Chernobyl and Fukushima

NASA Astrophysics Data System (ADS)

Nur Ain Sulaiman, Siti; Mohamed, Faizal; Rahim, Ahmad Nabil Ab

2018-01-01

Nuclear accidents that occurred in Chernobyl and Fukushima have initiated many research interests to understand the cause and mechanism of radioactive release within reactor compound and to the environment. Common types of radionuclide release are the fission products from the irradiated fuel rod itself. In case of nuclear accident, the focus of monitoring will be mostly on the release of noble gases, I-131 and Cs-137. As these are the only accidents have been rated within International Nuclear Events Scale (INES) Level 7, the radioactive release to the environment was one of the critical insights to be monitored. It was estimated that the release of radioactive material to the atmosphere due to Fukushima accident was approximately 10% of the Chernobyl accident. By referring to the previous reports using computational code systems to model the release rate, the release activity of I-131 and Cs-137 in Chernobyl was significantly higher compare to Fukushima. The simulation code also showed that Chernobyl had higher release rate of both radionuclides on the day of accident. Other factors affecting the radioactive release for Fukushima and Chernobyl accidents such as the current reactor technology and safety measures are also compared for discussion.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Little, M.P.; Muirhead, C.R.; Goossens, L.H.J.

1997-12-01

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M.

1998-04-01

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

Establishment and assessment of code scaling capability

NASA Astrophysics Data System (ADS)

Lim, Jaehyok

In this thesis, a method for using RELAP5/MOD3.3 (Patch03) code models is described to establish and assess the code scaling capability and to corroborate the scaling methodology that has been used in the design of the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR applications (PUMA-E) facility. It was sponsored by the United States Nuclear Regulatory Commission (USNRC) under the program "PUMA ESBWR Tests". PUMA-E facility was built for the USNRC to obtain data on the performance of the passive safety systems of the General Electric (GE) Nuclear Energy Economic Simplified Boiling Water Reactor (ESBWR). Similarities between the prototype plant and the scaled-down test facility were investigated for a Gravity-Driven Cooling System (GDCS) Drain Line Break (GDLB). This thesis presents the results of the GDLB test, i.e., the GDLB test with one Isolation Condenser System (ICS) unit disabled. The test is a hypothetical multi-failure small break loss of coolant (SB LOCA) accident scenario in the ESBWR. The test results indicated that the blow-down phase, Automatic Depressurization System (ADS) actuation, and GDCS injection processes occurred as expected. The GDCS as an emergency core cooling system provided adequate supply of water to keep the Reactor Pressure Vessel (RPV) coolant level well above the Top of Active Fuel (TAF) during the entire GDLB transient. The long-term cooling phase, which is governed by the Passive Containment Cooling System (PCCS) condensation, kept the reactor containment system that is composed of Drywell (DW) and Wetwell (WW) below the design pressure of 414 kPa (60 psia). In addition, the ICS continued participating in heat removal during the long-term cooling phase. A general Code Scaling, Applicability, and Uncertainty (CSAU) evaluation approach was discussed in detail relative to safety analyses of Light Water Reactor (LWR). The major components of the CSAU methodology that were highlighted particularly focused on the scaling issues of experiments and models and their applicability to the nuclear power plant transient and accidents. The major thermal-hydraulic phenomena to be analyzed were identified and the predictive models adopted in RELAP5/MOD3.3 (Patch03) code were briefly reviewed.

Status of VICTORIA: NRC peer review and recent code applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bixler, N.E.; Schaperow, J.H.

1997-12-01

VICTORIA is a mechanistic computer code designed to analyze fission product behavior within a nuclear reactor coolant system (RCS) during a severe accident. It provides detailed predictions of the release of radioactive and nonradioactive materials from the reactor core and transport and deposition of these materials within the RCS. A summary of the results and recommendations of an independent peer review of VICTORIA by the US Nuclear Regulatory Commission (NRC) is presented, along with recent applications of the code. The latter include analyses of a temperature-induced steam generator tube rupture sequence and post-test analyses of the Phebus FPT-1 test. Themore » next planned Phebus test, FTP-4, will focus on fission product releases from a rubble bed, especially those of the less-volatile elements, and on the speciation of the released elements. Pretest analyses using VICTORIA to estimate the magnitude and timing of releases are presented. The predicted release of uranium is a matter of particular importance because of concern about filter plugging during the test.« less

Interface requirements to couple thermal-hydraulic codes to severe accident codes: ATHLET-CD

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trambauer, K.

1997-07-01

The system code ATHLET-CD is being developed by GRS in cooperation with IKE and IPSN. Its field of application comprises the whole spectrum of leaks and large breaks, as well as operational and abnormal transients for LWRs and VVERs. At present the analyses cover the in-vessel thermal-hydraulics, the early phases of core degradation, as well as fission products and aerosol release from the core and their transport in the Reactor Coolant System. The aim of the code development is to extend the simulation of core degradation up to failure of the reactor pressure vessel and to cover all physically reasonablemore » accident sequences for western and eastern LWRs including RMBKs. The ATHLET-CD structure is highly modular in order to include a manifold spectrum of models and to offer an optimum basis for further development. The code consists of four general modules to describe the reactor coolant system thermal-hydraulics, the core degradation, the fission product core release, and fission product and aerosol transport. Each general module consists of some basic modules which correspond to the process to be simulated or to its specific purpose. Besides the code structure based on the physical modelling, the code follows four strictly separated steps during the course of a calculation: (1) input of structure, geometrical data, initial and boundary condition, (2) initialization of derived quantities, (3) steady state calculation or input of restart data, and (4) transient calculation. In this paper, the transient solution method is briefly presented and the coupling methods are discussed. Three aspects have to be considered for the coupling of different modules in one code system. First is the conservation of masses and energy in the different subsystems as there are fluid, structures, and fission products and aerosols. Second is the convergence of the numerical solution and stability of the calculation. The third aspect is related to the code performance, and running time.« less

Condensation model for the ESBWR passive condensers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Revankar, S. T.; Zhou, W.; Wolf, B.

2012-07-01

In the General Electric's Economic simplified boiling water reactor (GE-ESBWR) the passive containment cooling system (PCCS) plays a major role in containment pressure control in case of an loss of coolant accident. The PCCS condenser must be able to remove sufficient energy from the reactor containment to prevent containment from exceeding its design pressure following a design basis accident. There are three PCCS condensation modes depending on the containment pressurization due to coolant discharge; complete condensation, cyclic venting and flow through mode. The present work reviews the models and presents model predictive capability along with comparison with existing data frommore » separate effects test. The condensation models in thermal hydraulics code RELAP5 are also assessed to examine its application to various flow modes of condensation. The default model in the code predicts complete condensation well, and basically is Nusselt solution. The UCB model predicts through flow well. None of condensation model in RELAP5 predict complete condensation, cyclic venting, and through flow condensation consistently. New condensation correlations are given that accurately predict all three modes of PCCS condensation. (authors)« less

Pretest aerosol code comparisons for LWR aerosol containment tests LA1 and LA2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wright, A.L.; Wilson, J.H.; Arwood, P.C.

The Light-Water-Reactor (LWR) Aerosol Containment Experiments (LACE) are being performed in Richland, Washington, at the Hanford Engineering Development Laboratory (HEDL) under the leadership of an international project board and the Electric Power Research Institute. These tests have two objectives: (1) to investigate, at large scale, the inherent aerosol retention behavior in LWR containments under simulated severe accident conditions, and (2) to provide an experimental data base for validating aerosol behavior and thermal-hydraulic computer codes. Aerosol computer-code comparison activities are being coordinated at the Oak Ridge National Laboratory. For each of the six LACE tests, ''pretest'' calculations (for code-to-code comparisons) andmore » ''posttest'' calculations (for code-to-test data comparisons) are being performed. The overall goals of the comparison effort are (1) to provide code users with experience in applying their codes to LWR accident-sequence conditions and (2) to evaluate and improve the code models.« less

Review of Hybrid (Deterministic/Monte Carlo) Radiation Transport Methods, Codes, and Applications at Oak Ridge National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wagner, John C; Peplow, Douglas E.; Mosher, Scott W

2011-01-01

This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or moremore » localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(102-4), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M.

1997-12-01

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

Review of hydrogen accidents and incidents in NASA operations

NASA Technical Reports Server (NTRS)

Ordin, P. M.

1974-01-01

A number of the accidents/incidents with hydrogen in NASA operations are reviewed. The cause factors for the mishaps are reviewed and show that although few accidents occurred, the number could have been further reduced if the established NASA rules and regulations had been followed. Requirements for effective safety codes and areas of study for hydrogen safety information are included. The report concludes with a compilation of 96 hydrogen mishaps; a description of the accidents and their causes.

Development of Northeast Asia Nuclear Power Plant Accident Simulator.

PubMed

Kim, Juyub; Kim, Juyoul; Po, Li-Chi Cliff

2017-06-15

A conclusion from the lessons learned after the March 2011 Fukushima Daiichi accident was that Korea needs a tool to estimate consequences from a major accident that could occur at a nuclear power plant located in a neighboring country. This paper describes a suite of computer-based codes to be used by Korea's nuclear emergency response staff for training and potentially operational support in Korea's national emergency preparedness and response program. The systems of codes, Northeast Asia Nuclear Accident Simulator (NANAS), consist of three modules: source-term estimation, atmospheric dispersion prediction and dose assessment. To quickly assess potential doses to the public in Korea, NANAS includes specific reactor data from the nuclear power plants in China, Japan and Taiwan. The completed simulator is demonstrated using data for a hypothetical release. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Methodology, status and plans for development and assessment of Cathare code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bestion, D.; Barre, F.; Faydide, B.

1997-07-01

This paper presents the methodology, status and plans for the development, assessment and uncertainty evaluation of the Cathare code. Cathare is a thermalhydraulic code developed by CEA (DRN), IPSN, EDF and FRAMATOME for PWR safety analysis. First, the status of the code development and assessment is presented. The general strategy used for the development and the assessment of the code is presented. Analytical experiments with separate effect tests, and component tests are used for the development and the validation of closure laws. Successive Revisions of constitutive laws are implemented in successive Versions of the code and assessed. System tests ormore » integral tests are used to validate the general consistency of the Revision. Each delivery of a code Version + Revision is fully assessed and documented. A methodology is being developed to determine the uncertainty on all constitutive laws of the code using calculations of many analytical tests and applying the Discrete Adjoint Sensitivity Method (DASM). At last, the plans for the future developments of the code are presented. They concern the optimization of the code performance through parallel computing - the code will be used for real time full scope plant simulators - the coupling with many other codes (neutronic codes, severe accident codes), the application of the code for containment thermalhydraulics. Also, physical improvements are required in the field of low pressure transients and in the modeling for the 3-D model.« less

Trucks involved in fatal accidents codebook 2008.

DOT National Transportation Integrated Search

2011-01-01

This report provides documentation for UMTRIs file of Trucks Involved in Fatal Accidents : (TIFA), 2008, including distributions of the code values for each variable in the file. The 2008 : TIFA file is a census of all medium and heavy trucks invo...

Buses involved in fatal accidents codebook 2008.

DOT National Transportation Integrated Search

2011-03-01

This report provides documentation for UMTRIs file of Buses Involved in Fatal Accidents (BIFA), 2008, : including distributions of the code values for each variable in the file. The 2008 BIFA file is a census of all : buses involved in a fatal acc...

Buses involved in fatal accidents codebook 2007.

DOT National Transportation Integrated Search

2009-12-01

This report provides documentation for UMTRIs file of Buses Involved in Fatal Accidents (BIFA), 2007, : including distributions of the code values for each variable in the file. The 2007 BIFA file is a census of all : buses involved in a fatal acc...

Development of a three-dimensional transient code for reactivity-initiated events of BWRs (boiling water reactors) - Models and code verifications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Uematsu, Hitoshi; Yamamoto, Toru; Izutsu, Sadayuki

1990-06-01

A reactivity-initiated event is a design-basis accident for the safety analysis of boiling water reactors. It is defined as a rapid transient of reactor power caused by a reactivity insertion of over $1.0 due to a postulated drop or abnormal withdrawal of the control rod from the core. Strong space-dependent feedback effects are associated with the local power increase due to control rod movement. A realistic treatment of the core status in a transient by a code with a detailed core model is recommended in evaluating this event. A three-dimensional transient code, ARIES, has been developed to meet this need.more » The code simulates the event with three-dimensional neutronics, coupled with multichannel thermal hydraulics, based on a nonequilibrium separated flow model. The experimental data obtained in reactivity accident tests performed with the SPERT III-E core are used to verify the entire code, including thermal-hydraulic models.« less

Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baek J.; Diamond D.; Cuadra, A.

Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a modelmore » of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.« less

Review of Hybrid (Deterministic/Monte Carlo) Radiation Transport Methods, Codes, and Applications at Oak Ridge National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wagner, John C; Peplow, Douglas E.; Mosher, Scott W

2010-01-01

This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or moremore » localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(10{sup 2-4}), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications.« less

Trucks involved in fatal accidents codebook 2004 (Version March 23, 2007).

DOT National Transportation Integrated Search

2007-03-01

"This report provides documentation for UMTRIs file of Trucks Involved in Fatal Accidents (TIFA), : 2004, including distributions of the code values for each variable in the file. The 2004 TIFA file is : a census of all medium and heavy trucks inv...

Trucks involved in fatal accidents codebook 2010 (Version October 22, 2012).

DOT National Transportation Integrated Search

2012-11-01

This report provides documentation for UMTRIs file of Trucks Involved in Fatal Accidents : (TIFA), 2010, including distributions of the code values for each variable in the file. The 2010 : TIFA file is a census of all medium and heavy trucks invo...

Investigation of air cleaning system response to accident conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andrae, R.W.; Bolstad, J.W.; Foster, R.D.

1980-01-01

Air cleaning system response to the stress of accident conditions are being investigated. A program overview and hghlight recent results of our investigation are presented. The program includes both analytical and experimental investigations. Computer codes for predicting effects of tornados, explosions, fires, and material transport are described. The test facilities used to obtain supportive experimental data to define structural integrity and confinement effectiveness of ventilation system components are described. Examples of experimental results for code verification, blower response to tornado transients, and filter response to tornado and explosion transients are reported.

Novel Accident-Tolerant Fuel Meat and Cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robert D. Mariani; Pavel G Medvedev; Douglas L Porter

A novel accident-tolerant fuel meat and cladding are here proposed. The fuel meat design incorporates annular fuel with inserts and discs that are fabricated from a material having high thermal conductivity, for example niobium. The inserts are rods or tubes. Discs separate the fuel pellets. Using the BISON fuel performance code it was found that the peak fuel temperature can be lowered by more than 600 degrees C for one set of conditions with niobium metal as the thermal conductor. In addition to improved safety margin, several advantages are expected from the lower temperature such as decreased fission gas releasemore » and fuel cracking. Advantages and disadvantages are discussed. An enrichment of only 7.5% fully compensates the lost reactivity of the displaced UO2. Slightly higher enrichments, such as 9%, allow uprates and increased burnups to offset the initial costs for retooling. The design has applications for fast reactors and transuranic burning, which may accelerate its development. A zirconium silicide coating is also described for accident tolerant applications. A self-limiting degradation behavior for this coating is expected to produce a glassy, self-healing layer that becomes more protective at elevated temperature, with some similarities to MoSi2 and other silicides. Both the fuel and coating may benefit from the existing technology infrastructure and the associated wide expertise for a more rapid development in comparison to other, more novel fuels and cladding.« less

The Fukushima Daiichi Accident Study Information Portal

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shawn St. Germain; Curtis Smith; David Schwieder

This paper presents a description of The Fukushima Daiichi Accident Study Information Portal. The Information Portal was created by the Idaho National Laboratory as part of joint NRC and DOE project to assess the severe accident modeling capability of the MELCOR analysis code. The Fukushima Daiichi Accident Study Information Portal was created to collect, store, retrieve and validate information and data for use in reconstructing the Fukushima Daiichi accident. In addition to supporting the MELCOR simulations, the Portal will be the main DOE repository for all data, studies and reports related to the accident at the Fukushima Daiichi nuclear powermore » station. The data is stored in a secured (password protected and encrypted) repository that is searchable and accessible to researchers at diverse locations.« less

Road safety practices among commercial motorcyclists in a rural town in Nigeria: implications for health education.

PubMed

Amoran, O E; Eme, Owoaje; Giwa, O A; Gbolahan, O B

This cross-sectional, community-based study was carried out among commercial motorcyclists in Igboora. All the commercial motor parks in Igboora were visited and all the commercial motorcyclists who consented to participate in the study were interviewed. Information on the respondents' socio-demographic characteristics, and the practice of road safety measures was collected using an interviewer administered questionnaire. A total of 299 motorcyclists were interviewed. All (100%) of them were males. The mean age of the respondents was 27.4 +/- 7.4 years. One hundred eighty-two (60.7%) of the motorcyclists had the correct knowledge of the purpose of Highway Code. Only 70 (23.3%) could recognize more than half of the currently used road safety codes and 47 (15.7%) obey these road safety codes more than half of the time they see it. Only 183 (61.2%) of them had a driving license and 72 (24.1%) were able to produce these licenses on demand. All (100%) of the respondents did not use any protective helmet. Those who have longer years of working experience, higher level of education and higher knowledge of the safety codes practice it more regularly (r = 0.198, p = 0.001, chi2= 9.31, p = 0.025, and r = 0.28, p = 0.001 respectively). One hundred thirty-six (45.5%) have been involved in at least one accident in the preceding year. The overall incidence of road traffic accident was 2.16 per 1,000. There was however on statistically significant association between the practice of road safety codes and the occurrence of road traffic accidents (chi2= 0.176, p = 0.916). The study shows that the practice of road safety measures was low in this rural Nigerian community and was not associated with the incidence of road traffic accidents. Introducing road safety education particularly targeted at educating the motorcyclists on the importance and practice of road safety measures would lead to an increase in the practice of the safety measures and hopefully a reduction in the incidence of road traffic accidents.

The Euratom-Rosatom ERCOSAM-SAMARA projects on containment thermal-hydraulics of current and future LWRs for severe accident management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paladino, D.; Guentay, S.; Andreani, M.

2012-07-01

During a postulated severe accident with core degradation, hydrogen would form in the reactor pressure vessel mainly due to high temperatures zirconium-steam reaction and flow together with steam into the containment where it will mix with the containment atmosphere (steam-air). The hydrogen transport into the containment is a safety concern because it can lead to explosive mixtures through the associated phenomena of condensation, mixing and stratification. The ERCOSAM and SAMARA projects, co-financed by the European Union and the Russia, include various experiments addressing accident scenarios scaled down from existing plant calculations to different thermal-hydraulics facilities (TOSQAN, MISTRA, PANDA, SPOT). Themore » tests sequences aim to investigate hydrogen concentration build-up and stratification during a postulated accident and the effect of the activation of Severe Accident Management systems (SAMs), e.g. sprays, coolers and Passive Auto-catalytic Recombiners (PARs). Analytical activities, performed by the project participants, are an essential component of the projects, as they aim to improve and validate various computational methods. They accompany the projects in the various phases; plant calculations, scaling to generic containment and to the different facilities, planning pre-test and post-test simulations are performed. Code benchmark activities on the basis of conceptual near full scale HYMIX facility will finally provide a further opportunity to evaluate the applicability of the various methods to the study of scaling issues. (authors)« less

The Design of PSB-VVER Experiments Relevant to Accident Management

NASA Astrophysics Data System (ADS)

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

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

Fukushima Daiichi Unit 1 Uncertainty Analysis-Exploration of Core Melt Progression Uncertain Parameters-Volume II.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Denman, Matthew R.; Brooks, Dusty Marie

Sandia National Laboratories (SNL) has conducted an uncertainty analysi s (UA) on the Fukushima Daiichi unit (1F1) accident progression wit h the MELCOR code. Volume I of the 1F1 UA discusses the physical modeling details and time history results of the UA. Volume II of the 1F1 UA discusses the statistical viewpoint. The model used was developed for a previous accident reconstruction investigation jointly sponsored by the US Department of Energy (DOE) and Nuclear Regulatory Commission (NRC). The goal of this work was to perform a focused evaluation of uncertainty in core damage progression behavior and its effect on keymore » figures - of - merit (e.g., hydrogen production, fraction of intact fuel, vessel lower head failure) and in doing so assess the applicability of traditional sensitivity analysis techniques .« less

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

PubMed

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

2012-02-01

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

CHEMICAL STORAGE: MYTHS VERSUS REALITY

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simmons, F

A large number of resources explaining proper chemical storage are available. These resources include books, databases/tables, and articles that explain various aspects of chemical storage including compatible chemical storage, signage, and regulatory requirements. Another source is the chemical manufacturer or distributor who provides storage information in the form of icons or color coding schemes on container labels. Despite the availability of these resources, chemical accidents stemming from improper storage, according to recent reports (1) (2), make up almost 25% of all chemical accidents. This relatively high percentage of chemical storage accidents suggests that these publications and color coding schemes althoughmore » helpful, still provide incomplete information that may not completely mitigate storage risks. This manuscript will explore some ways published storage information may be incomplete, examine the associated risks, and suggest methods to help further eliminate chemical storage risks.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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

RADTRAD: A simplified model for RADionuclide Transport and Removal And Dose estimation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Humphreys, S.L.; Miller, L.A.; Monroe, D.K.

1998-04-01

This report documents the RADTRAD computer code developed for the U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear Reactor Regulation (NRR) to estimate transport and removal of radionuclides and dose at selected receptors. The document includes a users` guide to the code, a description of the technical basis for the code, the quality assurance and code acceptance testing documentation, and a programmers` guide. The RADTRAD code can be used to estimate the containment release using either the NRC TID-14844 or NUREG-1465 source terms and assumptions, or a user-specified table. In addition, the code can account for a reduction in themore » quantity of radioactive material due to containment sprays, natural deposition, filters, and other natural and engineered safety features. The RADTRAD code uses a combination of tables and/or numerical models of source term reduction phenomena to determine the time-dependent dose at user-specified locations for a given accident scenario. The code system also provides the inventory, decay chain, and dose conversion factor tables needed for the dose calculation. The RADTRAD code can be used to assess occupational radiation exposures, typically in the control room; to estimate site boundary doses; and to estimate dose attenuation due to modification of a facility or accident sequence.« less

30 CFR 50.20-5 - Criteria-MSHA Form 7000-1, Section B.

Code of Federal Regulations, 2010 CFR

2010-07-01

... following list which best defines the accident: Code 01—A death of an individual at a mine; Code 02—An injury to an individual at a mine which has a reasonable potential to cause death; Code 03—An entrapment... more than thirty minutes; and Code 12—An event at a mine which causes death or bodily injury to an...

Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murata, K.K.; Williams, D.C.; Griffith, R.O.

1997-12-01

The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of themore » input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.« less

Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baratta, A.J.

1997-07-01

To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts andmore » engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.« less

RMP Guidance for Warehouses - Appendix A/B: 40 CFR part 68/Selected NAICS Codes

EPA Pesticide Factsheets

These appendices contain the full text of 40 Code of Federal Regulations Part 68, Chemical Accident Prevention Provisions; which includes hazard assessment, emergency response, substance thresholds, reporting requirements, and the Risk Management Plan.

Why Data Linkage? The Importance of CODES (Crash Outcome Data Evaluation System)

DOT National Transportation Integrated Search

1996-06-01

This report briefly explains the computerized linked data system, Crash Outcome : Data Evaluation System (CODES) that provides greater depth accident data : analysis. The linking of data helps researchers to understand the nature of : traffic acciden...

Development of fission-products transport model in severe-accident scenarios for Scdap/Relap5

NASA Astrophysics Data System (ADS)

Honaiser, Eduardo Henrique Rangel

The understanding and estimation of the release of fission products during a severe accident became one of the priorities of the nuclear community after 1980, with the events of the Three-mile Island unit 2 (TMI-2), in 1979, and Chernobyl accidents, in 1986. Since this time, theoretical developments and experiments have shown that the primary circuit systems of light water reactors (LWR) have the potential to attenuate the release of fission products, a fact that had been neglected before. An advanced tool, compatible with nuclear thermal-hydraulics integral codes, is developed to predict the retention and physical evolution of the fission products in the primary circuit of LWRs, without considering the chemistry effects. The tool embodies the state-of-the-art models for the involved phenomena as well as develops new models. The capabilities acquired after the implementation of this tool in the Scdap/Relap5 code can be used to increase the accuracy of probability safety assessment (PSA) level 2, enhance the reactor accident management procedures and design new emergency safety features.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gougar, Hans

This document outlines the development of a high fidelity, best estimate nuclear power plant severe transient simulation capability that will complement or enhance the integral system codes historically used for licensing and analysis of severe accidents. As with other tools in the Risk Informed Safety Margin Characterization (RISMC) Toolkit, the ultimate user of Enhanced Severe Transient Analysis and Prevention (ESTAP) capability is the plant decision-maker; the deliverable to that customer is a modern, simulation-based safety analysis capability, applicable to a much broader class of safety issues than is traditional Light Water Reactor (LWR) licensing analysis. Currently, the RISMC pathway’s majormore » emphasis is placed on developing RELAP-7, a next-generation safety analysis code, and on showing how to use RELAP-7 to analyze margin from a modern point of view: that is, by characterizing margin in terms of the probabilistic spectra of the “loads” applied to systems, structures, and components (SSCs), and the “capacity” of those SSCs to resist those loads without failing. The first objective of the ESTAP task, and the focus of one task of this effort, is to augment RELAP-7 analyses with user-selected multi-dimensional, multi-phase models of specific plant components to simulate complex phenomena that may lead to, or exacerbate, severe transients and core damage. Such phenomena include: coolant crossflow between PWR assemblies during a severe reactivity transient, stratified single or two-phase coolant flow in primary coolant piping, inhomogeneous mixing of emergency coolant water or boric acid with hot primary coolant, and water hammer. These are well-documented phenomena associated with plant transients but that are generally not captured in system codes. They are, however, generally limited to specific components, structures, and operating conditions. The second ESTAP task is to similarly augment a severe (post-core damage) accident integral analyses code with high fidelity simulations that would allow investigation of multi-dimensional, multi-phase containment phenomena that are only treated approximately in established codes.« less

Construction safety program for the National Ignition Facility, July 30, 1999 (NIF-0001374-OC)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benjamin, D W

1999-07-30

These rules apply to all LLNL employees, non-LLNL employees (including contract labor, supplemental labor, vendors, personnel matrixed/assigned from other National Laboratories, participating guests, visitors and students) and contractors/subcontractors. The General Rules-Code of Safe Practices shall be used by management to promote accident prevention through indoctrination, safety and health training and on-the-job application. As a condition for contracts award, all contractors and subcontractors and their employees must certify on Form S and H A-l that they have read and understand, or have been briefed and understand, the National Ignition Facility OCIP Project General Rules-Code of Safe Practices. (An interpreter must briefmore » those employees who do not speak or read English fluently.) In addition, all contractors and subcontractors shall adopt a written General Rules-Code of Safe Practices that relates to their operations. The General Rules-Code of Safe Practices must be posted at a conspicuous location at the job site office or be provided to each supervisory employee who shall have it readily available. Copies of the General Rules-Code of Safe Practices can also be included in employee safety pamphlets.« less

Demonstration of fully coupled simplified extended station black-out accident simulation with RELAP-7

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhao, Haihua; Zhang, Hongbin; Zou, Ling

2014-10-01

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The RELAP-7 code develop-ment effort started in October of 2011 and by the end of the second development year, a number of physical components with simplified two phase flow capability have been de-veloped to support the simplified boiling water reactor (BWR) extended station blackout (SBO) analyses. The demonstration case includes the major components for the primary system of a BWR, as well as the safety system components for the safety relief valve (SRV), the reactor core isolation cooling (RCIC)more » system, and the wet well. Three scenar-ios for the SBO simulations have been considered. Since RELAP-7 is not a severe acci-dent analysis code, the simulation stops when fuel clad temperature reaches damage point. Scenario I represents an extreme station blackout accident without any external cooling and cooling water injection. The system pressure is controlled by automatically releasing steam through SRVs. Scenario II includes the RCIC system but without SRV. The RCIC system is fully coupled with the reactor primary system and all the major components are dynamically simulated. The third scenario includes both the RCIC system and the SRV to provide a more realistic simulation. This paper will describe the major models and dis-cuss the results for the three scenarios. The RELAP-7 simulations for the three simplified SBO scenarios show the importance of dynamically simulating the SRVs, the RCIC sys-tem, and the wet well system to the reactor safety during extended SBO accidents.« less

Amino acid codes in mitochondria as possible clues to primitive codes

NASA Technical Reports Server (NTRS)

Jukes, T. H.

1981-01-01

Differences between mitochondrial codes and the universal code indicate that an evolutionary simplification has taken place, rather than a return to a more primitive code. However, these differences make it evident that the universal code is not the only code possible, and therefore earlier codes may have differed markedly from the previous code. The present universal code is probably a 'frozen accident.' The change in CUN codons from leucine to threonine (Neurospora vs. yeast mitochondria) indicates that neutral or near-neutral changes occurred in the corresponding proteins when this code change took place, caused presumably by a mutation in a tRNA gene.

H-division quarterly report, October--December 1977. [Lawrence Livermore Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1978-02-10

The Theoretical EOS Group develops theoretical techniques for describing material properties under extreme conditions and constructs equation-of-state (EOS) tables for specific applications. Work this quarter concentrated on a Li equation of state, equation of state for equilibrium plasma, improved ion corrections to the Thomas--Fermi--Kirzhnitz theory, and theoretical estimates of high-pressure melting in metals. The Experimental Physics Group investigates properties of materials at extreme conditions of pressure and temperature, and develops new experimental techniques. Effort this quarter concerned the following: parabolic projectile distortion in the two-state light-gas gun, construction of a ballistic range for long-rod penetrators, thermodynamics and sound velocities inmore » liquid metals, isobaric expansion measurements in Pt, and calculation of the velocity--mass profile of a jet produced by a shaped charge. Code development was concentrated on the PELE code, a multimaterial, multiphase, explicit finite-difference Eulerian code for pool suppression dynamics of a hypothetical loss-of-coolant accident in a nuclear reactor. Activities of the Fluid Dynamics Group were directed toward development of a code to compute the equations of state and transport properties of liquid metals (e.g. Li) and partially ionized dense plasmas, jet stability in the Li reactor system, and the study and problem application of fluid dynamic turbulence theory. 19 figures, 5 tables. (RWR)« less

Mistranslation: from adaptations to applications.

PubMed

Hoffman, Kyle S; O'Donoghue, Patrick; Brandl, Christopher J

2017-11-01

The conservation of the genetic code indicates that there was a single origin, but like all genetic material, the cell's interpretation of the code is subject to evolutionary pressure. Single nucleotide variations in tRNA sequences can modulate codon assignments by altering codon-anticodon pairing or tRNA charging. Either can increase translation errors and even change the code. The frozen accident hypothesis argued that changes to the code would destabilize the proteome and reduce fitness. In studies of model organisms, mistranslation often acts as an adaptive response. These studies reveal evolutionary conserved mechanisms to maintain proteostasis even during high rates of mistranslation. This review discusses the evolutionary basis of altered genetic codes, how mistranslation is identified, and how deviations to the genetic code are exploited. We revisit early discoveries of genetic code deviations and provide examples of adaptive mistranslation events in nature. Lastly, we highlight innovations in synthetic biology to expand the genetic code. The genetic code is still evolving. Mistranslation increases proteomic diversity that enables cells to survive stress conditions or suppress a deleterious allele. Genetic code variants have been identified by genome and metagenome sequence analyses, suppressor genetics, and biochemical characterization. Understanding the mechanisms of translation and genetic code deviations enables the design of new codes to produce novel proteins. Engineering the translation machinery and expanding the genetic code to incorporate non-canonical amino acids are valuable tools in synthetic biology that are impacting biomedical research. This article is part of a Special Issue entitled "Biochemistry of Synthetic Biology - Recent Developments" Guest Editor: Dr. Ilka Heinemann and Dr. Patrick O'Donoghue. Copyright © 2017 Elsevier B.V. All rights reserved.

Verification and validation of RADMODL Version 1.0

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kimball, K.D.

1993-03-01

RADMODL is a system of linked computer codes designed to calculate the radiation environment following an accident in which nuclear materials are released. The RADMODL code and the corresponding Verification and Validation (V&V) calculations (Appendix A), were developed for Westinghouse Savannah River Company (WSRC) by EGS Corporation (EGS). Each module of RADMODL is an independent code and was verified separately. The full system was validated by comparing the output of the various modules with the corresponding output of a previously verified version of the modules. The results of the verification and validation tests show that RADMODL correctly calculates the transportmore » of radionuclides and radiation doses. As a result of this verification and validation effort, RADMODL Version 1.0 is certified for use in calculating the radiation environment following an accident.« less

Quality evaluation of official accident reports conducted by Labour Authorities in Andalusia (Spain).

PubMed

Salguero-Caparros, Francisco; Suarez-Cebador, Manuel; Carrillo-Castrillo, Jesús A; Rubio-Romero, Juan Carlos

2018-01-01

A public accident investigation is carried out when the consequences of the incident are significant or the accident has occurred in unusual circumstances. We evaluated the quality of the official accident investigations being conducted by Safety Specialists of the Labour Authorities in Andalusia. To achieve this objective, we analysed 98 occupational accident investigations conducted by the Labour Authorities in Andalusia in the last quarter of 2014. Various phases in the accident investigation process were examined, such as the use of the Eurostat variables within European Statistics on Accidents at Work (ESAW), detection of causes, determination of preventive measures, cost analysis of the accidents, identification of noncompliance with legal requirements or the investigation method used. The results of this study show that 77% of the official occupational accident investigation reports analysed were conducted in accordance with all the quality criteria recommended in the literature. To enhance glogal learning, and optimize allocation of resources, we propose the development of a harmonized European model for the public investigation of occupational accidents. Further it would be advisable to create a common classification and coding system for the causes of accidents for all European Union Member States.

MODELLING OF FUEL BEHAVIOUR DURING LOSS-OF-COOLANT ACCIDENTS USING THE BISON CODE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pastore, G.; Novascone, S. R.; Williamson, R. L.

2015-09-01

This work presents recent developments to extend the BISON code to enable fuel performance analysis during LOCAs. This newly developed capability accounts for the main physical phenomena involved, as well as the interactions among them and with the global fuel rod thermo-mechanical analysis. Specifically, new multiphysics models are incorporated in the code to describe (1) transient fission gas behaviour, (2) rapid steam-cladding oxidation, (3) Zircaloy solid-solid phase transition, (4) hydrogen generation and transport through the cladding, and (5) Zircaloy high-temperature non-linear mechanical behaviour and failure. Basic model characteristics are described, and a demonstration BISON analysis of a LWR fuel rodmore » undergoing a LOCA accident is presented. Also, as a first step of validation, the code with the new capability is applied to the simulation of experiments investigating cladding behaviour under LOCA conditions. The comparison of the results with the available experimental data of cladding failure due to burst is presented.« less

The SAS4A/SASSYS-1 Safety Analysis Code System, Version 5

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fanning, T. H.; Brunett, A. J.; Sumner, T.

The SAS4A/SASSYS-1 computer code is developed by Argonne National Laboratory for thermal, hydraulic, and neutronic analysis of power and flow transients in liquidmetal- cooled nuclear reactors (LMRs). SAS4A was developed to analyze severe core disruption accidents with coolant boiling and fuel melting and relocation, initiated by a very low probability coincidence of an accident precursor and failure of one or more safety systems. SASSYS-1, originally developed to address loss-of-decay-heat-removal accidents, has evolved into a tool for margin assessment in design basis accident (DBA) analysis and for consequence assessment in beyond-design-basis accident (BDBA) analysis. SAS4A contains detailed, mechanistic models of transientmore » thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, its coolant, fuel elements, and structural members to accident conditions. The core channel models in SAS4A provide the capability to analyze the initial phase of core disruptive accidents, through coolant heat-up and boiling, fuel element failure, and fuel melting and relocation. Originally developed to analyze oxide fuel clad with stainless steel, the models in SAS4A have been extended and specialized to metallic fuel with advanced alloy cladding. SASSYS-1 provides the capability to perform a detailed thermal/hydraulic simulation of the primary and secondary sodium coolant circuits and the balance-ofplant steam/water circuit. These sodium and steam circuit models include component models for heat exchangers, pumps, valves, turbines, and condensers, and thermal/hydraulic models of pipes and plena. SASSYS-1 also contains a plant protection and control system modeling capability, which provides digital representations of reactor, pump, and valve controllers and their response to input signal changes.« less

Analysis of construction accidents in Spain, 2003-2008.

PubMed

López Arquillos, Antonio; Rubio Romero, Juan Carlos; Gibb, Alistair

2012-12-01

The research objective for this paper is to obtain a new extended and updated insight to the likely causes of construction accidents in Spain, in order to identify suitable mitigating actions. The paper analyzes all construction sector accidents in Spain between 2003 and 2008. Ten variables were chosen and the influence of each variable is evaluated with respect to the severity of the accident. The descriptive analysis is based on a total of 1,163,178 accidents. Results showed that the severity of accidents was related to variables including age, CNAE (National Classification of Economic Activities) code, size of company, length of service, location of accident, day of the week, days of absence, deviation, injury, and climatic zones. According to data analyzed, a large company is not always necessarily safer than a small company in the aspect of fatal accidents, experienced workers do not have the best accident fatality rates, and accidents occurring away from the usual workplace had more severe consequences. Results obtained in this paper can be used by companies in their occupational safety strategies, and in their safety training programs. Copyright © 2012 National Safety Council and Elsevier Ltd. All rights reserved.

Analyses of transients for an 800 MW-class accelerator driven transmuter with fertile-free fuels

NASA Astrophysics Data System (ADS)

Maschek, Werner; Suzuki, Tohru; Chen, Xue-Nong; Rineiski, Andrei; Matzerath Boccaccini, Claudia; Mori, Magnus; Morita, Koji

2006-06-01

In the FUTURE Program, the development and application of fertile-free fuels for Accelerator Driven Transmuters (ADTs) has been advanced. To assess the reactor performance and safety behavior of an ADT with so-called dedicated fuels, various transient cases for an 800 MW-class Pb/Bi-cooled ADT were investigated using the SIMMER-III code. The FUTURE ADT also served as vehicle to develop and test ideas on a safety concept for such transmuters. After an extensive ranking procedure, a CERCER fuel with an MgO matrix and a CERMET fuel with a Mo-92 matrix were chosen. The transient scenarios shown here are: spurious beam trip (BT), unprotected loss of flow (ULOF) and unprotected blockage accident (UBA). Since the release of fission gas and helium after cladding failure could induce a significant positive reactivity, the gas-blowdown was investigated for the transient scenarios. The present analyses showed that power excursions could be avoided by the fuel sweep-out from the core under severe accident conditions.

RAMONA-3B application to Browns Ferry ATWS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Slovik, G.C.; Neymotin, L.Y.; Saha, P.

1985-01-01

The Anticipated Transient Without Scram (ATWS) is known to be a dominant accident sequence for possible core melt in a Boiling Water Reactor (BWR). A recent Probabilistic Risk Assessment (PRA) analysis for the Browns Ferry nuclear power plant indicates that ATWS is the second most dominant transient for core melt in BWR/4 with Mark I containment. The most dominant sequence being the failure of long term decay heat removal function of the Residual Heat Removal (RHR) system. Of all the various ATWS scenarios, the Main Steam Isolation Valve (MSIV) closure ATWS sequence was chosen for present analysis because of itsmore » relatively high frequency of occurrence and its challenge to the residual heat removal system and containment integrity. The objective of this paper is to discuss four MSIV closure ATWS calculations using the RAMONA-3B code. The paper is a summary of a report being prepared for the USNRC Severe Accident Sequence Analysis (SASA) program which should be referred to for details. 10 refs., 20 figs., 3 tabs.« less

Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Coryell, E.W.; Siefken, L.J.; Harvego, E.A.

1997-07-01

The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures.more » The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harmony, S.C.; Steiner, J.L.; Stumpf, H.J.

The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. As part of the preapplication and eventual design certification process, advanced reactor applicants are required to submit neutronic and thermal-hydraulic safety analyses over a sufficient range of normal operation, transient conditions, and specified accident sequences. Los Alamos is supporting the US Nuclear Regulatory Commission`s preapplication review of the PIUS reactor. A fully one-dimensional modelmore » of the PIUS reactor has been developed for the Transient Reactor Analysis Code, TRACPF1/MOD2. Early in 1992, ABB submitted a Supplemental Information Package describing recent design modifications. An important feature of the PIUS Supplement design was the addition of an active scram system that will function for most transient and accident conditions. A one-dimensional Transient Reactor Analysis Code baseline calculation of the PIUS Supplement design were performed for a break in the main steam line at the outlet nozzle of the loop 3 steam generator. Sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions following a main steam line break. The sensitivity study results provide insights into the robustness of the design.« less

Methodology, status, and plans for development and assessment of the RELAP5 code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johnson, G.W.; Riemke, R.A.

1997-07-01

RELAP/MOD3 is a computer code used for the simulation of transients and accidents in light-water nuclear power plants. The objective of the program to develop and maintain RELAP5 was and is to provide the U.S. Nuclear Regulatory Commission with an independent tool for assessing reactor safety. This paper describes code requirements, models, solution scheme, language and structure, user interface validation, and documentation. The paper also describes the current and near term development program and provides an assessment of the code`s strengths and limitations.

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

PubMed

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.

Thermodynamic consequences of hydrogen combustion within a containment of pressurized water reactor

NASA Astrophysics Data System (ADS)

Bury, Tomasz

2011-12-01

Gaseous hydrogen may be generated in a nuclear reactor system as an effect of the core overheating. This creates a risk of its uncontrolled combustion which may have a destructive consequences, as it could be observed during the Fukushima nuclear power plant accident. Favorable conditions for hydrogen production occur during heavy loss-of-coolant accidents. The author used an own computer code, called HEPCAL, of the lumped parameter type to realize a set of simulations of a large scale loss-of-coolant accidents scenarios within containment of second generation pressurized water reactor. Some simulations resulted in high pressure peaks, seemed to be irrational. A more detailed analysis and comparison with Three Mile Island and Fukushima accidents consequences allowed for withdrawing interesting conclusions.

Application of forensic image analysis in accident investigations.

PubMed

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.

Assessment of MSIV full closure for Santa Maria de Garona Nuclear Power Plant using TRAC-BF1 (G1J1)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crespo, J.L.; Fernandez, R.A.

1993-06-01

This document presents a spurious Main Steam Isolation Value (MSIV) closure analysis for Santa Maria de Garorta Nuclear Power Plan describing the problems found when comparing calculated and real data. The plant is a General Electric Boiling Water Reactor 3, containment type Mark 1. It is operated by NUCLENOR, S.A. and was connected to the grid in 1971. The analysis has been performed by the Apphed Physics Department from the University of Cantabria and the Analysis and Operation Section from NUCLENOR, S.A. as a part of an agreement for developing an engineering simulator of operational transients and accidents for Santamore » Maria de Gamma Power Plant. The analysis was performed using the frozen version of TRAC-BFI (GlJl) code and is the second of two NUCLENOR contributions to the International Code Applications and Assessment Program (ICAP). The code was run in a Cyber 932 with operating system NOS/VE, property of NUCLENOR, S.A.. A programming effort was carried out in order to provide suitable graphics from the output file.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andrews, Nathan; Faucett, Christopher; Haskin, Troy Christopher

Following the conclusion of the first phase of the crosswalk analysis, one of the key unanswered questions was whether or not the deviations found would persist during a partially recovered accident scenario, similar to the one that occurred in TMI - 2. In particular this analysis aims to compare the impact of core degradation morphology on quenching models inherent within the two codes and the coolability of debris during partially recovered accidents. A primary motivation for this study is the development of insights into how uncertainties in core damage progression models impact the ability to assess the potential for recoverymore » of a degraded core. These quench and core recovery models are of the most interest when there is a significant amount of core damage, but intact and degraded fuel still remain in the cor e region or the lower plenum. Accordingly this analysis presents a spectrum of partially recovered accident scenarios by varying both water injection timing and rate to highlight the impact of core degradation phenomena on recovered accident scenarios. This analysis uses the newly released MELCOR 2.2 rev. 966 5 and MAAP5, Version 5.04. These code versions, which incorporate a significant number of modifications that have been driven by analyses and forensic evidence obtained from the Fukushima - Daiichi reactor site.« less

Associations between task, training and social environmental factors and error types involved in rail incidents and accidents.

PubMed

Read, Gemma J M; Lenné, Michael G; Moss, Simon A

2012-09-01

Rail accidents can be understood in terms of the systemic and individual contributions to their causation. The current study was undertaken to determine whether errors and violations are more often associated with different local and organisational factors that contribute to rail accidents. The Contributing Factors Framework (CFF), a tool developed for the collection and codification of data regarding rail accidents and incidents, was applied to a sample of investigation reports. In addition, a more detailed categorisation of errors was undertaken. Ninety-six investigation reports into Australian accidents and incidents occurring between 1999 and 2008 were analysed. Each report was coded independently by two experienced coders. Task demand factors were significantly more often associated with skill-based errors, knowledge and training deficiencies significantly associated with mistakes, and violations significantly linked to social environmental factors. Copyright © 2012 Elsevier Ltd. All rights reserved.

Impact, Fire, and Fluid Spread Code Coupling for Complex Transportation Accident Environment Simulation.

PubMed

Brown, Alexander L; Wagner, Gregory J; Metzinger, Kurt E

2012-06-01

Transportation accidents frequently involve liquids dispersing in the atmosphere. An example is that of aircraft impacts, which often result in spreading fuel and a subsequent fire. Predicting the resulting environment is of interest for design, safety, and forensic applications. This environment is challenging for many reasons, one among them being the disparate time and length scales that are necessary to resolve for an accurate physical representation of the problem. A recent computational method appropriate for this class of problems has been described for modeling the impact and subsequent liquid spread. Because the environment is difficult to instrument and costly to test, the existing validation data are of limited scope and quality. A comparatively well instrumented test involving a rocket propelled cylindrical tank of water was performed, the results of which are helpful to understand the adequacy of the modeling methods. Existing data include estimates of drop sizes at several locations, final liquid surface deposition mass integrated over surface area regions, and video evidence of liquid cloud spread distances. Comparisons are drawn between the experimental observations and the predicted results of the modeling methods to provide evidence regarding the accuracy of the methods, and to provide guidance on the application and use of these methods.

Discussion on accuracy degree evaluation of accident velocity reconstruction model

NASA Astrophysics Data System (ADS)

Zou, Tiefang; Dai, Yingbiao; Cai, Ming; Liu, Jike

In order to investigate the applicability of accident velocity reconstruction model in different cases, a method used to evaluate accuracy degree of accident velocity reconstruction model is given. Based on pre-crash velocity in theory and calculation, an accuracy degree evaluation formula is obtained. With a numerical simulation case, Accuracy degrees and applicability of two accident velocity reconstruction models are analyzed; results show that this method is feasible in practice.

Validation of the CALSPAN gross-motion-simulation code with actually occurring injury patterns in aircraft accidents.

PubMed

Ballo, J M; Dunne, M J; McMeekin, R R

1978-01-01

Digital simulation of aircraft-accident kinematics has heretofore been used almost exclusively as a design tool to explore structural load limits, precalculate decelerative forces at various cabin stations, and describe the effect of protective devices in the crash environment. In an effort to determine the value of digital computer simulation of fatal aircraft accidents, a fatality involving an ejection-system failure (out-of-envelope ejection) was modeled, and the injuries actually incurred were compared to those predicted; good agreement was found. The simulation of fatal aircraft accidents is advantageous because of a well-defined endpoint (death), lack of therapeutic intervention, and a static anatomic situation that can be minutely investigated. Such simulation techniques are a useful tool in the study of experimental trauma.

The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 1: Basic Models

NASA Astrophysics Data System (ADS)

Mosunova, N. A.

2018-05-01

The article describes the basic models included in the EUCLID/V1 integrated code intended for safety analysis of liquid metal (sodium, lead, and lead-bismuth) cooled fast reactors using fuel rods with a gas gap and pellet dioxide, mixed oxide or nitride uranium-plutonium fuel under normal operation, under anticipated operational occurrences and accident conditions by carrying out interconnected thermal-hydraulic, neutronics, and thermal-mechanical calculations. Information about the Russian and foreign analogs of the EUCLID/V1 integrated code is given. Modeled objects, equation systems in differential form solved in each module of the EUCLID/V1 integrated code (the thermal-hydraulic, neutronics, fuel rod analysis module, and the burnup and decay heat calculation modules), the main calculated quantities, and also the limitations on application of the code are presented. The article also gives data on the scope of functions performed by the integrated code's thermal-hydraulic module, using which it is possible to describe both one- and twophase processes occurring in the coolant. It is shown that, owing to the availability of the fuel rod analysis module in the integrated code, it becomes possible to estimate the performance of fuel rods in different regimes of the reactor operation. It is also shown that the models implemented in the code for calculating neutron-physical processes make it possible to take into account the neutron field distribution over the fuel assembly cross section as well as other features important for the safety assessment of fast reactors.

Facility Targeting, Protection and Mission Decision Making Using the VISAC Code

NASA Technical Reports Server (NTRS)

Morris, Robert H.; Sulfredge, C. David

2011-01-01

The Visual Interactive Site Analysis Code (VISAC) has been used by DTRA and several other agencies to aid in targeting facilities and to predict the associated collateral effects for the go, no go mission decision making process. VISAC integrates the three concepts of target geometric modeling, damage assessment capabilities, and an event/fault tree methodology for evaluating accident/incident consequences. It can analyze a variety of accidents/incidents at nuclear or industrial facilities, ranging from simple component sabotage to an attack with military or terrorist weapons. For nuclear facilities, VISAC predicts the facility damage, estimated downtime, amount and timing of any radionuclides released. Used in conjunction with DTRA's HPAC code, VISAC also can analyze transport and dispersion of the radionuclides, levels of contamination of the surrounding area, and the population at risk. VISAC has also been used by the NRC to aid in the development of protective measures for nuclear facilities that may be subjected to attacks by car/truck bombs.

14 CFR 415.41 - Accident investigation plan.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 14 Aeronautics and Space 4 2010-01-01 2010-01-01 false Accident investigation plan. 415.41 Section... Launch Range § 415.41 Accident investigation plan. An applicant must file an accident investigation plan... reporting and responding to launch accidents, launch incidents, or other mishaps, as defined by § 401.5 of...

14 CFR 415.41 - Accident investigation plan.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 14 Aeronautics and Space 4 2011-01-01 2011-01-01 false Accident investigation plan. 415.41 Section... Launch Range § 415.41 Accident investigation plan. An applicant must file an accident investigation plan... reporting and responding to launch accidents, launch incidents, or other mishaps, as defined by § 401.5 of...

49 CFR 176.2 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-10-01

..., spillage, or other accident. INF cargo means packaged irradiated nuclear fuel, plutonium or high-level... Irradiated Nuclear Fuel, Plutonium and High-Level Radioactive Wastes on Board Ships” (INF Code) contained in...

Toxicological investigation in blood samples from suspected impaired driving cases in the Milan area: Possible loss of evidence due to late blood sampling.

PubMed

Ferrari, Davide; Manca, Monica; Premaschi, Simone; Banfi, Giuseppe; Locatelli, Massimo

2018-05-01

Driving under the influence of illicit drugs (DUID) represents a significant menace to public safety and is therefore sanctioned with severe fines and penalties such as driving disqualification or even arrest in case the accident has caused serious injury or death. In Italy, DUID is regulated by the article 187 of the National Street Code, however, the list of the substances to be searched and their threshold concentrations are left to the 20 Italian regional authorities. A further lack of legislative standardization concerns the type of detection methods and moreover the time gap between the car accident and blood sampling. This interval can be as high as 5h, enough to significantly reduce the concentration of drugs with fast pharmacokinetic. By analyzing 1258 blood tests performed on drivers involved in road traffic crashes in the Milan area between 2012 and 2016 we show that approximately 75% of such drivers who tested positive for THC and 15% of the drivers who tested positive for cocaine are at risk of misjudgment. Considering the severe sanctions associated with DUID, we emphasize the urgency of introducing a corrective factor that takes into account the time elapsed between the accident and blood sampling in order to avoid unfair treatment, including the unjust application of sanctions. Copyright © 2018 Elsevier B.V. All rights reserved.

75 FR 17439 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-06

... of an accident or that supports mitigation of an accident previously evaluated. The proposed... probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in...

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ritchie, L.T.; Johnson, J.D.; Blond, R.M.

The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems.

Applicability of health physics lessons learned from the Three Mile Island Unit 2 accident to the Fukushima Daiichi accident.

PubMed

Bevelacqua, J J

2012-02-01

The TMI-2 and Fukushima Daiichi accidents appear to be dissimilar because they involve different reactor types. However, the health physics related lessons learned from TMI-2 are applicable, and can enhance the Fukushima Daiichi recovery effort. Copyright © 2011 Elsevier Ltd. All rights reserved.

Simulation of hydrostatic water level measuring system for pressure vessels with the ATHLET-code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hampel, R.; Vandreier, B.; Kaestner, W.

1996-11-01

The static and dynamic behavior of measuring systems determine the value indicated by the measuring systems in relation to the true operating conditions. This paper demonstrates the necessity to involve the behavior of measuring systems in accident analysis with the thermohydraulic code ATHLET (developed by GRS Germany) by the example of hydrostatic water level measurement for horizontal steam generators on NPP (VVER). The modelling of a comparison vessel for the level measuring system with high sensitivity and a limited range of measurement (narrow-range level measuring system) by using ATHLET components and the checking of the function of the module weremore » realized. A good correspondence (maximal deviation 3%) between the measured and calculated narrow-range water level by the module was obtained for a realized post calculation of a measured operational transient in a NPP (VVER). The research carried out was sponsored by the Federal Ministry for Research and Technology within the projects {open_quotes}Basic research of process and system behaviour of NPP, control technique for accident management{close_quotes} (Project number 150 0855/7) and the project RS 978. The research work appertains to the theoretic and experimental work of institute {open_quotes}Institut fuer ProzeBtechnik, ProzeBautomatisierung und MeBtechnik (IPM){close_quotes} for accident analysis and accident management.« less

MELCOR Applications to SOARCA and Fukushima

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gauntt, Randall O.

2014-03-01

This PowerPoint presentation was organized as follows: Background; Overview of Fukushima Accidents; Comparisons of SOARCA Study with Fukushima accidents; Equipment functioning in real-world accidents; and, Conclusions.

Translation of the Department of Defense Disease and Injury Codes to the Eighth Revision International Classification of Diseases for use by the Military Services.

DTIC Science & Technology

1976-06-01

with, the DDDIC entity. 3. The ICDA-8 contracts groups of diseases or accidents which had been presented in expanded form in the DDDIC. Example: DDDIC...DDDIC. 4. The ICDA-8 expands groups of entities which had been presented in more condensed folin in the DDDIC. Example: DDDIC ICDA-8 Code Ntmuber Code...rapidly find a disease entity and all closely related entities. At the Naval Health Research Center (N11RC) a new code nunber was given to each

Multi-scale approach to the modeling of fission gas discharge during hypothetical loss-of-flow accident in gen-IV sodium fast reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Behafarid, F.; Shaver, D. R.; Bolotnov, I. A.

The required technological and safety standards for future Gen IV Reactors can only be achieved if advanced simulation capabilities become available, which combine high performance computing with the necessary level of modeling detail and high accuracy of predictions. The purpose of this paper is to present new results of multi-scale three-dimensional (3D) simulations of the inter-related phenomena, which occur as a result of fuel element heat-up and cladding failure, including the injection of a jet of gaseous fission products into a partially blocked Sodium Fast Reactor (SFR) coolant channel, and gas/molten sodium transport along the coolant channels. The computational approachmore » to the analysis of the overall accident scenario is based on using two different inter-communicating computational multiphase fluid dynamics (CMFD) codes: a CFD code, PHASTA, and a RANS code, NPHASE-CMFD. Using the geometry and time history of cladding failure and the gas injection rate, direct numerical simulations (DNS), combined with the Level Set method, of two-phase turbulent flow have been performed by the PHASTA code. The model allows one to track the evolution of gas/liquid interfaces at a centimeter scale. The simulated phenomena include the formation and breakup of the jet of fission products injected into the liquid sodium coolant. The PHASTA outflow has been averaged over time to obtain mean phasic velocities and volumetric concentrations, as well as the liquid turbulent kinetic energy and turbulence dissipation rate, all of which have served as the input to the core-scale simulations using the NPHASE-CMFD code. A sliding window time averaging has been used to capture mean flow parameters for transient cases. The results presented in the paper include testing and validation of the proposed models, as well the predictions of fission-gas/liquid-sodium transport along a multi-rod fuel assembly of SFR during a partial loss-of-flow accident. (authors)« less

Evaluation of radiological dispersion/consequence codes supporting DOE nuclear facility SARs

DOE Office of Scientific and Technical Information (OSTI.GOV)

O`Kula, K.R.; Paik, I.K.; Chung, D.Y.

1996-12-31

Since the early 1990s, the authorization basis documentation of many U.S. Department of Energy (DOE) nuclear facilities has been upgraded to comply with DOE orders and standards. In this process, many safety analyses have been revised. Unfortunately, there has been nonuniform application of software, and the most appropriate computer and engineering methodologies often are not applied. A DOE Accident Phenomenology and Consequence (APAC) Methodology Evaluation Program was originated at the request of DOE Defense Programs to evaluate the safety analysis methodologies used in nuclear facility authorization basis documentation and to define future cost-effective support and development initiatives. Six areas, includingmore » source term development (fire, spills, and explosion analysis), in-facility transport, and dispersion/ consequence analysis (chemical and radiological) are contained in the APAC program. The evaluation process, codes considered, key results, and recommendations for future model and software development of the Radiological Dispersion/Consequence Working Group are summarized in this paper.« less

RMP Guidance for Chemical Distributors - Appendix A: 40 CFR part 68/Selected NAICS Codes

EPA Pesticide Factsheets

The full text of Part 68, Chemical Accident Prevention provisions, includes hazard assessment, emergency response, threshold quantities for regulated substances, reporting requirements, and the Risk Management Plan.

BNL severe-accident sequence experiments and analysis program. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Greene, G.A.; Ginsberg, T.; Tutu, N.K.

1983-01-01

In the analysis of degraded core accidents, the two major sources of pressure loading on light water reactor containments are: steam generation from core debris-water thermal interactions; and molten core-concrete interactions. Experiments are in progress at BNL in support of analytical model development related to aspects of the above containment loading mechanisms. The work supports development and evaluation of the CORCON (Muir, 1981) and MARCH (Wooton, 1980) computer codes. Progress in the two programs is described.

Calculations of the skyshine gamma-ray dose rates from independent spent fuel storage installations (ISFSI) under worst case accident conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pace, J.V. III; Cramer, S.N.; Knight, J.R.

1980-09-01

Calculations of the skyshine gamma-ray dose rates from three spent fuel storage pools under worst case accident conditions have been made using the discrete ordinates code DOT-IV and the Monte Carlo code MORSE and have been compared to those of two previous methods. The DNA 37N-21G group cross-section library was utilized in the calculations, together with the Claiborne-Trubey gamma-ray dose factors taken from the same library. Plots of all results are presented. It was found that the dose was a strong function of the iron thickness over the fuel assemblies, the initial angular distribution of the emitted radiation, and themore » photon source near the top of the assemblies. 16 refs., 11 figs., 7 tabs.« less

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-23

... operators of Liquefied Natural Gas (LNG) facilities to conduct post- accident drug and alcohol tests of... reviewed, along with other applicable sections of Part 199: Under Sec. 199.105, post-accident drug tests of... administering the test. Covered employees must remain available for post-accident testing, but emergency...

10 CFR 71.73 - Hypothetical accident conditions.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Hypothetical accident conditions. 71.73 Section 71.73... Package, Special Form, and LSA-III Tests 2 § 71.73 Hypothetical accident conditions. (a) Test procedures. Evaluation for hypothetical accident conditions is to be based on sequential application of the tests...

10 CFR 50.67 - Accident source term.

Code of Federal Regulations, 2014 CFR

2014-01-01

... occupancy of the control room under accident conditions without personnel receiving radiation exposures in... 10 Energy 1 2014-01-01 2014-01-01 false Accident source term. 50.67 Section 50.67 Energy NUCLEAR... Conditions of Licenses and Construction Permits § 50.67 Accident source term. (a) Applicability. The...

10 CFR 50.67 - Accident source term.

Code of Federal Regulations, 2012 CFR

2012-01-01

... occupancy of the control room under accident conditions without personnel receiving radiation exposures in... 10 Energy 1 2012-01-01 2012-01-01 false Accident source term. 50.67 Section 50.67 Energy NUCLEAR... Conditions of Licenses and Construction Permits § 50.67 Accident source term. (a) Applicability. The...

10 CFR 50.67 - Accident source term.

Code of Federal Regulations, 2010 CFR

2010-01-01

... occupancy of the control room under accident conditions without personnel receiving radiation exposures in... 10 Energy 1 2010-01-01 2010-01-01 false Accident source term. 50.67 Section 50.67 Energy NUCLEAR... Conditions of Licenses and Construction Permits § 50.67 Accident source term. (a) Applicability. The...

10 CFR 50.67 - Accident source term.

Code of Federal Regulations, 2013 CFR

2013-01-01

... occupancy of the control room under accident conditions without personnel receiving radiation exposures in... 10 Energy 1 2013-01-01 2013-01-01 false Accident source term. 50.67 Section 50.67 Energy NUCLEAR... Conditions of Licenses and Construction Permits § 50.67 Accident source term. (a) Applicability. The...

10 CFR 50.67 - Accident source term.

Code of Federal Regulations, 2011 CFR

2011-01-01

... occupancy of the control room under accident conditions without personnel receiving radiation exposures in... 10 Energy 1 2011-01-01 2011-01-01 false Accident source term. 50.67 Section 50.67 Energy NUCLEAR... Conditions of Licenses and Construction Permits § 50.67 Accident source term. (a) Applicability. The...

Natural Circulation Level Optimization and the Effect during ULOF Accident in the SPINNOR Reactors

NASA Astrophysics Data System (ADS)

Abdullah, Ade Gafar; Su'ud, Zaki; Kurniadi, Rizal; Kurniasih, Neny; Yulianti, Yanti

2010-12-01

Natural circulation level optimization and the effect during loss of flow accident in the 250 MWt MOX fuelled small Pb-Bi Cooled non-refueling nuclear reactors (SPINNOR) have been performed. The simulation was performed using FI-ITB safety code which has been developed in ITB. The simulation begins with steady state calculation of neutron flux, power distribution and temperature distribution across the core, hot pool and cool pool, and also steam generator. When the accident is started due to the loss of pumping power the power distribution and the temperature distribution of core, hot pool and cool pool, and steam generator change. Then the feedback reactivity calculation is conducted, followed by kinetic calculation. The process is repeated until the optimum power distribution is achieved. The results show that the SPINNOR reactor has inherent safety capability against this accident.

Exploring inattention and distraction in the SafetyNet Accident Causation Database.

PubMed

Talbot, Rachel; Fagerlind, Helen; Morris, Andrew

2013-11-01

Distraction and inattention are considered to be very important and prevalent factors in the causation of road accidents. There have been many recent research studies which have attempted to understand the circumstances under which a driver becomes distracted or inattentive and how distraction/inattention can be prevented. Both factors are thought to have become more important in recent times partly due to the evolution of in-vehicle information and communication technology. This study describes a methodology that was developed to understand when factors such as distraction and inattention may have been contributors to crashes and also describes some of the consequences of distraction and inattention in terms of subsequent driver actions. The study uses data relating to distraction and inattention from the SafetyNet Accident Causation Database. This database was formulated as part of the SafetyNet project to address the lack of representative in-depth accident causation data within the European Union. Data were collected in 6 European countries using 'on-scene' and 'nearly on-scene' crash investigation methodologies. 32% of crashes recorded in the database, involved at least one driver, rider or pedestrian, who was determined to be 'Inattentive' or 'Distracted'. 212 of the drivers were assigned 'Distraction' and 140 drivers were given the code 'Inattention'. It was found that both distraction and inattention often lead to missed observations within the driving task and consequently 'Timing' or 'Direction' become critical events in the aetiology of crashes. In addition, the crash types and outcomes may differ according to the type and nature of the distraction and inattention as determined by the in-depth investigations. The development of accident coding methodology is described in this study as is its evolution into the Driver Reliability and Error Analysis Model (DREAM) version 3.0. Copyright © 2012 Elsevier Ltd. All rights reserved.

25 CFR 181.5 - How are applications ranked?

Code of Federal Regulations, 2010 CFR

2010-04-01

... resolution of the identified highway safety problem. (4) The number of traffic accidents occurring within the applicant's jurisdiction over the previous 3 years. (5) The number of alcohol-related traffic accidents occurring within the applicant's jurisdiction over the previous 3 years. (6) The number of reported traffic...

General RMP Guidance - Appendix A: 40 CFR 68

EPA Pesticide Factsheets

Here the full text of Chemical Accident Prevention Provisions and Risk Management Program is transcribed directly from the Code of Federal Regulations. Subparts include hazard assessment, regulated substances and thresholds, and risk management plan.

Investigation of accidents within construction zones in Louisiana.

DOT National Transportation Integrated Search

1981-07-01

This investigation is to analyze construction and maintenance work zone accidents by reviewing accident data to determine if deficiencies exist and recommend possible corrective measures for future traffic control applications. To accomplish this, a ...

SOARCA Peach Bottom Atomic Power Station Long-Term Station Blackout Uncertainty Analysis: Knowledge Advancement.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gauntt, Randall O.; Mattie, Patrick D.; Bixler, Nathan E.

2014-02-01

This paper describes the knowledge advancements from the uncertainty analysis for the State-of- the-Art Reactor Consequence Analyses (SOARCA) unmitigated long-term station blackout accident scenario at the Peach Bottom Atomic Power Station. This work assessed key MELCOR and MELCOR Accident Consequence Code System, Version 2 (MACCS2) modeling uncertainties in an integrated fashion to quantify the relative importance of each uncertain input on potential accident progression, radiological releases, and off-site consequences. This quantitative uncertainty analysis provides measures of the effects on consequences, of each of the selected uncertain parameters both individually and in interaction with other parameters. The results measure the modelmore » response (e.g., variance in the output) to uncertainty in the selected input. Investigation into the important uncertain parameters in turn yields insights into important phenomena for accident progression and off-site consequences. This uncertainty analysis confirmed the known importance of some parameters, such as failure rate of the Safety Relief Valve in accident progression modeling and the dry deposition velocity in off-site consequence modeling. The analysis also revealed some new insights, such as dependent effect of cesium chemical form for different accident progressions. (auth)« less

Development and Assessment of CFD Models Including a Supplemental Program Code for Analyzing Buoyancy-Driven Flows Through BWR Fuel Assemblies in SFP Complete LOCA Scenarios

NASA Astrophysics Data System (ADS)

Artnak, Edward Joseph, III

This work seeks to illustrate the potential benefits afforded by implementing aspects of fluid dynamics, especially the latest computational fluid dynamics (CFD) modeling approach, through numerical experimentation and the traditional discipline of physical experimentation to improve the calibration of the severe reactor accident analysis code, MELCOR, in one of several spent fuel pool (SFP) complete loss-ofcoolant accident (LOCA) scenarios. While the scope of experimental work performed by Sandia National Laboratories (SNL) extends well beyond that which is reasonably addressed by our allotted resources and computational time in accordance with initial project allocations to complete the report, these simulated case trials produced a significant array of supplementary high-fidelity solutions and hydraulic flow-field data in support of SNL research objectives. Results contained herein show FLUENT CFD model representations of a 9x9 BWR fuel assembly in conditions corresponding to a complete loss-of-coolant accident scenario. In addition to the CFD model developments, a MATLAB based controlvolume model was constructed to independently assess the 9x9 BWR fuel assembly under similar accident scenarios. The data produced from this work show that FLUENT CFD models are capable of resolving complex flow fields within a BWR fuel assembly in the realm of buoyancy-induced mass flow rates and that characteristic hydraulic parameters from such CFD simulations (or physical experiments) are reasonably employed in corresponding constitutive correlations for developing simplified numerical models of comparable solution accuracy.

46 CFR 4.03-1 - Marine casualty or accident.

Code of Federal Regulations, 2013 CFR

2013-10-01

... 46 Shipping 1 2013-10-01 2013-10-01 false Marine casualty or accident. 4.03-1 Section 4.03-1 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY PROCEDURES APPLICABLE TO THE PUBLIC MARINE CASUALTIES AND INVESTIGATIONS Definitions § 4.03-1 Marine casualty or accident. Marine casualty or accident means...

46 CFR 4.03-1 - Marine casualty or accident.

Code of Federal Regulations, 2014 CFR

2014-10-01

... 46 Shipping 1 2014-10-01 2014-10-01 false Marine casualty or accident. 4.03-1 Section 4.03-1 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY PROCEDURES APPLICABLE TO THE PUBLIC MARINE CASUALTIES AND INVESTIGATIONS Definitions § 4.03-1 Marine casualty or accident. Marine casualty or accident means...

46 CFR 4.03-1 - Marine casualty or accident.

Code of Federal Regulations, 2012 CFR

2012-10-01

... 46 Shipping 1 2012-10-01 2012-10-01 false Marine casualty or accident. 4.03-1 Section 4.03-1 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY PROCEDURES APPLICABLE TO THE PUBLIC MARINE CASUALTIES AND INVESTIGATIONS Definitions § 4.03-1 Marine casualty or accident. Marine casualty or accident means...

Investigations on optimization of accident management measures following a station blackout accident in a VVER-1000 pressurized water reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tusheva, P.; Schaefer, F.; Kliem, S.

2012-07-01

The reactor safety issues are of primary importance for preserving the health of the population and ensuring no release of radioactivity and fission products into the environment. A part of the nuclear research focuses on improvement of the safety of existing nuclear power plants. Studies, research and efforts are a continuing process at improving the safety and reliability of existing and newly developed nuclear power plants at prevention of a core melt accident. Station blackout (loss of AC power supply) is one of the dominant accidents taken into consideration at performing accident analysis. In case of multiple failures of safetymore » systems it leads to a severe accident. To prevent an accident to turn into a severe one or to mitigate the consequences, accident management measures must be performed. The present paper outlines possibilities for application and optimization of accident management measures following a station blackout accident. Assessed is the behaviour of the nuclear power plant during a station blackout accident without accident management measures and with application of primary/secondary side oriented accident management measures. Discussed are the possibilities for operators ' intervention and the influence of the performed accident management measures on the course of the accident. Special attention has been paid to the effectiveness of the passive feeding and physical phenomena having an influence on the system behaviour. The performed simulations show that the effectiveness of the secondary side feeding procedure can be limited due to an early evaporation or flashing effects in the feed water system. The analyzed cases show that the effectiveness of the accident management measures strongly depends on the initiation criteria applied for depressurization of the reactor coolant system. (authors)« less

Fukushima Daiichi Radionuclide Inventories

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cardoni, Jeffrey N.; Jankovsky, Zachary Kyle

Radionuclide inventories are generated to permit detailed analyses of the Fukushima Daiichi meltdowns. This is necessary information for severe accident calculations, dose calculations, and source term and consequence analyses. Inventories are calculated using SCALE6 and compared to values predicted by international researchers supporting the OECD/NEA's Benchmark Study on the Accident at Fukushima Daiichi Nuclear Power Station (BSAF). Both sets of inventory information are acceptable for best-estimate analyses of the Fukushima reactors. Consistent nuclear information for severe accident codes, including radionuclide class masses and core decay powers, are also derived from the SCALE6 analyses. Key nuclide activity ratios are calculated asmore » functions of burnup and nuclear data in order to explore the utility for nuclear forensics and support future decommissioning efforts.« less

Qualification of safety-related electrical equipment in France. Methods, approach and test facilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Raimondo, E.; Capman, J.L.; Herovard, M.

1985-05-01

Requirements for qualification of electrical equipment used in French-built nuclear power plants are stated in a national code, the RCC-E, or Regles de Construction et de Conception des Materiels Electriques. Under the RCC-E, safety related equipment is assigned to one of three different categories, according to location in the plant and anticipated normal, accident and post-accident behavior. Qualification tests differ for each category and procedures range in scope from the standard seismic test to the highly stringent VISA program, which specifies a predetermined sequence of aging, radiation, seismic and simulated accident testing. A network of official French test facilities wasmore » developed specifically to meet RCC-E requirements.« less

Summary and evaluation: fuel dynamics loss-of-flow experiments (tests L2, L3, and L4)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barts, E.W.; Deitrich, L.W.; Eberhart, J.G.

1975-09-01

Three similar experiments conducted to support the analyses of hypothetical LMFBR unprotected-loss-of-flow accidents are summarized and evaluated. The tests, designated L2, L3, and L4, provided experimental data against which accident-analysis codes could be compared, so as to guide further analysis and modeling of the initiating phases of the hypothetical accident. The tests were conducted using seven-pin bundles of mixed-oxide fuel pins in Mark-II flowing-sodium loops in the TREAT reactor. Test L2 used fresh fuel. Tests L3 and L4 used irradiated fuel pins having, respectively, ''intermediate-power'' (no central void) and ''high-power'' (fully developed central void) microstructure. 12 references. (auth)

Safer energetic materials by a nanotechnological approach

NASA Astrophysics Data System (ADS)

Siegert, Benny; Comet, Marc; Spitzer, Denis

2011-09-01

Energetic materials - explosives, thermites, populsive powders - are used in a variety of military and civilian applications. Their mechanical and electrostatic sensitivity is high in many cases, which can lead to accidents during handling and transport. These considerations limit the practical use of some energetic materials despite their good performance. For industrial applications, safety is one of the main criteria for selecting energetic materials. The sensitivity has been regarded as an intrinsic property of a substance for a long time. However, in recent years, several approaches to lower the sensitivity of a given substance, using nanotechnology and materials engineering, have been described. This feature article gives an overview over ways to prepare energetic (nano-)materials with a lower sensitivity.Energetic materials - explosives, thermites, populsive powders - are used in a variety of military and civilian applications. Their mechanical and electrostatic sensitivity is high in many cases, which can lead to accidents during handling and transport. These considerations limit the practical use of some energetic materials despite their good performance. For industrial applications, safety is one of the main criteria for selecting energetic materials. The sensitivity has been regarded as an intrinsic property of a substance for a long time. However, in recent years, several approaches to lower the sensitivity of a given substance, using nanotechnology and materials engineering, have been described. This feature article gives an overview over ways to prepare energetic (nano-)materials with a lower sensitivity. Electronic supplementary information (ESI) available: Experimental details for the preparation of the V2O5@CNF/Al nanothermite; X-ray diffractogram of the V2O5@CNF/Al combustion residue; installation instructions and source code for the nt-timeline program. See DOI: 10.1039/c1nr10292c

Analysis of Shahid Rajaee hospital administrative data on injuries resulting from car accidents in Shiraz, Iran: 2011-2014 data.

PubMed

Yadollahi, Mahnaz; Ghiassee, Aida; Anvar, Mehrdad; Ghaem, Hale; Farahmand, Mohammad

2017-02-01

The administrative data from trauma centers could serve as potential sources of invaluable information while studying epidemiologic features of car accidents. In this cross-sectional analysis of Shahid Rajaee hospital administrative data, we aimed to evaluate patients injured in car accidents in terms of age, gender, injury severity, injured body regions and hospitalization outcome in the recent four years (2011-2014). The hospital registry was accessed at Shiraz Trauma Research Center (Shiraz, Iran) and the admission's unit data were merged with the information gathered upon discharge. A total number of 27,222 car accident patients aged over 15 years with International Classification of Diseases 10th revision (ICD-10) external causes of injury codes (V40.9-V49.9) were analyzed. Injury severity score and injured body regions were determined based on converting ICD-10 injury codes to Abbreviated Injury Scale (AIS-98) severity codes using a domestically developed electronic algorithm. A binary logistic regression model was applied to the data to examine the contribution of all independent variables to in-hospital mortality. Men accounted for 68.9% of the injuries and the male to female ratio was 2.2:1. The age of the studied population was (34 ± 15) years, with more than 77.2% of the population located in the 15-45 years old age group. Head and neck was the most commonly injured body region (39.0%) followed by extremities (27.2%). Injury severity score (ISS) was calculated for 13,152 (48.3%) patients, of whom, 80.9% had severity scores less than 9. There were 332 patients (1.2%) admitted to the intensive care units and 422 in-hospital fatalities (1.5%) were recorded during the study period. Age above 65 years [OR = 7.4, 95% CI (5.0-10.9)], ISS above 16 [OR = 9.1, 95% CI (5.5-14.9)], sustaining a thoracic injury [OR = 7.4, 95% CI (4.6-11.9)] and head injury [OR = 4.9, 95% CI (3.1-7.6)] were the most important independent predictors of death following car accidents. Hospital administrative databases of this hospital could be used as reliable sources of information in providing epidemiologic reports of car accidents in terms of severity and outcomes. Improving the quality of recordings at hospital databases is an important initial step towards more comprehensive injury surveillance in Fars, Iran. Copyright © 2017. Production and hosting by Elsevier B.V.

The practical application of mishap data in Army aircraft system safety programs

NASA Technical Reports Server (NTRS)

Darrah, J. T., Jr.

1971-01-01

The means are discussed by which the the United States Army Board for Aviation Accident Research (USABAAR) now utilizes the vast store of historical accident data in the application of the system safety concept for developmental aircraft. USABAAR serves as the central agency for the Army Accident Prevention Program which includes the receipt, processing, and analysis of all data and information related to Army aircraft accident experience. It is pointed out that methods which served the cause of accident prevention so well in the past are no longer adequate and that traditional parameters used to measure mishap experience have become obsolete. USABAAR has developed, and recently put into use, completely revised accident reporting forms which greatly expand the scope and detail of information provided as a result of investigation. This and other factors which have resulted in an improved data system are discussed in detail.

Investigation of Containment Flooding Strategy for Mark-III Nuclear Power Plant with MAAP4

DOE Office of Scientific and Technical Information (OSTI.GOV)

Su Weinian; Wang, S.-J.; Chiang, S.-C

2005-06-15

Containment flooding is an important strategy for severe accident management of a conventional boiling water reactor (BWR) system. The purpose of this work is to investigate the containment flooding strategy of the Mark-III system after a reactor pressure vessel (RPV) breach. The Kuosheng Power Plant is a typical BWR-6 nuclear power plant (NPP) with Mark-III containment. The Severe Accident Management Guideline (SAMG) of the Kuosheng NPP has been developed based on the BWR Owners Group (BWROG) Emergency Procedure and Severe Accident Guidelines, Rev. 2. Therefore, the Kuosheng NPP is selected as the plant for study, and the MAAP4 code ismore » chosen as the tool for analysis. A postulated specific station blackout sequence for the Kuosheng NPP is cited as a reference case for this analysis. Because of the design features of Mark-III containment, the debris in the reactor cavity may not be submerged after an RPV breach when one follows the containment flooding strategy as suggested in the BWROG generic guideline, and the containment integrity could be challenged eventually. A more specific containment flooding strategy with drywell venting after an RPV breach is investigated, and a more stable plant condition is achieved with this strategy. Accordingly, the containment flooding strategy after an RPV breach will be modified for the Kuosheng SAMG, and these results are applicable to typical Mark-III plants with drywell vent path.« less

29 CFR Appendix A to Subpart Q of... - References to subpart Q of Part 1926

Code of Federal Regulations, 2013 CFR

2013-07-01

.... • Accident Prevention Manual for Industrial Operations; Eighth Edition; National Safety Council. • Building Code Requirements for Reinforced Concrete (ACI 318-83). • Formwork for Concrete (ACI SP-4...

29 CFR Appendix A to Subpart Q of... - References to subpart Q of Part 1926

Code of Federal Regulations, 2014 CFR

2014-07-01

.... • Accident Prevention Manual for Industrial Operations; Eighth Edition; National Safety Council. • Building Code Requirements for Reinforced Concrete (ACI 318-83). • Formwork for Concrete (ACI SP-4...

29 CFR Appendix A to Subpart Q of... - References to subpart Q of Part 1926

Code of Federal Regulations, 2011 CFR

2011-07-01

.... • Accident Prevention Manual for Industrial Operations; Eighth Edition; National Safety Council. • Building Code Requirements for Reinforced Concrete (ACI 318-83). • Formwork for Concrete (ACI SP-4...

29 CFR Appendix A to Subpart Q of... - References to subpart Q of Part 1926

Code of Federal Regulations, 2012 CFR

2012-07-01

.... • Accident Prevention Manual for Industrial Operations; Eighth Edition; National Safety Council. • Building Code Requirements for Reinforced Concrete (ACI 318-83). • Formwork for Concrete (ACI SP-4...

Flow diagram analysis of electrical fatalities in construction industry.

PubMed

Chi, Chia-Fen; Lin, Yuan-Yuan; Ikhwan, Mohamad

2012-01-01

The current study reanalyzed 250 electrical fatalities in the construction industry from 1996 to 2002 into seven patterns based on source of electricity (power line, energized equipment, improperly installed or damaged equipment), direct contact or indirect contact through some source of injury (boom vehicle, metal bar or pipe, and other conductive material). Each fatality was coded in terms of age, company size, experience, performing tasks, source of injury, accident cause and hazard pattern. The Chi-square Automatic Interaction Detector (CHAID) was applied to the coded data of the fatal electrocution to find a subset of predictors that might derive meaningful classifications or accidents scenarios. A series of Flow Diagrams was constructed based on CHAID result to illustrate the flow of electricity travelling from electrical source to human body. Each of the flow diagrams can be directly linked with feasible prevention strategies by cutting the flow of electricity.

Initial Neutronics Analyses for HEU to LEU Fuel Conversion of the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kontogeorgakos, D.; Derstine, K.; Wright, A.

2013-06-01

The purpose of the TREAT reactor is to generate large transient neutron pulses in test samples without over-heating the core to simulate fuel assembly accident conditions. The power transients in the present HEU core are inherently self-limiting such that the core prevents itself from overheating even in the event of a reactivity insertion accident. The objective of this study was to support the assessment of the feasibility of the TREAT core conversion based on the present reactor performance metrics and the technical specifications of the HEU core. The LEU fuel assembly studied had the same overall design, materials (UO 2more » particles finely dispersed in graphite) and impurities content as the HEU fuel assembly. The Monte Carlo N–Particle code (MCNP) and the point kinetics code TREKIN were used in the analyses.« less

Aircraft crashworthiness studies : findings in accidents involving an aerial application aircraft.

DOT National Transportation Integrated Search

1980-04-01

Aircraft crashworthiness features are presented, as others have done, in terms of packaging principles. Modern aerial application aircraft are recognized as being the most crashworthy in the civil aviation fleet. Eighteen accidents involving an aeria...

Multi-phase model development to assess RCIC system capabilities under severe accident conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kirkland, Karen Vierow; Ross, Kyle; Beeny, Bradley

The Reactor Core Isolation Cooling (RCIC) System is a safety-related system that provides makeup water for core cooling of some Boiling Water Reactors (BWRs) with a Mark I containment. The RCIC System consists of a steam-driven Terry turbine that powers a centrifugal, multi-stage pump for providing water to the reactor pressure vessel. The Fukushima Dai-ichi accidents demonstrated that the RCIC System can play an important role under accident conditions in removing core decay heat. The unexpectedly sustained, good performance of the RCIC System in the Fukushima reactor demonstrates, firstly, that its capabilities are not well understood, and secondly, that themore » system has high potential for extended core cooling in accident scenarios. Better understanding and analysis tools would allow for more options to cope with a severe accident situation and to reduce the consequences. The objectives of this project were to develop physics-based models of the RCIC System, incorporate them into a multi-phase code and validate the models. This Final Technical Report details the progress throughout the project duration and the accomplishments.« less

78 FR 6145 - Virgil C. Summer Nuclear Station, Units 2 and 3; Application and Amendment to Facility Combined...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-01-29

... increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a... proposed amendment involve a significant increase in the probability or consequences of an accident...

Application and Evaluation of Rumble Strips on Highways

DOT National Transportation Integrated Search

2001-01-01

In Utah, single vehicle accidents account for about 30 percent of total accidents each year. These single vehicle accidents are normally caused by vehicles first leaving the travel lane, and then either hitting various roadside objects or becoming ov...

Systems-based accident analysis in the led outdoor activity domain: application and evaluation of a risk management framework.

PubMed

Salmon, P; Williamson, A; Lenné, M; Mitsopoulos-Rubens, E; Rudin-Brown, C M

2010-08-01

Safety-compromising accidents occur regularly in the led outdoor activity domain. Formal accident analysis is an accepted means of understanding such events and improving safety. Despite this, there remains no universally accepted framework for collecting and analysing accident data in the led outdoor activity domain. This article presents an application of Rasmussen's risk management framework to the analysis of the Lyme Bay sea canoeing incident. This involved the development of an Accimap, the outputs of which were used to evaluate seven predictions made by the framework. The Accimap output was also compared to an analysis using an existing model from the led outdoor activity domain. In conclusion, the Accimap output was found to be more comprehensive and supported all seven of the risk management framework's predictions, suggesting that it shows promise as a theoretically underpinned approach for analysing, and learning from, accidents in the led outdoor activity domain. STATEMENT OF RELEVANCE: Accidents represent a significant problem within the led outdoor activity domain. This article presents an evaluation of a risk management framework that can be used to understand such accidents and to inform the development of accident countermeasures and mitigation strategies for the led outdoor activity domain.

Brief Overlook on the Occupational Accidents Occurring During the Geotechnical Site Works

NASA Astrophysics Data System (ADS)

Akboğa Kale, Özge; Eskişar, Tuğba

2017-10-01

The aim of this paper is to evaluate occupational accidents reported in geotechnical site works. Variables of the accidents are categorized as the year and month of accidents, the technical codes used for defining the scope of work trades, end use and project type and cost, nature and cause of accidents, occupation of the victims and finally the cause of fatality. As a result, it is seen that the majority of victims were construction laborers or in special trade constructors who were working on a new project or new additions to an existing project. The geotechnical phase of the projects was whether excavation, landfill, sewer-water treatment, pipeline construction, commercial building or road construction. As the outcomes of the study it is evaluated that excavation, trenching and installing pipe or pile driving were the main causes of the accidents while trench collapse, struck by a falling object / projectile and wall collapse were the main causes of fatality. Moreover, it is established that more than half of the fatalities were due to asphyxia followed by fracture. These findings show that accidents occurred in geotechnical works do not only have high frequency but also high severity. This study emphasizes project specific countermeasures should be taken regarding the nature, cost and importance of the project and the occupation variabilities working on the project.

Recent MELCOR and VICTORIA Fission Product Research at the NRC

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bixler, N.E.; Cole, R.K.; Gauntt, R.O.

1999-01-21

The MELCOR and VICTORIA severe accident analysis codes, which were developed at Sandia National Laboratories for the U. S. Nuclear Regulatory Commission, are designed to estimate fission product releases during nuclear reactor accidents in light water reactors. MELCOR is an integrated plant-assessment code that models the key phenomena in adequate detail for risk-assessment purposes. VICTORIA is a more specialized fission- product code that provides detailed modeling of chemical reactions and aerosol processes under the high-temperature conditions encountered in the reactor coolant system during a severe reactor accident. This paper focuses on recent enhancements and assessments of the two codes inmore » the area of fission product chemistry modeling. Recently, a model for iodine chemistry in aqueous pools in the containment building was incorporated into the MELCOR code. The model calculates dissolution of iodine into the pool and releases of organic and inorganic iodine vapors from the pool into the containment atmosphere. The main purpose of this model is to evaluate the effect of long-term revolatilization of dissolved iodine. Inputs to the model include dose rate in the pool, the amount of chloride-containing polymer, such as Hypalon, and the amount of buffering agents in the containment. Model predictions are compared against the Radioiodine Test Facility (RTF) experiments conduced by Atomic Energy of Canada Limited (AECL), specifically International Standard Problem 41. Improvements to VICTORIA's chemical reactions models were implemented as a result of recommendations from a peer review of VICTORIA that was completed last year. Specifically, an option is now included to model aerosols and deposited fission products as three condensed phases in addition to the original option of a single condensed phase. The three-condensed-phase model results in somewhat higher predicted fission product volatilities than does the single-condensed-phase model. Modeling of U02 thermochemistry was also improved, and results in better prediction of vaporization of uranium from fuel, which can react with released fission products to affect their volatility. This model also improves the prediction of fission product release rates from fuel. Finally, recent comparisons of MELCOR and VICTORIA with International Standard Problem 40 (STORM) data are presented. These comparisons focus on predicted therrnophoretic deposition, which is the dominant deposition mechanism. Sensitivity studies were performed with the codes to examine experimental and modeling uncertainties.« less

Classification scheme and prevention measures for caught-in-between occupational fatalities.

PubMed

Chi, Chia-Fen; Lin, Syuan-Zih

2018-04-01

The current study analyzed 312 caught-in-between fatalities caused by machinery and vehicles. A comprehensive and mutually exclusive coding scheme was developed to analyze and code each caught-in-between fatality in terms of age, gender, experience of the victim, type of industry, source of injury, and causes for these accidents. Boolean algebra analysis was applied on these 312 caught-in-between fatalities to derive minimal cut set (MCS) causes associated with each source of injury. Eventually, contributing factors and common accident patterns associated with (1) special process machinery including textile, printing, packaging machinery, (2) metal, woodworking, and special material machinery, (3) conveyor, (4) vehicle, (5) crane, (6) construction machinery, and (7) elevator can be divided into three major groups through Boolean algebra and MCS analysis. The MCS causes associated with conveyor share the same primary causes as those of the special process machinery including textile, printing, packaging and metal, woodworking, and special material machinery. These fatalities can be eliminated by focusing on the prevention measures associated with lack of safeguards, working on a running machine or process, unintentional activation, unsafe posture or position, unsafe clothing, and defective safeguards. Other precise and effective intervention can be developed based on the identified groups of accident causes associated with each source of injury. Copyright © 2017 Elsevier Ltd. All rights reserved.

Station Blackout at Browns Ferry Unit One - accident sequence analysis. Volume 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cook, D.H.; Harrington, R.M.; Greene, S.R.

1981-11-01

This study describes the predicted response of Unit 1 at the Browns Ferry Nuclear Plant to Station Blackout, defined as a loss of offsite power combined with failure of all onsite emergency diesel-generators to start and load. Every effort has been made to employ the most realistic assumptions during the process of defining the sequence of events for this hypothetical accident. DC power is assumed to remain available from the unit batteries during the initial phase and the operator actions and corresponding events during this period are described using results provided by an analysis code developed specifically for this purpose.more » The Station Blackout is assumed to persist beyond the point of battery exhaustion and the events during this second phase of the accident in which dc power would be unavailable were determined through use of the MARCH code. Without dc power, cooling water could no longer be injected into the reactor vessel and the events of the second phase include core meltdown and subsequent containment failure. An estimate of the magnitude and timing of the concomitant release of the noble gas, cesium, and iodine-based fission products to the environment is provided in Volume 2 of this report. 58 refs., 75 figs., 8 tabs.« less

40 CFR 68.20 - Applicability.

Code of Federal Regulations, 2010 CFR

2010-07-01

... Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Hazard Assessment § 68.20 Applicability. The owner or operator of a... § 68.25 of this part and complete the five-year accident history as provided in § 68.42. The owner or...

Mechanisms of and facility types involved in hazardous materials incidents.

PubMed Central

Kales, S N; Polyhronopoulos, G N; Castro, M J; Goldman, R H; Christiani, D C

1997-01-01

The purpose of this study was to systematically investigate hazardous materials (hazmat) releases and determine the mechanisms of these accidents, and the industries/activities and chemicals involved. We analyzed responses by Massachusetts' six district hazmat teams from their inception through May 1996. Information from incident reports was extracted onto standard coding sheets. The majority of hazardous materials incidents were caused by spills, leaks, or escapes of hazardous materials (76%) and occurred at fixed facilities (80%). Transportation-related accidents accounted for 20% of incidents. Eleven percent of hazardous materials incidents were at schools or health care facilities. Petroleum-derived fuels were involved in over half of transportation-related accidents, and these accounted for the majority of petroleum fuel releases. Chlorine derivatives were involved in 18% of all accidents and were associated with a wide variety of facility types and activities. In conclusion, systematic study of hazardous materials incidents allows the identification of preventable causes of these incidents. PMID:9300926

Thermal-hydraulic analysis of N Reactor graphite and shield cooling system performance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Low, J.O.; Schmitt, B.E.

1988-02-01

A series of bounding (worst-case) calculations were performed using a detailed hydrodynamic RELAP5 model of the N Reactor graphite and shield cooling system (GSCS). These calculations were specifically aimed to answer issues raised by the Westinghouse Independent Safety Review (WISR) committee. These questions address the operability of the GSCS during a worst-case degraded-core accident that requires the GDCS to mitigate the consequences of the accident. An accident scenario previously developed was designed as the hydrogen-mitigation design-basis accident (HMDBA). Previous HMDBA heat transfer analysis,, using the TRUMP-BD code, was used to define the thermal boundary conditions that the GSDS may bemore » exposed to. These TRUMP/HMDBA analysis results were used to define the bounding operating conditions of the GSCS during the course of an HMDBA transient. Nominal and degraded GSCS scenarios were investigated using RELAP5 within or at the bounds of the HMDBA transient. 10 refs., 42 figs., 10 tabs.« less

On-Board Failure-Protection Requirements for Railroad-Vehicle Equipment

DOT National Transportation Integrated Search

1979-03-01

An analysis of the 1975 railroad-equipment-caused accidents was made. Data reported to the FRA were the primary source of derailment information; however, data from other sources were also used. Individual cause codes were consolidated into groups wh...

Analysis of the FeCrAl Accident Tolerant Fuel Concept Benefits during BWR Station Blackout Accidents

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robb, Kevin R

2015-01-01

Iron-chromium-aluminum (FeCrAl) alloys are being considered for fuel concepts with enhanced accident tolerance. FeCrAl alloys have very slow oxidation kinetics and good strength at high temperatures. FeCrAl could be used for fuel cladding in light water reactors and/or as channel box material in boiling water reactors (BWRs). To estimate the potential safety gains afforded by the FeCrAl concept, the MELCOR code was used to analyze a range of postulated station blackout severe accident scenarios in a BWR/4 reactor employing FeCrAl. The simulations utilize the most recently known thermophysical properties and oxidation kinetics for FeCrAl. Overall, when compared to the traditionalmore » Zircaloy-based cladding and channel box, the FeCrAl concept provides a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. Finally, due to the slower oxidation kinetics, substantially less hydrogen is generated, and the generation is delayed in time. This decreases the amount of non-condensable gases in containment and the potential for deflagrations to inhibit the accident response.« less

[Fatal occupational accidents: updating of data from a mortality register].

PubMed

Mantero, Silvia; Baldasseroni, A; Chellini, Elisabetta; Giovanetti, Lucia

2005-01-01

In Italy, almost one thousand deaths due to occupational accidents are usually registered by INAIL each year. Case registration by INAIL has merely administrative purposes and therefore it is necessary to use other sources for case ascertainment in order to better estimate the real number of deaths related to occupational accidents, as shown also by previous papers. Evaluation of the contribution of another data source, namely the Tuscany Regional Mortality Registry, to obtain the correct figure for occupational accident deaths through the use of a place-of-occurrence notation on the death certificate. Cases that occurred in residents in Tuscany in 2000-2001 were considered. They were identified from : a) the Tuscany Regional Mortality Registry (RMR) using the E code of the ICD LX code of death, the year and place of occurrence; b) the INAIL archive using the year of event, the type of definition and management. The INAIL source was without doubt the most informative but was only 51% complete, whereas the RMR source, although less informative, was more complete (82.4%) and allowed identification of cases not registered by INAIL, that had occurred for instance in the Armed Forces and in the National Railway Company. However, the vast majority of RMR extra-cases occurred in subjects aged 65+, in agriculture and in the building industry. It is currently possible to plan a systematic linkage of the two sources due to the new possibilities that are available: the place-of-occurrence in the death certificate and the availability of individual data in the INAIL source.

United States Department of Energy severe accident research following the Fukushima Daiichi accidents

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farmer, M. T.; Corradini, M.; Rempe, J.

The U.S. Department of Energy (DOE) has played a major role in the U.S. response to the events at Fukushima Daiichi. During the first several weeks following the accident, U.S. assistance efforts were guided by results from a significant and diverse set of analyses. In the months that followed, a coordinated analysis activity aimed at gaining a more thorough understanding of the accident sequence was completed using laboratory-developed, system-level best-estimate accident analysis codes, while a parallel analysis was conducted by U.S. industry. A comparison of predictions for Unit 1 from these two studies indicated significant differences between MAAP and MELCORmore » results for key plant parameters, such as in-core hydrogen production. On that basis, a crosswalk was completed to determine the key modeling variations that led to these differences. In parallel with these activities, it became clear that there was a need to perform a technology gap evaluation on accident-tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist given the current state of light water reactor (LWR) severe accident research and augmented by insights from Fukushima. In addition, there is growing international recognition that data from Fukushima could significantly reduce uncertainties related to severe accident progression, particularly for boiling water reactors. On these bases, a group of U. S. experts in LWR safety and plant operations was convened by the DOE Office of Nuclear Energy (DOE-NE) to complete technology gap analysis and Fukushima forensics data needs identification activities. The results from these activities were used as the basis for refining DOE-NE's severe accident research and development (R&D) plan. Finally, this paper provides a high-level review of DOE-sponsored R&D efforts in these areas, including planned activities on accident-tolerant components and accident analysis methods.« less

United States Department of Energy severe accident research following the Fukushima Daiichi accidents

DOE PAGES

Farmer, M. T.; Corradini, M.; Rempe, J.; ...

2016-11-02

The U.S. Department of Energy (DOE) has played a major role in the U.S. response to the events at Fukushima Daiichi. During the first several weeks following the accident, U.S. assistance efforts were guided by results from a significant and diverse set of analyses. In the months that followed, a coordinated analysis activity aimed at gaining a more thorough understanding of the accident sequence was completed using laboratory-developed, system-level best-estimate accident analysis codes, while a parallel analysis was conducted by U.S. industry. A comparison of predictions for Unit 1 from these two studies indicated significant differences between MAAP and MELCORmore » results for key plant parameters, such as in-core hydrogen production. On that basis, a crosswalk was completed to determine the key modeling variations that led to these differences. In parallel with these activities, it became clear that there was a need to perform a technology gap evaluation on accident-tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist given the current state of light water reactor (LWR) severe accident research and augmented by insights from Fukushima. In addition, there is growing international recognition that data from Fukushima could significantly reduce uncertainties related to severe accident progression, particularly for boiling water reactors. On these bases, a group of U. S. experts in LWR safety and plant operations was convened by the DOE Office of Nuclear Energy (DOE-NE) to complete technology gap analysis and Fukushima forensics data needs identification activities. The results from these activities were used as the basis for refining DOE-NE's severe accident research and development (R&D) plan. Finally, this paper provides a high-level review of DOE-sponsored R&D efforts in these areas, including planned activities on accident-tolerant components and accident analysis methods.« less

Anatomy of an Accident.

ERIC Educational Resources Information Center

Mobley, Michael

1984-01-01

The findings of industrial safety engineers in the areas of accident causation and prevention are wholly applicable to adventure programs. Adventure education instructors can use safety engineering concepts to assess the risk in a particular activity, understand factors that cause accidents, and intervene to minimize injuries and damages if…

Squeal Those Tires! Automobile-Accident Reconstruction.

ERIC Educational Resources Information Center

Caples, Linda Griffin

1992-01-01

Methods use to reconstruct traffic accidents provide settings for real life applications for students in precalculus, mathematical analysis, or trigonometry. Described is the investigation of an accident in conjunction with the local Highway Patrol Academy integrating physics, vector, and trigonometry. Class findings were compared with those of…

10 CFR 70.24 - Criticality accident requirements.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Criticality accident requirements. 70.24 Section 70.24... Applications § 70.24 Criticality accident requirements. (a) Each licensee authorized to possess special nuclear...-sensitive radiation detectors which will energize clearly audible alarm signals if accidental criticality...

10 CFR 70.24 - Criticality accident requirements.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false Criticality accident requirements. 70.24 Section 70.24... Applications § 70.24 Criticality accident requirements. (a) Each licensee authorized to possess special nuclear...-sensitive radiation detectors which will energize clearly audible alarm signals if accidental criticality...

10 CFR 70.24 - Criticality accident requirements.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Criticality accident requirements. 70.24 Section 70.24... Applications § 70.24 Criticality accident requirements. (a) Each licensee authorized to possess special nuclear...-sensitive radiation detectors which will energize clearly audible alarm signals if accidental criticality...

10 CFR 70.24 - Criticality accident requirements.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Criticality accident requirements. 70.24 Section 70.24... Applications § 70.24 Criticality accident requirements. (a) Each licensee authorized to possess special nuclear...-sensitive radiation detectors which will energize clearly audible alarm signals if accidental criticality...

10 CFR 70.24 - Criticality accident requirements.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Criticality accident requirements. 70.24 Section 70.24... Applications § 70.24 Criticality accident requirements. (a) Each licensee authorized to possess special nuclear...-sensitive radiation detectors which will energize clearly audible alarm signals if accidental criticality...

49 CFR 225.3 - Applicability.

Code of Federal Regulations, 2010 CFR

2010-10-01

... TRANSPORTATION RAILROAD ACCIDENTS/INCIDENTS: REPORTS CLASSIFICATION, AND INVESTIGATIONS § 225.3 Applicability. (a... is in use; (iii) A bridge over a public road or waters used for commercial navigation; or (iv) A... injuries and illnesses and accountable rail equipment accidents/incidents found in § 225.25(a) through (g...

An Application of CICCT Accident Categories to Aviation Accidents in 1988-2004

NASA Technical Reports Server (NTRS)

Evans, Joni K.

2007-01-01

Interventions or technologies developed to improve aviation safety often focus on specific causes or accident categories. Evaluation of the potential effectiveness of those interventions is dependent upon mapping the historical aviation accidents into those same accident categories. To that end, the United States civil aviation accidents occurring between 1988 and 2004 (n=26,117) were assigned accident categories based upon the taxonomy developed by the CAST/ICAO Common Taxonomy Team (CICTT). Results are presented separately for four main categories of flight rules: Part 121 (large commercial air carriers), Scheduled Part 135 (commuter airlines), Non-Scheduled Part 135 (on-demand air taxi) and Part 91 (general aviation). Injuries and aircraft damage are summarized by year and by accident category.

Parametric Evaluation of SiC/SiC Composite Cladding with UO2 Fuel for LWR Applications: Fuel Rod Interactions and Impact of Nonuniform Power Profile in Fuel Rod

NASA Astrophysics Data System (ADS)

Singh, G.; Sweet, R.; Brown, N. R.; Wirth, B. D.; Katoh, Y.; Terrani, K.

2018-02-01

SiC/SiC composites are candidates for accident tolerant fuel cladding in light water reactors. In the extreme nuclear reactor environment, SiC-based fuel cladding will be exposed to neutron damage, significant heat flux, and a corrosive environment. To ensure reliable and safe operation of accident tolerant fuel cladding concepts such as SiC-based materials, it is important to assess thermo-mechanical performance under in-reactor conditions including irradiation and realistic temperature distributions. The effect of non-uniform dimensional changes caused by neutron irradiation with spatially varying temperatures, along with the closing of the fuel-cladding gap, on the stress development in the cladding over the course of irradiation were evaluated. The effect of non-uniform circumferential power profile in the fuel rod on the mechanical performance of the cladding is also evaluated. These analyses have been performed using the BISON fuel performance modeling code and the commercial finite element analysis code Abaqus. A constitutive model is constructed and solved numerically to predict the stress distribution in the cladding under normal operating conditions. The dependence of dimensions and thermophysical properties on irradiation dose and temperature has been incorporated into the models. Initial scoping results from parametric analyses provide time varying stress distributions in the cladding as well as the interaction of fuel rod with the cladding under different conditions of initial fuel rod-cladding gap and linear heat rate. It is found that a non-uniform circumferential power profile in the fuel rod may cause significant lateral bowing in the cladding, and motivates further analysis and evaluation.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Epiney, A.; Canepa, S.; Zerkak, O.

The STARS project at the Paul Scherrer Institut (PSI) has adopted the TRACE thermal-hydraulic (T-H) code for best-estimate system transient simulations of the Swiss Light Water Reactors (LWRs). For analyses involving interactions between system and core, a coupling of TRACE with the SIMULATE-3K (S3K) LWR core simulator has also been developed. In this configuration, the TRACE code and associated nuclear power reactor simulation models play a central role to achieve a comprehensive safety analysis capability. Thus, efforts have now been undertaken to consolidate the validation strategy by implementing a more rigorous and structured assessment approach for TRACE applications involving eithermore » only system T-H evaluations or requiring interfaces to e.g. detailed core or fuel behavior models. The first part of this paper presents the preliminary concepts of this validation strategy. The principle is to systematically track the evolution of a given set of predicted physical Quantities of Interest (QoIs) over a multidimensional parametric space where each of the dimensions represent the evolution of specific analysis aspects, including e.g. code version, transient specific simulation methodology and model "nodalisation". If properly set up, such environment should provide code developers and code users with persistent (less affected by user effect) and quantified information (sensitivity of QoIs) on the applicability of a simulation scheme (codes, input models, methodology) for steady state and transient analysis of full LWR systems. Through this, for each given transient/accident, critical paths of the validation process can be identified that could then translate into defining reference schemes to be applied for downstream predictive simulations. In order to illustrate this approach, the second part of this paper presents a first application of this validation strategy to an inadvertent blowdown event that occurred in a Swiss BWR/6. The transient was initiated by the spurious actuation of the Automatic Depressurization System (ADS). The validation approach progresses through a number of dimensions here: First, the same BWR system simulation model is assessed for different versions of the TRACE code, up to the most recent one. The second dimension is the "nodalisation" dimension, where changes to the input model are assessed. The third dimension is the "methodology" dimension. In this case imposed power and an updated TRACE core model are investigated. For each step in each validation dimension, a common set of QoIs are investigated. For the steady-state results, these include fuel temperatures distributions. For the transient part of the present study, the evaluated QoIs include the system pressure evolution and water carry-over into the steam line.« less

Computational modeling of driver speed control with its applications in developing intelligent transportation system to prevent speeding-related accidents.

DOT National Transportation Integrated Search

2013-08-01

Speeding is the leading contributing factor in fatal accidents in NY state, according to NY State Department of Motor : Vehicle Accidents Statistical Summary (2009). Understanding and modeling speeding and speed control is one of major : challenges i...

10 CFR 72.24 - Contents of application: Technical information.

Code of Federal Regulations, 2013 CFR

2013-01-01

... components provided for the prevention of accidents and the mitigation of the consequences of accidents... radiation exposures within the limits given in part 20 of this chapter, and for meeting the objective of... outside the controlled area from accidents or natural phenomena events that result in the release of...

10 CFR 72.24 - Contents of application: Technical information.

Code of Federal Regulations, 2014 CFR

2014-01-01

... components provided for the prevention of accidents and the mitigation of the consequences of accidents... radiation exposures within the limits given in part 20 of this chapter, and for meeting the objective of... outside the controlled area from accidents or natural phenomena events that result in the release of...

10 CFR 72.24 - Contents of application: Technical information.

Code of Federal Regulations, 2012 CFR

2012-01-01

... components provided for the prevention of accidents and the mitigation of the consequences of accidents... radiation exposures within the limits given in part 20 of this chapter, and for meeting the objective of... outside the controlled area from accidents or natural phenomena events that result in the release of...

Nuclear Power Plant Cyber Security Discrete Dynamic Event Tree Analysis (LDRD 17-0958) FY17 Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wheeler, Timothy A.; Denman, Matthew R.; Williams, R. A.

Instrumentation and control of nuclear power is transforming from analog to modern digital assets. These control systems perform key safety and security functions. This transformation is occurring in new plant designs as well as in the existing fleet of plants as the operation of those plants is extended to 60 years. This transformation introduces new and unknown issues involving both digital asset induced safety issues and security issues. Traditional nuclear power risk assessment tools and cyber security assessment methods have not been modified or developed to address the unique nature of cyber failure modes and of cyber security threat vulnerabilities.more » iii This Lab-Directed Research and Development project has developed a dynamic cyber-risk in- formed tool to facilitate the analysis of unique cyber failure modes and the time sequencing of cyber faults, both malicious and non-malicious, and impose those cyber exploits and cyber faults onto a nuclear power plant accident sequence simulator code to assess how cyber exploits and cyber faults could interact with a plants digital instrumentation and control (DI&C) system and defeat or circumvent a plants cyber security controls. This was achieved by coupling an existing Sandia National Laboratories nuclear accident dynamic simulator code with a cyber emulytics code to demonstrate real-time simulation of cyber exploits and their impact on automatic DI&C responses. Studying such potential time-sequenced cyber-attacks and their risks (i.e., the associated impact and the associated degree of difficulty to achieve the attack vector) on accident management establishes a technical risk informed framework for developing effective cyber security controls for nuclear power.« less

An uncertainty analysis of the hydrogen source term for a station blackout accident in Sequoyah using MELCOR 1.8.5

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gauntt, Randall O.; Bixler, Nathan E.; Wagner, Kenneth Charles

2014-03-01

A methodology for using the MELCOR code with the Latin Hypercube Sampling method was developed to estimate uncertainty in various predicted quantities such as hydrogen generation or release of fission products under severe accident conditions. In this case, the emphasis was on estimating the range of hydrogen sources in station blackout conditions in the Sequoyah Ice Condenser plant, taking into account uncertainties in the modeled physics known to affect hydrogen generation. The method uses user-specified likelihood distributions for uncertain model parameters, which may include uncertainties of a stochastic nature, to produce a collection of code calculations, or realizations, characterizing themore » range of possible outcomes. Forty MELCOR code realizations of Sequoyah were conducted that included 10 uncertain parameters, producing a range of in-vessel hydrogen quantities. The range of total hydrogen produced was approximately 583kg 131kg. Sensitivity analyses revealed expected trends with respected to the parameters of greatest importance, however, considerable scatter in results when plotted against any of the uncertain parameters was observed, with no parameter manifesting dominant effects on hydrogen generation. It is concluded that, with respect to the physics parameters investigated, in order to further reduce predicted hydrogen uncertainty, it would be necessary to reduce all physics parameter uncertainties similarly, bearing in mind that some parameters are inherently uncertain within a range. It is suspected that some residual uncertainty associated with modeling complex, coupled and synergistic phenomena, is an inherent aspect of complex systems and cannot be reduced to point value estimates. The probabilistic analyses such as the one demonstrated in this work are important to properly characterize response of complex systems such as severe accident progression in nuclear power plants.« less

Estimation Of 137Cs Using Atmospheric Dispersion Models After A Nuclear Reactor Accident

NASA Astrophysics Data System (ADS)

Simsek, V.; Kindap, T.; Unal, A.; Pozzoli, L.; Karaca, M.

2012-04-01

Nuclear energy will continue to have an important role in the production of electricity in the world as the need of energy grows up. But the safety of power plants will always be a question mark for people because of the accidents happened in the past. Chernobyl nuclear reactor accident which happened in 26 April 1986 was the biggest nuclear accident ever. Because of explosion and fire large quantities of radioactive material was released to the atmosphere. The release of the radioactive particles because of accident affected not only its region but the entire Northern hemisphere. But much of the radioactive material was spread over west USSR and Europe. There are many studies about distribution of radioactive particles and the deposition of radionuclides all over Europe. But this was not true for Turkey especially for the deposition of radionuclides released after Chernobyl nuclear reactor accident and the radiation doses received by people. The aim of this study is to determine the radiation doses received by people living in Turkish territory after Chernobyl nuclear reactor accident and use this method in case of an emergency. For this purpose The Weather Research and Forecasting (WRF) Model was used to simulate meteorological conditions after the accident. The results of WRF which were for the 12 days after accident were used as input data for the HYSPLIT model. NOAA-ARL's (National Oceanic and Atmospheric Administration Air Resources Laboratory) dispersion model HYSPLIT was used to simulate the 137Cs distrubition. The deposition values of 137Cs in our domain after Chernobyl Nuclear Reactor Accident were between 1.2E-37 Bq/m2 and 3.5E+08 Bq/m2. The results showed that Turkey was affected because of the accident especially the Black Sea Region. And the doses were calculated by using GENII-LIN which is multipurpose health physics code.

Qualification of CASMO5 / SIMULATE-3K against the SPERT-III E-core cold start-up experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grandi, G.; Moberg, L.

SIMULATE-3K is a three-dimensional kinetic code applicable to LWR Reactivity Initiated Accidents. S3K has been used to calculate several international recognized benchmarks. However, the feedback models in the benchmark exercises are different from the feedback models that SIMULATE-3K uses for LWR reactors. For this reason, it is worth comparing the SIMULATE-3K capabilities for Reactivity Initiated Accidents against kinetic experiments. The Special Power Excursion Reactor Test III was a pressurized-water, nuclear-research facility constructed to analyze the reactor kinetic behavior under initial conditions similar to those of commercial LWRs. The SPERT III E-core resembles a PWR in terms of fuel type, moderator,more » coolant flow rate, and system pressure. The initial test conditions (power, core flow, system pressure, core inlet temperature) are representative of cold start-up, hot start-up, hot standby, and hot full power. The qualification of S3K against the SPERT III E-core measurements is an ongoing work at Studsvik. In this paper, the results for the 30 cold start-up tests are presented. The results show good agreement with the experiments for the reactivity initiated accident main parameters: peak power, energy release and compensated reactivity. Predicted and measured peak powers differ at most by 13%. Measured and predicted reactivity compensations at the time of the peak power differ less than 0.01 $. Predicted and measured energy release differ at most by 13%. All differences are within the experimental uncertainty. (authors)« less

Application of CFX-10 to the Investigation of RPV Coolant Mixing in VVER Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moretti, Fabio; Melideo, Daniele; Terzuoli, Fulvio

2006-07-01

Coolant mixing phenomena occurring in the pressure vessel of a nuclear reactor constitute one of the main objectives of investigation by researchers concerned with nuclear reactor safety. For instance, mixing plays a relevant role in reactivity-induced accidents initiated by de-boration or boron dilution events, followed by transport of a de-borated slug into the vessel of a pressurized water reactor. Another example is constituted by temperature mixing, which may sensitively affect the consequences of a pressurized thermal shock scenario. Predictive analysis of mixing phenomena is strongly improved by the availability of computational tools able to cope with the inherent three-dimensionality ofmore » such problem, like system codes with three-dimensional capabilities, and Computational Fluid Dynamics (CFD) codes. The present paper deals with numerical analyses of coolant mixing in the reactor pressure vessel of a VVER-1000 reactor, performed by the ANSYS CFX-10 CFD code. In particular, the 'swirl' effect that has been observed to take place in the downcomer of such kind of reactor has been addressed, with the aim of assessing the capability of the codes to predict that effect, and to understand the reasons for its occurrence. Results have been compared against experimental data from V1000CT-2 Benchmark. Moreover, a boron mixing problem has been investigated, in the hypothesis that a de-borated slug, transported by natural circulation, enters the vessel. Sensitivity analyses have been conducted on some geometrical features, model parameters and boundary conditions. (authors)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salko, Robert K; Sung, Yixing; Kucukboyaci, Vefa

The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time stepmore » of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.« less

Delay Analysis of Car-to-Car Reliable Data Delivery Strategies Based on Data Mulling with Network Coding

NASA Astrophysics Data System (ADS)

Park, Joon-Sang; Lee, Uichin; Oh, Soon Young; Gerla, Mario; Lun, Desmond Siumen; Ro, Won Woo; Park, Joonseok

Vehicular ad hoc networks (VANET) aims to enhance vehicle navigation safety by providing an early warning system: any chance of accidents is informed through the wireless communication between vehicles. For the warning system to work, it is crucial that safety messages be reliably delivered to the target vehicles in a timely manner and thus reliable and timely data dissemination service is the key building block of VANET. Data mulling technique combined with three strategies, network codeing, erasure coding and repetition coding, is proposed for the reliable and timely data dissemination service. Particularly, vehicles in the opposite direction on a highway are exploited as data mules, mobile nodes physically delivering data to destinations, to overcome intermittent network connectivity cause by sparse vehicle traffic. Using analytic models, we show that in such a highway data mulling scenario the network coding based strategy outperforms erasure coding and repetition based strategies.

Tobit analysis of vehicle accident rates on interstate highways.

PubMed

Anastasopoulos, Panagiotis Ch; Tarko, Andrew P; Mannering, Fred L

2008-03-01

There has been an abundance of research that has used Poisson models and its variants (negative binomial and zero-inflated models) to improve our understanding of the factors that affect accident frequencies on roadway segments. This study explores the application of an alternate method, tobit regression, by viewing vehicle accident rates directly (instead of frequencies) as a continuous variable that is left-censored at zero. Using data from vehicle accidents on Indiana interstates, the estimation results show that many factors relating to pavement condition, roadway geometrics and traffic characteristics significantly affect vehicle accident rates.

Proceedings of the 21st DOE/NRC Nuclear Air Cleaning Conference; Sessions 1--8

DOE Office of Scientific and Technical Information (OSTI.GOV)

First, M.W.

1991-02-01

Separate abstracts have been prepared for the papers presented at the meeting on nuclear facility air cleaning technology in the following specific areas of interest: air cleaning technologies for the management and disposal of radioactive wastes; Canadian waste management program; radiological health effects models for nuclear power plant accident consequence analysis; filter testing; US standard codes on nuclear air and gas treatment; European community nuclear codes and standards; chemical processing off-gas cleaning; incineration and vitrification; adsorbents; nuclear codes and standards; mathematical modeling techniques; filter technology; safety; containment system venting; and nuclear air cleaning programs around the world. (MB)

Fundamentals, current state of the development of, and prospects for further improvement of the new-generation thermal-hydraulic computational HYDRA-IBRAE/LM code for simulation of fast reactor systems

NASA Astrophysics Data System (ADS)

Alipchenkov, V. M.; Anfimov, A. M.; Afremov, D. A.; Gorbunov, V. S.; Zeigarnik, Yu. A.; Kudryavtsev, A. V.; Osipov, S. L.; Mosunova, N. A.; Strizhov, V. F.; Usov, E. V.

2016-02-01

The conceptual fundamentals of the development of the new-generation system thermal-hydraulic computational HYDRA-IBRAE/LM code are presented. The code is intended to simulate the thermalhydraulic processes that take place in the loops and the heat-exchange equipment of liquid-metal cooled fast reactor systems under normal operation and anticipated operational occurrences and during accidents. The paper provides a brief overview of Russian and foreign system thermal-hydraulic codes for modeling liquid-metal coolants and gives grounds for the necessity of development of a new-generation HYDRA-IBRAE/LM code. Considering the specific engineering features of the nuclear power plants (NPPs) equipped with the BN-1200 and the BREST-OD-300 reactors, the processes and the phenomena are singled out that require a detailed analysis and development of the models to be correctly described by the system thermal-hydraulic code in question. Information on the functionality of the computational code is provided, viz., the thermalhydraulic two-phase model, the properties of the sodium and the lead coolants, the closing equations for simulation of the heat-mass exchange processes, the models to describe the processes that take place during the steam-generator tube rupture, etc. The article gives a brief overview of the usability of the computational code, including a description of the support documentation and the supply package, as well as possibilities of taking advantages of the modern computer technologies, such as parallel computations. The paper shows the current state of verification and validation of the computational code; it also presents information on the principles of constructing of and populating the verification matrices for the BREST-OD-300 and the BN-1200 reactor systems. The prospects are outlined for further development of the HYDRA-IBRAE/LM code, introduction of new models into it, and enhancement of its usability. It is shown that the program of development and practical application of the code will allow carrying out in the nearest future the computations to analyze the safety of potential NPP projects at a qualitatively higher level.

32 CFR 634.23 - Specified consent to impoundment.

Code of Federal Regulations, 2010 CFR

2010-07-01

... ENFORCEMENT AND CRIMINAL INVESTIGATIONS MOTOR VEHICLE TRAFFIC SUPERVISION Motor Vehicle Registration § 634.23... installation traffic code provide for the removal and temporary impoundment of privately owned motor vehicles..., creating a safety hazard, disabled by accident, left unattended in a restricted or control area, or...

TRAC-P1: an advanced best estimate computer program for PWR LOCA analysis. I. Methods, models, user information, and programming details

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1978-05-01

The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos Scientific Laboratory (LASL) to provide an advanced ''best estimate'' predictive capability for the analysis of postulated accidents in light water reactors (LWRs). TRAC-Pl provides this analysis capability for pressurized water reactors (PWRs) and for a wide variety of thermal-hydraulic experimental facilities. It features a three-dimensional treatment of the pressure vessel and associated internals; two-phase nonequilibrium hydrodynamics models; flow-regime-dependent constitutive equation treatment; 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.more » The TRAC-Pl User's Manual is composed of two separate volumes. Volume I gives a description of the thermal-hydraulic models and numerical solution methods used in the code. Detailed programming and user information is also provided. Volume II presents the results of the developmental verification calculations.« less

ITER's Tokamak Cooling Water System and the the Use of ASME Codes to Comply with French Regulations of Nuclear Pressure Equipment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Berry, Jan; Ferrada, Juan J; Curd, Warren

During inductive plasma operation of ITER, fusion power will reach 500 MW with an energy multiplication factor of 10. The heat will be transferred by the Tokamak Cooling Water System (TCWS) to the environment using the secondary cooling system. Plasma operations are inherently safe even under the most severe postulated accident condition a large, in-vessel break that results in a loss-of-coolant accident. A functioning cooling water system is not required to ensure safe shutdown. Even though ITER is inherently safe, TCWS equipment (e.g., heat exchangers, piping, pressurizers) are classified as safety important components. This is because the water is predictedmore » to contain low-levels of radionuclides (e.g., activated corrosion products, tritium) with activity levels high enough to require the design of components to be in accordance with French regulations for nuclear pressure equipment, i.e., the French Order dated 12 December 2005 (ESPN). ESPN has extended the practical application of the methodology established by the Pressure Equipment Directive (97/23/EC) to nuclear pressure equipment, under French Decree 99-1046 dated 13 December 1999, and Order dated 21 December 1999 (ESP). ASME codes and supplementary analyses (e.g., Failure Modes and Effects Analysis) will be used to demonstrate that the TCWS equipment meets these essential safety requirements. TCWS is being designed to provide not only cooling, with a capacity of approximately 1 GW energy removal, but also elevated temperature baking of first-wall/blanket, vacuum vessel, and divertor. Additional TCWS functions include chemical control of water, draining and drying for maintenance, and facilitation of leak detection/localization. The TCWS interfaces with the majority of ITER systems, including the secondary cooling system. U.S. ITER is responsible for design, engineering, and procurement of the TCWS with industry support from an Engineering Services Organization (ESO) (AREVA Federal Services, with support from Northrop Grumman, and OneCIS). ITER International Organization (ITER-IO) is responsible for design oversight and equipment installation in Cadarache, France. TCWS equipment will be fabricated using ASME design codes with quality assurance and oversight by an Agreed Notified Body (approved by the French regulator) that will ensure regulatory compliance. This paper describes the TCWS design and how U.S. ITER and fabricators will use ASME codes to comply with EU Directives and French Orders and Decrees.« less

Zero-state Markov switching count-data models: an empirical assessment.

PubMed

Malyshkina, Nataliya V; Mannering, Fred L

2010-01-01

In this study, a two-state Markov switching count-data model is proposed as an alternative to zero-inflated models to account for the preponderance of zeros sometimes observed in transportation count data, such as the number of accidents occurring on a roadway segment over some period of time. For this accident-frequency case, zero-inflated models assume the existence of two states: one of the states is a zero-accident count state, which has accident probabilities that are so low that they cannot be statistically distinguished from zero, and the other state is a normal-count state, in which counts can be non-negative integers that are generated by some counting process, for example, a Poisson or negative binomial. While zero-inflated models have come under some criticism with regard to accident-frequency applications - one fact is undeniable - in many applications they provide a statistically superior fit to the data. The Markov switching approach we propose seeks to overcome some of the criticism associated with the zero-accident state of the zero-inflated model by allowing individual roadway segments to switch between zero and normal-count states over time. An important advantage of this Markov switching approach is that it allows for the direct statistical estimation of the specific roadway-segment state (i.e., zero-accident or normal-count state) whereas traditional zero-inflated models do not. To demonstrate the applicability of this approach, a two-state Markov switching negative binomial model (estimated with Bayesian inference) and standard zero-inflated negative binomial models are estimated using five-year accident frequencies on Indiana interstate highway segments. It is shown that the Markov switching model is a viable alternative and results in a superior statistical fit relative to the zero-inflated models.

77 FR 50534 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-08-21

...) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously... statement of the alleged facts or expert opinion which support the contention and on which the requestor...

75 FR 30440 - Biweekly Notice: Applications and Amendments to Facility Operating Licenses Involving No...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-01

... consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of... alleged facts or expert opinion which support the contention and on which the requestor/ petitioner...

76 FR 67485 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-01

... [email protected] . Federal Rulemaking Web Site: Public comments and supporting materials related... increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a...

Application of the Life Change Unit model for the prevention of accident proneness among small to medium sized industries in Korea.

PubMed

Kang, Youngsig; Hahm, Hyojoon; Yang, Sunghwan; Kim, Taegu

2008-10-01

Behavior models have provided an accident proneness concept based on life change unit (LCU) factors. This paper describes the development of a Korean Life Change Unit (KLCU) model for workers and managers in fatal accident areas, as well as an evaluation of its application. Results suggest that death of parents is the highest stress-giving factor for employees of small and medium sized industries a rational finding the viewpoint of Korean culture. The next stress-giving factors were shown to be the death of a spouse or loved ones, followed by the death of close family members, the death of close friends, changes of family members' health, unemployment, and jail terms. It turned out that these factors have a serious effect on industrial accidents and work-related diseases. The death of parents and close friends are ranked higher in the KLCU model than that of Western society. Crucial information for industrial accident prevention in real fields will be provided and the provided information will be useful for safety management programs related to accident prevention.

Analysis on the Role of RSG-GAS Pool Cooling System during Partial Loss of Heat Sink Accident

NASA Astrophysics Data System (ADS)

Susyadi; Endiah, P. H.; Sukmanto, D.; Andi, S. E.; Syaiful, B.; Hendro, T.; Geni, R. S.

2018-02-01

RSG-GAS is a 30 MW reactor that is mostly used for radioisotope production and experimental activities. Recently, it is regularly operated at half of its capacity for efficiency reason. During an accident, especially loss of heat sink, the role of its pool cooling system is very important to dump decay heat. An analysis using single failure approach and partial modeling of RELAP5 performed by S. Dibyo, 2010 shows that there is no significant increase in the coolant temperature if this system is properly functioned. However lessons learned from the Fukushima accident revealed that an accident can happen due to multiple failures. Considering ageing of the reactor, in this research the role of pool cooling system is to be investigated for a partial loss of heat sink accident which is at the same time the protection system fails to scram the reactor when being operated at 15 MW. The purpose is to clarify the transient characteristics and the final state of the coolant temperature. The method used is by simulating the system in RELAP5 code. Calculation results shows the pool cooling systems reduce coolant temperature for about 1 K as compared without activating them. The result alsoreveals that when the reactor is being operated at half of its rated power, it is still in safe condition for a partial loss of heat sink accident without scram.

Probabilistic margin evaluation on accidental transients for the ASTRID reactor project

NASA Astrophysics Data System (ADS)

Marquès, Michel

2014-06-01

ASTRID is a technological demonstrator of Sodium cooled Fast Reactor (SFR) under development. The conceptual design studies are being conducted in accordance with the Generation IV reactor objectives, particularly in terms of improving safety. For the hypothetical events, belonging to the accidental category "severe accident prevention situations" having a very low frequency of occurrence, the safety demonstration is no more based on a deterministic demonstration with conservative assumptions on models and parameters but on a "Best-Estimate Plus Uncertainty" (BEPU) approach. This BEPU approach ispresented in this paper for an Unprotected Loss-of-Flow (ULOF) event. The Best-Estimate (BE) analysis of this ULOFt ransient is performed with the CATHARE2 code, which is the French reference system code for SFR applications. The objective of the BEPU analysis is twofold: first evaluate the safety margin to sodium boiling in taking into account the uncertainties on the input parameters of the CATHARE2 code (twenty-two uncertain input parameters have been identified, which can be classified into five groups: reactor power, accident management, pumps characteristics, reactivity coefficients, thermal parameters and head losses); secondly quantify the contribution of each input uncertainty to the overall uncertainty of the safety margins, in order to refocusing R&D efforts on the most influential factors. This paper focuses on the methodological aspects of the evaluation of the safety margin. At least for the preliminary phase of the project (conceptual design), a probabilistic criterion has been fixed in the context of this BEPU analysis; this criterion is the value of the margin to sodium boiling, which has a probability 95% to be exceeded, obtained with a confidence level of 95% (i.e. the M5,95percentile of the margin distribution). This paper presents two methods used to assess this percentile: the Wilks method and the Bootstrap method ; the effectiveness of the two methods is compared on the basis of 500 simulations performed with theCATHARE2 code. We conclude that, with only 100 simulations performed with the CATHARE2 code, which is a number of simulations workable in the conceptual design phase of the ASTRID project where the models and the hypothesis are often modified, it is best in order to evaluate the percentile M5,95 of the margin to sodium boiling to use the bootstrap method, which will provide a slightly conservative result. On the other hand, in order to obtain an accurate estimation of the percentileM5,95, for the safety report for example, it will be necessary to perform at least 300 simulations with the CATHARE2 code. In this case, both methods (Wilks and Bootstrap) would give equivalent results.

BESAFE II: Accident safety analysis code for MFE reactor designs

NASA Astrophysics Data System (ADS)

Sevigny, Lawrence Michael

The viability of controlled thermonuclear fusion as an alternative energy source hinges on its desirability from an economic and an environmental and safety standpoint. It is the latter which is the focus of this thesis. For magnetic fusion energy (MFE) devices, the safety concerns equate to a design's behavior during a worst-case accident scenario which is the loss of coolant accident (LOCA). In this dissertation, we examine the behavior of MFE devices during a LOCA and how this behavior relates to the safety characteristics of the machine; in particular the acute, whole-body, early dose. In doing so, we have produced an accident safety code, BESAFE II, now available to the fusion reactor design community. The Appendix constitutes the User's Manual for BESAFE II. The theory behind early dose calculations including the mobilization of activation products is presented in Chapter 2. Since mobilization of activation products is a strong function of temperature, it becomes necessary to calculate the thermal response of a design during a LOCA in order to determine the fraction of the activation products which are mobilized and thus become the source for the dose. The code BESAFE II is designed to determine the temperature history of each region of a design and determine the resulting mobilization of activation products at each point in time during the LOCA. The BESAFE II methodology is discussed in Chapter 4, followed by demonstrations of its use for two reference design cases: a PCA-Li tokamak and a SiC-He tokamak. Of these two cases, it is shown that the SiC-He tokamak is a better design from an accident safety standpoint than the PCA-Li tokamak. It is also found that doses derived from temperature-dependent mobilization data are different than those predicted using set mobilization categories such as those that involve Piet fractions. This demonstrates the need for more experimental data on fusion materials. The possibility for future improvements and modifications to BESAFE II is discussed in Chapter 6, for example, by adding additional environmental indices such as a waste disposal index. The biggest improvement to BESAFE II would be an increase in the database of activation product mobilization for a larger spectrum of fusion reactor materials. The ultimate goal we have is for BESAFE II to become part of a systems design program which would include economic factors and allow both safety and the cost of electricity to influence design.

LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carmack, William Jonathan; Barrett, Kristine Eloise; Chichester, Heather Jean MacLean

2015-09-01

The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirementsmore » for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.« less

Data Analysis Approaches for the Risk-Informed Safety Margins Characterization Toolkit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mandelli, Diego; Alfonsi, Andrea; Maljovec, Daniel P.

2016-09-01

In the past decades, several numerical simulation codes have been employed to simulate accident dynamics (e.g., RELAP5-3D, RELAP-7, MELCOR, MAAP). In order to evaluate the impact of uncertainties into accident dynamics, several stochastic methodologies have been coupled with these codes. These stochastic methods range from classical Monte-Carlo and Latin Hypercube sampling to stochastic polynomial methods. Similar approaches have been introduced into the risk and safety community where stochastic methods (such as RAVEN, ADAPT, MCDET, ADS) have been coupled with safety analysis codes in order to evaluate the safety impact of timing and sequencing of events. These approaches are usually calledmore » Dynamic PRA or simulation-based PRA methods. These uncertainties and safety methods usually generate a large number of simulation runs (database storage may be on the order of gigabytes or higher). The scope of this paper is to present a broad overview of methods and algorithms that can be used to analyze and extract information from large data sets containing time dependent data. In this context, “extracting information” means constructing input-output correlations, finding commonalities, and identifying outliers. Some of the algorithms presented here have been developed or are under development within the RAVEN statistical framework.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gauntt, Randall O.; Mattie, Patrick D.

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

STARDUST-U experiments on fluid-dynamic conditions affecting dust mobilization during LOVAs

NASA Astrophysics Data System (ADS)

Poggi, L. A.; Malizia, A.; Ciparisse, J. F.; Tieri, F.; Gelfusa, M.; Murari, A.; Del Papa, C.; Giovannangeli, I.; Gaudio, P.

2016-07-01

Since 2006 the Quantum Electronics and Plasma Physics (QEP) Research Group together with ENEA FusTech of Frascati have been working on dust re-suspension inside tokamaks and its potential capability to jeopardize the integrity of future fusion nuclear plants (i.e. ITER or DEMO) and to be a risk for the health of the operators. Actually, this team is working with the improved version of the "STARDUST" facility, i.e. "STARDUST-Upgrade". STARDUST-U facility has four new air inlet ports that allow the experimental replication of Loss of Vacuum Accidents (LOVAs). The experimental campaign to detect the different pressurization rates, local air velocity, temperature, have been carried out from all the ports in different accident conditions and the principal results will be analyzed and compared with the numerical simulations obtained through a CFD (Computational Fluid Dynamic) code. This preliminary thermo fluid-dynamic analysis of the accident is crucial for numerical model development and validation, and for the incoming experimental campaign of dust resuspension inside STARDUST-U due to well-defined accidents presented in this paper.

Graphical fault tree analysis for fatal falls in the construction industry.

PubMed

Chi, Chia-Fen; Lin, Syuan-Zih; Dewi, Ratna Sari

2014-11-01

The current study applied a fault tree analysis to represent the causal relationships among events and causes that contributed to fatal falls in the construction industry. Four hundred and eleven work-related fatalities in the Taiwanese construction industry were analyzed in terms of age, gender, experience, falling site, falling height, company size, and the causes for each fatality. Given that most fatal accidents involve multiple events, the current study coded up to a maximum of three causes for each fall fatality. After the Boolean algebra and minimal cut set analyses, accident causes associated with each falling site can be presented as a fault tree to provide an overview of the basic causes, which could trigger fall fatalities in the construction industry. Graphical icons were designed for each falling site along with the associated accident causes to illustrate the fault tree in a graphical manner. A graphical fault tree can improve inter-disciplinary discussion of risk management and the communication of accident causation to first line supervisors. Copyright © 2014 Elsevier Ltd. All rights reserved.

78 FR 60321 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-01

... sample selection. A steam generator tube rupture (SGTR) event is one of the design basis accidents that... in the design basis accident analysis. The proposed change will not cause the consequences of a SGTR... changes to the plant design basis or postulated accidents resulting from potential tube degradation. The...

Application of Near Infrared Spectroscopy, Intravascular Ultrasound and the Coronary Calcium Score to Predict Adverse Coronary Events

DTIC Science & Technology

2012-10-01

hospitalization 9. Emergence of rhythm disturbances requiring treatment 10. Development of acute coronary syndrome 11. Cerebrovascular accident Adverse...catheterization. These will include coronary injury including dissection, perforation or occlusion, death, cerebrovascular accident , myocardial... cerebrovascular accident , bleeding, infection, arrhythmia, access site damage, coronary dissection, coronary thrombosis and myocardial infarction, among

DOE Office of Scientific and Technical Information (OSTI.GOV)

PIEPHO, M.G.

Four bounding accidents postulated for the K West Basin integrated water treatment system are evaluated against applicable risk evaluation guidelines. The accidents are a spray leak during fuel retrieval, spray leak during backflushing a hydrogen explosion, and a fire breaching filter vessel and enclosure. Event trees and accident probabilities are estimated. In all cases, the unmitigated dose consequences are below the risk evaluation guidelines.

78 FR 77729 - Biweekly Notice: Applications and Amendments to Facility Operating Licenses and Combined Licenses...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-24

... amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident... alleged facts or expert opinion which support the contention and on which the requestor/ petitioner...

78 FR 15747 - Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-12

... consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of... contention and a concise statement of the alleged facts or expert opinion which support the contention and on...

75 FR 26287 - Notice; Applications and Amendments to Facility Operating Licenses Involving Proposed No...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-11

... amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident... contention and a concise statement of the alleged facts or expert opinion which support the contention and on...

78 FR 75379 - Entergy Operations, Inc.; Combined License Application for Grand Gulf Unit 3; Exemption From the...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-11

... accidents are increased. Therefore, there is no undue risk to public health and safety. Plant construction... at this time. Based on the nature of the requested exemption as described above, no new accident... a license decision is made, the probability of postulated accidents is not increased. Additionally...

Health Physics Code System for Evaluating Accidents Involving Radioactive Materials.

DOE Office of Scientific and Technical Information (OSTI.GOV)

2014-10-01

Version 03 The HOTSPOT Health Physics codes were created to provide Health Physics personnel with a fast, field-portable calculational tool for evaluating accidents involving radioactive materials. HOTSPOT codes provide a first-order approximation of the radiation effects associated with the atmospheric release of radioactive materials. The developer's website is: http://www.llnl.gov/nhi/hotspot/. Four general programs, PLUME, EXPLOSION, FIRE, and RESUSPENSION, calculate a downwind assessment following the release of radioactive material resulting from a continuous or puff release, explosive release, fuel fire, or an area contamination event. Additional programs deal specifically with the release of plutonium, uranium, and tritium to expedite an initial assessmentmore » of accidents involving nuclear weapons. The FIDLER program can calibrate radiation survey instruments for ground survey measurements and initial screening of personnel for possible plutonium uptake in the lung. The HOTSPOT codes are fast, portable, easy to use, and fully documented in electronic help files. HOTSPOT supports color high resolution monitors and printers for concentration plots and contours. The codes have been extensively used by the DOS community since 1985. Tables and graphical output can be directed to the computer screen, printer, or a disk file. The graphical output consists of dose and ground contamination as a function of plume centerline downwind distance, and radiation dose and ground contamination contours. Users have the option of displaying scenario text on the plots. HOTSPOT 3.0.1 fixes three significant Windows 7 issues: Executable installed properly under "Program Files/HotSpot 3.0". Installation package now smaller: removed dependency on older Windows DLL files which previously needed to; Forms now properly scale based on DPI instead of font for users who change their screen resolution to something other than 100%. This is a more common feature in Windows 7; Windows installer was starting everytime most users started the program, even after HotSpot was already installed. Now, after the program is installed the installer may come up once for each new user but only the first time they run HotSpot on a particular machine. So no user should see the installer come up more than once over many uses; and GPS capability updated to directly use a serial port through a USB connection. Non-USB connections should still work. Fixed table output inconsistencies for fire scenarios.« less

Structural testing of the Los Alamos National Laboratory Heat Source/Radioisotopic Thermoelectric Generator shipping container

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bronowski, D.R.; Madsen, M.M.

The Heat Source/Radioisotopic Thermoelectric Generator shipping container is a Type B packaging design currently under development by Los Alamos National Laboratory. Type B packaging for transporting radioactive material is required to maintain containment and shielding after being exposed to the normal and hypothetical accident environments defined in Title 10 Code of Federal Regulations Part 71. A combination of testing and analysis is used to verify the adequacy of this package design. This report documents the test program portion of the design verification, using several prototype packages. Four types of testing were performed: 30-foot hypothetical accident condition drop tests in threemore » orientations, 40-inch hypothetical accident condition puncture tests in five orientations, a 21 psi external overpressure test, and a normal conditions of transport test consisting of a water spray and a 4 foot drop test. 18 refs., 104 figs., 13 tabs.« less

Research subjects for analytical estimation of core degradation at Fukushima-Daiichi nuclear power plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nagase, F.; Ishikawa, J.; Kurata, M.

2013-07-01

Estimation of the accident progress and status inside the pressure vessels (RPV) and primary containment vessels (PCV) is required for appropriate conductance of decommissioning in the Fukushima-Daiichi NPP. For that, it is necessary to obtain additional experimental data and revised models for the estimation using computer codes with increased accuracies. The Japan Atomic Energy Agency (JAEA) has selected phenomena to be reviewed and developed, considering previously obtained information, conditions specific to the Fukushima-Daiichi NPP accident, and recent progress of experimental and analytical technologies. As a result, research and development items have been picked up in terms of thermal-hydraulic behavior inmore » the RPV and PCV, progression of fuel bundle degradation, failure of the lower head of RPV, and analysis of the accident. This paper introduces the selected phenomena to be reviewed and developed, research plans and recent results from the JAEA's corresponding research programs. (authors)« less

The role of usability in the evaluation of accidents: human error or design flaw?

PubMed

Correia, Walter; Soares, Marcelo; Barros, Marina; Campos, Fábio

2012-01-01

This article aims to highlight the role of consumer products companies in the heart and the extent of accidents involving these types of products, and as such undesired events take part as an agent in influencing decision making for the purchase of a product that nature on the part of consumers and users. The article demonstrates, by reference, interviews and case studies such as the development of poorly designed products and design errors of design can influence the usage behavior of users, thus leading to accidents, and also negatively affect the next image of a company. The full explanation of these types of questions aims to raise awareness, plan on a reliable usability, users and consumers in general about the safe use of consumer products, and also safeguard their rights before a legal system of consumer protection, even far away by the CDC--Code of Consumer Protection.

[Thrombosis and post-thrombotic syndrome as a consequence of an accident].

PubMed

Wahl, U; Hirsch, T

2015-10-01

Phlebothromboses represent alarming complications in accident victims since they can cause fatal pulmonary embolisms. More than half of those affected also develop post-thrombotic syndrome in the course of the illness. In addition to making clinical assessments, the traumatologist should also have fundamental knowledge about diagnostic methods and be familiar with interpreting internal findings. Colour-coded duplex sonography plays a central role in diagnosing thrombosis and in assessing functional limitations. Further information can be gathered from various phlebological procedures. The expert evaluation of the immediate, as well as the long-term consequences of an accident frequently require leg swelling to be classified. It is not uncommon for post-thrombotic syndrome to be diagnosed for the first time during this process. An additional vascular appraisal is often required. An appreciation of social-medical and insurance-related aspects means a high degree of responsibility is placed on the expert.

SBLOCA outside containment at Browns Ferry Unit One: accident sequence analysis. [Small break

DOE Office of Scientific and Technical Information (OSTI.GOV)

Condon, W.A.; Harrington, R.M.; Greene, S.R.

1982-11-01

This study describes the predicted response of Unit 1 at the Browns Ferry Nuclear Plant to a postulated small-break loss-of-coolant accident outside of the primary containment. The break has been assumed to occur in the scram discharge volume piping immediately following a reactor scram that cannot be reset. The events before core uncovering are discussed for both the worst-case accident sequence without operator action and for the more likely sequences with operator action. Without operator action, the events after core uncovering would include core meltdown and subsequent containment failure, and this event sequence has been determined through use of themore » MARCH code. An estimate of the magnitude and timing of the concomitant release of the noble gas, cesium, and iodine-based fission products to the environment is provided in Volume 2 of this report.« less

Markov switching multinomial logit model: An application to accident-injury severities.

PubMed

Malyshkina, Nataliya V; Mannering, Fred L

2009-07-01

In this study, two-state Markov switching multinomial logit models are proposed for statistical modeling of accident-injury severities. These models assume Markov switching over time between two unobserved states of roadway safety as a means of accounting for potential unobserved heterogeneity. The states are distinct in the sense that in different states accident-severity outcomes are generated by separate multinomial logit processes. To demonstrate the applicability of the approach, two-state Markov switching multinomial logit models are estimated for severity outcomes of accidents occurring on Indiana roads over a four-year time period. Bayesian inference methods and Markov Chain Monte Carlo (MCMC) simulations are used for model estimation. The estimated Markov switching models result in a superior statistical fit relative to the standard (single-state) multinomial logit models for a number of roadway classes and accident types. It is found that the more frequent state of roadway safety is correlated with better weather conditions and that the less frequent state is correlated with adverse weather conditions.

Integrating geographical information and augmented reality techniques for mobile escape guidelines on nuclear accident sites.

PubMed

Tsai, Ming-Kuan; Lee, Yung-Ching; Lu, Chung-Hsin; Chen, Mei-Hsin; Chou, Tien-Yin; Yau, Nie-Jia

2012-07-01

During nuclear accidents, when radioactive materials spread into the environment, the people in the affected areas should evacuate immediately. However, few information systems are available regarding escape guidelines for nuclear accidents. Therefore, this study constructs escape guidelines on mobile phones. This application is called Mobile Escape Guidelines (MEG) and adopts two techniques. One technique is the geographical information that offers multiple representations; the other is the augmented reality that provides semi-realistic information services. When this study tested the mobile escape guidelines, the results showed that this application was capable of identifying the correct locations of users, showing the escape routes, filtering geographical layers, and rapidly generating the relief reports. Users could evacuate from nuclear accident sites easily, even without relief personnel, since using slim devices to access the mobile escape guidelines is convenient. Overall, this study is a useful reference for a nuclear accident emergency response. Copyright © 2012 Elsevier Ltd. All rights reserved.

To amend the limitation on liability for certain passenger rail accidents or incidents under section 28103 of title 49, United States Code, and for other purposes.

THOMAS, 111th Congress

Rep. Gallegly, Elton [R-CA-24

2010-09-16

House - 09/17/2010 Referred to the Subcommittee on Railroads, Pipelines, and Hazardous Materials. (All Actions) Tracker: This bill has the status IntroducedHere are the steps for Status of Legislation:

Emergency Planning and Community Right-to-Know Act Section 312 Tier Two report forms

DOE Office of Scientific and Technical Information (OSTI.GOV)

Evans, R.A.

2000-02-01

The report contains forms for the chemical description, physical and health hazards, inventory volumes, and storage codes and locations for all hazardous chemicals located at the Y-12 Plant. These can be used by local emergency response teams in case of an accident.

Bleacher Safety: What Do We Look for? What Can We Do?

ERIC Educational Resources Information Center

IEA Environmental Consultant, 1999

1999-01-01

Discusses safety issues surrounding aging bleacher systems, highlighting the following three primary safety considerations: space between seats and footboards; guardrails; and the structural provisions of the 1997 Uniform Building Code. Tips for bleacher accident-prevention assessment and excerpts from federal and Minnesota legislation on bleacher…

Emergency Planning and Community Right-To-Know Act Section 312 Tier Two report forms

DOE Office of Scientific and Technical Information (OSTI.GOV)

Evans, R.A.

2000-02-01

The report contains forms for the chemical description, physical and health hazards, inventory volumes, and storage codes and locations for all hazardous chemicals located at the Y-12 Plant. These can be used by local emergency response teams in case of an accident.

Update of the Nuclear Criticality Slide Rule for the Emergency Response to a Nuclear Criticality Accident

NASA Astrophysics Data System (ADS)

Duluc, Matthieu; Bardelay, Aurélie; Celik, Cihangir; Heinrichs, Dave; Hopper, Calvin; Jones, Richard; Kim, Soon; Miller, Thomas; Troisne, Marc; Wilson, Chris

2017-09-01

AWE (UK), IRSN (France), LLNL (USA) and ORNL (USA) began a long term collaboration effort in 2015 to update the nuclear criticality Slide Rule for the emergency response to a nuclear criticality accident. This document, published almost 20 years ago, gives order of magnitude estimates of key parameters, such as number of fissions and doses (neutron and gamma), useful for emergency response teams and public authorities. This paper will present, firstly the motivation and the long term objectives for this update, then the overview of the initial configurations for updated calculations and preliminary results obtained with modern 3D codes.

Problems with numerical techniques: Application to mid-loop operation transients

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bryce, W.M.; Lillington, J.N.

1997-07-01

There has been an increasing need to consider accidents at shutdown which have been shown in some PSAs to provide a significant contribution to overall risk. In the UK experience has been gained at three levels: (1) Assessment of codes against experiments; (2) Plant studies specifically for Sizewell B; and (3) Detailed review of modelling to support the plant studies for Sizewell B. The work has largely been carried out using various versions of RELAP5 and SCDAP/RELAP5. The paper details some of the problems that have needed to be addressed. It is believed by the authors that these kinds ofmore » problems are probably generic to most of the present generation system thermal-hydraulic codes for the conditions present in mid-loop transients. Thus as far as possible these problems and solutions are proposed in generic terms. The areas addressed include: condensables at low pressure, poor time step calculation detection, water packing, inadequate physical modelling, numerical heat transfer and mass errors. In general single code modifications have been proposed to solve the problems. These have been very much concerned with means of improving existing models rather than by formulating a completely new approach. They have been produced after a particular problem has arisen. Thus, and this has been borne out in practice, the danger is that when new transients are attempted, new problems arise which then also require patching.« less

A first principles study of the electronic structure, elastic and thermal properties of UB2

NASA Astrophysics Data System (ADS)

Jossou, Ericmoore; Malakkal, Linu; Szpunar, Barbara; Oladimeji, Dotun; Szpunar, Jerzy A.

2017-07-01

Uranium diboride (UB2) has been widely deployed for refractory use and is a proposed material for Accident Tolerant Fuel (ATF) due to its high thermal conductivity. However, the applicability of UB2 towards high temperature usage in a nuclear reactor requires the need to investigate the thermomechanical properties, and recent studies have failed in highlighting applicable properties. In this work, we present an in-depth theoretical outlook of the structural and thermophysical properties of UB2, including but not limited to elastic, electronic and thermal transport properties. These calculations were performed within the framework of Density Functional Theory (DFT) + U approach, using Quantum ESPRESSO (QE) code considering the addition of Coulomb correlations on the uranium atom. The phonon spectra and elastic constant analysis show the dynamic and mechanical stability of UB2 structure respectively. The electronic structure of UB2 was investigated using full potential linear augmented plane waves plus local orbitals method (FP-LAPW+lo) as implemented in WIEN2k code. The absence of a band gap in the total and partial density of states confirms the metallic nature while the valence electron density plot reveals the presence of covalent bond between adjacent B-B atoms. We predicted the lattice thermal conductivity (kL) by solving Boltzmann Transport Equation (BTE) using ShengBTE. The second order harmonic and third-order anharmonic interatomic force constants required as input to ShengBTE was calculated using the Density-functional perturbation theory (DFPT). However, we predicted the electronic thermal conductivity (kel) using Wiedemann-Franz law as implemented in Boltztrap code. We also show that the sound velocity along 'a' and 'c' axes exhibit high anisotropy, which accounts for the anisotropic thermal conductivity of UB2.

GEN-IV Benchmarking of Triso Fuel Performance Models under accident conditions modeling input data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collin, Blaise Paul

This document presents the benchmark plan for the calculation of particle fuel performance on safety testing experiments that are representative of operational accidental transients. The benchmark is dedicated to the modeling of fission product release under accident conditions by fuel performance codes from around the world, and the subsequent comparison to post-irradiation experiment (PIE) data from the modeled heating tests. The accident condition benchmark is divided into three parts: • The modeling of a simplified benchmark problem to assess potential numerical calculation issues at low fission product release. • The modeling of the AGR-1 and HFR-EU1bis safety testing experiments. •more » The comparison of the AGR-1 and HFR-EU1bis modeling results with PIE data. The simplified benchmark case, thereafter named NCC (Numerical Calculation Case), is derived from “Case 5” of the International Atomic Energy Agency (IAEA) Coordinated Research Program (CRP) on coated particle fuel technology [IAEA 2012]. It is included so participants can evaluate their codes at low fission product release. “Case 5” of the IAEA CRP-6 showed large code-to-code discrepancies in the release of fission products, which were attributed to “effects of the numerical calculation method rather than the physical model” [IAEA 2012]. The NCC is therefore intended to check if these numerical effects subsist. The first two steps imply the involvement of the benchmark participants with a modeling effort following the guidelines and recommendations provided by this document. The third step involves the collection of the modeling results by Idaho National Laboratory (INL) and the comparison of these results with the available PIE data. The objective of this document is to provide all necessary input data to model the benchmark cases, and to give some methodology guidelines and recommendations in order to make all results suitable for comparison with each other. The participants should read this document thoroughly to make sure all the data needed for their calculations is provided in the document. Missing data will be added to a revision of the document if necessary. 09/2016: Tables 6 and 8 updated. AGR-2 input data added« less

Generation IV benchmarking of TRISO fuel performance models under accident conditions: Modeling input data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collin, Blaise P.

2014-09-01

This document presents the benchmark plan for the calculation of particle fuel performance on safety testing experiments that are representative of operational accidental transients. The benchmark is dedicated to the modeling of fission product release under accident conditions by fuel performance codes from around the world, and the subsequent comparison to post-irradiation experiment (PIE) data from the modeled heating tests. The accident condition benchmark is divided into three parts: the modeling of a simplified benchmark problem to assess potential numerical calculation issues at low fission product release; the modeling of the AGR-1 and HFR-EU1bis safety testing experiments; and, the comparisonmore » of the AGR-1 and HFR-EU1bis modeling results with PIE data. The simplified benchmark case, thereafter named NCC (Numerical Calculation Case), is derived from ''Case 5'' of the International Atomic Energy Agency (IAEA) Coordinated Research Program (CRP) on coated particle fuel technology [IAEA 2012]. It is included so participants can evaluate their codes at low fission product release. ''Case 5'' of the IAEA CRP-6 showed large code-to-code discrepancies in the release of fission products, which were attributed to ''effects of the numerical calculation method rather than the physical model''[IAEA 2012]. The NCC is therefore intended to check if these numerical effects subsist. The first two steps imply the involvement of the benchmark participants with a modeling effort following the guidelines and recommendations provided by this document. The third step involves the collection of the modeling results by Idaho National Laboratory (INL) and the comparison of these results with the available PIE data. The objective of this document is to provide all necessary input data to model the benchmark cases, and to give some methodology guidelines and recommendations in order to make all results suitable for comparison with each other. The participants should read this document thoroughly to make sure all the data needed for their calculations is provided in the document. Missing data will be added to a revision of the document if necessary.« less

CRAC2 model description

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ritchie, L.T.; Alpert, D.J.; Burke, R.P.

1984-03-01

The CRAC2 computer code is a revised version of CRAC (Calculation of Reactor Accident Consequences) which was developed for the Reactor Safety Study. This document provides an overview of the CRAC2 code and a description of each of the models used. Significant improvements incorporated into CRAC2 include an improved weather sequence sampling technique, a new evacuation model, and new output capabilities. In addition, refinements have been made to the atmospheric transport and deposition model. Details of the modeling differences between CRAC2 and CRAC are emphasized in the model descriptions.

Comparison of MELCOR and SCDAP/RELAP5 results for a low-pressure, short-term station blackout at Browns Ferry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carbajo, J.J.

1995-12-31

This study compares results obtained with two U.S. Nuclear Regulatory Commission (NRC)-sponsored codes, MELCOR version 1.8.3 (1.8PQ) and SCDAP/RELAP5 Mod3.1 release C, for the same transient - a low-pressure, short-term station blackout accident at the Browns Ferry nuclear plant. This work is part of MELCOR assessment activities to compare core damage progression calculations of MELCOR against SCDAP/RELAP5 since the two codes model core damage progression very differently.

Developpement d'une methode de Monte Carlo dependante du temps et application au reacteur de type CANDU-6

NASA Astrophysics Data System (ADS)

Mahjoub, Mehdi

La resolution de l'equation de Boltzmann demeure une etape importante dans la prediction du comportement d'un reacteur nucleaire. Malheureusement, la resolution de cette equation presente toujours un defi pour une geometrie complexe (reacteur) tout comme pour une geometrie simple (cellule). Ainsi, pour predire le comportement d'un reacteur nucleaire,un schema de calcul a deux etapes est necessaire. La premiere etape consiste a obtenir les parametres nucleaires d'une cellule du reacteur apres une etape d'homogeneisation et condensation. La deuxieme etape consiste en un calcul de diffusion pour tout le reacteur en utilisant les resultats de la premiere etape tout en simplifiant la geometrie du reacteur a un ensemble de cellules homogenes le tout entoure de reflecteur. Lors des transitoires (accident), ces deux etapes sont insuffisantes pour pouvoir predire le comportement du reacteur. Comme la resolution de l'equation de Boltzmann dans sa forme dependante du temps presente toujours un defi de taille pour tous types de geometries,un autre schema de calcul est necessaire. Afin de contourner cette difficulte, l'hypothese adiabatique est utilisee. Elle se concretise en un schema de calcul a quatre etapes. La premiere et deuxieme etapes demeurent les memes pour des conditions nominales du reacteur. La troisieme etape se resume a obtenir les nouvelles proprietes nucleaires de la cellule a la suite de la perturbation pour les utiliser, au niveau de la quatrieme etape, dans un nouveau calcul de reacteur et obtenir l'effet de la perturbation sur le reacteur. Ce projet vise a verifier cette hypothese. Ainsi, un nouveau schema de calcul a ete defini. La premiere etape de ce projet a ete de creer un nouveau logiciel capable de resoudre l'equation de Boltzmann dependante du temps par la methode stochastique Monte Carlo dans le but d'obtenir des sections efficaces qui evoluent dans le temps. Ce code a ete utilise pour simuler un accident LOCA dans un reacteur nucleaire de type CANDU-6. Les sections efficaces dependantes du temps ont ete par la suite utilisees dans un calcul de diffusion espace-temps pour un reacteur CANDU-6 subissant un accident de type LOCA affectant la moitie du coeur afin d'observer son comportement durant toutes les phases de la perturbation. Dans la phase de developpement, nous avons choisi de demarrer avec le code OpenMC, developpe au MIT,comme plateforme initiale de developpement. L'introduction et le traitement des neutrons retardes durant la simulation ont presente un grand defi a surmonter. Il est important de noter que le code developpe utilisant la methode Monte Carlo peut etre utilise a grande echelle pour la simulation de tous les types des reacteurs nucleaires si les supports informatiques sont disponibles.

Strategies for dealing with resistance to recommendations from accident investigations.

PubMed

Lundberg, Jonas; Rollenhagen, Carl; Hollnagel, Erik; Rankin, Amy

2012-03-01

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

Validity and consistency assessment of accident analysis methods in the petroleum industry.

PubMed

Ahmadi, Omran; Mortazavi, Seyed Bagher; Khavanin, Ali; Mokarami, Hamidreza

2017-11-17

Accident analysis is the main aspect of accident investigation. It includes the method of connecting different causes in a procedural way. Therefore, it is important to use valid and reliable methods for the investigation of different causal factors of accidents, especially the noteworthy ones. This study aimed to prominently assess the accuracy (sensitivity index [SI]) and consistency of the six most commonly used accident analysis methods in the petroleum industry. In order to evaluate the methods of accident analysis, two real case studies (process safety and personal accident) from the petroleum industry were analyzed by 10 assessors. The accuracy and consistency of these methods were then evaluated. The assessors were trained in the workshop of accident analysis methods. The systematic cause analysis technique and bowtie methods gained the greatest SI scores for both personal and process safety accidents, respectively. The best average results of the consistency in a single method (based on 10 independent assessors) were in the region of 70%. This study confirmed that the application of methods with pre-defined causes and a logic tree could enhance the sensitivity and consistency of accident analysis.

Attendance at a hospital emergency department by drivers involved in automobile accidents in Italy.

PubMed

Pileggi, C; Nicotera, G; Angelillo, I F

2005-04-01

This study investigated the profile of drivers involved in automobile accidents attending a hospital emergency department (ED) in Catanzaro (Italy). Car drivers involved in automobile accidents who were registered for emergency care between May 2003 and February 2004 were included in the study. Demographics and details of the accident were collected immediately after admittance, before examination by the medical staff. For each patient, the medical staff completed a form including diagnostic investigations and medical/surgical examination in the ED. Of a total of 424 drivers included in the study 27.4% had conditions that were definitely non-urgent problems. Multiple logistic regression analysis indicated that the use of the ED as a source of non-urgent care was significantly higher among patients who were driving at a lower speed when the accident occurred, among those who presented to the ED before the implementation of the new Italian traffic code, and among those who underwent fewer diagnostic investigations and medical/surgical examinations in the ED. Most of the automobile related lesions occurred in the neck (43.9%) followed by multiple body regions (12.5%) and the upper extremities (10.4%). According to the nature of the injury a third were contusions (34%), followed by pain without physical signs and symptoms (28.8%), and dislocation, sprains, and strains (22.9%). Development of health promotion and education campaigns is required to prevent the use of the ED as a source of non-urgent care by those involved in automobile accidents.

Heat up and potential failure of BWR upper internals during a severe accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robb, Kevin R

2015-01-01

In boiling water reactors, the steam dome, steam separators, and dryers above the core are comprised of approximately 100 tons of stainless steel. During a severe accident in which the coolant boils away and exothermic oxidation of zirconium occurs, gases (steam and hydrogen) are superheated in the core region and pass through the upper internals. Historically, the upper internals have been modeled using severe accident codes with relatively simple approximations. The upper internals are typically modeled in MELCOR as two lumped volumes with simplified heat transfer characteristics, with no structural integrity considerations, and with limited ability to oxidize, melt, andmore » relocate. The potential for and the subsequent impact of the upper internals to heat up, oxidize, fail, and relocate during a severe accident was investigated. A higher fidelity representation of the shroud dome, steam separators, and steam driers was developed in MELCOR v1.8.6 by extending the core region upwards. This modeling effort entailed adding 45 additional core cells and control volumes, 98 flow paths, and numerous control functions. The model accounts for the mechanical loading and structural integrity, oxidation, melting, flow area blockage, and relocation of the various components. The results indicate that the upper internals can reach high temperatures during a severe accident; they are predicted to reach a high enough temperature such that they lose their structural integrity and relocate. The additional 100 tons of stainless steel debris influences the subsequent in-vessel and ex-vessel accident progression.« less

The evolution of the genetic code: Impasses and challenges.

PubMed

Kun, Ádám; Radványi, Ádám

2018-02-01

The origin of the genetic code and translation is a "notoriously difficult problem". In this survey we present a list of questions that a full theory of the genetic code needs to answer. We assess the leading hypotheses according to these criteria. The stereochemical, the coding coenzyme handle, the coevolution, the four-column theory, the error minimization and the frozen accident hypotheses are discussed. The integration of these hypotheses can account for the origin of the genetic code. But experiments are badly needed. Thus we suggest a host of experiments that could (in)validate some of the models. We focus especially on the coding coenzyme handle hypothesis (CCH). The CCH suggests that amino acids attached to RNA handles enhanced catalytic activities of ribozymes. Alternatively, amino acids without handles or with a handle consisting of a single adenine, like in contemporary coenzymes could have been employed. All three scenarios can be tested in in vitro compartmentalized systems. Copyright © 2017 Elsevier B.V. All rights reserved.

Real-world aspects of the nuclear criticality safety program at the University of Tennessee-Knoxville

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bentley, C.L.; Dunn, M.E.; Goluoglu, S.

1996-12-31

The nuclear criticality safety (NCS) program at the University of Tennessee-Knoxville (UTK) emphasizes the {open_quotes}real world{close_quotes} in the NCS courses that are offered and also the NCS research that is conducted. Two NCS courses are offered at UTK. The first course is an introduction to the NCS field, which uses the text by Knief and includes an overview of criticality accidents that have actually happened, standards that are currently in use and being developed, and state-of-the-art computer methods and codes. The students learn the same codes, including both theory and application, that are used by most professionals in the NCSmore » field. Thus, if a student accepts a job offer in the NCS area after graduation, he or she is capable of doing productive NCS work the first day on the job. Subcritical limits, hand-calculation methods, current regulations [both U.S. Department of Energy (DOE) and U.S. Nuclear Regulatory Commission (NRC)] and current practices are also discussed in the introductory course. The second course emphasizes real world experience and is taught by five instructors with over 100 years of combined experience.« less

Summary of papers on current and anticipated uses of thermal-hydraulic codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Caruso, R.

1997-07-01

The author reviews a range of recent papers which discuss possible uses and future development needs for thermal/hydraulic codes in the nuclear industry. From this review, eight common recommendations are extracted. They are: improve the user interface so that more people can use the code, so that models are easier and less expensive to prepare and maintain, and so that the results are scrutable; design the code so that it can easily be coupled to other codes, such as core physics, containment, fission product behaviour during severe accidents; improve the numerical methods to make the code more robust and especiallymore » faster running, particularly for low pressure transients; ensure that future code development includes assessment of code uncertainties as integral part of code verification and validation; provide extensive user guidelines or structure the code so that the `user effect` is minimized; include the capability to model multiple fluids (gas and liquid phase); design the code in a modular fashion so that new models can be added easily; provide the ability to include detailed or simplified component models; build on work previously done with other codes (RETRAN, RELAP, TRAC, CATHARE) and other code validation efforts (CSAU, CSNI SET and IET matrices).« less

General RMP Guidance - Table of Contents

EPA Pesticide Factsheets

The Risk Management Programs for Chemical Accident Prevention (40 CFR Part 68) guidance is in chapters; each covering topics such as applicability of the rule, and requirements for reporting five-year accident history and offsite consequence analysis.

Protecting the AG pilot.

DOT National Transportation Integrated Search

1966-09-01

Summarized data resulting from a detailed review of fatal and non-fatal aerial applicator accidents are presented with an analysis of the factors underlying the accidents. The role of toxic substance is evaluated. Preventive and newer therapeutic app...

Towards a Consolidated Approach for the Assessment of Evaluation Models of Nuclear Power Reactors

DOE PAGES

Epiney, A.; Canepa, S.; Zerkak, O.; ...

2016-11-02

The STARS project at the Paul Scherrer Institut (PSI) has adopted the TRACE thermal-hydraulic (T-H) code for best-estimate system transient simulations of the Swiss Light Water Reactors (LWRs). For analyses involving interactions between system and core, a coupling of TRACE with the SIMULATE-3K (S3K) LWR core simulator has also been developed. In this configuration, the TRACE code and associated nuclear power reactor simulation models play a central role to achieve a comprehensive safety analysis capability. Thus, efforts have now been undertaken to consolidate the validation strategy by implementing a more rigorous and structured assessment approach for TRACE applications involving eithermore » only system T-H evaluations or requiring interfaces to e.g. detailed core or fuel behavior models. The first part of this paper presents the preliminary concepts of this validation strategy. The principle is to systematically track the evolution of a given set of predicted physical Quantities of Interest (QoIs) over a multidimensional parametric space where each of the dimensions represent the evolution of specific analysis aspects, including e.g. code version, transient specific simulation methodology and model "nodalisation". If properly set up, such environment should provide code developers and code users with persistent (less affected by user effect) and quantified information (sensitivity of QoIs) on the applicability of a simulation scheme (codes, input models, methodology) for steady state and transient analysis of full LWR systems. Through this, for each given transient/accident, critical paths of the validation process can be identified that could then translate into defining reference schemes to be applied for downstream predictive simulations. In order to illustrate this approach, the second part of this paper presents a first application of this validation strategy to an inadvertent blowdown event that occurred in a Swiss BWR/6. The transient was initiated by the spurious actuation of the Automatic Depressurization System (ADS). The validation approach progresses through a number of dimensions here: First, the same BWR system simulation model is assessed for different versions of the TRACE code, up to the most recent one. The second dimension is the "nodalisation" dimension, where changes to the input model are assessed. The third dimension is the "methodology" dimension. In this case imposed power and an updated TRACE core model are investigated. For each step in each validation dimension, a common set of QoIs are investigated. For the steady-state results, these include fuel temperatures distributions. For the transient part of the present study, the evaluated QoIs include the system pressure evolution and water carry-over into the steam line.« less

The COSIMA experiments and their verification, a data base for the validation of two phase flow computer codes

NASA Astrophysics Data System (ADS)

Class, G.; Meyder, R.; Stratmanns, E.

1985-12-01

The large data base for validation and development of computer codes for two-phase flow, generated at the COSIMA facility, is reviewed. The aim of COSIMA is to simulate the hydraulic, thermal, and mechanical conditions in the subchannel and the cladding of fuel rods in pressurized water reactors during the blowout phase of a loss of coolant accident. In terms of fuel rod behavior, it is found that during blowout under realistic conditions only small strains are reached. For cladding rupture extremely high rod internal pressures are necessary. The behavior of fuel rod simulators and the effect of thermocouples attached to the cladding outer surface are clarified. Calculations performed with the codes RELAP and DRUFAN show satisfactory agreement with experiments. This can be improved by updating the phase separation models in the codes.

A bill to amend the limitation on liability for certain passenger rail accidents or incidents under section 28103 of title 49, United States Code, and for other purposes.

THOMAS, 111th Congress

Sen. Feinstein, Dianne [D-CA

2010-11-29

Senate - 11/29/2010 Read twice and referred to the Committee on Commerce, Science, and Transportation. (All Actions) Tracker: This bill has the status IntroducedHere are the steps for Status of Legislation:

76 FR 20611 - Electronic On-Board Recorders and Hours of Service Supporting Documents

Federal Register 2010, 2011, 2012, 2013, 2014

2011-04-13

..., used, and disseminated (e.g., in post- accident litigation or in personal litigation such as divorce proceedings). Based on the factors above, the Agency has determined that the statute requires it to protect... Doc. 2011-8789 Filed 4-12-11; 8:45 am] BILLING CODE 4910-EX-P ...

Performance modeling of Deep Burn TRISO fuel using ZrC as a load-bearing layer and an oxygen getter

NASA Astrophysics Data System (ADS)

Wongsawaeng, Doonyapong

2010-01-01

The effects of design choices for the TRISO particle fuel were explored in order to determine their contribution to attaining high-burnup in Deep Burn modular helium reactor fuels containing transuranics from light water reactor spent fuel. The new design features were: (1) ZrC coating substituted for the SiC, allowing the fuel to survive higher accident temperatures; (2) pyrocarbon/SiC "alloy" substituted for the inner pyrocarbon coating to reduce layer failure and (3) pyrocarbon seal coat and thin ZrC oxygen getter coating on the kernel to eliminate CO. Fuel performance was evaluated using General Atomics Company's PISA code. The only acceptable design has a 200-μm kernel diameter coupled with at least 150-μm thick, 50% porosity buffer, a 15-μm ZrC getter over a 10-μm pyrocarbon seal coat on the kernel, an alloy inner pyrocarbon, and ZrC substituted for SiC. The code predicted that during a 1600 °C postulated accident at 70% FIMA, the ZrC failure probability is <10-4.

Transient analysis of ”2 inch Direct Vessel Injection line break” in SPES-2 facility by using TRACE code

NASA Astrophysics Data System (ADS)

D'Amico, S.; Lombardo, C.; Moscato, I.; Polidori, M.; Vella, G.

2015-11-01

In the past few decades a lot of theoretical and experimental researches have been done to understand the physical phenomena characterizing nuclear accidents. In particular, after the Three Miles Island accident, several reactors have been designed to handle successfully LOCA events. This paper presents a comparison between experimental and numerical results obtained for the “2 inch Direct Vessel Injection line break” in SPES-2. This facility is an integral test facility built in Piacenza at the SIET laboratories and simulating the primary circuit, the relevant parts of the secondary circuits and the passive safety systems typical of the AP600 nuclear power plant. The numerical analysis here presented was performed by using TRACE and CATHARE thermal-hydraulic codes with the purpose of evaluating their prediction capability. The main results show that the TRACE model well predicts the overall behaviour of the plant during the transient, in particular it is able to simulate the principal thermal-hydraulic phenomena related to all passive safety systems. The performance of the presented CATHARE noding has suggested some possible improvements of the model.

Developing an ontological explosion knowledge base for business continuity planning purposes.

PubMed

Mohammadfam, Iraj; Kalatpour, Omid; Golmohammadi, Rostam; Khotanlou, Hasan

2013-01-01

Industrial accidents are among the most known challenges to business continuity. Many organisations have lost their reputation following devastating accidents. To manage the risks of such accidents, it is necessary to accumulate sufficient knowledge regarding their roots, causes and preventive techniques. The required knowledge might be obtained through various approaches, including databases. Unfortunately, many databases are hampered by (among other things) static data presentations, a lack of semantic features, and the inability to present accident knowledge as discrete domains. This paper proposes the use of Protégé software to develop a knowledge base for the domain of explosion accidents. Such a structure has a higher capability to improve information retrieval compared with common accident databases. To accomplish this goal, a knowledge management process model was followed. The ontological explosion knowledge base (EKB) was built for further applications, including process accident knowledge retrieval and risk management. The paper will show how the EKB has a semantic feature that enables users to overcome some of the search constraints of existing accident databases.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, Young-Ho; Byun, Thak Sang

Accident-tolerant fuels are expected to have considerably longer coping time to respond to the loss of active cooling under severe accidents and, at the same time, have comparable or improved fuel performance during normal operation. The wear resistance of accident tolerant fuels, therefore, needs to be examined to determine the applicability of these cladding candidates to the current operating PWRs because the most common failure of nuclear fuel claddings is still caused by grid-to-rod fretting during normal operations. In this study, reciprocating sliding wear tests on three kinds of cladding candidates for accident-tolerant fuels have been performed to investigate themore » tribological compatibilities of selfmated cladding candidates and to determine the direct applicability of conventional Zirconium-based alloys as supporting structural materials. The friction coefficients of the cladding candidates are strongly influenced by the test environments and coupled materials. The wear test results under water lubrication conditions indicate that the supporting structural materials for the cladding candidates of accident-tolerant fuels need to be replaced with the same cladding materials instead of using conventional Zirconium-based alloys.« less

Methods for Ensuring High Quality of Coding of Cause of Death. The Mortality Register to Follow Southern Urals Populations Exposed to Radiation.

PubMed

Startsev, N; Dimov, P; Grosche, B; Tretyakov, F; Schüz, J; Akleyev, A

2015-01-01

To follow up populations exposed to several radiation accidents in the Southern Urals, a cause-of-death registry was established at the Urals Center capturing deaths in the Chelyabinsk, Kurgan and Sverdlovsk region since 1950. When registering deaths over such a long time period, quality measures need to be in place to maintain quality and reduce the impact of individual coders as well as quality changes in death certificates. To ensure the uniformity of coding, a method for semi-automatic coding was developed, which is described here. Briefly, the method is based on a dynamic thesaurus, database-supported coding and parallel coding by two different individuals. A comparison of the proposed method for organizing the coding process with the common procedure of coding showed good agreement, with, at the end of the coding process, 70  - 90% agreement for the three-digit ICD -9 rubrics. The semi-automatic method ensures a sufficiently high quality of coding by at the same time providing an opportunity to reduce the labor intensity inherent in the creation of large-volume cause-of-death registries.

Station blackout calculations for Browns Ferry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ott, L.J.; Weber, C.F.; Hyman, C.R.

1985-01-01

This paper presents the results of calculations performed with the ORNL SASA code suite for the Station Blackout Severe Accident Sequence at Browns Ferry. The accident is initiated by a loss of offsite power combined with failure of all onsite emergency diesel generators to start and load. The Station Blackout is assumed to persist beyond the point of battery exhaustion (at six hours) and without DC power, cooling water could no longer be injected into the reactor vessel. Calculations are continued through the period of core degradation and melting, reactor vessel failure, and the subsequent containment failure. An estimate ofmore » the magnitude and timing of the concomitant fission product releases is also provided.« less

Recent plant studies using Victoria 2.0

DOE Office of Scientific and Technical Information (OSTI.GOV)

BIXLER,NATHAN E.; GASSER,RONALD D.

2000-03-08

VICTORIA 2.0 is a mechanistic computer code designed to analyze fission product behavior within the reactor coolant system (RCS) during a severe nuclear reactor accident. It provides detailed predictions of the release of radioactive and nonradioactive materials from the reactor core and transport and deposition of these materials within the RCS and secondary circuits. These predictions account for the chemical and aerosol processes that affect radionuclide behavior. VICTORIA 2.0 was released in early 1999; a new version VICTORIA 2.1, is now under development. The largest improvements in VICTORIA 2.1 are connected with the thermochemical database, which is being revised andmore » expanded following the recommendations of a peer review. Three risk-significant severe accident sequences have recently been investigated using the VICTORIA 2.0 code. The focus here is on how various chemistry options affect the predictions. Additionally, the VICTORIA predictions are compared with ones made using the MELCOR code. The three sequences are a station blackout in a GE BWR and steam generator tube rupture (SGTR) and pump-seal LOCA sequences in a 3-loop Westinghouse PWR. These sequences cover a range of system pressures, from fully depressurized to full system pressure. The chief results of this study are the fission product fractions that are retained in the core, RCS, secondary, and containment and the fractions that are released into the environment.« less

Investigation of technology needs for avoiding helicopter pilot error related accidents

NASA Technical Reports Server (NTRS)

Chais, R. I.; Simpson, W. E.

1985-01-01

Pilot error which is cited as a cause or related factor in most rotorcraft accidents was examined. Pilot error related accidents in helicopters to identify areas in which new technology could reduce or eliminate the underlying causes of these human errors were investigated. The aircraft accident data base at the U.S. Army Safety Center was studied as the source of data on helicopter accidents. A randomly selected sample of 110 aircraft records were analyzed on a case-by-case basis to assess the nature of problems which need to be resolved and applicable technology implications. Six technology areas in which there appears to be a need for new or increased emphasis are identified.

TRAC-PF1/MOD1 support calculations for the MIST/OTIS program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fujita, R.K.; Knight, T.D.

1984-01-01

We are using the Transient Reactor Analysis Code (TRAC), specifically version TRAC-PF1/MOD1, to perform analyses in support of the MultiLoop Integral-System Test (MIST) and the Once-Through Integral-System (OTIS) experiment program. We have analyzed Geradrohr Dampferzeuger Anlage (GERDA) Test 1605AA to benchmark the TRAC-PF1/MOD1 code against phenomena expected to occur in a raised-loop B and W plant during a small-break loss-of-coolant accident (SBLOCA). These results show that the code can calculate both single- and two-phase natural circulation, flow interruption, boiler-condenser-mode (BCM) heat transfer, and primary-system refill in a B and W-type geometry with low-elevation auxiliary feedwater. 19 figures, 7 tables.

The relationship between risk factors and aeronautical decision making in the flight training environment

NASA Astrophysics Data System (ADS)

Wetmore, Michael J.

The purpose of this applied dissertation was to investigate the relationship between risk factors and aeronautical decision making in the flight training environment using a quantitative, non-experimental, ex post facto research design. All 75 of the flight training accidents that involved a fatality from the years 2001-2003 were selected for study from the National Transportation Safety Board (NTSB) aviation accident database. Objective evidence from the Factual Reports was used to construct accident chains and to code and quantify total risk factors and total poor aeronautical decisions. The data were processed using correlational statistical tests at the 1% significance level. There was a statistically significant relationship between total risk factors per flight and poor decisions per flight. Liveware risks were the most prevalent risk factor category. More poor decisions were made during preflight than any other phase of flight. Pilots who made multiple poor decisions per flight had significantly higher risk factors per flight. A risk factor threat to decision making chart is presented for use by flight instructors and/or flight training organizations. The main threat to validity of this study was the NTSB accident investigation team investigative equality assumption.

Traffic Accident Investigation: A Suitable Theme for Teaching Mechanics.

ERIC Educational Resources Information Center

Tao, P. K.

1987-01-01

Suggests the development of curriculum materials on the applications of physics to traffic accident investigations as a theme for teaching mechanics. Describes several standard investigation techniques and the physics principles involved, along with some sample exercises. (TW)

Second-generation antipsychotics and risk of cerebrovascular accidents in the elderly.

PubMed

Percudani, Mauro; Barbui, Corrado; Fortino, Ida; Tansella, Michele; Petrovich, Lorenzo

2005-10-01

Concern has been recently raised for risperidone and olanzapine, possibly associated with cerebrovascular events in placebo-controlled trials conducted in elderly subjects with dementia. We investigated the relationship between exposure to second-generation antipsychotics (SGAs) and occurrence of cerebrovascular accidents in the elderly. From the regional database of hospital admissions of Lombardy, Italy, we extracted all patients aged 65 or older with cerebrovascular-related outcomes for the year 2002. From the regional database of prescriptions reimbursed by the National Health Service, we extracted all patients aged 65 or older who received antipsychotic prescriptions during 2001. The 2 databases were linked anonymously using the individual patient code. The proportions of cerebrovascular accidents were 3.31% (95% confidence interval, 2.95-3.69) in elderly subjects exclusively exposed to SGAs and 2.37% (95% confidence interval, 2.19-2.57) in elderly subjects exclusively exposed to first-generation antipsychotics. After background group differences were controlled for, exposure to SGAs significantly increased the risk of accidents. The analysis of cerebrovascular events in elderly subjects exposed to each individual SGA, in comparison with exposure to haloperidol, showed a significantly increased risk for risperidone only (adjusted odds ratio, 1.43; 95% confidence interval, 1.12-1.93). These data provide preliminary epidemiological evidence that exposure to SGAs, in comparison with exposure to first-generation antipsychotics, significantly increased the risk of cerebrovascular accidents in the elderly.

PRELIMINARY EVALUATION OF FeCrAl CLADDING AND U-Si FUEL FOR ACCIDENT TOLERANT FUEL CONCEPTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hales, J. D.; Gamble, K. A.

2015-09-01

Since the accident at the Fukushima Daiichi Nuclear Power Station, enhancing the accident tolerance of light water reactors (LWRs) has become an important research topic. In particular, the community is actively developing enhanced fuels and cladding for LWRs to improve safety in the event of accidents in the reactor or spent fuel pools. Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system, can tolerate loss of active cooling in the reactor core for a considerably longer time period during design-basis and beyond design-basis events while maintaining or improving the fuel performance during normalmore » operations and operational transients. This paper presents early work in developing thermal and mechanical models for two materials that may have promise: U-Si for fuel, and FeCrAl for cladding. These materials would not necessarily be used together in the same fuel system, but individually have promising characteristics. BISON, the finite element-based fuel performance code in development at Idaho National Laboratory, was used to compare results from normal operation conditions with Zr-4/UO2 behavior. In addition, sensitivity studies are presented for evaluating the relative importance of material parameters such as ductility and thermal conductivity in FeCrAl and U-Si in order to provide guidance on future experiments for these materials.« less

Integrated Speed Limiter and Fatigue Analyzer System

NASA Astrophysics Data System (ADS)

Pranoto, Hadi; Leman, A. M.; Wahab, Abdi; Sebayang, Darwin

2018-03-01

The traffic accident increase in line with the growth of the vehicle, so the safety system must be developed to decrease the accident. This paper will purpose the integrated between speed limiter and fatigue analyser to improve the safety for vehicle, and also to analyse if there is an accident. The device and the software or application are developed and then integrated into one system. The testing held to prove the integrated between device and the application, and it show the system can work well. The next improvement for this system can be developing the server to collect data from internet, so the driver and the vehicle owner can monitor the system by internet.

ARAMIS project: a comprehensive methodology for the identification of reference accident scenarios in process industries.

PubMed

Delvosalle, Christian; Fievez, Cécile; Pipart, Aurore; Debray, Bruno

2006-03-31

In the frame of the Accidental Risk Assessment Methodology for Industries (ARAMIS) project, this paper aims at presenting the work carried out in the part of the project devoted to the definition of accident scenarios. This topic is a key-point in risk assessment and serves as basis for the whole risk quantification. The first result of the work is the building of a methodology for the identification of major accident hazards (MIMAH), which is carried out with the development of generic fault and event trees based on a typology of equipment and substances. The term "major accidents" must be understood as the worst accidents likely to occur on the equipment, assuming that no safety systems are installed. A second methodology, called methodology for the identification of reference accident scenarios (MIRAS) takes into account the influence of safety systems on both the frequencies and possible consequences of accidents. This methodology leads to identify more realistic accident scenarios. The reference accident scenarios are chosen with the help of a tool called "risk matrix", crossing the frequency and the consequences of accidents. This paper presents both methodologies and an application on an ethylene oxide storage.

The MELTSPREAD Code for Modeling of Ex-Vessel Core Debris Spreading Behavior, Code Manual – Version3-beta

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farmer, M. T.

MELTSPREAD3 is a transient one-dimensional computer code that has been developed to predict the gravity-driven flow and freezing behavior of molten reactor core materials (corium) in containment geometries. Predictions can be made for corium flowing across surfaces under either dry or wet cavity conditions. The spreading surfaces that can be selected are steel, concrete, a user-specified material (e.g., a ceramic), or an arbitrary combination thereof. The corium can have a wide range of compositions of reactor core materials that includes distinct oxide phases (predominantly Zr, and steel oxides) plus metallic phases (predominantly Zr and steel). The code requires input thatmore » describes the containment geometry, melt “pour” conditions, and cavity atmospheric conditions (i.e., pressure, temperature, and cavity flooding information). For cases in which the cavity contains a preexisting water layer at the time of RPV failure, melt jet breakup and particle bed formation can be calculated mechanistically given the time-dependent melt pour conditions (input data) as well as the heatup and boiloff of water in the melt impingement zone (calculated). For core debris impacting either the containment floor or previously spread material, the code calculates the transient hydrodynamics and heat transfer which determine the spreading and freezing behavior of the melt. The code predicts conditions at the end of the spreading stage, including melt relocation distance, depth and material composition profiles, substrate ablation profile, and wall heatup. Code output can be used as input to other models such as CORQUENCH that evaluate long term core-concrete interaction behavior following the transient spreading stage. MELTSPREAD3 was originally developed to investigate BWR Mark I liner vulnerability, but has been substantially upgraded and applied to other reactor designs (e.g., the EPR), and more recently to the plant accidents at Fukushima Daiichi. The most recent round of improvements that are documented in this report have been specifically implemented to support industry in developing Severe Accident Water Management (SAWM) strategies for Boiling Water Reactors.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohr, C.L.; Rausch, W.N.; Hesson, G.M.

The LOCA Simulation Program in the NRU reactor is the first set of experiments to provide data on the behavior of full-length, nuclear-heated PWR fuel bundles during the heatup, reflood, and quench phases of a loss-of-coolant accident (LOCA). This paper compares the temperature time histories of 4 experimental test cases with 4 computer codes: CE-THERM, FRAP-T5, GT3-FLECHT, and TRUMP-FLECHT. The preliminary comparisons between prediction and experiment show that the state-of-the art fuel codes have large uncertainties and are not necessarily conservative in predicting peak temperatures, turn around times, and bundle quench times.

Evaluation of Emerging Technologies for Traffic Crash Reporting

DOT National Transportation Integrated Search

1998-02-01

A traffic accident records system is a necessity for a cost-effective safety program at any level of government. The more complete the system, the more potential exists for the application of scarce resources to those accident countermeasures that wi...

Input-output model for MACCS nuclear accident impacts estimation¹

DOE Office of Scientific and Technical Information (OSTI.GOV)

Outkin, Alexander V.; Bixler, Nathan E.; Vargas, Vanessa N

Since the original economic model for MACCS was developed, better quality economic data (as well as the tools to gather and process it) and better computational capabilities have become available. The update of the economic impacts component of the MACCS legacy model will provide improved estimates of business disruptions through the use of Input-Output based economic impact estimation. This paper presents an updated MACCS model, bases on Input-Output methodology, in which economic impacts are calculated using the Regional Economic Accounting analysis tool (REAcct) created at Sandia National Laboratories. This new GDP-based model allows quick and consistent estimation of gross domesticmore » product (GDP) losses due to nuclear power plant accidents. This paper outlines the steps taken to combine the REAcct Input-Output-based model with the MACCS code, describes the GDP loss calculation, and discusses the parameters and modeling assumptions necessary for the estimation of long-term effects of nuclear power plant accidents.« less

Preliminary risks associated with postulated tritium release from production reactor operation

DOE Office of Scientific and Technical Information (OSTI.GOV)

O'Kula, K.R.; Horton, W.H.

1988-01-01

The Probabilistic Risk Assessment (PRA) of Savannah River Plant (SRP) reactor operation is assessing the off-site risk due to tritium releases during postulated full or partial loss of heavy water moderator accidents. Other sources of tritium in the reactor are less likely to contribute to off-site risk in non-fuel melting accident scenarios. Preliminary determination of the frequency of average partial moderator loss (including incidents with leaks as small as .5 kg) yields an estimate of /approximately/1 per reactor year. The full moderator loss frequency is conservatively chosen as 5 /times/ 10/sup /minus/3/ per reactor year. Conditional consequences, determined with amore » version of the MACCS code modified to handle tritium, are found to be insignificant. The 95th percentile individual cancer risk is 4 /times/ 10/sup /minus/8/ per reactor year within 16 km of the release point. The full moderator loss accident contributes about 75% of the evaluated risks. 13 refs., 4 figs., 5 tabs.« less

Enhancing the management response to oil spills in the Tuscany Archipelago through operational modelling.

PubMed

Janeiro, João; Zacharioudaki, Anna; Sarhadi, Ehsan; Neves, Augusto; Martins, Flávio

2014-08-30

A new approach towards the management of oil pollution accidents in marine sensitive areas is presented in this work. A set of nested models in a downscaling philosophy was implemented, externally forced by existing regional operational products. The 3D hydrodynamics, turbulence and the oil transport/weathering models are all linked in the same system, sharing the same code, exchanging information in real time and improving its ability to correctly reproduce the spill. A wind-generated wave model is also implemented using the same downscaling philosophy. Observations from several sources validated the numerical components of the system. The results obtained highlight the good performance of the system and its ability to be applied for oil spill forecasts in the region. The success of the methodology described in this paper was underline during the Costa Concordia accident, where a high resolution domain was rapidly created and deployed inside the system covering the accident site. Copyright © 2014 Elsevier Ltd. All rights reserved.

Heat up and failure of BWR upper internals during a severe accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robb, Kevin R.

In boiling water reactors, the shroud dome, separators, and dryers above the core are made of approximately 100,000 kg of stainless steel. During a severe accident in which the coolant boils away and exothermic oxidation of zirconium occurs, gases (steam and hydrogen) are superheated in the core region and pass through the upper internals. In this scenario, the upper internals can also be heated by thermal radiation from the hot degrading core. Historically, models of the upper internals have been relatively simple in severe accident codes. The upper internals are typically modeled in MELCOR as two lumped volumes with simplifiedmore » heat transfer characteristics and no structural integrity considerations, and with limited ability to oxidize, melt, and relocate. The potential for and the subsequent impact of the upper internals to heat up, oxidize, fail, and relocate during a severe accident was investigated. A higher fidelity representation of the shroud dome, steam separators, and steam driers was developed in MELCOR v1.8.6 by extending the core region upwards. The MELCOR modeling effort entailed adding 45 additional core cells and control volumes, 98 flow paths, and numerous control functions. The model accounts for the mechanical loading and structural integrity, oxidation, melting, flow area blockage, and relocation of the various components. Consistent with a previous study, the results indicate that the upper internals can reach high temperatures during a severe accident sufficient to lose their structural integrity and relocate. Finally, the additional 100 metric tons of stainless steel debris influences the subsequent in-vessel and ex-vessel accident progression.« less

Heat up and failure of BWR upper internals during a severe accident

DOE PAGES

Robb, Kevin R.

2017-02-21

In boiling water reactors, the shroud dome, separators, and dryers above the core are made of approximately 100,000 kg of stainless steel. During a severe accident in which the coolant boils away and exothermic oxidation of zirconium occurs, gases (steam and hydrogen) are superheated in the core region and pass through the upper internals. In this scenario, the upper internals can also be heated by thermal radiation from the hot degrading core. Historically, models of the upper internals have been relatively simple in severe accident codes. The upper internals are typically modeled in MELCOR as two lumped volumes with simplifiedmore » heat transfer characteristics and no structural integrity considerations, and with limited ability to oxidize, melt, and relocate. The potential for and the subsequent impact of the upper internals to heat up, oxidize, fail, and relocate during a severe accident was investigated. A higher fidelity representation of the shroud dome, steam separators, and steam driers was developed in MELCOR v1.8.6 by extending the core region upwards. The MELCOR modeling effort entailed adding 45 additional core cells and control volumes, 98 flow paths, and numerous control functions. The model accounts for the mechanical loading and structural integrity, oxidation, melting, flow area blockage, and relocation of the various components. Consistent with a previous study, the results indicate that the upper internals can reach high temperatures during a severe accident sufficient to lose their structural integrity and relocate. Finally, the additional 100 metric tons of stainless steel debris influences the subsequent in-vessel and ex-vessel accident progression.« less

Applications of the RELAP5 code to the station blackout transients at the Browns Ferry Unit One Plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schultz, R.R.; Wagoner, S.R.

1983-01-01

As a part of the charter of the Severe Accident Sequence Analysis (SASA) Program, station blackout transients have been analyzed using a RELAP5 model of the Browns Ferry Unit 1 Plant. The task was conducted as a partial fulfillment of the needs of the US Nuclear Regulatory Commission in examining the Unresolved Safety Issue A-44: Station Blackout (1) the station blackout transients were examined (a) to define the equipment needed to maintain a well cooled core, (b) to determine when core uncovery would occur given equipment failure, and (c) to characterize the behavior of the vessel thermal-hydraulics during the stationmore » blackout transients (in part as the plant operator would see it). These items are discussed in the paper. Conclusions and observations specific to the station blackout are presented.« less

Repair, Evaluation, Maintenance, and Rehabilitation Research Program. Lock Accident Study

DTIC Science & Technology

1990-09-01

ZIP Code) 10 . SOURCE OF FUNDIN6 NUMBERS -- . ;_ PROGRAM PROJECT TASK WORK UNIT Washington, DC 20314-1000 ELEMENT NO. NO. NO. . NO. 11. TITLE (1 eNy...miwcrwA; I ’+an na SECURITY CLASSIFICATION OF THIS PAGE 10 . WORK UNIT ACCESSION NO. (Continued). Funding provided by Repair, Evaluation, Maintenance, and... 10 PM S ............................................................... 10 District Records

Walk-Rally Support System Using Two-Dimensional Codes and Mobile Phones

ERIC Educational Resources Information Center

Miyagawa, Tetsuya; Yamagishi, Yoshio; Mizuno, Shun

2013-01-01

"Walk Rally" (WR), an orienteering-like recreation game, is common, especially in Japan. Numerous trials to combine WR with educational activities are being carried out by some educators. However, participants are always at the risk of straying and subjected to various accidents during the WR. We developed a WR support system based on…

Reactor vessel lower head integrity

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rubin, A.M.

1997-02-01

On March 28, 1979, the Three Mile Island Unit 2 (TMI-2) nuclear power plant underwent a prolonged small break loss-of-coolant accident that resulted in severe damage to the reactor core. Post-accident examinations of the TMI-2 reactor core and lower plenum found that approximately 19,000 kg (19 metric tons) of molten material had relocated onto the lower head of the reactor vessel. Results of the OECD TMI-2 Vessel Investigation Project concluded that a localized hot spot of approximately 1 meter diameter had existed on the lower head. The maximum temperature on the inner surface of the reactor pressure vessel (RPV) inmore » this region reached 1100{degrees}C and remained at that temperature for approximately 30 minutes before cooling occurred. Even under the combined loads of high temperature and high primary system pressure, the TMI-2 RPV did not fail. (i.e. The pressure varied from about 8.5 to 15 MPa during the four-hour period following the relocation of melt to the lower plenum.) Analyses of RPV failure under these conditions, using state-of-the-art computer codes, predicted that the RPV should have failed via local or global creep rupture. However, the vessel did not fail; and it has been hypothesized that rapid cooling of the debris and the vessel wall by water that was present in the lower plenum played an important role in maintaining RPV integrity during the accident. Although the exact mechanism(s) of how such cooling occurs is not known, it has been speculated that cooling in a small gap between the RPV wall and the crust, and/or in cracks within the debris itself, could result in sufficient cooling to maintain RPV integrity. Experimental data are needed to provide the basis to better understand these phenomena and improve models of RPV failure in severe accident codes.« less

TRANSURANUS: a fuel rod analysis code ready for use

NASA Astrophysics Data System (ADS)

Lassmann, K.

1992-06-01

TRANSURANUS is a computer program for the thermal and mechanical analysis of fuel rods in nuclear reactors and was developed at the European Institute for Transuranium Elements (TUI). The TRANSURANUS code consists of a clearly defined mechanical-mathematical framework into which physical models can easily be incorporated. Besides its flexibility for different fuel rod designs the TRANSURANUS code can deal with very different situations, as given for instance in an experiment, under normal, off-normal and accident conditions. The time scale of the problems to be treated may range from milliseconds to years. The code has a comprehensive material data bank for oxide, mixed oxide, carbide and nitride fuels, Zircaloy and steel claddings and different coolants. During its development great effort was spent on obtaining an extremely flexible tool which is easy to handle, exhibiting very fast running times. The total development effort is approximately 40 man-years. In recent years the interest to use this code grew and the code is in use in several organisations, both research and private industry. The code is now available to all interested parties. The paper outlines the main features and capabilities of the TRANSURANUS code, its validation and treats also some practical aspects.

TRAC-PF1 code verification with data from the OTIS test facility. [Once-Through Intergral System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Childerson, M.T.; Fujita, R.K.

1985-01-01

A computer code (TRAC-PF1/MOD1) developed for predicting transient thermal and hydraulic integral nuclear steam supply system (NSSS) response was benchmarked. Post-small break loss-of-coolant accident (LOCA) data from a scaled, experimental facility, designated the One-Through Integral System (OTIS), were obtained for the Babcock and Wilcox NSSS and compared to TRAC predictions. The OTIS tests provided a challenging small break LOCA data set for TRAC verification. The major phases of a small break LOCA observed in the OTIS tests included pressurizer draining and loop saturation, intermittent reactor coolant system circulation, boiler-condenser mode, and the initial stages of refill. The TRAC code wasmore » successful in predicting OTIS loop conditions (system pressures and temperatures) after modification of the steam generator model. In particular, the code predicted both pool and auxiliary-feedwater initiated boiler-condenser mode heat transfer.« less

77 FR 75499 - Reasonable Charges for Medical Care or Services; V3.12, 2013 Calendar Year Update and National...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-12-20

... automobile accident reparations insurance. The charge tables and supplemental tables that are applicable to...-connected disability incurred as a result of a motor vehicle accident in a State that requires automobile...

Design of a smart, survivable sensor system for rapid transit applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hogan, J.R.; Mitchell, J.L.

1994-08-01

An application of smart sensor technology developed by Sandia National Laboratories has been proposed for real-time monitoring and tracking in the transportation industry. Its primary purpose is to reduce operating costs by improving preventative maintenance scheduling, reducing the number, severity and consequence of accidents and by reducing losses due to theft. The concept uses a strap-on sensor package, the Green Box, that can be attached to any vehicle. The Green Box is designed as a valued-added component, integrated into existing transportation industry systems and standards. The device, designed to provide advanced warning of component failures, would be capable of survivingmore » most typical accidents. In an accident, the system would send a distress signal notifying authorities of the location and condition of the cargo; permitting them to respond in the most effective manner. In addition, the Green Box is adaptable for use as a notification/locator system to enhance the security of operators and passengers for various modes of public transportation. The modular architecture which facilitates system integration in a number of different applications is discussed. A test plan for evaluating performance in both normal and abnormal operating and accident conditions is described.« less

Design and application of a tool for structuring, capitalizing and making more accessible information and lessons learned from accidents involving machinery.

PubMed

Sadeghi, Samira; Sadeghi, Leyla; Tricot, Nicolas; Mathieu, Luc

2017-12-01

Accident reports are published in order to communicate the information and lessons learned from accidents. An efficient accident recording and analysis system is a necessary step towards improvement of safety. However, currently there is a shortage of efficient tools to support such recording and analysis. In this study we introduce a flexible and customizable tool that allows structuring and analysis of this information. This tool has been implemented under TEEXMA®. We named our prototype TEEXMA®SAFETY. This tool provides an information management system to facilitate data collection, organization, query, analysis and reporting of accidents. A predefined information retrieval module provides ready access to data which allows the user to quickly identify the possible hazards for specific machines and provides information on the source of hazards. The main target audience for this tool includes safety personnel, accident reporters and designers. The proposed data model has been developed by analyzing different accident reports.

Construction safety monitoring based on the project's characteristic with fuzzy logic approach

NASA Astrophysics Data System (ADS)

Winanda, Lila Ayu Ratna; Adi, Trijoko Wahyu; Anwar, Nadjadji; Wahyuni, Febriana Santi

2017-11-01

Construction workers accident is the highest number compared with other industries and falls are the main cause of fatal and serious injuries in high rise projects. Generally, construction workers accidents are caused by unsafe act and unsafe condition that can occur separately or together, thus a safety monitoring system based on influencing factors is needed to achieve zero accident in construction industry. The dynamic characteristic in construction causes high mobility for workers while doing the task, so it requires a continuously monitoring system to detect unsafe condition and to protect workers from potential hazards. In accordance with the unique nature of project, fuzzy logic approach is one of the appropriate methods for workers safety monitoring on site. In this study, the focus of discussion is based on the characteristic of construction projects in analyzing "potential hazard" and the "protection planning" to be used in accident prevention. The data have been collected from literature review, expert opinion and institution of safety and health. This data used to determine hazard identification. Then, an application model is created using Delphi programming. The process in fuzzy is divided into fuzzification, inference and defuzzification, according to the data collection. Then, the input and final output data are given back to the expert for assessment as a validation of application model. The result of the study showed that the potential hazard of construction workers accident could be analysed based on characteristic of project and protection system on site and fuzzy logic approach can be used for construction workers accident analysis. Based on case study and the feedback assessment from expert, it showed that the application model can be used as one of the safety monitoring tools.

Main steam line break accident simulation of APR1400 using the model of ATLAS facility

NASA Astrophysics Data System (ADS)

Ekariansyah, A. S.; Deswandri; Sunaryo, Geni R.

2018-02-01

A main steam line break simulation for APR1400 as an advanced design of PWR has been performed using the RELAP5 code. The simulation was conducted in a model of thermal-hydraulic test facility called as ATLAS, which represents a scaled down facility of the APR1400 design. The main steam line break event is described in a open-access safety report document, in which initial conditions and assumptionsfor the analysis were utilized in performing the simulation and analysis of the selected parameter. The objective of this work was to conduct a benchmark activities by comparing the simulation results of the CESEC-III code as a conservative approach code with the results of RELAP5 as a best-estimate code. Based on the simulation results, a general similarity in the behavior of selected parameters was observed between the two codes. However the degree of accuracy still needs further research an analysis by comparing with the other best-estimate code. Uncertainties arising from the ATLAS model should be minimized by taking into account much more specific data in developing the APR1400 model.

Human factors analysis and classification system applied to civil aircraft accidents in India.

PubMed

Gaur, Deepak

2005-05-01

The Human Factors Analysis and Classification System (HFACS) has gained wide acceptance as a tool to classify human factors in aircraft accidents and incidents. This study on application of HFACS to civil aircraft accident reports at Directorate General Civil of Aviation (DGCA), India, was conducted to ascertain the practicability of applying HFACS to existing investigation reports and to analyze the trends of human factor causes of civil aircraft accidents. Accident investigation reports held at DGCA, New Delhi, for the period 1990--99 were scrutinized. In all, 83 accidents occurred during this period, of which 48 accident reports were evaluated in this study. One or more human factors contributed to 37 of the 48 (77.1%) accidents. The commonest unsafe act was 'skill based errors' followed by 'decision errors.' Violations of laid down rules were contributory in 16 cases (33.3%). 'Preconditions for unsafe acts' were seen in 23 of the 48 cases (47.9%). A fairly large number (52.1%) had 'organizational influences' contributing to the accident. These results are in consonance with larger studies of accidents in the U.S. Navy and general aviation. Such a high percentage of 'organizational influences' has not been reported in other studies. This is a healthy sign for Indian civil aviation, provided effective remedial action for the same is undertaken.

UV-induced occupational skin cancer: possibilities of secondary individual prevention in the "Dermatologist's Procedure".

PubMed

Elsner, Peter; Blome, Otto; Diepgen, Thomas Ludwig

2013-07-01

Invasive squamous cell carcinoma (SCC) as a "quasi occupational disease" according to §9 Section 2 of the German Social Code Book (SGB) VII typically develops on chronically UV-damaged skin from actinic keratoses. After the Medical Scientific Committee of the Federal Ministry of Labor and Social Affairs has confirmed the legal criteria for acknowledging UV-induced SCC as an occupational disease, it is expected that the condition will be added to the official list of occupational diseases issued by the Federal Government in the near future. The Social Accident Insurance is required by law (§3 Occupational Disease Regulation) to prevent these tumors by "all appropriate means". There are excellent therapeutic and preventive measures for the management of actinic keratoses to avoid the development of SCC. The "Dermatologist's Procedure" according to §§ 41-43 of the agreement between the Social Accident Insurance and the Federal Medical Association was established in Germany in 1972 to take preventive measures in insured persons with skin lesions possibly developing into an occupational disease, or worsening it, or leading to a recurrence of it This procedure proved to be very successful in the prevention of severe and/or recurring skin diseases forcing a worker to leave his job. On the basis of this agreement, the Social Accident Insurance has the instruments to independently provide preventive measures for the new occupational skin disease SCC induced by natural UV light according to §9 Section 2 of the German Social Code Book (SGB) VII. © The Authors • Journal compilation © Blackwell Verlag GmbH, Berlin.

Assessment of environmental public exposure from a hypothetical nuclear accident for Unit-1 Bushehr nuclear power plant.

PubMed

Sohrabi, M; Ghasemi, M; Amrollahi, R; Khamooshi, C; Parsouzi, Z

2013-05-01

Unit-1 of the Bushehr nuclear power plant (BNPP-1) is a VVER-type reactor with 1,000-MWe power constructed near Bushehr city at the coast of the Persian Gulf, Iran. The reactor has been recently operational to near its full power. The radiological impact of nuclear power plant (NPP) accidents is of public concern, and the assessment of radiological consequences of any hypothetical nuclear accident on public exposure is vital. The hypothetical accident scenario considered in this paper is a design-basis accident, that is, a primary coolant leakage to the secondary circuit. This scenario was selected in order to compare and verify the results obtained in the present paper with those reported in the Final Safety Analysis Report (FSAR 2007) of the BNPP-1 and to develop a well-proven methodology that can be used to study other and more severe hypothetical accident scenarios for this reactor. In the present study, the version 2.01 of the PC COSYMA code was applied. In the early phase of the accidental releases, effective doses (from external and internal exposures) as well as individual and collective doses (due to the late phase of accidental releases) were evaluated. The surrounding area of the BNPP-1 within a radius of 80 km was subdivided into seven concentric rings and 16 sectors, and distribution of population and agricultural products was calculated for this grid. The results show that during the first year following the modeled hypothetical accident, the effective doses do not exceed the limit of 5 mSv, for the considered distances from the BNPP-1. The results obtained in this study are in good agreement with those in the FSAR-2007 report. The agreement obtained is in light of many inherent uncertainties and variables existing in the two modeling procedures applied and proves that the methodology applied here can also be used to model other severe hypothetical accident scenarios of the BNPP-1 such as a small and large break in the reactor coolant system as well as beyond design-basis accidents. Such scenarios are planned to be studied in the near future, for this reactor.

[Maculopathy caused by Nd:YAG laser accident].

PubMed

Blümel, C; Brosig, J

1999-02-01

Since the construction of the first laser in the sixties and the extended use in medicine, technology and hobby the number of accidents has increased. Appreciated to therapy concepts are missing at the time. A 19 year-old-man was hit by the impulse of an military hand-held rangefinder (Nd:YAG with a wavelength of 1064 nm) on the right eye. The visual acuity dropped to 1/35 and a central scotoma with metamorphopsia occurred immediatly after the accident. The ophthalmological findings showed a distinct submacular hemorrhage. The therapy with Prednisolon intravenous and daily parabulbar, vitamin C, indomethacin systemical and lokal application resulted in an increase of visual acuity up to 0.4 and a reduction of central scotoma from 8 degrees to 2 degrees. Systemical and local use of antiphlogistic and antiinflamatoric substances may partially reduce the vision limitating scar formation. Application of antioxidants to neutralize the toxic radicals that arise by tissue decay should be given additionally to the cyclopegic medication. Special attention should be payed to the prevention of such laser accidents.

Visualization of Traffic Accidents

NASA Technical Reports Server (NTRS)

Wang, Jie; Shen, Yuzhong; Khattak, Asad

2010-01-01

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

Sleepiness, driving, and motor vehicle accidents: A questionnaire-based survey.

PubMed

Zwahlen, Daniel; Jackowski, Christian; Pfäffli, Matthias

2016-11-01

In Switzerland, the prevalence of an excessive daytime sleepiness (EDS) in drivers undergoing a driving capacity assessment is currently not known. In this study, private and professional drivers were evaluated by means of a paper-based questionnaire, including Epworth Sleepiness Scale, Berlin Questionnaire, and additional questions to sleepiness-related accidents, near-miss accidents, health issues, and demographic data. Of the 435 distributed questionnaires, 128 completed were returned. The response rate was 29%. The mean age of the investigated drivers was 42.5 years (20-85 years). According to the Epworth Sleepiness Scale, 9% of the participants are likely to suffer from excessive daytime sleepiness. An equal percentage has a high risk for obstructive sleep apnea syndrome based on the Berlin Questionnaire. 16% admitted an involuntary nodding off while driving a motor vehicle. This subset of the participants scored statistically significant higher on the Epworth Sleepiness Scale (p = 0.036). 8% of the participants already suffered an accident because of being sleepy while driving. An equal number experienced a sleepiness-related near-miss accident on the road. The study shows that a medical workup of excessive daytime sleepiness is highly recommended in each driver undergoing a driving capacity assessment. Routine application of easily available and time-saving assessment tools such as the Epworth Sleepiness Scale questionnaire could prevent accidents in a simple way. The applicability of the Berlin Questionnaire to screen suspected fatal sleepiness-related motor vehicle accidents is discussed. Copyright © 2016 Elsevier Ltd and Faculty of Forensic and Legal Medicine. All rights reserved.

49 CFR Appendix A to Part 99 - Categories of Financial Interests Exempted From the Prohibitions of Section 208(a) of Title 18...

Code of Federal Regulations, 2010 CFR

2010-10-01

..., health, or accident insurance plan or other employee welfare or benefit plan that is maintained by a... Transportation Office of the Secretary of Transportation EMPLOYEE RESPONSIBILITIES AND CONDUCT Pt. 99, App. A... States Code, because they are too remote or too inconsequential to affect the integrity of an employee's...

75 FR 54940 - Agency Information Collection (IC) Activities; Revision of an Approved IC; Accident Recordkeeping...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-09

... include the Agency name and the docket number for this Notice. Note that DOT posts all comments received... underlying this IC is 49 CFR 390.15, ``Assistance in investigations and special studies.'' It requires motor... Information Technology. [FR Doc. 2010-22456 Filed 9-8-10; 8:45 am] BILLING CODE 4910-EX-P ...

Deprivation and mortality in non-metropolitan areas of England and Wales.

PubMed Central

Jessop, E G

1996-01-01

OBJECTIVE: To test the hypothesis that the relationship between deprivation and mortality is weaker among residents of non-metropolitan areas of England and Wales than among residents of metropolitan areas. DESIGN: This study compared mortality, expressed as standardised mortality ratios (SMRs), in residents of metropolitan and non-metropolitan districts at three levels of deprivation classified by an electoral ward deprivation score and by home and car ownership. SMRs were computed for all causes of death, for bronchitis and asthma (ICD9 codes 490-493), and for accident, violence, and poisoning (ICD9 codes 800-999). SETTING: England and Wales. PARTICIPANTS: Members of the longitudinal study of the Office of Population Censuses and Surveys, a quasi-random 1% sample of the population of England and Wales. MAIN RESULTS: There was an association between deprivation and mortality which was clear for all cause mortality, more noticeable for respiratory disease, and less clear for deaths from accident, violence, and poison. In general, the results showed a remarkable similarity between metropolitan and non-metropolitan areas. CONCLUSIONS: This study does not support the hypothesis that the relationship between mortality and deprivation differs between residents of metropolitan and non-metropolitan areas of England and Wales. PMID:8944858

Deprivation and mortality in non-metropolitan areas of England and Wales.

PubMed

Jessop, E G

1996-10-01

To test the hypothesis that the relationship between deprivation and mortality is weaker among residents of non-metropolitan areas of England and Wales than among residents of metropolitan areas. This study compared mortality, expressed as standardised mortality ratios (SMRs), in residents of metropolitan and non-metropolitan districts at three levels of deprivation classified by an electoral ward deprivation score and by home and car ownership. SMRs were computed for all causes of death, for bronchitis and asthma (ICD9 codes 490-493), and for accident, violence, and poisoning (ICD9 codes 800-999). England and Wales. Members of the longitudinal study of the Office of Population Censuses and Surveys, a quasi-random 1% sample of the population of England and Wales. There was an association between deprivation and mortality which was clear for all cause mortality, more noticeable for respiratory disease, and less clear for deaths from accident, violence, and poison. In general, the results showed a remarkable similarity between metropolitan and non-metropolitan areas. This study does not support the hypothesis that the relationship between mortality and deprivation differs between residents of metropolitan and non-metropolitan areas of England and Wales.

Containment Sodium Chemistry Models in MELCOR.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Louie, David; Humphries, Larry L.; Denman, Matthew R

To meet regulatory needs for sodium fast reactors’ future development, including licensing requirements, Sandia National Laboratories is modernizing MELCOR, a severe accident analysis computer code developed for the U.S. Nuclear Regulatory Commission (NRC). Specifically, Sandia is modernizing MELCOR to include the capability to model sodium reactors. However, Sandia’s modernization effort primarily focuses on the containment response aspects of the sodium reactor accidents. Sandia began modernizing MELCOR in 2013 to allow a sodium coolant, rather than water, for conventional light water reactors. In the past three years, Sandia has been implementing the sodium chemistry containment models in CONTAIN-LMR, a legacy NRCmore » code, into MELCOR. These chemistry models include spray fire, pool fire and atmosphere chemistry models. Only the first two chemistry models have been implemented though it is intended to implement all these models into MELCOR. A new package called “NAC” has been created to manage the sodium chemistry model more efficiently. In 2017 Sandia began validating the implemented models in MELCOR by simulating available experiments. The CONTAIN-LMR sodium models include sodium atmosphere chemistry and sodium-concrete interaction models. This paper presents sodium property models, the implemented models, implementation issues, and a path towards validation against existing experimental data.« less

Road Traffic Accident Victims’ Experiences of Return to Normal Life: A Qualitative Study

PubMed Central

Pashaei Sabet, Fatemeh; Norouzi Tabrizi, Kian; Khankeh, Hamid Reza; Saadat, Soheil; Abedi, Heidar Ali; Bastami, Alireza

2016-01-01

Background Road traffic accident (RTA) victims also suffer from different types of injuries and disabilities, which can affect their quality of life. They usually face with various physical, mental, and social problems. Most traffic accident victims had difficulty to return to normal life. Objectives This study aimed to understand the experiences of return to normal life in RTA victims. Patients and Methods This qualitative study with content analysis approach was conducted on 18 Iranian patients with disability in the upper or lower limbs caused by traffic accidents, who had passed a time between 3 months till 2 years. A purposeful sampling method was applied until reaching data saturation. Data were collected using semi-structured interviews. Afterwards, the gathered data were analyzed through conventional content analysis. Results By analyzing 498 primary codes, four main categories, including supportive needs, adaptation to the new situation, seeking information, and transition from functional limitation, were extracted from traffic accident victims’ experiences of reintegration to normal life. Conclusions The results of this study may help policy-makers to take steps toward health promotion and recovery of RTA victims. Considering the results of this study, it is a need for further research to investigate RTAs victims’ needs for reintegration to home and community. Access to training and supportive facilities like strong therapeutic, nursing and social support, and the possibility to participate in self-care activities is essential for reintegration to community in RTA victims. PMID:27275399

Hazard Prevention Regarding Occupational Accidents Involving Blue-Collar Foreign Workers: A Perspective of Taiwanese Manpower Agencies

PubMed Central

Chang, Huan-Cheng; Wang, Mei-Chin; Liao, Hung-Chang; Cheng, Shu-Fang; Wang, Ya-huei

2016-01-01

Since 1989, blue-collar foreign workers have been permitted to work in Taiwanese industries. Most blue-collar foreign workers apply for jobs in Taiwan through blue-collar foreign workers’ agencies. Because blue-collar foreign workers are not familiar with the language and culture in Taiwan, in occupational accident education and hazard prevention, the agencies play an important role in the coordination and translation between employees and blue-collar foreign workers. The purpose of this study is to establish the agencies’ role in the occupational accidents education and hazard prevention for blue-collar foreign workers in Taiwan. This study uses a qualitative method—grounded theory—to collect, code, and analyze the data in order to understand the agencies’ role in occupational accident education and hazard prevention for blue-collar foreign workers in Taiwan. The results show that the duty of agencies in occupational accident education and hazard prevention includes selecting appropriate blue-collar foreign workers, communicating between employees and blue-collar foreign workers, collecting occupational safety and health information, assisting in the training of occupational safety and health, and helping blue-collar foreign workers adapt to their lives in Taiwan. Finally, this study suggests seven important points and discusses the implementation process necessary to improve governmental policies. The government and employees should pay attention to the education/training of occupational safety and health for blue-collar foreign workers to eliminate unsafe behavior in order to protect the lives of blue-collar foreign workers. PMID:27420085

Hazard Prevention Regarding Occupational Accidents Involving Blue-Collar Foreign Workers: A Perspective of Taiwanese Manpower Agencies.

PubMed

Chang, Huan-Cheng; Wang, Mei-Chin; Liao, Hung-Chang; Cheng, Shu-Fang; Wang, Ya-Huei

2016-07-13

Since 1989, blue-collar foreign workers have been permitted to work in Taiwanese industries. Most blue-collar foreign workers apply for jobs in Taiwan through blue-collar foreign workers' agencies. Because blue-collar foreign workers are not familiar with the language and culture in Taiwan, in occupational accident education and hazard prevention, the agencies play an important role in the coordination and translation between employees and blue-collar foreign workers. The purpose of this study is to establish the agencies' role in the occupational accidents education and hazard prevention for blue-collar foreign workers in Taiwan. This study uses a qualitative method-grounded theory-to collect, code, and analyze the data in order to understand the agencies' role in occupational accident education and hazard prevention for blue-collar foreign workers in Taiwan. The results show that the duty of agencies in occupational accident education and hazard prevention includes selecting appropriate blue-collar foreign workers, communicating between employees and blue-collar foreign workers, collecting occupational safety and health information, assisting in the training of occupational safety and health, and helping blue-collar foreign workers adapt to their lives in Taiwan. Finally, this study suggests seven important points and discusses the implementation process necessary to improve governmental policies. The government and employees should pay attention to the education/training of occupational safety and health for blue-collar foreign workers to eliminate unsafe behavior in order to protect the lives of blue-collar foreign workers.

Preliminary Three-Dimensional Simulation of Sediment and Cesium Transport in the Ogi Dam Reservoir using FLESCOT – Task 6, Subtask 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Onishi, Yasuo; Kurikami, Hiroshi; Yokuda, Satoru T.

2014-03-28

After the accident at the Fukushima Daiichi Nuclear Power Plant in March 2011, the Japan Atomic Energy Agency and the Pacific Northwest National Laboratory initiated a collaborative project on environmental restoration. In October 2013, the collaborative team started a task of three-dimensional modeling of sediment and cesium transport in the Fukushima environment using the FLESCOT (Flow, Energy, Salinity, Sediment Contaminant Transport) code. As the first trial, we applied it to the Ogi Dam Reservoir that is one of the reservoirs in the Japan Atomic Energy Agency’s (JAEA’s) investigation project. Three simulation cases under the following different temperature conditions were studied:more » • incoming rivers and the Ogi Dam Reservoir have the same water temperature • incoming rivers have lower water temperature than that of the reservoir • incoming rivers have higher water temperature than that of the reservoir. The preliminary simulations suggest that seasonal temperature changes influence the sediment and cesium transport. The preliminary results showed the following: • Suspended sand, and cesium adsorbed by sand, coming into the reservoirs from upstream rivers is deposited near the reservoir entrance. • Suspended silt, and cesium adsorbed by silt, is deposited farther in the reservoir. • Suspended clay, and cesium adsorbed by clay, travels the farthest into the reservoir. With sufficient time, the dissolved cesium reaches the downstream end of the reservoir. This preliminary modeling also suggests the possibility of a suitable dam operation to control the cesium migration farther downstream from the dam. JAEA has been sampling in the Ogi Dam Reservoir, but these data were not yet available for the current model calibration and validation for this reservoir. Nonetheless these preliminary FLESCOT modeling results were qualitatively valid and confirmed the applicability of the FLESCOT code to the Ogi Dam Reservoir, and in general to other reservoirs in the Fukushima environment. The issues to be addressed in future are the following: • Validate the simulation results by comparison with the investigation data. • Confirm the applicability of the FLESCOT code to Fukushima coastal areas. • Increase computation speed by parallelizing the FLESCOT code.« less

Simulation of internal contamination screening with dose rate meters

NASA Astrophysics Data System (ADS)

Fonseca, T. C. F.; Mendes, B. M.; Hunt, J. G.

2017-11-01

Assessing the intake of radionuclides after an accident in a nuclear power plant or after the intentional release of radionuclides in public places allows dose calculations and triage actions to be carried out for members of the public and for emergency response teams. Gamma emitters in the lung, thyroid or the whole body may be detected and quantified by making dose rate measurements at the surface of the internally contaminated person. In an accident scenario, quick measurements made with readily available portable equipment are a key factor for success. In this paper, the Monte Carlo program Visual Monte Carlo (VMC) and MCNPx code are used in conjunction with voxel phantoms to calculate the dose rate at the surface of a contaminated person due to internally deposited radionuclides. A whole body contamination with 137Cs and a thyroid contamination with 131I were simulated and the calibration factors in kBq per μSv/h were calculated. The calculated calibration factors were compared with real data obtained from the Goiania accident in the case of 137Cs and the Chernobyl accident in terms of the 131I. The close comparison of the calculated and real measurements indicates that the method may be applied to other radionuclides. Minimum detectable activities are discussed.

OSL properties of three commonly available salt brands in India for its use in accident dosimetry

NASA Astrophysics Data System (ADS)

Singh, A. K.; Menon, S. N.; Kadam, S. Y.; Koul, D. K.; Datta, D.

2018-03-01

Thermally stimulated luminescence (TL) and Optically Stimulated Luminescence (OSL) characterization of three commonly available salt brands in India were undertaken for their application in accident dosimetry. The investigations showed that the luminescence properties differed to some extent with that reported in literature. Dosimetric properties of these salt samples showed that these can be useful in accident dosimetry. Based on the sensitization and fading behaviour of the samples a Single Aliquot Regenerative (SAR) protocol has been proposed for dose estimation.

Measurement of long-lived radionuclides in surface soil around F1NPP accident site by Accelerator Mass Spectrometry

NASA Astrophysics Data System (ADS)

Miyake, Yasuto; Matsuzaki, Hiroyuki; Sasa, Kimikazu; Takahashi, Tsutomu

2015-10-01

In March 2011, vast amounts of radionuclides were released into the environment due to the Fukushima Daiichi Nuclear Power Plant (F1NPP) accident. However, very little work has been done concerning accident-derived long-lived nuclides such as 129I (T1/2 = 1.57 × 107 year) and 36Cl (T1/2 = 3.01 × 105 year). 129I and 131I are both produced by 235U fission in nuclear reactors. Being isotopes of iodine, these nuclides are expected to behave similarly in the environment. This makes 129I useful for retrospective reconstruction of 131I distribution during the initial stages of the accident. On the other hand, 36Cl is generated during reactor operation via neutron capture reaction of 35Cl, an impurity in the coolant or reactor component. Resulting 36Cl/Cl ratio within the reactor is thus much higher compared to that in environment. Similar to 129I, 36Cl is expected to have leaked out during the accident and it is important to evaluate its effects. In this study, 129I concentrations were determined in several surface soil samples collected around F1NPP. Average 129I/131I ratio was estimated to be 26.1 ± 5.8 as of March 11, 2011, consistent with calculations using ORIGEN2 code and other published data. 36Cl/Cl ratios in some of the soil samples were likewise measured and ranged from 1.1 × 10-12 to 2.6 × 10-11. These are higher compared to ratios measured around F1NPP before the accident. A positive correlation between 36Cl and 129I concentration was observed.

45 CFR 2553.43 - What cost reimbursements are provided to RSVP volunteers?

Code of Federal Regulations, 2014 CFR

2014-10-01

...: (1) Accident insurance. Accident insurance covers RSVP volunteers for personal injury during travel...) Excess automobile liability insurance. (i) For RSVP volunteers who drive in connection with their service... volunteers carry on their own automobiles; or (B) The limits of applicable state financial responsibility law...

45 CFR 2553.43 - What cost reimbursements are provided to RSVP volunteers?

Code of Federal Regulations, 2010 CFR

2010-10-01

...: (1) Accident insurance. Accident insurance covers RSVP volunteers for personal injury during travel...) Excess automobile liability insurance. (i) For RSVP volunteers who drive in connection with their service... volunteers carry on their own automobiles; or (B) The limits of applicable state financial responsibility law...

45 CFR 2553.43 - What cost reimbursements are provided to RSVP volunteers?

Code of Federal Regulations, 2013 CFR

2013-10-01

...: (1) Accident insurance. Accident insurance covers RSVP volunteers for personal injury during travel...) Excess automobile liability insurance. (i) For RSVP volunteers who drive in connection with their service... volunteers carry on their own automobiles; or (B) The limits of applicable state financial responsibility law...

45 CFR 2553.43 - What cost reimbursements are provided to RSVP volunteers?

Code of Federal Regulations, 2011 CFR

2011-10-01

...: (1) Accident insurance. Accident insurance covers RSVP volunteers for personal injury during travel...) Excess automobile liability insurance. (i) For RSVP volunteers who drive in connection with their service... volunteers carry on their own automobiles; or (B) The limits of applicable state financial responsibility law...

45 CFR 2553.43 - What cost reimbursements are provided to RSVP volunteers?

Code of Federal Regulations, 2012 CFR

2012-10-01

...: (1) Accident insurance. Accident insurance covers RSVP volunteers for personal injury during travel...) Excess automobile liability insurance. (i) For RSVP volunteers who drive in connection with their service... volunteers carry on their own automobiles; or (B) The limits of applicable state financial responsibility law...

Safe Driving Knowledge Dissemination and Testing Techniques. Volume II: Final Report.

ERIC Educational Resources Information Center

McKnight, James; Green, Molly A.

In order to determine the effectiveness of improved information dissemination and assessment techniques in reducing highway accidents, a set of seven targeted driver license manuals and tests were developed for the following groups of drivers: new drivers, youthful drivers, renewal applicants, older drivers, traffic violators, accident repeaters,…

Safe Driving Knowledge Dissemination and Testing Techniques. Volume 1: General Findings.

ERIC Educational Resources Information Center

McKnight, James; Green, Molly A.

In order to determine the effectiveness of improved information dissemination and assessment techniques in reducing highway accidents, a set of seven targeted driver license manuals and tests were developed for the following groups of drivers: new drivers, youthful drivers, renewal applicants, older drivers, traffic violators, accident repeaters,…

The Role of Trust and Interaction in Global Positioning System Related Accidents

NASA Technical Reports Server (NTRS)

Johnson, Chris W.; Shea, Christine; Holloway, C. Michael

2008-01-01

The Global Positioning System (GPS) uses a network of satellites to calculate the position of a receiver over time. This technology has revolutionized a wide range of safety-critical industries and leisure applications. These systems provide diverse benefits; supplementing the users existing navigation skills and reducing the uncertainty that often characterizes many route planning tasks. GPS applications can also help to reduce workload by automating tasks that would otherwise require finite cognitive and perceptual resources. However, the operation of these systems has been identified as a contributory factor in a range of recent accidents. Users often come to rely on GPS applications and, therefore, fail to notice when they develop faults or when errors occur in the other systems that use the data from these systems. Further accidents can stem from the over confidence that arises when users assume automated warnings will be issued when they stray from an intended route. Unless greater attention is paid to the role of trust and interaction in GPS applications then there is a danger that we will see an increasing number of these failures as positioning technologies become integral in the functioning of increasing numbers of applications.

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

PubMed

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.

[Accidents with biological material in health care workers in 2 primary health care areas (1990-1999)].

PubMed

Hernández Navarrete, M J; Montes Villameriel, F J; Solano Bernad, V M; Sánchez Matienzo, D; del Val García, J L; Gil Montalbán, E; Arribas Llorente, J L

2001-09-15

To find out the exposures with biological material in health care workers in primary health care, registered in the biological accidents database from Preventive Medicine Service in Miguel Servet Universitary Hospital of Zaragoza. Descriptive study of a retrospective cohort. SITE: Primary health care, Areas II and V of Zaragoza.Participants. Workers in this areas, distributed by: physician, nursing staff, auxiliary, orderly, housekeeping staff, others. Data of: workers, accident, serologic source, worker protection and vaccinal status of hepatitis B. The incidence of accidents was 26 (period 1997-1999). Most proportion of accidents were declared by nursing (78%). The highest occupational incidence was in auxiliary (63 ). In 90,1% of the cases, the accident was needlestick injury. The source was known in 67,7% of cases. The accidents occurred in hands in 96,8% of cases, and only one third of workers carried gloves. Results obtained are similar with previous studies about this event. We must insist on the need to declare these accidents, providing more information and accessibility for the declaration to worker. Moreover, we must insist on the correct application in the health care field of the standard precautions, because almost 50% of accidents are evitable, and to increase hepatitis B vaccination covertures.

Precursors of dangerous substances formed in the loss of control of chemical systems.

PubMed

Cozzani, V; Zanelli, S

1999-03-01

Article 2 of Directive 96/82/EC on the control of major accident hazards caused by dangerous substances requires to consider also the hazards due to the dangerous substances "which it is believed may be generated during loss of control of an industrial chemical process", although no generally accepted guidelines are available for the identification of these substances. In the present study, the accidents involving the unwanted formation of dangerous substances as a consequence of the loss of control of chemical systems were investigated. A specifically developed database was used, containing data on more than 400 of these accidents and on the substances involved. The hazardous substances formed in the accidents and the precursors of these substances were identified. The influence of accident characteristics on the substances formed was investigated. In the context of the application of Directive 96/82/EC, an accident severity index and a hazard rating of the precursors of dangerous substances formed in the accidents were proposed. A lumping approach was used in order to develop schemes for the preliminary identification of substances that may be formed in the loss of control of chemical system. The results of accident analysis were used to test the schemes developed.

Categorizing accident sequences in the external radiotherapy for risk analysis

PubMed Central

2013-01-01

Purpose This study identifies accident sequences from the past accidents in order to help the risk analysis application to the external radiotherapy. Materials and Methods This study reviews 59 accidental cases in two retrospective safety analyses that have collected the incidents in the external radiotherapy extensively. Two accident analysis reports that accumulated past incidents are investigated to identify accident sequences including initiating events, failure of safety measures, and consequences. This study classifies the accidents by the treatments stages and sources of errors for initiating events, types of failures in the safety measures, and types of undesirable consequences and the number of affected patients. Then, the accident sequences are grouped into several categories on the basis of similarity of progression. As a result, these cases can be categorized into 14 groups of accident sequence. Results The result indicates that risk analysis needs to pay attention to not only the planning stage, but also the calibration stage that is committed prior to the main treatment process. It also shows that human error is the largest contributor to initiating events as well as to the failure of safety measures. This study also illustrates an event tree analysis for an accident sequence initiated in the calibration. Conclusion This study is expected to provide sights into the accident sequences for the prospective risk analysis through the review of experiences. PMID:23865005

Development and evaluation of a Naïve Bayesian model for coding causation of workers' compensation claims.

PubMed

Bertke, S J; Meyers, A R; Wurzelbacher, S J; Bell, J; Lampl, M L; Robins, D

2012-12-01

Tracking and trending rates of injuries and illnesses classified as musculoskeletal disorders caused by ergonomic risk factors such as overexertion and repetitive motion (MSDs) and slips, trips, or falls (STFs) in different industry sectors is of high interest to many researchers. Unfortunately, identifying the cause of injuries and illnesses in large datasets such as workers' compensation systems often requires reading and coding the free form accident text narrative for potentially millions of records. To alleviate the need for manual coding, this paper describes and evaluates a computer auto-coding algorithm that demonstrated the ability to code millions of claims quickly and accurately by learning from a set of previously manually coded claims. The auto-coding program was able to code claims as a musculoskeletal disorders, STF or other with approximately 90% accuracy. The program developed and discussed in this paper provides an accurate and efficient method for identifying the causation of workers' compensation claims as a STF or MSD in a large database based on the unstructured text narrative and resulting injury diagnoses. The program coded thousands of claims in minutes. The method described in this paper can be used by researchers and practitioners to relieve the manual burden of reading and identifying the causation of claims as a STF or MSD. Furthermore, the method can be easily generalized to code/classify other unstructured text narratives. Published by Elsevier Ltd.

Terrorism-related injuries versus road traffic accident-related trauma: 5 years of experience in Israel.

PubMed

Peleg, Kobi; Savitsky, Bella

2009-12-01

Terrorism victims comprise the minority among trauma injured people, but this small population imposes a burden on the health care system. Thirty percent of the population injured in terrorist activities experienced severe trauma (injury severity score > or =16), more than half of them need a surgical procedure, and 25% of the population affected by terrorism had been admitted to intensive care. Furthermore, compared with patients with non-terrorism-related trauma, victims of terrorism often arrive in bulk, as part of a mass casualty event. This poses a sudden load on hospital resources and requires special organization and preparedness. The present study compared terrorism-related and road accident-related injuries and examined clinical characteristics of both groups of patients. This study is a retrospective study of all patients injured through terrorist acts and road traffic accidents from September 29, 2000 to December 31, 2005, and recorded in the Israel Trauma Registry. Data on the nature of injuries, treatment, and outcome were obtained from the registry. Medical diagnoses were extracted from the registry and classified based on International Classification of Diseases coding. Diagnoses were grouped to body regions, based on the Barell Injury Diagnosis Matrix. The study includes 2197 patients with terrorism-related injuries and 30,176 patients injured in road traffic accidents. All in all, 27% of terrorism-related casualties suffered severe to critical injuries, comparing to 17% among road traffic accident-related victims. Glasgow Coma Scale scores

A review of human factors causations in commercial air transport accidents and incidents: From to 2000-2016

NASA Astrophysics Data System (ADS)

Kharoufah, Husam; Murray, John; Baxter, Glenn; Wild, Graham

2018-05-01

Human factors have been defined by the International Civil Aviation Organization (ICAO) as "about people in their living and working situations; about their relationship with machines, with procedures and with the environment about them; and about their relationships with other people (at work)". Human factors contribute to approximately 75% of aircraft accidents and incidents. As such, understanding their influence is essential to improve safety in the aviation industry. This study examined the different human factors causations in a random sample of over 200 commercial air transport accidents and incidents from 2000 to 2016. The main objective of this study was to identify the principal human factor contributions to aviation accidents and incidents. An exploratory research design was utilised. The qualitative data were recorded in a database, and were coded into categories about the flights (including date, manufacturer, carrier, state of occurrence, etc). These categories were then analysed using Chi-Squared tests to determine which were statistically significant in terms of having an influence on the accidents/incidents. The most significant human factor was found to be situational awareness followed by non-adherence to procedures. In addition, charter operations proved to have a significantly higher rate of human factor related occurrence as compared to other type of operations. A significant finding was that Africa has a high rate of accidents/incidents relative to the amount of traffic and aircraft movements. These findings reflect some of the more noteworthy incidents that have received significant media attention, including Air Asia 8501 on the 28th of December 2014, TransAsia Airways 235 on the 4th of February 2015, and Air France 447 on the 1st of June 2009; these accidents resulted in a significant loss of lives where situational awareness and non-adherence to procedures were significant contributing factors.

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

PubMed

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

2000-01-01

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

Wheelchair-related accidents: relationship with wheelchair-using behavior in active community wheelchair users.

PubMed

Chen, Wan-Yin; Jang, Yuh; Wang, Jung-Der; Huang, Wen-Ni; Chang, Chan-Chia; Mao, Hui-Fen; Wang, Yen-Ho

2011-06-01

To report the prevalence, mechanisms, self-perceived causes, consequences, and wheelchair-using behaviors associated with wheelchair-related accidents. A case-control study. Community. A sample of experienced, community-dwelling, active manual and powered wheelchair users (N=95) recruited from a hospital assistive technology service center. Not applicable. Wheelchair-using behaviors, wheelchair-related accidents over a 3-year period, and the mechanisms and consequences of the accidents. Among the 95 participants, 52 (54.7%) reported at least 1 accident and 16 (16.8%) reported 2 or more accidents during the 3 years prior to the interview. A total of 74 accidents, were categorized into tips and falls (87.8%), accidental contact (6.8%), and dangerous operations (5.4%). A logistic regression found individuals who failed to maintain their wheelchairs regularly (odds ratio [OR]=11.28; 95% confidence interval [CI], 2.62-48.61) and used a wheelchair not prescribed by professionals (OR=4.31; 95% CI, 1.10-16.82) had significantly greater risks of accidents. In addition to the risk factor, lack of regular wheelchair maintenance, the Poisson regression corroborated the other risk factor, seat belts not used (incident rate ratio=2.14; 95% CI, 1.08-4.14), for wheelchair-related accidents. Wheelchair-related accidents are closely related to their wheelchair-using behaviors. Services including professional evaluation, repair, maintenance, and an educational program on proper wheelchair use may decrease the risks of wheelchair accidents. Copyright © 2011 American Congress of Rehabilitation Medicine. Published by Elsevier Inc. All rights reserved.

NSRD-10: Leak Path Factor Guidance Using MELCOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Louie, David; Humphries, Larry L.

Estimates of the source term from a U.S. Department of Energy (DOE) nuclear facility requires that the analysts know how to apply the simulation tools used, such as the MELCOR code, particularly for a complicated facility that may include an air ventilation system and other active systems that can influence the environmental pathway of the materials released. DOE has designated MELCOR 1.8.5, an unsupported version, as a DOE ToolBox code in its Central Registry, which includes a leak-path-factor guidance report written in 2004 that did not include experimental validation data. To continue to use this MELCOR version requires additional verificationmore » and validations, which may not be feasible from a project cost standpoint. Instead, the recent MELCOR should be used. Without any developer support and lack of experimental data validation, it is difficult to convince regulators that the calculated source term from the DOE facility is accurate and defensible. This research replaces the obsolete version in the 2004 DOE leak path factor guidance report by using MELCOR 2.1 (the latest version of MELCOR with continuing modeling development and user support) and by including applicable experimental data from the reactor safety arena and from applicable experimental data used in the DOE-HDBK-3010. This research provides best practice values used in MELCOR 2.1 specifically for the leak path determination. With these enhancements, the revised leak-path-guidance report should provide confidence to the DOE safety analyst who would be using MELCOR as a source-term determination tool for mitigated accident evaluations.« less

Renewed Commitment to Excellence: An Assessment of the NASA Agency-Wide Applicability of the Columbia Accident Investigation Board Report

NASA Technical Reports Server (NTRS)

2004-01-01

The Space Shuttle fleet has been grounded since the Columbia accident. As a result, 'Return to Flight' has become not just a phrase but a program and the global of virtually everyone associated with NASA. Even those who are not affiliated with the Shuttle Program are looking forward to the safe and successful completion of the next Shuttle mission. In this recovery process, NASA will be guided by the Report of the Columbia Accident Investigation Board (CAIB). The CAIB was an investigating body, convened by NASA Administrator O'Keefe the day of the Columbia accident, according to procedures established after the loss of Space Challenger.

Improvement of COBRA-TF for modeling of PWR cold- and hot-legs during reactor transients

NASA Astrophysics Data System (ADS)

Salko, Robert K.

COBRA-TF is a two-phase, three-field (liquid, vapor, droplets) thermal-hydraulic modeling tool that has been developed by the Pacific Northwest Laboratory under sponsorship of the NRC. The code was developed for Light Water Reactor analysis starting in the 1980s; however, its development has continued to this current time. COBRA-TF still finds wide-spread use throughout the nuclear engineering field, including nuclear-power vendors, academia, and research institutions. It has been proposed that extension of the COBRA-TF code-modeling region from vessel-only components to Pressurized Water Reactor (PWR) coolant-line regions can lead to improved Loss-of-Coolant Accident (LOCA) analysis. Improved modeling is anticipated due to COBRA-TF's capability to independently model the entrained-droplet flow-field behavior, which has been observed to impact delivery to the core region[1]. Because COBRA-TF was originally developed for vertically-dominated, in-vessel, sub-channel flow, extension of the COBRA-TF modeling region to the horizontal-pipe geometries of the coolant-lines required several code modifications, including: • Inclusion of the stratified flow regime into the COBRA-TF flow regime map, along with associated interfacial drag, wall drag and interfacial heat transfer correlations, • Inclusion of a horizontal-stratification force between adjacent mesh cells having unequal levels of stratified flow, and • Generation of a new code-input interface for the modeling of coolant-lines. The sheer number of COBRA-TF modifications that were required to complete this work turned this project into a code-development project as much as it was a study of thermal-hydraulics in reactor coolant-lines. The means for achieving these tasks shifted along the way, ultimately leading the development of a separate, nearly completely independent one-dimensional, two-phase-flow modeling code geared toward reactor coolant-line analysis. This developed code has been named CLAP, for Coolant-Line-Analysis Package. Versions were created that were both coupled to COBRA-TF and standalone, with the most recent version being a standalone code. This code performs a separate, simplified, 1-D solution of the conservation equations while making special considerations for coolant-line geometry and flow phenomena. The end of this project saw a functional code package that demonstrates a stable numerical solution and that has gone through a series of Validation and Verification tests using the Two-Phase Testing Facility (TPTF) experimental data[2]. The results indicate that CLAP is under-performing RELAP5-MOD3 in predicting the experimental void of the TPTF facility in some cases. There is no apparent pattern, however, to point to a consistent type of case that the code fails to predict properly (e.g., low-flow, high-flow, discharging to full vessel, or discharging to empty vessel). Pressure-profile predictions are sometimes unrealistic, which indicates that there may be a problem with test-case boundary conditions or with the coupling of continuity and momentum equations in the solution algorithm. The code does predict the flow regime correctly for all cases with the stratification-force model off. Turning the stratification model on can cause the low-flow case void profiles to over-react to the force and the flow regime to transition out of stratified flow. The code would benefit from an increased amount of Validation & Verification testing. The development of CLAP was significant, as it is a cleanly written, logical representation of the reactor coolant-line geometry. It is stable and capable of modeling basic flow physics in the reactor coolant-line. Code development and debugging required the temporary removal of the energy equation and mass-transfer terms in governing equations. The reintroduction of these terms will allow future coupling to RELAP and re-coupling with COBRA-TF. Adding in more applicable entrainment and de-entrainment models would allow the capture of more advanced physics in the coolant-line that can be expected during Loss-of-Coolant Accident. One of the package's benefits is its ability to be used as a platform for future coolant-line model development and implementation, including capturing of the important de-entrainment behavior in reactor hot-legs (steam-binding effect) and flow convection in the upper-plenum region of the vessel.

A Police and Insurance Joint Management System Based on High Precision BDS/GPS Positioning

PubMed Central

Zuo, Wenwei; Guo, Chi; Liu, Jingnan; Peng, Xuan; Yang, Min

2018-01-01

Car ownership in China reached 194 million vehicles at the end of 2016. The traffic congestion index (TCI) exceeds 2.0 during rush hour in some cities. Inefficient processing for minor traffic accidents is considered to be one of the leading causes for road traffic jams. Meanwhile, the process after an accident is quite troublesome. The main reason is that it is almost always impossible to get the complete chain of evidence when the accident happens. Accordingly, a police and insurance joint management system is developed which is based on high precision BeiDou Navigation Satellite System (BDS)/Global Positioning System (GPS) positioning to process traffic accidents. First of all, an intelligent vehicle rearview mirror terminal is developed. The terminal applies a commonly used consumer electronic device with single frequency navigation. Based on the high precision BDS/GPS positioning algorithm, its accuracy can reach sub-meter level in the urban areas. More specifically, a kernel driver is built to realize the high precision positioning algorithm in an Android HAL layer. Thus the third-party application developers can call the general location Application Programming Interface (API) of the original standard Global Navigation Satellite System (GNSS) to get high precision positioning results. Therefore, the terminal can provide lane level positioning service for car users. Next, a remote traffic accident processing platform is built to provide big data analysis and management. According to the big data analysis of information collected by BDS high precision intelligent sense service, vehicle behaviors can be obtained. The platform can also automatically match and screen the data that uploads after an accident to achieve accurate reproduction of the scene. Thus, it helps traffic police and insurance personnel to complete remote responsibility identification and survey for the accident. Thirdly, a rapid processing flow is established in this article to meet the requirements to quickly handle traffic accidents. The traffic police can remotely identify accident responsibility and the insurance personnel can remotely survey an accident. Moreover, the police and insurance joint management system has been carried out in Wuhan, Central China’s Hubei Province, and Wuxi, Eastern China’s Jiangsu Province. In a word, a system is developed to obtain and analyze multisource data including precise positioning and visual information, and a solution is proposed for efficient processing of traffic accidents. PMID:29320406

A Police and Insurance Joint Management System Based on High Precision BDS/GPS Positioning.

PubMed

Zuo, Wenwei; Guo, Chi; Liu, Jingnan; Peng, Xuan; Yang, Min

2018-01-10

Car ownership in China reached 194 million vehicles at the end of 2016. The traffic congestion index (TCI) exceeds 2.0 during rush hour in some cities. Inefficient processing for minor traffic accidents is considered to be one of the leading causes for road traffic jams. Meanwhile, the process after an accident is quite troublesome. The main reason is that it is almost always impossible to get the complete chain of evidence when the accident happens. Accordingly, a police and insurance joint management system is developed which is based on high precision BeiDou Navigation Satellite System (BDS)/Global Positioning System (GPS) positioning to process traffic accidents. First of all, an intelligent vehicle rearview mirror terminal is developed. The terminal applies a commonly used consumer electronic device with single frequency navigation. Based on the high precision BDS/GPS positioning algorithm, its accuracy can reach sub-meter level in the urban areas. More specifically, a kernel driver is built to realize the high precision positioning algorithm in an Android HAL layer. Thus the third-party application developers can call the general location Application Programming Interface (API) of the original standard Global Navigation Satellite System (GNSS) to get high precision positioning results. Therefore, the terminal can provide lane level positioning service for car users. Next, a remote traffic accident processing platform is built to provide big data analysis and management. According to the big data analysis of information collected by BDS high precision intelligent sense service, vehicle behaviors can be obtained. The platform can also automatically match and screen the data that uploads after an accident to achieve accurate reproduction of the scene. Thus, it helps traffic police and insurance personnel to complete remote responsibility identification and survey for the accident. Thirdly, a rapid processing flow is established in this article to meet the requirements to quickly handle traffic accidents. The traffic police can remotely identify accident responsibility and the insurance personnel can remotely survey an accident. Moreover, the police and insurance joint management system has been carried out in Wuhan, Central China's Hubei Province, and Wuxi, Eastern China's Jiangsu Province. In a word, a system is developed to obtain and analyze multisource data including precise positioning and visual information, and a solution is proposed for efficient processing of traffic accidents.

System Simulation of Nuclear Power Plant by Coupling RELAP5 and Matlab/Simulink

DOE Office of Scientific and Technical Information (OSTI.GOV)

Meng Lin; Dong Hou; Zhihong Xu

2006-07-01

Since RELAP5 code has general and advanced features in thermal-hydraulic computation, it has been widely used in transient and accident safety analysis, experiment planning analysis, and system simulation, etc. So we wish to design, analyze, verify a new Instrumentation And Control (I and C) system of Nuclear Power Plant (NPP) based on the best-estimated code, and even develop our engineering simulator. But because of limited function of simulating control and protection system in RELAP5, it is necessary to expand the function for high efficient, accurate, flexible design and simulation of I and C system. Matlab/Simulink, a scientific computation software, justmore » can compensate the limitation, which is a powerful tool in research and simulation of plant process control. The software is selected as I and C part to be coupled with RELAP5 code to realize system simulation of NPPs. There are two key techniques to be solved. One is the dynamic data exchange, by which Matlab/Simulink receives plant parameters and returns control results. Database is used to communicate the two codes. Accordingly, Dynamic Link Library (DLL) is applied to link database in RELAP5, while DLL and S-Function is applied in Matlab/Simulink. The other problem is synchronization between the two codes for ensuring consistency in global simulation time. Because Matlab/Simulink always computes faster than RELAP5, the simulation time is sent by RELAP5 and received by Matlab/Simulink. A time control subroutine is added into the simulation procedure of Matlab/Simulink to control its simulation advancement. Through these ways, Matlab/Simulink is dynamically coupled with RELAP5. Thus, in Matlab/Simulink, we can freely design control and protection logic of NPPs and test it with best-estimated plant model feedback. A test will be shown to illuminate that results of coupling calculation are nearly the same with one of single RELAP5 with control logic. In practice, a real Pressurized Water Reactor (PWR) is modeled by RELAP5 code, and its main control and protection system is duplicated by Matlab/Simulink. Some steady states and transients are calculated under control of these I and C systems, and the results are compared with the plant test curves. The application showed that it can do exact system simulation of NPPs by coupling RELAP5 and Matlab/Simulink. This paper will mainly focus on the coupling method, plant thermal-hydraulic model, main control logics, test and application results. (authors)« less

Could driving safety be compromised by noise exposure at work and noise-induced hearing loss?

PubMed

Picard, Michel; Girard, Serge André; Courteau, Marilène; Leroux, Tony; Larocque, Richard; Turcotte, Fernand; Lavoie, Michel; Simard, Marc

2008-10-01

A study was conducted to verify if there is an association between occupational noise exposure, noise-induced hearing loss and driving safety expanding on previous findings by Picard, et al. (2008) that the two factors did increase accident risk in the workplace. This study was made possible when driving records of all Quebec drivers were made available by the Societe de l'assurance automobile du Quebec (SAAQ is the state monopoly responsible for the provision of motor vehicle insurance and the compensation of victims of traffic accidents). These records were linked with personal records maintained by the Quebec National Institute of Public Health as part of its mission to prevent noise induced hearing loss in the workplace. Individualized information on occupational noise exposure and hearing sensitivity was available for 46,030 male workers employed in noisy industries who also held a valid driver's permit. The observation period is of five years duration, starting with the most recent audiometric examination. The associations between occupational noise exposure levels, hearing status, and personal driving record were examined by log-binomial regression on data adjusted for age and duration of exposure. Daily noise exposures and bilateral average hearing threshold levels at 3, 4, and 6 kHz were used as independent variables while the dependent variables were 1) the number of motor vehicle accidents experienced by participants during the study period and 2) participants' records of registered traffic violations of the highway safety code. The findings are reported as prevalence ratios (PRs) with their 95% confidence intervals (CIs). Attributable numbers of events were computed with the relevant PRs, lesser-noise, exposed workers and those with normal hearing levels making the group of reference. Adjusting for age confirmed that experienced workers had fewer traffic accidents. The data show that occupational noise exposure and hearing loss have the same effect on driving safety record than that reported on the risk of accident in noisy industrial settings. Specifically, the risk of traffic accident (PR = 1.07 (CI 95% [1.01; 1.15]) is significantly associated with the daily occupational noise exposures >or= 100 dBA. For participants having a bilateral average hearing loss ranging from 16 to 30 dB, the PR of traffic accident is 1.06 (CI 95% [1.01; 1.11]) and reaches 1.31 (CI 95% [1.2; 1.42]) when the hearing loss exceeds of 50 dB. A reduction in the number of speeding violations occurred among workers occupationally exposed to noise levels >or= 90 dBA and those with noise-induced hearing loss >or=16 dB. By contrast, the same individuals had an increase in other violations of the Highway safety code. This suggests that noise-exposed workers might be less vigilant to other traffic hazards. Daily occupational noise exposures >or= 100 dBA and noise-induced hearing losses-even when just barely noticeable-may interfere with the safe operation of motor vehicles.

A novel framework for improvement of road accidents considering decision-making styles of drivers in a large metropolitan area.

PubMed

Azadeh, Ali; Zarrin, Mansour; Hamid, Mehdi

2016-02-01

Road accidents can be caused by different factors such as human factors. Quality of the decision-making process of drivers could have a considerable impact on preventing disasters. The main objective of this study is the analysis of factors affecting road accidents by considering the severity of accidents and decision-making styles of drivers. To this end, a novel framework is proposed based on data envelopment analysis (DEA) and statistical methods (SMs) to assess the factors affecting road accidents. In this study, for the first time, dominant decision-making styles of drivers with respect to severity of injuries are identified. To show the applicability of the proposed framework, this research employs actual data of more than 500 samples in Tehran, Iran. The empirical results indicate that the flexible decision style is the dominant style for both minor and severe levels of accident injuries. Copyright © 2015 Elsevier Ltd. All rights reserved.

Empirical Bayesian Geographical Mapping of Occupational Accidents among Iranian Workers.

PubMed

Vahabi, Nasim; Kazemnejad, Anoshirvan; Datta, Somnath

2017-05-01

Work-related accidents are believed to be a serious preventable cause of mortality and disability worldwide. This study aimed to provide Bayesian geographical maps of occupational injury rates among workers insured by the Iranian Social Security Organization. The participants included all insured workers in the Iranian Social Security Organization database in 2012. One of the applications of the Bayesian approach called the Poisson-Gamma model was applied to estimate the relative risk of occupational accidents. Data analysis and mapping were performed using R 3.0.3, Open-Bugs 3.2.3 rev 1012 and ArcMap9.3. The majority of all 21,484 investigated occupational injury victims were male (98.3%) including 16,443 (76.5%) single workers aged 20 - 29 years. The accidents were more frequent in basic metal, electric, and non-electric machining jobs. About 0.4% (96) of work-related accidents led to death, 2.2% (457) led to disability (partial and total), 4.6% (980) led to fixed compensation, and 92.8% (19,951) of the injured victims recovered completely. The geographical maps of estimated relative risk of occupational accidents were also provided. The results showed that the highest estimations pertained to provinces which were mostly located along mountain chains, some of which are categorized as deprived provinces in Iran. The study revealed the need for further investigation of the role of economic and climatic factors in high risk areas. The application of geographical mapping together with statistical approaches can provide more accurate tools for policy makers to make better decisions in order to prevent and reduce the risks and adverse outcomes of work-related accidents.

BISON Modeling of Reactivity-Initiated Accident Experiments in a Static Environment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Folsom, Charles P.; Jensen, Colby B.; Williamson, Richard L.

2016-09-01

In conjunction with the restart of the TREAT reactor and the design of test vehicles, modeling and simulation efforts are being used to model the response of Accident Tolerant Fuel (ATF) concepts under reactivity insertion accident (RIA) conditions. The purpose of this work is to model a baseline case of a 10 cm long UO2-Zircaloy fuel rodlet using BISON and RELAP5 over a range of energy depositions and with varying reactor power pulse widths. The results show the effect of varying the pulse width and energy deposition on both thermal and mechanical parameters that are important for predicting failure ofmore » the fuel rodlet. The combined BISON/RELAP5 model captures coupled thermal and mechanical effects on the fuel-to-cladding gap conductance, cladding-to-coolant heat transfer coefficient and water temperature and pressure that would not be capable in each code individually. These combined effects allow for a more accurate modeling of the thermal and mechanical response in the fuel rodlet and thermal-hydraulics of the test vehicle.« less

NSRD-15:Computational Capability to Substantiate DOE-HDBK-3010 Data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Louie, David; Bignell, John; Dingreville, Remi Philippe Michel

Safety basis analysts throughout the U.S. Department of Energy (DOE) complex rely heavily on the information provided in the DOE Handbook, DOE-HDBK-3010, Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities, to determine radionuclide source terms from postulated accident scenarios. In calculating source terms, analysts tend to use the DOE Handbook’s bounding values on airborne release fractions (ARFs) and respirable fractions (RFs) for various categories of insults (representing potential accident release categories). This is typically due to both time constraints and the avoidance of regulatory critique. Unfortunately, these bounding ARFs/RFs represent extremely conservative values. Moreover, they were derived frommore » very limited small-scale bench/laboratory experiments and/or from engineered judgment. Thus, the basis for the data may not be representative of the actual unique accident conditions and configurations being evaluated. The goal of this research is to develop a more accurate and defensible method to determine bounding values for the DOE Handbook using state-of-art multi-physics-based computer codes.« less

Characterization of thermal-hydraulic and ignition phenomena in prototypic, full-length boiling water reactor spent fuel pool assemblies after a complete loss-of-coolant accident.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lindgren, Eric Richard; Durbin, Samuel G

2007-04-01

The objective of this project was to provide basic thermal-hydraulic data associated with a SFP complete loss-of-coolant accident. The accident conditions of interest for the SFP were simulated in a full-scale prototypic fashion (electrically-heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate code validation (primarily MELCOR) and reduce questions associated with interpretation of the experimental results. It was necessary to simulate a cluster of assemblies to represent a higher decay (younger) assembly surrounded by older, lower-power assemblies. Specifically, this program providedmore » data and analysis confirming: (1) MELCOR modeling of inter-assembly radiant heat transfer, (2) flow resistance modeling and the natural convective flow induced in a fuel assembly as it heats up in air, (3) the potential for and nature of thermal transient (i.e., Zircaloy fire) propagation, and (4) mitigation strategies concerning fuel assembly management.« less

CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kotas, J.F.; Stroh, K.R.

1983-01-01

The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident thatmore » simulates a control-rod withdrawal at full power.« less

A new system for digital image acquisition, storage and presentation in an accident and emergency department

PubMed Central

Clegg, G; Roebuck, S; Steedman, D

2001-01-01

Objectives—To develop a computer based storage system for clinical images—radiographs, photographs, ECGs, text—for use in teaching, training, reference and research within an accident and emergency (A&E) department. Exploration of methods to access and utilise the data stored in the archive. Methods—Implementation of a digital image archive using flatbed scanner and digital camera as capture devices. A sophisticated coding system based on ICD 10. Storage via an "intelligent" custom interface. Results—A practical solution to the problems of clinical image storage for teaching purposes. Conclusions—We have successfully developed a digital image capture and storage system, which provides an excellent teaching facility for a busy A&E department. We have revolutionised the practice of the "hand-over meeting". PMID:11435357

LARC-1: a Los Alamos release calculation program for fission product transport in HTGRs during the LOFC accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carruthers, L.M.; Lee, C.E.

1976-10-01

The theoretical and numerical data base development of the LARC-1 code is described. Four analytical models of fission product release from an HTGR core during the loss of forced circulation accident are developed. Effects of diffusion, adsorption and evaporation of the metallics and precursors are neglected in this first LARC model. Comparison of the analytic models indicates that the constant release-renormalized model is adequate to describe the processes involved. The numerical data base for release constants, temperature modeling, fission product release rates, coated fuel particle failure fraction and aged coated fuel particle failure fractions is discussed. Analytic fits and graphicmore » displays for these data are given for the Ft. St. Vrain and GASSAR models.« less

41 CFR 102-39.60 - What restrictions and prohibitions apply to the exchange/sale of personal property?

Code of Federal Regulations, 2010 CFR

2010-07-01

... for FSC Class 1005, Guns through 30mm. Deviations are not required for Department of Defense (DoD... this title. (e) Property with a condition code of scrap, as defined at FMR 102-36.40, except: (1... to damage (e.g., accident or natural disaster); or (3) Scrap gold for fine gold. (f) Property that...

Calculations of skyshine from an intense portable electron linac

DOE Office of Scientific and Technical Information (OSTI.GOV)

Estes, G.P.; Hughes, H.G.; Fry, D.A.

1994-12-31

The MCNP Monte carlo code has been used at Los Alamos to calculate skyshine and terrain albedo efects from an intense portable electron linear accelerator that is to be used by the Russian Federation to radiograph nuclear weapons that may have been damaged by accidents. Relative dose rate profiles have been calculated. The design of the accelerator, along with a diagram, is presented.

45 CFR 3.4 - False reports and reports of injury or damage.

Code of Federal Regulations, 2010 CFR

2010-10-01

... PERSONS AND TRAFFIC ON THE NATIONAL INSTITUTES OF HEALTH FEDERAL ENCLAVE General § 3.4 False reports and... accident or violation of the regulations of this part or any applicable Federal or Maryland statute to any person properly investigating an accident or alleged violation. All incidents resulting in injury to...

26 CFR 1.72-15 - Applicability of section 72 to accident or health plans.

Code of Federal Regulations, 2010 CFR

2010-04-01

... retirement and the payment of an earlier pension in the event of permanent disability. This section will also... presumed that the disability pension is provided by employer contributions, unless the plan expressly... or inclusion of accident or health benefits under sections 104 and 105. For example, the investment...

77 FR 67679 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-13

... diesel generator surveillance requirements. Margin of safety is related to the ability of the fission... surveillance tests, allowing testing in any MODE of operation. The Division 3 AC sources, including the diesel generator (DG) and its associated emergency loads are accident mitigating features, not accident initiators...

[Company health promotion as a possible preparatory stage of effective rehabilitation].

PubMed

Köpke, K-H

2012-01-01

In Section 20a of Book 5 of the German Social Code (SGB V), the legislator obliged the health insurance funds to carry out company health promotion in cooperation with the accident insurance agency in charge. He thus created a foundation for more health and secure earnings or employment potential in the companies. At the same time company health promotion permits detecting threats to this potential. That helps to identify a possible need for rehabilitation at an earlier stage and to take appropriate action.To verify whether and how this instrument of preventive health policy is being used, an empirical study explored the actual application of that legal provision in small and medium-sized enterprises in particular. The law, administrative measures and company everyday evidence were set against each other under legal and de facto aspects, which showed obvious deficits in applying the law. Proposals for better company health promotion are derived from these findings. In the first place, actors in administration and self-management of the statutory health and accident insurance schemes are addressed to this end, in anticipation of enhanced implementation. A premature reduction of earning capacity could thus be counteracted. Pension insurants could retain gainful employment for a longer time, companies would have a more reliable employee basis. Social insurance carriers, notably the health and pension insurance schemes, would have to spend less in the end. A development like that would be a benefit for all--including the state. © Georg Thieme Verlag KG Stuttgart · New York.

Formulating accident occurrence as a survival process.

PubMed

Chang, H L; Jovanis, P P

1990-10-01

A conceptual framework for accident occurrence is developed based on the principle of the driver as an information processor. The framework underlies the development of a modeling approach that is consistent with the definition of exposure to risk as a repeated trial. Survival theory is proposed as a statistical technique that is consistent with the conceptual structure and allows the exploration of a wide range of factors that contribute to highway operating risk. This survival model of accident occurrence is developed at a disaggregate level, allowing safety researchers to broaden the scope of studies which may be limited by the use of traditional aggregate approaches. An application of the approach to motor carrier safety is discussed as are potential applications to a variety of transportation industries. Lastly, a typology of highway safety research methodologies is developed to compare the properties of four safety methodologies: laboratory experiments, on-the-road studies, multidisciplinary accident investigations, and correlational studies. The survival theory formulation has a mathematical structure that is compatible with each safety methodology, so it may facilitate the integration of findings across methodologies.

Analyses in support of risk-informed natural gas vehicle maintenance facility codes and standards :

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ekoto, Isaac W.; Blaylock, Myra L.; LaFleur, Angela Christine

2014-03-01

Safety standards development for maintenance facilities of liquid and compressed gas fueled large-scale vehicles is required to ensure proper facility design and operation envelopes. Standard development organizations are utilizing risk-informed concepts to develop natural gas vehicle (NGV) codes and standards so that maintenance facilities meet acceptable risk levels. The present report summarizes Phase I work for existing NGV repair facility code requirements and highlights inconsistencies that need quantitative analysis into their effectiveness. A Hazardous and Operability study was performed to identify key scenarios of interest. Finally, scenario analyses were performed using detailed simulations and modeling to estimate the overpressure hazardsmore » from HAZOP defined scenarios. The results from Phase I will be used to identify significant risk contributors at NGV maintenance facilities, and are expected to form the basis for follow-on quantitative risk analysis work to address specific code requirements and identify effective accident prevention and mitigation strategies.« less

Task 6 - Subtask 1: PNNL Visit by JAEA Researchers to Evaluate the Feasibility of the FLESCOT Code for the Future JAEA Use for the Fukushima Surface Water Environmental Assessment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Onishi, Yasuo

Four Japan Atomic Energy Agency (JAEA) researchers visited Pacific Northwest National Laboratory (PNNL) for seven working days and have evaluated the suitability and adaptability of FLESCOT to a JAEA’s supercomputer system to effectively simulate cesium behavior in dam reservoirs, river mouths, and coastal areas in Fukushima contaminated by the Fukushima Daiichi nuclear accident. PNNL showed the following to JAEA visitors during the seven-working day period: FLESCOT source code; User’s manual; FLESCOT description – Program structure – Algorism – Solver – Boundary condition handling – Data definition – Input and output methods – How to run. During the visit, JAEA hadmore » access to FLESCOT to run with an input data set to evaluate the capacity and feasibility of adapting it to a JAEA super computer with massive parallel processors. As a part of this evaluation, PNNL ran FLESCOT for sample cases of the contaminant migration simulation to further describe FLESCOT in action. JAEA and PNNL researchers also evaluated time spent for each subroutine of FLESCOT, and the JAEA researcher implemented some initial parallelization schemes to FLESCOT. Based on this code evaluation, JAEA and PNNL determined that FLESCOT is: applicable to Fukushima lakes/dam reservoirs, river mouth areas, and coastal water; and feasible to implement parallelization for the JAEA supercomputer. In addition, PNNL and JAEA researchers discussed molecular modeling approaches on cesium adsorption mechanisms to enhance the JAEA molecular modeling activities. PNNL and JAEA also discussed specific collaboration of molecular and computational modeling activities.« less

SecPop Version 4: Sector Population Land Fraction and Economic Estimation Program: Users? Guide Model Manual and Verification Report.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Weber, Scott; Bixler, Nathan E.; McFadden, Katherine Letizia

In 1973 the U.S. Environmental Protection Agency (EPA) developed SecPop to calculate population estimates to support a study on air quality. The Nuclear Regulatory Commission (NRC) adopted this program to support siting reviews for nuclear power plant construction and license applications. Currently SecPop is used to prepare site data input files for offsite consequence calculations with the MELCOR Accident Consequence Code System (MACCS). SecPop enables the use of site-specific population, land use, and economic data for a polar grid defined by the user. Updated versions of SecPop have been released to use U.S. decennial census population data. SECPOP90 was releasedmore » in 1997 to use 1990 population and economic data. SECPOP2000 was released in 2003 to use 2000 population data and 1997 economic data. This report describes the current code version, SecPop version 4.3, which uses 2010 population data and both 2007 and 2012 economic data. It is also compatible with 2000 census and 2002 economic data. At the time of this writing, the current version of SecPop is 4.3.0, and that version is described herein. This report contains guidance for the installation and use of the code as well as a description of the theory, models, and algorithms involved. This report contains appendices which describe the development of the 2010 census file, 2007 county file, and 2012 county file. Finally, an appendix is included that describes the validation assessments performed.« less

Application of systems and control theory-based hazard analysis to radiation oncology.

PubMed

Pawlicki, Todd; Samost, Aubrey; Brown, Derek W; Manger, Ryan P; Kim, Gwe-Ya; Leveson, Nancy G

2016-03-01

Both humans and software are notoriously challenging to account for in traditional hazard analysis models. The purpose of this work is to investigate and demonstrate the application of a new, extended accident causality model, called systems theoretic accident model and processes (STAMP), to radiation oncology. Specifically, a hazard analysis technique based on STAMP, system-theoretic process analysis (STPA), is used to perform a hazard analysis. The STPA procedure starts with the definition of high-level accidents for radiation oncology at the medical center and the hazards leading to those accidents. From there, the hierarchical safety control structure of the radiation oncology clinic is modeled, i.e., the controls that are used to prevent accidents and provide effective treatment. Using STPA, unsafe control actions (behaviors) are identified that can lead to the hazards as well as causal scenarios that can lead to the identified unsafe control. This information can be used to eliminate or mitigate potential hazards. The STPA procedure is demonstrated on a new online adaptive cranial radiosurgery procedure that omits the CT simulation step and uses CBCT for localization, planning, and surface imaging system during treatment. The STPA procedure generated a comprehensive set of causal scenarios that are traced back to system hazards and accidents. Ten control loops were created for the new SRS procedure, which covered the areas of hospital and department management, treatment design and delivery, and vendor service. Eighty three unsafe control actions were identified as well as 472 causal scenarios that could lead to those unsafe control actions. STPA provides a method for understanding the role of management decisions and hospital operations on system safety and generating process design requirements to prevent hazards and accidents. The interaction of people, hardware, and software is highlighted. The method of STPA produces results that can be used to improve safety and prevent accidents and warrants further investigation.

Rotating shift work, sleep, and accidents related to sleepiness in hospital nurses

NASA Technical Reports Server (NTRS)

Gold, D. R.; Rogacz, S.; Bock, N.; Tosteson, T. D.; Baum, T. M.; Speizer, F. E.; Czeisler, C. A.

1992-01-01

A hospital-based survey on shift work, sleep, and accidents was carried out among 635 Massachusetts nurses. In comparison to nurses who worked only day/evening shifts, rotators had more sleep/wake cycle disruption and nodded off more at work. Rotators had twice the odds of nodding off while driving to or from work and twice the odds of a reported accident or error related to sleepiness. Application of circadian principles to the design of hospital work schedules may result in improved health and safety for nurses and patients.

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

NASA Astrophysics Data System (ADS)

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.

Development of MPS Method for Analyzing Melt Spreading Behavior and MCCI in Severe Accidents

NASA Astrophysics Data System (ADS)

Yamaji, Akifumi; Li, Xin

2016-08-01

Spreading of molten core (corium) on reactor containment vessel floor and molten corium-concrete interaction (MCCI) are important phenomena in the late phase of a severe accident for assessment of the containment integrity and managing the severe accident. The severe accident research at Waseda University has been advancing to show that simulations with moving particle semi-implicit (MPS) method (one of the particle methods) can greatly improve the analytical capability and mechanical understanding of the melt behavior in severe accidents. MPS models have been developed and verified regarding calculations of radiation and thermal field, solid-liquid phase transition, buoyancy, and temperature dependency of viscosity to simulate phenomena, such as spreading of corium, ablation of concrete by the corium, crust formation and cooling of the corium by top flooding. Validations have been conducted against experiments such as FARO L26S, ECOKATS-V1, Theofanous, and SPREAD for spreading, SURC-2, SURC-4, SWISS-1, and SWISS-2 for MCCI. These validations cover melt spreading behaviors and MCCI by mixture of molten oxides (including prototypic UO2-ZrO2), metals, and water. Generally, the analytical results show good agreement with the experiment with respect to the leading edge of spreading melt and ablation front history of concrete. The MPS results indicate that crust formation may play important roles in melt spreading and MCCI. There is a need to develop a code for two dimensional MCCI experiment simulation with MPS method as future study, which will be able to simulate anisotropic ablation of concrete.

Willingness to use safety belt and levels of injury in car accidents.

PubMed

de Lapparent, Matthieu

2008-05-01

In this article, we develop a bivariate ordered Probit model to analyze the decision to fasten the safety belt in a car and the resulting severity of accidents if it happens. The approach takes into account the fact that the decision to fasten the safety belt has a direct causal effect on the category of injury if an accident happens. Our application to a sample drawn from the database of French accident reports in 2003 for three populations of car users (drivers, front passengers, rear passengers) shows that fastening the safety belt is significantly related to a decrease in severe injuries but it shows also that these car users compensate partly for this safety benefit. Furthermore, it is observed that demographic characteristics of car users, as well as transport facilities, play important roles in decisions to fasten safety belts and in the eventual resulting accident injuries.

Building of communication system for nuclear accident emergency disposal based on IP multimedia subsystem

NASA Astrophysics Data System (ADS)

Wang, Kang; Gao, Guiqing; Qin, Yuanli; He, Xiangyong

2018-05-01

The nuclear accident emergency disposal must be supported by an efficient, real-time modularization and standardization communication system. Based on the analysis of communication system for nuclear accident emergency disposal which included many functions such as the internal and external communication, multiply access supporting and command center. Some difficult problems of the communication system were discussed such as variety access device type, complex composition, high mobility, set up quickly, multiply business support, and so on. Taking full advantages of the IP Multimedia Subsystem (IMS), a nuclear accident emergency communication system was build based on the IMS. It was studied and implemented that some key unit and module functions of communication system were included the system framework implementation, satellite access, short-wave access, load/vehicle-mounted communication units. The application tests showed that the system could provide effective communication support for the nuclear accident emergency disposal, which was of great practical value.

Hazmat transport: a methodological framework for the risk analysis of marshalling yards.

PubMed

Cozzani, Valerio; Bonvicini, Sarah; Spadoni, Gigliola; Zanelli, Severino

2007-08-17

A methodological framework was outlined for the comprehensive risk assessment of marshalling yards in the context of quantified area risk analysis. Three accident typologies were considered for yards: (i) "in-transit-accident-induced" releases; (ii) "shunting-accident-induced" spills; and (iii) "non-accident-induced" leaks. A specific methodology was developed for the assessment of expected release frequencies and equivalent release diameters, based on the application of HazOp and Fault Tree techniques to reference schemes defined for the more common types of railcar vessels used for "hazmat" transportation. The approach was applied to the assessment of an extended case-study. The results evidenced that "non-accident-induced" leaks in marshalling yards represent an important contribution to the overall risk associated to these zones. Furthermore, the results confirmed the considerable role of these fixed installations to the overall risk associated to "hazmat" transportation.

Trespassing on the tracks: a review of railway pedestrian safety research.

PubMed

Lobb, Brenda

2006-01-01

Train-pedestrian collisions have been shown to be the leading cause of fatality in train-related accidents worldwide, yet there is remarkably little research in this area. In this paper, the major types of railway transportation accident research are briefly highlighted to indicate the general context of research concerning train-pedestrian collisions, which are then reviewed. Themes emerging from the diverse research are identified, the various strategies that have been proposed for prevention of railway pedestrian accidents are discussed, and the empirical evidence for their efficacy examined in the light of the much more extensive literature on road pedestrian accidents. Finally, it is proposed that application of current theory in behavioral and cognitive psychology may usefully inform future research in transportation safety.

The Model 9977 Radioactive Material Packaging Primer

DOE Office of Scientific and Technical Information (OSTI.GOV)

Abramczyk, G.

2015-10-09

The Model 9977 Packaging is a single containment drum style radioactive material (RAM) shipping container designed, tested and analyzed to meet the performance requirements of Title 10 the Code of Federal Regulations Part 71. A radioactive material shipping package, in combination with its contents, must perform three functions (please note that the performance criteria specified in the Code of Federal Regulations have alternate limits for normal operations and after accident conditions): Containment, the package must “contain” the radioactive material within it; Shielding, the packaging must limit its users and the public to radiation doses within specified limits; and Subcriticality, themore » package must maintain its radioactive material as subcritical« less

Marine ecological risk assessment methods for radiation accidents.

PubMed

Ye, Sufen; Zhang, Luoping; Feng, Huan

2017-12-01

Ecological risk assessment (ERA) is a powerful technical tool that can be used to analyze potential and extreme adverse environmental impacts. With the rapid development of nuclear power plants in coastal areas around the world, the establishment of approaches and methodologies for marine ERA with a focus on radiation accidents is an urgent requirement for marine environmental management. In this study, the approaches and methodologies for ERA pertaining to marine radiation accidents (MRA) are discussed and summarized with applications in case studies, such as the nuclear accident in Fukushima, Japan, and a hypothetical accident in Daya Bay, China. The concepts of ERA and Risk Degree of ERA on MRA are defined for the first time to optimize the ERA system. The results of case studies show that the ERA approach and methodology for MRA are scientifically sound and effective in both the early and late stage of MRAs along with classic ERA Approach and the ERICA Integrated Approach. The results can be useful in the decision-making processes and the risk management at the beginning of accident as well as the ecological restoration after the accident. Copyright © 2017 Elsevier Ltd. All rights reserved.

Development and evaluation of a Naïve Bayesian model for coding causation of workers’ compensation claims☆

PubMed Central

Bertke, S. J.; Meyers, A. R.; Wurzelbacher, S. J.; Bell, J.; Lampl, M. L.; Robins, D.

2015-01-01

Introduction Tracking and trending rates of injuries and illnesses classified as musculoskeletal disorders caused by ergonomic risk factors such as overexertion and repetitive motion (MSDs) and slips, trips, or falls (STFs) in different industry sectors is of high interest to many researchers. Unfortunately, identifying the cause of injuries and illnesses in large datasets such as workers’ compensation systems often requires reading and coding the free form accident text narrative for potentially millions of records. Method To alleviate the need for manual coding, this paper describes and evaluates a computer auto-coding algorithm that demonstrated the ability to code millions of claims quickly and accurately by learning from a set of previously manually coded claims. Conclusions The auto-coding program was able to code claims as a musculoskeletal disorders, STF or other with approximately 90% accuracy. Impact on industry The program developed and discussed in this paper provides an accurate and efficient method for identifying the causation of workers’ compensation claims as a STF or MSD in a large database based on the unstructured text narrative and resulting injury diagnoses. The program coded thousands of claims in minutes. The method described in this paper can be used by researchers and practitioners to relieve the manual burden of reading and identifying the causation of claims as a STF or MSD. Furthermore, the method can be easily generalized to code/classify other unstructured text narratives. PMID:23206504

Enhancing Public Helicopter Safety as a Component of Homeland Security

DTIC Science & Technology

2016-12-01

Risk Assessment Tool GPS Global Positioning System IFR instrument flight rules ILS instrument landing system IMC instrument meteorological...flight rules ( IFR ) flying and the lack of a pre-flight risk assessment. Pilot fatigue is a factor that appeared in two of the accident reports (New...three common factors that emerged from the qualitative analysis of coding: inadequate proficiency of IFR flying, lack of a pre- flight risk assessment

Solid Rocket Launch Vehicle Explosion Environments

NASA Technical Reports Server (NTRS)

Richardson, E. H.; Blackwood, J. M.; Hays, M. J.; Skinner, T.

2014-01-01

Empirical explosion data from full scale solid rocket launch vehicle accidents and tests were collected from all available literature from the 1950s to the present. In general data included peak blast overpressure, blast impulse, fragment size, fragment speed, and fragment dispersion. Most propellants were 1.1 explosives but a few were 1.3. Oftentimes the data from a single accident was disjointed and/or missing key aspects. Despite this fact, once the data as a whole was digitized, categorized, and plotted clear trends appeared. Particular emphasis was placed on tests or accidents that would be applicable to scenarios from which a crew might need to escape. Therefore, such tests where a large quantity of high explosive was used to initiate the solid rocket explosion were differentiated. Also, high speed ground impacts or tests used to simulate such were also culled. It was found that the explosions from all accidents and applicable tests could be described using only the pressurized gas energy stored in the chamber at the time of failure. Additionally, fragmentation trends were produced. Only one accident mentioned the elusive "small" propellant fragments, but upon further analysis it was found that these were most likely produced as secondary fragments when larger primary fragments impacted the ground. Finally, a brief discussion of how this data is used in a new launch vehicle explosion model for improving crew/payload survival is presented.

SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

1978-02-01

The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocitymore » and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.« less

MELCOR Peer Review

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boyack, B.E.; Dhir, V.K.; Gieseke, J.A.

1992-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. The newest version of MELCOR is Version 1.8.1, July 1991. MELCOR development has reached the point that the United States Nuclear Regulatory Commission sponsored a broad technical review by recognized experts to determine or confirm the technical adequacy of the code for the serious and complex analyses it is expected to perform. For this purpose, an eight-member MELCOR Peer Review Committee was organized. The Committee has completed its review of the MELCOR code: the review process and findingsmore » of the MELCOR Peer Review Committee are documented in this report. The Committee has determined that recommendations in five areas are appropriate: (1) MELCOR numerics, (2) models missing from MELCOR Version 1.8.1, (3) existing MELCOR models needing revision, (4) the need for expanded MELCOR assessment, and (5) documentation.« less

Why is Probabilistic Seismic Hazard Analysis (PSHA) still used?

NASA Astrophysics Data System (ADS)

Mulargia, Francesco; Stark, Philip B.; Geller, Robert J.

2017-03-01

Even though it has never been validated by objective testing, Probabilistic Seismic Hazard Analysis (PSHA) has been widely used for almost 50 years by governments and industry in applications with lives and property hanging in the balance, such as deciding safety criteria for nuclear power plants, making official national hazard maps, developing building code requirements, and determining earthquake insurance rates. PSHA rests on assumptions now known to conflict with earthquake physics; many damaging earthquakes, including the 1988 Spitak, Armenia, event and the 2011 Tohoku, Japan, event, have occurred in regions relatively rated low-risk by PSHA hazard maps. No extant method, including PSHA, produces reliable estimates of seismic hazard. Earthquake hazard mitigation should be recognized to be inherently political, involving a tradeoff between uncertain costs and uncertain risks. Earthquake scientists, engineers, and risk managers can make important contributions to the hard problem of allocating limited resources wisely, but government officials and stakeholders must take responsibility for the risks of accidents due to natural events that exceed the adopted safety criteria.

Training Presentation for NASA Civil Helicopter Safety Website

NASA Technical Reports Server (NTRS)

Iseler, Laura

2002-01-01

NASA civil helicopter safety News & Updates include the following: Mar. 2002. The Air Medical Operations Survey has been completed! Check it out! Also accessible via the Mission pages under Air Medical Mission. Air Medical and Law Enforcement Mission pages have been added. They are accessible via the Mission pages. The Public Use, Personal, Offshore, Law Enforcement, External Load, Business and Gyro accident pages (accessable via the Mission page) have been updated. Feb. 2002. A Words of Wisdom section has been added. You can access it by clicking the Library button. A link to a Corporate Accident Response Plan has been added to the Accident page. The AMs, Aerial Application and Instruction accident pages (accessable via the Mission page) have been updated. Jan. 2002. A new searchable safety article database has been added. You can access it by clicking the Library button. The 2001 accident summaries have been updated and the statistics have been compiled - check it out by clicking the accident tab to the left. Dec. 2001. Please read the FAA Administrator's memo regarding the latest FBI warning. 3ee the FAA column - Fall 2001 Read it now!

Preliminary topical report on comparison reactor disassembly calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

McLaughlin, T.P.

1975-11-01

Preliminary results of comparison disassembly calculations for a representative LMFBR model (2100-l voided core) and arbitrary accident conditions are described. The analytical methods employed were the computer programs: FX2- POOL, PAD, and VENUS-II. The calculated fission energy depositions are in good agreement, as are measures of the destructive potential of the excursions, kinetic energy, and work. However, in some cases the resulting fuel temperatures are substantially divergent. Differences in the fission energy deposition appear to be attributable to residual inconsistencies in specifying the comparison cases. In contrast, temperature discrepancies probably stem from basic differences in the energy partition models inherentmore » in the codes. Although explanations of the discrepancies are being pursued, the preliminary results indicate that all three computational methods provide a consistent, global characterization of the contrived disassembly accident. (auth)« less

Decay Heat Removal from a GFR Core by Natural Convection

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williams, Wesley C.; Hejzlar, Pavel; Driscoll, Michael J.

2004-07-01

One of the primary challenges for Gas-cooled Fast Reactors (GFR) is decay heat removal after a loss of coolant accident (LOCA). Due to the fact that thermal gas cooled reactors currently under design rely on passive mechanisms to dissipate decay heat, there is a strong motivation to accomplish GFR core cooling through natural phenomena. This work investigates the potential of post-LOCA decay heat removal from a GFR core to a heat sink using an external convection loop. A model was developed in the form of the LOCA-COLA (Loss of Coolant Accident - Convection Loop Analysis) computer code as a meansmore » for 1D steady state convective heat transfer loop analysis. The results show that decay heat removal by means of gas cooled natural circulation is feasible under elevated post-LOCA containment pressure conditions. (authors)« less

Analysis of the SL-1 Accident Using RELAPS5-3D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Francisco, A.D. and Tomlinson, E. T.

2007-11-08

On January 3, 1961, at the National Reactor Testing Station, in Idaho Falls, Idaho, the Stationary Low Power Reactor No. 1 (SL-1) experienced a major nuclear excursion, killing three people, and destroying the reactor core. The SL-1 reactor, a 3 MW{sub t} boiling water reactor, was shut down and undergoing routine maintenance work at the time. This paper presents an analysis of the SL-1 reactor excursion using the RELAP5-3D thermal-hydraulic and nuclear analysis code, with the intent of simulating the accident from the point of reactivity insertion to destruction and vaporization of the fuel. Results are presented, along with amore » discussion of sensitivity to some reactor and transient parameters (many of the details are only known with a high level of uncertainty).« less

Ecological Associations of Alcohol Outlets with Underage and Young Adult Injuries

PubMed Central

Gruenewald, Paul J.; Freisthler, Bridget; Remer, Lillian; LaScala, Elizabeth A.; Treno, Andrew J.; Ponicki, William R.

2010-01-01

Objective This paper argues that associations between rates of three specific problems related to alcohol (i.e., accidents, traffic crashes, and assaults) should be differentially related to densities of off-premise outlets among underage youth and young adults based upon age related-patterns of alcohol outlet use. Methods Zip code-level population models assessed local and distal effects of alcohol outlets upon rates of hospital discharges for these outcomes. Results Densities of off-premise alcohol outlets were significantly related to injuries from accidents, assaults, and traffic crashes for both underage youth and young adults. Densities of bars were associated with more assaults and densities of restaurants were associated with more traffic crash injuries for young adults. Conclusions The distribution of alcohol-related injuries relative to alcohol outlets reflect patterns of alcohol outlet use. PMID:20028361

Radiation dose distributions due to sudden ejection of cobalt device.

PubMed

Abdelhady, Amr

2016-09-01

The evaluation of the radiation dose during accident in a nuclear reactor is of great concern from the viewpoint of safety. One of important accident must be analyzed and may be occurred in open pool type reactor is the rejection of cobalt device. The study is evaluating the dose rate levels resulting from upset withdrawal of co device especially the radiation dose received by the operator in the control room. Study of indirect radiation exposure to the environment due to skyshine effect is also taken into consideration in order to evaluate the radiation dose levels around the reactor during the ejection trip. Microshield, SHLDUTIL, and MCSky codes were used in this study to calculate the radiation dose profiles during cobalt device ejection trip inside and outside the reactor building. Copyright © 2016 Elsevier Ltd. All rights reserved.

Considerations regarding the implementation of EPR dosimetry for the population in the vicinity of Semipalatinsk nuclear test site based on experience from other radiation accidents.

PubMed

Skvortsov, Valeriy; Ivannikov, Alexander; Tikunov, Dimitri; Stepanenko, Valeriy; Borysheva, Natalie; Orlenko, Sergey; Nalapko, Mikhail; Hoshi, Masaharu

2006-02-01

General aspects of applying the method of retrospective dose estimation by electron paramagnetic resonance spectroscopy of human tooth enamel (EPR dosimetry) to the population residing in the vicinity of the Semipalatinsk nuclear test site are analyzed and summarized. The analysis is based on the results obtained during 20 years of investigations conducted in the Medical Radiological Research Center regarding the development and practical application of this method for wide-scale dosimetrical investigation of populations exposed to radiation after the Chernobyl accident and other radiation accidents.

Application of Gaussian Process Modeling to Analysis of Functional Unreliability

DOE Office of Scientific and Technical Information (OSTI.GOV)

R. Youngblood

2014-06-01

This paper applies Gaussian Process (GP) modeling to analysis of the functional unreliability of a “passive system.” GPs have been used widely in many ways [1]. The present application uses a GP for emulation of a system simulation code. Such an emulator can be applied in several distinct ways, discussed below. All applications illustrated in this paper have precedents in the literature; the present paper is an application of GP technology to a problem that was originally analyzed [2] using neural networks (NN), and later [3, 4] by a method called “Alternating Conditional Expectations” (ACE). This exercise enables a multifacetedmore » comparison of both the processes and the results. Given knowledge of the range of possible values of key system variables, one could, in principle, quantify functional unreliability by sampling from their joint probability distribution, and performing a system simulation for each sample to determine whether the function succeeded for that particular setting of the variables. Using previously available system simulation codes, such an approach is generally impractical for a plant-scale problem. It has long been recognized, however, that a well-trained code emulator or surrogate could be used in a sampling process to quantify certain performance metrics, even for plant-scale problems. “Response surfaces” were used for this many years ago. But response surfaces are at their best for smoothly varying functions; in regions of parameter space where key system performance metrics may behave in complex ways, or even exhibit discontinuities, response surfaces are not the best available tool. This consideration was one of several that drove the work in [2]. In the present paper, (1) the original quantification of functional unreliability using NN [2], and later ACE [3], is reprised using GP; (2) additional information provided by the GP about uncertainty in the limit surface, generally unavailable in other representations, is discussed; (3) a simple forensic exercise is performed, analogous to the inverse problem of code calibration, but with an accident management spin: given an observation about containment pressure, what can we say about the system variables? References 1. For an introduction to GPs, see (for example) Gaussian Processes for Machine Learning, C. E. Rasmussen and C. K. I. Williams (MIT, 2006). 2. Reliability Quantification of Advanced Reactor Passive Safety Systems, J. J. Vandenkieboom, PhD Thesis (University of Michigan, 1996). 3. Z. Cui, J. C. Lee, J. J. Vandenkieboom, and R. W. Youngblood, “Unreliability Quantification of a Containment Cooling System through ACE and ANN Algorithms,” Trans. Am. Nucl. Soc. 85, 178 (2001). 4. Risk and Safety Analysis of Nuclear Systems, J. C. Lee and N. J. McCormick (Wiley, 2011). See especially §11.2.4.« less

High burden of hospital resource utilization in patients with hidradenitis suppurativa in England: a retrospective cohort study using hospital episode statistics.

PubMed

Desai, N; Shah, P

2017-04-01

The health resource utilization associated with managing patients with hidradenitis suppurativa (HS) in England is unknown. To describe the characteristics of patients with HS and hospital resource use associated with management of HS in England. A retrospective cohort study using Hospital Episode Statistics data. Patients with a primary diagnostic code for HS (ICD-10 code L73·2) during an inpatient admission (n = 11 359) between 1 April 2007 and 31 December 2013 were identified; patients with code L73·2 attending only as outpatients were excluded. Data for all inpatient, outpatient and accident and emergency admissions during the study period were extracted. Of the 11 359 patients, 10 832 had a first recorded inpatient HS diagnostic code (index spell) during the study period (female 7569, 69·9%). The mean age at the index spell was 39 ± 13·1 years in men and 36 ± 11·7 years in women. There were 65 544 inpatient spells during the study period; 7202 (63·4%) patients underwent nonelective spells, 4128 (36.3%) elective spells and 9790 (86·2%) day-case attendances. There were 43 773 accident and emergency attendances during the study period in 8716 (76·7%) patients. There were 303 204 outpatient appointments in 11 203 patients (mean 27·1 per patient); 4827 (42·5%) of the study population attended dermatology, 8087 (71·2%) general surgery and 4111 (36·2%) plastic surgery. Based on the mean number of spells per patient per year, the mean hospital resource utilization cost for a patient with HS was £2027 per patient per year. HS is associated with a large burden of hospital attendances for young patients of working age and high National Health Service resource costs. © 2016 British Association of Dermatologists.

Light Water Reactor Sustainability Program Reactor Safety Technologies Pathway Technical Program Plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Corradini, M. L.; Peko, D.; Farmer, M.

In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safetymore » initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.« less

Light Water Reactor Sustainability Program: Reactor Safety Technologies Pathway Technical Program Plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Corradini, M. L.

In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safetymore » initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.« less

Influencing factors for early acute cerebrovascular accidents in patients with stroke history following off-pump coronary artery bypass grafting.

PubMed

Wang, Bin; Jia, Ming; Jia, Shijie; Wan, Jiuhe; Zhou, Xiao; Luo, Zhimin; Zhou, Ye; Zhang, Jianqun

2014-06-01

To analyse risk factors for early acute cerebrovascular accidents following off-pump coronary artery bypass grafting (OPCAB) in patients with stroke history, and to propose preventive measures to reduce the incidence of these events. A total of 468 patients with a history of stroke underwent OPCAB surgery in Beijing Anzhen Hospital of China from January 2010 to September 2012. They were retrospectively divided into two groups according to the occurrence of early acute cerebrovascular accidents within 48 hours following OPCAB. Multivariate logistic regression analysis was used to find risk or protective factors for early acute cerebrovascular accidents following the OPCAB. Fifty-two patients (11.1%) suffered from early acute cerebrovascular accidents in 468 patients, including 39 cases of cerebral infarction, two cases of cerebral haemorrhage, 11 cases of transient ischaemic attack (TIA). There were significant differences between the two groups in preoperative left ventricular ejection fraction ≤ 35%, severe bilateral carotid artery stenosis, poorly controlled hypertension, intraoperative application of Enclose® II proximal anastomotic device, postoperative acute myocardial infarction, atrial fibrillation, hypotension, ventilation time > 48h, ICU duration >48h and mortality. Multivariate logistic regression analysis showed that preoperative severe bilateral carotid stenosis (OR=6.378, 95%CI: 2.278-20.987) and preoperative left ventricular ejection fraction ≤ 35% (OR=2.737, 95%CI: 1.267-6.389), postoperative acute myocardial infarction (OR=3.644, 95%CI: 1.928-6.876), postoperative atrial fibrillation (OR=3.104, 95%CI:1.135∼8.016) and postoperative hypotension (OR=4.173, 95%CI: 1.836∼9.701) were independent risk factors for early acute cerebrovascular accidents in patients with a history of stroke following OPCAB procedures, while intraoperative application of Enclose® II proximal anastomotic device was protective factor (OR=0.556, 95%CI: 0.337-0.925). This study indicated that patients with severe bilateral carotid stenosis, the left ventricular ejection fraction ≤35%, the postoperative acute myocardial infarction, postoperative atrial fibrillation and postoperative hypotension were more likely to suffer from early acute cerebrovascular accidents when they received OPCAB. Application of Enclose® II proximal anastomotic device may decrease the incidence of early acute cerebrovascular accidents during OPCAB. Copyright © 2014 Australian and New Zealand Society of Cardiac and Thoracic Surgeons (ANZSCTS) and the Cardiac Society of Australia and New Zealand (CSANZ). Published by Elsevier B.V. All rights reserved.

Best Practices for Optimizing DoD Contractor Safety and Occupational Health Program Performance

DTIC Science & Technology

2012-12-01

such as Accident Prevention Plan (APP), Activity Hazard Analysis (AHA), Quality Assurance Surveillance Plans (QASP), etc. Contract administration...technology support, medical , and maintenance of equipment and facilities. The DoD Guidebook for the Acquisition of Services, provides acquisition...OSHA regulations and perform in accordance with an applicable accident prevention program that complies with State and Federal requirements. The

Bayesian networks for maritime traffic accident prevention: benefits and challenges.

PubMed

Hänninen, Maria

2014-12-01

Bayesian networks are quantitative modeling tools whose applications to the maritime traffic safety context are becoming more popular. This paper discusses the utilization of Bayesian networks in maritime safety modeling. Based on literature and the author's own experiences, the paper studies what Bayesian networks can offer to maritime accident prevention and safety modeling and discusses a few challenges in their application to this context. It is argued that the capability of representing rather complex, not necessarily causal but uncertain relationships makes Bayesian networks an attractive modeling tool for the maritime safety and accidents. Furthermore, as the maritime accident and safety data is still rather scarce and has some quality problems, the possibility to combine data with expert knowledge and the easy way of updating the model after acquiring more evidence further enhance their feasibility. However, eliciting the probabilities from the maritime experts might be challenging and the model validation can be tricky. It is concluded that with the utilization of several data sources, Bayesian updating, dynamic modeling, and hidden nodes for latent variables, Bayesian networks are rather well-suited tools for the maritime safety management and decision-making. Copyright © 2014 Elsevier Ltd. All rights reserved.

The development of an inherent safety approach to the prevention of domino accidents.

PubMed

Cozzani, Valerio; Tugnoli, Alessandro; Salzano, Ernesto

2009-11-01

The severity of industrial accidents in which a domino effect takes place is well known in the chemical and process industry. The application of an inherent safety approach for the prevention of escalation events leading to domino accidents was explored in the present study. Reference primary scenarios were analyzed and escalation vectors were defined. Inherent safety distances were defined and proposed as a metric to express the intensity of the escalation vectors. Simple rules of thumb were presented for a preliminary screening of these distances. Swift reference indices for layout screening with respect to escalation hazard were also defined. Two case studies derived from existing layouts of oil refineries were selected to understand the potentialities coming from the application in the methodology. The results evidenced that the approach allows a first comparative assessment of the actual domino hazard in a layout, and the identification of critical primary units with respect to escalation events. The methodology developed also represents a useful screening tool to identify were to dedicate major efforts in the design of add-on measures, optimizing conventional passive and active measures for the prevention of severe domino accidents.

SINGLE PHASE ANALYTICAL MODELS FOR TERRY TURBINE NOZZLE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhao, Haihua; Zhang, Hongbin; Zou, Ling

All BWR RCIC (Reactor Core Isolation Cooling) systems and PWR AFW (Auxiliary Feed Water) systems use Terry turbine, which is composed of the wheel with turbine buckets and several groups of fixed nozzles and reversing chambers inside the turbine casing. The inlet steam is accelerated through the turbine nozzle and impacts on the wheel buckets, generating work to drive the RCIC pump. As part of the efforts to understand the unexpected “self-regulating” mode of the RCIC systems in Fukushima accidents and extend BWR RCIC and PWR AFW operational range and flexibility, mechanistic models for the Terry turbine, based on Sandiamore » National Laboratories’ original work, has been developed and implemented in the RELAP-7 code to simulate the RCIC system. RELAP-7 is a new reactor system code currently under development with the funding support from U.S. Department of Energy. The RELAP-7 code is a fully implicit code and the preconditioned Jacobian-free Newton-Krylov (JFNK) method is used to solve the discretized nonlinear system. This paper presents a set of analytical models for simulating the flow through the Terry turbine nozzles when inlet fluid is pure steam. The implementation of the models into RELAP-7 will be briefly discussed. In the Sandia model, the turbine bucket inlet velocity is provided according to a reduced-order model, which was obtained from a large number of CFD simulations. In this work, we propose an alternative method, using an under-expanded jet model to obtain the velocity and thermodynamic conditions for the turbine bucket inlet. The models include both adiabatic expansion process inside the nozzle and free expansion process out of the nozzle to reach the ambient pressure. The combined models are able to predict the steam mass flow rate and supersonic velocity to the Terry turbine bucket entrance, which are the necessary input conditions for the Terry Turbine rotor model. The nozzle analytical models were validated with experimental data and benchmarked with CFD simulations. The analytical models generally agree well with the experimental data and CFD simulations. The analytical models are suitable for implementation into a reactor system analysis code or severe accident code as part of mechanistic and dynamical models to understand the RCIC behaviors. The cases with two-phase flow at the turbine inlet will be pursued in future work.« less

Alternative method of highway traffic safety analysis for developing countries using delphi technique and Bayesian network.

PubMed

Mbakwe, Anthony C; Saka, Anthony A; Choi, Keechoo; Lee, Young-Jae

2016-08-01

Highway traffic accidents all over the world result in more than 1.3 million fatalities annually. An alarming number of these fatalities occurs in developing countries. There are many risk factors that are associated with frequent accidents, heavy loss of lives, and property damage in developing countries. Unfortunately, poor record keeping practices are very difficult obstacle to overcome in striving to obtain a near accurate casualty and safety data. In light of the fact that there are numerous accident causes, any attempts to curb the escalating death and injury rates in developing countries must include the identification of the primary accident causes. This paper, therefore, seeks to show that the Delphi Technique is a suitable alternative method that can be exploited in generating highway traffic accident data through which the major accident causes can be identified. In order to authenticate the technique used, Korea, a country that underwent similar problems when it was in its early stages of development in addition to the availability of excellent highway safety records in its database, is chosen and utilized for this purpose. Validation of the methodology confirms the technique is suitable for application in developing countries. Furthermore, the Delphi Technique, in combination with the Bayesian Network Model, is utilized in modeling highway traffic accidents and forecasting accident rates in the countries of research. Copyright © 2016 Elsevier Ltd. All rights reserved.

NASA Standard for Models and Simulations (M and S): Development Process and Rationale

NASA Technical Reports Server (NTRS)

Zang, Thomas A.; Blattnig, Steve R.; Green, Lawrence L.; Hemsch, Michael J.; Luckring, James M.; Morison, Joseph H.; Tripathi, Ram K.

2009-01-01

After the Columbia Accident Investigation Board (CAIB) report. the NASA Administrator at that time chartered an executive team (known as the Diaz Team) to identify the CAIB report elements with Agency-wide applicability, and to develop corrective measures to address each element. This report documents the chronological development and release of an Agency-wide Standard for Models and Simulations (M&S) (NASA Standard 7009) in response to Action #4 from the report, "A Renewed Commitment to Excellence: An Assessment of the NASA Agency-wide Applicability of the Columbia Accident Investigation Board Report, January 30, 2004".

Preliminary Modeling of Accident Tolerant Fuel Concepts under Accident Conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gamble, Kyle A.; Hales, Jason D.

2016-12-01

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. Thus, the United States Department of Energy through its NEAMS (Nuclear Energy Advanced Modeling and Simulation) program has funded an Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The ATF HIP is funded for a three-year period. The purpose of the HIP is to perform research into two potential accident tolerant concepts and provide an in-depth report to the Advanced Fuels Campaign (AFC) describing the behavior of themore » concepts, both of which are being considered for inclusion in a lead test assembly scheduled for placement into a commercial reactor in 2022. The initial focus of the HIP is on uranium silicide fuel and iron-chromium-aluminum (FeCrAl) alloy cladding. Utilizing the expertise of three national laboratory participants (INL, LANL, and ANL) a comprehensive mulitscale approach to modeling is being used including atomistic modeling, molecular dynamics, rate theory, phase-field, and fuel performance simulations. In this paper, we present simulations of two proposed accident tolerant fuel systems: U3Si2 fuel with Zircaloy-4 cladding, and UO2 fuel with FeCrAl cladding. The simulations investigate the fuel performance response of the proposed ATF systems under Loss of Coolant and Station Blackout conditions using the BISON code. 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). Early results indicate that each concept has significant advantages as well as areas of concern. Further work is required prior to formulating the proposition report for the Advanced Fuels Campaign.« less

Motor vehicle crashes in roadway construction workzones: an analysis using narrative text from insurance claims.

PubMed

Sorock, G S; Ranney, T A; Lehto, M R

1996-01-01

Motor vehicle travel through roadway construction workzones has been shown to increase the risk of a crash. The number of workzones has increased due to recent congressional funding in 1991 for expanded roadway maintenance and repair. In this paper, we describe the characteristics and costs of motor vehicle crashes in roadway construction workzones. As opposed to using standard accident codes to identify accident types, automobile insurance claims files from 1990-93 were searched to identify records with the keyword "construction" in the accident narrative field. A total of 3,686 claims were used for the analysis of crashes. Keywords from the accident narrative field were used to identify five pre-crash vehicle activities and five crash types. We evaluated misclassification error by reading 560 randomly selected claims and found it to be only 5%. For each of four years, 1990-93, there was a total of 648,996,977 and 1,065 crashes, respectively. There was a 70% increase in the crash rate per 10,000 personal insured vehicles from 1990-93 (2.1-3.6). Most crashes (26%) involved a stopped or slowing vehicle in the workzone. The most common crash (31%) was a rear-end collision. The most costly pre-crash activity was a major judgment error on the part of a driver (n = 120, median cost = $2,628). An overturned vehicle was the most costly crash type (n = 16, median cost = $4,745). In summary, keyword text analysis of accident narrative data used in this study demonstrated its utility and potential for enhancing injury epidemiology. The results suggest interventions are needed to respond to growing traffic hazards in construction workzones.

Criticality Calculations with MCNP6 - Practical Lectures

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Forrest B.; Rising, Michael Evan; Alwin, Jennifer Louise

2016-11-29

These slides are used to teach MCNP (Monte Carlo N-Particle) usage to nuclear criticality safety analysts. The following are the lecture topics: course information, introduction, MCNP basics, criticality calculations, advanced geometry, tallies, adjoint-weighted tallies and sensitivities, physics and nuclear data, parameter studies, NCS validation I, NCS validation II, NCS validation III, case study 1 - solution tanks, case study 2 - fuel vault, case study 3 - B&W core, case study 4 - simple TRIGA, case study 5 - fissile mat. vault, criticality accident alarm systems. After completion of this course, you should be able to: Develop an input modelmore » for MCNP; Describe how cross section data impact Monte Carlo and deterministic codes; Describe the importance of validation of computer codes and how it is accomplished; Describe the methodology supporting Monte Carlo codes and deterministic codes; Describe pitfalls of Monte Carlo calculations; Discuss the strengths and weaknesses of Monte Carlo and Discrete Ordinants codes; The diffusion theory model is not strictly valid for treating fissile systems in which neutron absorption, voids, and/or material boundaries are present. In the context of these limitations, identify a fissile system for which a diffusion theory solution would be adequate.« less

The Self-Powered Detector Simulation `MATiSSe' Toolbox applied to SPNDs for severe accident monitoring in PWRs

NASA Astrophysics Data System (ADS)

Barbot, Loïc; Villard, Jean-François; Fourrez, Stéphane; Pichon, Laurent; Makil, Hamid

2018-01-01

In the framework of the French National Research Agency program on nuclear safety and radioprotection, the `DIstributed Sensing for COrium Monitoring and Safety' project aims at developing innovative instrumentation for corium monitoring in case of severe accident in a Pressurized Water nuclear Reactor. Among others, a new under-vessel instrumentation based on Self-Powered Neutron Detectors is developed using a numerical simulation toolbox, named `MATiSSe'. The CEA Instrumentation Sensors and Dosimetry Lab developed MATiSSe since 2010 for Self-Powered Neutron Detectors material selection and geometry design, as well as for their respective partial neutron and gamma sensitivity calculations. MATiSSe is based on a comprehensive model of neutron and gamma interactions which take place in Selfpowered neutron detector components using the MCNP6 Monte Carlo code. As member of the project consortium, the THERMOCOAX SAS Company is currently manufacturing some instrumented pole prototypes to be tested in 2017. The full severe accident monitoring equipment, including the standalone low current acquisition system, will be tested during a joined CEA-THERMOCOAX experimental campaign in some realistic irradiation conditions, in the Slovenian TRIGA Mark II research reactor.

Analysis of Accidents at the Pakistan Research Reactor-1 Using Proposed Mixed-Fuel (HEU and LEU) Core

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bokhari, Ishtiaq H.

2004-12-15

The Pakistan Research Reactor-1 (PARR-1) was converted from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel in 1991. The reactor is running successfully, with an upgraded power level of 10 MW. To save money on the purchase of costly fresh LEU fuel elements, the use of less burnt HEU spent fuel elements along with the present LEU fuel elements is being considered. The proposal calls for the HEU fuel elements to be placed near the thermal column to gain the required excess reactivity. In the present study the safety analysis of a proposed mixed-fuel core has been carried outmore » at a calculated steady-state power level of 9.8 MW. Standard computer codes and correlations were employed to compute various parameters. Initiating events in reactivity-induced accidents involve various modes of reactivity insertion, namely, start-up accident, accidental drop of a fuel element on the core, flooding of a beam tube with water, and removal of an in-pile experiment during reactor operation. For each of these transients, time histories of reactor power, energy released, temperature, and reactivity were determined.« less

Ocean Thermal Energy Conversion power system development. Phase I: preliminary design. Final report. [ODSP-3 code; OTEC Steady-State Analysis Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1978-12-04

The following appendices are included; Dynamic Simulation Program (ODSP-3); sample results of dynamic simulation; trip report - NH/sub 3/ safety precautions/accident records; trip report - US Coast Guard Headquarters; OTEC power system development, preliminary design test program report; medium turbine generator inspection point program; net energy analysis; bus bar cost of electricity; OTEC technical specifications; and engineer drawings. (WHK)

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

PubMed

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.

An approach to accidents modeling based on compounds road environments.

PubMed

Fernandes, Ana; Neves, Jose

2013-04-01

The most common approach to study the influence of certain road features on accidents has been the consideration of uniform road segments characterized by a unique feature. However, when an accident is related to the road infrastructure, its cause is usually not a single characteristic but rather a complex combination of several characteristics. The main objective of this paper is to describe a methodology developed in order to consider the road as a complete environment by using compound road environments, overcoming the limitations inherented in considering only uniform road segments. The methodology consists of: dividing a sample of roads into segments; grouping them into quite homogeneous road environments using cluster analysis; and identifying the influence of skid resistance and texture depth on road accidents in each environment by using generalized linear models. The application of this methodology is demonstrated for eight roads. Based on real data from accidents and road characteristics, three compound road environments were established where the pavement surface properties significantly influence the occurrence of accidents. Results have showed clearly that road environments where braking maneuvers are more common or those with small radii of curvature and high speeds require higher skid resistance and texture depth as an important contribution to the accident prevention. Copyright © 2013 Elsevier Ltd. All rights reserved.

UTSG-2; A theoretical model describing the transient behavior of a pressurized water reactor natural circulation U-tube steam generator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hold, A.

An advanced nonlinear transient model for calculating steady-state and dynamic behaviors of characteristic parameters of a Kraftwerk Union-type vertical natural-circulation U-tube steam generator and its main steam system is presented. This model has been expanded due to the increasing need for safety-related accident research studies. It now takes into consideration the possibilities of dryout and superheating along the secondary side of the steam generator. The resulting theoretical model is the basis of the digital code UTSG-2, which can be used both by itself and in combination with other pressurized water reactor transient codes, such as ALMOD-3.4, AMOD-4, and ATHLET.

Development of posture-specific computational phantoms using motion capture technology and application to radiation dose-reconstruction for the 1999 Tokai-Mura nuclear criticality accident

NASA Astrophysics Data System (ADS)

Vazquez, Justin A.; Caracappa, Peter F.; Xu, X. George

2014-09-01

The majority of existing computational phantoms are designed to represent workers in typical standing anatomical postures with fixed arm and leg positions. However, workers found in accident-related scenarios often assume varied postures. This paper describes the development and application of two phantoms with adjusted postures specified by data acquired from a motion capture system to simulate unique human postures found in a 1999 criticality accident that took place at a JCO facility in Tokai-Mura, Japan. In the course of this accident, two workers were fatally exposed to extremely high levels of radiation. Implementation of the emergent techniques discussed produced more accurate and more detailed dose estimates for the two workers than were reported in previous studies. A total-body dose of 6.43 and 26.38 Gy was estimated for the two workers, who assumed a crouching and a standing posture, respectively. Additionally, organ-specific dose estimates were determined, including a 7.93 Gy dose to the thyroid and 6.11 Gy dose to the stomach for the crouching worker and a 41.71 Gy dose to the liver and a 37.26 Gy dose to the stomach for the standing worker. Implications for the medical prognosis of the workers are discussed, and the results of this study were found to correlate better with the patient outcome than previous estimates, suggesting potential future applications of such methods for improved epidemiological studies involving next-generation computational phantom tools.

Development of posture-specific computational phantoms using motion capture technology and application to radiation dose-reconstruction for the 1999 Tokai-Mura nuclear criticality accident.

PubMed

Vazquez, Justin A; Caracappa, Peter F; Xu, X George

2014-09-21

The majority of existing computational phantoms are designed to represent workers in typical standing anatomical postures with fixed arm and leg positions. However, workers found in accident-related scenarios often assume varied postures. This paper describes the development and application of two phantoms with adjusted postures specified by data acquired from a motion capture system to simulate unique human postures found in a 1999 criticality accident that took place at a JCO facility in Tokai-Mura, Japan. In the course of this accident, two workers were fatally exposed to extremely high levels of radiation. Implementation of the emergent techniques discussed produced more accurate and more detailed dose estimates for the two workers than were reported in previous studies. A total-body dose of 6.43 and 26.38 Gy was estimated for the two workers, who assumed a crouching and a standing posture, respectively. Additionally, organ-specific dose estimates were determined, including a 7.93 Gy dose to the thyroid and 6.11 Gy dose to the stomach for the crouching worker and a 41.71 Gy dose to the liver and a 37.26 Gy dose to the stomach for the standing worker. Implications for the medical prognosis of the workers are discussed, and the results of this study were found to correlate better with the patient outcome than previous estimates, suggesting potential future applications of such methods for improved epidemiological studies involving next-generation computational phantom tools.

An In vitro evaluation of the reliability of QR code denture labeling technique.

PubMed

Poovannan, Sindhu; Jain, Ashish R; Krishnan, Cakku Jalliah Venkata; Chandran, Chitraa R

2016-01-01

Positive identification of the dead after accidents and disasters through labeled dentures plays a key role in forensic scenario. A number of denture labeling methods are available, and studies evaluating their reliability under drastic conditions are vital. This study was conducted to evaluate the reliability of QR (Quick Response) Code labeled at various depths in heat-cured acrylic blocks after acid treatment, heat treatment (burns), and fracture in forensics. It was an in vitro study. This study included 160 specimens of heat-cured acrylic blocks (1.8 cm × 1.8 cm) and these were divided into 4 groups (40 samples per group). QR Codes were incorporated in the samples using clear acrylic sheet and they were assessed for reliability under various depths, acid, heat, and fracture. Data were analyzed using Chi-square test, test of proportion. The QR Code inclusion technique was reliable under various depths of acrylic sheet, acid (sulfuric acid 99%, hydrochloric acid 40%) and heat (up to 370°C). Results were variable with fracture of QR Code labeled acrylic blocks. Within the limitations of the study, by analyzing the results, it was clearly indicated that the QR Code technique was reliable under various depths of acrylic sheet, acid, and heat (370°C). Effectiveness varied in fracture and depended on the level of distortion. This study thus suggests that QR Code is an effective and simpler denture labeling method.

CTF Theory Manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Avramova, Maria N.; Salko, Robert K.

Coolant-Boiling in Rod Arrays|Two Fluids (COBRA-TF) is a thermal/ hydraulic (T/H) simulation code designed for light water reactor (LWR) vessel analysis. It uses a two-fluid, three-field (i.e. fluid film, fluid drops, and vapor) modeling approach. Both sub-channel and 3D Cartesian forms of 9 conservation equations are available for LWR modeling. The code was originally developed by Pacific Northwest Laboratory in 1980 and had been used and modified by several institutions over the last few decades. COBRA-TF also found use at the Pennsylvania State University (PSU) by the Reactor Dynamics and Fuel Management Group (RDFMG) and has been improved, updated, andmore » subsequently re-branded as CTF. As part of the improvement process, it was necessary to generate sufficient documentation for the open-source code which had lacked such material upon being adopted by RDFMG. This document serves mainly as a theory manual for CTF, detailing the many two-phase heat transfer, drag, and important accident scenario models contained in the code as well as the numerical solution process utilized. Coding of the models is also discussed, all with consideration for updates that have been made when transitioning from COBRA-TF to CTF. Further documentation outside of this manual is also available at RDFMG which focus on code input deck generation and source code global variable and module listings.« less

Preliminary Analysis of SiC BWR Channel Box Performance under Normal Operation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wirth, Brian; Singh, Gyanender P.; Gorton, Jacob

SiC-SiC composites are being considered for applications in the core components, including BWR channel box and fuel rod cladding, of light water reactors to improve accident tolerance. In the extreme nuclear reactor environment, core components like the BWR channel box will be exposed to neutron damage and a corrosive environment. To ensure reliable and safe operation of a SiC channel box, it is important to assess its deformation behavior under in-reactor conditions including the expected neutron flux and temperature distributions. In particular, this work has evaluated the effect of non-uniform dimensional changes caused by spatially varying neutron flux and temperaturesmore » on the deformation behavior of the channel box over the course of one cycle of irradiation. These analyses have been performed using the fuel performance modeling code BISON and the commercial finite element analysis code Abaqus, based on fast flux and temperature boundary conditions have been calculated using the neutronics and thermal-hydraulics codes Serpent2 and COBRA-TF, respectively. The dependence of dimensions and thermophysical properties on fast flux and temperature has been incorporated into the material models. These initial results indicate significant bowing of the channel box with a lateral displacement greater than 6.5mm. The channel box bowing behavior is time dependent, and driven by the temperature dependence of the SiC irradiation-induced swelling and the neutron flux/fluence gradients. The bowing behavior gradually recovers during the course of the operating cycle as the swelling of the SiC-SiC material saturates. However, the bending relaxation due to temperature gradients does not fully recover and residual bending remains after the swelling saturates in the entire channel box.« less

QR code for medical information uses.

PubMed

Fontelo, Paul; Liu, Fang; Ducut, Erick G

2008-11-06

We developed QR code online tools, simulated and tested QR code applications for medical information uses including scanning QR code labels, URLs and authentication. Our results show possible applications for QR code in medicine.

Simulation of a small cold-leg-break experiment at the PMK-2 test facility using the RELAP5 and ATHLET codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ezsoel, G.; Guba, A.; Perneczky, L.

Results of a small-break loss-of-coolant accident experiment, conducted on the PMK-2 integral-type test facility are presented. The experiment simulated a 1% break in the cold leg of a VVER-440-type reactor. The main phenomena of the experiment are discussed, and in the case of selected events, a more detailed interpretation with the help of measured void fraction, obtained by a special measurement device, is given. Two thermohydraulic computer codes, RELAP5 and ATHLET, are used for posttest calculations. The aim of these calculations is to investigate the code capability for modeling natural circulation phenomena in VVER-440-type reactors. Therefore, the results of themore » experiment and both calculations are compared. Both codes predict most of the transient events well, with the exception that RELAP5 fails to predict the dryout period in the core. In the experiment, the hot- and cold-leg loop-seal clearing is accompanied by natural circulation instabilities, which can be explained by means of the ATHLET calculation.« less

Dosimetric comparison of Monte Carlo codes (EGS4, MCNP, MCNPX) considering external and internal exposures of the Zubal phantom to electron and photon sources.

PubMed

Chiavassa, S; Lemosquet, A; Aubineau-Lanièce, I; de Carlan, L; Clairand, I; Ferrer, L; Bardiès, M; Franck, D; Zankl, M

2005-01-01

This paper aims at comparing dosimetric assessments performed with three Monte Carlo codes: EGS4, MCNP4c2 and MCNPX2.5e, using a realistic voxel phantom, namely the Zubal phantom, in two configurations of exposure. The first one deals with an external irradiation corresponding to the example of a radiological accident. The results are obtained using the EGS4 and the MCNP4c2 codes and expressed in terms of the mean absorbed dose (in Gy per source particle) for brain, lungs, liver and spleen. The second one deals with an internal exposure corresponding to the treatment of a medullary thyroid cancer by 131I-labelled radiopharmaceutical. The results are obtained by EGS4 and MCNPX2.5e and compared in terms of S-values (expressed in mGy per kBq and per hour) for liver, kidney, whole body and thyroid. The results of these two studies are presented and differences between the codes are analysed and discussed.

Accidents caused by Bothrops and Bothropoides in the State of Paraiba: epidemiological and clinical aspects.

PubMed

Oliveira, Fagner Neves; Brito, Monalisa Taveira; Morais, Isabel Cristina Oliveira de; Fook, Sayonara Maria Lia; Albuquerque, Helder Neves de

2010-01-01

Bothrops and Bothropoides snakes cause 70% of the ophidic accidents in Brazil. The species that cause ophidic accidents in State of Paraíba are Bothropoides erythromelas, Bothrops leucurus and Bothropoides neuwiedi. This is a prospective and transverse study, following a quantitative approach of accidents involving Bothrops and Bothropoides admitted to the Toxicological Assistance and Information Centers of Campina Grande and João Pessoa (Ceatox-CG and Ceatox-JP), aimed at identifying the epidemiological and clinical profile of such accidents. All of the patients admitted had medical diagnoses and were monitored at Ceatox-CG or Ceatox-JP. The genera Bothrops and Bothropoides caused 91.7% of the ophidic accidents reported. Snake bites were frequent in men (75.1%), rural workers (65.1%), literate individuals (69%) between 11 and 20 years-old (21.7%), and toes the most common area attacked (52.7%). Most (86.6%) patients were admitted within 6 hours after the accident/bite, with a predominance of mild cases (64.6%). The annual occurrence in Paraíba was 5.5 accidents/100,000 inhabitants and lethality was 0.2%. Positive changes in the profiles of these accidents were verified, such as the non-application of inadequate solutions, including the use of tourniquet, coffee grounds, garlic, suction and/or cutting the bitten area. Moreover, the Itinerant Laboratory project, linked to Paraíba State University in partnership with Ceatox-CG, has contributed positively, providing several cities of the state with information regarding the prevention of accidents involving venomous animals. The local press has also contributed, reporting the educational work developed by the centers.

Current and anticipated uses of the thermal hydraulics codes at the NRC

DOE Office of Scientific and Technical Information (OSTI.GOV)

Caruso, R.

1997-07-01

The focus of Thermal-Hydraulic computer code usage in nuclear regulatory organizations has undergone a considerable shift since the codes were originally conceived. Less work is being done in the area of {open_quotes}Design Basis Accidents,{close_quotes}, and much more emphasis is being placed on analysis of operational events, probabalistic risk/safety assessment, and maintenance practices. All of these areas need support from Thermal-Hydraulic computer codes to model the behavior of plant fluid systems, and they all need the ability to perform large numbers of analyses quickly. It is therefore important for the T/H codes of the future to be able to support thesemore » needs, by providing robust, easy-to-use, tools that produce easy-to understand results for a wider community of nuclear professionals. These tools need to take advantage of the great advances that have occurred recently in computer software, by providing users with graphical user interfaces for both input and output. In addition, reduced costs of computer memory and other hardware have removed the need for excessively complex data structures and numerical schemes, which make the codes more difficult and expensive to modify, maintain, and debug, and which increase problem run-times. Future versions of the T/H codes should also be structured in a modular fashion, to allow for the easy incorporation of new correlations, models, or features, and to simplify maintenance and testing. Finally, it is important that future T/H code developers work closely with the code user community, to ensure that the code meet the needs of those users.« less

77 FR 137 - Applications and Amendments to Facility Operating Licenses Involving Proposed No Significant...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-01-03

... the LSCS, Cycle 15, operation. Cycle 15 will be the first cycle of operation with a mixed core... methodologies. The analyses for LSCS, Unit 1, Cycle 15 have concluded that a two-loop MCPR SL of >= 1.13, based... accident from any accident previously evaluated? Response: No. The GNF2 fuel to be used in Cycle 15 is of a...

Quantification of color vision using a tablet display.

PubMed

Chacon, Alicia; Rabin, Jeff; Yu, Dennis; Johnston, Shawn; Bradshaw, Timothy

2015-01-01

Accurate color vision is essential for optimal performance in aviation and space environments using nonredundant color coding to convey critical information. Most color tests detect color vision deficiency (CVD) but fail to diagnose type or severity of CVD, which are important to link performance to occupational demands. The computer-based Cone Contrast Test (CCT) diagnoses type and severity of CVD. It is displayed on a netbook computer for clinical application, but a more portable version may prove useful for deployments, space and aviation cockpits, as well as accident and sports medicine settings. Our purpose was to determine if the CCT can be conducted on a tablet display (Windows 8, Microsoft, Seattle, WA) using touch-screen response input. The CCT presents colored letters visible only to red (R), green (G), and blue (B) sensitive retinal cones to determine the lowest R, G, and B cone contrast visible to the observer. The CCT was measured in 16 color vision normals (CVN) and 16 CVDs using the standard netbook computer and a Windows 8 tablet display calibrated to produce equal color contrasts. Both displays showed 100% specificity for confirming CVN and 100% sensitivity for detecting CVD. In CVNs there was no difference between scores on netbook vs. tablet displays. G cone CVDs showed slightly lower G cone CCT scores on the tablet. CVD can be diagnosed with a tablet display. Ease-of-use, portability, and complete computer capabilities make tablets ideal for multiple settings, including aviation, space, military deployments, accidents and rescue missions, and sports vision. Chacon A, Rabin J, Yu D, Johnston S, Bradshaw T. Quantification of color vision using a tablet display.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Guntay, Salih; Dehbi, Abdel; Suckow, Detlef

2002-07-01

Steam generator tube rupture (SGTR) incidents, such as those, which occurred in various operating pressurized, water reactors in the past, are serious operational concerns and remain among the most risk-dominant events. Although considerable efforts have been spent to understand tube degradation processes, develop improved modes of operation, and take preventative and corrective measures, SGTR incidents cannot be completely ruled out. Under certain conditions, high releases of radionuclides to the environment are possible during design basis accidents (DBA) and severe accidents. The severe accident codes' models for aerosol retention in the secondary side of a steam generator (SG) have not beenmore » assessed against any experimental data, which means that the uncertainties in the source term following an un-isolated SGTR concurrent with a severe accident are not currently quantified. The accident management (AM) procedures aim at avoiding or minimizing the release of fission products from the SG. The enhanced retention of activity within the SG defines the effectiveness of the accident management actions for the specific hardware characteristics and accident conditions of concern. A sound database on aerosol retention due to natural processes in the SG is not available, nor is an assessment of the effect of management actions on these processes. Hence, the effectiveness of the AM in SGTR events is not presently known. To help reduce uncertainties relating to SGTR issues, an experimental project, ARTIST (Aerosol Trapping In a Steam generator), has been initiated at the Paul Scherrer Institut to address aerosol and droplet retention in the various parts of the SG. The test section is comprised of a scaled-down tube bundle, a full-size separator and a full-size dryer unit. The project will study phenomena at the separate effect and integral levels and address AM issues in seven distinct phases: Aerosol retention in 1) the broken tube under dry secondary side conditions, 2) the near field close to break under dry conditions, 3) the bundle far-field under dry conditions, 4) the separator and dryer under dry conditions, 5) the bundle section under wet conditions, 6) droplet retention in the separator and dryer sections and 7) the overall SG (integral tests). Prototypical test parameters are selected to cover the range of conditions expected in severe accident as well as DBA scenarios. This paper summarizes the relevant issues and introduces the ARTIST facility and the provisional test program which will run between 2003 and 2007. (authors)« less

A case study analysis to examine motorcycle crashes in Bogota, Colombia.

PubMed

Jimenez, Adriana; Bocarejo, Juan Pablo; Zarama, Roberto; Yerpez, Joël

2015-02-01

Contributory factors to motorcycle crashes vary among populations depending on several aspects such as the users' profiles, the composition and density of traffic, and the infrastructure features. A better understanding of local motorcycle crashes can be reached in those places where a comprehensive analysis is performed. This paper presents the results obtained from a case study analysis of 400 police records of accidents involving motorcycles in Bogota. To achieve a deeper level of understanding of how these accidents occur, we propose a systemic approach that uses available crash data. The methodology is inspired by accident prototypical scenarios, a tool for analysis developed in France. When grouping cases we identified three categories: solo motorcycle accidents, motorcyclist and pedestrian accidents, and accidents involving a motorcycle and another vehicle. Within these categories we undertook in-depth analyses of 32 groups of accidents obtaining valuable information to better comprehend motorcyclists' road crashes in a local context. Recurrent contributory factors in the groups of accidents include: inexperienced motorcyclists, wide urban roads that incite speeding and risky overtaking maneuvers, flowing urban roads that encourage high speed and increased interaction between vehicles, and lack of infrastructure maintenance. The results obtained are a valuable asset to define measures that will be conveniently adapted to the group of accident on which we want to act. The methodology exposed in this paper is applicable to the study of road crashes that involve all types of actors, not only the motorcyclists, and in contexts different than those presented in Bogota. Copyright © 2014 National Safety Council and Elsevier Ltd. All rights reserved.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sattison, M.B.; Schroeder, J.A.; Russell, K.D.

The Idaho National Engineering Laboratory (INEL) over the past year has created 75 plant-specific Accident Sequence Precursor (ASP) models using the SAPHIRE suite of PRA codes. Along with the new models, the INEL has also developed a new module for SAPHIRE which is tailored specifically to the unique needs of ASP evaluations. These models and software will be the next generation of risk tools for the evaluation of accident precursors by both NRR and AEOD. This paper presents an overview of the models and software. Key characteristics include: (1) classification of the plant models according to plant response with amore » unique set of event trees for each plant class, (2) plant-specific fault trees using supercomponents, (3) generation and retention of all system and sequence cutsets, (4) full flexibility in modifying logic, regenerating cutsets, and requantifying results, and (5) user interface for streamlined evaluation of ASP events.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sattison, M.B.; Schroeder, J.A.; Russell, K.D.

The Idaho National Engineering Laboratory (INEL) over the past year has created 75 plant-specific Accident Sequence Precursor (ASP) models using the SAPHIRE suite of PRA codes. Along with the new models, the INEL has also developed a new module for SAPHIRE which is tailored specifically to the unique needs of conditional core damage probability (CCDP) evaluations. These models and software will be the next generation of risk tools for the evaluation of accident precursors by both NRR and AEOD. This paper presents an overview of the models and software. Key characteristics include: (1) classification of the plant models according tomore » plant response with a unique set of event trees for each plant class, (2) plant-specific fault trees using supercomponents, (3) generation and retention of all system and sequence cutsets, (4) full flexibility in modifying logic, regenerating cutsets, and requantifying results, and (5) user interface for streamlined evaluation of ASP events.« less

The ability of flexible car bonnets to mitigate the consequences of frontal impact with pedestrians

NASA Astrophysics Data System (ADS)

Stanisławek, Sebastian; Niezgoda, Tadeusz

2018-01-01

The paper presents the results of numerical research on a vehicle representing a Toyota Yaris passenger sedan hitting a pedestrian. A flexible car body is suggested as an interesting way to increase safety. The authors present a simple low-cost bonnet buffer concept that may mitigate the effects of frontal impact. Computer simulation was the method chosen to solve the problem efficiently. The Finite Element Method (FEM) implemented in the LS-DYNA commercial code was used. The testing procedure was based on the Euro NCAP protocol. A flexible bonnet buffer shows its usefulness in preventing casualties in typical accidents. In the best scenario, the HIC15 parameter is only 380 when such a buffer is installed. In comparison, an accident involving a car without any protection produces an HIC15 of 970, which is very dangerous for pedestrians.

[Medical and legal considerations in whiplash injury].

PubMed

Castillo-Chávez, Miguel Angel

2013-01-01

Whiplash injury usually occurs in people who suffered an automobile accident, but also occurs as a result of physical assault and other mechanisms. Diagnosis and initial management of the patient by the emergency physician or orthopedist, and prescribing indications, are taken into account by two forensic intervention specialists. One of these is the medical officer, who, through analysis of the injury mechanism, establishes a cause-effect relationship and concludes whether the accident suffered by a worker it is related to work or not, determines how long the worker will remain disabled and if the injury caused permanent disability under Federal Labor Law. The medical examiner by injury classification assists the Public Ministry so that it can frame the crime of injury to the Criminal Code of Federal District. For these reasons a review of medical information about the mechanism of injury, diagnosis, treatment and healing time was performed to help both specialists to standardize their approach in their daily activities.

Rate theory scenarios study on fission gas behavior of U 3 Si 2 under LOCA conditions in LWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miao, Yinbin; Gamble, Kyle A.; Andersson, David

Fission gas behavior of U3Si2 under various loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs) was simulated using rate theory. A rate theory model for U3Si2 that covers both steady-state operation and power transients was developed for the GRASS-SST code based on existing research reactor/ion irradiation experimental data and theoretical predictions of density functional theory (DFT) calculations. The steady-state and LOCA condition parameters were either directly provided or inspired by BISON simulations. Due to the absence of in-pile experiment data for U3Si2's fuel performance under LWR conditions at this stage of accident tolerant fuel (ATF) development, a variety ofmore » LOCA scenarios were taken into consideration to comprehensively and conservatively evaluate the fission gas behavior of U3Si2 during a LOCA.« less

Investigating accident causation through information network modelling.

PubMed

Griffin, T G C; Young, M S; Stanton, N A

2010-02-01

Management of risk in complex domains such as aviation relies heavily on post-event investigations, requiring complex approaches to fully understand the integration of multi-causal, multi-agent and multi-linear accident sequences. The Event Analysis of Systemic Teamwork methodology (EAST; Stanton et al. 2008) offers such an approach based on network models. In this paper, we apply EAST to a well-known aviation accident case study, highlighting communication between agents as a central theme and investigating the potential for finding agents who were key to the accident. Ultimately, this work aims to develop a new model based on distributed situation awareness (DSA) to demonstrate that the risk inherent in a complex system is dependent on the information flowing within it. By identifying key agents and information elements, we can propose proactive design strategies to optimize the flow of information and help work towards avoiding aviation accidents. Statement of Relevance: This paper introduces a novel application of an holistic methodology for understanding aviation accidents. Furthermore, it introduces an ongoing project developing a nonlinear and prospective method that centralises distributed situation awareness and communication as themes. The relevance of findings are discussed in the context of current ergonomic and aviation issues of design, training and human-system interaction.

Multiple external hazards compound level 3 PSA methods research of nuclear power plant

NASA Astrophysics Data System (ADS)

Wang, Handing; Liang, Xiaoyu; Zhang, Xiaoming; Yang, Jianfeng; Liu, Weidong; Lei, Dina

2017-01-01

2011 Fukushima nuclear power plant severe accident was caused by both earthquake and tsunami, which results in large amount of radioactive nuclides release. That accident has caused the radioactive contamination on the surrounding environment. Although this accident probability is extremely small, once such an accident happens that is likely to release a lot of radioactive materials into the environment, and cause radiation contamination. Therefore, studying accidents consequences is important and essential to improve nuclear power plant design and management. Level 3 PSA methods of nuclear power plant can be used to analyze radiological consequences, and quantify risk to the public health effects around nuclear power plants. Based on multiple external hazards compound level 3 PSA methods studies of nuclear power plant, and the description of the multiple external hazards compound level 3 PSA technology roadmap and important technical elements, as well as taking a coastal nuclear power plant as the reference site, we analyzed the impact of off-site consequences of nuclear power plant severe accidents caused by multiple external hazards. At last we discussed the impact of off-site consequences probabilistic risk studies and its applications under multiple external hazards compound conditions, and explained feasibility and reasonableness of emergency plans implementation.

Ex-Vessel Core Melt Modeling Comparison between MELTSPREAD-CORQUENCH and MELCOR 2.1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robb, Kevin R.; Farmer, Mitchell; Francis, Matthew W.

System-level code analyses by both United States and international researchers predict major core melting, bottom head failure, and corium-concrete interaction for Fukushima Daiichi Unit 1 (1F1). Although system codes such as MELCOR and MAAP are capable of capturing a wide range of accident phenomena, they currently do not contain detailed models for evaluating some ex-vessel core melt behavior. However, specialized codes containing more detailed modeling are available for melt spreading such as MELTSPREAD as well as long-term molten corium-concrete interaction (MCCI) and debris coolability such as CORQUENCH. In a preceding study, Enhanced Ex-Vessel Analysis for Fukushima Daiichi Unit 1: Meltmore » Spreading and Core-Concrete Interaction Analyses with MELTSPREAD and CORQUENCH, the MELTSPREAD-CORQUENCH codes predicted the 1F1 core melt readily cooled in contrast to predictions by MELCOR. The user community has taken notice and is in the process of updating their systems codes; specifically MAAP and MELCOR, to improve and reduce conservatism in their ex-vessel core melt models. This report investigates why the MELCOR v2.1 code, compared to the MELTSPREAD and CORQUENCH 3.03 codes, yield differing predictions of ex-vessel melt progression. To accomplish this, the differences in the treatment of the ex-vessel melt with respect to melt spreading and long-term coolability are examined. The differences in modeling approaches are summarized, and a comparison of example code predictions is provided.« less

Stochastic Plume Simulations for the Fukushima Accident and the Deep Water Horizon Oil Spill

NASA Astrophysics Data System (ADS)

Coelho, E.; Peggion, G.; Rowley, C.; Hogan, P.

2012-04-01

The Fukushima Dai-ichi power plant suffered damage leading to radioactive contamination of coastal waters. Major issues in characterizing the extent of the affected waters were a poor knowledge of the radiation released to the coastal waters and the rather complex coastal dynamics of the region, not deterministically captured by the available prediction systems. Equivalently, during the Gulf of Mexico Deep Water Horizon oil platform accident in April 2010, significant amounts of oil and gas were released from the ocean floor. For this case, issues in mapping and predicting the extent of the affected waters in real-time were a poor knowledge of the actual amounts of oil reaching the surface and the fact that coastal dynamics over the region were not deterministically captured by the available prediction systems. To assess the ocean regions and times that were most likely affected by these accidents while capturing the above sources of uncertainty, ensembles of the Navy Coastal Ocean Model (NCOM) were configured over the two regions (NE Japan and Northern Gulf of Mexico). For the Fukushima case tracers were released on each ensemble member; their locations at each instant provided reference positions of water volumes where the signature of water released from the plant could be found. For the Deep Water Horizon oil spill case each ensemble member was coupled with a diffusion-advection solution to estimate possible scenarios of oil concentrations using perturbed estimates of the released amounts as the source terms at the surface. Stochastic plumes were then defined using a Risk Assessment Code (RAC) analysis that associates a number from 1 to 5 to each grid point, determined by the likelihood of having tracer particle within short ranges (for the Fukushima case), hence defining the high risk areas and those recommended for monitoring. For the Oil Spill case the RAC codes were determined by the likelihood of reaching oil concentrations as defined in the Bonn Agreement Oil Appearance Code. The likelihoods were taken in both cases from probability distribution functions derived from the ensemble runs. Results were compared with a control-deterministic solution and checked against available reports to assess their skill in capturing the actual observed plumes and other in-situ data, as well as their relevance for planning surveys and reconnaissance flights for both cases.

Mathematical fundamentals for the noise immunity of the genetic code.

PubMed

Fimmel, Elena; Strüngmann, Lutz

2018-02-01

Symmetry is one of the essential and most visible patterns that can be seen in nature. Starting from the left-right symmetry of the human body, all types of symmetry can be found in crystals, plants, animals and nature as a whole. Similarly, principals of symmetry are also some of the fundamental and most useful tools in modern mathematical natural science that play a major role in theory and applications. As a consequence, it is not surprising that the desire to understand the origin of life, based on the genetic code, forces us to involve symmetry as a mathematical concept. The genetic code can be seen as a key to biological self-organisation. All living organisms have the same molecular bases - an alphabet consisting of four letters (nitrogenous bases): adenine, cytosine, guanine, and thymine. Linearly ordered sequences of these bases contain the genetic information for synthesis of proteins in all forms of life. Thus, one of the most fascinating riddles of nature is to explain why the genetic code is as it is. Genetic coding possesses noise immunity which is the fundamental feature that allows to pass on the genetic information from parents to their descendants. Hence, since the time of the discovery of the genetic code, scientists have tried to explain the noise immunity of the genetic information. In this chapter we will discuss recent results in mathematical modelling of the genetic code with respect to noise immunity, in particular error-detection and error-correction. We will focus on two central properties: Degeneracy and frameshift correction. Different amino acids are encoded by different quantities of codons and a connection between this degeneracy and the noise immunity of genetic information is a long standing hypothesis. Biological implications of the degeneracy have been intensively studied and whether the natural code is a frozen accident or a highly optimised product of evolution is still controversially discussed. Symmetries in the structure of degeneracy of the genetic code are essential and give evidence of substantial advantages of the natural code over other possible ones. In the present chapter we will present a recent approach to explain the degeneracy of the genetic code by algorithmic methods from bioinformatics, and discuss its biological consequences. The biologists recognised this problem immediately after the detection of the non-overlapping structure of the genetic code, i.e., coding sequences are to be read in a unique way determined by their reading frame. But how does the reading head of the ribosome recognises an error in the grouping of codons, caused by e.g. insertion or deletion of a base, that can be fatal during the translation process and may result in nonfunctional proteins? In this chapter we will discuss possible solutions to the frameshift problem with a focus on the theory of so-called circular codes that were discovered in large gene populations of prokaryotes and eukaryotes in the early 90s. Circular codes allow to detect a frameshift of one or two positions and recently a beautiful theory of such codes has been developed using statistics, group theory and graph theory. Copyright © 2017 Elsevier B.V. All rights reserved.

Forgotten Digital Tourniquet: Salvage of an Ischaemic Finger by Application of Medicinal Leeches

PubMed Central

Durrant, C; Townley, WA; Ramkumar, S; Khoo, CTK

2006-01-01

Individual finger tourniquets are appropriate to the management of a wide range of conditions presenting to an accident and emergency department. They are simpler and more comfortable to use than upper arm pneumatic tourniquets and commercially available digital tourniquets are not readily available in the accident and emergency unit. However, if a finger tourniquet is overlooked, ischaemia of the digit results, and gangrene may follow if the problem is not defused early enough, leading to potential disaster.1–3 We present one case where a digit was salvaged after 4 days of tourniquet application, using medicinal leeches. PMID:17002851

Safety Psychology Applicating on Coal Mine Safety Management Based on Information System

NASA Astrophysics Data System (ADS)

Hou, Baoyue; Chen, Fei

In recent years, with the increase of intensity of coal mining, a great number of major accidents happen frequently, the reason mostly due to human factors, but human's unsafely behavior are affected by insecurity mental control. In order to reduce accidents, and to improve safety management, with the help of application security psychology, we analyse the cause of insecurity psychological factors from human perception, from personality development, from motivation incentive, from reward and punishment mechanism, and from security aspects of mental training , and put forward countermeasures to promote coal mine safety production,and to provide information for coal mining to improve the level of safety management.

Severe Accident Scoping Simulations of Accident Tolerant Fuel Concepts for BWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robb, Kevin R.

2015-08-01

Accident-tolerant fuels (ATFs) are fuels and/or cladding that, in comparison with the standard uranium dioxide Zircaloy system, can tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operations [1]. It is important to note that the currently used uranium dioxide Zircaloy fuel system tolerates design basis accidents (and anticipated operational occurrences and normal operation) as prescribed by the US Nuclear Regulatory Commission. Previously, preliminary simulations of the plant response have been performed under a range of accident scenarios using various ATF cladding concepts and fully ceramicmore » microencapsulated fuel. Design basis loss of coolant accidents (LOCAs) and station blackout (SBO) severe accidents were analyzed at Oak Ridge National Laboratory (ORNL) for boiling water reactors (BWRs) [2]. Researchers have investigated the effects of thermal conductivity on design basis accidents [3], investigated silicon carbide (SiC) cladding [4], as well as the effects of ATF concepts on the late stage accident progression [5]. These preliminary analyses were performed to provide initial insight into the possible improvements that ATF concepts could provide and to identify issues with respect to modeling ATF concepts. More recently, preliminary analyses for a range of ATF concepts have been evaluated internationally for LOCA and severe accident scenarios for the Chinese CPR1000 [6] and the South Korean OPR-1000 [7] pressurized water reactors (PWRs). In addition to these scoping studies, a common methodology and set of performance metrics were developed to compare and support prioritizing ATF concepts [8]. A proposed ATF concept is based on iron-chromium-aluminum alloys (FeCrAl) [9]. With respect to enhancing accident tolerance, FeCrAl alloys have substantially slower oxidation kinetics compared to the zirconium alloys typically employed. During a severe accident, FeCrAl would tend to generate heat and hydrogen from oxidation at a slower rate compared to the zirconium-based alloys in use today. The previous study, [2], of the FeCrAl ATF concept during station blackout (SBO) severe accident scenarios in BWRs was based on simulating short term SBO (STSBO), long term SBO (LTSBO), and modified SBO scenarios occurring in a BWR-4 reactor with MARK-I containment. The analysis indicated that FeCrAl had the potential to delay the onset of fuel failure by a few hours depending on the scenario, and it could delay lower head failure by several hours. The analysis demonstrated reduced in-vessel hydrogen production. However, the work was preliminary and was based on limited knowledge of material properties for FeCrAl. Limitations of the MELCOR code were identified for direct use in modeling ATF concepts. This effort used an older version of MELCOR (1.8.5). Since these analyses, the BWR model has been updated for use in MELCOR 1.8.6 [10], and more representative material properties for FeCrAl have been modeled. Sections 2 4 present updated analyses for the FeCrAl ATF concept response during severe accidents in a BWR. The purpose of the study is to estimate the potential gains afforded by the FeCrAl ATF concept during BWR SBO scenarios.« less

BNL program in support of LWR degraded-core accident analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ginsberg, T.; Greene, G.A.

1982-01-01

Two major sources of loading on dry watr reactor containments are steam generatin from core debris water thermal interactions and molten core-concrete interactions. Experiments are in progress at BNL in support of analytical model development related to aspects of the above containment loading mechanisms. The work supports development and evaluation of the CORCON (Muir, 1981) and MARCH (Wooton, 1980) computer codes. Progress in the two programs is described in this paper. 8 figures.

United States Coast Guard 2010 Posture Statement: With 2011 Budget in Brief

DTIC Science & Technology

2010-02-01

marine animals and plants, and prevents foreign poaching out to 200 miles offshore. Did you know? Coast Guard Hero Petty Offi cer 3rd Class Caleb S...Code). Did you know? U.S. Coast Guard Posture Statement • 17 Coast Guard Missions hazardous substance accidents and reduce their impact on the...inspects all vessels’ ballast water before they enter the Great Lakes to prevent invasive species from inhabiting the ecosystem . Photo by PA3 William

Racial Differences in Accidental and Violent Deaths Among U.S. Navy Personnel

DTIC Science & Technology

1984-10-01

for accidental injuries than whites; i O bjective The objective of this study was to examine the death rates of black and white enlisted personnel...pre-servIce soclo- cultural factors may play a role in accounting for the high death rates among older black .males. It is suggested, however, that...all deaths of personnel diagnosed as having injuries due to accidents, poisonings and violence (ICDA-8 codes 800.0 -. 999.9). Death rates were

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Montgomery, Robert; Tomé, Carlos; Liu, Wenfeng

Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. CASL has endeavored to improve upon this approach by incorporating a microstructurally-based, atomistically-informed, zirconium alloy mechanical deformation analysis capability into the BISON-CASL engineering scale fuel performance code. Specifically, the viscoplastic self-consistent (VPSC) polycrystal plasticity modeling approach, developed bymore » Lebensohn and Tome´ [2], has been coupled with BISON-CASL to represent the mechanistic material processes controlling the deformation behavior of the cladding. A critical component of VPSC is the representation of the crystallographic orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON-CASL and provides initial results utilizing the coupled functionality.« less

Thermodynamic properties of gaseous ruthenium species.

PubMed

Miradji, Faoulat; Souvi, Sidi; Cantrel, Laurent; Louis, Florent; Vallet, Valérie

2015-05-21

The review of thermodynamic data of ruthenium oxides reveals large uncertainties in some of the standard enthalpies of formation, motivating the use of high-level relativistic correlated quantum chemical methods to reduce the level of discrepancies. The reaction energies leading to the formation of ruthenium oxides RuO, RuO2, RuO3, and RuO4 have been calculated for a series of reactions. The combination of different quantum chemical methods has been investigated [DFT, CASSCF, MRCI, CASPT2, CCSD(T)] in order to predict the geometrical parameters, the energetics including electronic correlation and spin-orbit coupling. The most suitable method for ruthenium compounds is the use of TPSSh-5%HF for geometry optimization, followed by CCSD(T) with complete basis set (CBS) extrapolations for the calculation of the total electronic energies. SO-CASSCF seems to be accurate enough to estimate spin-orbit coupling contributions to the ground-state electronic energies. This methodology yields very accurate standard enthalpies of formations of all species, which are either in excellent agreement with the most reliable experimental data or provide an improved estimate for the others. These new data will be implemented in the thermodynamical databases that are used by the ASTEC code (accident source term evaluation code) to build models of ruthenium chemistry behavior in severe nuclear accident conditions. The paper also discusses the nature of the chemical bonds both from molecular orbital and topological view points.

Numerical investigation on super-cooled large droplet icing of fan rotor blade in jet engine

NASA Astrophysics Data System (ADS)

Isobe, Keisuke; Suzuki, Masaya; Yamamoto, Makoto

2014-10-01

Icing (or ice accretion) is a phenomenon in which super-cooled water droplets impinge and accrete on a body. It is well known that ice accretion on blades and vanes leads to performance degradation and has caused severe accidents. Although various anti-icing and deicing systems have been developed, such accidents still occur. Therefore, it is important to clarify the phenomenon of ice accretion on an aircraft and in a jet engine. However, flight tests for ice accretion are very expensive, and in the wind tunnel it is difficult to reproduce all climate conditions where ice accretion can occur. Therefore, it is expected that computational fluid dynamics (CFD), which can estimate ice accretion in various climate conditions, will be a useful way to predict and understand the ice accretion phenomenon. On the other hand, although the icing caused by super-cooled large droplets (SLD) is very dangerous, the numerical method has not been established yet. This is why SLD icing is characterized by splash and bounce phenomena of droplets and they are very complex in nature. In the present study, we develop an ice accretion code considering the splash and bounce phenomena to predict SLD icing, and the code is applied to a fan rotor blade. The numerical results with and without the SLD icing model are compared. Through this study, the influence of the SLD icing model is numerically clarified.

Revised catalog of types of CODES applications implemented using linked state data : crash outcome data evaluation system (CODES)

DOT National Transportation Integrated Search

2000-06-01

The purpose of the Revised Catalog of Types of CODES Applications Implemented Using Linked : State Data (CODES) is to inspire the development of new applications for linked data that support : efforts to reduce death, disability, severity, and health...

Application of Electron Microscopy Techniques to the Investigation of Space Shuttle Columbia Accident

NASA Technical Reports Server (NTRS)

Shah, Sandeep

2005-01-01

This viewgraph presentation gives an overview of the investigation into the breakup of the Space Shuttle Columbia, and addresses the importance of a failure analysis strategy for the investigation of the Columbia accident. The main focus of the presentation is on the usefulness of electron microscopy for analyzing slag deposits from the tiles and reinforced carbon-carbon (RCC) wing panels of the Columbia orbiter.

Human error analysis of commercial aviation accidents: application of the Human Factors Analysis and Classification system (HFACS).

PubMed

Wiegmann, D A; Shappell, S A

2001-11-01

The Human Factors Analysis and Classification System (HFACS) is a general human error framework originally developed and tested within the U.S. military as a tool for investigating and analyzing the human causes of aviation accidents. Based on Reason's (1990) model of latent and active failures, HFACS addresses human error at all levels of the system, including the condition of aircrew and organizational factors. The purpose of the present study was to assess the utility of the HFACS framework as an error analysis and classification tool outside the military. The HFACS framework was used to analyze human error data associated with aircrew-related commercial aviation accidents that occurred between January 1990 and December 1996 using database records maintained by the NTSB and the FAA. Investigators were able to reliably accommodate all the human causal factors associated with the commercial aviation accidents examined in this study using the HFACS system. In addition, the classification of data using HFACS highlighted several critical safety issues in need of intervention research. These results demonstrate that the HFACS framework can be a viable tool for use within the civil aviation arena. However, additional research is needed to examine its applicability to areas outside the flight deck, such as aircraft maintenance and air traffic control domains.

Application of the DG-1199 methodology to the ESBWR and ABWR.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kalinich, Donald A.; Gauntt, Randall O.; Walton, Fotini

2010-09-01

Appendix A-5 of Draft Regulatory Guide DG-1199 'Alternative Radiological Source Term for Evaluating Design Basis Accidents at Nuclear Power Reactors' provides guidance - applicable to RADTRAD MSIV leakage models - for scaling containment aerosol concentration to the expected steam dome concentration in order to preserve the simplified use of the Accident Source Term (AST) in assessing containment performance under assumed design basis accident (DBA) conditions. In this study Economic and Safe Boiling Water Reactor (ESBWR) and Advanced Boiling Water Reactor (ABWR) RADTRAD models are developed using the DG-1199, Appendix A-5 guidance. The models were run using RADTRAD v3.03. Low Populationmore » Zone (LPZ), control room (CR), and worst-case 2-hr Exclusion Area Boundary (EAB) doses were calculated and compared to the relevant accident dose criteria in 10 CFR 50.67. For the ESBWR, the dose results were all lower than the MSIV leakage doses calculated by General Electric/Hitachi (GEH) in their licensing technical report. There are no comparable ABWR MSIV leakage doses, however, it should be noted that the ABWR doses are lower than the ESBWR doses. In addition, sensitivity cases were evaluated to ascertain the influence/importance of key input parameters/features of the models.« less

Injury patterns among various age and gender groups of trauma patients in southern Iran

PubMed Central

Bolandparvaz, Shahram; Yadollahi, Mahnaz; Abbasi, Hamid Reza; Anvar, Mehrdad

2017-01-01

Abstract Administrative data from trauma referral centers are useful sources while studying epidemiologic aspects of injuries. We aimed to provide a hospital-based view of injuries in Shiraz considering victims’ age and gender, using administrative data from trauma research center. A cross-sectional registry-based study of adult trauma patients (age ≥15 years) sustaining injury through traffic accidents, violence, and unintentional incidents was conducted. Information was retrieved from 3 hospital administrative databases. Data on demographics, injury mechanisms, injured body regions, and injury descriptions; outcomes of hospitalization; and development of nosocomial infections were recorded. Injury Severity Score (ISS) was calculated by crosswalking from ICD-10 (International Classification of Diseases) injury diagnosis codes to AIS-98 (Abbreviated Injury Scale) severity codes. Patients were compared based on age groups and gender differences. A total of 47,295 trauma patients with a median age of 30 (interquartile range: 24–44 years) were studied, of whom 73.1% were male and the remaining 26.9% were female (M/F = 2.7:1.0). The most common injury mechanisms in the male group were car and motorcycle accidents whereas females were mostly victims of falls and pedestrian accidents (P < .01). As age increased, a shift from transportation-related to unintentionally caused injuries occurred. Overall, young men had their most severe injuries on head, whereas elderly women suffered more severe extremity injuries. Injury severity was similar between men and women; however, elderly had a significantly higher ISS. Although incidence of nosocomial infections was independent of victims’ age and gender, elderly men had a significantly higher mortality rate. Based on administrative data from our trauma center, male gender and age >65 years are associated with increased risk of injury incidence, prolonged hospitalizations, and in-hospital death following trauma. Development of a regional trauma surveillance system may provide further opportunities for studying injuries and evaluating preventive actions. PMID:29019874

Injury patterns among various age and gender groups of trauma patients in southern Iran: A cross-sectional study.

PubMed

Bolandparvaz, Shahram; Yadollahi, Mahnaz; Abbasi, Hamid Reza; Anvar, Mehrdad

2017-10-01

Administrative data from trauma referral centers are useful sources while studying epidemiologic aspects of injuries. We aimed to provide a hospital-based view of injuries in Shiraz considering victims' age and gender, using administrative data from trauma research center.A cross-sectional registry-based study of adult trauma patients (age ≥15 years) sustaining injury through traffic accidents, violence, and unintentional incidents was conducted. Information was retrieved from 3 hospital administrative databases. Data on demographics, injury mechanisms, injured body regions, and injury descriptions; outcomes of hospitalization; and development of nosocomial infections were recorded. Injury Severity Score (ISS) was calculated by crosswalking from ICD-10 (International Classification of Diseases) injury diagnosis codes to AIS-98 (Abbreviated Injury Scale) severity codes. Patients were compared based on age groups and gender differences.A total of 47,295 trauma patients with a median age of 30 (interquartile range: 24-44 years) were studied, of whom 73.1% were male and the remaining 26.9% were female (M/F = 2.7:1.0). The most common injury mechanisms in the male group were car and motorcycle accidents whereas females were mostly victims of falls and pedestrian accidents (P < .01). As age increased, a shift from transportation-related to unintentionally caused injuries occurred. Overall, young men had their most severe injuries on head, whereas elderly women suffered more severe extremity injuries. Injury severity was similar between men and women; however, elderly had a significantly higher ISS. Although incidence of nosocomial infections was independent of victims' age and gender, elderly men had a significantly higher mortality rate.Based on administrative data from our trauma center, male gender and age >65 years are associated with increased risk of injury incidence, prolonged hospitalizations, and in-hospital death following trauma. Development of a regional trauma surveillance system may provide further opportunities for studying injuries and evaluating preventive actions.

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

PubMed

Dźwiarek, Marek; Latała, Agata

2016-01-01

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

Upon the reconstruction of accidents triggered by tire explosion. Analytical model and case study

NASA Astrophysics Data System (ADS)

Gaiginschi, L.; Agape, I.; Talif, S.

2017-10-01

Accident Reconstruction is important in the general context of increasing road traffic safety. In the casuistry of traffic accidents, those caused by tire explosions are critical under the severity of consequences, because they are usually happening at high speeds. Consequently, the knowledge of the running speed of the vehicle involved at the time of the tire explosion is essential to elucidate the circumstances of the accident. The paper presents an analytical model for the kinematics of a vehicle which, after the explosion of one of its tires, begins to skid, overturns and rolls. The model consists of two concurent approaches built as applications of the momentum conservation and energy conservation principles, and allows determination of the initial speed of the vehicle involved, by running backwards the sequences of the road event. The authors also aimed to both validate the two distinct analytical approaches by calibrating the calculation algorithms on a case study

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

PubMed Central

Dźwiarek, Marek; Latała, Agata

2016-01-01

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

Application of composite small calibration objects in traffic accident scene photogrammetry.

PubMed

Chen, Qiang; Xu, Hongguo; Tan, Lidong

2015-01-01

In order to address the difficulty of arranging large calibration objects and the low measurement accuracy of small calibration objects in traffic accident scene photogrammetry, a photogrammetric method based on a composite of small calibration objects is proposed. Several small calibration objects are placed around the traffic accident scene, and the coordinate system of the composite calibration object is given based on one of them. By maintaining the relative position and coplanar relationship of the small calibration objects, the local coordinate system of each small calibration object is transformed into the coordinate system of the composite calibration object. The two-dimensional direct linear transformation method is improved based on minimizing the reprojection error of the calibration points of all objects. A rectified image is obtained using the nonlinear optimization method. The increased accuracy of traffic accident scene photogrammetry using a composite small calibration object is demonstrated through the analysis of field experiments and case studies.

The application of coded excitation technology in medical ultrasonic Doppler imaging

NASA Astrophysics Data System (ADS)

Li, Weifeng; Chen, Xiaodong; Bao, Jing; Yu, Daoyin

2008-03-01

Medical ultrasonic Doppler imaging is one of the most important domains of modern medical imaging technology. The application of coded excitation technology in medical ultrasonic Doppler imaging system has the potential of higher SNR and deeper penetration depth than conventional pulse-echo imaging system, it also improves the image quality, and enhances the sensitivity of feeble signal, furthermore, proper coded excitation is beneficial to received spectrum of Doppler signal. Firstly, this paper analyzes the application of coded excitation technology in medical ultrasonic Doppler imaging system abstractly, showing the advantage and bright future of coded excitation technology, then introduces the principle and the theory of coded excitation. Secondly, we compare some coded serials (including Chirp and fake Chirp signal, Barker codes, Golay's complementary serial, M-sequence, etc). Considering Mainlobe Width, Range Sidelobe Level, Signal-to-Noise Ratio and sensitivity of Doppler signal, we choose Barker codes as coded serial. At last, we design the coded excitation circuit. The result in B-mode imaging and Doppler flow measurement coincided with our expectation, which incarnated the advantage of application of coded excitation technology in Digital Medical Ultrasonic Doppler Endoscope Imaging System.

Lessons from Fukushima for Improving the Safety of Nuclear Reactors

NASA Astrophysics Data System (ADS)

Lyman, Edwin

2012-02-01

The March 2011 accident at the Fukushima Daiichi nuclear power plant has revealed serious vulnerabilities in the design, operation and regulation of nuclear power plants. While some aspects of the accident were plant- and site-specific, others have implications that are broadly applicable to the current generation of nuclear plants in operation around the world. Although many of the details of the accident progression and public health consequences are still unclear, there are a number of lessons that can already be drawn. The accident demonstrated the need at nuclear plants for robust, highly reliable backup power sources capable of functioning for many days in the event of a complete loss of primary off-site and on-site electrical power. It highlighted the importance of detailed planning for severe accident management that realistically evaluates the capabilities of personnel to carry out mitigation operations under extremely hazardous conditions. It showed how emergency plans rooted in the assumption that only one reactor at a multi-unit site would be likely to experience a crisis fail miserably in the event of an accident affecting multiple reactor units simultaneously. It revealed that alternate water injection following a severe accident could be needed for weeks or months, generating large volumes of contaminated water that must be contained. And it reinforced the grim lesson of Chernobyl: that a nuclear reactor accident could lead to widespread radioactive contamination with profound implications for public health, the economy and the environment. While many nations have re-examined their policies regarding nuclear power safety in the months following the accident, it remains to be seen to what extent the world will take the lessons of Fukushima seriously and make meaningful changes in time to avert another, and potentially even worse, nuclear catastrophe.

Fukushima Daiichi Unit 1 Ex-Vessel Prediction: Core Concrete Interaction

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robb, Kevin R; Farmer, Mitchell; Francis, Matthew W

Lower head failure and corium concrete interaction were predicted to occur at Fukushima Daiichi Unit 1 (1F1) by several different system-level code analyses, including MELCOR v2.1 and MAAP5. Although these codes capture a wide range of accident phenomena, they do not contain detailed models for ex-vessel core melt behavior. However, specialized codes exist for analysis of ex-vessel melt spreading (e.g., MELTSPREAD) and long-term debris coolability (e.g., CORQUENCH). On this basis, an analysis was carried out to further evaluate ex-vessel behavior for 1F1 using MELTSPREAD and CORQUENCH. Best-estimate melt pour conditions predicted by MELCOR v2.1 and MAAP5 were used as input.more » MELTSPREAD was then used to predict the spatially dependent melt conditions and extent of spreading during relocation from the vessel. The results of the MELTSPREAD analysis are reported in a companion paper. This information was used as input for the long-term debris coolability analysis with CORQUENCH.« less

Fukushima Daiichi Unit 1 ex-vessel prediction: Core melt spreading

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farmer, M. T.; Robb, K. R.; Francis, M. W.

Lower head failure and corium-concrete interaction were predicted to occur at Fukushima Daiichi Unit 1 (1F1) by several different system-level code analyses, including MELCOR v2.1 and MAAP5. Although these codes capture a wide range of accident phenomena, they do not contain detailed models for ex-vessel core melt behavior. However, specialized codes exist for analysis of ex-vessel melt spreading (e.g., MELTSPREAD) and long-term debris coolability (e.g., CORQUENCH). On this basis, an analysis has been carried out to further evaluate ex-vessel behavior for 1F1 using MELTSPREAD and CORQUENCH. Best-estimate melt pour conditions predicted by MELCOR v2.1 and MAAP5 were used as input.more » MELTSPREAD was then used to predict the spatially-dependent melt conditions and extent of spreading during relocation from the vessel. Lastly, this information was then used as input for the long-term debris coolability analysis with CORQUENCH that is reported in a companion paper.« less

Fukushima Daiichi Unit 1 ex-vessel prediction: Core melt spreading

DOE PAGES

Farmer, M. T.; Robb, K. R.; Francis, M. W.

2016-10-31

Lower head failure and corium-concrete interaction were predicted to occur at Fukushima Daiichi Unit 1 (1F1) by several different system-level code analyses, including MELCOR v2.1 and MAAP5. Although these codes capture a wide range of accident phenomena, they do not contain detailed models for ex-vessel core melt behavior. However, specialized codes exist for analysis of ex-vessel melt spreading (e.g., MELTSPREAD) and long-term debris coolability (e.g., CORQUENCH). On this basis, an analysis has been carried out to further evaluate ex-vessel behavior for 1F1 using MELTSPREAD and CORQUENCH. Best-estimate melt pour conditions predicted by MELCOR v2.1 and MAAP5 were used as input.more » MELTSPREAD was then used to predict the spatially-dependent melt conditions and extent of spreading during relocation from the vessel. Lastly, this information was then used as input for the long-term debris coolability analysis with CORQUENCH that is reported in a companion paper.« less

Verification of combined thermal-hydraulic and heat conduction analysis code FLOWNET/TRUMP

NASA Astrophysics Data System (ADS)

Maruyama, Soh; Fujimoto, Nozomu; Kiso, Yoshihiro; Murakami, Tomoyuki; Sudo, Yukio

1988-09-01

This report presents the verification results of the combined thermal-hydraulic and heat conduction analysis code, FLOWNET/TRUMP which has been utilized for the core thermal hydraulic design, especially for the analysis of flow distribution among fuel block coolant channels, the determination of thermal boundary conditions for fuel block stress analysis and the estimation of fuel temperature in the case of fuel block coolant channel blockage accident in the design of the High Temperature Engineering Test Reactor(HTTR), which the Japan Atomic Energy Research Institute has been planning to construct in order to establish basic technologies for future advanced very high temperature gas-cooled reactors and to be served as an irradiation test reactor for promotion of innovative high temperature new frontier technologies. The verification of the code was done through the comparison between the analytical results and experimental results of the Helium Engineering Demonstration Loop Multi-channel Test Section(HENDEL T(sub 1-M)) with simulated fuel rods and fuel blocks.

The impacts of speed cameras on road accidents: an application of propensity score matching methods.

PubMed

Li, Haojie; Graham, Daniel J; Majumdar, Arnab

2013-11-01

This paper aims to evaluate the impacts of speed limit enforcement cameras on reducing road accidents in the UK by accounting for both confounding factors and the selection of proper reference groups. The propensity score matching (PSM) method is employed to do this. A naïve before and after approach and the empirical Bayes (EB) method are compared with the PSM method. A total of 771 sites and 4787 sites for the treatment and the potential reference groups respectively are observed for a period of 9 years in England. Both the PSM and the EB methods show similar results that there are significant reductions in the number of accidents of all severities at speed camera sites. It is suggested that the propensity score can be used as the criteria for selecting the reference group in before-after control studies. Speed cameras were found to be most effective in reducing accidents up to 200 meters from camera sites and no evidence of accident migration was found. Copyright © 2013 Elsevier Ltd. All rights reserved.

Risk analysis of urban gas pipeline network based on improved bow-tie model

NASA Astrophysics Data System (ADS)

Hao, M. J.; You, Q. J.; Yue, Z.

2017-11-01

Gas pipeline network is a major hazard source in urban areas. In the event of an accident, there could be grave consequences. In order to understand more clearly the causes and consequences of gas pipeline network accidents, and to develop prevention and mitigation measures, the author puts forward the application of improved bow-tie model to analyze risks of urban gas pipeline network. The improved bow-tie model analyzes accident causes from four aspects: human, materials, environment and management; it also analyzes the consequences from four aspects: casualty, property loss, environment and society. Then it quantifies the causes and consequences. Risk identification, risk analysis, risk assessment, risk control, and risk management will be clearly shown in the model figures. Then it can suggest prevention and mitigation measures accordingly to help reduce accident rate of gas pipeline network. The results show that the whole process of an accident can be visually investigated using the bow-tie model. It can also provide reasons for and predict consequences of an unfortunate event. It is of great significance in order to analyze leakage failure of gas pipeline network.

Modification of knowledge on occupational accidents among schoolchildren who resided in a landfill impacted area.

PubMed

Gomes, Rosangela Maiara Vindoura; Câmara, Volney de Magalhães; Souza, Delma Perpétua Oliveira de

2016-01-01

The prevalence of occupational accidents is very high in Brazil, having impacts on the health system and social security. This requires prevention, which must start with students of the Basic Education. The knowledge on this kind of accidents among children and adolescents studying in an area near a sanitary landfill was evaluated, before and after the development of activities on health education. A cross-sectional study was conducted in 2013 and included the application of the same questionnaire among students from a school in Cuiabá-MT, Brazil, before and after educational health activities related to the definition of occupational accidents. Univariate analyses of absolute and relative frequencies and bivariate analyses using the χ2 Test and Fisher's Exact Test were performed with a significance level of 0.05 and 95%CI. There was a statistically significant increase of the knowledge on these types of accidents after the educational activities (p < 0.05). The activities carried out indicate that schools are important for the development and systematization of knowledge arising from reality.

24 CFR 92.251 - Property standards.

Code of Federal Regulations, 2011 CFR

2011-04-01

..., as applicable, one of three model codes: Uniform Building Code (ICBO), National Building Code (BOCA), Standard (Southern) Building Code (SBCCI); or the Council of American Building Officials (CABO) one or two...) Housing that is constructed or rehabilitated with HOME funds must meet all applicable local codes...

24 CFR 92.251 - Property standards.

Code of Federal Regulations, 2013 CFR

2013-04-01

..., as applicable, one of three model codes: Uniform Building Code (ICBO), National Building Code (BOCA), Standard (Southern) Building Code (SBCCI); or the Council of American Building Officials (CABO) one or two...) Housing that is constructed or rehabilitated with HOME funds must meet all applicable local codes...

24 CFR 92.251 - Property standards.

Code of Federal Regulations, 2012 CFR

2012-04-01

..., as applicable, one of three model codes: Uniform Building Code (ICBO), National Building Code (BOCA), Standard (Southern) Building Code (SBCCI); or the Council of American Building Officials (CABO) one or two...) Housing that is constructed or rehabilitated with HOME funds must meet all applicable local codes...

24 CFR 92.251 - Property standards.

Code of Federal Regulations, 2010 CFR

2010-04-01

..., as applicable, one of three model codes: Uniform Building Code (ICBO), National Building Code (BOCA), Standard (Southern) Building Code (SBCCI); or the Council of American Building Officials (CABO) one or two...) Housing that is constructed or rehabilitated with HOME funds must meet all applicable local codes...

Explorative spatial analysis of traffic accident statistics and road mortality among the provinces of Turkey.

PubMed

Erdogan, Saffet

2009-10-01

The aim of the study is to describe the inter-province differences in traffic accidents and mortality on roads of Turkey. Two different risk indicators were used to evaluate the road safety performance of the provinces in Turkey. These indicators are the ratios between the number of persons killed in road traffic accidents (1) and the number of accidents (2) (nominators) and their exposure to traffic risk (denominator). Population and the number of registered motor vehicles in the provinces were used as denominators individually. Spatial analyses were performed to the mean annual rate of deaths and to the number of fatal accidents that were calculated for the period of 2001-2006. Empirical Bayes smoothing was used to remove background noise from the raw death and accident rates because of the sparsely populated provinces and small number of accident and death rates of provinces. Global and local spatial autocorrelation analyses were performed to show whether the provinces with high rates of deaths-accidents show clustering or are located closer by chance. The spatial distribution of provinces with high rates of deaths and accidents was nonrandom and detected as clustered with significance of P<0.05 with spatial autocorrelation analyses. Regions with high concentration of fatal accidents and deaths were located in the provinces that contain the roads connecting the Istanbul, Ankara, and Antalya provinces. Accident and death rates were also modeled with some independent variables such as number of motor vehicles, length of roads, and so forth using geographically weighted regression analysis with forward step-wise elimination. The level of statistical significance was taken as P<0.05. Large differences were found between the rates of deaths and accidents according to denominators in the provinces. The geographically weighted regression analyses did significantly better predictions for both accident rates and death rates than did ordinary least regressions, as indicated by adjusted R(2) values. Geographically weighted regression provided values of 0.89-0.99 adjusted R(2) for death and accident rates, compared with 0.88-0.95, respectively, by ordinary least regressions. Geographically weighted regression has the potential to reveal local patterns in the spatial distribution of rates, which would be ignored by the ordinary least regression approach. The application of spatial analysis and modeling of accident statistics and death rates at provincial level in Turkey will help to identification of provinces with outstandingly high accident and death rates. This could help more efficient road safety management in Turkey.

Posttest analysis of international standard problem 10 using RELAP4/MOD7. [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hsu, M.; Davis, C.B.; Peterson, A.C. Jr.

RELAP4/MOD7, a best estimate computer code for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This paper evaluates the capability of RELAP4/MOD7 to calculate refill/reflood phenomena. This evaluation uses the data of International Standard Problem 10, which is based on West Germany's KWU PKL refill/reflood experiment K9A. The PKL test facility represents a typical West German four-loop, 1300 MW pressurized water reactor (PWR) in reduced scale while maintaining prototypical volume-to-power ratio. The PKL facility was designed to specifically simulate the refill/reflood phase of amore » hypothetical loss-of-coolant accident (LOCA).« less

Primary system fission product release and transport: A state-of-the-art report to the committee on the safety of nuclear installations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wright, A.L.

This report presents a summary of the status of research activities associated with fission product behavior (release and transport) under severe accident conditions within the primary systems of water-moderated and water-cooled nuclear reactors. For each of the areas of fission product release and fission product transport, the report summarizes relevant information on important phenomena, major experiments performed, relevant computer models and codes, comparisons of computer code calculations with experimental results, and general conclusions on the overall state of the art. Finally, the report provides an assessment of the overall importance and knowledge of primary system release and transport phenomena andmore » presents major conclusions on the state of the art.« less

Linear chirp phase perturbing approach for finding binary phased codes

NASA Astrophysics Data System (ADS)

Li, Bing C.

2017-05-01

Binary phased codes have many applications in communication and radar systems. These applications require binary phased codes to have low sidelobes in order to reduce interferences and false detection. Barker codes are the ones that satisfy these requirements and they have lowest maximum sidelobes. However, Barker codes have very limited code lengths (equal or less than 13) while many applications including low probability of intercept radar, and spread spectrum communication, require much higher code lengths. The conventional techniques of finding binary phased codes in literatures include exhaust search, neural network, and evolutionary methods, and they all require very expensive computation for large code lengths. Therefore these techniques are limited to find binary phased codes with small code lengths (less than 100). In this paper, by analyzing Barker code, linear chirp, and P3 phases, we propose a new approach to find binary codes. Experiments show that the proposed method is able to find long low sidelobe binary phased codes (code length >500) with reasonable computational cost.

Probabilistic dose assessment of normal operations and accident conditions for an assured isolation facility in Texas

NASA Astrophysics Data System (ADS)

Arno, Matthew Gordon

Texas is investigating building a long-term waste storage facility, also known as an Assured Isolation Facility. This is an above-ground low-level radioactive waste storage facility that is actively maintained and from which waste may be retrieved. A preliminary, scoping-level analysis has been extended to consider more complex scenarios of radiation streaming and skyshine by using the computer code Monte Carlo N-Particle (MCNP) to model the facility in greater detail. Accidental release scenarios have been studied in more depth to better assess the potential dose to off-site individuals. Using bounding source term assumptions, the projected radiation doses and dose rates are estimated to exceed applicable limits by an order of magnitude. By altering the facility design to fill in the hollow cores of the prefabricated concrete slabs used in the roof over the "high-gamma rooms," where the waste with the highest concentration of gamma emitting radioactive material is stored, dose rates outside the facility decrease by an order of magnitude. With the modified design, the annual dose at the site fenceline is estimated at 86 mrem, below the 100 mrem annual limit for exposure of the public. Within the site perimeter, the dose rates are lowered sufficiently such that it is not necessary to categorize many workers and contractor personnel as radiation workers, saving on costs as well as being advisable under ALARA principles. A detailed analysis of bounding accidents incorporating information on the local meteorological conditions indicate that the maximum committed effective dose equivalent from the passage of a plume of material released in an accident at any of the cities near the facility is 59 :rem in the city of Eunice, NM based on the combined day and night meteorological conditions. Using the daytime meteorological conditions, the maximum dose at any city is 7 :rem, also in the city of Eunice. The maximum dose at the site boundary was determined to be 230 mrem using the combined day and night meteorological conditions and 33 mrem using the daytime conditions.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vandenhove, Hildegarde

The accident at the Fukushima Daiichi Nuclear Power Plant has raised questions about the accumulation of radionuclides in soils, the transfer in the food chain and the possibility of continued restricted future land use. This paper summarizes what is generally understood about the application of agricultural countermeasures as a land management option to reduce the radionuclides transfer in the food chain and to facilitate the return of potentially affected soils to agricultural practices in areas impacted by a nuclear accident. (authors)

Application of Non-destructive Methods of Stress-strain State at Hazardous Production Facilities

NASA Astrophysics Data System (ADS)

Shram, V.; Kravtsova, Ye; Selsky, A.; Bezborodov, Yu; Lysyannikova, N.; Lysyannikov, A.

2016-06-01

The paper deals with the sources of accidents in distillation columns, on the basis of which the most dangerous defects are detected. The analysis of the currently existing methods of non-destructive testing of the stress-strain state is performed. It is proposed to apply strain and acoustic emission techniques to continuously monitor dangerous objects, which helps prevent the possibility of accidents, as well as reduce the work.

An In vitro evaluation of the reliability of QR code denture labeling technique

PubMed Central

Poovannan, Sindhu; Jain, Ashish R.; Krishnan, Cakku Jalliah Venkata; Chandran, Chitraa R.

2016-01-01

Statement of Problem: Positive identification of the dead after accidents and disasters through labeled dentures plays a key role in forensic scenario. A number of denture labeling methods are available, and studies evaluating their reliability under drastic conditions are vital. Aim: This study was conducted to evaluate the reliability of QR (Quick Response) Code labeled at various depths in heat-cured acrylic blocks after acid treatment, heat treatment (burns), and fracture in forensics. It was an in vitro study. Materials and Methods: This study included 160 specimens of heat-cured acrylic blocks (1.8 cm × 1.8 cm) and these were divided into 4 groups (40 samples per group). QR Codes were incorporated in the samples using clear acrylic sheet and they were assessed for reliability under various depths, acid, heat, and fracture. Data were analyzed using Chi-square test, test of proportion. Results: The QR Code inclusion technique was reliable under various depths of acrylic sheet, acid (sulfuric acid 99%, hydrochloric acid 40%) and heat (up to 370°C). Results were variable with fracture of QR Code labeled acrylic blocks. Conclusion: Within the limitations of the study, by analyzing the results, it was clearly indicated that the QR Code technique was reliable under various depths of acrylic sheet, acid, and heat (370°C). Effectiveness varied in fracture and depended on the level of distortion. This study thus suggests that QR Code is an effective and simpler denture labeling method. PMID:28123284

Expert system for maintenance management of a boiling water reactor power plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hong Shen; Liou, L.W.; Levine, S.

1992-01-01

An expert system code has been developed for the maintenance of two boiling water reactor units in Berwick, Pennsylvania, that are operated by the Pennsylvania Power and Light Company (PP and L). The objective of this expert system code, where the knowledge of experienced operators and engineers is captured and implemented, is to support the decisions regarding which components can be safely and reliably removed from service for maintenance. It can also serve as a query-answering facility for checking the plant system status and for training purposes. The operating and maintenance information of a large number of support systems, whichmore » must be available for emergencies and/or in the event of an accident, is stored in the data base of the code. It identifies the relevant technical specifications and management rules for shutting down any one of the systems or removing a component from service to support maintenance. Because of the complexity and time needed to incorporate a large number of systems and their components, the first phase of the expert system develops a prototype code, which includes only the reactor core isolation coolant system, the high-pressure core injection system, the instrument air system, the service water system, and the plant electrical system. The next phase is scheduled to expand the code to include all other systems. This paper summarizes the prototype code and the design concept of the complete expert system code for maintenance management of all plant systems and components.« less

A model for the release, dispersion and environmental impact of a postulated reactor accident from a submerged commercial nuclear power plant

NASA Astrophysics Data System (ADS)

Bertch, Timothy Creston

1998-12-01

Nuclear power plants are inherently suitable for submerged applications and could provide power to the shore power grid or support future underwater applications. The technology exists today and the construction of a submerged commercial nuclear power plant may become desirable. A submerged reactor is safer to humans because the infinite supply of water for heat removal, particulate retention in the water column, sedimentation to the ocean floor and inherent shielding of the aquatic environment would significantly mitigate the effects of a reactor accident. A better understanding of reactor operation in this new environment is required to quantify the radioecological impact and to determine the suitability of this concept. The impact of release to the environment from a severe reactor accident is a new aspect of the field of marine radioecology. Current efforts have been centered on radioecological impacts of nuclear waste disposal, nuclear weapons testing fallout and shore nuclear plant discharges. This dissertation examines the environmental impact of a severe reactor accident in a submerged commercial nuclear power plant, modeling a postulated site on the Atlantic continental shelf adjacent to the United States. This effort models the effects of geography, decay, particle transport/dispersion, bioaccumulation and elimination with associated dose commitment. The use of a source term equivalent to the release from Chernobyl allows comparison between the impacts of that accident and the postulated submerged commercial reactor plant accident. All input parameters are evaluated using sensitivity analysis. The effect of the release on marine biota is determined. Study of the pathways to humans from gaseous radionuclides, consumption of contaminated marine biota and direct exposure as contaminated water reaches the shoreline is conducted. The model developed by this effort predicts a significant mitigation of the radioecological impact of the reactor accident release with a submerged commercial nuclear power plant. The two box models predict the most of the radio-ecological impact occurs during the first eight days after release. The most significant risk to humans is from consumption of biota. The reduction in impact to humans from a large radioactive release makes the concept worthy of further study.

Active numerical model of human body for reconstruction of falls from height.

PubMed

Milanowicz, Marcin; Kędzior, Krzysztof

2017-01-01

Falls from height constitute the largest group of incidents out of approximately 90,000 occupational accidents occurring each year in Poland. Reconstruction of the exact course of a fall from height is generally difficult due to lack of sufficient information from the accident scene. This usually results in several contradictory versions of an incident and impedes, for example, determination of the liability in a judicial process. In similar situations, in many areas of human activity, researchers apply numerical simulation. They use it to model physical phenomena to reconstruct their real course over time; e.g. numerical human body models are frequently used for investigation and reconstruction of road accidents. However, they are validated in terms of specific road traffic accidents and are considerably limited when applied to the reconstruction of other types of accidents. The objective of the study was to develop an active numerical human body model to be used for reconstruction of accidents associated with falling from height. Development of the model involved extension and adaptation of the existing Pedestrian human body model (available in the MADYMO package database) for the purposes of reconstruction of falls from height by taking into account the human reaction to the loss of balance. The model was developed by using the results of experimental tests of the initial phase of the fall from height. The active numerical human body model covering 28 sets of initial conditions related to various human reactions to the loss of balance was developed. The application of the model was illustrated by using it to reconstruct a real fall from height. From among the 28 sets of initial conditions, those whose application made it possible to reconstruct the most probable version of the incident was selected. The selection was based on comparison of the results of the reconstruction with information contained in the accident report. Results in the form of estimated injuries overlap with the real injuries sustained by the casualty. Copyright © 2016 Elsevier Ireland Ltd. All rights reserved.

A novel grey-fuzzy-Markov and pattern recognition model for industrial accident forecasting

NASA Astrophysics Data System (ADS)

Edem, Inyeneobong Ekoi; Oke, Sunday Ayoola; Adebiyi, Kazeem Adekunle

2017-10-01

Industrial forecasting is a top-echelon research domain, which has over the past several years experienced highly provocative research discussions. The scope of this research domain continues to expand due to the continuous knowledge ignition motivated by scholars in the area. So, more intelligent and intellectual contributions on current research issues in the accident domain will potentially spark more lively academic, value-added discussions that will be of practical significance to members of the safety community. In this communication, a new grey-fuzzy-Markov time series model, developed from nondifferential grey interval analytical framework has been presented for the first time. This instrument forecasts future accident occurrences under time-invariance assumption. The actual contribution made in the article is to recognise accident occurrence patterns and decompose them into grey state principal pattern components. The architectural framework of the developed grey-fuzzy-Markov pattern recognition (GFMAPR) model has four stages: fuzzification, smoothening, defuzzification and whitenisation. The results of application of the developed novel model signify that forecasting could be effectively carried out under uncertain conditions and hence, positions the model as a distinctly superior tool for accident forecasting investigations. The novelty of the work lies in the capability of the model in making highly accurate predictions and forecasts based on the availability of small or incomplete accident data.

Key findings and remaining questions in the areas of core-concrete interaction and debris coolability

DOE PAGES

Farmer, M. T.; Gerardi, C.; Bremer, N.; ...

2016-10-31

The reactor accidents at Fukushima-Dai-ichi have rekindled interest in late phase severe accident behavior involving reactor pressure vessel breach and discharge of molten core melt into the containment. Two technical issues of interest in this area include core-concrete interaction and the extent to which the core debris may be quenched and rendered coolable by top flooding. The OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) programs at Argonne National Laboratory included the conduct of large scale reactor material experiments and associated analysis with the objectives of resolving the ex-vessel debris coolability issue, and to address remaining uncertainties related to long-term two-dimensionalmore » molten core-concrete interactions under both wet and dry cavity conditions. These tests provided a broad database to support accident management planning, as well as the development and validation of models and codes that can be used to extrapolate the experiment results to plant conditions. This paper provides a high level overview of the key experiment results obtained during the program. Finally, a discussion is also provided that describes technical gaps that remain in this area, several of which have arisen based on the sequence of events and operator actions during Fukushima.« less

Thermal Stratification Analysis for Sodium Fast Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schneider, James; Anderson, Mark; Baglietto, Emilio

The sodium fast reactor (SFR) is the most mature reactor concept of all the generation-IV nuclear systems and is a promising reactor design that is currently under development by several organizations. The majority of sodium fast reactor designs utilize a pool type arrangement which incorporates the primary coolant pumps and intermediate heat exchangers within the sodium pool. These components typically protrude into the pool thus reducing the risk and severity of a loss of coolant accidents. To further ensure safe operation under even the most severe transients a more comprehensive understanding of key thermal hydraulic phenomena in this pool ismore » desired. One of the key technology gaps identified for SFR safety is determining the extent and the effects of thermal stratification developing in the pool during postulated accident scenarios such as a protected or unprotected loss of flow incident. In an effort to address these issues, detailed flow models of transient stratification in the pool during an accident can be developed. However, to develop the calculation models, and ensure they can reproduce the underlying physics, highly spatially resolved data is needed. This data can be used in conjunction with advanced computational fluid dynamic calculations to aid in the development of simple reduced dimensional models for systems codes such as SAM and SAS4A/SASSYS-1.« less

Underestimates of unintentional firearm fatalities: comparing Supplementary Homicide Report data with the National Vital Statistics System

PubMed Central

Barber, C; Hemenway, D; Hochstadt, J; Azrael, D

2002-01-01

Objective: A growing body of evidence suggests that the nation's vital statistics system undercounts unintentional firearm deaths that are not self inflicted. This issue was examined by comparing how unintentional firearm injuries identified in police Supplementary Homicide Report (SHR) data were coded in the National Vital Statistics System. Methods: National Vital Statistics System data are based on death certificates and divide firearm fatalities into six subcategories: homicide, suicide, accident, legal intervention, war operations, and undetermined. SHRs are completed by local police departments as part of the FBI's Uniform Crime Reports program. The SHR divides homicides into two categories: "murder and non-negligent manslaughter" (type A) and "negligent manslaughter" (type B). Type B shooting deaths are those that are inflicted by another person and that a police investigation determined were inflicted unintentionally, as in a child killing a playmate after mistaking a gun for a toy. In 1997, the SHR classified 168 shooting victims this way. Using probabilistic matching, 140 of these victims were linked to their death certificate records. Results: Among the 140 linked cases, 75% were recorded on the death certificate as homicides and only 23% as accidents. Conclusion: Official data from the National Vital Statistics System almost certainly undercount firearm accidents when the victim is shot by another person. PMID:12226128

Surface Movement Incidents Reported to the NASA Aviation Safety Reporting System

NASA Technical Reports Server (NTRS)

Connell, Linda J.; Hubener, Simone

1997-01-01

Increasing numbers of aircraft are operating on the surface of airports throughout the world. Airport operations are forecast to grow by more that 50%, by the year 2005. Airport surface movement traffic would therefore be expected to become increasingly congested. Safety of these surface operations will become a focus as airport capacity planning efforts proceed toward the future. Several past events highlight the prevailing risks experienced while moving aircraft during ground operations on runways, taxiways, and other areas at terminal, gates, and ramps. The 1994 St. Louis accident between a taxiing Cessna crossing an active runway and colliding with a landing MD-80 emphasizes the importance of a fail-safe system for airport operations. The following study explores reports of incidents occurring on an airport surface that did not escalate to an accident event. The Aviation Safety Reporting System has collected data on surface movement incidents since 1976. This study sampled the reporting data from June, 1993 through June, 1994. The coding of the data was accomplished in several categories. The categories include location of airport, phase of ground operation, weather /lighting conditions, ground conflicts, flight crew characteristics, human factor considerations, and airport environment. These comparisons and distributions of variables contributing to surface movement incidents can be invaluable to future airport planning, accident prevention efforts, and system-wide improvements.

Key findings and remaining questions in the areas of core-concrete interaction and debris coolability

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farmer, M. T.; Gerardi, C.; Bremer, N.

The reactor accidents at Fukushima-Dai-ichi have rekindled interest in late phase severe accident behavior involving reactor pressure vessel breach and discharge of molten core melt into the containment. Two technical issues of interest in this area include core-concrete interaction and the extent to which the core debris may be quenched and rendered coolable by top flooding. The OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) programs at Argonne National Laboratory included the conduct of large scale reactor material experiments and associated analysis with the objectives of resolving the ex-vessel debris coolability issue, and to address remaining uncertainties related to long-term two-dimensionalmore » molten core-concrete interactions under both wet and dry cavity conditions. These tests provided a broad database to support accident management planning, as well as the development and validation of models and codes that can be used to extrapolate the experiment results to plant conditions. This paper provides a high level overview of the key experiment results obtained during the program. Finally, a discussion is also provided that describes technical gaps that remain in this area, several of which have arisen based on the sequence of events and operator actions during Fukushima.« less

Large Scale Gas Mixing and Stratification Triggered by a Buoyant Plume With and Without Occurrence of Condensation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paladino, Domenico; Auban, Olivier; Zboray, Robert

The benefits of using codes with 3-D capabilities to address safety issues of LWRs will be applicable to both the current generation of nuclear reactors as well to future ALWRs. The phenomena governing the containment response in case of some postulated severe accident scenarios include gas (air, hydrogen, steam) stratification in the containment, gas distribution between containment compartments, wall condensation, etc. These phenomena are driven by buoyant high momentum injection (jets) and/or low momentum injection (plumes). For instance, mixing in the immediate vicinity of the postulated line break is mainly dominated by very high velocity efflux, while low-momentum flows aremore » responsible for most of the transport processes within the containment. A project named SETH is currently in progress under the auspices of 15 OECD countries, with the aim of creating an experimental database suitable to assess the 3-D code capabilities in analyzing key-physical phenomena relevant for LWR safety analysis. This paper describes some results of two SETH tests, performed in the PANDA facility (located at PSI in Switzerland), focusing on plumes flowing near a containment wall. The plumes are generated by injecting a constant amount of steam in one of two interconnected vessels initially filled with air. In one of the two tests the temperature of the injected steam and the initial containment wall and fluid temperatures allowed for condensation during the test. (authors)« less

Once-through integral system (OTIS): Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gloudemans, J R

1986-09-01

A scaled experimental facility, designated the once-through integral system (OTIS), was used to acquire post-small break loss-of-coolant accident (SBLOCA) data for benchmarking system codes. OTIS was also used to investigate the application of the Abnormal Transient Operating Guidelines (ATOG) used in the Babcock and Wilcox (B and W) designed nuclear steam supply system (NSSS) during the course of an SBLOCA. OTIS was a single-loop facility with a plant to model power scale factor of 1686. OTIS maintained the key elevations, approximate component volumes, and loop flow resistances, and simulated the major component phenomena of a B and W raised-loop nuclearmore » plant. A test matrix consisting of 15 tests divided into four categories was performed. The largest group contained 10 tests and was defined to parametrically obtain an extensive set of plant-typical experimental data for code benchmarking. Parameters such as leak size, leak location, and high-pressure injection (HPI) shut-off head were individually varied. The remaining categories were specified to study the impact of the ATOGs (2 tests), to note the effect of guard heater operation on observed phenomena (2 tests), and to provide a data set for comparison with previous test experience (1 test). A summary of the test results and a detailed discussion of Test 220100 is presented. Test 220100 was the nominal or reference test for the parametric studies. This test was performed with a scaled 10-cm/sup 2/ leak located in the cold leg suction piping.« less

Off-road truck-related accidents in U.S. mines

PubMed Central

Dindarloo, Saeid R.; Pollard, Jonisha P.; Siami-Irdemoosa, Elnaz

2016-01-01

Introduction Off-road trucks are one of the major sources of equipment-related accidents in the U.S. mining industries. A systematic analysis of all off-road truck-related accidents, injuries, and illnesses, which are reported and published by the Mine Safety and Health Administration (MSHA), is expected to provide practical insights for identifying the accident patterns and trends in the available raw database. Therefore, appropriate safety management measures can be administered and implemented based on these accident patterns/trends. Methods A hybrid clustering-classification methodology using K-means clustering and gene expression programming (GEP) is proposed for the analysis of severe and non-severe off-road truck-related injuries at U.S. mines. Using the GEP sub-model, a small subset of the 36 recorded attributes was found to be correlated to the severity level. Results Given the set of specified attributes, the clustering sub-model was able to cluster the accident records into 5 distinct groups. For instance, the first cluster contained accidents related to minerals processing mills and coal preparation plants (91%). More than two-thirds of the victims in this cluster had less than 5 years of job experience. This cluster was associated with the highest percentage of severe injuries (22 severe accidents, 3.4%). Almost 50% of all accidents in this cluster occurred at stone operations. Similarly, the other four clusters were characterized to highlight important patterns that can be used to determine areas of focus for safety initiatives. Conclusions The identified clusters of accidents may play a vital role in the prevention of severe injuries in mining. Further research into the cluster attributes and identified patterns will be necessary to determine how these factors can be mitigated to reduce the risk of severe injuries. Practical application Analyzing injury data using data mining techniques provides some insight into attributes that are associated with high accuracies for predicting injury severity. PMID:27620937

Road traffic accidents: Global Burden of Disease study, Brazil and federated units, 1990 and 2015.

PubMed

Ladeira, Roberto Marini; Malta, Deborah Carvalho; Morais, Otaliba Libânio de; Montenegro, Marli de Mesquita Silva; Soares, Adauto Martins; Vasconcelos, Cíntia Honório; Mooney, Meghan; Naghavi, Mohsen

2017-05-01

To describe the global burden of disease due to road traffic accidents in Brazil and federated units in 1990 and 2015. This is an analysis of secondary data from the 2015 Global Burden of Disease study estimates. The following estimates were used: standardized mortality rates and years of life lost by death or disability, potential years of life lost due to premature death, and years of unhealthy living conditions. The Mortality Information System was the main source of death data. Underreporting and redistribution of ill-defined causes and nonspecific codes were corrected. Around 52,326 deaths due to road traffic accidents were estimated in Brazil in 2015. From 1990 to 2015, mortality rates decreased from 36.9 to 24.8/100 thousand people, a reduction of 32.8%. Tocantins and Piauí have the highest mortality risks among the federated units (FU), with 41.7/100 and 33.1/100 thousand people, respectively. They both present the highest rates of potential years of life lost due to premature deaths. Road traffic accidents are a public health problem. Using death- or disability-adjusted life years in studies of these causes is important because there are still no sources to know the magnitude of sequelae, as well as the weight of early deaths. Since its data are updated every year, the Global Burden of Disease study may provide evidence to formulate traffic security and health attention policies, which are guided to the needs of the federated units and of different groups of traffic users.

What are the differences in injury patterns of young and elderly traffic accident fatalities considering death on scene and death in hospital?

PubMed

Heinrich, Daniela; Holzmann, Christopher; Wagner, Anja; Fischer, Anja; Pfeifer, Roman; Graw, Matthias; Schick, Sylvia

2017-07-01

Older traffic participants have higher risks of injury than the population up to 65 years in case of comparable road traffic accidents and further, higher mortality rates at comparable injury severities. Rib fractures as risk factors are currently discussed. However, death on scene is associated with hardly survivable injuries and might not be a matter of neither rib fractures nor age. As 60% of traffic accident fatalities are estimated to die on scene, they are not captured in hospital-based trauma registries and injury patterns remain unknown. Our database comprises 309 road traffic fatalities, autopsied at the Institute of Legal Medicine Munich in 2004 and 2005. Injuries are coded according to Abbreviated Injury Scale, AIS© 2005 update 2008 [1]. Data used for this analysis are age, sex, site of death, site of accident, traffic participation mode, measures of injury severity, and rib fractures. The injury patterns of elderly, aged 65+ years, are compared to the younger ones divided by their site of death. Elderly with death on scene more often show serious thorax injuries and pelvic fractures than the younger. Some hints point towards older fatalities showing less frequently serious abdominal injuries. In hospital, elderly fatalities show lower Injury Severity Scores (ISSs) compared to the younger. The number of rib fractures is significantly higher for the elderly but is not the reason for death. Results show that young and old fatalities have different injury patterns and reveal first hints towards the need to analyze death on scene more in-depth.

Knowledge of and attitude towards road traffic codes among commercial motorcycle riders in Anambra State.

PubMed

Adogu, O U; Ilika, A L

2006-12-01

Road traffic accidents (rtas) represent a major epidemic of non communicable disease in the country and has since escalated with the introduction of the new phenomenon of commercial motorcycle transportation such as is found in the two urban towns of nnewi and Awka of Anambra state, Nigeria. making use of a pre-tested, semi structured, interviewer administered questionnaire, relevant data on socio demographic and motorcycle characteristics were collected from a sample of commercial motorcyclists selected by systematic sampling technique. their knowledge of and attitude towards road traffic and safety codes were elicited. The result showed that the all-male commercial motorcyclists had a mean age of 30+8.9 years. one hundred and seventy six (32.6%) possessed good knowledge of road traffic codes and safety, while 35 (6.5%) exhibited good attitude towards them. both knowledge of and attitude towards traffic codes and safety improved with increase in educational level (p<0.005, p<0.001 respectively). the younger motorcyclists also possessed statistically significant better knowledge of traffic codes than their older counterparts (p<0.025). attitude to traffic codes and safety had no association with age of the motorcyclists (p>0.25). the study has provided useful information on the knowledge of and attitude towards road traffic and safety codes among commercial motorcyclists in nigeria. pursuit of knowledge through formal and informal education should run pari pasu with efforts to improve the nigerian economy in order to ensure a sustainable positive attitudinal change towards road traffic codes and safety among commercial motorcyclists.

40 CFR 68.90 - Applicability.

Code of Federal Regulations, 2010 CFR

2010-07-01

... Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Emergency Response § 68.90 Applicability. (a) Except as provided in... processes shall comply with the requirements of § 68.95. (b) The owner or operator of stationary source...

75 FR 38597 - Qualification of Drivers; Exemption Applications; Diabetes Mellitus

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-02

... hypoglycemia, significant complications, or inability to manage diabetes; also, any involvement in an accident...-2010-0083] Qualification of Drivers; Exemption Applications; Diabetes Mellitus AGENCY: Federal Motor... diabetes mellitus (ITDM) from operating commercial motor vehicles (CMVs) in interstate commerce. The...

Posttest analysis of MIST Test 3109AA using TRAC-PF1/MOD1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Steiner, J.L.; Siebe, D.A.; Boyack, B.E.

This document discusses a posttest calculation and analysis of Multi-loop Integral System Test (MIST) 3109AA as the nominal test for the MIST program. It is a test of a small-break loss-of-coolant accident (SBLOCA) with a scaled 10-cm{sup 2} break in the B1 cold leg. The test exhibited the major post-SBLOCA phenomena, as expected, including depressurization to saturation, intermittent and interrupted loop flow, boiler-condenser mode cooling, refill, and postrefill cooldown. Full high-pressure injection and auxiliary feedwater were available, reactor coolant pumps were not available, and reactor-vessel vent valves and guard heaters were automatically controlled. Constant level control in the steam-generator secondariesmore » was used after steam-generator secondary refill and symmetric steam-generator pressure control was used. We performed the calculation using TRAC-PF1/MODI. Agreement between test data and the calculation was generally reasonable. All major trends and phenomena were correctly predicted. It is believed that the correct conclusions about trends and phenomena will be reached if the code is used in similar applications.« less

Posttest analysis of MIST Test 3109AA using TRAC-PF1/MOD1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Steiner, J.L.; Siebe, D.A.; Boyack, B.E.

This document discusses a posttest calculation and analysis of Multi-loop Integral System Test (MIST) 3109AA as the nominal test for the MIST program. It is a test of a small-break loss-of-coolant accident (SBLOCA) with a scaled 10-cm[sup 2] break in the B1 cold leg. The test exhibited the major post-SBLOCA phenomena, as expected, including depressurization to saturation, intermittent and interrupted loop flow, boiler-condenser mode cooling, refill, and postrefill cooldown. Full high-pressure injection and auxiliary feedwater were available, reactor coolant pumps were not available, and reactor-vessel vent valves and guard heaters were automatically controlled. Constant level control in the steam-generator secondariesmore » was used after steam-generator secondary refill and symmetric steam-generator pressure control was used. We performed the calculation using TRAC-PF1/MODI. Agreement between test data and the calculation was generally reasonable. All major trends and phenomena were correctly predicted. It is believed that the correct conclusions about trends and phenomena will be reached if the code is used in similar applications.« less

Hazardous sign detection for safety applications in traffic monitoring

NASA Astrophysics Data System (ADS)

Benesova, Wanda; Kottman, Michal; Sidla, Oliver

2012-01-01

The transportation of hazardous goods in public streets systems can pose severe safety threats in case of accidents. One of the solutions for these problems is an automatic detection and registration of vehicles which are marked with dangerous goods signs. We present a prototype system which can detect a trained set of signs in high resolution images under real-world conditions. This paper compares two different methods for the detection: bag of visual words (BoW) procedure and our approach presented as pairs of visual words with Hough voting. The results of an extended series of experiments are provided in this paper. The experiments show that the size of visual vocabulary is crucial and can significantly affect the recognition success rate. Different code-book sizes have been evaluated for this detection task. The best result of the first method BoW was 67% successfully recognized hazardous signs, whereas the second method proposed in this paper - pairs of visual words and Hough voting - reached 94% of correctly detected signs. The experiments are designed to verify the usability of the two proposed approaches in a real-world scenario.

Application Side Casing on Open Deck RoRo to Improve Ship Stability

NASA Astrophysics Data System (ADS)

Hasanudin; K. A. P Utama, I.; Chen, Jeng-Horng

2018-03-01

RoRo is a vessel that can transport passengers, cargo, container and cars. Open Car Deck is favourite RoRo Vessel in developing countries due to its small GT, small tax and spacious car deck, but it has poor survival of stability. Many accident involve Open Car Deck RoRo which cause fatalities and victim. In order to ensure the safety of the ship, IMO had applied intact stability criteria IS Code 2008 which adapted from Rahola’s Research, but since 2008 IMO improved criteria become probabilistic damage stability SOLAS 2009. The RoRo type Open Car Deck has wide Breadth (B), small Draft (D) and small freeboard. It has difficulties to satisfy the ship’s stability criteria. Side Casings which has been applied in some RoRo have be known reduce freeboard or improve ship’s safety. In this paper investigated the effect side casings to survival of intact dan damage ship’s stability. Calculation has been conducted for four ships without, existing and full side casings. The investigation results shows that defect stability of Open Deck RoRo can be reduce with fitting side casing.

Modelling of LOCA Tests with the BISON Fuel Performance Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williamson, Richard L; Pastore, Giovanni; Novascone, Stephen Rhead

2016-05-01

BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculationsmore » are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.« less

Simulation of Containment Atmosphere Mixing and Stratification Experiment in the ThAI Facility with a CFD Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Babic, Miroslav; Kljenak, Ivo; Mavko, Borut

2006-07-01

The CFD code CFX4.4 was used to simulate an experiment in the ThAI facility, which was designed for investigation of thermal-hydraulic processes during a severe accident inside a Light Water Reactor containment. In the considered experiment, air was initially present in the vessel, and helium and steam were injected during different phases of the experiment at various mass flow rates and at different locations. The main purpose of the proposed work was to assess the capabilities of the CFD code to reproduce the atmosphere structure with a three-dimensional model, coupled with condensation models proposed by the authors. A three-dimensional modelmore » of the ThAI vessel for the CFX4.4 code was developed. The flow in the simulation domain was modeled as single-phase. Steam condensation on vessel walls was modeled as a sink of mass and energy using a correlation that was originally developed for an integral approach. A simple model of bulk phase change was also included. Calculated time-dependent variables together with temperature and volume fraction distributions at the end of different experiment phases are compared to experimental results. (authors)« less

Anthropotechnological analysis of industrial accidents in Brazil.

PubMed Central

Binder, M. C.; de Almeida, I. M.; Monteau, M.

1999-01-01

The Brazilian Ministry of Labour has been attempting to modify the norms used to analyse industrial accidents in the country. For this purpose, in 1994 it tried to make compulsory use of the causal tree approach to accident analysis, an approach developed in France during the 1970s, without having previously determined whether it is suitable for use under the industrial safety conditions that prevail in most Brazilian firms. In addition, opposition from Brazilian employers has blocked the proposed changes to the norms. The present study employed anthropotechnology to analyse experimental application of the causal tree method to work-related accidents in industrial firms in the region of Botucatu, São Paulo. Three work-related accidents were examined in three industrial firms representative of local, national and multinational companies. On the basis of the accidents analysed in this study, the rationale for the use of the causal tree method in Brazil can be summarized for each type of firm as follows: the method is redundant if there is a predominance of the type of risk whose elimination or neutralization requires adoption of conventional industrial safety measures (firm representative of local enterprises); the method is worth while if the company's specific technical risks have already largely been eliminated (firm representative of national enterprises); and the method is particularly appropriate if the firm has a good safety record and the causes of accidents are primarily related to industrial organization and management (multinational enterprise). PMID:10680249