National Library of Energy BETA

Sample records for transportation accident involving

  1. Accident resistant transport container

    DOE Patents [OSTI]

    Andersen, John A. (Albuquerque, NM); Cole, James K. (Albuquerque, NM)

    1980-01-01

    The invention relates to a container for the safe air transport of plutonium having several intermediate wood layers and a load spreader intermediate an inner container and an outer shell for mitigation of shock during a hypothetical accident.

  2. Temperature of aircraft cargo flame exposure during accidents involving fuel spills

    SciTech Connect (OSTI)

    Mansfield, J.A.

    1993-01-01

    This report describes an evaluation of flame exposure temperatures of weapons contained in alert (parked) bombers due to accidents that involve aircraft fuel fires. The evaluation includes two types of accident, collisions into an alert aircraft by an aircraft that is on landing or take-off, and engine start accidents. Both the B-1B and B-52 alert aircraft are included in the evaluation.

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

    E-Print Network [OSTI]

    Aldrich, David C.

    1979-01-01

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted ...

  4. Fatal accidents involving roof falls in coal mining, 1996--1998

    SciTech Connect (OSTI)

    Not Available

    1999-01-01

    This publication presents information on fatalities involving roof and rib falls that occurred in coal mining operations from January 1996 through December 1998. It includes statistics for the fatalities, as well as abstracts, best practices and illustrations. Conclusion statements have been substituted for best practices where no Title 30 Code of Regulations violations were cited during the accident investigation. From January 1996 through December 1998, 36 miners died at coal operations from accidents classified as roof falls. The information in the report is based on statistics taken from the 1996 through 1998 MSHA Fatal Illustration Programs: Roof Fall Fatalities by District.

  5. Fatal accidents involving roof falls in coal mining, 1996--1998

    SciTech Connect (OSTI)

    NONE

    1999-11-01

    This publication presents information on fatalities involving roof and rib falls that occurred in coal mining operations from January 1996 through December 1998. It includes statistics for the fatalities, as well as abstracts, best practices and illustrations. Conclusion statements have been substituted for best practices where no Title 30 Code of Regulations violations were cited during the accident investigation. From January 1996 through December 1998, 36 miners died at coal operations from accidents classified as roof falls. The information in the report is based on statistics taken from the 1996 through 1998 MSHA Fatal Illustration Programs: Roof Fall Fatalities by District.

  6. Emergency Response to a Transportation Accident Involving Radioactive Material

    Broader source: Energy.gov [DOE]

    The purpose of this User’s Guide is to provide instructors with an overview of the key points covered in the video.  The Student Handout portion of this Guide is designed to assist the instructor...

  7. Decontamination Dressdown at a Transportation Accident Involving Radioactive Material

    Broader source: Energy.gov [DOE]

    The purpose of this User’s Guide is to provide instructors with an overview of the key points covered in the video. The Student Handout portion of this Guide is designed to assist the instructor in...

  8. Trans-oceanic transport of 137 Cs from the Fukushima nuclear accident

    E-Print Network [OSTI]

    Yu, Peter K.N.

    Trans-oceanic transport of 137 Cs from the Fukushima nuclear accident and impact of hypothetical Fukushima-like events of future nuclear plants in Southern China Ka-Ming Wai a,b, , Peter K.N. Yu b. · Observed and modeled 137 Cs concentrations were comparable for the Fukushima accident. · The maximum

  9. Abstract--Petroleum transportation accidents often cause losses of millions of dollars and take human lives. SIGA

    E-Print Network [OSTI]

    Virginia, University of

    Abstract-- Petroleum transportation accidents often cause losses of millions of dollars and take that was designed by the largest petroleum transportation company in Brazil, to help remediate these accidents. SIGA of highly combustible chemical products, the petroleum industry is an extremely high-risk industry in terms

  10. Full-Scale Accident Testing in Support of Used Nuclear Fuel Transportation.

    SciTech Connect (OSTI)

    Durbin, Samuel G.; Lindgren, Eric R.; Rechard, Rob P.; Sorenson, Ken B.

    2014-09-01

    The safe transport of spent nuclear fuel and high-level radioactive waste is an important aspect of the waste management system of the United States. The Nuclear Regulatory Commission (NRC) currently certifies spent nuclear fuel rail cask designs based primarily on numerical modeling of hypothetical accident conditions augmented with some small scale testing. However, NRC initiated a Package Performance Study (PPS) in 2001 to examine the response of full-scale rail casks in extreme transportation accidents. The objectives of PPS were to demonstrate the safety of transportation casks and to provide high-fidelity data for validating the modeling. Although work on the PPS eventually stopped, the Blue Ribbon Commission on America’s Nuclear Future recommended in 2012 that the test plans be re-examined. This recommendation was in recognition of substantial public feedback calling for a full-scale severe accident test of a rail cask to verify evaluations by NRC, which find that risk from the transport of spent fuel in certified casks is extremely low. This report, which serves as the re-assessment, provides a summary of the history of the PPS planning, identifies the objectives and technical issues that drove the scope of the PPS, and presents a possible path for moving forward in planning to conduct a full-scale cask test. Because full-scale testing is expensive, the value of such testing on public perceptions and public acceptance is important. Consequently, the path forward starts with a public perception component followed by two additional components: accident simulation and first responder training. The proposed path forward presents a series of study options with several points where the package performance study could be redirected if warranted.

  11. Sensitivity and uncertainty analysis of accident management strategies involving multiple decisions

    SciTech Connect (OSTI)

    Jae, Moosung; Milici, A.D.; Kastenberg, W.E.; Apostolakis, G.E. (Univ. of California, Los Angeles, CA (United States))

    1993-10-01

    A framework for assessing severe accident management strategies is presented using a new analytical tool, namely, influence diagrams. This framework includes multiple and sequential decisions, sensitivity analysis, and uncertainty propagation, and is applied to a proposed set of strategies for a pressurized water reactor station blackout sequence. The influence diagram associated with these strategies is constructed and evaluated. Each decision variable, represented by a node in the influence diagram, has an uncertainty distribution associated with it. Using the mean value of these distributions, a best estimate assessment is performed, and each strategy is ranked with respect to the conditional frequency of early containment failure (ECF). For the preferred alternative, the sensitivity of the results to values of the input variables is investigated. The sensitivity of the ranking itself is then considered. The distributions of the uncertain variables are also propagated through the influence diagram to rank the alternatives with respect to the uncertainty associated with the calculated conditional frequency of ECF. Finally, the sensitivity of the variance of the output distribution, given the preferred decision alternative, to the uncertainty of the input variables is investigated.

  12. A1D04: Committee on Public Involvement in Transportation State of the Practice: White Paper on Public Involvement

    E-Print Network [OSTI]

    Handy, Susan L.

    -West Gateway Coordinating Council Public involvement is the process of two-way communication between citizen the passage of the Intermodal Surface Transportation Efficiency Act of 1991 (ISTEA), there has been underserved populations. ISTEA's directive was reinforced by the passage of the Transportation Equity Act

  13. Co-relation of Variables Involved in the Occurrence of Crane Accidents in U.S. through Logit Modeling. 

    E-Print Network [OSTI]

    Bains, Amrit Anoop Singh

    2010-10-12

    technical challenges, which has lead to escalation of danger on a construction site. Data from OSHA show that crane accidents have increased rapidly from 2000 to 2004. By analyzing the characteristics of all the crane accident inspections, we can better...

  14. Transportation of Dangerous Goods Anyone involved with the Transportation of Dangerous Goods must be trained. This includes shipping

    E-Print Network [OSTI]

    TDG Transportation of Dangerous Goods Anyone involved with the Transportation of Dangerous Goods must be trained. This includes shipping and receiving. All receiving of Dangerous Goods for the Science in Departmental Offices or labs. How do you know if the package is a dangerous goods shipment? Parcels containing

  15. Accident information needs

    SciTech Connect (OSTI)

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

    1992-12-31

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

  16. Accident information needs

    SciTech Connect (OSTI)

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

    1992-01-01

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

  17. Using Software Development Standards to Analyse Accidents Involving Electrical, Electronic or Programmable Electronic Systems: The Blade Mill Case Study

    E-Print Network [OSTI]

    Johnson, C.W.

    Johnson,C.W. Bowell,M. In C.J. Hayhurst, C.M.Holloway and B. Strauch (eds), Proceedings of the 2nd Workshop on the Investigation and Reporting of Incidents and Accidents, Published as NASA Technical Report NASA/CP-2003-212642 pp 111-128

  18. Pore scale modeling of reactive transport involved in geologic CO2 sequestration

    SciTech Connect (OSTI)

    Kang, Qinjin [Los Alamos National Laboratory; Lichtner, Peter C [Los Alamos National Laboratory; Viswanathan, Hari S [Los Alamos National Laboratory; Abdel-fattah, Amr I [Los Alamos National Laboratory

    2009-01-01

    We apply a multi-component reactive transport lattice Boltzmann model developed in previolls studies to modeling the injection of a C02 saturated brine into various porous media structures at temperature T=25 and 80 C. The porous media are originally consisted of calcite. A chemical system consisting of Na+, Ca2+, Mg2+, H+, CO2(aq), and CI-is considered. The fluid flow, advection and diHusion of aqueous species, homogeneous reactions occurring in the bulk fluid, as weB as the dissolution of calcite and precipitation of dolomite are simulated at the pore scale. The effects of porous media structure on reactive transport are investigated. The results are compared with continuum scale modeling and the agreement and discrepancy are discussed. This work may shed some light on the fundamental physics occurring at the pore scale for reactive transport involved in geologic C02 sequestration.

  19. Causal Reasoning About Aircraft Accidents Peter B. Ladkin

    E-Print Network [OSTI]

    Ladkin, Peter B.

    , rigorous causal reasoning in the analysis of air transportation accidents can improve our understanding

  20. Substrate binding by a bacterial ABC transporter involved in polysaccharide export

    SciTech Connect (OSTI)

    Cuthbertson, Leslie; Kimber, Matthew S.; Whitfield, Chris (Guelph)

    2008-04-02

    ATP-binding-cassette (ABC) transporters are responsible for the export of a wide variety of cell-surface glycoconjugates in both Gram-positive and Gram-negative bacteria. These include the O-antigenic polysaccharide (O-PS) portion of lipopolysaccharide, a crucial virulence determinant in Gram-negative pathogens. O-PSs are synthesized by one of two fundamentally different pathways. Escherichia coli O serotypes O8 and O9a provide the prototype systems for studying O-PS export via ABC transporters. The transporter is composed of the transmembrane component Wzm and the nucleotide-binding component Wzt. Although the N-terminal domain of Wzt is a conventional ABC protein, the C-terminal domain of Wzt (C-Wzt) is a unique structural element that determines the specificity of the transporter for either the O8 or O9a O-PS. We show here that the two domains of Wzt can function when expressed as separate polypeptides; both are essential for export. In vitro, C-Wzt binds its cognate O-PS by recognizing a residue located at the nonreducing end of the polymer. The crystal structure of C-WztO9a is reported here and reveals a {beta} sandwich with an immunoglobulin-like topology that contains the O-PS-binding pocket. Substrate interactions with nucleotide-binding domains have been demonstrated in an ABC exporter previously. However, to our knowledge substrate binding by a discrete, cytoplasmic accessory domain in an extended nucleotide-binding domain polypeptide has not previously been demonstrated. Elucidation of the substrate-recognition system involved in O-PS export provides insight into the mechanism that coordinates polymer biosynthesis, termination, and export.

  1. Consent Versus Consensus - Stakehold Involvement in the Identification of Necessary and Sufficient Transportation

    SciTech Connect (OSTI)

    Fawcett, Ricky Lee; Kramer, George Leroy Jr.

    2003-03-01

    Transportation (DOT) and the Nuclear Regulatory Commission (NRC) to provide for the protection of the public and the environment; historically these regulations have proven quite sufficient. Even so, when the Department of Energy (DOE) makes radioactive materials shipments, that are deemed to be a major federal activity, regulations under the National Environmental Policy Act require that public input on safety issues be sought. This requirement leads to interactions with State, Tribal and local stakeholders that often result in the imposition of extra-regulatory requirements – requirements beyond those prescribed by DOT and NRC regulations. Unfortunately, these additional requirements virtually always increase costs and delay schedules, and usually do so without significantly increasing, and possibly even decreasing overall transportation safety. We believe that this problem arises because of efforts to achieve stakeholder consensus rather than stakeholder consent, where “consensus” connotes universal agreement with all aspects of the program, while “consent”, as used here, is simple agreement with the overall course of action. Gaining consensus entails extensive negotiations because all aspects and requirements of the project must be agreed to by each stakeholder. Gaining consent, on the other hand, requires only that stakeholders be satisfied that the project, as planned, provides adequately for their safety needs. This article addresses the issue of consent versus consensus and proposes a systematic, decision science process for reaching consent. Key steps in this proposed process are early identification and involvement of stakeholders, compilation of their concerns, perceptions, needs, causes, and translation of that information into an appropriate set of “derived requirements.” These derived requirements, along with already-established DOT and NRC regulatory requirements, form the necessary and sufficient conditions for safe transportation and for obtaining stakeholder consent.

  2. Accident management for severe accidents

    SciTech Connect (OSTI)

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

    1988-01-01

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

  3. Type B Accident Investigation Board Report on the February 27, 1998, Shipping Violations Involving the Corehole 8 Project at the Oak Ridge National Laboratory, Oak Ridge, Tennesee

    Broader source: Energy.gov [DOE]

    This report is an independent product of the Type B Investigation Board appointed by James C. Hall, Manager, Oak Ridge Operations Office, U.S. Department of Energy. The Board was appointed to perform a Type B investigation of these incidents and to prepare an investigation report in accordance with DOE Order 225.1A, Accident Investigations.

  4. UNIVERSITY OF TRENTO ACCIDENT INSURANCE

    E-Print Network [OSTI]

    or freezing; · electrocution; · sunstrokes, hot or cold strokes; · lesions caused by muscular efforts. Sanitary transportation expenses (always included) The Company refunds, up to a liability limit of 500 for the transport from the place of the accident to an equipped Healthcare Institute, the transport among Health

  5. Accident Investigations

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1996-04-26

    To prescribe requirements for conducting investigations of certain accidents occurring at Department of Energy (DOE) operations and sites; to improve the environment, safety and health for DOE, contractors, and the public; and to prevent the recurrence of such accidents. Chg 2, 4-26-96

  6. Accident Investigations

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1995-10-26

    To prescribe requirements for conducting investigations of certain accidents occurring at Department of Energy (DOE) operations and sites; to improve the environment , safety and health for DOE, contractors, and the public; and to prevent the recurrence of such accidents. Chg 1, 10-26-95. Cancels parts of DOE 5484.1

  7. Accident management information needs

    SciTech Connect (OSTI)

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

    1990-04-01

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

  8. Accident Investigations

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2011-03-04

    This Order prescribes organizational responsibilities, authorities, and requirements for conducting investigations of certain accidents occurring at DOE sites, facilities, areas, operations, and activities. Supersedes DOE O 225.1A. Cancels DOE G 225.1A-1.

  9. The MacArthur Maze Fire and Roadway Collapse: A "Worst Case Scenario" for Spent Nuclear Fuel Transportation?

    SciTech Connect (OSTI)

    Bajwa, Christopher S.; Easton, Earl P.; Adkins, Harold E.; Cuta, Judith M.; Klymyshyn, Nicholas A.; Suffield, Sarah R.

    2012-07-06

    In 2007, a severe transportation accident occurred near Oakland, California, at the interchange known as the "MacArthur Maze." The accident involved a double tanker truck of gasoline overturning and bursting into flames. The subsequent fire reduced the strength of the supporting steel structure of an overhead interstate roadway causing the collapse of portions of that overpass onto the lower roadway in less than 20 minutes. The US Nuclear Regulatory Commission has analyzed what might have happened had a spent nuclear fuel transportation package been involved in this accident, to determine if there are any potential regulatory implications of this accident to the safe transport of spent nuclear fuel in the United States. This paper provides a summary of this effort, presents preliminary results and conclusions, and discusses future work related to the NRC's analysis of the consequences of this type of severe accident.

  10. Risk Estimation Methodology for Launch Accidents.

    SciTech Connect (OSTI)

    Clayton, Daniel James; Lipinski, Ronald J.; Bechtel, Ryan D.

    2014-02-01

    As compact and light weight power sources with reliable, long lives, Radioisotope Power Systems (RPSs) have made space missions to explore the solar system possible. Due to the hazardous material that can be released during a launch accident, the potential health risk of an accident must be quantified, so that appropriate launch approval decisions can be made. One part of the risk estimation involves modeling the response of the RPS to potential accident environments. Due to the complexity of modeling the full RPS response deterministically on dynamic variables, the evaluation is performed in a stochastic manner with a Monte Carlo simulation. The potential consequences can be determined by modeling the transport of the hazardous material in the environment and in human biological pathways. The consequence analysis results are summed and weighted by appropriate likelihood values to give a collection of probabilistic results for the estimation of the potential health risk. This information is used to guide RPS designs, spacecraft designs, mission architecture, or launch procedures to potentially reduce the risk, as well as to inform decision makers of the potential health risks resulting from the use of RPSs for space missions.

  11. Commercial SNF Accident Release Fractions

    SciTech Connect (OSTI)

    J. Schulz

    2004-11-05

    The purpose of this analysis is to specify and document the total and respirable fractions for radioactive materials that could be potentially released from an accident at the repository involving commercial spent nuclear fuel (SNF) in a dry environment. The total and respirable release fractions are used to support the preclosure licensing basis for the repository. The total release fraction is defined as the fraction of total commercial SNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. Radionuclides are released from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses; this subset of the total release fraction is referred to as the respirable release fraction. Accidents may involve waste forms characterized as: (1) bare unconfined intact fuel assemblies, (2) confined intact fuel assemblies, or (3) canistered failed commercial SNF. Confined intact commercial SNF assemblies at the repository are contained in shipping casks, canisters, or waste packages. Four categories of failed commercial SNF are identified: (1) mechanically and cladding-penetration damaged commercial SNF, (2) consolidated/reconstituted assemblies, (3) fuel rods, pieces, and debris, and (4) nonfuel components. It is assumed that failed commercial SNF is placed into waste packages with a mesh screen at each end (CRWMS M&O 1999). In contrast to bare unconfined fuel assemblies, the container that confines the fuel assemblies could provide an additional barrier for diminishing the total release fraction should the fuel rod cladding breach during an accident. This analysis, however, does not take credit for the additional barrier and establishes only the total release fractions for bare unconfined intact commercial SNF assemblies, which may be conservatively applied to confined intact commercial I SNF assemblies.

  12. Type B Accident Investigation Report of the October 28, 2004...

    Office of Environmental Management (EM)

    of the October 28, 2004, Burn Injuries Sustained During an Office of Secure Transportation Joint Training Exercise at Fort Hunter-Liggett, CA Type B Accident Investigation Report...

  13. Simulation of transportation of low enriched uranium solutions

    SciTech Connect (OSTI)

    Hope, E.P.; Ades, M.J.

    1996-08-01

    A simulation of the transportation by truck of low enriched uranium solutions has been completed for NEPA purposes at the Savannah River Site. The analysis involves three distinct source terms, and establishes the radiological risks of shipment to three possible destinations. Additionally, loading accidents were analyzed to determine the radiological consequences of mishaps during handling and delivery. Source terms were developed from laboratory measurements of chemical samples from low enriched uranium feed materials being stored at SRS facilities, and from manufacturer data on transport containers. The transportation simulations were accomplished over the INTERNET using the DOE TRANSNET system at Sandia National Laboratory. The HIGHWAY 3.3 code was used to analyze routing scenarios, and the RADTRAN 4 code was used to analyze incident free and accident risks of transporting radiological materials. Loading accidents were assessed using the Savannah River Site AXAIR89Q and RELEASE 2 codes.

  14. REAC/TS Radiation Accident Registry: An Overview

    SciTech Connect (OSTI)

    Doran M. Christensen, DO, REAC /TS Associate Director and Staff Physician Becky Murdock, REAC/TS Registry and Health Physics Technician

    2012-12-12

    Over the past four years, REAC/TS has presented a number of case reports from its Radiation Accident Registry. Victims of radiological or nuclear incidents must meet certain dose criteria for an incident to be categorized as an “accident” and be included in the registry. Although the greatest numbers of “accidents” in the United States that have been entered into the registry involve radiation devices, the greater percentage of serious accidents have involved sealed sources of one kind or another. But if one looks at the kinds of accident scenarios that have resulted in extreme consequence, i.e., death, the greater share of deaths has occurred in medical settings.

  15. Utilizing an encroachment probability benefit-cost model to estimate accident reduction factors 

    E-Print Network [OSTI]

    Hayes, Carolyn A

    1997-01-01

    Improving safety on Texas roadways is a major public concern. Over the years, the Texas Department of Transportation and other highway agencies have become interested in reducing society's accident cost while maximizing returns on accident...

  16. Relating geometric design consistency and accident experience on two-lane rural highways 

    E-Print Network [OSTI]

    Glascock, Stephen Wade

    1991-01-01

    to provide more efficient transportation on the one hand, and this same compulsion pushes man to work and devise methods of reducing the probability and severity of accidents. For so long as we have humans at the control of the transport vehicle..., the driver, and the vehicle contribute to the occurrence and severity of accidents. Seldom is information included in accident analyses that addresses all of these factors. Even with more complete accident data bases, however, researchers often are unable...

  17. Interpreting Accident Statistics

    E-Print Network [OSTI]

    Ferreira, Joseph Jr.

    Accident statistics have often been used to support the argument that an abnormally small proportion of drivers account for a large proportion of the accidents. This paper compares statistics developed from six-year data ...

  18. Accident Analysis and Prevention 42 (2010) 364371 Contents lists available at ScienceDirect

    E-Print Network [OSTI]

    Boggess, May M.

    2010-01-01

    number of studies reported age-specific accident rates for heavy vehicles for the spec- trum of driver' shaped curve indicates a higher risk of accident involvement for both younger and older drivers. More

  19. Accidents, engineering and history at NASA: 1967-2003

    E-Print Network [OSTI]

    Brown, Alexander F. G. (Alexander Frederic Garder), 1970-

    2009-01-01

    The manned spaceflight program of the National Aeronautics and Space Administration (NASA) has suffered three fatal accidents: one in the Apollo program and two in the Space Transportation System (the Shuttle). These were ...

  20. COGEMA operating experience in the transportation of spent fuel, nuclear materials and radioactive waste

    SciTech Connect (OSTI)

    Bernard, H. [COGEMA, Velizy-Villacoublay (France)

    1993-12-31

    Were a spent fuel transportation accident to occur, no matter how insignificant, the public outcry could jeopardize both reprocessing operations and power plant operations for utilities that have elected to reprocess their spent fuel. Aware of this possibility, COGEMA has become deeply involved in spent fuel transportation to ensure that it is performed according to the highest standards of transportation safety. Spent fuel transportation is a vital link between the reactor site and the reprocessing plant. This paper gives an overview of COGEMA`s experience in the transportation of spent fuel.

  1. Release fractions for Rocky Flats specific accidents

    SciTech Connect (OSTI)

    Weiss, R.C.

    1992-09-01

    As Rocky Flats and other DOE facilities begin the transition process towards decommissioning, the nature of the scenarios to be studied in safety analysis will change. Whereas the previous emphasis in safety accidents related to production, now the emphasis is shifting to accidents related tc decommissioning and waste management. Accident scenarios of concern at Rocky Flats now include situations of a different nature and different scale than are represented by most of the existing experimental accident data. This presentation will discuss approaches@to use for applying the existing body of release fraction data to this new emphasis. Mention will also be made of ongoing efforts to produce new data and improve the understanding of physical mechanisms involved.

  2. Summary of a workshop on severe accident management for PWRs

    SciTech Connect (OSTI)

    Kastenberg, W.E. [ed.; Apostolakis, G.; Jae, M.; Milici, T.; Park, H.; Xing, L.; Dhir, V.K.; Lim, H.; Okrent, D.; Swider, J.; Yu, D. [California Univ., Los Angeles, CA (United States). Dept. of Mechanical, Aerospace and Nuclear Engineering

    1991-11-01

    Severe accident management can be defined as the use of existing and/or alternative resources, systems and actions to prevent or mitigate a core-melt accident. For each accident sequence and each combination of strategy, there may be several options available to the operator; and each involves phenomenological and operational considerations regarding uncertainty. Operational uncertainty includes operator, system and instrument behavior during severe accidents. During the period May 15--17, 1990 a workshop was held at the University of California, Los Angeles, to address these uncertainties for pressurized water reactors (PWRs). This report contains a summary of the workshop proceedings.

  3. Summary of a workshop on severe accident management for BWRs

    SciTech Connect (OSTI)

    Kastenberg, W.E. [ed.; Apostolakis, G.; Jae, M.; Milici, T.; Park, H.; Xing, L.; Dhir, V.K.; Lim, H.; Okrent, D.; Swider, J.; Yu, D. [California Univ., Los Angeles, CA (United States). Dept. of Mechanical, Aerospace and Nuclear Engineering

    1991-11-01

    Severe accident management can be defined as the use of existing and/or alternative resources, systems and actions to prevent or mitigate a core-melt accident. For each accident sequence and each combination of strategies there may be several options available to the operator; and each involves phenomenological and operational considerations regarding uncertainty. Operational uncertainties include operator, system and instrument behavior during an accident. During the period September 26--28, 1990, a workshop was held at the University of California, Los Angeles, to address these uncertainties for Boiling Water Reactors (BWRs). This report contains a summary of the workshop proceedings.

  4. Uncertainties and severe-accident management

    SciTech Connect (OSTI)

    Kastenberg, W.E. (Univ. of California, Los Angeles (United States))

    1991-01-01

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

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

    SciTech Connect (OSTI)

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

    1993-09-01

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

  6. Accident causation study on roadways with limited sight distance crest vertical curves 

    E-Print Network [OSTI]

    Stoddard, Angela May

    1994-01-01

    reflect the driver and vehicle population currently on the transportation network. An accident causation study was conducted to determine if roadways with limited stopping sight distance present a safety hazard for the transportation network. Rural two...

  7. Transportation scenarios for risk analysis.

    SciTech Connect (OSTI)

    Weiner, Ruth F.

    2010-09-01

    Transportation risk, like any risk, is defined by the risk triplet: what can happen (the scenario), how likely it is (the probability), and the resulting consequences. This paper evaluates the development of transportation scenarios, the associated probabilities, and the consequences. The most likely radioactive materials transportation scenario is routine, incident-free transportation, which has a probability indistinguishable from unity. Accident scenarios in radioactive materials transportation are of three different types: accidents in which there is no impact on the radioactive cargo, accidents in which some gamma shielding may be lost but there is no release of radioactive material, and accident in which radioactive material may potentially be released. Accident frequencies, obtainable from recorded data validated by the U.S. Department of Transportation, are considered equivalent to accident probabilities in this study. Probabilities of different types of accidents are conditional probabilities, conditional on an accident occurring, and are developed from event trees. Development of all of these probabilities and the associated highway and rail accident event trees are discussed in this paper.

  8. Management of severe accidents

    SciTech Connect (OSTI)

    Jankowski, M.W.

    1988-01-01

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

  9. Transportation

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home RoomPreservationBio-Inspired SolarAbout / Transforming Y-12Capacity-Forum Sign InTransportation

  10. Accident motivates scholarship recipient

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    accident, college wasn't on my mind. I was all about sports," said Leyba, who played football, basketball, and ran track at the small Northern New Mexico school. "I had no idea...

  11. Accident Tolerant Fuel Analysis

    SciTech Connect (OSTI)

    Curtis Smith; Heather Chichester; Jesse Johns; Melissa Teague; Michael Tonks; Robert Youngblood

    2014-09-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for risk-informed margins management by improving economics and reliability, and sustaining safety, of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. In order to carry out the R&D needed for the Pathway, the Idaho National Laboratory is performing a series of case studies that will explore methods- and tools-development issues, in addition to being of current interest in their own right. One such study is a comparative analysis of safety margins of plants using different fuel cladding types: specifically, a comparison between current-technology Zircaloy cladding and a notional “accident-tolerant” (e.g., SiC-based) cladding. The present report begins the process of applying capabilities that are still under development to the problem of assessing new fuel designs. The approach and lessons learned from this case study will be included in future Technical Basis Guides produced by the RISMC Pathway. These guides will be the mechanism for developing the specifications for RISMC tools and for defining how plant decision makers should propose and evaluate margin recovery strategies.

  12. Accident tolerant fuel analysis

    SciTech Connect (OSTI)

    Smith, Curtis; Chichester, Heather; Johns, Jesse; Teague, Melissa; Tonks, Michael Idaho National Laboratory; Youngblood, Robert

    2014-09-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for risk-informed margins management by improving economics and reliability, and sustaining safety, of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced ''RISMC toolkit'' that enables more accurate representation of NPP safety margin. In order to carry out the R&D needed for the Pathway, the Idaho National Laboratory is performing a series of case studies that will explore methods- and tools-development issues, in addition to being of current interest in their own right. One such study is a comparative analysis of safety margins of plants using different fuel cladding types: specifically, a comparison between current-technology Zircaloy cladding and a notional ''accident-tolerant'' (e.g., SiC-based) cladding. The present report begins the process of applying capabilities that are still under development to the problem of assessing new fuel designs. The approach and lessons learned from this case study will be included in future Technical Basis Guides produced by the RISMC Pathway. These guides will be the mechanism for developing the specifications for RISMC tools and for defining how plant decision makers should propose and evaluate margin recovery strategies.

  13. Accident motivates scholarship recipient

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home Room News Publications Traditional Knowledge KiosksAbout UsAboutWeb PoliciesAccidentAccident

  14. Severe accident progression perspectives based on IPE results

    SciTech Connect (OSTI)

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

    1996-08-01

    Accident progression perspectives were gathered from the level 2 PRA analyses (the analysis of the accident after core damage has occurred involving the containment performance and the radionuclide release from the containment) described in the IPE submittals. Insights related to the containment failure modes, the releases associated with those failure modes, and the factors responsible for the types of containment failures and release sizes reported were obtained. Complete results are discussed in NUREG-1560 and summarized here.

  15. http://travel.state.gov/travel/about/who/who_1245.html AT THE SCENE OF AN ACCIDENT FORM

    E-Print Network [OSTI]

    Tullos, Desiree

    -737-7252 or risk@oregonstate.edu IMMEDIATELY if this was a serious accident (i.e. ambulance involved, vehicle towed). If this is an OSU Motor Pool vehicle, also call 541-737-4141. If OSU Risk not available, LEAVE MESSAGE. You may also Accident and Insurance Report" (required for accidents with ANY injury, when a vehicle is towed, and

  16. Electrical shock accident investigation

    SciTech Connect (OSTI)

    Not Available

    1994-09-30

    This report documents results of the accident investigation of an electrical shock received by two subcontractor employees on May 13, 1994, at the Pinellas Plant. The direct cause of the electrical shock was worker contact with a cut ``hot`` wire and a grounded panelboard (PPA) enclosure. Workers presumed that all wires in the enclosure were dead at the time of the accident and did not perform thorough Lockout/Tagout (LO/TO). Three contributing causes were identified. First, lack of guidance in the drawing for the modification performed in 1987 allowed the PPA panel to be used as a junction box. The second contributing cause is that Environmental, Safety and Health (ES&H) procedures do not address multiple electrical sources in an enclosure. Finally, the workers did not consider the possibility of multiple electrical sources. The root cause of the electrical shock was the inadequacy of administrative controls, including construction requirement and LO/TO requirements, and subcontractor awareness regarding multiple electrical sources. Recommendations to prevent further reoccurrence of this type of accident include revision of ES&H Standard 2.00, Electrical Safety Program Manual, to document requirements for multiple electrical sources in a single enclosure to specify a thorough visual inspection as part of the voltage check process. In addition, the formality of LO/TO awareness training for subcontractor electricians should be increased.

  17. Analysis of RADTRAN transportation model

    SciTech Connect (OSTI)

    Resnikoff, M. [Radioactive Waste Management Associates, New York, NY (United States)

    1990-10-01

    This paper reviews the RADTRAN computer model, used to estimate the risks of transporting waste to a high-level repository. Ignoring high consequence accidents, human error, sabotage, realistic accident scenarios and recent health effects data leads to an underestimate of potential health effects. Potential economic costs following an accident in a rural area are low compared to a scale-up of the Palomares, Spain clean-up and other estimates. The author states that these costs can vary widely depending on the accident locale and whether certain direct and indirect costs are included.

  18. Safety Analysis: Evaluation of Accident Risks in the Transporation of Hazardous Materials by Truck and Rail at the Savannah River Plant

    SciTech Connect (OSTI)

    Blanchard, A.

    1999-04-15

    This report presents an analysis of the consequences and risks of accidents resulting from hazardous material transportation at the Savannah River Plant.

  19. Accident Investigation of the June 17, 2012, Construction Accident...

    Energy Savers [EERE]

    June 17, 2012, Construction Accident - Structural Steel Collapse at The Over pack Storage Expansion 2 at the Naval Reactors Facility at the Idaho National Laboratory, Idaho Falls,...

  20. Sandia Energy - Severe Accident Modeling

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    nuclear energy efforts by developing risk margins, creating risk assessments, sequencing nuclear reactor accident progression, and performing reactor consequence modeling. Severe...

  1. Emergency response to a highway accident in Springfield, Massachusetts, on December 16, 1991

    SciTech Connect (OSTI)

    Not Available

    1992-06-01

    On December 16, 1991, a truck carrying unirradiated (fresh) nuclear fuel was involved in an accident on US Interstate 91, in Springfield, Massachusetts. This report describes the emergency response measures undertaken by local, State, Federal, and private parties. The report also discusses ``lessons learned`` from the response to the accident and suggests areas where improvements might be made.

  2. Emergency response to a highway accident in Springfield, Massachusetts, on December 16, 1991

    SciTech Connect (OSTI)

    Not Available

    1992-06-01

    On December 16, 1991, a truck carrying unirradiated (fresh) nuclear fuel was involved in an accident on US Interstate 91, in Springfield, Massachusetts. This report describes the emergency response measures undertaken by local, State, Federal, and private parties. The report also discusses lessons learned'' from the response to the accident and suggests areas where improvements might be made.

  3. CNG buses fire safety: learnings from recent accidents in France and Germany

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    CNG buses fire safety: learnings from recent accidents in France and Germany Lionel PERRETTE Saarland Holding, Sulzbach Saar/ Germany ABSTRACT The use of CNG in bus and private vehicles is growing steadily. Recent fire accidents involving CNG buses have shown that tanks may explode though compliant

  4. Industry program needed for nuclear accident management

    SciTech Connect (OSTI)

    Klopp, G.T

    1989-05-01

    This paper addresses the need for a management program for nuclear power accidents. According to the author, the tools and technology for severe accident management exist. The need for a clear, realistic definition of nuclear accident program requirements is discussed.

  5. Fuel removal, transport, and storage

    SciTech Connect (OSTI)

    Reno, H.W.

    1986-01-01

    The March 1979 accident at Unit 2 of the Three Mile Island Nuclear Power Station (TMI-2) which damaged the core of the reactor resulted in numerous scientific and technical challenges. Some of those challenges involve removing the core debris from the reactor, packaging it into canisters, loading canisters into a rail cask, and transporting the debris to the Idaho National Engineering Laboratory (INEL) for storage, examination, and preparation for final disposal. This paper highlights how some challenges were resolved, including lessons learned and benefits derived therefrom. Key to some success at TMI was designing, testing, fabricating, and licensing two rail casks, which each provide double containment of the damaged fuel. 10 refs., 12 figs.

  6. First Responders and Criticality Accidents

    SciTech Connect (OSTI)

    Valerie L. Putman; Douglas M. Minnema

    2005-11-01

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

  7. Method of assessing severe accident management strategies

    SciTech Connect (OSTI)

    Kastenberg, W.E.; Apostolakis, G.; Dhir, V.K.; Okrent, D.; Jae, M.; Lim, H.; Milici, T.; Park, H.; Swider, J.; Xing, L.; Yu, D. (Univ. of California, Los Angeles (United States))

    1991-01-01

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

  8. Hanford waste tank bump accident analysis

    SciTech Connect (OSTI)

    MALINOVIC, B.

    2003-03-21

    This report provides a new evaluation of the Hanford tank bump accident analysis (HNF-SD-Wh4-SAR-067 2001). The purpose of the new evaluation is to consider new information and to support new recommendations for final safety controls. This evaluation considers historical data, industrial failure modes, plausible accident scenarios, and system responses. A tank bump is a postulated event in which gases, consisting mostly of water vapor, are suddenly emitted from the waste and cause tank headspace pressurization. A tank bump is distinguished from a gas release event in two respects: First, the physical mechanism for release involves vaporization of locally superheated liquid, and second, gases emitted to the head space are not flammable. For this reason, a tank bump is often called a steam bump. In this report, even though non-condensible gases may be considered in bump models, flammability and combustion of emitted gases are not. The analysis scope is safe storage of waste in its current configuration in single-shell tanks (SSTs) and double-shell tanks (DSTs). The analysis considers physical mechanisms for tank bump to formulate criteria for bump potential, application of the criteria to the tanks, and accident analysis of bump scenarios. The result of consequence analysis is the mass of waste released from tanks for specific scenarios where bumps are credible; conversion to health consequences is performed elsewhere using standard Hanford methods (Cowley et al. 2000). The analysis forms a baseline for future extension to consider waste retrieval.

  9. Iodine behavior in containment under LWR accident conditions

    SciTech Connect (OSTI)

    Wisbey, S.J.; Beahm, E.C.; Shockley, W.E.; Wang, Y.M.

    1986-01-01

    The description of containment iodine behavior in reactor accident sequences requires an understanding of iodine volatility effects, deposition and revaporization/resuspension (from surfaces and aerosols), chemical changes between species, and mass transport. The experimental work in this program has largely centered on the interactions of iodine in or with water pools. The formation of volatile iodine, as I/sub 2/ or organic iodides, is primarily dependent on radiation and solution pH. Lower pH results in increased formation of volatile iodine species; thus, for example, a pH of 3.05 resulted in a conversion of I/sup -/ to I/sub 2/ that was more than two orders of magnitude greater than tests run at pH 6.1 or 6.8. The formation or organic iodides involving water pools has been linked to the presence of iodine as I/sub 2/, the solution/gas contact, and to the type of organic material.

  10. Effects of the MacArthur Maze Fire and Roadway Collapse on a Spent Nuclear Fuel Transportation Package

    SciTech Connect (OSTI)

    Bajwa, Christopher S.; Easton, Earl P.; Adkins, Harold E.; Cuta, Judith M.; Klymyshyn, Nicholas A.; Suffield, Sarah R.

    2011-03-03

    In 2007, a severe transportation accident occurred near Oakland, California, on a section of Interstate 880 known as the "MacArthur Maze," involving a tractor trailer carrying gasoline which impacted an overpass support column and burst into flames. The subsequent fire caused the collapse of portions of the Interstate 580 overpass onto the remains of the tractor-trailer in less than 20 minutes, due to a reduction of strength in the structural steel exposed to the fire. The US Nuclear Regulatory Commission is in the process of examining the impacts of this accident on the performance of a spent nuclear fuel transportation package, using detailed analysis models, in order to determine the potential regulatory implications related to the safe transport of spent nuclear fuel in the United States. This paper will provide a summary of this ongoing effort and present some preliminary results and conclusions.

  11. Parent Involvement 

    E-Print Network [OSTI]

    Howard, Jeff W.

    2005-05-10

    To be successful, a 4-H program must have parent involvement. Although 4-H leaders and Extension agents may interest young people in becoming members, they need the parents' goodwill and support to keep them interested, ...

  12. Computerized Accident Incident Reporting System | Department...

    Office of Environmental Management (EM)

    Incident Reporting System Computerized Accident Incident Reporting System CAIRS Database The Computerized AccidentIncident Reporting System is a database used to collect and...

  13. Benchmarking of finite element codes for radioactive material transportation packages

    SciTech Connect (OSTI)

    Ammerman, D.J.

    1996-10-01

    The increased power of computers and computer codes makes the use of nonlinear dynamic finite element analyses attractive for use as a tool used in the design and certification of radioactive material transportation packages. For this analysis technique to be acceptable it must be demonstrated. The technique has the ability to accurately capture the response of the packages to accident environments required by the regulations. The best method of demonstrating this ability is via a series of benchmark analyses. In this paper three benchmark problems involving significant inelastic deformations will be discussed. One of the problems has been analyzed using many different finite element codes. The other two problems involve comparison of finite element calculations to the results form physical tests. The ability of the finite element method to accurately capture the response in these three problems indicates the method should be acceptable for radioactive material transportation package design and certification.

  14. Bibliography [AAIC96] Japan Aircraft Accident Investigation Commission, Ministry of Trans-

    E-Print Network [OSTI]

    Ladkin, Peter B.

    .S. National Transportation Safety Board. Factual data [in the cali accident]. Available from [Lad], Dec 1995. [Boa96] US National Transportation Safety Board. Safety recommendation (in- cluding a-96-90 through a, Chicago, 1996. [HC96] G.E. Hughes and M. Creswell. A New Introduction to Modal Logic. Routledge, London

  15. Going the Distance? NRC's Response to the National Academy of Science's Transportation Study

    SciTech Connect (OSTI)

    Easton, E.P.; Bajwa, C.S.

    2008-07-01

    In February 2006, the National Academy of Sciences (NAS) published the results of a 3 1/2-year study, titled Going the Distance, that examined the safety of transporting spent nuclear fuel (SNF) and high level waste (HLW) in the United States. NAS initiated this study to address what it perceived to be a national need for an independent, objective, and authoritative analysis of SNF and HLW transport in the United States. The study was co-sponsored by the U.S. Nuclear Regulatory Commission (NRC), the U.S. Department of Energy (DOE), the U.S. Department of Transportation (DOT), the Electric Power Research Institute and the National Cooperative Highway Research Program. This paper addresses some of the recommendations made in the NAS study related to the performance of SNF transportation casks in long duration fires, the use of full-scale package testing, and the need for an independent review of transportation security prior to the commencement of large scale shipping campaigns to an interim storage site or geologic repository. In conclusion: The NRC believes that the current regulations in 10 CFR Part 71 for the design of SNF and HLW transportation packages provide a very high level of protection to the public for very severe accidents and credible threat scenarios. As recommended by the NAS study, additional studies of accidents involving severe fires have been completed. These studies have confirmed that spent fuel casks would be expected to withstand very severe fires without the release of any fission products from the spent fuel. Additionally, changes in rail operating procedures such as the use of dedicated trains and prohibition on the co-location of SNF and flammable liquids in rail tunnels can further reduce the already low probability of severe rail accident fires involving SNF and HLW. (authors)

  16. Summary of the SRS Severe Accident Analysis Program, 1987--1992

    SciTech Connect (OSTI)

    Long, T.A.; Hyder, M.L.; Britt, T.E.; Allison, D.K.; Chow, S.; Graves, R.D.; DeWald, A.B. Jr.; Monson, P.R. Jr.; Wooten, L.A.

    1992-11-01

    The Severe Accident Analysis Program (SAAP) is a program of experimental and analytical studies aimed at characterizing severe accidents that might occur in the Savannah River Site Production Reactors. The goals of the Severe Accident Analysis Program are: To develop an understanding of severe accidents in SRS reactors that is adequate to support safety documentation for these reactors, including the Safety Analysis Report (SAR), the Probabilistic Risk Assessment (PRA), and other studies evaluating the safety of reactor operation; To provide tools and bases for the evaluation of existing or proposed safety related equipment in the SRS reactors; To provide bases for the development of accident management procedures for the SRS reactors; To develop and maintain on the site a sufficient body of knowledge, including documents, computer codes, and cognizant engineers and scientists, that can be used to authoritatively resolve questions or issues related to reactor accidents. The Severe Accident Analysis Program was instituted in 1987 and has already produced a substantial amount of information, and specialized calculational tools. Products of the Severe Accident Analysis Program (listed in Section 9 of this report) have been used in the development of the Safety Analysis Report (SAR) and the Probabilistic Risk Assessment (PRA), and in the development of technical specifications for the SRS reactors. A staff of about seven people is currently involved directly in the program and in providing input on severe accidents to other SRS activities.

  17. Use of inelastic analysis to determine the response of packages to puncture accidents

    SciTech Connect (OSTI)

    Ammerman, D.J.; Ludwigsen, J.S.

    1996-08-01

    The accurate analytical determination of the response of radioactive material transportation packages to the hypothetical puncture accident requires inelastic analysis techniques. Use of this improved analysis method recudes the reliance on empirical and approximate methods to determine the safety for puncture accidents. This paper will discuss how inelastic analysis techniques can be used to determine the stresses, strains and deformations resulting from puncture accidents for thin skin materials with different backing materials. A method will be discussed to assure safety for all of these types of packages.

  18. Graphite Oxidation Simulation in HTR Accident Conditions

    SciTech Connect (OSTI)

    El-Genk, Mohamed

    2012-10-19

    Massive air and water ingress, following a pipe break or leak in steam-generator tubes, is a design-basis accident for high-temperature reactors (HTRs). Analysis of these accidents in both prismatic and pebble bed HTRs requires state-of-the-art capability for predictions of: 1) oxidation kinetics, 2) air ?helium gas mixture stratification and diffusion into the core following the depressurization, 3) transport of multi-species gas mixture, and 4) graphite corrosion. This project will develop a multi-dimensional, comprehensive oxidation kinetics model of graphite in HTRs, with diverse capabilities for handling different flow regimes. The chemical kinetics/multi-species transport model for graphite burning and oxidation will account for temperature-related changes in the properties of graphite, oxidants (O2, H2O, CO), reaction products (CO, CO2, H2, CH4) and other gases in the mixture (He and N2). The model will treat the oxidation and corrosion of graphite in geometries representative of HTR core component at temperatures of 900°C or higher. The developed chemical reaction kinetics model will be user-friendly for coupling to full core analysis codes such as MELCOR and RELAP, as well as computational fluid dynamics (CFD) codes such as CD-adapco. The research team will solve governing equations for the multi-dimensional flow and the chemical reactions and kinetics using Simulink, an extension of the MATLAB solver, and will validate and benchmark the model's predictions using reported experimental data. Researchers will develop an interface to couple the validated model to a commercially available CFD fluid flow and thermal-hydraulic model of the reactor , and will perform a simulation of a pipe break in a prismatic core HTR, with the potential for future application to a pebble-bed type HTR.

  19. Environment/Health/Safety (EHS): Monthly Accident Statistics

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Personal Protective Equipment (PPE) Injury Review & Analysis Worker Safety and Health Program: PUB-3851 Monthly Accident Statistics Latest Accident Statistics Accident...

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

    SciTech Connect (OSTI)

    Sauvage, E. C.; Prior, R.; Coffey, K. [AREVA, FRAMATOME-ANP SAS, Paris, 92084 La Defense (France); Mazurkiewicz, S. M. [AREVA, FRAMATOME-ANP Inc, Lynchburg, VA 24506-0935 (United States)

    2006-07-01

    There is a recognized need to provide nuclear power plant technical staff with structured guidance for response to a potential severe accident condition involving core damage and potential release of fission products to the environment. Over the past ten years, many plants worldwide have implemented such guidance for their emergency technical support center teams either by following one of the generic approaches, or by developing fully independent approaches. There are many lessons to be learned from the experience of the past decade, in developing, implementing, and validating severe accident management guidance. Also, though numerous basic approaches exist which share common principles, there are differences in the methodology and application of the guidelines. AREVA/Framatome-ANP is developing an optimized approach to severe accident management guidance in a project called OSSA ('Operating Strategies for Severe Accidents'). There are still numerous operating power plants which have yet to implement severe accident management programs. For these, the option to use an updated approach which makes full use of lessons learned and experience, is seen as a major advantage. Very few of the current approaches covers all operating plant states, including shutdown states with the primary system closed and open. Although it is not necessary to develop an entirely new approach in order to add this capability, the opportunity has been taken to develop revised full scope guidance covering all plant states in addition to the fuel in the fuel building. The EPR includes at the design phase systems and measures to minimize the risk of severe accident and to mitigate such potential scenarios. This presents a difference in comparison with existing plant, for which severe accidents where not considered in the design. Thought developed for all type of plants, OSSA will also be applied on the EPR, with adaptations designed to take into account its favourable situation in that field. This revised approach will incorporate a number of new features which will simplify and streamline the guidance material while ensuring comprehensive guidance for response to any severe accident. Examples of such features include : - Identification of severe accident challenges based on plant specific studies. - Revision of the split of responsibilities between operations and technical support center staff. - Fixed setpoint entry conditions, ensuring that the transition from emergency procedures takes place at a consistent core/fuel condition (regardless of scenario), and which fixes the time window available to attempt ultimate preventive measures. - A safety function concept for monitoring plant conditions (in the control room). - An integrated graphic-based diagnostic tool including entry condition, challenge prioritization, and exit condition monitoring to be used by the technical support team. This paper describes the basic features of OSSA, and project status. (authors)

  1. Accident Conditions versus Regulatory Test for NRC-Approved UF6 Packages

    SciTech Connect (OSTI)

    MILLS, G. SCOTT; AMMERMAN, DOUGLAS J.; LOPEZ, CARLOS

    2003-01-01

    The Nuclear Regulatory Commission (NRC) approves new package designs for shipping fissile quantities of UF{sub 6}. Currently there are three packages approved by the NRC for domestic shipments of fissile quantities of UF{sub 6}: NCI-21PF-1; UX-30; and ESP30X. For approval by the NRC, packages must be subjected to a sequence of physical tests to simulate transportation accident conditions as described in 10 CFR Part 71. The primary objective of this project was to relate the conditions experienced by these packages in the tests described in 10 CFR Part 71 to conditions potentially encountered in actual accidents and to estimate the probabilities of such accidents. Comparison of the effects of actual accident conditions to 10 CFR Part 71 tests was achieved by means of computer modeling of structural effects on the packages due to impacts with actual surfaces, and thermal effects resulting from test and other fire scenarios. In addition, the likelihood of encountering bodies of water or sufficient rainfall to cause complete or partial immersion during transport over representative truck routes was assessed. Modeled effects, and their associated probabilities, were combined with existing event-tree data, plus accident rates and other characteristics gathered from representative routes, to derive generalized probabilities of encountering accident conditions comparable to the 10 CFR Part 71 conditions. This analysis suggests that the regulatory conditions are unlikely to be exceeded in real accidents, i.e. the likelihood of UF{sub 6} being dispersed as a result of accident impact or fire is small. Moreover, given that an accident has occurred, exposure to water by fire-fighting, heavy rain or submersion in a body of water is even less probable by factors ranging from 0.5 to 8E-6.

  2. Analysis of accidents during flashing operations 

    E-Print Network [OSTI]

    Obermeyer, Michael Edward

    1993-01-01

    . In this thesis, the relative impacts of flashing signal operation versus regular signal operation were evaluated in several cities and towns in the State of Texas. Analysis were conducted to determine whether an increase in accidents and accident severity...

  3. Specific topics in severe accident management

    SciTech Connect (OSTI)

    Meyer, J.F.; Chung, D.T.; Panciera, V.W.; Traver, L.E.; Humphries, D.S. (SCIENTECH, Inc., Rockville, MD (USA))

    1991-05-01

    This report examines five topical areas of concern to severe accident management. These areas are as follows: Human Factors, Accident Management During Shutdown, Information Needs, Long-term Implications, and Uncertainties. The objective of this report is to assist the NRC in performing its research function and to provide guidance to the industry on accident management strategies, as well as to accident management programs in general. 47 refs., 4 figs., 5 tabs.

  4. Office of Inspector General report on inspection of selected issues regarding the Department of Energy accident investigation program

    SciTech Connect (OSTI)

    NONE

    1999-04-01

    One method used by the Department of Energy (DOE) to promote worker safety is through the Department`s accident investigation program. The objectives of the program are, among other things, to enhance safety and health of employees, to prevent the recurrence of accidents, and to reduce accident fatality rates and promote a downward trend in the number and severity of accidents. The Assistant Secretary, Office of Environment, Safety and Health (EH), through the EH Office of the Deputy Assistant Secretary for Oversight, is responsible for implementation of the Department`s accident investigation program. As part of the inspection, the authors reviewed an April 1997 EH accident investigation report regarding an accident involving a Lockheed Martin Energy Systems (LMES) welder, who suffered fatal burns when his clothing caught fire while he was using a cutting torch at the Oak Ridge K-25 Site. They also reviewed reports of other accident investigations conducted by EH and DOE field organizations. Based on the review of these reports, the authors identified issues concerning the adequacy of the examination and reporting by accident investigation boards of specific management systems and organizations as a possible accident root cause. The inspection also identified issues concerning worker safety that they determined required immediate management attention, such as whether occurrences were being reported in the appropriate management systems and whether prompt consideration was being given to implementing revisions of national standards when the revisions increased worker safety.

  5. The Nevada railroad system: Physical, operational, and accident characteristics

    SciTech Connect (OSTI)

    1991-09-01

    This report provides a description of the operational and physical characteristics of the Nevada railroad system. To understand the dynamics of the rail system, one must consider the system`s physical characteristics, routing, uses, interactions with other systems, and unique operational characteristics, if any. This report is presented in two parts. The first part is a narrative description of all mainlines and major branchlines of the Nevada railroad system. Each Nevada rail route is described, including the route`s physical characteristics, traffic type and volume, track conditions, and history. The second part of this study provides a more detailed analysis of Nevada railroad accident characteristics than was presented in the Preliminary Nevada Transportation Accident Characterization Study (DOE, 1990).

  6. Accident management for indian pressurized heavy water reactors

    SciTech Connect (OSTI)

    Hajela, S.; Grover, R.; Ghadge, S.G.; Bajaj, S.S. [Directorate of Safety, Nuclear Power Corporation of India Limited Nabhikiya Urja Bhawan, Anushakti Nagar, Mumbai-400 094 (India)

    2006-07-01

    Indian nuclear power program as of now is mainly based on Pressurized Heavy Water Reactors (PHWRs). Operating Procedures for normal power operation and Emergency Operating Procedures for operational transients and accidents within design basis exist for all Indian PHWRs. In addition, on-site and off-site emergency response procedures are also available for these NPPs. The guidelines needed for severe accidents mitigation are now formally being documented for Indian PHWRs. Also, in line with International trend of having symptom based emergency handling, the work is in advanced stage for preparation of symptom-based emergency operating procedures. Following a plant upset condition; a number of alarms distributed in different information systems appear in the control room to aid operator to identify the nature of the event. After identifying the event, appropriate intervention in the form of event based emergency operating procedure is put into use by the operating staff. However, if the initiating event cannot be unambiguously identified or after the initial event some other failures take place, then the selected event based emergency operating procedure will not be optimal. In such a case, reactor safety is ensured by monitoring safety functions (depicted by selected plant parameters grouped together) throughout the event handling so that the barriers to radioactivity release namely, fuel and fuel cladding, primary heat transport system integrity and containment remain intact. Simultaneous monitoring of all these safety functions is proposed through status trees and this concept will be implemented through a computer-based system. For beyond design basis accidents, event sequences are identified which may lead to severe core damage. As part of this project, severe accident mitigation guidelines are being finalized for the selected event sequences. The paper brings out the details of work being carried out for Indian PHWRs for symptom based event handling and severe accident management. (authors)

  7. Accident analysis and DOE criteria

    SciTech Connect (OSTI)

    Graf, J.M.; Elder, J.C.

    1982-01-01

    In analyzing the radiological consequences of major accidents at DOE facilities one finds that many facilities fall so far below the limits of DOE Order 6430 that compliance is easily demonstrated by simple analysis. For those cases where the amount of radioactive material and the dispersive energy available are enough for accident consequences to approach the limits, the models and assumptions used become critical. In some cases the models themselves are the difference between meeting the criteria or not meeting them. Further, in one case, we found that not only did the selection of models determine compliance but the selection of applicable criteria from different chapters of Order 6430 also made the difference. DOE has recognized the problem of different criteria in different chapters applying to one facility, and has proceeded to make changes for the sake of consistency. We have proposed to outline the specific steps needed in an accident analysis and suggest appropriate models, parameters, and assumptions. As a result we feed DOE siting and design criteria will be more fairly and consistently applied.

  8. Review of U.S. and European Regional Modeling Studies of Policies Intended to Reduce Transportation Greenhouse Gas Emissions

    E-Print Network [OSTI]

    Johnston, Robert A.

    2008-01-01

    emissions of pollutants and greenhouse gases are possible (accidents, noise, and greenhouse gas emissions. We reviewto Reduce Transportation Greenhouse Gas Emissions Robert A.

  9. ARGX-87: Accident Response Group Exercise, 1987: A Broken Arrow mini exercise. [Training

    SciTech Connect (OSTI)

    Schuld, E.P.; Cruff, D.F.

    1987-07-01

    A Broken Arrow mini exercise dubbed ''Accident Response Group Exercise - 1987'' (ARGX-87) was conducted on June 1, 1987 at the Lawrence Livermore National Laboratory (LLNL) and Sandia National Laboratories, Livermore (SNLL). The exercise started at 0445 PDT with a call from the Department of Energy (DOE) - EOC in Washington, DC, to the Albuquerque Operations (AL - ) - EOC. AL, in turn, called the Laboratory off-hour emergency number (Fire Dispatcher), who called the Laboratory Emergency Duty Officer (LEDO). The LEDO then contacted the Accident Response Group (ARG) Senior Scientific Advisor. Calls were placed to assemble appropriate members of the ARG in the ALERT Center. No phone number for SNLL was available at the Albuquerque Operations EOC, so a controller injected a message to SNLL to get them involved in the exercise. The messages received at the Laboratory identified the Air Force line item weapon system involved in the accident and the accident location. As people arrived at the ALERT Center they began discussing the details of the accident. They also started working the deployment logistics and other issues. Travel arrangements for the HOT SPOT equipment and ARG personnel were made for immediate deployment to the accident site in North Dakota. The exercise was terminated at 0840 as planned. While certain procedural deficiencies were noted, the exercise was considered a valuable learning experience. The results and observations from this experience will be used to refine the operating procedures and the training program.

  10. Synthesis of VERCORS and Phebus data in severe accident codes and applications.

    SciTech Connect (OSTI)

    Gauntt, Randall O.

    2010-04-01

    The Phebus and VERCORS data have played an important role in contemporary understanding and modeling of fission product release and transport from damaged LWR fuel. The data from these test programs have allowed improvement of MELCOR modeling of release and transport processes for both low enrichment uranium fuel as well as high burnup and MOX fuels. The following paper describes the derivation, testing and incorporation of improved radionuclide release models into the MELCOR severe accident code.

  11. DOE Accident Prevention and Investigation Program | Department...

    Broader source: Energy.gov (indexed) [DOE]

    for improvement of our integrated safety management system. The techniques and tools utilized in the investigation of "accidents" can be valuable in looking at leading...

  12. Transportation of Hazardous Evidentiary Material.

    SciTech Connect (OSTI)

    Osborn, Douglas.

    2005-06-01

    This document describes the specimen and transportation containers currently available for use with hazardous and infectious materials. A detailed comparison of advantages, disadvantages, and costs of the different technologies is included. Short- and long-term recommendations are also provided.3 DraftDraftDraftExecutive SummaryThe Federal Bureau of Investigation's Hazardous Materials Response Unit currently has hazardous material transport containers for shipping 1-quart paint cans and small amounts of contaminated forensic evidence, but the containers may not be able to maintain their integrity under accident conditions or for some types of hazardous materials. This report provides guidance and recommendations on the availability of packages for the safe and secure transport of evidence consisting of or contaminated with hazardous chemicals or infectious materials. Only non-bulk containers were considered because these are appropriate for transport on small aircraft. This report will addresses packaging and transportation concerns for Hazardous Classes 3, 4, 5, 6, 8, and 9 materials. If the evidence is known or suspected of belonging to one of these Hazardous Classes, it must be packaged in accordance with the provisions of 49 CFR Part 173. The anthrax scare of several years ago, and less well publicized incidents involving unknown and uncharacterized substances, has required that suspicious substances be sent to appropriate analytical laboratories for analysis and characterization. Transportation of potentially hazardous or infectious material to an appropriate analytical laboratory requires transport containers that maintain both the biological and chemical integrity of the substance in question. As a rule, only relatively small quantities will be available for analysis. Appropriate transportation packaging is needed that will maintain the integrity of the substance, will not allow biological alteration, will not react chemically with the substance being shipped, and will otherwise maintain it as nearly as possible in its original condition.The recommendations provided are short-term solutions to the problems of shipping evidence, and have considered only currently commercially available containers. These containers may not be appropriate for all cases. Design, testing, and certification of new transportation containers would be necessary to provide a container appropriate for all cases.Table 1 provides a summary of the recommendations for each class of hazardous material.Table 1: Summary of RecommendationsContainerCost1-quart paint can with ArmlockTM seal ringLabelMaster(r)%242.90 eachHazard Class 3, 4, 5, 8, or 9 Small ContainersTC Hazardous Material Transport ContainerCurrently in Use4 DraftDraftDraftTable 1: Summary of Recommendations (continued)ContainerCost55-gallon open or closed-head steel drumsAll-Pak, Inc.%2458.28 - %2473.62 eachHazard Class 3, 4, 5, 8, or 9 Large Containers95-gallon poly overpack LabelMaster(r)%24194.50 each1-liter glass container with plastic coatingLabelMaster(r)%243.35 - %243.70 eachHazard Class 6 Division 6.1 Poisonous by Inhalation (PIH) Small ContainersTC Hazardous Material Transport ContainerCurrently in Use20 to 55-gallon PIH overpacksLabelMaster(r)%24142.50 - %24170.50 eachHazard Class 6 Division 6.1 Poisonous by Inhalation (PIH) Large Containers65 to 95-gallon poly overpacksLabelMaster(r)%24163.30 - %24194.50 each1-liter transparent containerCurrently in UseHazard Class 6 Division 6.2 Infectious Material Small ContainersInfectious Substance ShipperSource Packaging of NE, Inc.%24336.00 eachNone Commercially AvailableN/AHazard Class 6 Division 6.2 Infectious Material Large ContainersNone Commercially Available N/A5

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

    SciTech Connect (OSTI)

    Baek J.; Diamond D.; Cuadra, A.; Hanson, A.L.; Cheng, L-Y.; Brown, N.R.

    2012-09-30

    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 model 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.

  14. SILENE Benchmark Critical Experiments for Criticality Accident Alarm Systems

    SciTech Connect (OSTI)

    Miller, Thomas Martin; Reynolds, Kevin H.

    2011-01-01

    In October 2010 a series of benchmark experiments was conducted at the Commissariat a Energie Atomique et aux Energies Alternatives (CEA) Valduc SILENE [1] facility. These experiments were a joint effort between the US Department of Energy (DOE) and the French CEA. The purpose of these experiments was to create three benchmarks for the verification and validation of radiation transport codes and evaluated nuclear data used in the analysis of criticality accident alarm systems (CAASs). This presentation will discuss the geometric configuration of these experiments and the quantities that were measured and will present some preliminary comparisons between the measured data and calculations. This series consisted of three single-pulsed experiments with the SILENE reactor. During the first experiment the reactor was bare (unshielded), but during the second and third experiments it was shielded by lead and polyethylene, respectively. During each experiment several neutron activation foils and thermoluminescent dosimeters (TLDs) were placed around the reactor, and some of these detectors were themselves shielded from the reactor by high-density magnetite and barite concrete, standard concrete, and/or BoroBond. All the concrete was provided by CEA Saclay, and the BoroBond was provided by Y-12 National Security Complex. Figure 1 is a picture of the SILENE reactor cell configured for pulse 1. Also included in these experiments were measurements of the neutron and photon spectra with two BICRON BC-501A liquid scintillators. These two detectors were provided and operated by CEA Valduc. They were set up just outside the SILENE reactor cell with additional lead shielding to prevent the detectors from being saturated. The final detectors involved in the experiments were two different types of CAAS detectors. The Babcock International Group provided three CIDAS CAAS detectors, which measured photon dose and dose rate with a Geiger-Mueller tube. CIDAS detectors are currently in use at Y-12 in the newly constructed Highly Enriched Uranium Materials Facility. The second CAAS detector used a {sup 6}LiF TLD to absorb neutrons and a silicon detector to count the charge particles released by these absorption events. Lawrence Livermore National Laboratory provided four of these detectors, which had formerly been used at the Rocky Flats facility in the United States.

  15. Does Daylight Savings Time Affect Traffic Accidents

    E-Print Network [OSTI]

    Deen, Sophia 1988-

    2012-04-20

    This paper studies the effect of changes in accident pattern due to Daylight Savings Time (DST). The extension of the DST in 2007 provides a natural experiment to determine whether the number of traffic accidents is affected by shifts in hours...

  16. Light-water reactor accident classification

    SciTech Connect (OSTI)

    Washburn, B.W.

    1980-02-01

    The evolution of existing classifications and definitions of light-water reactor accidents is considered. Licensing practice and licensing trends are examined with respect to terms of art such as Class 8 and Class 9 accidents. Interim definitions, consistent with current licensing practice and the regulations, are proposed for these terms of art.

  17. THERMAL ANALYSIS OF A 9975 PACKAGE IN A FACILITY FIRE ACCIDENT

    SciTech Connect (OSTI)

    Gupta, N.

    2011-02-14

    Surplus plutonium bearing materials in the U.S. Department of Energy (DOE) complex are stored in the 3013 containers that are designed to meet the requirements of the DOE standard DOE-STD-3013. The 3013 containers are in turn packaged inside 9975 packages that are designed to meet the NRC 10 CFR Part 71 regulatory requirements for transporting the Type B fissile materials across the DOE complex. The design requirements for the hypothetical accident conditions (HAC) involving a fire are given in 10 CFR 71.73. The 9975 packages are stored at the DOE Savannah River Site in the K-Area Material Storage (KAMS) facility for long term of up to 50 years. The design requirements for safe storage in KAMS facility containing multiple sources of combustible materials are far more challenging than the HAC requirements in 10 CFR 71.73. While the 10 CFR 71.73 postulates an HAC fire of 1475 F and 30 minutes duration, the facility fire calls for a fire of 1500 F and 86 duration. This paper describes a methodology and the analysis results that meet the design limits of the 9975 component and demonstrate the robustness of the 9975 package.

  18. The Fukushima Daiichi Accident Study Information Portal

    SciTech Connect (OSTI)

    Shawn St. Germain; Curtis Smith; David Schwieder; Cherie Phelan

    2012-11-01

    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 power station. The data is stored in a secured (password protected and encrypted) repository that is searchable and accessible to researchers at diverse locations.

  19. Comments on a paper tilted `The sea transport of vitrified high-level radioactive wastes: Unresolved safety issues`

    SciTech Connect (OSTI)

    Sprung, J.L.; McConnell, P.E.; Nigrey, P.J.; Ammerman, D.J. [and others

    1997-05-01

    The cited paper estimates the consequences that might occur should a purpose-built ship transporting Vitrified High Level Waste (VHLW) be involved in a severe collision that causes the VHLW canisters in one Type-B package to spill onto the floor of a major ocean fishing region. Release of radioactivity from VHLW glass logs, failure of elastomer cask seals, failure of VHLW canisters due to stress corrosion cracking (SCC), and the probabilities of the hypothesized accident scenario, of catastrophic cask failure, and of cask recovery from the sea are all discussed.

  20. Accident Investigation of the February 5, 2014, Underground Salt...

    Office of Environmental Management (EM)

    Accident Investigation of the February 5, 2014, Underground Salt Haul Truck Fire at the Waste Isolation Pilot Plant, Carlsbad NM Accident Investigation of the February 5, 2014,...

  1. Development of Light Water Reactor Fuels with Enhanced Accident...

    Energy Savers [EERE]

    Development of Light Water Reactor Fuels with Enhanced Accident Tolerance - Report to Congress Development of Light Water Reactor Fuels with Enhanced Accident Tolerance - Report to...

  2. Type B Accident Investigation Board Report on the October 8,...

    Energy Savers [EERE]

    Bettis Atomic Power Laboratory Type B Accident Investigation of the Arc Flash at Brookhaven National Laboratory, April 14, 2006 Type B Accident Investigation Board Report on...

  3. Type B Accident Investigation Board Report on the November 1...

    Energy Savers [EERE]

    Type B Accident Investigation Board Report on the November 1, 1999, Construction Injury at the Monticello Mill Tailings Remedial Action Site, Monticello, Utah Type B Accident...

  4. Naval Spent Fuel Rail Shipment Accident Exercise Objectives

    Office of Environmental Management (EM)

    NAVAL SPENT FUEL RAIL SHIPMENT ACCIDENT EXERCISE OBJECTIVES * Familiarize stakeholders with the Naval spent fuel ACCIDENT EXERCISE OBJECTIVES Familiarize stakeholders with the...

  5. Accident Investigation of the February 7, 2013, Scissor Lift...

    Energy Savers [EERE]

    Lift Accident in the West Hackberry Brine Tank-14 Resulting in Injury, Strategic Petroleum Reserve West Hackberry, LA Accident Investigation of the February 7, 2013, Scissor...

  6. Type B Accident Investigation of the July 12, 2007, Forklift...

    Energy Savers [EERE]

    2, 2007, Forklift and Pedestrian Accident at the Paducah Gaseous Diffusion Plant, PortsmouthPaducah Project Office Type B Accident Investigation of the July 12, 2007, Forklift and...

  7. ORISE: The Medical Basis for Radiation-Accident Preparedness...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    The Medical Basis for Radiation-Accident Preparedness: Medical Management Proceedings of the Fifth International REACTS Symposium on the Medical Basis for Radiation-Accident...

  8. A methodology for estimating the residual contamination contribution to the source term in a spent-fuel transport cask

    SciTech Connect (OSTI)

    Sanders, T.L. ); Jordan, H. . Rocky Flats Plant); Pasupathi, V. ); Mings, W.J. ); Reardon, P.C. )

    1991-09-01

    This report describes the ranges of the residual contamination that may build up in spent-fuel transport casks. These contamination ranges are calculated based on data taken from published reports and from previously unpublished data supplied by cask transporters. The data involve dose rate measurements, interior smear surveys, and analyses of water flushed out of cask cavities during decontamination operations. A methodology has been developed to estimate the effect of residual contamination on spent-fuel cask containment requirements. Factors in estimating the maximum permissible leak rates include the form of the residual contamination; possible release modes; internal gas-borne depletion; and the temperature, pressure, and vibration characteristics of the cask during transport under normal and accident conditions. 12 refs., 9 figs., 4 tabs.

  9. COMPARING THE IDENTIFICATION OF RECOMMENDATIONS BY DIFFERENT ACCIDENT

    E-Print Network [OSTI]

    Johnson, Chris

    will be identified for similar incidents. Accident analysis methods can also help to reduce individual bias

  10. Crediting Tritium Deposition in Accident Analysis

    SciTech Connect (OSTI)

    Murphy, C.E. Jr.

    2001-06-20

    This paper describes the major aspects of tritium dispersion phenomenology, summarizes deposition attributes of the computer models used in the DOE Complex for tritium dispersion, and recommends an approach to account for deposition in accident analysis.

  11. A systems approach to food accident analysis

    E-Print Network [OSTI]

    Helferich, John D

    2011-01-01

    Food borne illnesses lead to 3000 deaths per year in the United States. Some industries, such as aviation, have made great strides increasing safety through careful accident analysis leading to changes in industry practices. ...

  12. Structural assessment of accident loads

    SciTech Connect (OSTI)

    Wagenblast, G.R., Westinghouse Hanford

    1996-05-28

    Structural assessments were made for specific accident loads for specific catch, receiver, and storage tanks. The evaluation herein represents level-of-effort order-of-magnitude estimates of limiting loads that would lead to collapse or rupture of the tank and unmitigated loss of confinement for the waste. Structural capacities were established using failure criteria. Compliance with codes such as ACI, ASCE, ASME, RCRA, UBC, WAC, and DOE Orders was `NOT` maintained. Normal code practice is to prevent failure with margins consistent with expected variations in loads and strengths and confidence in analysis techniques. The evaluation herein represent estimates of code limits without code load factors or code strength reduction factors, and loading beyond such a limit is considered as an onset of some failure mode. The exact nature of the failure mode and its relation to a safe condition is a judgment of the analyst. Consequently, these `RESULTS SHALL NOT BE USED TO ESTABLISH OPERATING OR SAFETY LOAD LIMITS FOR THESE TANKS`.

  13. Accident management to prevent containment failure and reduce fission product release

    SciTech Connect (OSTI)

    Lehner, J.R.; Lin, C.C.; Luckas, W.J.; Pratt, W.T.

    1991-01-01

    Brookhaven National Laboratory, under the auspices of the US Nuclear Regulatory Commission, is investigating accident management strategies which could help preserve containment integrity or minimize releases during a severe accident. The strategies considered make use of existing plant systems and equipment in innovative ways to reduce the likelihood of containment failure or to mitigate the release of fission products to the environment if failure cannot be prevented. Many of these strategies would be implemented during the later stages of a severe accident, i.e. after vessel breach, and sizable uncertainties exist regarding some of the phenomena involved. The identification and assessment process for containment and release strategies is described, and some insights derived from its application to specific containment types are presented. 2 refs., 5 figs., 2 tabs.

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

    SciTech Connect (OSTI)

    Sdouz, Gert [ARC Seibersdorf Research GmbH, Viktor Kaplan-Strasse 2, 2700 Wr. Neustadt (Austria)

    2006-07-01

    The goal of this work is the investigation of the influence of different accident management strategies on the thermal-hydraulics in the containment during a Large Break Loss of Coolant Accident with a large containment leak from the beginning of the accident. The increasing relevance of terrorism suggests a closer look at this kind of severe accidents. Normally the course of severe accidents and their associated phenomena are investigated with the assumption of an intact containment from the beginning of the accident. This intact containment has the ability to retain a large part of the radioactive inventory. In these cases there is only a release via a very small leakage due to the un-tightness of the containment up to cavity bottom melt through. This paper represents the last part of a comprehensive study on the influence of accident management strategies on the source term of VVER-1000 reactors. Basically two different accident sequences were investigated: the 'Station Blackout'- sequence and the 'Large Break LOCA'. In a first step the source term calculations were performed assuming an intact containment from the beginning of the accident and no accident management action. In a further step the influence of different accident management strategies was studied. The last part of the project was a repetition of the calculations with the assumption of a damaged containment from the beginning of the accident. This paper concentrates on the last step in the case of a Large Break LOCA. To be able to compare the results with calculations performed years ago the calculations were performed using the Source Term Code Package (STCP), hydrogen explosions are not considered. In this study four different scenarios have been investigated. The main parameter was the switch on time of the spray systems. One of the results is the influence of different accident management strategies on the source term. In the comparison with the sequence with intact containment it was demonstrated that the accident management measures have quite lower consequences. In addition it was shown that in the case of a 'Large Break LOCA'-sequence the intact containment retains the nuclides up to a factor of 20 000. This is much more than in the case of a 'Station Blackout'-sequence. Within the frame of the study 17 source terms have been generated to evaluate in detail accident management strategies for VVER-1000 reactors. (authors)0.

  15. Spent Fuel Transportation Package Performance Study - Experimental Design Challenges

    SciTech Connect (OSTI)

    Snyder, A. M.; Murphy, A. J.; Sprung, J. L.; Ammerman, D. J.; Lopez, C.

    2003-02-25

    Numerous studies of spent nuclear fuel transportation accident risks have been performed since the late seventies that considered shipping container design and performance. Based in part on these studies, NRC has concluded that the level of protection provided by spent nuclear fuel transportation package designs under accident conditions is adequate. [1] Furthermore, actual spent nuclear fuel transport experience showcase a safety record that is exceptional and unparalleled when compared to other hazardous materials transportation shipments. There has never been a known or suspected release of the radioactive contents from an NRC-certified spent nuclear fuel cask as a result of a transportation accident. In 1999 the United States Nuclear Regulatory Commission (NRC) initiated a study, the Package Performance Study, to demonstrate the performance of spent fuel and spent fuel packages during severe transportation accidents. NRC is not studying or testing its current regulations, a s the rigorous regulatory accident conditions specified in 10 CFR Part 71 are adequate to ensure safe packaging and use. As part of this study, NRC currently plans on using detailed modeling followed by experimental testing to increase public confidence in the safety of spent nuclear fuel shipments. One of the aspects of this confirmatory research study is the commitment to solicit and consider public comment during the scoping phase and experimental design planning phase of this research.

  16. Precursors to potential severe core damage accidents, 1986: A status report: Main report and Appendixes A,B, and C

    SciTech Connect (OSTI)

    Minarick, J W; Harris, J D; Austin, P N; Cletcher, J W; Hagen, E W

    1988-05-01

    The Accident Sequence Precursor Program reviews licensee event reports of operational events that have occurred at LWRs to identify and categorize precursors to potential severe core-damage accidents. Accident sequences considered in the study are those associated with inadequate core cooling. Accident sequence precursors are events that are important elements in such sequences. Such precursors could be infrequent initiating events or equipment failures that, when coupled with one or more postulated events, could result in a plant condition with inadequate core cooling. Originally proposed in the Risk Assessment Review Group Report (Lewis Committee report) in 1978, the study - subsequently named the Accident Sequence Precursor Program - was initiated at the Nuclear Operations Analysis Center in 1979. Earlier reports by the program involved assessment of events that occurred in 1969-1981 and 1984-1985. The present report involves the assessment of events that occurred during 1986. A nuclear plant has safety systems for mitigating the consequences of accidents or off-normal initiating events that may occur during the course of plant operation. These systems are built to high-quality standards and are redundant; nonetheless, they have a nonzero probability of failing or being in a failed state when required to operate. This report uses LERs and other plant data, estimated system unavailabilities, the expected average frequency of initiating events (LOFWs, LOOPs, LOCAs), and event details to evaluate the potential impact of the following two situations.

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

    SciTech Connect (OSTI)

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

    1992-12-31

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

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

    SciTech Connect (OSTI)

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

    1992-01-01

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

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

    SciTech Connect (OSTI)

    Paul A Demkowicz; Paul Demkowicz; David V Laug

    2010-10-01

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

  20. Sustainable Transport

    E-Print Network [OSTI]

    Webber, Melvin

    2006-01-01

    THOUGHT PIECE Sustainable Transport by Melvin M. Webberwant to sustain any mode of transport only if we judge it todraconian in rejecting transport modes that have failed in

  1. Estimate of radionuclide release characteristics into containment under severe accident conditions. Final report

    SciTech Connect (OSTI)

    Nourbakhsh, H.P.

    1993-11-01

    A detailed review of the available light water reactor source term information is presented as a technical basis for development of updated source terms into the containment under severe accident conditions. Simplified estimates of radionuclide release and transport characteristics are specified for each unique combination of the reactor coolant and containment system combinations. A quantitative uncertainty analysis in the release to the containment using NUREG-1150 methodology is also presented.

  2. ATMOSPHERIC MODELING IN SUPPORT OF A ROADWAY ACCIDENT

    SciTech Connect (OSTI)

    Buckley, R.; Hunter, C.

    2010-10-21

    The United States Forest Service-Savannah River (USFS) routinely performs prescribed fires at the Savannah River Site (SRS), a Department of Energy (DOE) facility located in southwest South Carolina. This facility covers {approx}800 square kilometers and is mainly wooded except for scattered industrial areas containing facilities used in managing nuclear materials for national defense and waste processing. Prescribed fires of forest undergrowth are necessary to reduce the risk of inadvertent wild fires which have the potential to destroy large areas and threaten nuclear facility operations. This paper discusses meteorological observations and numerical model simulations from a period in early 2002 of an incident involving an early-morning multicar accident caused by poor visibility along a major roadway on the northern border of the SRS. At the time of the accident, it was not clear if the limited visibility was due solely to fog or whether smoke from a prescribed burn conducted the previous day just to the northwest of the crash site had contributed to the visibility. Through use of available meteorological information and detailed modeling, it was determined that the primary reason for the low visibility on this night was fog induced by meteorological conditions.

  3. A Review of Criticality Accidents 2000 Revision

    SciTech Connect (OSTI)

    Thomas P. McLaughlin; Shean P. Monahan; Norman L. Pruvost; Vladimir V. Frolov; Boris G. Ryazanov; Victor I. Sviridov

    2000-05-01

    Criticality accidents and the characteristics of prompt power excursions are discussed. Sixty accidental power excursions are reviewed. Sufficient detail is provided to enable the reader to understand the physical situation, the chemistry and material flow, and when available the administrative setting leading up to the time of the accident. Information on the power history, energy release, consequences, and causes are also included when available. For those accidents that occurred in process plants, two new sections have been included in this revision. The first is an analysis and summary of the physical and neutronic features of the chain reacting systems. The second is a compilation of observations and lessons learned. Excursions associated with large power reactors are not included in this report.

  4. Enhanced Accident Tolerant LWR Fuels: Metrics Development

    SciTech Connect (OSTI)

    Shannon Bragg-Sitton; Lori Braase; Rose Montgomery; Chris Stanek; Robert Montgomery; Lance Snead; Larry Ott; Mike Billone

    2013-09-01

    The Department of Energy (DOE) Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) is conducting research and development on enhanced Accident Tolerant Fuels (ATF) for light water reactors (LWRs). This mission emphasizes the development of novel fuel and cladding concepts to replace the current zirconium alloy-uranium dioxide (UO2) fuel system. The overall mission of the ATF research is to develop advanced fuels/cladding with improved performance, reliability and safety characteristics during normal operations and accident conditions, while minimizing waste generation. The initial effort will focus on implementation in operating reactors or reactors with design certifications. To initiate the development of quantitative metrics for ATR, a LWR Enhanced Accident Tolerant Fuels Metrics Development Workshop was held in October 2012 in Germantown, MD. This paper summarizes the outcome of that workshop and the current status of metrics development for LWR ATF.

  5. Accident progression event tree analysis for postulated severe accidents at N Reactor

    SciTech Connect (OSTI)

    Wyss, G.D.; Camp, A.L.; Miller, L.A.; Dingman, S.E.; Kunsman, D.M. (Sandia National Labs., Albuquerque, NM (USA)); Medford, G.T. (Science Applications International Corp., Albuquerque, NM (USA))

    1990-06-01

    A Level II/III probabilistic risk assessment (PRA) has been performed for N Reactor, a Department of Energy (DOE) production reactor located on the Hanford reservation in Washington. The accident progression analysis documented in this report determines how core damage accidents identified in the Level I PRA progress from fuel damage to confinement response and potential releases the environment. The objectives of the study are to generate accident progression data for the Level II/III PRA source term model and to identify changes that could improve plant response under accident conditions. The scope of the analysis is comprehensive, excluding only sabotage and operator errors of commission. State-of-the-art methodology is employed based largely on the methods developed by Sandia for the US Nuclear Regulatory Commission in support of the NUREG-1150 study. The accident progression model allows complex interactions and dependencies between systems to be explicitly considered. Latin Hypecube sampling was used to assess the phenomenological and systemic uncertainties associated with the primary and confinement system responses to the core damage accident. The results of the analysis show that the N Reactor confinement concept provides significant radiological protection for most of the accident progression pathways studied.

  6. Fuel performance during severe accidents. [PWR

    SciTech Connect (OSTI)

    Buescher, B.J.; Gruen, G.E.; MacDonald, P.E.

    1982-01-01

    As a result of the Three Mile Island Unit-2 (TMI-2) accident, the Nuclear Regulatory Commission has initiated a severe fuel damage test program to evaluate fuel rod and core response during severe accidents similar to TMI-2. This program is underway in the Power Burst Facility at the Idaho National Engineering Laboratory. In preparation for the first test, predictions have been performed using the TRAC-BD1 computer. This paper presents the calculated results showing a slow heatup to 2400 K over 5 hours, and the analysis includes accelerated oxidation of the zirconium cladding at temperatures above 1850 K.

  7. The temporal effect of traffic violations and accidents on accident occurrence 

    E-Print Network [OSTI]

    McKemie, Martha Susan

    1979-01-01

    THE TEMPORAL EFFECT OF TRAFFIC VIOLATIONS AND ACCIDENTS ON ACCIDENT OCCURRENCE A Thesis by . 1artha Susan McKemie Submitted to the Graduate College of Texas A&M University in partial fulfillment of the requirement for the degree of MASTER... OF SCIENCE December 1979 Major Subject: Industrial Engineering THE TEMPORAL El'FECT OF TRAI'FIC VIOIATIONS AND ACCIDENTS ON XCCIDENT OCCURPEENCE A Thesis by Martha Susan McKemie Approved as to style and content by: / ~J' (Chairman of Commi tee...

  8. Decontamination analysis of the NUWAX-83 accident site using DECON

    SciTech Connect (OSTI)

    Tawil, J.J.

    1983-11-01

    This report presents an analysis of the site restoration options for the NUWAX-83 site, at which an exercise was conducted involving a simulated nuclear weapons accident. This analysis was performed using a computer program deveoped by Pacific Northwest Laboratory. The computer program, called DECON, was designed to assist personnel engaged in the planning of decontamination activities. The many features of DECON that are used in this report demonstrate its potential usefulness as a site restoration planning tool. Strategies that are analyzed with DECON include: (1) employing a Quick-Vac option, under which selected surfaces are vacuumed before they can be rained on; (2) protecting surfaces against precipitation; (3) prohibiting specific operations on selected surfaces; (4) requiring specific methods to be used on selected surfaces; (5) evaluating the trade-off between cleanup standards and decontamination costs; and (6) varying of the cleanup standards according to expected exposure to surface.

  9. Type B Accident Investigation, Response to the 24 Command Wildland...

    Energy Savers [EERE]

    Type B Accident Investigation, Response to the 24 Command Wildland Fire on the Hanford Site, June 27-July 1, 2000 Type B Accident Investigation, Response to the 24 Command Wildland...

  10. PNNL Results from 2009 Silene Criticality Accident Dosimeter Intercomparison Exercise

    SciTech Connect (OSTI)

    Hill, Robin L.; Conrady, Matthew M.

    2010-06-30

    This document reports the results of testing of the Hanford Personnel Nuclear Accident Dosimeter (PNAD) during a criticality accident dosimeter intercomparison exercise at the CEA Valduc Center on October 13, 14, and 15, 2009.

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

    SciTech Connect (OSTI)

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

    1986-01-01

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

  12. ACCIDENT ANALYSIS AND HAZARD ANALYSIS FOR HUMAN AND ORGANIZATIONAL FACTORS

    E-Print Network [OSTI]

    Leveson, Nancy

    culpable. An accident analysis method is needed that will guide the work, aid in the analysis of the role

  13. INTERNATIONAL STUDENT & SCHOLAR Accident & Sickness Insurance Plan

    E-Print Network [OSTI]

    Bordenstein, Seth

    and scholars participating in international educational programs outside of the United States. It is strongly an accident and sickness insurance plan for international students and scholars studying in the United States. The International Student & Scholar plan has a low monthly rate of $70 per person. WE'VE GOT YOU COVERED

  14. Severe Accident Test Station Activity Report

    SciTech Connect (OSTI)

    Pint, Bruce A.; Terrani, Kurt A.

    2015-06-01

    Enhancing safety margins in light water reactor (LWR) severe accidents is currently the focus of a number of international R&D programs. The current UO2/Zr-based alloy fuel system is particularly susceptible since the Zr-based cladding experiences rapid oxidation kinetics in steam at elevated temperatures. Therefore, alternative cladding materials that offer slower oxidation kinetics and a smaller enthalpy of oxidation can significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident. In the U.S. program, the high temperature steam oxidation performance of accident tolerant fuel (ATF) cladding solutions has been evaluated in the Severe Accident Test Station (SATS) at Oak Ridge National Laboratory (ORNL) since 2012. This report summarizes the capabilities of the SATS and provides an overview of the oxidation kinetics of several candidate cladding materials. A suggested baseline for evaluating ATF candidates is a two order of magnitude reduction in the steam oxidation resistance above 1000ºC compared to Zr-based alloys. The ATF candidates are categorized based on the protective external oxide or scale that forms during exposure to steam at high temperature: chromia, alumina, and silica. Comparisons are made to literature and SATS data for Zr-based alloys and other less-protective materials.

  15. Characterization of a nuclear accident dosimeter 

    E-Print Network [OSTI]

    Burrows, Ronald Allen

    1995-01-01

    The 23rd nuclear accident dosimetry intercomparison was held during the week of June 12-16, 1995 at Los Alamos National Laboratory. This report presents the results of this event, referred to as NAD 23, as related to the performance of Sandia...

  16. Accident or Design Taeil A. Bai

    E-Print Network [OSTI]

    Bai, Taeil

    to this principle, a life-giving factor lies at the center of the whole machinery and design of the world. Here weAccident or Design Taeil A. Bai Stanford University, Stanford, CA 94305 In a companion article find the word "design," which has been expelled from biological scholarship by the biologists. Here we

  17. L'accident la centrale nuclaire de Quelques explications scientifiques

    E-Print Network [OSTI]

    Skorobogatiy, Maksim

    L'accident à la centrale nucléaire de Fukushima Quelques explications scientifiques G. Marleau, J´eal, 18 mars 2011 L'accident `a la centrale nucl´eaire de Fukushima ­ 1/29 Accident de Fukushima 1 Contenu de Fukushima. 3. La puissance résiduelle. 4. Perte de refroidissement et conséquences. 5

  18. Louisiana Forest Products Lab 1 Accidents in the Primary &

    E-Print Network [OSTI]

    Louisiana Forest Products Lab 1 Accidents in the Primary & Secondary Forest Products Industry Center #12;Louisiana Forest Products Lab 2 Abitibi Paper Co. Camp 40 Thunder Bay, Ontario #12;Louisiana Forest Products Lab 3 Accidents in Forest Products Industry Accident Statistics Primary industry

  19. Transportation fuels from wood

    SciTech Connect (OSTI)

    Baker, E.G.; Elliott, D.C.; Stevens, D.J.

    1980-01-01

    The various methods of producing transportation fuels from wood are evaluated in this paper. These methods include direct liquefaction schemes such as hydrolysis/fermentation, pyrolysis, and thermochemical liquefaction. Indirect liquefaction techniques involve gasification followed by liquid fuels synthesis such as methanol synthesis or the Fischer-Tropsch synthesis. The cost of transportation fuels produced by the various methods are compared. In addition, three ongoing programs at Pacific Northwest Laboratory dealing with liquid fuels from wood are described.

  20. Use of influence diagrams for evaluation of severe accident management strategies

    SciTech Connect (OSTI)

    Jae, M.; Apostolakis, G.E. (Univ. of California, Los Angeles (United States))

    1991-01-01

    This paper presents a new approach for developing and assessing severe accident management strategies under uncertainty in nuclear power plants. The strategy of flooding the reactor cavity during the TMLB{prime} sequence as a means to prevent vessel breach is used as an example. The modeling of complex decision problems, such as those encountered in severe accident management, involves a large number of random variables. While the state of the art relies on decision trees, influence diagrams have been proposed as an alternative. Large decision trees cannot be displayed except in pieces, but influence diagrams (as suggested in this paper) can depict much larger and more complicated models, such as those required for the development of strategies for managing severe accidents in nuclear power plants. The advantages of influence diagrams include a compact and unambiguous representation of probabilistic dependencies of various events or processes and good communication of the structure of a decision model. Furthermore, influence diagrams allow for the rapid identification of important variables and are easily modified in case the decision maker wants to add or remove some nodes or reverse arcs, making the influence diagrams good tools for developing, as well as evaluating, severe accident management strategies. The superiority of this model is clear in more complicated situations, such as multidecision problems.

  1. Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 3: Fission-Product Transport and Dose PIRTs

    SciTech Connect (OSTI)

    Morris, Robert Noel [ORNL

    2008-03-01

    This Fission Product Transport (FPT) Phenomena Identification and Ranking Technique (PIRT) report briefly reviews the high-temperature gas-cooled reactor (HTGR) FPT mechanisms and then documents the step-by-step PIRT process for FPT. The panel examined three FPT modes of operation: (1) Normal operation which, for the purposes of the FPT PIRT, established the fission product circuit loading and distribution for the accident phase. (2) Anticipated transients which were of less importance to the panel because a break in the pressure circuit boundary is generally necessary for the release of fission products. The transients can change the fission product distribution within the circuit, however, because temperature changes, flow perturbations, and mechanical vibrations or shocks can result in fission product movement. (3) Postulated accidents drew the majority of the panel's time because a breach in the pressure boundary is necessary to release fission products to the confinement. The accidents of interest involved a vessel or pipe break, a safety valve opening with or without sticking, or leak of some kind. Two generic scenarios were selected as postulated accidents: (1) the pressurized loss-of-forced circulation (P-LOFC) accident, and (2) the depressurized loss-of-forced circulation (D-LOFC) accidents. FPT is not an accident driver; it is the result of an accident, and the PIRT was broken down into a two-part task. First, normal operation was seen as the initial starting point for the analysis. Fission products will be released by the fuel and distributed throughout the reactor circuit in some fashion. Second, a primary circuit breach can then lead to their release. It is the magnitude of the release into and out of the confinement that is of interest. Depending on the design of a confinement or containment, the impact of a pressure boundary breach can be minimized if a modest, but not excessively large, fission product attenuation factor can be introduced into the release path. This exercise has identified a host of material properties, thermofluid states, and physics models that must be collected, defined, and understood to evaluate this attenuation factor. The assembled PIRT table underwent two iterations with extensive reorganization between meetings. Generally, convergence was obtained on most issues, but different approaches to the specific physics and transport paths shade the answers accordingly. The reader should be cautioned that merely selecting phenomena based on high importance and low knowledge may not capture the true uncertainty of the situation. This is because a transport path is composed of several serial linkages, each with its own uncertainty. The propagation of a chain of modest uncertainties can lead to a very large uncertainty at the end of a long path, resulting in a situation that is of little regulatory guidance.

  2. Broken Arrows: Radiological hazards from nuclear warhead accidents (the Minot USAF base nuclear weapons incident)

    E-Print Network [OSTI]

    Liolios, Theodore

    2009-01-01

    According to numerous press reports, in 2007 at Minot US Air Force Base six AGM-129 Advanced Cruise Missiles mistakenly armed with W80-1 thermonuclear warheads were loaded on a B-52H heavy bomber in place of six unarmed AGM-129 missiles that were awaiting transport to Barksdale US Air Force Base for disposal. The live nuclear missiles were not reported missing, and stood unsecured and unguarded while mounted to the aircraft for a period of 36 hours. The present work investigates the radiological hazards associated with a worst-case postulated accident that would disperse the nuclear material of the six warheads in large metropolitan cities. Using computer simulations approximate estimates are derived for the ensuing cancer mortality and land contamination after the accident. Health, decontamination and evacuation costs are also estimated in the framework of the linear risk model.

  3. Lessons Learned from Three Mile Island Packaging, Transportation and Disposition that Apply to Fukushima Daiichi Recovery

    SciTech Connect (OSTI)

    Layne Pincock; Wendell Hintze; Dr. Koji Shirai

    2012-07-01

    Following the massive earthquake and resulting tsunami damage in March of 2011 at the Fukushima Daiichi nuclear power plant in Japan, interest was amplified for what was done for recovery at the Three Mile Island Unit 2 (TMI-2) in the United States following its meltdown in 1979. Many parallels could be drawn between to two accidents. This paper presents the results of research done into the TMI-2 recovery effort and its applicability to the Fukushima Daiichi cleanup. This research focused on three topics: packaging, transportation, and disposition. This research work was performed as a collaboration between Japan’s Central Research Institute of Electric Power Industry (CRIEPI) and the Idaho National Laboratory (INL). Hundreds of TMI-2 related documents were searched and pertinent information was gleaned from these documents. Other important information was also obtained by interviewing employees who were involved first hand in various aspects of the TMI-2 cleanup effort. This paper is organized into three main sections: (1) Transport from Three Mile Island to Central Facilities Area at INL, (2) Transport from INL Central Receiving Facility to INL Test Area North (TAN) and wet storage at TAN, and (3) Transport from TAN to INL Idaho Nuclear Technology and Engineering Center (INTEC) and Dry Storage at INTEC. Within each of these sections, lessons learned from performing recovery activities are presented and their applicability to the Fukushima Daiichi nuclear power plant cleanup are outlined.

  4. Evaluation Metrics Applied to Accident Tolerant Fuels

    SciTech Connect (OSTI)

    Shannon M. Bragg-Sitton; Jon Carmack; Frank Goldner

    2014-10-01

    The safe, reliable, and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and have yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. One of the current missions of the U.S. Department of Energy’s (DOE) Office of Nuclear Energy (NE) is to develop nuclear fuels and claddings with enhanced accident tolerance for use in the current fleet of commercial LWRs or in reactor concepts with design certifications (GEN-III+). Accident tolerance became a focus within advanced LWR research upon direction from Congress following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal of ATF development is to identify alternative fuel system technologies to further enhance the safety, competitiveness and economics of commercial nuclear power. Enhanced accident tolerant fuels would endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system while maintaining or improving performance during normal operations. The U.S. DOE is supporting multiple teams to investigate a number of technologies that may improve fuel system response and behavior in accident conditions, with team leadership provided by DOE national laboratories, universities, and the nuclear industry. Concepts under consideration offer both evolutionary and revolutionary changes to the current nuclear fuel system. Mature concepts will be tested in the Advanced Test Reactor at Idaho National Laboratory beginning in Summer 2014 with additional concepts being readied for insertion in fiscal year 2015. This paper provides a brief summary of the proposed evaluation process that would be used to evaluate and prioritize the candidate accident tolerant fuel concepts currently under development.

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

    SciTech Connect (OSTI)

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

    2010-03-01

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

  6. SARNET: Integrating Severe Accident Research in Europe - Safety Issues in the Source Term Area

    SciTech Connect (OSTI)

    Haste, T. [Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Giordano, P.; Micaelli, J.-C. [Institut de Radioprotection et de S et Nucl ire, IRSN, BP 3 13115 St Paul lez Durance Cedex (France); Herranz, L. [Centro de Investigaciones Energeticas Medio Ambientales y Tecnologica, CIEMAT, Avda. Complutense 22, 28040 Madrid (Spain)

    2006-07-01

    SARNET (Severe Accident Research Network) is a Network of Excellence of the EU 6. Framework Programme that integrates in a sustainable manner the research capabilities of about fifty European organisations to resolve important remaining uncertainties and safety issues concerning existing and future nuclear plant, especially water-cooled reactors, under hypothetical severe accident conditions. It emphasises integrating activities, spreading of excellence (including knowledge transfer) and jointly-executed research. This paper summarises the main results obtained at the middle of the current 4-year term, highlighting those concerning radioactive release to the environment. Integration is pursued through different methods: the ASTEC integral computer code for severe accident modelling, development of PSA level 2 methods, a means for definition, updating and resolution of safety issues, and development of a web database for storing experimental results. These activities are helped by an evolving Advanced Communication Tool, easing communication amongst partners. Concerning spreading of excellence, educational courses covering severe accident analysis methodology and level 2 PSA have been organised for early 2006. A text book on Severe Accident Phenomenology is being written. A mobility programme for students and young researchers has started. Results are disseminated mainly through open conference proceedings, with journal publications planned. The 1. European Review Meeting on Severe Accidents in November 2005 covered SARNET activities during its first 18 months. Jointly executed research activities concern key issues grouped in the Corium, Containment and Source Term areas. In Source Term, behaviour of the highly radio-toxic ruthenium under oxidising conditions, including air ingress, is investigated. Models are proposed for fuel and ruthenium oxidation. Experiments on transport of oxide ruthenium species are performed. Reactor scenario studies assist in defining conditions for new experiments. Regarding predictability of iodine species exiting the Reactor Coolant System (RCS), which affects the amount entering the containment, iodine behaviour in the circuit and silver-indium-cadmium (SIC) release have been reviewed. New experiments are being discussed and performed, and SIC degradation and release models are being improved. For the radioactive aerosol source term, work is conducted in the risk-relevant areas of steam generator (SG) tube rupture, transport through cracks in containment walls and revaporization from previous deposits in the RCS that could lead to a delayed source term. Models for aerosol retention in containment cracks and interpretation of data on retention in the SG secondary side are proposed. For radioactive iodine release to the environment, many physical and chemical processes affect the iodine concentration in the containment atmosphere; of these effects, mass transfer phenomena and radiolytic oxidation are being investigated first. (authors)

  7. US Department of Energy Chernobyl accident bibliography

    SciTech Connect (OSTI)

    Kennedy, R A; Mahaffey, J A; Carr, F Jr

    1992-04-01

    This bibliography has been prepared by Pacific Northwest Laboratory (PNL) for the US Department of Energy (DOE) Office of Health and Environmental Research to provide bibliographic information in a usable format for research studies relating to the Chernobyl nuclear accident that occurred in the Ukrainian Republic, USSR in 1986. This report is a product of the Chernobyl Database Management project. The purpose of this project is to produce and maintain an information system that is the official United States repository for information related to the accident. Two related products prepared for this project are the Chernobyl Bibliographic Search System (ChernoLit{trademark}) and the Chernobyl Radiological Measurements Information System (ChernoDat). This report supersedes the original release of Chernobyl Bibliography (Carr and Mahaffey, 1989). The original report included about 2200 references. Over 4500 references and an index of authors and editors are included in this report.

  8. Mitigative techniques and analysis of generic site conditions for ground-water contamination associated with severe accidents

    SciTech Connect (OSTI)

    Shafer, J.M.; Oberlander, P.L.; Skaggs, R.L.

    1984-04-01

    The purpose of this study is to evaluate the feasibility of using ground-water contaminant mitigation techniques to control radionuclide migration following a severe commercial nuclear power reactor accident. The two types of severe commercial reactor accidents investigated are: (1) containment basemat penetration of core melt debris which slowly cools and leaches radionuclides to the subsurface environment, and (2) containment basemat penetration of sump water without full penetration of the core mass. Six generic hydrogeologic site classifications are developed from an evaluation of reported data pertaining to the hydrogeologic properties of all existing and proposed commercial reactor sites. One-dimensional radionuclide transport analyses are conducted on each of the individual reactor sites to determine the generic characteristics of a radionuclide discharge to an accessible environment. Ground-water contaminant mitigation techniques that may be suitable, depending on specific site and accident conditions, for severe power plant accidents are identified and evaluated. Feasible mitigative techniques and associated constraints on feasibility are determined for each of the six hydrogeologic site classifications. The first of three case studies is conducted on a site located on the Texas Gulf Coastal Plain. Mitigative strategies are evaluated for their impact on contaminant transport and results show that the techniques evaluated significantly increased ground-water travel times. 31 references, 118 figures, 62 tables.

  9. Transportation of medical isotopes

    SciTech Connect (OSTI)

    Nielsen, D.L.

    1997-11-19

    A Draft Technical Information Document (HNF-1855) is being prepared to evaluate proposed interim tritium and medical isotope production at the Fast Flux Test Facility (FFTF). This assessment examines the potential health and safety impacts of transportation operations associated with the production of medical isotopes. Incident-free and accidental impacts are assessed using bounding source terms for the shipment of nonradiological target materials to the Hanford Site, the shipment of irradiated targets from the FFTF to the 325 Building, and the shipment of medical isotope products from the 325 Building to medical distributors. The health and safety consequences to workers and the public from the incident-free transportation of targets and isotope products would be within acceptable levels. For transportation accidents, risks to works and the public also would be within acceptable levels. This assessment is based on best information available at this time. As the medical isotope program matures, this analysis will be revised, if necessary, to support development of a final revision to the Technical Information Document.

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

    SciTech Connect (OSTI)

    Tusheva, P.; Schaefer, F.; Kliem, S. [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstrasse 400, D-01328 Dresden (Germany)

    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 safety 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)

  11. Source term evaluation for postulated UF{sub 6} release accidents in gaseous diffusion plants -- Summer ventilation mode (non-seismic cases)

    SciTech Connect (OSTI)

    Kim, S.H.; Chen, N.C.J.; Taleyarkhan, R.P.; Wendel, M.W.; Keith, K.D.; Schmidt, R.W. [Oak Ridge National Lab., TN (United States); Carter, J.C. [J.C. Carter Associates, Inc., Knoxville, TN (United States); Dyer, R.H. [Dyer Enterprises, Harriman, TN (United States)

    1996-12-30

    Computer models have been developed to simulate the transient behavior of aerosols and vapors as a result of a postulated accident involving the release of uranium hexafluoride (UF{sub 6}) into the process building of a gaseous diffusion plant. For the current study, gaseous UF{sub 6} is assumed to get released in the cell housing atmosphere through B-line break at 58.97 kg/s for 10 min and 30 min duration at the Paducah and Portsmouth Gaseous Diffusion Plants. The released UF{sub 6} undergoes an exothermic chemical reaction with moisture (H{sub 2}O) in the air to form hydrogen fluoride (HF) and radioactive uranyl fluoride (UO{sub 2}F{sub 2}) while it disperses throughout the process building. As part of a facility-wide safety evaluation, this study evaluated source terms consisting of UO{sub 2}F{sub 2} as well as HF during a postulated UF{sub 6} release accident in a process building. UO{sub 2}F{sub 2} mainly remains as airborne-solid particles (aerosols), and HF is in a vapor form. Some UO{sub 2}F{sub 2} aerosols are removed from the air flow due to gravitational settling. The HF and the remaining UO{sub 2}F{sub 2} are mixed with air and exhausted through the building ventilation system. The MELCOR computer code was selected for simulating aerosols and vapor transport in the process building. To characterize leakage flow through the cell housing wall, 3-D CFD tool (CFDS-FLOW3D) was used. About 57% of UO{sub 2}F{sub 2} was predicted to be released into the environment. Since HF was treated as vapor, close to 100% was estimated to get released into the environment.

  12. Spent Fuel Transportation Cask Response to the Caldecott Tunnel Fire Scenario

    SciTech Connect (OSTI)

    Adkins, Harold E.; Koeppel, Brian J.; Cuta, Judith M.

    2007-01-01

    On April 7, 1982, a tank truck and trailer carrying 8,800 gallons of gasoline was involved in an accident in the Caldecott tunnel on State Route 24 near Oakland, California. The tank trailer overturned and subsequently caught fire. The United States Nuclear Regulatory Commission (USNRC), one of the agencies responsible for ensuring the safe transportation of radioactive materials in the United States, undertook analyses to determine the possible regulatory implications of this particular event for the transportation of spent nuclear fuel by truck. The Fire Dynamics Simulator (FDS) code developed by National Institute of Standards and Technology (NIST) was used to determine the thermal environment in the Caldecott tunnel during the fire. The FDS results were used to define boundary conditions for a thermal transient model of a truck transport cask containing spent nuclear fuel. The Nuclear Assurance Corporation (NAC) Legal Weight Truck (LWT) transportation cask was selected for this evaluation, as it represents a typical truck (over-the-road) cask, and can be used to transport a wide variety of spent nuclear fuels. Detailed analysis of the cask response to the fire was performed using the ANSYS® computer code to evaluate the thermal performance of the cask design in this fire scenario. This report describes the methods and approach used to assess the thermal response of the selected cask design to the conditions predicted in the Caldecott tunnel fire. The results of the analysis are presented in detail, with an evaluation of the cask response to the fire. The staff concluded that some components of smaller transportation casks resembling the NAC LWT, despite placement within an ISO container, could degrade significantly. Small transportation casks similar to the NAC LWT would probably experience failure of seals in this severe accident scenario. USNRC staff evaluated the radiological consequences of the cask response to the Caldecott tunnel fire. Although some components heated up beyond their service temperatures, the staff determined that there would be no significant release as a result of the fire for the NAC LWT and similar casks.

  13. The MacArthur Maze Fire and Roadway Collapse: Consequences for SNF Transportation - 12476

    SciTech Connect (OSTI)

    Bajwa, Christopher S.; Easton, Earl P. [U.S. Nuclear Regulatory Commission, Washington, DC 20005 (United States); Adkins, Harold; Cuta, Judith; Klymyshyn, Nicholas; Suffield, Sarah [Pacific Northwest National Laboratory, Richland, Washington 99352 (United States)

    2012-07-01

    In 2007, a severe transportation accident occurred near Oakland, California, on a section of Interstate 880 known as the 'MacArthur Maze', involving a tractor trailer carrying gasoline which impacted an overpass support column and burst into flames. The subsequent fire caused the collapse of portions of the Interstate 580 overpass onto the remains of the tractor-trailer in less than 20 minutes, due to a reduction of strength in the structural steel exposed to the fire. The US Nuclear Regulatory Commission is in the process of examining the impacts of this accident on the performance of a spent nuclear fuel transportation package, using detailed analysis models, in order to determine the potential regulatory implications related to the safe transport of spent nuclear fuel in the United States. This paper will provide a summary of this effort and present results and conclusions. The detailed thermal models of the MacArthur Maze fire scenario with ANSYS and COBRASFS have produced preliminary results indicating that in a fire of this severity, the peak fuel cladding temperature would almost certainly exceed the short-term limit of 570 deg. C (1058 deg. F), and would likely exceed the Zircaloy burst temperature limit of 750 deg. C (1382 deg. F) assumed in previous transportation studies. Additional work is needed to refine and verify some of the details of these complex models, but the overall results are consistent with previous fire analyses with similar models, and with the results obtained for the HAC fire evaluations with these models. These results as well as future results produced by these models can therefore be considered as reliable estimates of the temperatures that would be experienced in fire conditions of the severity of the MacArthur Maze fire scenario. The structural analyses show that the GA-4 package is robust enough to withstand the impact of the overhead span without suffering major damage or deformation to the containment boundary. The greatest potential for local package damage in this scenario appears to be at the bolted closure end. The thermal expansion response of the closure bolts and impact limiter attachment bolts were evaluated in a separate analysis, and it was determined that both the lid and impact limiters are expected to remain in place. The response of the closure seal is currently being evaluated separately in the context of the accident's overall potential to release radioactive material in the environment. (authors)

  14. Uncertainties in source term estimates for a station blackout accident in a BWR with Mark I containment

    SciTech Connect (OSTI)

    Lee, M.; Cazzoli, E.; Liu, Y.; Davis, R.; Nourbakhsh, H.; Schmidt, E.; Unwin, S.; Khatib-Rahbar, M.

    1988-01-01

    In this paper, attention is limited to a single accident progression sequence, namly a station blackout accident in a BWR with a Mark I containment building. Identified as an important accident in the draft version of NUREG-1150 a station blackout involves loss of both off-site power and dc power resulting in failure of the diesels to start and in the unavailability of the high pressure injection and core isolation cooling systems. This paper illustrates the calculated uncertainties (Probability Density Functions) associated with the radiological releases into the environment for the nine fission product groups at 10 hours following the initiation of core-concrete interactions. Also shown are the results ofthe STCP base case simulation. 5 refs., 1 fig., 1 tab.

  15. Chernobyl Studies Project: Working group 7.0, Environmental transport and health effects. Progress report, March--September 1994

    SciTech Connect (OSTI)

    Anspaugh, L.R.; Hendrickson, S.M.

    1994-12-01

    In April 1988, the US and the former-USSR signed a Memorandum of Cooperation (MOC) for Civilian Nuclear Reactor Safety; this MOC was a direct result of the accident at the Chernobyl Nuclear Power Plant Unit 4 and the following efforts by the two countries to implement a joint program to improve the safety of nuclear power plants and to understand the implications of environmental releases. A Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS) was formed to implement the MOC. The JCCCNRS established many working groups; most of these were the responsibility of the Nuclear Regulatory Commission, as far as the US participation was concerned. The lone exception was Working Group 7 on Environmental Transport and Health Effects, for which the US participation was the responsibility of the US Department of Energy (DOE). The purpose of Working Group 7 was succintly stated to be, ``To develop jointly methods to project rapidly the health effects of any future nuclear reactor accident.`` To implement the work DOE then formed two subworking groups: 7.1 to address Environmental Transport and 7.2 to address Health Effects. Thus, the DOE-funded Chernobyl Studies Project began. The majority of the initial tasks for this project are completed or near completion. The focus is now turned to the issue of health effects from the Chernobyl accident. Currently, we are involved in and making progress on the case-control and co-hort studies of thyroid diseases among Belarussian children. Dosimetric aspects are a fundamental part of these studies. We are currently working to implement similar studies in Ukraine. A major part of the effort of these projects is supporting these studies, both by providing methods and applications of dose reconstruction and by providing support and equipment for the medical teams.

  16. Georgia Hosts Multi-Agency Waste Isolation Pilot Plant Transportation Exercise

    Broader source: Energy.gov [DOE]

    COVINGTON, Ga. – Emergency personnel throughout the U.S. who respond in the event of a potential accident involving radioactive waste shipments take part in mock training scenarios to help them prepare for an actual incident.

  17. Accident source terms for boiling water reactors with high burnup cores.

    SciTech Connect (OSTI)

    Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas

    2007-11-01

    The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.

  18. Plutonium contamination twenty years after the nuclear weapons accident in Spain

    SciTech Connect (OSTI)

    Iranzo, E.; Richmond, C.R.

    1987-01-01

    An accident involving two US Air Force planes engaged in a refueling operation occurred at 0922 GMT on January 17, 1966 over the town of Palomares in southeastern Spain. Three of the bombs, one intact, were found on land, in or near Palomares while the fourth was removed from the Mediterranean Sea. The parachutes of two of the bombs did not deploy resulting in the detonation of their conventional explosives and release of fissile material upon impact. Partial burning of the fissile material formed an aerosol that contaminated approximately 226 hectares of uncultivated, farmed, and urban land. The objective of this study was to determine the magnitude of the risk from internal contamination of the area inhabitants immediately after the accident and during the emergency phase and to determine the short, medium and long-term risk of internal contamination for the inhabitants of Palomares and its environs and to those who consume planet products cultivated in that area.

  19. A methodology for generating dynamic accident progression event trees for level-2 PRA

    SciTech Connect (OSTI)

    Hakobyan, A.; Denning, R.; Aldemir, T. [Ohio State Univ., Nuclear Engineering Program, 650 Ackerman Road, Columbus, OH 43202 (United States); Dunagan, S.; Kunsman, D. [Sandia National Laboratory, Albuquerque, NM 87185 (United States)

    2006-07-01

    Currently, the development and analysis of Accident Progression Event Trees (APETs) are performed in a manner that is computationally time consuming, difficult to reproduce and also can be phenomenologically inconsistent. A software tool (ADAPT) is described for automated APET generation using the concept of dynamic event trees. The tool determines the branching times from a severe accident analysis code based on user specified criteria for branching. It assigns user specified probabilities to every branch, tracks the total branch probability, and truncates branches based on the given pruning/truncation rules to avoid an unmanageable number of scenarios. While the software tool could be applied to any systems analysis code, the MELCOR code is used for this illustration. A case study is presented involving station blackout with the loss of auxiliary feedwater system for a pressurized water reactor. (authors)

  20. Natural Circulation in the Blanket Heat Removal System During a Loss-of-Pumping Accident (LOFA) Based on Initial Conceptual Design

    SciTech Connect (OSTI)

    Hamm, L.L.

    1998-10-07

    A transient natural convection model of the APT blanket primary heat removal (HR) system was developed to demonstrate that the blanket could be cooled for a sufficient period of time for long term cooling to be established following a loss-of-flow accident (LOFA). The particular case of interest in this report is a complete loss-of-pumping accident. For the accident scenario in which pumps are lost in both the target and blanket HR systems, natural convection provides effective cooling of the blanket for approximately 68 hours, and, if only the blanket HR systems are involved, natural convection is effective for approximately 210 hours. The heat sink for both of these accident scenarios is the assumed stagnant fluid and metal on the secondary sides of the heat exchangers.

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

    SciTech Connect (OSTI)

    Hanson, D.J.; Johnson, S.P.; Blackman, H.S.; Stewart, M.A. (EG and G Idaho, Inc., Idaho Falls, ID (United States))

    1992-07-01

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

  2. Assessment of light water reactor accident management programs and experience

    SciTech Connect (OSTI)

    Hammersley, R.J. [Fauske and Associates, Inc., Burr Ridge, IL (United States)

    1992-03-01

    The objective of this report is to provide an assessment of the current light water reactor experience regarding accident management programs and associated technology developments. This assessment for light water reactor (LWR) designs is provided as a resource and reference for the development of accident management capabilities for the production reactors at the Savannah River Site. The specific objectives of this assessment are as follows: 1. Perform a review of the NRC, utility, and industry (NUMARC, EPRI) accident management programs and implementation experience. 2. Provide an assessment of the problems and opportunities in developing an accident management program in conjunction or following the Individual Plant Examination process. 3. Review current NRC, utility, and industry technological developments in the areas of computational tools, severe accident predictive tools, diagnostic aids, and severe accident training and simulation.

  3. Non-Tracial Free Transport and Applications

    E-Print Network [OSTI]

    Nelson, Brent Andrew

    2015-01-01

    tracial transport . . . . . . . . . . . . . . . . . . . .the transport element . . . . . . . . . . . . . .Free Transport . . . . . . . . . . . .

  4. Use of probabilistic safety analyses in severe accident management

    SciTech Connect (OSTI)

    Neogy, P.; Lehner, J.

    1991-01-01

    An important consideration in the development and assessment of severe accident management strategies is that while the strategies are often built on the knowledge base of Probabilistic Safety Analyses (PSA), they must be interpretable and meaningful in terms of the control room indicators. In the following, the relationships between PSA and severe accident management are explored using ex-vessel accident management at a PWR ice-condenser plant as an example. 2 refs., 1 fig., 3 tabs.

  5. Baseline requirements of the proposed action for the Transportation Management Division routing models

    SciTech Connect (OSTI)

    Johnson, P.E.; Joy, D.S.

    1995-02-01

    The potential impacts associated with the transportation of hazardous materials are important to shippers, carriers, and the general public. This is particularly true for shipments of radioactive material. The shippers are primarily concerned with safety, security, efficiency, and equipment requirements. The carriers are concerned with the potential impact that radioactive shipments may have on their operations--particularly if such materials are involved in an accident. The general public has also expressed concerns regarding the safety of transporting radioactive and other hazardous materials through their communities. Because transportation routes are a central concern in hazardous material transport, the prediction of likely routes is the first step toward resolution of these issues. In response to these routing needs, several models have been developed over the past fifteen years at Oak Ridge National Laboratory (ORNL). The HIGHWAY routing model is used to predict routes for truck transportation, the INTERLINE routing model is used to predict both rail and barge routes, and the AIRPORT locator model is used to determine airports with specified criteria near a specific location. As part of the ongoing improvement of the US Department of Energy`s (DOE) Environmental Management Transportation Management Division`s (EM-261) computer systems and development efforts, a Baseline Requirements Assessment Session on the HIGHWAY, INTERLINE, and AIRPORT models was held at ORNL on April 27, 1994. The purpose of this meeting was to discuss the existing capabilities of the models and data bases and to review enhancements of the models and data bases to expand their usefulness. The results of the Baseline Requirements Assessment Section will be discussed in this report. The discussions pertaining to the different models are contained in separate sections.

  6. Type B Accident Investigation Board Report of the Brookhaven...

    Energy Savers [EERE]

    of the Brookhaven National Laboratory Employee Injury at Building 1005H on October 9, 2009 Type B Accident Investigation Board Report of the Brookhaven National Laboratory Employee...

  7. Type A Accident Investigation of the March 16, 2000, Plutonium...

    Energy Savers [EERE]

    Multiple Intake Event at the Plutonium Facility, Los Alamos National Laboratory, New Mexico Type A Accident Investigation of the March 16, 2000, Plutonium-238 Multiple Intake...

  8. Type B Accident Investigation Report on the Exertional Heat Illnesses...

    Energy Savers [EERE]

    Heat Illnesses during SPOTC 2006 at the National Training Center in Albuquerque, New Mexico, July 13, 2006 Type B Accident Investigation Report on the Exertional Heat Illnesses...

  9. Type B Accident Investigation Board Report of the September 29...

    Energy Savers [EERE]

    of the September 29, 2010, Radiological Contamination Event at the Separations Process Research Unit (SPRU), Building H2 Demolition, in Niskayuna, New, York Type B Accident...

  10. Accident Investigation of the August 21, 2012, Contamination...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    August 21, 2012, Contamination Incident at the Los Alamos Neutron Science Center at the Los Alamos National Laboratory Accident Investigation of the August 21, 2012, Contamination...

  11. Type B Accident Investigation of the October 9, 2008 Employee...

    Energy Savers [EERE]

    October 9, 2008 Employee Injured when Rocket Motor Unexpectedly Fired at the Sandia National Laboratories Technical Area III Sled Track, Sandia Site Office Type B Accident...

  12. Type B Accident Investigation Board Report for the January 11...

    Broader source: Energy.gov (indexed) [DOE]

    a serious injury to his right hand while operating a table saw. In conducting its investigation, the Accident Investigation Board (the Board) used various analytical techniques,...

  13. Type B Accident Investigation Board Report on the October 15...

    Energy Savers [EERE]

    15, 2001, Grout Injection Operator Injury at the Cold Test Pit South, Idaho National Engineering and Environmental Laboratory Type B Accident Investigation Board Report on the...

  14. Accident Investigation of the October 1, 2013, Tice Electric...

    Broader source: Energy.gov (indexed) [DOE]

    22, 2013 On October 2, 2013, at the request of the Bonneville Power Administration (BPA) Chief Safety Officer, a Level I Accident Investigation was convened to investigate an...

  15. Accident Investigation of the September 20, 2012 Fatal Fall from...

    Energy Savers [EERE]

    20, 2012 Fatal Fall from the Dworshak-Taft 1 Transmission Tower, at the Bonneville Power Marketing Administration Accident Investigation of the September 20, 2012 Fatal Fall...

  16. Type B Accident Investigation At Washington Closure Hanford,...

    Office of Environmental Management (EM)

    part of a Washington Closure Hanford, LLC (WCH) team of craft personnel preparing a bridge crane for removal from the 336 Building. Type B Accident Investigation At Washington...

  17. Type B Accident Investigation Board Report on the October 15...

    Office of Environmental Management (EM)

    Type B Accident Investigation Board Report on the October 15, 2001, Grout Injection Operator Injury at the Cold Test Pit South, Idaho National Engineering and Environmental...

  18. Type A Accident Investigation Board Report on the February 20...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    February 20, 1996, Fall Fatality at the Radioactive Waste Management Complex Transuranic Storage Area - Retrieval Enclosure, Idaho National Engineering Laboratory Type A Accident...

  19. Type B Accident Investigation Board Report, May 8, 2004, Exothermic...

    Broader source: Energy.gov (indexed) [DOE]

    August 17, 2004 On May 8, 2004, at approximately 11:00 am, an exothermic metal reaction (exothermic reaction) accident occurred during heating of surplus activated sodium shields...

  20. Estimating Pedestrian Accident Exposure: Automated Pedestrian Counting Devices Report

    E-Print Network [OSTI]

    Bu, Fanping; Greene-Roesel, Ryan; Diogenes, Mara Chagas; Ragland, David R

    2007-01-01

    pp. 283-291. Estimating Pedestrian Accident Exposure: Draftand J. Thiran. Counting Pedestrians in Video Sequences UsingPartnership (CLP) Automatic Pedestrian Counting Trial. Stage

  1. Type A Accident Investigation of the June 21, 2001, Drilling...

    Office of Environmental Management (EM)

    June 21, 2001, Drilling Rig Operator Injury at the Fermi National Accelerator Laboratory, August 2001 Type A Accident Investigation of the June 21, 2001, Drilling Rig Operator...

  2. Improvement of Design Codes to Account for Accident Thermal Effects...

    Office of Environmental Management (EM)

    IMPROVEMENT OF DESIGN CODES TO ACCOUNT FOR ACCIDENT THERMAL EFFECTS ON SEISMIC PERFORMANCE Amit H. Varma, Kadir Sener, Saahas Bhardwaj Purdue University Andrew Whittaker: Univ. of...

  3. Type B Accident Investigation Board Report of the Bechtel Jacobs...

    Office of Environmental Management (EM)

    Jacobs Company, LLC Employee Fall Injury on January 3, 2006, at the K-25 Building, East Tennessee Technology Park, Oak Ridge, Tennessee Type B Accident Investigation Board...

  4. Hazard Categorization and Accident Analysis Techniques for Compliance...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports by Diane Johnson he purpose of this DOE Standard is to...

  5. Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

    SciTech Connect (OSTI)

    Su'ud, Zaki; Anshari, Rio

    2012-06-06

    Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.

  6. PARTICIPANT!INFORMATION!SHEET! Transportation*of*children*with*bicycle*seats,*trailers,*and*other*carriers:*

    E-Print Network [OSTI]

    Balasuriya, Sanjeeva

    ! ! PARTICIPANT!INFORMATION!SHEET! Transportation*of*children*with*bicycle%Adelaide%and%the%Monash%University%Accident%Research%Centre%(MUARC).% The%aim%of%the%project%is%to%explore%the%issues%surrounding%the%transportation%of%young%children%on%bicycles%type%of%carrier%designed%for%passengers%("dinkying"%on%the% handlebars%does%not%count).%% % Bicycle%carriers%such%as%frontKmounted%seats

  7. Heat transport within the Earth

    E-Print Network [OSTI]

    Herndon, J Marvin

    2011-01-01

    Numerous attempts have been made to interpret Earth's dynamic processes based upon heat transport concepts derived from ordinary experience. But, ordinary experience can be misleading, especially when underlain by false assumptions. Geodynamic considerations traditionally have embraced three modes of heat transport: conduction, convection, and radiation. Recently, I introduced a fourth, "mantle decompression thermal tsunami" that, I submit, is responsible for emplacing heat at the base of the Earth's crust. Here, I review thermal transport within the Earth and speculate that there might be a fifth mode: "heat channeling", involving heat transport from the core to "hot-spots" such as those that power the Hawaiian Islands and Iceland.

  8. Assessment of two BWR accident management strategies

    SciTech Connect (OSTI)

    Hodge, S.A.; Petek, M.

    1991-01-01

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

  9. Material Selection for Accident Tolerant Fuel Cladding

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    none,

    2014-07-01

    Alternative cladding materials are being investigated for accident tolerance, which can be defined as >100X improvement (compared to current Zr-based alloys) in oxidation resistance in steam environments at ?1200°C for short (?4 h) times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases and FeCrAl alloys. Recently reported low mass losses for Mo in steam at 800°C could not be reproduced. Both FeCrAl and MAX phase Ti2AlC form a protective alumina scale in steam. However, commercial Ti2AlC that is not single phase, formed a much thicker oxide at 1200°C in steam and significant TiO2, and therefore may be challenging to use as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1475°C, while reducing its Cr content to minimize susceptibility to irradiation assisted Cr-rich ?’ formation. The composition effects and critical limits to retaining protective scale formation at >1400°C are still being evaluated. Keywords: Accident tolerant LWR Fuel cladding, FeCrAl, Mo, Ti2AlC, Al2O3, high temperature steam oxidation resistance

  10. Web Based Course: SAF-230DE, Accident Investigation Overview Promotional Video

    Broader source: Energy.gov [DOE]

    This course that provides an overview of the fundamentals of accident investigation. The course is intended to meet the every five year refresher training requirement for DOE Federal Accident Investigators under DOE O 225.1B, Accident Investigations.

  11. Enforcement Guidance Supplement 98-02: DOE Enforcement Activities where Off-site Transportation Issues are also Present.

    Broader source: Energy.gov [DOE]

    Recently several questions have arisen regarding the scope of Price-Anderson enforcement when transportation issues are directly or indirectly involved in an incident. These questions can be separated into two areas, (1) transportation issues that involve on-site transportation typically not regulated by the Department of Transportation (DOT), and (2) transportation issues that involve off-site transportation. This guidance addresses off-site transportation that is regulated by DOT and other state and federal agencies.

  12. CASE STUDY FOR ENHANCED ACCIDENT TOLERANCE DESIGN CHANGES

    SciTech Connect (OSTI)

    Prescott, Steven; Smith, Curtis; Koonce, Tony

    2014-09-01

    The ability to better characterize and quantify safety margin is important to improved decision making about Light Water Reactor (LWR) design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margin management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. In addition, as research and development in the LWR Sustainability (LWRS) Program and other collaborative efforts yield new data, sensors, and improved scientific understanding of physical processes that govern the aging and degradation of plant SSCs needs and opportunities to better optimize plant safety and performance will become known. To support decision making related to economics, readability, and safety, the Risk Informed Safety Margin Characterization (RISMC) Pathway provides methods and tools that enable mitigation options known as risk informed margins management (RIMM) strategies. The methods and tools provided by RISMC are essential to a comprehensive and integrated RIMM approach that supports effective preservation of margin for both active and passive SSCs. In this report, we discuss the methods and technologies behind RIMM for an application focused on enhanced accident tolerance design changes for a representative nuclear power plant. We look at a variety of potential plant modifications and evaluate, using the RISMC approach, the implications to safety margin for the various strategies.

  13. OFFICE OF RISK MANAGEMENT STUDENT ACCIDENT INVESTIGATION REPORT

    E-Print Network [OSTI]

    Arnold, Elizabeth A.

    OFFICE OF RISK MANAGEMENT STUDENT ACCIDENT INVESTIGATION REPORT James Madison University, Office of Risk Management, 131 West Grace Street, MSC 6703 Harrisonburg, VA 22807, Phone: 540-568-7812, Fax: 540/Injury: Student's Signature: Date: Return Original Form to: #12;OFFICE OF RISK MANAGEMENT STUDENT ACCIDENT

  14. Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

    SciTech Connect (OSTI)

    CROWE, R.D.; PIEPHO, M.G.

    2000-03-23

    This document provided the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report''. All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

  15. Cold Vacuum Drying (CVD) Facility Design Basis Accident Analysis Documentation

    SciTech Connect (OSTI)

    PIEPHO, M.G.

    1999-10-20

    This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report, ''Cold Vacuum Drying Facility Final Safety Analysis Report (FSAR).'' All assumptions, parameters and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR.

  16. BWR containment failure analysis during degraded-core accidents

    SciTech Connect (OSTI)

    Yue, D.D.

    1982-06-06

    This paper presents a containment failure mode analysis during a spectrum of postulated degraded core accident sequences in a typical 1000-MW(e) boiling water reactor (BWR) with a Mark-I wetwell containment. Overtemperature failure of containment electric penetration assemblies (CEPAs) has been found to be the major failure mode during such accidents.

  17. Accident source terms for Light-Water Nuclear Power Plants. Final report

    SciTech Connect (OSTI)

    Soffer, L.; Burson, S.B.; Ferrell, C.M.; Lee, R.Y.; Ridgely, J.N.

    1995-02-01

    In 1962 tile US Atomic Energy Commission published TID-14844, ``Calculation of Distance Factors for Power and Test Reactors`` which specified a release of fission products from the core to the reactor containment for a postulated accident involving ``substantial meltdown of the core``. This ``source term``, tile basis for tile NRC`s Regulatory Guides 1.3 and 1.4, has been used to determine compliance with tile NRC`s reactor site criteria, 10 CFR Part 100, and to evaluate other important plant performance requirements. During the past 30 years substantial additional information on fission product releases has been developed based on significant severe accident research. This document utilizes this research by providing more realistic estimates of the ``source term`` release into containment, in terms of timing, nuclide types, quantities and chemical form, given a severe core-melt accident. This revised ``source term`` is to be applied to the design of future light water reactors (LWRs). Current LWR licensees may voluntarily propose applications based upon it.

  18. The Fuel Accident Condition Simulator (FACS) furnace system for high temperature performance testing of VHTR fuel

    SciTech Connect (OSTI)

    Paul A. Demkowicz; David V. Laug; Dawn M. Scates; Edward L. Reber; Lyle G. Roybal; John B. Walter; Jason M. Harp; Robert N. Morris

    2012-10-01

    The AGR-1 irradiation of TRISO-coated particle fuel specimens was recently completed and represents the most successful such irradiation in US history, reaching peak burnups of greater than 19% FIMA with zero failures out of 300,000 particles. An extensive post-irradiation examination (PIE) campaign will be conducted on the AGR-1 fuel in order to characterize the irradiated fuel properties, assess the in-pile fuel performance in terms of coating integrity and fission metals release, and determine the fission product retention behavior during high temperature safety testing. A new furnace system has been designed, built, and tested to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000 degrees C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, and Eu), iodine, and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator furnace system and the associated fission gas monitoring system, as well as preliminary system calibration results.

  19. Microsoft Word - 2015.06.22 - Report to Congress - Accident Tolerant...

    Energy Savers [EERE]

    OF LWR FUELS WITH ENHANCED ACCIDENT TOLERANCE Page i Development of Light Water Reactor Fuels with Enhanced Accident Tolerance Report to Congress April 2015 United States...

  20. MELCOR accident analysis for ARIES-ACT

    SciTech Connect (OSTI)

    Paul W. Humrickhouse; Brad J. Merrill

    2012-08-01

    We model a loss of flow accident (LOFA) in the ARIES-ACT1 tokamak design. ARIES-ACT1 features an advanced SiC blanket with LiPb as coolant and breeder, a helium cooled steel structural ring and tungsten divertors, a thin-walled, helium cooled vacuum vessel, and a room temperature water-cooled shield outside the vacuum vessel. The water heat transfer system is designed to remove heat by natural circulation during a LOFA. The MELCOR model uses time-dependent decay heats for each component determined by 1-D modeling. The MELCOR model shows that, despite periodic boiling of the water coolant, that structures are kept adequately cool by the passive safety system.

  1. Big Rock Point severe accident management strategies

    SciTech Connect (OSTI)

    Brogan, B.A. [Consumers Power Co., Charlevoix, MI (United States); Gabor, J.R. [Dames and Moore, Westmont, IL (United States)

    1996-07-01

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

  2. Guidelines for accident prevention and emergency preparedness

    SciTech Connect (OSTI)

    Fthenakis, V.M.; Morris, S.C.; Moskowitz, P.D.

    1993-05-01

    This report reviews recent developments in the guidelines on chemical accident prevention, risk assessment, and management of chemical emergencies, principally in the United States and Europe, and discusses aspects of their application to developing countries. Such guidelines are either in the form of laws or regulations promulgated by governments, or of recommendations from international, professional, or non governmental organizations. In many cases, these guidelines specify lists of materials of concern and methods for evaluating safe usage of these materials and recommend areas of responsibility for different organizations; procedures to be included in planning, evaluation, and response; and appropriate levels of training for different classes of workers. Guidelines frequently address the right of communities to be informed of potential hazards and address ways for them to participate in planning and decision making.

  3. Sustainable Transportation

    SciTech Connect (OSTI)

    2012-09-01

    This document highlights DOE's Office of Energy Efficiency and Renewable Energy's advancements in transportation technologies, alternative fuels, and fuel cell technologies.

  4. Calculation notes in support of TWRS FSAR spray leak accident analysis

    SciTech Connect (OSTI)

    Hall, B.W.

    1996-09-25

    This document contains the detailed calculations that support the spray leak accident analysis in the Tank Waste Remediation System (TWRS) Final Safety Analysis Report (FSAR). The consequence analyses in this document form the basis for the selection of controls to mitigate or prevent spray leaks throughout TWRS. Pressurized spray leaks can occur due to a breach in containment barriers along transfer routes, during waste transfers. Spray leaks are of particular safety concern because, depending on leak dimensions, and waste pressure, they can be relatively efficient generators of dispersible sized aerosols that can transport downwind to onsite and offsite receptors. Waste is transferred between storage tanks and between processing facilities and storage tanks in TWRS through a system of buried transfer lines. Pumps for transferring waste and jumpers and valves for rerouting waste are located inside below grade pits and structures that are normally covered. Pressurized spray leaks can emanate to the atmosphere due to breaches in waste transfer associated equipment inside these structures should the structures be uncovered at the time of the leak. Pressurized spray leaks can develop through holes or cracks in transfer piping, valve bodies or pump casings caused by such mechanisms as corrosion, erosion, thermal stress, or water hammer. Leaks through degraded valve packing, jumper gaskets, or pump seals can also result in pressurized spray releases. Mechanisms that can degrade seals, packing and gaskets include aging, radiation hardening, thermal stress, etc. An1782other common cause for spray leaks inside transfer enclosures are misaligned jumpers caused by human error. A spray leak inside a DST valve pit during a transfer of aging waste was selected as the bounding, representative accident for detailed analysis. Sections 2 through 5 below develop this representative accident using the DOE- STD-3009 format. Sections 2 describes the unmitigated and mitigated accident scenarios evaluated to determine the need for safety class SSCs or TSR controls. Section 3 develops the source terms associated with the unmitigated and mitigated accident scenarios. Section 4 estimates the radiological and toxicological consequences for the unmitigated and mitigated scenarios. Section 5 compares the radiological and toxicological consequences against the TWRS evaluation guidelines. Section 6 extrapolates from the representative accident case to other represented spray leak sites to assess the conservatism in using the representative case to define controls for other postulated spray leak sites throughout TWRS. Section 7 discusses the sensitivities of the consequence analyses to the key parameters and assumptions used in the analyses. Conclusions are drawn in Section 8. The analyses herein pertain to spray leaks initiated due to internal mechanisms (e.g., corrosion, erosion, thermal stress, etc). External initiators of spray leaks (e.g., excavation accidents), and natural phenomena initiators (e.g., seismic events) are to be covered in separate accident analyses.

  5. Accident management then and now: Progress since Di Salvo's work

    SciTech Connect (OSTI)

    Shotkin, L.M.

    1991-01-01

    The nuclear industry is now initiating a serious effort to define the elements of an accident management program at each utility with an operating reactor, which is a significant change in conditions from those in 1985, when the work of Di Salvo et al. was published. Each utility is now conducting an individual plant examination (IPE) to uncover plant vulnerabilities to severe accidents. In conjunction with the IPE program, the Nuclear Utility Management and Resources committee, the Electric Power Research Institute, and owners' groups are developing an accident management program. This program is emphasizing the management program. This program is emphasizing the management of severe accidents (i.e., accidents that proceed to significant core melt) including strategies for managing ex-vessel events. Attention is also being paid to interfacing any severe accident management strategies with existing emergency operating procedures already in place at utilities. The industry program is addressing the five elements define by the US Nuclear Regulatory Commission (NRC): (1) strategies; (2) instrumentation; (3) guidance and computational aids; (4) organization and decision making; and (5) training. It will also be able to accept new information as it becomes available from ongoing efforts to better understand severe accidents and how to manage them effectively.

  6. Post-accident inhalation exposure and experience with plutonium

    SciTech Connect (OSTI)

    Shinn, J

    1998-06-01

    This paper addresses the issue of inhalation exposure immediately afterward and for a long time following a nuclear accident. For the cases where either a nuclear weapon burns or explodes prior to nuclear fission, or at locations close to a nuclear reactor accident containing fission products, a major concern is the inhalation of aerosolized plutonium (Pu) particles producing alpha-radiation. We have conducted field studies of Pu- contaminated real and simulated accident sites at Bikini, Johnston Atoll, Tonopah (Nevada), Palomares (Spain), Chernobyl, and Maralinga (Australia).

  7. The Accident at Fukushima: What Happened?

    SciTech Connect (OSTI)

    Fujie, Takao

    2012-07-01

    At 2:46 PM, on the coast of the Pacific Ocean in eastern Japan, people were spending an ordinary afternoon. The earthquake had a magnitude of 9.0, the fourth largest ever recorded in the world. Avery large number of aftershocks were felt after the initial earthquake. More than 100 of them had a magnitude of over 6.0. There were very few injured or dead at this point. The large earthquake caused by this enormous crustal deformation spawned a rare and enormous tsunami that crashed down 30-40 minutes later. It easily cleared the high levees, washing away cars and houses and swallowing buildings of up to three stories in height. The largest tsunami reading taken from all regions was 40 meters in height. This tsunami reached the West Coast of the United States and the Pacific coast of South America, with wave heights of over two meters. It was due to this tsunami that the disaster became one of a not imaginable scale, which saw the number of dead or missing reach about 20,000 persons. The enormous tsunami headed for 15 nuclear power plants on the Pacific coast, but 11 power plants withstood the tsunami and attained cold shutdown. The flood height of the tsunami that struck each power station ranged to a maximum of 15 meters. The Fukushima Daiichi Nuclear Power Plant Units experienced the largest and the cores of three reactors suffered meltdown. As a result, more than 160,000 residents were forced to evacuate, and are still living in temporary accommodation. The main focus of this presentation is on what happened at the Fukushima Daiichi, and how station personnel responded to the accident, with considerable international support. A year after the Fukushima Daiichi accident, Japan is in the process of leveraging the lessons learned from the accident to further improve the safety of nuclear power facilities and regain the trust of society. In this connection, not only international organizations, including IAEA, and WANO, but also governmental organizations and nuclear industry representatives from various countries, have been evaluating what happened at Fukushima Daiichi. Support from many countries has contributed to successfully stabilizing the Fukushima Daiichi Nuclear Power Station. International cooperation is required as Japan started along the long road to decommissioning the reactors. Such cooperation with the international community would achieve the decommissioning of the damaged reactors. Finally, recovery plans by the Japanese government to decontaminate surrounding regions have been started in order to get residents back to their homes as early as possible. Looking at the world's nuclear power industry, there are currently approximately 440 reactors in operation and 60 under construction. Despite the dramatic consequences of the Fukushima Daiichi catastrophe it is expected that the importance of nuclear power generation will not change in the years to come. Newly accumulated knowledge and capabilities must be passed on to the next generation. This is the duty put upon us and which is one that we must embrace.

  8. Exploring transport and phase behavior in nanoporous carbon materials

    E-Print Network [OSTI]

    Shimizu, Steven (Steven Franklin Esau)

    2015-01-01

    Understanding transport and phase behavior in nanopores has a substantial impact on applications involving membrane fabrication, single-molecule detection, oil reservoir modeling, and drug delivery. While transport and ...

  9. Virus transport in physically and geochemically heterogeneous subsurface porous media

    E-Print Network [OSTI]

    Ryan, Joe

    Virus transport in physically and geochemically heterogeneous subsurface porous media Subir for virus transport in physically and geochemically heterogeneous subsurface porous media is presented. The model involves solution of the advection­dispersion equation, which additionally considers virus

  10. Risk Assessment of Railway Transportation Systems using Timed Fault Trees

    E-Print Network [OSTI]

    Miller, Alice

    by the GRAIL (GNSS Introduction in the RAIL sector) project that is under development cooperated with ERTMS difference between the GRAIL and current railway systems is that it involves unmanned operation. Trains to eliminate the fault, to prevent the accident. There are still many unsolved problems related to GRAIL

  11. Type B Accident Investigation Board Report of the Savannah River...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Savannah River Site Hand Injury at the Salt Waste Processing Facility on October 6, 2009 Type B Accident Investigation Board Report of the Savannah River Site Hand Injury at the...

  12. Accident Investigation of the February 5, 2014, Underground Salt...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    5, 2014, Underground Salt Haul Truck Fire at the Waste Isolation Pilot Plant, Carlsbad NM Accident Investigation of the February 5, 2014, Underground Salt Haul Truck Fire at the...

  13. Type B Accident Investigation of the Savannah River Site Arc...

    Energy Savers [EERE]

    H2 Demolition, in Niskayuna, New, York Type B Accident Investigation Board Report of the Savannah River Site Hand Injury at the Salt Waste Processing Facility on October 6, 2009...

  14. Type B Accident Investigation of the Arc Flash at Brookhaven...

    Broader source: Energy.gov (indexed) [DOE]

    event and causal factor analysis. Type B Accident Investigation of the Arc Flash at Brookhaven National Laboratory, April 14, 2006 More Documents & Publications DOE-HDBK-1092-1998...

  15. Type B Accident Investigation of the January 10, 2006, Flash...

    Energy Savers [EERE]

    Review, Savannah River National Laboratory - January 2012 Type B Accident Investigation of the Arc Flash at Brookhaven National Laboratory, April 14, 2006 Audit Report: OAS-L-06-15...

  16. Type B Accident Investigation of the Acid Vapor Inhalation on...

    Energy Savers [EERE]

    of the Acid Vapor Inhalation on June 7, 2005, in TA-48, Building RC-1 Room 402 at the Los Alamos National Laboratory Type B Accident Investigation of the Acid Vapor Inhalation on...

  17. The Effect of Removing Accidents Repeaters From the Road

    E-Print Network [OSTI]

    Ferreira, Joseph Jr.

    In newspaper editorials, public commentaries and the like, licensing authorities are often advised to solve the "accident problem" by taking the "nut behind the wheel" off the road. This paper uses six-year driver records ...

  18. Failsafe : living with man-made disaster and accident

    E-Print Network [OSTI]

    Higgins, Saoirse, 1966-

    2004-01-01

    "There is no progress with out progress of the catastrophe." Virilio. This thesis project proposes that technological solutions in the design of our systems are not enough to prevent 'man-made' accident. Social, organisational ...

  19. The 2011 Tohoku earthquake, tsunami, and Fukushima nuclear accident

    E-Print Network [OSTI]

    Ferrari, Silvia

    The 2011 Tohoku earthquake, tsunami, and Fukushima nuclear accident: the Risk Policy Aftermath 3 #12;Personal experience in March 2011 Tsukuba 170km Tokyo 230km Fukushima Daiichi nuclear power

  20. Accident Investigation of the July 30, 2013, Electrical Fatality...

    Energy Savers [EERE]

    July 30, 2013, Electrical Fatality on the Bandon-Rogue No. 1 115kV Line at the Bonneville Power Administration Accident Investigation of the July 30, 2013, Electrical Fatality on...

  1. Type B Accident Investigation Board Report Grout Injection Operator...

    Energy Savers [EERE]

    and no damage to any structures inside the calvareum (i.e., no evidence of brain injury). Page 16 2.4. Investigation Readiness and Accident Scene Preservation The...

  2. Some methods of estimating uncertainty in accident reconstruction

    E-Print Network [OSTI]

    Milan Batista

    2011-07-20

    In the paper four methods for estimating uncertainty in accident reconstruction are discussed: total differential method, extreme values method, Gauss statistical method, and Monte Carlo simulation method. The methods are described and the program solutions are given.

  3. Modeling control room crews for accident sequence analysis

    E-Print Network [OSTI]

    Huang, Y. (Yuhao)

    1991-01-01

    This report describes a systems-based operating crew model designed to simulate the behavior of an nuclear power plant control room crew during an accident scenario. This model can lead to an improved treatment of potential ...

  4. Type B Accident Investigation Board Report BNFL, Inc. Employee...

    Broader source: Energy.gov (indexed) [DOE]

    17, 2003, at approximately 7:15 a.m., an accident occurred at the U.S. Department of Energy (DOE) East Tennessee Technology Park, Building K-31. An employee (Pipefitter) of...

  5. Material selection for accident tolerant fuel cladding

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Pint, B. A.; Terrani, K. A.; Yamamoto, Y.; Snead, L. L.

    2015-09-14

    Alternative cladding materials are being investigated for accident tolerance, which can be defined as >100X improvement (compared to current Zr-based alloys) in oxidation resistance in steam environments at ?1200°C for short (?4 h) times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases and FeCrAl alloys. Recently reported low mass losses for Mo in steam at 800°C could not be reproduced. Both FeCrAl and MAX phase Ti2AlC form a protective alumina scale in steam. Therefore, commercial Ti2AlC that is not single phase, formed a much thicker oxide at 1200°C in steammore »and significant TiO2, and therefore may be challenging to use as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1475°C, while reducing its Cr content to minimize susceptibility to irradiation assisted Cr-rich ?’ formation. The composition effects and critical limits to retaining protective scale formation at >1400°C are still being evaluated.« less

  6. A commentary on the 1995 DOT/NRC amendments to the U.S. nuclear transportation regulations

    SciTech Connect (OSTI)

    Grella, A.

    1996-07-01

    This article discusses the major revisions (1995 DOT/NRC ammendments) to the US Nuclear Transportation regulations and their probable impacts on transportation. Areas covered include the following: the LSA and SCO definitions and packaging; radiation protection programs; mandatory use of SI units; changes an additions to the table of A1/A2 radionuclide values; and additional type B package hypothetical accident parameters.

  7. Study on drywell cooler applicability to severe accident management

    SciTech Connect (OSTI)

    Nakagawa, Takahiro [Information and manufacturing systems division, Toshiba Plant Systems and Services Corporation, 8, Shinsugita-cho, Isogo-ku, Yokohama, 235-8523 (Japan); Akinaga, Makoto [Power and Industrial Systems R and D Center, Toshiba Corporation, 4-1 Ukishima-cho, Kawasaki-ku, Kawasaki, 210-0862 (Japan); Hamazaki, Ryoichi [Isogo Nuclear Engineering Center, Toshiba Corporation, 8, Shinsugita-cho, Isogo-ku, Yokohama, 235-8523 (Japan); Matsuo, Toshihiro [Nuclear Power Engineering Department, Tokyo Electric Power Company, 1-3 Uchisaiwai-cho 1-chome, Chiyoda-ku, Tokyo 100-0011 (Japan); Hashimoto, Kouji [Nuclear Plant Engineering Department, HITACHI, Ltd., 1-1, Saiwai-cho, 3-chome, Hitachi-shi, Ibaraki-ken, 317-8511 (Japan)

    2004-07-01

    This paper concerns applicability of drywell cooler (DWC) heat removal under severe accident condition in BWR plants. Newly developed heat removal models based on DWC heat removal experiments were built into the MAAP3 code. And then, two types of Japanese BWR were selected to evaluate DWC heat removal performance under typical severe accident scenarios. According to the results of the evaluation, DWC delays or prevents containment failure or venting. (authors)

  8. Extension of emergency operating procedures for severe accident management

    SciTech Connect (OSTI)

    Chiang, S.C. (Taiwan Power Company, Taipei (Taiwan, Province of China))

    1992-01-01

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

  9. Transportation Plan 

    E-Print Network [OSTI]

    Boreo, Andrea; Li, Wei; Wunnenbuger, Douglas; Giusti, Cecilia; Cooper, John T.; Masterson, Jaimie

    2015-01-01

    Mobility throughout a community ensures freedom of movement and enhances quality of life. Traffic congestion, pollution, urban sprawl, social exclusion, safety and health can decrease mobility and should be a part of a sustainable transportation...

  10. electrifyingthefuture transportation

    E-Print Network [OSTI]

    Birmingham, University of

    programme of electrification and the potential introduction of diesel hybrids. The Department for Transport vehicles Wind turbine systems Industrial equipment The lab has full ethernet capability which will enable

  11. Type B Accident Investigation Board Report on the March 27, 1998, Rotating Shaft Accident at the Ames Laboratory, Ames, Iowa

    Broader source: Energy.gov [DOE]

    This report is an independent product of the Type B Accident Investigation Board appointed by John Kennedy, Acting Manager, Chicago Operations Office, U.S. Department of Energy (DOE).

  12. Multiscale thermal transport.

    SciTech Connect (OSTI)

    Graham, Samuel Jr. (; .); Wong, C. C.; Piekos, Edward Stanley

    2004-02-01

    A concurrent computational and experimental investigation of thermal transport is performed with the goal of improving understanding of, and predictive capability for, thermal transport in microdevices. The computational component involves Monte Carlo simulation of phonon transport. In these simulations, all acoustic modes are included and their properties are drawn from a realistic dispersion relation. Phonon-phonon and phonon-boundary scattering events are treated independently. A new set of phonon-phonon scattering coefficients are proposed that reflect the elimination of assumptions present in earlier analytical work from the simulation. The experimental component involves steady-state measurement of thermal conductivity on silicon films as thin as 340nm at a range of temperatures. Agreement between the experiment and simulation on single-crystal silicon thin films is excellent, Agreement for polycrystalline films is promising, but significant work remains to be done before predictions can be made confidently. Knowledge gained from these efforts was used to construct improved semiclassical models with the goal of representing microscale effects in existing macroscale codes in a computationally efficient manner.

  13. Policy on Programs Involving Minors Policy on Programs Involving

    E-Print Network [OSTI]

    Sridhar, Srinivas

    Policy on Programs Involving Minors 6/4/2014 Policy on Programs Involving Minors I. Purpose in University-sponsored activities that take place off campus. This policy applies to University or facilities that are operated or sponsored by third parties. This policy provides program registration

  14. Human factors review for Severe Accident Sequence Analysis (SASA)

    SciTech Connect (OSTI)

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

    1984-01-01

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

  15. Molecular Mechanism of Biological Proton Transport

    SciTech Connect (OSTI)

    Pomes, R.

    1998-09-01

    Proton transport across lipid membranes is a fundamental aspect of biological energy transduction (metabolism). This function is mediated by a Grotthuss mechanism involving proton hopping along hydrogen-bonded networks embedded in membrane-spanning proteins. Using molecular simulations, the authors have explored the structural, dynamic, and thermodynamic properties giving rise to long-range proton translocation in hydrogen-bonded networks involving water molecules, or water wires, which are emerging as ubiquitous H{sup +}-transport devices in biological systems.

  16. Characteristics and Behaviour of High Burnup Fuel that May Affect the Containment Source Terms for Cask Accidents

    SciTech Connect (OSTI)

    Einziger, Robert E.; Beyer, Carl E.

    2007-08-01

    Current risk assessments of spent fuel in storage and transportation casks use the properties of LWR fuel below 45 GWd/MTU. Fuel is being driven to higher burnups that may influence the source term in cask accidents. To achieve these burnups the manufacturers are introducing new assembly designs and cladding alloys. As a result, at the higher burnups (? 50 GWd/MTU) some of the characteristics of the fuel pellets, cladding, and assembly design used in the safety analysis have changed. The fuel pellet has developed a fine grained, Pu rich rim zone on its exterior surface. The source term may increase by 1 – 3 orders of magnitude depending on the fracture characteristics of the rim. The cladding may acquire hydrogen contents up to 700 wppm during the increased exposure. Embrittlement with subsequent lose of ductility may occur, especially if there is hydride reorientation. As a result, there may be a greater propensity for fracture of the rods upon impact with subsequent release of fuel particulate and gas. Significantly improved source terms can be developed if additional data on fuel rim fracture as a function of impact energy, the dependence of cladding ductility for Zircaloy and the newer cladding alloys as a function of hydride reorientation, and release characteristics for fractured rods were obtained. CRUD spallation characteristics only make a significant contribution to the source term if the rods do not fracture in the accident or if a fire only accident occurs.

  17. RESEARCH OR TEACHING INVOLVING ANIMALS

    E-Print Network [OSTI]

    Victoria, University of

    Page 1 RESEARCH OR TEACHING INVOLVING ANIMALS University Policy No: RH8110 Classification: Research Approving Authority: Vice-President Research Effective date: June, 2014 Supersedes: December, 2010 Last for Conducting Research or Teaching Involving Animals Animal Care Committee Terms of Reference PURPOSE 1

  18. Radiological Impact Assessment (RIA) following a postulated accident in PHWRS

    SciTech Connect (OSTI)

    Soni, N.; Kansal, M.; Rammohan, H. P.; Malhotra, P. K.

    2012-07-01

    Radiological Impact Assessment (RIA) following postulated accident i.e Loss of Coolant Accident (LOCA) with failed Emergency Core Cooling System (ECCS), performed as part of the reactor safety analysis of a typical 700 MWe Indian Pressurized Heavy Water Reactor(PHWR). The rationale behind the assessment is that the public needs to be protected in the event that the postulated accident results in radionuclide release outside containment. Radionuclides deliver dose to the human body through various pathways namely, plume submersion, exposure due to ground deposition, inhalation and ingestion. The total exposure dose measured in terms of total effective dose equivalent (TEDE) is the sum of doses to a hypothetical adult human at exclusion zone boundary by all the exposure pathways. The analysis provides the important inputs to decide upon the type of emergency counter measures to be adopted during the postulated accident. The importance of the various pathways in terms of contribution to the total effective dose equivalent(TEDE) is also assessed with respect to time of exposure. Inhalation and plume gamma dose are the major contributors towards TEDE during initial period of accident whereas ingestion and ground shine dose start dominating in TEDE in the extended period of exposure. Moreover, TEDE is initially dominated by I-131, Kr-88, Te-132, I-133 and Sr-89, whereas, as time progresses, Xe-133,I-131 and Te-132 become the main contributors. (authors)

  19. Accident Sequence Evaluation Program: Human reliability analysis procedure

    SciTech Connect (OSTI)

    Swain, A.D.

    1987-02-01

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

  20. Cognitive Radio will revolutionize American transportation

    ScienceCinema (OSTI)

    None

    2013-12-06

    Cognitive Radio will revolutionize American transportation. Through smart technology, it will anticipate user needs; detect available bandwidths and frequencies then seamlessly connect vehicles, infrastructures, and consumer devices; and it will support the Department of Transportation IntelliDrive Program, helping researchers, auto manufacturers, and Federal and State officials advance the connectivity of US transportation systems for improved safety, mobility, and environmental conditions. Using cognitive radio, a commercial vehicle will know its driver, onboard freight and destination route. Drivers will save time and resources communicating with automatic toll booths and know ahead of time whether to stop at a weigh station or keep rolling. At accident scenes, cognitive radio sensors on freight and transportation modes can alert emergency personnel and measure on-site, real-time conditions such as a chemical leak. The sensors will connect freight to industry, relaying shipment conditions and new delivery schedules. For industry or military purposes, cognitive radio will enable real-time freight tracking around the globe and its sensory technology can help prevent cargo theft or tampering by alerting shipper and receiver if freight is tampered with while en route. For the average consumer, a vehicle will tailor the transportation experience to the passenger such as delivering age-appropriate movies via satellite. Cognitive radio will enhance transportation safety by continually sensing what is important to the user adapting to its environment and incoming information, and proposing solutions that improve mobility and quality of life.

  1. Transportation needs assessment: Emergency response section

    SciTech Connect (OSTI)

    1989-05-01

    The transportation impacts of moving high level nuclear waste (HLNW) to a repository at Yucca Mountain in Nevada are of concern to the residents of the State as well as to the residents of other states through which the nuclear wastes might be transported. The projected volume of the waste suggests that shipments will occur on a daily basis for some period of time. This will increase the risk of accidents, including a catastrophic incident. Furthermore, as the likelihood of repository construction and operation and waste shipments increase, so will the attention given by the national media. This document is not to be construed as a willingness to accept the HLNW repository on the part of the State. Rather it is an initial step in ensuring that the safety and well-being of Nevada residents and visitors and the State`s economy will be adequately addressed in federal decision-making pertaining to the transportation of HLNW into and across Nevada for disposal in the proposed repository. The Preferred Transportation System Needs Assessment identifies critical system design elements and technical and social issues that must be considered in conducting a comprehensive transportation impact analysis. Development of the needs assessment and the impact analysis is especially complex because of the absence of information and experience with shipping HLNW and because of the ``low probability, high consequence`` aspect of the transportation risk.

  2. A Transportation Risk Assessment Tool for Analyzing the Transport of Spent Nuclear Fuel and High-Level Radioactive Waste to the Proposed Yucca Mountain Repository

    SciTech Connect (OSTI)

    Ralph Best; T. Winnard; S. Ross; R. Best

    2001-08-17

    The Yucca Mountain Transportation Database was developed as a data management tool for assembling and integrating data from multiple sources to compile the potential transportation impacts presented in the Draft Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada (DEIS). The database uses the results from existing models and codes such as RADTRAN, RISKIND, INTERLINE, and HIGHWAY to estimate transportation-related impacts of transporting spent nuclear fuel and high-level radioactive waste from commercial reactors and U. S. Department of Energy (DOE) facilities to Yucca Mountain. The source tables in the database are compendiums of information from many diverse sources including: radionuclide quantities for each waste type; route and route characteristics for rail, legal-weight truck, heavy haul. truck, and barge transport options; state-specific accident and fatality rates for routes selected for analysis; packaging and shipment data by waste type; unit risk factors; the complex behavior of the packaged waste forms in severe transport accidents; and the effects of exposure to radiation or the isotopic specific effects of radionclides should they be released in severe transportation accidents. The database works together with the codes RADTRAN (Neuhauser, et al, 1994) and RISKlND (Yuan, et al, 1995) to calculate incident-free dose and accident risk. For the incident-free transportation scenario, the database uses RADTRAN and RISKIND-generated data to calculate doses to offlink populations, onlink populations, people at stops, crews, inspectors, workers at intermodal transfer stations, guards at overnight stops, and escorts, as well as non-radioactive pollution health effects. For accident scenarios, the database uses RADTRAN-generated data to calculate dose risks based on ingestion, inhalation, resuspension, immersion (cloudshine), and groundshine as well as non-radioactive traffic fatalities. The Yucca Mountain EIS Transportation Database was developed using Microsoft Access 97{trademark} software and the Microsoft Windows NT{trademark} operating system. The database consists of tables for storing data, forms for selecting data for querying, and queries for retrieving the data in a predefined format. Database queries retrieve records based on input parameters and are used to calculate incident-free and accident doses using unit risk factors obtained from RADTRAN results. The next section briefly provides some background that led to the development of the database approach used in preparing the Yucca Mountain DEIS. Subsequent sections provide additional details on the database structure and types of impacts calculated using the database.

  3. Compendium of the Environmental Measurements Laboratory's research projects related to the Chernobyl nuclear accident

    SciTech Connect (OSTI)

    Volchok, H L; Chieco, N [comps.

    1986-10-01

    Following the accident at the Chernobyl nuclear reactor power station in the USSR on April 26, 1986, the Environmental Measurements Laboratory (EML) initiated a number of research projects as follows: (1) selected sites in both the Deposition and Surface Air networks were alerted and their sampling protocols adjusted to accommodate the anticipated arrival times and activity concentrations of the Chernobyl debris; (2) a number of cooperative programs involving field work, sampling, analysis and data interpretation were set up with institutions and scientists in other countries; (3) EML's Regional Baseline Station at Chester, NJ, as well as the roof of the Laboratory in New York City, provided bases for sampling and measurements to study the radionuclide concentrations, radiation levels, physical characteristics and potential biological implications of the Chernobyl fallout on the northeastern United States; and (4) the resulting fallout from the Chernobyl accident provided an 'experiment of opportunity' in that it enabled us to study fresh fission product deposition using collection systems resurrected from the 1950's and 1960's for comparison with current state-of-the-art methodology. The 13 reports of this volume have been entered separately into the data base.

  4. ATWS at Browns Ferry Unit One - accident sequence analysis

    SciTech Connect (OSTI)

    Harrington, R.M.; Hodge, S.A.

    1984-07-01

    This study describes the predicted response of Unit One at the Browns Ferry Nuclear Plant to a postulated complete failure to scram following a transient occurrence that has caused closure of all Main Steam Isolation Valves (MSIVs). This hypothetical event constitutes the most severe example of the type of accident classified as Anticipated Transient Without Scram (ATWS). Without the automatic control rod insertion provided by scram, the void coefficient of reactivity and the mechanisms by which voids are formed in the moderator/coolant play a dominant role in the progression of the accident. Actions taken by the operator greatly influence the quantity of voids in the coolant and the effect is analyzed in this report. The progression of the accident sequence under existing and under recommended procedures is discussed. For the extremely unlikely cases in which equipment failure and wrongful operator actions might lead to severe core damage, the sequence of emergency action levels and the associated timing of events are presented.

  5. Effect of shape reactivity on the rod-ejection accident

    SciTech Connect (OSTI)

    Neogy, P.; Carew, J.F.

    1982-09-01

    The shape reactivity has a significant influence on the rod ejection accident. After the control rod is fully ejected from the core, the neutron flux undergoes a large reduction at the ejected rod location. The corresponding effect on the control reactivity is comparable in magnitude to the Doppler reactivity, and makes a significant contribution to limiting the power excursion during the transient. The neglect of this effect in point kinetics and space time synthesis analyses of the rod ejection accident may account in part for the large degree of conservatism usually associated with these analyses.

  6. Truck transport of RAM: Risk effects of avoiding metropolitan areas

    SciTech Connect (OSTI)

    Mills, G.S.; Neuhauser, K.S.

    1997-11-01

    In the transport of radioactive material (RAM), e.g., spent nuclear fuel (SNF), stakeholders are generally most concerned about risks in high population density areas along transportation routes because of the perceived high consequences of potential accidents. The most significant portions of a transcontinental route and an alternative examined previously were evaluated again using population density data derived from US Census Block data. This method of characterizing population that adjoins route segments offers improved resolution of population density variations, especially in high population density areas along typical transport routes. Calculated incident free doses and accident dose risks for these routes, and the rural, suburban and urban segments are presented for comparison of their relative magnitudes. The results indicate that modification of this route to avoid major metropolitan areas through use of non-Interstate highways increases total risk yet does not eliminate a relatively small urban component of the accident dose risk. This conclusion is not altered by improved resolution of route segments adjoining high density populations.

  7. Preface: Nonclassical Transport

    E-Print Network [OSTI]

    Bolshov, L.

    2010-01-01

    models of solute transport in highly heterogeneous geologicSemenov. 2008b. Nonclassical transport processes in geologicand L. Matveev. 2008. Transport regimes and concentration

  8. Intelligent Transport Systems

    E-Print Network [OSTI]

    Deakin, Elizabeth; Frick, Karen Trapenberg; Skabardonis, Alexander

    2009-01-01

    in Sustainable Urban Transport: City Interview Synthesis (of Leeds, Institute for Transport Studies, forthcoming.I NTELLIGENT TRANSPORT SYSTEMS LINKING TECHNOLOGY AND

  9. Sustainability and Transport

    E-Print Network [OSTI]

    Gilbert, Richard

    2006-01-01

    Gilbert is a Toronto-based transport and energy consultantof the forthcoming book Transport Revolutions: Making theand substantial transition to transport systems based on

  10. Sandia Energy - Transportation Safety

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Transportation Safety Home Stationary Power Nuclear Fuel Cycle Nuclear Energy Safety Technologies Risk and Safety Assessment Transportation Safety Transportation SafetyTara...

  11. A planning problem combining calculus of variations and optimal transport

    E-Print Network [OSTI]

    Lachapelle, Aimé

    problem as cost. The decoupled case is more involved since it is nonlinear in the transport plan. Keywords/assignment problems have their modern roots in planning prob- lems (optimally transporting coal from mines to steel and will compare them. The coupled problem thus amounts to solve a standard optimal transport where the cost

  12. Simulations of Thermal and Oxygen Transport in UO2 Fuels Marius Stan1

    E-Print Network [OSTI]

    Mihaila, Bogdan

    Simulations of Thermal and Oxygen Transport in UO2 Fuels Marius Stan1 and Bogdan Mihaila2 1 Argonne properties of nuclear fuels and predicting the behavior of nuclear fuel elements (ceramic fuel pellets or metallic fuel rods) under normal and accident conditions are major challenges for the nuclear fuel

  13. Emergency response planning for railroad transportation related spills of oil or other hazardous materials 

    E-Print Network [OSTI]

    Reeder, Geoffrey Benton

    1995-01-01

    In December 1984 an unintentional release of poison gas from a chemical plant in Bhopal, India killed over 2,500 people. Thousands of others were injured. Although this material was not in transportation at the time, this accident raised public...

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

    SciTech Connect (OSTI)

    Paladino, D.; Guentay, S.; Andreani, M. [Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland); Tkatschenko, I.; Brinster, J.; Dabbene, F. [CEA, DEN, STMF, F-91191 Gif-sur-Yvette Cedex (France); Kelm, S.; Allelein, H. J. [Forschungszentrum Juelich, 52425 Juelich (Germany); Visser, D. C. [Nuclear Research and Consultancy Group, 1755 Le Petten (Netherlands); Benz, S.; Jordan, T. [Karlsruher Institut fuer Technologie, 76131 Karlsruhe (Germany); Liang, Z. [Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, ON L5K 1B2 (Canada); Porcheron, E.; Malet, J.; Bentaib, A. [Institut de Radioprotection et de Surete Nucleaire, 92269 Fontenay-aux-Roses (France); Kiselev, A.; Yudina, T.; Filippov, A. [Nuclear Safety Inst., Russian Academy of Sciences, Moscow 115191 (Russian Federation); Khizbullin, A.; Kamnev, M. [JSC Afrikantov OKB Mechanical Engineering, Nizhny Novgorod, 603074 (Russian Federation); Zaytsev, A.; Loukianov, A. [SSC RF-IPPE, 1 Bondarenko Sq., Obninsk, Kaluga Region, 249033 (Russian Federation)

    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). The 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)

  15. Type A Accident Investigation Board Report on the July 1, 2008...

    Office of Environmental Management (EM)

    July 1, 2008, of the Vehicle Fatality Accident-Western Area Power Marketing Administration Type A Accident Investigation Board Report on the July 1, 2008, of the Vehicle Fatality...

  16. Type A Accident Investigation Board Report on the April 3, 1995...

    Broader source: Energy.gov (indexed) [DOE]

    1995 The accident under investigation occurred on April 3, 1995, at approximately 10:46 a.m. As a result of the accident, a Wackenhut Services, Incorporated-Savannah River Site...

  17. The Effect of the 18-Year Old Drinking Age on Auto Accidents

    E-Print Network [OSTI]

    Cucchiaro, Stephen

    The effect of Massachusetts' reduced drinking age on auto accidents is examined by employing an interrupted time series analysis of monthly accident data covering the period January, 1969, through September 1973. The data ...

  18. Three dimensional effects in analysis of PWR steam line break accident

    E-Print Network [OSTI]

    Tsai, Chon-Kwo

    A steam line break accident is one of the possible severe abnormal transients in a pressurized water reactor. It is required to present an analysis of a steam line break accident in the Final Safety Analysis Report (FSAR) ...

  19. Calculation notes for surface leak resulting in pool, TWRS FSAR accident analysis

    SciTech Connect (OSTI)

    Hall, B.W.

    1996-09-25

    This document includes the calculations performed to quantify the risk associated with the unmitigated and mitigated accident scenarios described in the TWRS FSAR for the accident analysis titled: Surface Leaks Resulting in Pool.

  20. Calculation Notes for Subsurface Leak Resulting in Pool, TWRS FSAR Accident Analysis

    SciTech Connect (OSTI)

    Hall, B.W.

    1996-09-25

    This document includes the calculations performed to quantify the risk associated with the unmitigated and mitigated accident scenarios described in the TWRS FSAR for the accident analysis titled: Subsurface Leaks Resulting in Pool.

  1. Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

    SciTech Connect (OSTI)

    Mourao, Rogerio P.; Leite da Silva, Luiz; Miranda, Carlos A.; Mattar Neto, Miguel; Quintana, Jose F.A.; Saliba, Roberto O.; Novara, Oscar E.

    2013-07-01

    Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away from reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)

  2. MSIV leakage airborne iodine transport

    SciTech Connect (OSTI)

    Cline, J.E. (Cline Associates Inc., Rockville, MD (United States))

    1993-01-01

    Gaseous iodine deposits on surfaces exposed to vapors. Basic chemical and physical principles predict this behavior, and several laboratory and in-plant measurements demonstrate the characteristic. An empirical model was developed that describes the deposition, resuspension, and transformation of airborne radioiodine molecular species as a stream containing these forms moves along its pathway. The model uses a data base of measured values of deposition and resuspension rates in its application and describes the conversion of the more reactive inorganic iodine species I[sub 2] to the less reactive organic species CH[sub 3]I as the iodine deposits and resuspends along the path. It also considers radioactive decay and chemical surface bonding during residence on surfaces. For the 8-day [sup 131]I, decay during the airborne portion of the transport is negligible. Verification of the model included measurement tests of long gaseous-activity sampling lines of different diameters, operated at different flow rates and stream temperatures. The model was applied to the streams at a boiling water reactor nuclear power plant to describe the transport through leaking main steam isolation valves (MSIVs), following a loss-of-coolant accident.

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

    SciTech Connect (OSTI)

    Dukelow, J S [Pacific Northwest Lab., Richland, WA (United States); Harrison, D G [Jason Associates, Idaho Falls, ID (United States); Morgenstern, M [Battelle Human Affairs Research Center, Seattle, WA (United States)

    1992-03-01

    This reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core degradation and prior to RPV failure. A parallel project at BNL has been considering strategies applicable to the ex-vessel phase of PWR severe accidents.

  4. An analysis to determine correlations of freeway traffic accidents with specific geometric design features 

    E-Print Network [OSTI]

    Smith, Frank Miller

    1960-01-01

    was followed to select an accident frequency index was the development and evaluation of several experimental indices. Factors considered, Measurement of accident exposure based on volumes and the number of accidents which occurred served as the basic... of these conditions, the length of the through lanes, in each direction from the ramp terminal, over which accidents were counted was a variable which was considered for several indices. In the application of this factor, through-lane acci- dents were grouped...

  5. Uncertainty and sensitivity analysis of a fire-induced accident scenario involving binary variables and mechanistic codes

    E-Print Network [OSTI]

    Minton, Mark A. (Mark Aaron)

    2010-01-01

    In response to the transition by the United States Nuclear Regulatory Commission (NRC) to a risk-informed, performance-based fire protection rulemaking standard, Fire Probabilistic Risk Assessment (PRA) methods have been ...

  6. SUSTAINABLE TRANSPORTATION ENERGY PATHWAYS

    E-Print Network [OSTI]

    California at Davis, University of

    SUSTAINABLE TRANSPORTATION ENERGY PATHWAYS A Research Summary for Decision Makers Edited by Joan Ogden and Lorraine Anderson #12;SUSTAINABLE TRANSPORTATION ENERGY PATHWAYS #12;SUSTAINABLE;6 SUSTAINABLE TRANSPORTATION ENERGY PATHWAYS ACKNOWLEDGEMENTS #12;1 SUSTAINABLE TRANSPORTATION ENERGY PATHWAYS

  7. Accident/Injury Reporting, Investigation, & Basic First Aid Plan

    E-Print Network [OSTI]

    Long, Nicholas

    . It is designed to help reduce injuries by reducing unsafe or hazardous conditions and discouraging accident causing unsafe acts or practices. It applies to all SFASU employees and campus locations conditions to your supervisor and the Safety Department by filling out a Report of Safety or Health Hazard

  8. FORMALISM HELPS IN DESCRIBING ACCIDENTS Peter Ladkin, Universitt Bielefeld, Germany

    E-Print Network [OSTI]

    Ladkin, Peter B.

    , such reasoning engineering is both essential and non-trivial. Accident reports in aviation present careful disagreement systems resulting from maintenance-induced damage leading to the separation of the No. 1 engine maintenance procedures which led to failure of the pylon structure. We shall analyse this statement

  9. Nuclear fuel cycle facility accident analysis handbook

    SciTech Connect (OSTI)

    1998-03-01

    The purpose of this Handbook is to provide guidance on how to calculate the characteristics of releases of radioactive materials and/or hazardous chemicals from nonreactor nuclear facilities. In addition, the Handbook provides guidance on how to calculate the consequences of those releases. There are four major chapters: Hazard Evaluation and Scenario Development; Source Term Determination; Transport Within Containment/Confinement; and Atmospheric Dispersion and Consequences Modeling. These chapters are supported by Appendices, including: a summary of chemical and nuclear information that contains descriptions of various fuel cycle facilities; details on how to calculate the characteristics of source terms for releases of hazardous chemicals; a comparison of NRC, EPA, and OSHA programs that address chemical safety; a summary of the performance of HEPA and other filters; and a discussion of uncertainties. Several sample problems are presented: a free-fall spill of powder, an explosion with radioactive release; a fire with radioactive release; filter failure; hydrogen fluoride release from a tankcar; a uranium hexafluoride cylinder rupture; a liquid spill in a vitrification plant; and a criticality incident. Finally, this Handbook includes a computer model, LPF No.1B, that is intended for use in calculating Leak Path Factors. A list of contributors to the Handbook is presented in Chapter 6. 39 figs., 35 tabs.

  10. An analysis of accident experience at entrance ramps within construction work zones at long-term freeway reconstruction projects in Texas 

    E-Print Network [OSTI]

    Casteel, David Bryan

    1991-01-01

    , severe accidents, daytime accidents, and multi-vehicle accidents (other than rear-end accidents) increased disproportionately in entrance ramp areas during construction. Conversely, accident frequencies did not increase significantly (a =0. 05... in Virginia, found that accidents in construction zones during 1977 were less severe than normal highway accidents. A reported 35 percent of work zone accidents were rear-end collisions. This may have contributed to the decrease in severity of reported work...

  11. Assessment of the amount of cesium-137 released into the Pacific Ocean after the Fukushima accident

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Assessment of the amount of cesium-137 released into the Pacific Ocean after the Fukushima accident into the Pacific Ocean after the Fukushima accident and analysis of its dispersion in Japanese coastal waters, J into the ocean from the Fukushima Daiichi nuclear power plant (NPP) after the accident in March 2011 and to gain

  12. Why System Safety Professionals Should Read Accident Reports C. M. Holloway*, C. W. Johnson

    E-Print Network [OSTI]

    Johnson, Chris

    Why System Safety Professionals Should Read Accident Reports C. M. Holloway*, C. W. Johnson *NASA, who regularly read accident reports reap important benefits. These benefits include an improved accident reports regularly. This is a shame. People from many different disciplines have much to gain

  13. K West Basin Integrated Water Treatment System (IWTS) E-F Annular Filter Vessel Accident Calculations

    SciTech Connect (OSTI)

    PIEPHO, M.G.

    2000-01-10

    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.

  14. K West Basin Integrated Water Treatment System (IWTS) E-F Annular Filter Vessel Accident Calculations

    SciTech Connect (OSTI)

    RITTMANN, P.D.

    1999-10-07

    Three bounding accidents postdated 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, and a hydrogen explosion. Event trees and accident probabilities are estimated. In all cases, the unmitigated dose consequences are below the risk evaluation guidelines.

  15. Modelling of Stochastic Hybrid Systems with Applications to Accident Risk Assessment

    E-Print Network [OSTI]

    Del Moral , Pierre

    Modelling of Stochastic Hybrid Systems with Applications to Accident Risk Assessment #12;The SYSTEMS WITH APPLICATIONS TO ACCIDENT RISK ASSESSMENT DISSERTATION to obtain the doctor's degree promotor Prof. dr. A. Bagchi #12;Contents 1 Introduction 3 1.1 Accident risk assessment

  16. Waste Isolation Pilot Plant Transportation Security | Department...

    Office of Environmental Management (EM)

    Accident Investigation of the February 5, 2014, Underground Salt Haul Truck Fire at the Waste Isolation Pilot Plant, Carlsbad NM Accident Investigation Report - Fire...

  17. Phase 1A Final Report for the AREVA Team Enhanced Accident Tolerant Fuels Concepts

    SciTech Connect (OSTI)

    Morrell, Mike E.

    2015-03-19

    In response to the Department of Energy (DOE) funded initiative to develop and deploy lead fuel assemblies (LFAs) of Enhanced Accident Tolerant Fuel (EATF) into a US reactor within 10 years, AREVA put together a team to develop promising technologies for improved fuel performance during off normal operations. This team consisted of the University of Florida (UF) and the University of Wisconsin (UW), Savannah River National Laboratory (SRNL), Duke Energy and Tennessee Valley Authority (TVA). This team brought broad experience and expertise to bear on EATF development. AREVA has been designing; manufacturing and testing nuclear fuel for over 50 years and is one of the 3 large international companies supplying fuel to the nuclear industry. The university and National Laboratory team members brought expertise in nuclear fuel concepts and materials development. Duke and TVA brought practical utility operating experience. This report documents the results from the initial “discovery phase” where the team explored options for EATF concepts that provide enhanced accident tolerance for both Design Basis (DB) and Beyond Design Basis Events (BDB). The main driver for the concepts under development were that they could be implemented in a 10 year time frame and be economically viable and acceptable to the nuclear fuel marketplace. The economics of fuel design make this DOE funded project very important to the nuclear industry. Even incremental changes to an existing fuel design can cost in the range of $100M to implement through to LFAs. If this money is invested evenly over 10 years then it can take the fuel vendor several decades after the start of the project to recover their initial investment and reach a breakeven point on the initial investment. Step or radical changes to a fuel assembly design can cost upwards of $500M and will take even longer for the fuel vendor to recover their investment. With the projected lifetimes of the current generation of nuclear power plants large scale investment by the fuel vendors is difficult to justify. Specific EATF enhancements considered by the AREVA team were; Improved performance in DB and BDB conditions; Reduced release to the environment in a catastrophic accident; Improved performance during normal operating conditions; Improved performance if US reactors start to load follow; Equal or improved economics of the fuel; and Improvements to the fuel behavior to support future transportation and storage of the used nuclear fuel (UNF). In pursuit of the above enhancements, EATF technology concepts that our team considered were; Additives to the fuel pellets which included; Chromia doping to increase fission gas retention. Chromia doping has the potential to improve load following characteristics, improve performance of the fuel pellet during clad failure, and potentially lock up cesium into the fuel matrix; Silicon Carbide (SiC) Fibers to improve thermal heat transfer in normal operating conditions which also improves margin in accident conditions and the potential to lock up iodine into the fuel matrix; Nano-diamond particles to enhance thermal conductivity; Coatings on the fuel cladding; and Nine coatings on the existing Zircaloy cladding to increase coping time and reduce clad oxidation and hydrogen generation during accident conditions, as well as reduce hydrogen pickup and mitigate hydride reorientation in the cladding. To facilitate the development process AREVA adopted a formal “Gate Review Process” (GR) that was used to review results and focus resources onto promising technologies to reduce costs and identify the technologies that would potentially be carried forward to LFAs within a 10 year period. During the initial discovery phase of the project AREVA took the decision to be relatively hands off and allow our university and National Laboratory partners to be free thinking and consider options that would not be constrained by preconceived ideas from the fuel vendor. To counter this and to keep the partners focused, the GR process was utilized. During this GR process each

  18. Audit of the Department of Energy's Transportation Accident Resistant Container Program, IG-0380

    Broader source: Energy.gov (indexed) [DOE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirleyEnergyTher i n c i p a l De p uBUS SERVICE SUBSIDIES AT THE IDAHO1997 MEMORANDUM51997Site

  19. Beam Transport

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home Room News PublicationsAudits & Inspections AuditsBarbara McClintockSecurityBeam Transport Beam

  20. Stochastic Transport

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power AdministrationRobust,Field-effect Photovoltaics -7541C.3X-rays3 Prepared by:'!Transport in PPCD Discharges by

  1. Greening Transportation

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home Room NewsInformation Current HABFESOpportunitiesNERSC GettingGraphene's 3DGreenGreenTransportation

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

    SciTech Connect (OSTI)

    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 traditional 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.

  3. Some Ramsey Problems Involving Triangles -

    E-Print Network [OSTI]

    Radziszowski, Stanislaw P.

    Some Ramsey Problems Involving Triangles - Computational Approach Stanislaw Pawel Radziszowski Department of Computer Science Rochester Institute of Technology, NY ramsey@dimacs, 28 may 2009 1/40 #12;Outline - Triangles Everywhere or avoiding K3 in some/most colors 1 Ramsey Numbers - Two Colors Some known

  4. Radiological science involves technologists use

    E-Print Network [OSTI]

    Sibille, Etienne

    English Composition II Mathematics MATH 0110 Fundamentals of Mathematics THE HUMAN EXPERIENCE Students new Clinical Experience Radiographic Procedures II Radiation Protection II Radiographic Film ProcessingRadiological science involves technologists use X-rays to see the inside of the human body. Health

  5. Type B Accident Investigation Board Report of the January 20, 1998, Electrical Accident at the Casa Grande Substation,South of Phoenix, Arizona

    Broader source: Energy.gov [DOE]

    This report is an independent product of the Type-B Accident Investigation Board appointed by Michael S.Cowan, Chief Program Officer, Western Area Power Administration.

  6. An accident analysis of the physical plant of the Agricultural and Mechanical College of Texas 

    E-Print Network [OSTI]

    Allen, Gary James

    1963-01-01

    to reduce the fre- quency and/or severity of future accidents. 2 I. THE PROBLEM Statement of the problem. The problem was to make an analysis of the accident records of the Physical Plant Department National Safety Council, Accident Prevention Manual..., based on the analysis of rollected data, as to what type and where safety coz zectiors were most needed to reduce the frequency az d/or severity of accidents. Significance of the problem. The importance of accident prevention was probably best...

  7. Mitigation of Severe Accident Consequences Using Inherent Safety Principles

    SciTech Connect (OSTI)

    R. A. Wigeland; J. E. Cahalan

    2009-12-01

    Sodium-cooled fast reactors are designed to have a high level of safety. Events of high probability of occurrence are typically handled without consequence through reliable engineering systems and good design practices. For accidents of lower probability, the initiating events are characterized by larger and more numerous challenges to the reactor system, such as failure of one or more major engineered systems and can also include a failure to scram the reactor in response. As the initiating conditions become more severe, they have the potential for creating serious consequences of potential safety significance, including fuel melting, fuel pin disruption and recriticality. If the progression of such accidents is not mitigated by design features of the reactor, energetic events and dispersal of radioactive materials may result. For severe accidents, there are several approaches that can be used to mitigate the consequences of such severe accident initiators, which typically include fuel pin failures and core disruption. One approach is to increase the reliability of the reactor protection system so that the probability of an ATWS event is reduced to less than 1 x 10-6 per reactor year, where larger accident consequences are allowed, meeting the U.S. NRC goal of relegating such accident consequences as core disruption to these extremely low probabilities. The main difficulty with this approach is to convincingly test and guarantee such increased reliability. Another approach is to increase the redundancy of the reactor scram system, which can also reduce the probability of an ATWS event to a frequency of less than 1 x 10-6 per reactor year or lower. The issues with this approach are more related to reactor core design, with the need for a greater number of control rod positions in the reactor core and the associated increase in complexity of the reactor protection system. A third approach is to use the inherent reactivity feedback that occurs in a fast reactor to automatically respond to the change in reactor conditions and to result in a benign response to these events. This approach has the advantage of being relatively simple to implement, and does not face the issue of reliability since only fundamental physical phenomena are used in a passive manner, not active engineered systems. However, the challenge is to present a convincing case that such passive means can be implemented and used. The purpose of this paper is to describe this third approach in detail, the technical basis and experimental validation for the approach, and the resulting reactor performance that can be achieved for ATWS events.

  8. Enhanced Accident Tolerant LWR Fuels National Metrics Workshop Report

    SciTech Connect (OSTI)

    Lori Braase

    2013-01-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), in collaboration with the nuclear industry, has been conducting research and development (R&D) activities on advanced Light Water Reactor (LWR) fuels for the last few years. The emphasis for these activities was on improving the fuel performance in terms of increased burnup for waste minimization and increased power density for power upgrades, as well as collaborating with industry on fuel reliability. After the events at the Fukushima Nuclear Power Plant in Japan in March 2011, enhancing the accident tolerance of LWRs became a topic of serious discussion. In the Consolidated Appropriations Act, 2012, Conference Report 112-75, the U.S. Congress directed DOE-NE to: • Give “priority to developing enhanced fuels and cladding for light water reactors to improve safety in the event of accidents in the reactor or spent fuel pools.” • Give “special technical emphasis and funding priority…to activities aimed at the development and near-term qualification of meltdown-resistant, accident-tolerant nuclear fuels that would enhance the safety of present and future generations of light water reactors.” • Report “to the Committee, within 90 days of enactment of this act, on its plan for development of meltdown-resistant fuels leading to reactor testing and utilization by 2020.” Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, and operational transients, as well as design-basis and beyond design-basis events. The overall draft strategy for development and demonstration is comprised of three phases: Feasibility Assessment and Down-selection; Development and Qualification; and Commercialization. The activities performed during the feasibility assessment phase include laboratory scale experiments; fuel performance code updates; and analytical assessment of economic, operational, safety, fuel cycle, and environmental impacts of the new concepts. The development and qualification stage will consist of fuel fabrication and large scale irradiation and safety basis testing, leading to qualification and ultimate NRC licensing of the new fuel. The commercialization phase initiates technology transfer to industry for implementation. Attributes for fuels with enhanced accident tolerance include improved reaction kinetics with steam and slower hydrogen generation rate, while maintaining acceptable cladding thermo-mechanical properties; fuel thermo-mechanical properties; fuel-clad interactions; and fission-product behavior. These attributes provide a qualitative guidance for parameters that must be considered in the development of fuels and cladding with enhanced accident tolerance. However, quantitative metrics must be developed for these attributes. To initiate the quantitative metrics development, a Light Water Reactor Enhanced Accident Tolerant Fuels Metrics Development Workshop was held October 10-11, 2012, in Germantown, Maryland. This document summarizes the structure and outcome of the two-day workshop. Questions regarding the content can be directed to Lori Braase, 208-526-7763, lori.braase@inl.gov.

  9. Markov Model of Severe Accident Progression and Management

    SciTech Connect (OSTI)

    Bari, R.A.; Cheng, L.; Cuadra,A.; Ginsberg,T.; Lehner,J.; Martinez-Guridi,G.; Mubayi,V.; Pratt,W.T.; Yue, M.

    2012-06-25

    The earthquake and tsunami that hit the nuclear power plants at the Fukushima Daiichi site in March 2011 led to extensive fuel damage, including possible fuel melting, slumping, and relocation at the affected reactors. A so-called feed-and-bleed mode of reactor cooling was initially established to remove decay heat. The plan was to eventually switch over to a recirculation cooling system. Failure of feed and bleed was a possibility during the interim period. Furthermore, even if recirculation was established, there was a possibility of its subsequent failure. Decay heat has to be sufficiently removed to prevent further core degradation. To understand the possible evolution of the accident conditions and to have a tool for potential future hypothetical evaluations of accidents at other nuclear facilities, a Markov model of the state of the reactors was constructed in the immediate aftermath of the accident and was executed under different assumptions of potential future challenges. This work was performed at the request of the U.S. Department of Energy to explore 'what-if' scenarios in the immediate aftermath of the accident. The work began in mid-March and continued until mid-May 2011. The analysis had the following goals: (1) To provide an overall framework for describing possible future states of the damaged reactors; (2) To permit an impact analysis of 'what-if' scenarios that could lead to more severe outcomes; (3) To determine approximate probabilities of alternative end-states under various assumptions about failure and repair times of cooling systems; (4) To infer the reliability requirements of closed loop cooling systems needed to achieve stable core end-states and (5) To establish the importance for the results of the various cooling system and physical phenomenological parameters via sensitivity calculations.

  10. Ion irradiation testing of Improved Accident Tolerant Cladding Materials

    SciTech Connect (OSTI)

    Anderoglu, Osman [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tesmer, Joseph R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Maloy, Stuart A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-01-14

    This report summarizes the results of ion irradiations conducted on two FeCrAl alloys (named as ORNL A&B) for improving the accident tolerance of LWR nuclear fuel cladding. After irradiation with 1.5 MeV protons to ~0.5 to ~1 dpa and 300°C nanoindentations were performed on the cross-sections along the ion range. An increase in hardness was observed in both alloys. Microstructural analysis shows radiation induced defects.

  11. TABLE OF CONTENTS Accident Prevention Signs, Tags, Labels, Signals,

    E-Print Network [OSTI]

    US Army Corps of Engineers

    to meet or exceed ANSI and/or OSHA requirements. USACE facilities shall use signs based upon and contractors may opt to use signs meeting either the OSHA or ANSI standards for temporary use during the life.200; Accident Prevention Signs and Tags; e. ANSI/IEEE C95.2; f. ANSI Z136.1; g. ANSI Z535.1; h. ANSI Z535.2; i

  12. Phase II Accident Investigation Board Briefing | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankADVANCED MANUFACTURINGEnergy Bills andOrder 422.1, CONDUCTCritical Materials UsePhase II Accident

  13. Transporting particulate material

    DOE Patents [OSTI]

    Aldred, Derek Leslie (North Hollywood, CA); Rader, Jeffrey A. (North Hollywood, CA); Saunders, Timothy W. (North Hollywood, CA)

    2011-08-30

    A material transporting system comprises a material transporting apparatus (100) including a material transporting apparatus hopper structure (200, 202), which comprises at least one rotary transporting apparatus; a stationary hub structure (900) constraining and assisting the at least one rotary transporting apparatus; an outlet duct configuration (700) configured to permit material to exit therefrom and comprising at least one diverging portion (702, 702'); an outlet abutment configuration (800) configured to direct material to the outlet duct configuration; an outlet valve assembly from the material transporting system venting the material transporting system; and a moving wall configuration in the material transporting apparatus capable of assisting the material transporting apparatus in transporting material in the material transporting system. Material can be moved from the material transporting apparatus hopper structure to the outlet duct configuration through the at least one rotary transporting apparatus, the outlet abutment configuration, and the outlet valve assembly.

  14. KERENA safety concept in the context of the Fukushima accident

    SciTech Connect (OSTI)

    Zacharias, T.; Novotny, C.; Bielor, E.

    2012-07-01

    Within the last three years AREVA NP and E.On KK finalized the basic design of KERENA which is a medium sized innovative boiling water reactor, based on the operational experience of German BWR nuclear power plants (NPPs). It is a generation III reactor design with a net electrical output of about 1250 MW. It combines active safety equipment of service-proven designs with new passive safety components, both safety classified. The passive systems utilize basic laws of physics, such as gravity and natural convection, enabling them to function without electric power. Even actuation of these systems is performed thanks to basic physic laws. The degree of diversity in component and system design, achieved by combining active and passive equipment, results in a very low core damage frequency. The Fukushima accident enhanced the world wide discussion about the safety of operating nuclear power plants. World wide stress tests for operating nuclear power plants are being performed embracing both natural and man made hazards. Beside the assessment of existing power plants, also new designs are analyzed regarding the system response to beyond design base accidents. KERENA's optimal combination of diversified cooling systems (active and passive) allows passing efficiently such tests, with a high level of confidence. This paper describes the passive safety components and the KERENA reactor behavior after a Fukushima like accident. (authors)

  15. Cofrentes NPP activities on PSA and severe accident analysis

    SciTech Connect (OSTI)

    Suarez, J.; Borondo, L. [IBERDROLA, Madrid (Spain); Garcia, P.J. [UITESA, Madrid (Spain). Nuclear Dept.

    1996-07-01

    Cofrentes NPP (CNPP) has developed a Level 1 PSA with the following scope: analysis of internal events, with the reactor initially operating at power, internal and external flooding risk analysis; internal fire risk analysis; reliability analysis of the containment heat removal and containment isolation systems. Level 1 CNPP-PSA results reveal that total core damage frequency in CNPP is less than other similar BWR/6 plants. The CNPP-PSA related activities and applications being carried out currently are: adjusting of MAAP 3.0B, revision 10, on VAX and PC; acquisition of MAAP 4; development of Level1/Level2-PSA interface; seismic site categorization for the IPEEE; prioritization of motor operated valves related to GL-89/10, complementary analysis for exemption to some 10CFR50 App. J requirements; Q-List grading; reliability-centered maintenance; maintenance rule support; on-line maintenance support, off-line risk-monitor development, PSA applicability to the 10CFR50 App. R requirements, analysis of the frequency of mis-oriented fuel bundle event, etc. About severe accident management, CNPP, as part of the Spanish-BWROG, is currently analyzing the generic products of the US-BWROG AMG in order to generate their specific ones. Also, in this group BWR, the development of tools to simulate accident scenarios beyond core damage will be studied and a training process oriented to warrant the optimum use of new EOP/AMG in accident scenarios will be implemented.

  16. Analysis of PWR RCS Injection Strategy During Severe Accident

    SciTech Connect (OSTI)

    Wang, S.-J. [Institute of Nuclear Energy Research, Taiwan (China); Chiang, K.-S. [Institute of Nuclear Energy Research, Taiwan (China); Chiang, S.-C. [Taiwan Power Company, Taiwan (China)

    2004-05-15

    Reactor coolant system (RCS) injection is an important strategy for severe accident management of a pressurized water reactor (PWR) system. Maanshan is a typical Westinghouse PWR nuclear power plant (NPP) with large, dry containment. The severe accident management guideline (SAMG) of Maanshan NPP is developed based on the Westinghouse Owners Group (WOG) SAMG.The purpose of this work is to analyze the RCS injection strategy of PWR system in an overheated core condition. Power is assumed recovered as the vessel water level drops to the bottom of active fuel. The Modular Accident Analysis Program version 4.0.4 (MAAP4) code is chosen as a tool for analysis. A postulated station blackout sequence for Maanshan NPP is cited as a reference case for this analysis. The hot leg creep rupture occurs during the mitigation action with immediate injection after power recovery according to WOG SAMG, which is not desired. This phenomenon is not considered while developing the WOG SAMG. Two other RCS injection methods are analyzed by using MAAP4. The RCS injection strategy is modified in the Maanshan SAMG. These results can be applied for typical PWR NPPs.

  17. Enhanced Accident Tolerant Fuels for LWRS - A Preliminary Systems Analysis

    SciTech Connect (OSTI)

    Gilles Youinou; R. Sonat Sen

    2013-09-01

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

  18. Transportation Data Programs:Transportation Energy Data Book...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Transportation Data Programs:Transportation Energy Data Book,Vehicle Technologies Market Report, and VT Fact of the Week Transportation Data Programs:Transportation Energy Data...

  19. Erosion and Optimal Transport

    E-Print Network [OSTI]

    Birnir, Bjorn; Rowlett, Julie

    2010-01-01

    383 pp. EROSION AND OPTIMAL TRANSPORT [23] I. Ekeland and T.and D. Simons, Sediment transport capacity of overland ?ow,measure spaces via optimal transport, Ann. of Math. (2),

  20. TRANSPORT CHARACTERISTICS OF SELECTED PWR LOCA GENERATED DEBRIS.

    SciTech Connect (OSTI)

    A. K. MAJI; B. MARSHALL; ET AL

    2000-10-01

    In the unlikely event of a Loss of Coolant Accident (LOCA) in a pressurized water reactor (PWR), break jet impingement would dislodge thermal insulation from nearby piping, as well as other materials within the containment, such as paint chips, concrete dust, and fire barrier materials. Steam/water flows induced by the break and by the containment sprays would transport debris to the containment floor. Subsequently, debris would likely transport to and accumulate on the suction sump screens of the emergency core cooling system (ECCS) pumps, thereby potentially degrading ECCS performance and possibly even failing the ECCS. In 1998, the U. S. Nuclear Regulatory Commission (NRC) initiated a generic study (Generic Safety Issue-191) to evaluate the potential for the accumulation of LOCA related debris on the PWR sump screen and the consequent loss of ECCS pump net positive suction head (NPSH). Los Alamos National Laboratory (LANL), supporting the resolution of GSI-191, was tasked with developing a method for estimating debris transport in PWR containments to estimate the quantity of debris that would accumulate on the sump screen for use in plant specific evaluations. The analytical method proposed by LANL, to predict debris transport within the water that would accumulate on the containment floor, is to use computational fluid dynamics (CFD) combined with experimental debris transport data to predict debris transport and accumulation on the screen. CFD simulations of actual plant containment designs would provide flow data for a postulated accident in that plant, e.g., three-dimensional patterns of flow velocities and flow turbulence. Small-scale experiments would determine parameters defining the debris transport characteristics for each type of debris. The containment floor transport methodology will merge debris transport characteristics with CFD results to provide a reasonable and conservative estimate of debris transport within the containment floor pool and subsequent accumulation of debris on the sump screen. The complete methodology will, of course, include a means of estimating debris generation, transport to the containment floor, transport to the sump screen, and the resulting loss of NPSH.

  1. Annotated bibliography of literature relating to wind transport of plutonium-contaminated soils at the Nevada Test Site

    SciTech Connect (OSTI)

    Lancaster, N.; Bamford, R.

    1993-12-01

    During the period from 1954 through 1963, a number of tests were conducted on the Nevada Test Site (NTS) and Tonopah Test Range (TTR) to determine the safety of nuclear devices with respect to storage, handling, transport, and accidents. These tests were referred to as ``safety shots.`` ``Safety`` in this context meant ``safety against fission reaction.`` The safety tests were comprised of chemical high explosive detonations with components of nuclear devices. The conduct of these tests resulted in the dispersion of plutonium, and some americium over areas ranging from several tens to several hundreds of hectares. Of the various locations used for safety tests, the site referred to as ``Plutonium Valley`` was subject to a significant amount of plutonium contamination. Plutonium Valley is located in Area 11 on the eastern boundary of the NTS at an elevation of about 1036 m (3400 ft). Plutonium Valley was the location of four safety tests (A,B,C, and D) conducted during 1956. A major environmental, health, and safety concern is the potential for inhalation of Pu{sup 239,240} by humans as a result of airborne dust containing Pu particles. Thus, the wind transport of Pu{sup 239,240} particles has been the subject of considerable research. This annotated bibliography was created as a reference guide to assist in the better understanding of the environmental characteristics of Plutonium Valley, the safety tests performed there, the processes and variables involved with the wind transport of dust, and as an overview of proposed clean-up procedures.

  2. Transporting Hazardous Materials

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Transporting Hazardous Materials The procedures given below apply to all materials that are considered to be hazardous by the U.S. Department of Transportation (DOT). Consult your...

  3. Sustainability and Transport

    E-Print Network [OSTI]

    Gilbert, Richard

    2006-01-01

    2005. Integrating Sustainability into the Trans- portationTHOUGHT PIECE Sustainability and Transport by Richardof the concept of sustainability to transport planning. In

  4. Key Parameters for Operator Diagnosis of BWR Plant Condition during a Severe Accident

    SciTech Connect (OSTI)

    Clayton, Dwight A.; Poore, III, Willis P.

    2015-01-01

    The objective of this research is to examine the key information needed from nuclear power plant instrumentation to guide severe accident management and mitigation for boiling water reactor (BWR) designs (specifically, a BWR/4-Mark I), estimate environmental conditions that the instrumentation will experience during a severe accident, and identify potential gaps in existing instrumentation that may require further research and development. This report notes the key parameters that instrumentation needs to measure to help operators respond to severe accidents. A follow-up report will assess severe accident environmental conditions as estimated by severe accident simulation model analysis for a specific US BWR/4-Mark I plant for those instrumentation systems considered most important for accident management purposes.

  5. Improved dose assessment in a nuclear reactor accident using the old and new ICRP methodologies 

    E-Print Network [OSTI]

    Yoon, Suk-Chul

    1987-01-01

    using WASH-1400 methodologies will be presented later with discussion of the previous work. THEORY AND MODELING Atmospheric Dispersion Radionuclides leaking from the containment during a severe nuclear accident are dispersed continually... generic set of accident releases that could provide useful insights into improved safety action. A set of five groups of source terms were proposed (BI82) to encompass the full spectrum of severe accident release possibilities, these were called Siting...

  6. Human factors review for nuclear power plant severe accident sequence analysis

    SciTech Connect (OSTI)

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

    1985-01-01

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

  7. Simulation of a small break loss of coolant accident conducted at the BETHSY Integral Test Facility 

    E-Print Network [OSTI]

    Bott, Charles Patrick

    1992-01-01

    systems where the interaction mechanisms themselves are complex and contain uncertainties. Early analysis of reactor accidents concentrated on severe accident scenarios in- volving large piping ruptures in the coolant system. These were seen... as the design basis accident for the systems, since, if the system could survive such a, scenario and maintain fuel cladding integrity, any smaller piping break would be a less severe ac- cident of one already evaluated. The duration of these large break loss...

  8. Descriptions of selected accidents that have occurred at nuclear reactor facilities

    SciTech Connect (OSTI)

    Bertini, H.W.

    1980-04-01

    This report was prepared at the request of the President's Commission on the Accident at Three Mile Island to provide the members of the Commission with some insight into the nature and significance of accidents that have occurred at nuclear reactor facilities in the past. Toward that end, this report presents a brief description of 44 accidents which have occurred throughout the world and which meet at least one of the severity criteria that were established.

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

    SciTech Connect (OSTI)

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

    2010-04-01

    In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs2MoO4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU analyses. Additionally, current analyses suggest that the NUREG-1465 release fractions are conservative by about a factor of 2 in terms of release fractions and that release durations for in-vessel and late in-vessel release periods are in fact longer than the NUREG-1465 durations. It is currently planned that a subsequent report will further characterize these results using more refined statistical methods, permitting a more precise reformulation of the NUREG-1465 alternative source term for both LBU and HBU fuels, with the most important finding being that the NUREG-1465 formula appears to embody significant conservatism compared to current best-estimate analyses.

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

    SciTech Connect (OSTI)

    Blaise Collin

    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 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.

  11. A comparative analysis of accident risks in fossil, hydro, and nuclear energy chains

    SciTech Connect (OSTI)

    Burgherr, P.; Hirschberg, S.

    2008-07-01

    This study presents a comparative assessment of severe accident risks in the energy sector, based on the historical experience of fossil (coal, oil, natural gas, and LPG (Liquefied Petroleum Gas)) and hydro chains contained in the comprehensive Energy-related Severe Accident Database (ENSAD), as well as Probabilistic Safety Assessment (PSA) for the nuclear chain. Full energy chains were considered because accidents can take place at every stage of the chain. Comparative analyses for the years 1969-2000 included a total of 1870 severe ({>=} 5 fatalities) accidents, amounting to 81,258 fatalities. Although 79.1% of all accidents and 88.9% of associated fatalities occurred in less developed, non-OECD countries, industrialized OECD countries dominated insured losses (78.0%), reflecting their substantially higher insurance density and stricter safety regulations. Aggregated indicators and frequency-consequence (F-N) curves showed that energy-related accident risks in non-OECD countries are distinctly higher than in OECD countries. Hydropower in non-OECD countries and upstream stages within fossil energy chains are most accident-prone. Expected fatality rates are lowest for Western hydropower and nuclear power plants; however, the maximum credible consequences can be very large. Total economic damages due to severe accidents are substantial, but small when compared with natural disasters. Similarly, external costs associated with severe accidents are generally much smaller than monetized damages caused by air pollution.

  12. Type B Accident Investigation on the August 5, 2003, Pu-238 Multiple...

    Energy Savers [EERE]

    Board concluded that the direct cause of the accident was the release of airborne contamination from a degraded package that contained cellulose material and plutonium-238...

  13. Integrating accident management issues in the design of future reactors EDF tentative approach

    SciTech Connect (OSTI)

    Berbey, P.; Vidard, M. [EDF-SEPTEN, Villeurbanne (France)

    1997-12-01

    In next generation plants, Severe Accidents will be explicitly considered at a very early stage of the design, and risk significant challenges will be addressed as appropriate. As design provisions could be considered to address some of these challenges, the need for Severe Accident Management (SAM) could be debated. For EDF, Accident Management (AM) in general, and SAM in particular, will remain a cornerstone of plant safety. Design provisions, if any, will have to be such as not to preclude any SAM measure with the potential for preventing accident sequences from progressing further. 3 refs., 2 figs.

  14. Level 1 Accident Report of the March 1, 2010 Bobcat Fatality...

    Energy Savers [EERE]

    Bluffs Substation March 31, 2010 On March 2, 2010 at the request of the Bonneville Power Administration (BPA) Chief Safety Officer, a Level I Accident Investigation was...

  15. Order Module--DOE Order 225.1B, ACCIDENT INVESTIGATIONS

    Office of Energy Efficiency and Renewable Energy (EERE)

    DOE O 225.1B prescribes organizational responsibilities, authorities, and requirements for conducting investigations of certain accidents occurring at DOE sites, facilities, areas, operations, and...

  16. Safety evaluation of MHTGR licensing basis accident scenarios

    SciTech Connect (OSTI)

    Kroeger, P.G.

    1989-04-01

    The safety potential of the Modular High-Temperature Gas Reactor (MHTGR) was evaluated, based on the Preliminary Safety Information Document (PSID), as submitted by the US Department of Energy to the US Nuclear Regulatory Commission. The relevant reactor safety codes were extended for this purpose and applied to this new reactor concept, searching primarily for potential accident scenarios that might lead to fuel failures due to excessive core temperatures and/or to vessel damage, due to excessive vessel temperatures. The design basis accident scenario leading to the highest vessel temperatures is the depressurized core heatup scenario without any forced cooling and with decay heat rejection to the passive Reactor Cavity Cooling System (RCCS). This scenario was evaluated, including numerous parametric variations of input parameters, like material properties and decay heat. It was found that significant safety margins exist, but that high confidence levels in the core effective thermal conductivity, the reactor vessel and RCCS thermal emissivities and the decay heat function are required to maintain this safety margin. Severe accident extensions of this depressurized core heatup scenario included the cases of complete RCCS failure, cases of massive air ingress, core heatup without scram and cases of degraded RCCS performance due to absorbing gases in the reactor cavity. Except for no-scram scenarios extending beyond 100 hr, the fuel never reached the limiting temperature of 1600/degree/C, below which measurable fuel failures are not expected. In some of the scenarios, excessive vessel and concrete temperatures could lead to investment losses but are not expected to lead to any source term beyond that from the circulating inventory. 19 refs., 56 figs., 11 tabs.

  17. Markov Model of Accident Progression at Fukushima Daiichi

    SciTech Connect (OSTI)

    Cuadra A.; Bari R.; Cheng, L-Y; Ginsberg, T.; Lehner, J.; Martinez-Guridi, G.; Mubayi, V.; Pratt, T.; Yue, M.

    2012-11-11

    On March 11, 2011, a magnitude 9.0 earthquake followed by a tsunami caused loss of offsite power and disabled the emergency diesel generators, leading to a prolonged station blackout at the Fukushima Daiichi site. After successful reactor trip for all operating reactors, the inability to remove decay heat over an extended period led to boil-off of the water inventory and fuel uncovery in Units 1-3. A significant amount of metal-water reaction occurred, as evidenced by the quantities of hydrogen generated that led to hydrogen explosions in the auxiliary buildings of the Units 1 & 3, and in the de-fuelled Unit 4. Although it was assumed that extensive fuel damage, including fuel melting, slumping, and relocation was likely to have occurred in the core of the affected reactors, the status of the fuel, vessel, and drywell was uncertain. To understand the possible evolution of the accident conditions at Fukushima Daiichi, a Markov model of the likely state of one of the reactors was constructed and executed under different assumptions regarding system performance and reliability. The Markov approach was selected for several reasons: It is a probabilistic model that provides flexibility in scenario construction and incorporates time dependence of different model states. It also readily allows for sensitivity and uncertainty analyses of different failure and repair rates of cooling systems. While the analysis was motivated by a need to gain insight on the course of events for the damaged units at Fukushima Daiichi, the work reported here provides a more general analytical basis for studying and evaluating severe accident evolution over extended periods of time. This work was performed at the request of the U.S. Department of Energy to explore 'what-if' scenarios in the immediate aftermath of the accidents.

  18. Using design structure matrices to improve decentralization urban transportation systems

    E-Print Network [OSTI]

    Dunn, Travis P

    2005-01-01

    Management of large, complex, urban transportation systems involves numerous stakeholders due to the decentralized ownership and operation of distinct pieces of the physical network. In order to deliver better service to ...

  19. Facilitated Transport of Small Carbohydrates through Plasticized Cellulose Triacetate

    E-Print Network [OSTI]

    Smith, Bradley D.

    Facilitated Transport of Small Carbohydrates through Plasticized Cellulose Triacetate Membranes.7,8 This report describes plasticized cellulose triacetate mem- branes that incorporate a large involved plasticized films that are homogeneous mixtures of cellulose triacetate (CTA), plasticizer (e

  20. Analysis of Three Mile Island-Unit 2 accident

    SciTech Connect (OSTI)

    Not Available

    1980-03-01

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

  1. Advance plant severe accident/thermal hydraulic issues for ACRS

    SciTech Connect (OSTI)

    Kress, T.S.

    1994-09-01

    The ACRS has been reviewing various advance plant designs for certification. The most active reviews have been for the ABWR, AP600, and System 80+. We have completed the reviews for ABWR and System 80+ and are presently concentrating on AP600. The ACRS gave essentially unqualified certification approval for the two completed reviews, yet,,during the process of review a number of issues arose and the plant designs changed somewhat to accommodate some of the ACRS concerns. In this talk, I will describe some of the severe accident and thermal hydraulic related issues we discussed in our reviews.

  2. Locations of criticality alarms and nuclear accident dosimeters at Hanford

    SciTech Connect (OSTI)

    Not Available

    1992-08-01

    Hanford facilities that contain fissionable materials capable of achieving critical mass are monitored with nuclear accident dosimeters (NADS) in compliance with the requirements of DOE Order 5480.11, Chapter XI, Section 4.c. (DOE 1988). The US Department of Energy (DOE) Richland Field Office (RL) has assigned the responsibility for maintaining and evaluating the Hanford NAD system to the Instrumentation and External Dosimetry (I ED) Section of Pacific Northwest Laboratory's (PNL's) Health Physics Department. This manual provides a description of the Hanford NAD, criteria and instructions for proper NAD placement, and the locations of these dosimeters onsite.

  3. Fission product release from irradiated LWR fuel under accident conditions

    SciTech Connect (OSTI)

    Strain, R.V.; Sanecki, J.E.; Osborne, M.F.

    1984-01-01

    Fission product release from irradiated LWR fuel is being studied by heating fuel rod segments in flowing steam and an inert carrier gas to simulate accident conditions. Fuels with a range of irradiation histories are being subjected to several steam flow rates over a wide range of temperatures. Fission product release during each test is measured by gamma spectroscopy and by detailed examination of the collection apparatus after the test has been completed. These release results are complemented by a detailed posttest examination of samples of the fuel rod segment. Results of release measurements and fuel rod characterizations for tests at 1400 through 2000/sup 0/C are presented in this paper.

  4. Accident Investigation Report - Fire Report | Department of Energy

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (BillionProvedTravel TravelChallenges | Department of Energy ASHRAEUs About UstheAccident

  5. In a mining accident, first responders are working against

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration would likeUniverseIMPACT EVALUATION PLAN FOR THE SITE-218 58ImprovingIna mining accident, first

  6. Transport Equations Thomas Hillen

    E-Print Network [OSTI]

    Hillen, Thomas

    Transport Equations Thomas Hillen supported by NSERC University of Alberta, Edmonton Transport V , V compact and symmetric. Transport Equations ­ p.2/33 #12;Directed Movement The equation pt(t, x of v. Transport Equations ­ p.3/33 #12;With Directional Changes µ: turning rate. T(v, v ): probability

  7. Motor Transport Co. 

    E-Print Network [OSTI]

    Unknown

    2011-09-05

    METROPOLITAN TRANSPORTATION PLAN Prepared by: The Longview Metropolitan Planning Organization In cooperation with: o City of Longview o City of White Oak o Gregg County o Harrison County o Texas Department of Transportation o U.S. Department... of Transportation o Federal Highway Administration o Federal Transit Administration Adopted November 12, 2009 TRANSPORTATION 2035 TABLE OF CONTENTS INTRODUCTION...

  8. IDAHO NATIONAL LABORATORY TRANSPORTATION TASK REPORT ON ACHIEVING MODERATOR EXCLUSION AND SUPPORTING STANDARDIZED TRANSPORTATION

    SciTech Connect (OSTI)

    D.K. Morton

    2011-09-01

    Following the defunding of the Yucca Mountain Project, it is reasonable to assume that commercial used fuel will remain in storage for the foreseeable future. This report proposes supplementing the ongoing research and development work related to potential degradation of used fuel, baskets, poisons, and storage canisters during an extended period of storage with a parallel path. This parallel path can assure criticality safety during transportation by implementing a concept that achieves moderator exclusion (no in-leakage of moderator into the used fuel cavity). Using updated risk assessment insights for additional technical justification and relying upon a component inside of the transportation cask that provides a watertight function, a strong argument can be made that moderator intrusion is not credible and should not be a required assumption for criticality evaluations during normal conditions of transportation. A demonstrating testing program supporting a detailed analytical effort as well as updated risk assessment insights can provide the basis for moderator exclusion during hypothetical accident conditions. This report also discusses how this engineered concept can support the goal of standardized transportation.

  9. Quantitative transportation risk analysis based on available data/databases: decision support tools for hazardous materials transportation 

    E-Print Network [OSTI]

    Qiao, Yuanhua

    2007-09-17

    raised questions about the potential consequences of incidents involving LNG transportation. One of those consequences, rapid phase transition (RPT), is studied in this dissertation. The incidents and experiments of LNG-water RPT and theoretical analysis...

  10. Introduction Transport in disordered graphene

    E-Print Network [OSTI]

    Fominov, Yakov

    Introduction Transport in disordered graphene Summary Ballistic transport in disordered graphene P, Gornyi, Mirlin Ballistic transport in disordered graphene #12;Introduction Transport in disordered graphene Summary Outline 1 Introduction Model Experimental motivation Transport in clean graphene 2

  11. Advanced sodium fast reactor accident source terms : research needs.

    SciTech Connect (OSTI)

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

    2010-09-01

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

  12. Novel Accident-Tolerant Fuel Meat and Cladding

    SciTech Connect (OSTI)

    Robert D. Mariani; Pavel G Medvedev; Douglas L Porter; Steven L Hayes; James I. Cole; Xian-Ming Bai

    2013-09-01

    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 release 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.

  13. Thermohydraulic and Safety Analysis for CARR Under Station Blackout Accident

    SciTech Connect (OSTI)

    Wenxi Tian; Suizheng Qiu; Guanghui Su; Dounan Jia [Xi'an Jiaotong University, 28 Xianning Road, Xi'an 710049 (China); Xingmin Liu - China Institute of Atomic Energy

    2006-07-01

    A thermohydraulic and safety analysis code (TSACC) has been developed using Fortran 90 language to evaluate the transient thermohydraulic behaviors and safety characteristics of the China Advanced Research Reactor(CARR) under Station Blackout Accident(SBA). For the development of TSACC, a series of corresponding mathematical and physical models were considered. Point reactor neutron kinetics model was adopted for solving reactor power. All possible flow and heat transfer conditions under station blackout accident were considered and the optional models were supplied. The usual Finite Difference Method (FDM) was abandoned and a new model was adopted to evaluate the temperature field of core plate type fuel element. A new simple and convenient equation was proposed for the resolution of the transient behaviors of the main pump instead of the complicated four-quadrant model. Gear method and Adams method were adopted alternately for a better solution to the stiff differential equations describing the dynamic behaviors of the CARR. The computational result of TSACC showed the enough safety margin of CARR under SBA. For the purpose of Verification and Validation (V and V), the simulated results of TSACC were compared with those of Relap5/Mdo3. The V and V result indicated a good agreement between the results by the two codes. Because of the adoption of modular programming techniques, this analysis code is expected to be applied to other reactors by easily modifying the corresponding function modules. (authors)

  14. Type B Accident Investigation Board Report on the August 5, 1998, Load Haul Dump Accident at U16b Tunnel, Nevada Test Site

    Broader source: Energy.gov [DOE]

    Thisis theType B Accident Investigation Board report of an industrial accident at the Nevada Test site (NTS), U16b tunnel in which a Bechtel Nevada (BN) employee suffered a compressed skull fracture as a result of being struck onthe head by a valve and fitting assembly on the end of a hose whichhad been broken from a water pipe by a moving piece of construction equipment.

  15. Potential Threats from a Likely Nuclear Power Plant Accident: a Climatological Trajectory Analysis

    E-Print Network [OSTI]

    Chen, Shu-Hua

    in the near future as insecure nuclear power plants with a high risk of accidents remain in the regionPotential Threats from a Likely Nuclear Power Plant Accident: a Climatological Trajectory Analysis at the Metsamor Nuclear Power Plant would influence all of Turkey. Furthermore, vulnerable regions in Turkey after

  16. Recovery sequences for a station blackout accident at the Grand Gulf Nuclear Station

    SciTech Connect (OSTI)

    Carbajo, J.J. [Martin Marietta Energy Systems, Oak Ridge, TN (United States)

    1995-12-31

    Recovery sequences for a low-pressure, short term, station blackout severe accident at the Grand Gulf power plant have been investigated using the computer code MELCOR, version 1.8.3 PN. This paper investigates the effect of reflood timing and mass flow rate on accident recovery.

  17. Secure Transportation Management

    SciTech Connect (OSTI)

    Gibbs, P. W.

    2014-10-15

    Secure Transport Management Course (STMC) course provides managers with information related to procedures and equipment used to successfully transport special nuclear material. This workshop outlines these procedures and reinforces the information presented with the aid of numerous practical examples. The course focuses on understanding the regulatory framework for secure transportation of special nuclear materials, identifying the insider and outsider threat(s) to secure transportation, organization of a secure transportation unit, management and supervision of secure transportation units, equipment and facilities required, training and qualification needed.

  18. An evaluation of operating speed reduction as a surrogate measure for accident experience on horizontal curves on two-lane rural highways 

    E-Print Network [OSTI]

    Anderson, Ingrid Bernice

    1993-01-01

    of accidents is relatively small. This problem is especially significant at rural locations where low traffic volumes require several years to establish a sufficient accident rate. A small sample of accident data may not accurately reveal the safety... to accident records. " One geometric measure which several studies have shown to have a relationship with accident experience is degree of curvature. Glennon (7) cites five studies which concluded that accident experience increased with increasing degree...

  19. Wilderness Accidents as Complex Life Events in Cognitive, Social, and Recreational Life Domains: Application of a Stress-Coping Model 

    E-Print Network [OSTI]

    McMahan, Kelli

    2015-08-12

    A wilderness accident can be an unfortunate outcome for outdoor recreationists participating in risk-related recreation or outdoor adventure pursuits. Some pursuits in certain environments or conditions increase the likelihood of accidents...

  20. Chemical accident databases: what they tell us and how they can be improved to establish national safety goals 

    E-Print Network [OSTI]

    McCray, Eboni Trevette

    2000-01-01

    The objectives of this research are to examine and critique eight chemical accident databases, document any trends in accident occurrences, develop a strategy for improving current databases, and to establish national safety goals on the basis...

  1. On an integral involving the digamma function

    E-Print Network [OSTI]

    Donal F. Connon

    2012-12-09

    We consider several possible approaches to evaluating an integral involving the digamma function and a related logarithmic series.

  2. Water Transport Within the STack: Water Transport Exploratory...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Within the STack: Water Transport Exploratory Studies Water Transport Within the STack: Water Transport Exploratory Studies Part of a 100 million fuel cell award announced by DOE...

  3. Study of Air Ingress Across the Duct During the Accident Conditions

    SciTech Connect (OSTI)

    Hassan, Yassin

    2013-05-06

    The goal of this project is to study the fundamental physical phenoena associated with air ingress in very high temperature reactors (VHTRs). Air ingress may occur due to a nupture of primary piping and a subsequent breach in the primary pressure boundary in helium-cooled and graphite-moderated VHTRs. Significant air ingress is a concern because it introduces potential to expose the fuel, graphite support rods, and core to a risk of severe graphite oxidation. Two of the most probable air ingress scenarios involve rupture of a control rod or fuel access standpipe, and rupture in the main coolant pipe on the lower part of the reactor pressure vessel. Therefor, establishing a fundamental understanding of air ingress phenomena is critical in order to rationally evaluate safety of existing VHTRs and develop new designs that mimimize these risks. But despite this importance, progress toward development these predictive capabilities has been slowed by the complex nature of the underlaying phenomena. The combination of interdiffusion among multiple species, molecular diffusion, natural convection, and complex geometries, as well as the multiple chemical reactions involved, impose significant roadblocks to both modeling and experiment design. The project team will employ a coordinated experimental and computational effort that will help gain a deeper understanding of multiphased air ingress phenomena. THis project will enhance advanced modeling and simulation methods, enabling calculation of nuclear power plant transients and accident scenarios with a high degree of confidence. The following are the project tasks: Perform particle image velocimetry measurement of multiphase air ingresses Perform computational fluid dynamics analysis of air ingress phenomena

  4. Theoretical and Experimental Simulation of Accident Scenarios of the JET Cryogenic Components Part II: The JET LHCD Cryopump

    E-Print Network [OSTI]

    Theoretical and Experimental Simulation of Accident Scenarios of the JET Cryogenic Components Part II: The JET LHCD Cryopump

  5. Type B Accident Investigation Board Report Subcontractor Radioactive...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    August 17, 2004 On Friday, May 14, 2004, at approximately 11:30 a.m., a dump truck transporting mixing tank T-12 (Tank T-12) from the New Hydrofracture Facility (NHF)...

  6. Midwestern Radioactive Materials Transportation Committee Agenda...

    Office of Environmental Management (EM)

    Midwestern Radioactive Materials Transportation Committee Agenda Midwestern Radioactive Materials Transportation Committee Agenda Midwestern Radioactive Materials Transportation...

  7. A Perspective on Long-Term Recovery Following the Fukushima Nuclear Accident - 12075

    SciTech Connect (OSTI)

    Chen, S.Y. [Environmental Science Division, Argonne National Laboratory, Argonne, IL (United States)

    2012-07-01

    The tragic events at the Fukushima Daiichi Nuclear Power Station began occurring on March 11, 2011, following Japan's unprecedented earthquake and tsunami. The subsequent loss of external power and on-site cooling capacity severely compromised the plant's safety systems, and subsequently, led to core melt in the affected reactors and damage to spent nuclear fuel in the storage pools. Together with hydrogen explosions, this resulted in a substantial release of radioactive material to the environment (mostly Iodine-131 and Cesium- 137), prompting an extensive evacuation effort. The latest release estimate places the event at the highest severity level (Level 7) on the International Nuclear Event Scale, the same as the Chernobyl accident of 1986. As the utility owner endeavored to stabilize the damaged facility, environmental contamination continued to propagate and affect every aspect of daily life in the affected region of Japan. Elevated levels of radioactivity (mostly dominated by Cs-137 with the passage of time) were found in soil, drinking water, vegetation, produce, seafood, and other foodstuffs. An estimated 80,000 to 90,000 people were evacuated; more evacuations are being contemplated months after the accident, and a vast amount of land has become contaminated. Early actions were taken to ban the shipment and sale of contaminated food and drinking water, followed by later actions to ban the shipment and sale of contaminated beef, mushrooms, and seafood. As the event continues to evolve toward stabilization, the long-term recovery effort needs to commence - a process that doubtless will involve rather complex decision-making interactions between various stakeholders. Key issues that may be encountered and considered in such a process include (1) socio-political factors, (2) local economic considerations, (3) land use options, (4) remediation approaches, (5) decontamination methods, (6) radioactive waste management, (7) cleanup levels and options, and (8) government policies, among others. This paper offers a perspective on this likely long and arduous journey toward establishing a 'new normal' that will ultimately take shape. Toward this end, it is important to evaluate the 'optimization' process advocated by the international community in achieving long-term recovery from this particularly fateful event in Fukushima. In the process, experience and lessons learned from past events will be fully evaluated and considered. (author)

  8. Packaging and Transportation Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1995-09-27

    Establishes safety requirements for the proper packaging and transportation of Department of Energy (DOE) offsite shipments and onsite transfers of hazardous materials and for modal transport. Canceled by DOE 460.1A

  9. Packaging and Transportation Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2010-05-14

    The order establishes safety requirements for the proper packaging and transportation of DOE, including NNSA, offsite shipments and onsite transfers of radioactive and other hazardous materials and for modal transportation. Supersedes DOE O 460.1B.

  10. Sustainability and Transport

    E-Print Network [OSTI]

    Gilbert, Richard

    2006-01-01

    also known there as sustainable mobility. This de?nition wasfor De?ning Sustainable Transport and Mobility. [cited 13Sustainable transporta- tion is de?ned as a means to satisfy current transport and mobility

  11. Transportation Management Workshop: Proceedings

    SciTech Connect (OSTI)

    Not Available

    1993-10-01

    This report is a compilation of discussions presented at the Transportation Management Workshop held in Gaithersburg, Maryland. Topics include waste packaging, personnel training, robotics, transportation routing, certification, containers, and waste classification.

  12. Linear Motor Powered Transportation

    E-Print Network [OSTI]

    Thornton, Richard D.

    This special issue on linear-motor powered transportation covers both supporting technologies and innovative transport systems in various parts of the World, as this technology moves from the lab to commercial operations. ...

  13. Transportation Conference Speakers - 4 

    E-Print Network [OSTI]

    Unknown

    2011-08-17

    Laboratory experiments were performed to study and improve longshore sediment transport rate predictions. Measured total longshore transport in the laboratory was approximately three times greater for plunging breakers than spilling breakers. Three...

  14. Transportation Energy Futures Study

    Office of Energy Efficiency and Renewable Energy (EERE)

    Transportation accounts for 71% of total U.S. petroleum consumption and 33% of total greenhouse gas emissions. The Transportation Energy Futures (TEF) study examines underexplored oil-savings and...

  15. Transportation Market Distortions

    E-Print Network [OSTI]

    Litman, Todd

    2006-01-01

    roads and parking facilities is exempt from rent and taxes,road transport relative to rail (which pays rent and taxesroad tolls, parking fees, and Litman, Transportation Market Distortions higher fuel taxes

  16. Introduction to Transportation Planning

    E-Print Network [OSTI]

    Tipple, Brett

    Introduction to Transportation Planning CMP 4710/6710 Fall 2012 3 Credit Hours Room: ARCH 229 of City & Metropolitan Planning; Associate Dean, College of Architecture + Planning; former associate, social equity, fiscal health, and public health. Unfortunately, most transportation planning processes

  17. Transportation Conference Speakers - 1 

    E-Print Network [OSTI]

    Unknown

    2011-08-17

    Thiamin transport in Escherichia coli is a model system to establish the tolerance of derivatives for transport into the cell. Since little is known about what types of thiamin derivatives may be successfully taken into the cell through...

  18. Packaging and Transportation Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1995-09-27

    Establishes safety requirements for the proper packaging and transportation of offsite shipments and onsite transfers of hazardous materials andor modal transport. Cancels DOE 1540.2 and DOE 5480.3

  19. Packaging and Transportation Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1996-10-02

    Establishes safety requirements for the proper packaging and transportation of Department of Energy (DOE) offsite shipments and onsite transfers of hazardous materials and for modal transport. Cancels DOE O 460.1.

  20. Biofuels and Transportation

    E-Print Network [OSTI]

    Minnesota, University of

    Biofuels and Transportation Impacts and Uncertainties Some Observations of a Reformed Ethanol and Logistics Symposium 3 Topics · Why Biofuels · Ethanol Economics · Ethanol Transportation Equipment Biofuels? · National Security · Reduce Imports of oil · Peak Oil · Replace Fossil Resources

  1. Technical evaluation: 300 Area steam line valve accident

    SciTech Connect (OSTI)

    Not Available

    1993-08-01

    On June 7, 1993, a journeyman power operator (JPO) was severely burned and later died as a result of the failure of a 6-in. valve that occurred when he attempted to open main steam supply (MSS) valve MSS-25 in the U-3 valve pit. The pit is located northwest of Building 331 in the 300 Area of the Hanford Site. Figure 1-1 shows a layout of the 300 Area steam piping system including the U-3 steam valve pit. Figure 1-2 shows a cutaway view of the approximately 10- by 13- by 16-ft-high valve pit with its various steam valves and connecting piping. Valve MSS-25, an 8-in. valve, is located at the bottom of the pit. The failed 6-in. valve was located at the top of the pit where it branched from the upper portion of the 8-in. line at the 8- by 8- by 6-in. tee and was then ``blanked off`` with a blind flange. The purpose of this technical evaluation was to determine the cause of the accident that led to the failure of the 6-in. valve. The probable cause for the 6-in. valve failure was determined by visual, nondestructive, and destructive examination of the failed valve and by metallurgical analysis of the fractured region of the valve. The cause of the accident was ultimately identified by correlating the observed failure mode to the most probable physical phenomenon. Thermal-hydraulic analyses, component stress analyses, and tests were performed to verify that the probable physical phenomenon could be reasonably expected to produce the failure in the valve that was observed.

  2. Indianapolis Public Transportation Corporation

    SciTech Connect (OSTI)

    Not Available

    2004-12-01

    Fact sheet describes the National Renewable Energy Laboratory's evaluation of Indianapolis Public Transportation Corporation's (IndyGo's) hybrid electric buses.

  3. Parking & Transportation Services Sustainability &

    E-Print Network [OSTI]

    Minnesota, University of

    : 2011 #12;As a long-time leader in the areas of waste abatement, pollution reduction, energy management Metro Commuter Services Infinity Award ­ in recognition of alternative transportation programs. · 1996 to maintaining impressive and viable alternative transportation programs. TRANSPORTATION SYSTEM DESIGNS

  4. Arrival condition of spent fuel after storage, handling, and transportation

    SciTech Connect (OSTI)

    Bailey, W.J.; Pankaskie, P.J.; Langstaff, D.C.; Gilbert, E.R.; Rising, K.H.; Schreiber, R.E.

    1982-11-01

    This report presents the results of a study conducted to determine the probable arrival condition of spent light-water reactor (LWR) fuel after handling and interim storage in spent fuel storage pools and subsequent handling and accident-free transport operations under normal or slightly abnormal conditions. The objective of this study was to provide information on the expected condition of spent LWR fuel upon arrival at interim storage or fuel reprocessing facilities or at disposal facilities if the fuel is declared a waste. Results of a literature survey and data evaluation effort are discussed. Preliminary threshold limits for storing, handling, and transporting unconsolidated spent LWR fuel are presented. The difficulty in trying to anticipate the amount of corrosion products (crud) that may be on spent fuel in future shipments is also discussed, and potential areas for future work are listed. 95 references, 3 figures, 17 tables.

  5. TYPE A FISSILE PACKAGING FOR AIR TRANSPORT PROJECT OVERVIEW

    SciTech Connect (OSTI)

    Eberl, K.; Blanton, P.

    2013-10-11

    This paper presents the project status of the Model 9980, a new Type A fissile packaging for use in air transport. The Savannah River National Laboratory (SRNL) developed this new packaging to be a light weight (<150-lb), drum-style package and prepared a Safety Analysis for Packaging (SARP) for submission to the DOE/EM. The package design incorporates unique features and engineered materials specifically designed to minimize packaging weight and to be in compliance with 10CFR71 requirements. Prototypes were fabricated and tested to evaluate the design when subjected to Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC). An overview of the design details, results of the regulatory testing, and lessons learned from the prototype fabrication for the 9980 will be presented.

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

    SciTech Connect (OSTI)

    Zhegang Ma

    2013-09-01

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

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

    SciTech Connect (OSTI)

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

    1992-12-31

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

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

    SciTech Connect (OSTI)

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

    1992-10-01

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

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

    SciTech Connect (OSTI)

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

    1992-01-01

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

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

    SciTech Connect (OSTI)

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

    1992-10-01

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

  11. Analysis of Kuosheng Station Blackout Accident Using MELCOR 1.8.4

    SciTech Connect (OSTI)

    Wang, S.-J.; Chien, C.-S.; Wang, T.-C.; Chiang, K.-S

    2000-11-15

    The MELCOR code, developed by Sandia National Laboratories, is a fully integrated, relatively fast-running code that models the progression of severe accidents in commercial light water nuclear power plants (NPPs).A specific station blackout (SBO) accident for Kuosheng (BWR-6) NPP is simulated using the MELCOR 1.8.4 code. The MELCOR input deck for Kuosheng NPP is established based on Kuosheng NPP design data and the MELCOR users' guides. The initial steady-state conditions are generated with a developed self-initialization algorithm. The main severe accident phenomena and the fission product release fractions associated with the SBO accident were simulated. The predicted results are plausible and as expected in light of current understanding of severe accident phenomena. The uncertainty of this analysis is briefly discussed. The important features of the MELCOR 1.8.4 are described. The estimated results provide useful information for the probabilistic risk assessment (PRA) of Kuosheng NPP. This tool will be applied to the PRA, the severe accident analysis, and the severe accident management study of Kuosheng NPP in the near future.

  12. Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 2: Accident and Thermal Fluids Analysis PIRTs

    SciTech Connect (OSTI)

    Ball, Sydney J; Corradini, M.; Fisher, Stephen Eugene; Gauntt, R.; Geffraye, G.; Gehin, Jess C; Hassan, Y.; Moses, David Lewis; Renier, John-Paul; Schultz, R.; Wei, T.

    2008-03-01

    An accident, thermal fluids, and reactor physics phenomena identification and ranking process was conducted by a panel of experts on the next generation nuclear plant (NGNP) design (consideration given to both pebble-bed and prismatic gas-cooled reactor configurations). Safety-relevant phenomena, importance, and knowledge base were assessed for the following event classes: (1) normal operation (including some reactor physics aspects), (2) general loss of forced circulation (G-LOFC), (3) pressurized loss-of-forced circulation (P-LOFC), (4) depressurized loss-of-forced circulation (D-LOFC), (5) air ingress (following D-LOFC), (6) reactivity transients - including anticipated transients without scram (ATWS), (7) processes coupled via intermediate heat exchanger (IHX) (IHX failure with molten salt), and (8) steam/water ingress. The panel's judgment of the importance ranking of a given phenomenon (or process) was based on the effect it had on one or more figures of merit or evaluation criteria. These included public and worker dose, fuel failure, and primary (and other safety) system integrity. The major phenomena of concern that were identified and categorized as high importance combined with medium to low knowledge follow: (1) core coolant bypass flows (normal operation), (2) power/flux profiles (normal operation), (3) outlet plenum flows (normal operation), (4) reactivity-temperature feedback coefficients for high-plutonium-content cores (normal operation and accidents), (5) fission product release related to the transport of silver (normal operation), (6)emissivity aspects for the vessel and reactor cavity cooling system (G-LOFC), (7) reactor vessel cavity air circulation and heat transfer (G-LOFC), and (8)convection/radiation heating of upper vessel area (P-LOFC).

  13. An analysis of factors contributing to train-involved crashes 

    E-Print Network [OSTI]

    Cooner, Scott Allen

    1995-01-01

    , light conditions, driver race and sex, accident severity, location type, and roadway class were statistically different compared to the corresponding national and statewide frequency distributions. The analysis of the research hypotheses showed...

  14. Subwavelength Transportation of Light with Atomic Resonances

    E-Print Network [OSTI]

    Chui, Siu-Tat; Jo, Gyu-Boong

    2015-01-01

    We propose and investigate a new type of optical waveguide made by an array of atoms without involving conventional Bragg scattering or total internal reflection. A finite chain of atoms collectively coupled through their intrinsic resonance supports a propagating mode with minimal radiative loss when the array spacing $a$ is around 0.6$\\lambda_0/2\\pi$ where $\\lambda_0$ is the wavelength of the nearly resonant optical transition. We find that the transportation is robust with respect to position fluctuation and remains possible when the atoms are placed on a circle. Our result paves the way to implement the subwavelength transportation of light in integrated optical circuits with cold atoms.

  15. Assessment of the risk of transporting liquid chlorine by rail

    SciTech Connect (OSTI)

    Andrews, W.B.

    1980-03-01

    This report presents the risk of shipping liquid chlorine by rail. While chlorine is not an energy material, there are several benefits to studying chlorine transportation risks. First, chlorine, like energy materials, is widely used as a feedstock to industry. Second, it is the major purification agent in municipal water treatment systems and therefore, provides direct benefits to the public. Finally, other risk assessments have been completed for liquid chlorine shipments in the US and Europe, which provide a basis for comparison with this study. None of the previous PNL energy material risk assessments have had other studies for comparison. For these reasons, it was felt that a risk assessment of chlorine transportation by rail could provide information on chlorine risk levels, identify ways to reduce these risks and use previous studies on chlorine risks to assess the strengths and weaknesses of the PNL risk assessment methodology. The risk assessment methodology used in this study is summarized. The methodology is presented in the form of a risk assessment model which is constructed for ease of periodic updating of the data base so that the risk may be reevaluated as additional data become available. The report is sectioned to correspond to specific analysis steps identified in the model. The transport system and accident environment are described. The response of the transport system to accident environments is described. Release sequences are postulated and evaluated to determine both the likelihood and possible consequences of a release. Supportive data and analyses are given in the appendices. The risk assessment results are related to the year 1985 to allow a direct comparison with other reports in this series.

  16. Electronic spectroscopy of intermediates involved in the conversion of methane to methanol by FeO

    E-Print Network [OSTI]

    Metz, Ricardo B.

    Electronic spectroscopy of intermediates involved in the conversion of methane to methanol by Fe.1063/1.1448489 I. INTRODUCTION The direct oxidation of methane to an easily transport- able liquid such as methanol process and as the simplest model for alkane oxidation.1,2 Although no direct, efficient methane­methanol

  17. DWPF (Defense Waste Processing Facility) canister impact testing and analyses for the Transportation Technology Center

    SciTech Connect (OSTI)

    Farnsworth, R.K.; Mishima, J.

    1988-12-01

    A legal weight truck cask design has been developed for the US Department of Energy by GA Technologies, Inc. The cask will be used to transport defense high-level waste canisters produced by the Defense Waste Processing Facility (DWPF) at the Savannah River Plant. The development of the cask required the collection of impact data for the DWPF canisters. The Materials Characterization Center (MCC) performed this work under the guidance of the Transportation Technology Center (TTC) at Sandia National Laboratories. Two full-scale DWPF canisters filled with nonradioactive borosilicate glass were impacted under ''normal'' and ''hypothetical'' accident conditions. Two canisters, supplied by the DWPF, were tested. Each canister was vertically dropped on the bottom end from a height of either 0.3 m or 9.1 m (for normal or hypothetical accident conditions, respectively). The structural integrity of each canister was then examined using helium leak and dye penetrant testing. The canisters' diameters and heights, which had been previously measured, were then remeasured to determine how the canister dimensions had changed. Following structural integrity testing, the canisters were flaw leak tested. For transportation flaw leak testing, four holes were fabricated into the shell of canister A-27 (0.3 m drop height). The canister was then transported a total distance of 2069 miles. During transport, the waste form material that fell from each flaw was collected to determine the amount of size distribution of each flaw release. 2 refs., 8 figs., 12 tabs.

  18. Soviet-American relations in light of the Chernobyl nuclear accident 

    E-Print Network [OSTI]

    Beck, Erick Keith

    1993-01-01

    to an ignorance of the role of the Chernobyl accident in ending the Cold In this thesis I will analyze the impact of the Chernobyl accident on Soviet-American relations. I will examine from several different angles the roles which Chernobyl played in shaping...SOVIET-AMERICAN RELATIONS IN LIGHT OF THE CHERNOBYL NUCLEAR ACCIDENT A Thesis by ERICK KEITH BECK Submitted to the Office of Graduate Studies of Texas A&M University in partial fulfillment of the requirements for the degree of MASTER OF ARTS...

  19. The response of BWR Mark II containments to station blackout severe accident sequences

    SciTech Connect (OSTI)

    Greene, S.R.; Hodge, S.A.; Hyman, C.R.; Tobias, M.L. (Oak Ridge National Lab., TN (USA))

    1991-05-01

    This report describes the results of a series of calculations conducted to investigate the response of BWR Mark 2 containments to short-term and long-term station blackout severe accident sequences. The BWR-LTAS, BWRSAR, and MELCOR codes were employed to conduct quantitative accident sequence progression and containment response analyses for several station blackout scenarios. The accident mitigation effectiveness of automatic depressurization system actuation, drywell flooding via containment spray operation, and debris quenching in Mark 2 suppression pools is assessed. 27 refs., 16 figs., 21 tabs.

  20. Study on the Accidental Rupture of Hot Leg or Surge Line in SBO Accident

    SciTech Connect (OSTI)

    Kun Zhang; Xuewu Cao [Shanghai Jiaotong University, Shanghai (China)

    2006-07-01

    The postulated total station blackout accident (SBO) of PWR NPP with 600 MWe in China is analyzed as the base case using SCDAP/RELAP5 code. Then the hot leg or surge line are assumed to rupture before the lower head of Reactor Pressure Vessel (RPV) ruptures, and the progressions are analyzed in detail comparing with the base case. The results show that the accidental rupture of hot leg or surge line will greatly influence the progression of accident. The probability of hot leg or surge line rupture in intentional depressurization is also studied in this paper, which provides a suggestion to the development of Severe Accident Management Guidelines (SAMG). (authors)

  1. Examination of Risk Analysis Methods for MOX Land Transport in Japan

    SciTech Connect (OSTI)

    HOHNSTREITER, GLENN FREDRICK; PIERCE, JIM D.

    2003-04-01

    This report presents background information and methodology for a risk assessment of mixed oxide (MOX) reactor fuel transport in the nation of Japan to support their nuclear energy program. This work includes an extensive literature review, a review of other MOX activities worldwide, a survey of the statutory requirements for transporting nuclear materials, a discussion of risk assessment methodology, and calculation results for specific examples. Typical risk evaluations are given to provide guidance for later risk analyses specific to MOX fuel transport in Japan. This report also includes specific information that will be required for routes, cask types, accident-rate statistics, and population densities along specified routes, along with other detailed information needed for risk analysis studies pertinent to MOX transport in Japan. This information will be used in future specific risk studies.

  2. EBS Radionuclide Transport Abstraction

    SciTech Connect (OSTI)

    J. Prouty

    2006-07-14

    The purpose of this report is to develop and analyze the engineered barrier system (EBS) radionuclide transport abstraction model, consistent with Level I and Level II model validation, as identified in Technical Work Plan for: Near-Field Environment and Transport: Engineered Barrier System: Radionuclide Transport Abstraction Model Report Integration (BSC 2005 [DIRS 173617]). The EBS radionuclide transport abstraction (or EBS RT Abstraction) is the conceptual model used in the total system performance assessment (TSPA) to determine the rate of radionuclide releases from the EBS to the unsaturated zone (UZ). The EBS RT Abstraction conceptual model consists of two main components: a flow model and a transport model. Both models are developed mathematically from first principles in order to show explicitly what assumptions, simplifications, and approximations are incorporated into the models used in the TSPA. The flow model defines the pathways for water flow in the EBS and specifies how the flow rate is computed in each pathway. Input to this model includes the seepage flux into a drift. The seepage flux is potentially split by the drip shield, with some (or all) of the flux being diverted by the drip shield and some passing through breaches in the drip shield that might result from corrosion or seismic damage. The flux through drip shield breaches is potentially split by the waste package, with some (or all) of the flux being diverted by the waste package and some passing through waste package breaches that might result from corrosion or seismic damage. Neither the drip shield nor the waste package survives an igneous intrusion, so the flux splitting submodel is not used in the igneous scenario class. The flow model is validated in an independent model validation technical review. The drip shield and waste package flux splitting algorithms are developed and validated using experimental data. The transport model considers advective transport and diffusive transport from a breached waste package. Advective transport occurs when radionuclides that are dissolved or sorbed onto colloids (or both) are carried from the waste package by the portion of the seepage flux that passes through waste package breaches. Diffusive transport occurs as a result of a gradient in radionuclide concentration and may take place while advective transport is also occurring, as well as when no advective transport is occurring. Diffusive transport is addressed in detail because it is the sole means of transport when there is no flow through a waste package, which may dominate during the regulatory compliance period in the nominal and seismic scenarios. The advective transport rate, when it occurs, is generally greater than the diffusive transport rate. Colloid-facilitated advective and diffusive transport is also modeled and is presented in detail in Appendix B of this report.

  3. Ramsey Numbers Involving Cycles Stanislaw P. Radziszowski

    E-Print Network [OSTI]

    Radziszowski, Stanislaw P.

    Ramsey Numbers Involving Cycles Stanislaw P. Radziszowski Department of Computer Science Rochester and data on Ramsey numbers involving cycles. This survey is based on the author's 2009 revi- sion #12 of the Dynamic Survey DS1, "Small Ramsey Numbers", at the Electronic Journal of Combinatorics. Table of Contents

  4. RESEARCH INVOLVING PRISONERS DEFINITIONS AND EXAMPLES

    E-Print Network [OSTI]

    1 RESEARCH INVOLVING PRISONERS 05/30/2008 DEFINITIONS AND EXAMPLES The DUHS IRB reviews and approves research involving prisoners in compliance with 45 CFR 46 Subpart C and other applicable regulations and laws. The provisions of Subpart C apply whenever the research targets prisoners as subjects

  5. ACCIDENT ANALYSES & CONTROL OPTIONS IN SUPPORT OF THE SLUDGE WATER SYSTEM SAFETY ANALYSIS

    SciTech Connect (OSTI)

    WILLIAMS, J.C.

    2003-11-15

    This report documents the accident analyses and nuclear safety control options for use in Revision 7 of HNF-SD-WM-SAR-062, ''K Basins Safety Analysis Report'' and Revision 4 of HNF-SD-SNF-TSR-001, ''Technical Safety Requirements - 100 KE and 100 KW Fuel Storage Basins''. These documents will define the authorization basis for Sludge Water System (SWS) operations. This report follows the guidance of DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports'', for calculating onsite and offsite consequences. The accident analysis summary is shown in Table ES-1 below. While this document describes and discusses potential control options to either mitigate or prevent the accidents discussed herein, it should be made clear that the final control selection for any accident is determined and presented in HNF-SD-WM-SAR-062.

  6. Quantitative similarity analysis of small-break loss-of-coolant accident scenarios

    SciTech Connect (OSTI)

    Prosek, A.; Kljenak, I.; Mavko, B. [Univ. of Ljubljana (Slovenia). Jozef Stefan Inst.

    1996-07-01

    Classifications of small-break loss-of-coolant accidents based on objective quantitative similarity analysis are proposed. Accident scenarios were simulated in a two-loop pressurized water reactor plant with the RELAP5/MOD3.1 computer code for break sizes ranging from 1.27 cm (0.5 in.) to 15.2 cm (6 in.), with different availability of auxiliary feedwater system or reactor coolant pump trip delay. Similarities between different accident simulations were evaluated by comparing relevant time-dependent parameters with fast Fourier transform and correlation methods. Quantification of similarity between accident simulations could eventually lead to further development of the Code Scaling, Applicability and Uncertainty methodology.

  7. Effect of helium injection on diffusion dominated air ingress accidents in pebble bed reactors

    E-Print Network [OSTI]

    Yurko, Joseph Paul

    2010-01-01

    The primary objective of this thesis was to validate the sustained counter air diffusion (SCAD) method at preventing natural circulation onset in diffusion dominated air ingress accidents. The analysis presented in this ...

  8. Considerations for severe-accident management strategies in a pressurized water reactor

    SciTech Connect (OSTI)

    Carbajo, J.J.; Carter, J.C. (IT Corp., Oak Ridge, TN (USA))

    1988-01-01

    This paper presents results of a sensitivity study for potential recovery actions during a station blackout severe accident in a pressurized water reactor (PWR). The accident progression for each of the recovery actions was calculated by a modified version of the severe-accident integrated analysis code MAAP 3.0B. According to MAAP calculations for a station blackout in a PWR, vessel failure can be avoided if the power is recovered between 9500 s (2.6h) and 9800 s (2.7 h) after accident initiation (depending on reflood flow rate) and the core is immediately reflooded. If the power is recovered after 10 000 s (2.8 h), vessel failure cannot be avoided. Finally, there is a time window of 26 h between vessel failure and containment failure for actions that could prevent containment failure.

  9. Developing and assessing accident management plans for nuclear power plants, Development process and criteria. Volume 1

    SciTech Connect (OSTI)

    Hanson, D.J.; Blackman, H.S.; Meyer, O.R.; Ward, L.W. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

    1992-08-01

    This document is the first volume of a two-volume NUREG/CR. It describes a four-phase approach for developing criteria that can be used for assessing the adequacy of severe accident management plans for nuclear power plants. The general attributes of accident management plans (Phase 1) are identified, and a process for developing and implementing severe accident management plans (Phase 2) is described. This process is based on a prototype process described in NUREG/CR-5543. The prototype process was revised using results from an evaluation of this process (Phase 3), which is documented in Volume 2. General criteria for assessing the adequacy of accident management plans are also presented (Phase 4). These criteria were based on process specific criteria presented in Volume 2 and NUREG/CR-5543.

  10. Developing and assessing accident management plans for nuclear power plants, Development process and criteria

    SciTech Connect (OSTI)

    Hanson, D.J.; Blackman, H.S.; Meyer, O.R.; Ward, L.W. (Idaho National Engineering Lab., Idaho Falls, ID (United States))

    1992-08-01

    This document is the first volume of a two-volume NUREG/CR. It describes a four-phase approach for developing criteria that can be used for assessing the adequacy of severe accident management plans for nuclear power plants. The general attributes of accident management plans (Phase 1) are identified, and a process for developing and implementing severe accident management plans (Phase 2) is described. This process is based on a prototype process described in NUREG/CR-5543. The prototype process was revised using results from an evaluation of this process (Phase 3), which is documented in Volume 2. General criteria for assessing the adequacy of accident management plans are also presented (Phase 4). These criteria were based on process specific criteria presented in Volume 2 and NUREG/CR-5543.

  11. The Adequacy of DOE Natural Phenomena Hazards Performance Goals from an Accident Analysis Perspective

    Office of Energy Efficiency and Renewable Energy (EERE)

    The Adequacy of DOE Natural Phenomena Hazards Performance Goals from an Accident Analysis Perspective Jeff Kimball Defense Nuclear Facilities Safety Board Staff Department of Energy NPH Conference October 26, 2011

  12. Overview of the U.S. DOE Accident Tolerant Fuel Development Program

    SciTech Connect (OSTI)

    Jon Carmack; Frank Goldner; Shannon M. Bragg-Sitton; Lance L. Snead

    2013-09-01

    The United States Fuel Cycle Research and Development Advanced Fuels Campaign has been given the responsibility to conduct research and development on enhanced accident tolerant fuels with the goal of performing a lead test assembly or lead test rod irradiation in a commercial reactor by 2022. The Advanced Fuels Campaign has defined fuels with enhanced accident tolerance as those that, in comparison with the standard UO2-Zircaloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations and operational transients, as well as design-basis and beyond design-basis events. This paper provides an overview of the FCRD Accident Tolerant Fuel program. The ATF attributes will be presented and discussed. Attributes identified as potentially important to enhance accident tolerance include reduced hydrogen generation (resulting from cladding oxidation), enhanced fission product retention under severe accident conditions, reduced cladding reaction with high-temperature steam, and improved fuel-cladding interaction for enhanced performance under extreme conditions. To demonstrate the enhanced accident tolerance of candidate fuel designs, metrics must be developed and evaluated using a combination of design features for a given LWR design, potential improvements to that design, and the design of an advanced fuel/cladding system. The aforementioned attributes provide qualitative guidance for parameters that will be considered for fuels with enhanced accident tolerance. It may be unnecessary to improve in all attributes and it is likely that some attributes or combination of attributes provide meaningful gains in accident tolerance, while others may provide only marginal benefits. Thus, an initial step in program implementation will be the development of quantitative metrics. A companion paper in these proceedings provides an update on the status of establishing these quantitative metrics for accident tolerant LWR fuel.1 The United States FCRD Advanced Fuels Campaign has embarked on an aggressive schedule for development of enhanced accident tolerant LWR fuels. The goal of developing such a fuel system that can be deployed in the U.S. LWR fleet in the next 10 to 20 years supports the sustainability of clean nuclear power generation in the United States.

  13. Accident Performance of Light Water Reactor Cladding Materials

    SciTech Connect (OSTI)

    Nelson, Andrew T.

    2012-07-24

    During a loss of coolant accident as experienced at Fukushima, inadequate cooling of the reactor core forces component temperatures ever higher where they must withstand aggressive chemical environments. Conventional zirconium cladding alloys will readily oxidize in the presence of water vapor at elevated temperatures, rapidly degrading and likely failing. A cladding breach removes the critical barrier between actinides and fission products and the coolant, greatly increasing the probability of the release of radioactivity in the event of a containment failure. These factors have driven renewed international interest in both study and improvement of the materials used in commercial light water reactors. Characterization of a candidate cladding alloy or oxidation mitigation technique requires understanding of both the oxidation kinetics and hydrogen production as a function of temperature and atmosphere conditions. Researchers in the MST division supported by the DOE-NE Fuel Cycle Research and Development program are working to evaluate and quantify these parameters across a wide range of proposed cladding materials. The primary instrument employed is a simultaneous thermal analyzer (STA) equipped with a specialized water vapor furnace capable of maintaining temperatures above 1200 C in a range of atmospheres and water vapor contents. The STA utilizes thermogravimetric analysis and a coupled mass spectrometer to measure in situ oxidation and hydrogen production of candidate materials. This capability is unprecedented in study of materials under consideration for reactor cladding use, and is currently being expanded to investigate proposed coating techniques as well as the effect of coating defects on corrosion resistance.

  14. INTERCOMPARISON OF RESULTS FOR A PWR ROD EJECTION ACCIDENT

    SciTech Connect (OSTI)

    DIAMOND,D.J.; ARONSON,A.; JO,J.; AVVAKUMOV,A.; MALOFEEV,V.; SIDOROV,V.; FERRARESI,P.; GOUIN,C.; ANIEL,S.; ROYER,M.E.

    1999-10-01

    This study is part of an overall program to understand the uncertainty in best-estimate calculations of the local fuel enthalpy during the rod ejection accident. Local fuel enthalpy is used as the acceptance criterion for this design-basis event and can also be used to estimate fuel damage for the purpose of determining radiological consequences. The study used results from neutron kinetics models in PARCS, BARS, and CRONOS2, codes developed in the US, the Russian Federation, and France, respectively. Since BARS uses a heterogeneous representation of the fuel assembly as opposed to the homogeneous representations in PARCS and CRONOS, the effect of the intercomparison was primarily to compare different intra-assembly models. Quantitative comparisons for core power, reactivity, assembly fuel enthalpy and pin power were carried out. In general the agreement between methods was very good providing additional confidence in the codes and providing a starting point for a quantitative assessment of the uncertainty in calculated fuel enthalpy using best-estimate methods.

  15. Development of LWR Fuels with Enhanced Accident Tolerance

    SciTech Connect (OSTI)

    Lahoda, Edward J.; Boylan, Frank A.

    2015-10-30

    Significant progress was made on the technical, licensing, and business aspects of the Westinghouse Electric Company’s Enhanced Accident Tolerant Fuel (ATF) by the Westinghouse ATF team. The fuel pellet options included waterproofed U15N and U3Si2 and the cladding options SiC composites and zirconium alloys with surface treatments. Technology was developed that resulted in U3Si2 pellets with densities of >94% being achieved at the Idaho National Laboratory (INL). The use of U3Si2 will represent a 15% increase in U235 loadings over those in UO2 fuel pellets. This technology was then applied to manufacture pellets for 6 test rodlets which were inserted in the Advanced Test Reactor (ATR) in early 2015 in zirconium alloy cladding. The first of these rodlets are expected to be removed in about 2017. Key characteristics to be determined include verification of the centerline temperature calculations, thermal conductivity, fission gas release, swelling and degree of amorphization

  16. Regional Transportation Coordination Study 

    E-Print Network [OSTI]

    Golden Crescent Regional Planning Commission

    2006-01-01

    stream_source_info Golden Crescent Regional Transportation Coordination Study.pdf.txt stream_content_type text/plain stream_size 357268 Content-Encoding ISO-8859-1 stream_name Golden Crescent Regional Transportation Coordination... Study.pdf.txt Content-Type text/plain; charset=ISO-8859-1 Golden Crescent Regional Transit i Regional Transportation Coordination Study: 7-County Golden Crescent Region Regional...

  17. FUEL CELLS FOR TRANSPORTATION

    E-Print Network [OSTI]

    for Fuel Cells for Transportation Energy Efficiency and Renewable Energy Office of Transportation............................................................................................. 101 A. R&D of a 50-kW, High-Efficiency, High-Power-Density, CO-Tolerant PEM Fuel Cell Stack SystemFUEL CELLS FOR TRANSPORTATION 2 0 0 1 A N N U A L P R O G R E S S R E P O R T U.S. Department

  18. Description of interview data regarding Pittsburgh and confluence toxic chemical accidents

    SciTech Connect (OSTI)

    Rogers, G.O.; Shumpert, B.L.; Sorensen, J.H.

    1990-11-01

    Evacuation is the protective action most often recommended in response to chemical releases in the United States. The appropriateness of a decision to evacuate depends on whether the affected areas can be cleared of residents before it is contaminated by the chemical release. In determining whether an evacuation can be completed in time, emergency officials must consider both technical and behavioral aspects. The technical components can be readily conceived and quantified. In contrast, the behavioral components are much more abstract and more difficult to estimate. This report summarizes the univariate analysis of responses to surveys conducted in two communities where evacuation was recommended following train derailments involving hazardous chemicals. The surveys were designed to identify the actions taken by residents upon receiving the emergency warning; determine when people received the warning, decided to take action, and implemented the action; and ascertain factors that might explain the nature and timing of their actions. The surveys were conducted in the Bloomfield section of Pittsburgh, Pennsylvania, and in the town of Confluence, Pennsylvania. The study confirms that compliance with an emergency warning to evacuate varies and that potentially dangerous delays can be expected. Significant differences were noted, however, in the rate and speed of compliance in the two communities. The surveys provide information on several factors that may be useful in determining the reasons for differences in the responses from the two communities as well as differences among individual respondents. Such factors include the time of day when the accident occurred, where the respondent was at the time, whether the family was together, previous disaster experience, pet ownership, the content of the warning message, and demographic characteristics. 4 refs., 4 figs., 18 tabs.

  19. Estimating Loss-of-Coolant Accident Frequencies for the Standardized Plant Analysis Risk Models

    SciTech Connect (OSTI)

    S. A. Eide; D. M. Rasmuson; C. L. Atwood

    2008-09-01

    The U.S. Nuclear Regulatory Commission maintains a set of risk models covering the U.S. commercial nuclear power plants. These standardized plant analysis risk (SPAR) models include several loss-of-coolant accident (LOCA) initiating events such as small (SLOCA), medium (MLOCA), and large (LLOCA). All of these events involve a loss of coolant inventory from the reactor coolant system. In order to maintain a level of consistency across these models, initiating event frequencies generally are based on plant-type average performance, where the plant types are boiling water reactors and pressurized water reactors. For certain risk analyses, these plant-type initiating event frequencies may be replaced by plant-specific estimates. Frequencies for SPAR LOCA initiating events previously were based on results presented in NUREG/CR-5750, but the newest models use results documented in NUREG/CR-6928. The estimates in NUREG/CR-6928 are based on historical data from the initiating events database for pressurized water reactor SLOCA or an interpretation of results presented in the draft version of NUREG-1829. The information in NUREG-1829 can be used several ways, resulting in different estimates for the various LOCA frequencies. Various ways NUREG-1829 information can be used to estimate LOCA frequencies were investigated and this paper presents two methods for the SPAR model standard inputs, which differ from the method used in NUREG/CR-6928. In addition, results obtained from NUREG-1829 are compared with actual operating experience as contained in the initiating events database.

  20. Improved assessment of population doses and risk factors for a nuclear power plant under accident conditions 

    E-Print Network [OSTI]

    Meyer, Christopher Martin

    1985-01-01

    result of an accident with the severity of those postulated in WASH-1400. These changes are nearly impossible to predict and even more difficult to quantify. For the purposes of this study, calculations will primarily be restricted to individual...IMPROVED ASSESSMENT OF POPULATION DOSES AND RISK FACTORS FOR A NUCLEAR PONER PLANT UNDER ACCIDENT CONDITIONS A Thesis by CHRISTOPHER MARTIN NEVER Submitted to the Graduate College of Texas A&M University in partial fulfillment...

  1. Development of a severe-accident simulator with a visual plant behavior display

    SciTech Connect (OSTI)

    Okabe, K.; Yamagishi, M.; Yoshiki, R.; Miyake, S.

    1989-01-01

    Severe-accident management is one of the important safety concerns of the nuclear industry and regulatory organizations. Mitsubishi Atomic Power and Mitsubishi Heavy Industries in Japan have developed a severe-accident simulator with the ability to display plant thermal-hydraulic behavior visually in order to develop operating guidelines and to use as an education and training tool. The main features of this simulator are described.

  2. Design consistency and driver error as reflected by driver workload and accident rates 

    E-Print Network [OSTI]

    Wooldridge, Mark Douglas

    1992-01-01

    DESIGN CONSISTENCY AND DRIVER ERROR AS REFLECTED BY DRIVER WORKLOAD AND ACCIDENT RATES A Thesis by MARK DOUGLAS WOOLDRIDGE Approved as to style and content by: Daniel B. Fambro (Chair of Committee) Raymond A. Krammes (Member) Olga J.... Pendleton (Member) James T. P. Yao (Head of Department) May 1992 ABSTRACT Design Consistency and Driver Error as Reflected by Driver Workload and Accident Rates (May 1992) Mark Douglas Wooldridge, B. S. , Texas A&M University Chair of Advisory...

  3. An investigative study of automobile traffic accidents utilizing biorhythm theory for predictive value 

    E-Print Network [OSTI]

    Egly, Oscar Henry

    1977-01-01

    literature showed that there were several cases where Biorhythm theory could tend to explain or account for accident occurrences. Although a number of studies have been successfully conducted over a span of time, the role of Biorhythm theory... for the degree of MASTER OF SCIENCE May 1977 Major Subject: Industrial Engineering AN INVESTIGATIVE STUDY OF AUTOMOBILE TRAFFIC ACCIDENTS UTILIZING BIORHYTHM THEORY FOR PREDICTIVE VALUE A Thesis by OSCAR HENRY EGLY JR. Approved as to style and content...

  4. Creep behavior of a nuclear pressure vessel under severe accident conditions

    SciTech Connect (OSTI)

    Beghini, M.; Bertini, L.; Vitale, E.

    1996-12-31

    The results of a study on the creep behavior of the vessel lower head under severe accident conditions are reported. An experimental program aimed at the evaluation of the creep properties of A533grB steel at high temperature (800--1,100 C) and under biaxial loading is summarized and the main results reported. A Finite Element simulation of the lower head under severe accident conditions allows to show the effect of the main parameters affecting the time to rupture.

  5. ELECTROCHEMICAL POWER FOR TRANSPORTATION

    E-Print Network [OSTI]

    Cairns, Elton J.

    2012-01-01

    of the Fuel Cell in Transportation Applications Workshop,practical fuel cell for commercial or consumer applicationfuel cell system engineer- ing is made, vehicle applications

  6. National Transportation Stakeholders Forum

    Office of Environmental Management (EM)

    Nuclear Fuel Grand BC and High-Level Radioactive Waste - Jeff Williams, Director, Nuclear Fuel Storage and Transportation Planning Project, DOEOffice of Nuclear Energy National...

  7. Radioactive Material Transportation Practices

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2002-09-23

    Establishes standard transportation practices for Departmental programs to use in planning and executing offsite shipments of radioactive materials including radioactive waste. Does not cancel other directives.

  8. Transportation Energy Futures Snapshot

    Office of Energy Efficiency and Renewable Energy (EERE)

    This snapshot is a summary of the EERE reports that provide a detailed analysis of opportunities and challenges along the path to a more sustainable transportation energy future.

  9. Transportation and its Infrastructure

    E-Print Network [OSTI]

    2007-01-01

    gas vehicles Annual road tax differentiated by vintageand charges for road transport Tax/pricing measure Optimalannual circulation taxes, tolls and road charges and parking

  10. Transportation Energy Futures

    E-Print Network [OSTI]

    Sperling, Daniel

    1989-01-01

    TRANSPORTATION ment of Oil Shale Technology. Washing- ton,interest and investments in oil shale, ethanol, coal liquidsbiomass materials, coal, oil shale, tar sands, natural gas,

  11. UZ Colloid Transport Model

    SciTech Connect (OSTI)

    M. McGraw

    2000-04-13

    The UZ Colloid Transport model development plan states that the objective of this Analysis/Model Report (AMR) is to document the development of a model for simulating unsaturated colloid transport. This objective includes the following: (1) use of a process level model to evaluate the potential mechanisms for colloid transport at Yucca Mountain; (2) Provide ranges of parameters for significant colloid transport processes to Performance Assessment (PA) for the unsaturated zone (UZ); (3) Provide a basis for development of an abstracted model for use in PA calculations.

  12. Transportation and its Infrastructure

    E-Print Network [OSTI]

    2007-01-01

    of reduction options/ AERO. Ministry of Transport, Publicfrom aviation with the AERO modeling system Part I.from aviation with the AERO modeling system. Montreal,

  13. Natural Gas Transportation Resiliency

    Broader source: Energy.gov (indexed) [DOE]

    Transportation Resiliency Anders Johnson Director Pipeline System Design April 29, 2014 Confidential and Illustrative for discussion purposes only. The views expressed in this...

  14. Transportation Energy Futures Snapshot

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    modes, manage the demand for transportation, and shift the fuel mix to more sustainable sources necessary to reach these significant outcomes. Coordinating a...

  15. WASTE PACKAGE TRANSPORTER DESIGN

    SciTech Connect (OSTI)

    D.C. Weddle; R. Novotny; J. Cron

    1998-09-23

    The purpose of this Design Analysis is to develop preliminary design of the waste package transporter used for waste package (WP) transport and related functions in the subsurface repository. This analysis refines the conceptual design that was started in Phase I of the Viability Assessment. This analysis supports the development of a reliable emplacement concept and a retrieval concept for license application design. The scope of this analysis includes the following activities: (1) Assess features of the transporter design and evaluate alternative design solutions for mechanical components. (2) Develop mechanical equipment details for the transporter. (3) Prepare a preliminary structural evaluation for the transporter. (4) Identify and recommend the equipment design for waste package transport and related functions. (5) Investigate transport equipment interface tolerances. This analysis supports the development of the waste package transporter for the transport, emplacement, and retrieval of packaged radioactive waste forms in the subsurface repository. Once the waste containers are closed and accepted, the packaged radioactive waste forms are termed waste packages (WP). This terminology was finalized as this analysis neared completion; therefore, the term disposal container is used in several references (i.e., the System Description Document (SDD)) (Ref. 5.6). In this analysis and the applicable reference documents, the term ''disposal container'' is synonymous with ''waste package''.

  16. Sustainable Transportation (Fact Sheet)

    SciTech Connect (OSTI)

    Not Available

    2012-09-01

    This document highlights DOE's Office of Energy Efficiency and Renewable Energy's advancements in transportation technologies, alternative fuels, and fuel cell technologies.

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

    SciTech Connect (OSTI)

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

    1996-05-01

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

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

    SciTech Connect (OSTI)

    Putman, V.L.

    1995-09-01

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

  19. CALIFORNIA CENTER FOR INNOVATIVE TRANSPORTATION INSTITUTE OF TRANSPORTATION STUDIES

    E-Print Network [OSTI]

    CALIFORNIA CENTER FOR INNOVATIVE TRANSPORTATION INSTITUTE OF TRANSPORTATION STUDIES UNIVERSITY Investigator CCIT Research Report UCB-ITS-CWP-2011-2 The California Center for Innovative Transportation works;CALIFORNIA CENTER FOR INNOVATIVE TRANSPORTATION INSTITUTE OF TRANSPORTATION STUDIES UNIVERSITY OF CALIFORNIA

  20. Transportation of radionuclides in urban environs: draft environmental assessment

    SciTech Connect (OSTI)

    Finley, N.C.; Aldrich, D.C.; Daniel, S.L.; Ericson, D.M.; Henning-Sachs, C.; Kaestner, P.C.; Ortiz, N.R.; Sheldon, D.D.; Taylor, J.M.

    1980-07-01

    This report assesses the environmental consequences of the transportation of radioactive materials in densely populated urban areas, including estimates of the radiological, nonradiological, and social impacts arising from this process. The chapters of the report and the appendices which follow detail the methodology and results for each of four causative event categories: incident free transport, vehicular accidents, human errors or deviations from accepted quality assurance practices, and sabotage or malevolent acts. The numerical results are expressed in terms of the expected radiological and economic impacts from each. Following these discussions, alternatives to the current transport practice are considered. Then, the detailed analysis is extended from a limited area of New York city to other urban areas. The appendices contain the data bases and specific models used to evaluate these impacts, as well as discussions of chemical toxicity and the social impacts of radioactive material transport in urban areas. The latter are evaluated for each causative event category in terms of psychological, sociological, political, legal, and organizational impacts. The report is followed by an extensive bibliography covering the many fields of study which were required in performing the analysis.

  1. Ethnic differences in parent involvement are moderated by type of involvement scale 

    E-Print Network [OSTI]

    Wong, Shuk Wa

    2006-10-30

    This study examines ethnic group differences on different dimensions of parentrated and teacher-rated parent involvement after adjusting for the influence of family socioeconomic factors, and the role of involvement scale ...

  2. Transportation risk assessment for ethanol transport 

    E-Print Network [OSTI]

    Shelton Davis, Anecia Delaine

    2008-10-10

    This research is aimed at assessing the quantitative risks involved with an ethanol pipeline. Pipelines that run from the Midwest, where the vast majority of ethanol is produced, to the target areas where reformulated gasoline is required...

  3. Transportation risk assessment for ethanol transport 

    E-Print Network [OSTI]

    Shelton Davis, Anecia Delaine

    2009-05-15

    This research is aimed at assessing the quantitative risks involved with an ethanol pipeline. Pipelines that run from the Midwest, where the vast majority of ethanol is produced, to the target areas where reformulated gasoline is required...

  4. Transportation Todd Litman

    E-Print Network [OSTI]

    consumer cost savings. #12;11 Per Capita Transport EnergyPer Capita Transport Energy 0 0.5 1 1.5 2 2.5 D Problem? · Traffic congestion? · Road construction costs? · Parking congestion or costs? · Excessive costs to consumers? · Government costs? · Traffic crashes? · Lack of mobility for non-drivers? · Poor freight

  5. Expert systems in transportation

    SciTech Connect (OSTI)

    O'Leary, K.P.

    1988-01-01

    The 5 papers in the report deal with the following areas: Knowledge representation and software selection for expert-systems design; Expert-system architecture for retaining-wall design; Development of expert-systems technology in the California Department of Transportation; Development of an expert system to assist in the interactive graphic transit system design process; Expert systems development for contingency transportation planing.

  6. Packaging and Transportation Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2003-04-04

    To establish safety requirements for the proper packaging and transportation of Department of Energy (DOE)/National Nuclear Security Administration (NNSA) offsite shipments and onsite transfers of hazardous materials and for modal transport. Cancels DOE O 460.1A. Canceled by DOE O 460.1C.

  7. Transport Layer Cornell University

    E-Print Network [OSTI]

    Low, Steven H.

    Transport Layer Ao Tang Cornell University Ithaca, NY 14853 Lachlan L. H. Andrew California. Low California Institute of Technology Pasadena, CA 91125 I. INTRODUCTION The Internet has evolved of the physical layer, the link layer, the network layer, the transport layer and the application layer1 . See

  8. TRANSPORTATION ENERGY RESEARCH PIER Transportation Research

    E-Print Network [OSTI]

    . The project also tested a Caterpillar C15 engine certified to 2007 U.S. Environmental Protection Agency.energy.ca.gov/research/ transportation/ January 2011 Heavy-Duty Vehicle Emissions and Fuel Consumption Improvement Illustration of a heavy-duty tractor-trailer modified to meet the SmartWayTM Equipment Standards for lower fuel

  9. Bounds on Some Ramsey Numbers Involving Quadrilateral +

    E-Print Network [OSTI]

    Radziszowski, Stanislaw P.

    Bounds on Some Ramsey Numbers Involving Quadrilateral + Xiaodong Xu Guangxi Academy of Sciences@cs.rit.edu Abstract. For graphs G 1 , G 2 , · · · , Gm , the Ramsey number R(G 1 , G 2 , · · · , Gm ) is defined Ramsey numbers involving quadrilateral C 4 , including R(C 4 , K 9 ) # 32, 19 # R(C 4 , C 4 , K 4 ) # 22

  10. Type B Accident Investigation of the Subcontractor Employee Injuries from a November 15, 2000, Fall Accident at the Oak Ridge National Laboratory

    Broader source: Energy.gov [DOE]

    On November 15, 2000, an accident occurred at the U. S. Department of Energy (DOE) Oak Ridge National Laboratory located in Oak Ridge, Tennessee. An employee of Decon and Recovery Services of Oak Ridge, LLC (DRS), working on an Oak Ridge Operations Office (ORO) Environmental Management decommissioning and demolition project received serious injuries from a fall (approximately 13 feet) from a fixed ladder.

  11. Analysis of Loss-of-Coolant Accidents in the NBSR

    SciTech Connect (OSTI)

    Baek J. S.; Cheng L.; Diamond, D.

    2014-05-23

    This report documents calculations of the fuel cladding temperature during loss-of-coolant accidents in the NBSR. The probability of a pipe failure is small and procedures exist to minimize the loss of water and assure emergency cooling water flows into the reactor core during such an event. Analysis in the past has shown that the emergency cooling water would provide adequate cooling if the water filled the flow channels within the fuel elements. The present analysis is to determine if there is adequate cooling if the water drains from the flow channels. Based on photographs of how the emergency water flows into the fuel elements from the distribution pan, it can be assumed that this water does not distribute uniformly across the flow channels but rather results in a liquid film flowing downward on the inside of one of the side plates in each fuel element and only wets the edges of the fuel plates. An analysis of guillotine breaks shows the cladding temperature remains below the blister temperature in fuel plates in the upper section of the fuel element. In the lower section, the fuel plates are also cooled by water outside the element that is present due to the hold-up pan and temperatures are lower than in the upper section. For small breaks, the simulation results show that the fuel elements are always cooled on the outside even in the upper section and the cladding temperature cannot be higher than the blister temperature. The above results are predicated on assumptions that are examined in the study to see their influence on fuel temperature.

  12. Environmental remediation following the Fukushima-Daiichi accident

    SciTech Connect (OSTI)

    Tagawa, A.; Miyahara, K.; Nakayama, S.

    2013-07-01

    A wide area of Fukushima Prefecture was contaminated with radioactivity released by the Fukushima Daiichi nuclear accident. The decontamination pilot projects conducted by JAEA aimed at demonstrating the applicability of different techniques to rehabilitate affected areas. As most radioactive cesium is concentrated at the top of the soil column and strongly bound to mineral surfaces, there are 3 options left to decrease the gamma dose rate (usually measured 1 m above the ground surface): the stripping of the contaminated topsoil (i.e. direct removal of cesium), the dilution by mixing and the soil profile inversion. The last two options do not generate waste. As the half-distance of {sup 137}Cs gammas in soil is in the order of 5-6 cm (depending on density and water content), the shielding by 50 cm of uncontaminated deep soil would theoretically reduce gamma doses by about 3 orders of magnitude. Which option is employed depends basically on the Cesium concentration in the topsoil, averaged over a 15-cm thickness. The JAEA's decontamination pilot projects focus on soil profile inversion and topsoil stripping. Two different techniques have been tested for the soil profile inversion: one is the reversal tillage by which surface soil of thickness of several tens of cm is reversed by using a tractor plough and the other is the complete interchanging of contaminated topsoil with uncontaminated subsoil by using a back-hoe. Reversal tillage with a tractor plough cost about 30 yen/m{sup 2}, which is an order of magnitude lower than that of topsoil-subsoil interchange (about 300 yen/m{sup 2}). Topsoil stripping is significantly more costly (between 550 yen/m{sup 2} and 690 yen/m{sup 2} according to the equipment used)

  13. Optimal Design for Problems Involving Flow and Transport Phenomena in Saturated

    E-Print Network [OSTI]

    of injection wells p set of model parameters q Darcy velocity, or volumetric ux p uid pressure q y ; q z maximum injection rate Q T total net pumping rate rate Q e T total extraction rate Q max T maximum total

  14. Functional characterization and expression of molluscan detoxification enzymes and transporters involved in dietary allelochemical resistance

    E-Print Network [OSTI]

    Whalen, Kristen Elizabeth

    2008-01-01

    Understanding how organisms deal with potentially toxic or fitness-reducing allelochemicals is important for understanding patterns of predation and herbivory in the marine environment. The ability of marine consumers to ...

  15. STRUCTURAL REQUIREMENTS OF ORGANIC ANION TRANSPORTING POLYPEPTIDE MEDIATED TRANSPORT

    E-Print Network [OSTI]

    Weaver, Yi Miao

    2010-04-12

    The organic anion transporting polypeptides (human: OATP; other: Oatp) form a mammalian transporter superfamily that mediates the transport of structurally unrelated compounds across the cell membrane. Members in this superfamily participate...

  16. Essays on Urban Transportation and Transportation Energy Policy

    E-Print Network [OSTI]

    Kim, Chun Kon

    2008-01-01

    and Transportation Energy Policy Chun Kon Kim University of California,California Goyang, KOREA viii P???????????? ??? W?????? P????? The Impacts of Transportation EnergyCalifornia Transportation Center (UCTC) Regents’ Dissertation Fellowship University of California, Irvine California Energy

  17. Developing and assessing accident management plans for nuclear power plants. Evaluation of a prototype process: Volume 2

    SciTech Connect (OSTI)

    Hanson, D.J.; Johnson, S.P.; Blackman, H.S.; Stewart, M.A. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1992-07-01

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

  18. Assessment of potential doses to workers during postulated accident conditions at the Waste Isolation Pilot Plant

    SciTech Connect (OSTI)

    Hoover, M.D.; Newton, G.J. [Inhalation Toxicology Research Inst., Albuquerque, NM (United States); Farrell, R.F. [Dept. of Energy, Carlsbad, NM (United States)

    1996-06-01

    This qualitative hazard evaluation systematically assessed potential doses to workers during postulated accident conditions at the U.S. Department of Energy`s Waste Isolation Pilot Plant (WIPP). Postulated accidents included the spontaneous ignition of a waste drum, puncture of a waste drum by a forklift, dropping of a waste drum from a forklift, and simultaneous dropping of seven drums during a crane failure. The descriptions and estimated frequencies of occurrence for these accidents were developed by the Hazard and Operability Study for CH TRU Waste Handling System (WCAP 14312). The estimated materials at risk, damage ratios, airborne release fractions and respirable fractions for these accidents were taken from the 1995 Safety Analysis Report (SAR) update and from the DOE handbook Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities (DOE-HDBK-3010-94). A Monte Carlo simulation was used to estimate the range of worker exposures that could result from each accident. Guidelines for evaluating the adequacy of defense-in-depth for worker protection at WIPP were adopted from a scheme presented by the International Commission on Radiological Protection in its publication on Protection from Potential Exposure: A Conceptual Framework (ICRP Publication 64). Probabilities of exposures greater than 5, 50, and 300 rem were less than 10{sup -2}, 10{sup -4}, and 10{sup -6} per year, respectively. In conformance with the guidance of DOE standard 3009-94, Appendix A (draft), we emphasize that use of these evaluation guidelines is not intended to imply that these numbers constitute acceptable limits for worker exposure under accident conditions. However, in conjunction with the extensive safety assessment in the 1995 SAR update, these results indicate that the Carlsbad Area Office strategy for the assessment of hazards and accidents assures the protection of workers, as well as members of the public and the environment.

  19. Study on release and transport of aerial radioactive materials in reprocessing plants

    SciTech Connect (OSTI)

    Amano, Y.; Tashiro, S.; Uchiyama, G.; Abe, H.; Yamane, Y.; Yoshida, K.; Kodama, T.

    2013-07-01

    The release and transport characteristics of radioactive materials at a boiling accident of the high active liquid waste (HALW) in a reprocessing plant have been studied for improving experimental data of source terms of the boiling accident. In the study, a heating test and a thermogravimetry and differential thermal analysis (TG-DTA) test were conducted. In the heating test using a simulated HALW, it was found that ruthenium was mainly released into the air in the form of gas and that non-volatile elements were released into the air in the form of mist. In the TG-DTA test, the rate constants and reaction heat of thermal decomposition of ruthenium nitrosyl nitrate were obtained from TG and DTA curves. (authors)

  20. Implementing Advances in Transport Security Technologies | Department...

    Office of Environmental Management (EM)

    Implementing Advances in Transport Security Technologies Implementing Advances in Transport Security Technologies Implementing Advances in Transport Security Technologies More...

  1. Transportation Storage Interface | Department of Energy

    Office of Environmental Management (EM)

    Storage Interface Transportation Storage Interface Regulation of Future Extended Storage and Transportation. Transportation Storage Interface More Documents & Publications Gap...

  2. Automated Transportation Logistics and Analysis System (ATLAS...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Packaging and Transportation Automated Transportation Logistics and Analysis System (ATLAS) Automated Transportation Logistics and Analysis System (ATLAS) The Department of...

  3. Primary system fission product release and transport: A state-of-the-art report to the committee on the safety of nuclear installations

    SciTech Connect (OSTI)

    Wright, A.L.

    1994-06-01

    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 and presents major conclusions on the state of the art.

  4. Transport Services (TAPS) BOF plan

    E-Print Network [OSTI]

    Welzl, Michael

    Transport Services (TAPS) BOF plan T. Moncaster, M. Welzl, D. Ros: dra5-moncaster-tsvwg-transport-services-00 h Reducing Internet Transport Latency Michael Welzl, with help from (alphabe/cal): Anna

  5. Improving Efficiency and Equity in Transportation Finance

    E-Print Network [OSTI]

    Watts, Michael

    2006-01-01

    Fueling Transportation Finance. ” Ian W. H. Parry andFueling Transportation Finance. ” Transportation ResearchFueling Transportation Finance: A Primer on the Gas Tax •

  6. A ThreeDimensional Finite Element Simulation for Transport of Nuclear Waste Contamination in Porous Media \\Lambda

    E-Print Network [OSTI]

    Ewing, Richard E.

    A Three­Dimensional Finite Element Simulation for Transport of Nuclear Waste Contamination for transport of nuclear­waste contamination in three­dimensional porous media are presented with a description waste contamination involving the transport of fluid, heat, brine, and trace­species radionuclides

  7. Type B Accident Investigation Board Report on the May 24, 1998, Electrical Arc Blast at the Kansas City Plant

    Office of Energy Efficiency and Renewable Energy (EERE)

    This report is a product of an accident investigation board appointed by Bruce G. Twining, Manager, Albuquerque Operations Office, Department of Energy.

  8. Superconnections and Parallel Transport

    E-Print Network [OSTI]

    Dumitrescu, Florin

    2007-01-01

    This note addresses the construction of a notion of parallel transport along superpaths arising from the concept of a superconnection on a vector bundle over a manifold $M$. A superpath in $M$ is, loosely speaking, a path in $M$ together with an odd vector field in $M$ along the path. We also develop a notion of parallel transport associated with a connection (a.k.a. covariant derivative) on a vector bundle over a \\emph{supermanifold} which is a direct generalization of the classical notion of parallel transport for connections over manifolds.

  9. Lagrangian transport through surfaces in volume-preserving flows

    E-Print Network [OSTI]

    Daniel Karrasch

    2015-07-24

    Advective transport of scalar quantities through surfaces is of fundamental importance in many scientific applications. From the Eulerian perspective of the surface it can be quantified by the well-known integral of the flux density. The recent development of highly accurate semi-Lagrangian methods for solving scalar conservation laws and of Lagrangian approaches to coherent structures in turbulent (geophysical) fluid flows necessitate a new approach to transport from the (Lagrangian) material perspective. We present a Lagrangian framework for calculating transport of conserved quantities through a given surface in $n$-dimensional, fully aperiodic, volume-preserving flows. Our approach does not involve any dynamical assumptions on the surface or its boundary.

  10. EBS Radionuclide Transport Abstraction

    SciTech Connect (OSTI)

    J.D. Schreiber

    2005-08-25

    The purpose of this report is to develop and analyze the engineered barrier system (EBS) radionuclide transport abstraction model, consistent with Level I and Level II model validation, as identified in ''Technical Work Plan for: Near-Field Environment and Transport: Engineered Barrier System: Radionuclide Transport Abstraction Model Report Integration'' (BSC 2005 [DIRS 173617]). The EBS radionuclide transport abstraction (or EBS RT Abstraction) is the conceptual model used in the total system performance assessment for the license application (TSPA-LA) to determine the rate of radionuclide releases from the EBS to the unsaturated zone (UZ). The EBS RT Abstraction conceptual model consists of two main components: a flow model and a transport model. Both models are developed mathematically from first principles in order to show explicitly what assumptions, simplifications, and approximations are incorporated into the models used in the TSPA-LA. The flow model defines the pathways for water flow in the EBS and specifies how the flow rate is computed in each pathway. Input to this model includes the seepage flux into a drift. The seepage flux is potentially split by the drip shield, with some (or all) of the flux being diverted by the drip shield and some passing through breaches in the drip shield that might result from corrosion or seismic damage. The flux through drip shield breaches is potentially split by the waste package, with some (or all) of the flux being diverted by the waste package and some passing through waste package breaches that might result from corrosion or seismic damage. Neither the drip shield nor the waste package survives an igneous intrusion, so the flux splitting submodel is not used in the igneous scenario class. The flow model is validated in an independent model validation technical review. The drip shield and waste package flux splitting algorithms are developed and validated using experimental data. The transport model considers advective transport and diffusive transport from a breached waste package. Advective transport occurs when radionuclides that are dissolved or sorbed onto colloids (or both) are carried from the waste package by the portion of the seepage flux that passes through waste package breaches. Diffusive transport occurs as a result of a gradient in radionuclide concentration and may take place while advective transport is also occurring, as well as when no advective transport is occurring. Diffusive transport is addressed in detail because it is the sole means of transport when there is no flow through a waste package, which may dominate during the regulatory compliance period in the nominal and seismic scenarios. The advective transport rate, when it occurs, is generally greater than the diffusive transport rate. Colloid-facilitated advective and diffusive transport is also modeled and is presented in detail in Appendix B of this report.

  11. Transportation and Greenhouse Gas Mitigation

    E-Print Network [OSTI]

    Lutsey, Nicholas P.; Sperling, Dan

    2008-01-01

    natural gas and liquefied petroleum gas have continued to make small contributions to transportation,transportation actions include electric power sector actions, eg coal to natural gas

  12. Office of Secure Transportation Activities

    Office of Environmental Management (EM)

    6th, 2012 WIPP Knoxville, TN OFFICE OF SECURE TRANSPORTATION Agency Integration Briefing Our Mission To provide safe and secure ground and air transportation of nuclear weapons,...

  13. Spent Fuel Transportation Risk Assessment

    Office of Environmental Management (EM)

    Spent Fuel Transportation Risk Assessment (SFTRA) Draft NUREG-2125 Overview for National Transportation Stakeholders Forum John Cook Division of Spent Fuel Storage and...

  14. Attenuation mechanisms in the transport of in-vessel radiological source term fission products in an LMFBR

    SciTech Connect (OSTI)

    Woodley, R.E.; Nguyen, D.H.; Hunter, C.W.

    1983-12-01

    Quantifying the release of radiological source term fission products from an LMFBR reactor vessel (RV) is a necessary input to the containment analysis. To estimate this initial source term value, the distribution of the fission products and actinides inside the RV, prior to release, must be known. The in-vessel source term fission product distribution and transport behavior is also essential in assessing and mitigating the plant contamination and cleanup problems which occur from any significant core disruption. This paper attempts to summarize the current knowledge on the behavior of several radioisotopes in different environments created by the accident, without dealing with the modeling of the transport process itself.

  15. Iodine transport analysis in the ESBWR.

    SciTech Connect (OSTI)

    Kalinich, Donald A.; Gauntt, Randall O.; Young, Michael Francis; Longmire, Pamela

    2009-03-01

    A simplified ESBWR MELCOR model was developed to track the transport of iodine released from damaged reactor fuel in a hypothesized core damage accident. To account for the effects of iodine pool chemistry, radiolysis of air and cable insulation, and surface coatings (i.e., paint) the iodine pool model in MELCOR was activated. Modifications were made to MELCOR to add sodium pentaborate as a buffer in the iodine pool chemistry model. An issue of specific interest was whether iodine vapor removed from the drywell vapor space by the PCCS heat exchangers would be sequestered in water pools or if it would be rereleased as vapor back into the drywell. As iodine vapor is not included in the deposition models for diffusiophoresis or thermophoresis in current version of MELCOR, a parametric study was conducted to evaluate the impact of a range of iodine removal coefficients in the PCCS heat exchangers. The study found that higher removal coefficients resulted in a lower mass of iodine vapor in the drywell vapor space.

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

    SciTech Connect (OSTI)

    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, 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. 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.

  17. Insights into the behavior of nuclear power plant containments during severe accidents

    SciTech Connect (OSTI)

    Horschel, D.S.; Ludwigsen, J.S.; Parks, M.B.; Lambert, L.D. [Sandia National Labs., Albuquerque, NM (United States); Dameron, R.A.; Rashid, Y.R. [ANATECH Research Corp., San Diego, CA (United States)

    1993-06-01

    The containment building surrounding a nuclear reactor offers the last barrier to the release of radioactive materials from a severe accident into the environment. The loading environment of the containment under severe accident conditions may include much greater than design pressures and temperatures. Investigations into the performance of containments subject to ultimate or failure pressure and temperature conditions have been performed over the last several years through a program administered by the Nuclear Regulatory Commission (NRC). These NRC sponsored investigations are subsequently discussed. Reviewed are the results of large scale experiments on reinforced concrete, prestressed concrete, and steel containment models pressurized to failure. In conjunction with these major tests, the results of separate effect testing on many of the critical containment components; that is, aged and unaged seals, a personnel air lock and electrical penetration assemblies subjected to elevated temperature and pressure have been performed. An objective of the NRC program is to gain an understanding of the behavior of typical existing and planned containment designs subject to postulated severe accident conditions. This understanding has led to the development of experimentally verified analytical tools that can be applied to accurately predict their ultimate capacities useful in developing severe accident mitigation schemes. Finally, speculation on the response of containments subjected to severe accident conditions is presented.

  18. Automating Risk Assessments of Hazardous Material Shipments for Transportation Routes and Mode Selection

    SciTech Connect (OSTI)

    Barbara H. Dolphin; William D. RIchins; Stephen R. Novascone

    2010-10-01

    The METEOR project at Idaho National Laboratory (INL) successfully addresses the difficult problem in risk assessment analyses of combining the results from bounding deterministic simulation results with probabilistic (Monte Carlo) risk assessment techniques. This paper describes a software suite designed to perform sensitivity and cost/benefit analyses on selected transportation routes and vehicles to minimize risk associated with the shipment of hazardous materials. METEOR uses Monte Carlo techniques to estimate the probability of an accidental release of a hazardous substance along a proposed transportation route. A METEOR user selects the mode of transportation, origin and destination points, and charts the route using interactive graphics. Inputs to METEOR (many selections built in) include crash rates for the specific aircraft, soil/rock type and population densities over the proposed route, and bounding limits for potential accident types (velocity, temperature, etc.). New vehicle, materials, and location data are added when available. If the risk estimates are unacceptable, the risks associated with alternate transportation modes or routes can be quickly evaluated and compared. Systematic optimizing methods will provide the user with the route and vehicle selection identified with the lowest risk of hazardous material release. The effects of a selected range of potential accidents such as vehicle impact, fire, fuel explosions, excessive containment pressure, flooding, etc. are evaluated primarily using hydrocodes capable of accurately simulating the material response of critical containment components. Bounding conditions that represent credible accidents (i.e; for an impact event, velocity, orientations, and soil conditions) are used as input parameters to the hydrocode models yielding correlation functions relating accident parameters to component damage. The Monte Carlo algorithms use random number generators to make selections at the various decision points such as; crash, location, etc. For each pass through the routines, when a crash is randomly selected, crash parameters are then used to determine if failure has occurred using either external look up tables, correlations functions from deterministic calculations, or built in data libraries. The effectiveness of the software was recently demonstrated in safety analyses of the transportation of radioisotope systems for the US Dept. of Energy. These methods are readily adaptable to estimating risks associated with a variety of hazardous shipments such as spent nuclear fuel, explosives, and chemicals.

  19. DEVELOPMENT OF THE HS99 AIR TRANSPORT TYPE A FISSILE PACKAGE

    SciTech Connect (OSTI)

    Blanton, P.; Eberl, K.

    2012-07-10

    An air-transport Type A Fissile radioactive shipping package for the transport of special form uranium sources has been developed by the Savannah River National Laboratory (SRNL) for the Department of Homeland Security. The Package model number is HS99 for Homeland Security Model 99. This paper presents the major design features of the HS99 and highlights engineered materials necessary for meeting the design requirements for this light-weight Type AF packaging. A discussion is provided demonstrating how the HS99 complies with the regulatory safety requirements of the Nuclear Regulatory Commission. The paper summarizes the results of structural testing to specified in 10 CFR 71 for Normal Conditions of Transport and Hypothetical Accident Conditions events. Planned and proposed future missions for this packaging are also addressed.

  20. Transportation and its Infrastructure

    E-Print Network [OSTI]

    2007-01-01

    IEA personnel (WBCSD, 2004b), the WEO 2004 and Mobility 2030are quite similar. The WEO 2006 (IEA, 2006b) includes higherwhile the IEA’s more recent WEO 2006 projects transport

  1. Optimization of Maritime Transportation

    E-Print Network [OSTI]

    Grossmann, Ignacio E.

    and Technology Management Norwegian University of Science and Technology, Trondheim, Norway Enterprise and D. Ronen (2007). Maritime transportation. Handbooks in Operations Research and Management Science (consolidation in the manufacturing sector, increasing competition, profit margins reduced, mergers and pooling

  2. Transportation Baseline Report

    SciTech Connect (OSTI)

    Fawcett, Ricky Lee; Kramer, George Leroy Jr.

    1999-12-01

    The National Transportation Program 1999 Transportation Baseline Report presents data that form a baseline to enable analysis and planning for future Department of Energy (DOE) Environmental Management (EM) waste and materials transportation. In addition, this Report provides a summary overview of DOE’s projected quantities of waste and materials for transportation. Data presented in this report were gathered as a part of the IPABS Spring 1999 update of the EM Corporate Database and are current as of July 30, 1999. These data were input and compiled using the Analysis and Visualization System (AVS) which is used to update all stream-level components of the EM Corporate Database, as well as TSD System and programmatic risk (disposition barrier) information. Project (PBS) and site-level IPABS data are being collected through the Interim Data Management System (IDMS). The data are presented in appendices to this report.

  3. Atmospheric Transport of Radionuclides

    SciTech Connect (OSTI)

    Crawford, T.V.

    2003-03-03

    The purpose of atmospheric transport and diffusion calculations is to provide estimates of concentration and surface deposition from routine and accidental releases of pollutants to the atmosphere. This paper discusses this topic.

  4. National Transportation Stakeholders Forum

    Office of Environmental Management (EM)

    next webinar is scheduled to occur in June 2013 TRIBAL NATIONS CAUCUS UPDATE WIKI AND NTSF WEB SITES ntsf.wikidot.com www.em.doe.govPagesNationalTransportationForum.aspx...

  5. Transport in granular systems

    E-Print Network [OSTI]

    Wendell, Dawn M. (Dawn Marie), 1983-

    2011-01-01

    There are many situations in which a continuum view of granular systems does not fully capture the relevant mechanics. In order for engineers to be able to design systems for transporting granular materials, there needs ...

  6. Transportation Storage Interface

    Office of Environmental Management (EM)

    in above- ground bunkers, each of which is about the size of a one-car garage. Spent Fuel Storage: Dual Purpose Cask Systems 8 Spent Fuel Storage and Transportation: Framework...

  7. PBA Transportation Websites

    Broader source: Energy.gov [DOE]

    PBA Transportation Websites presented to the DOE Systems Analysis Workshop held in Washington, D.C. July 28-29, 2004 to discuss and define role of systems analysis in DOE Hydrogen Program.

  8. Getting the lead out: Citizen involvement in the Williamsburg Bridge lead paint removal project

    SciTech Connect (OSTI)

    Forker, T.R.

    1997-08-01

    This paper examines the process and results of citizen involvement in developing new environmental control and compliance procedures used by the New York City Department of Transportation (NYC DOT) for lead paint removal on the Williamsburg Bridge and other structures. As a case study of the effects of public involvement in environmental decision-making, the study identifies and discusses the factors that produced failures or successes in satisfying the citizen`s concerns about health risks and the effectiveness of the selected pollution control technology.

  9. Fluid transport container

    DOE Patents [OSTI]

    DeRoos, B.G.; Downing, J.P. Jr.; Neal, M.P.

    1995-11-14

    An improved fluid container for the transport, collection, and dispensing of a sample fluid that maintains the fluid integrity relative to the conditions of the location at which it is taken. More specifically, the invention is a fluid sample transport container that utilizes a fitting for both penetrating and sealing a storage container under controlled conditions. Additionally, the invention allows for the periodic withdrawal of portions of the sample fluid without contamination or intermixing from the environment surrounding the sample container. 13 figs.

  10. Membrane Transport Chloride Transport Across Vesicle and Cell

    E-Print Network [OSTI]

    Smith, Bradley D.

    Membrane Transport Chloride Transport Across Vesicle and Cell Membranes by Steroid-Based Receptors-established that molecules which transport cations across cell membranes (cationophores) can have potent biological effects of biological activity. Indeed, chloride transporters have direct medical potential as treatments for cystic

  11. Parking and Transport Policy Page 1 Parking and Transport Policy

    E-Print Network [OSTI]

    Mucina, Ladislav

    Parking and Transport Policy Page 1 Parking and Transport Policy Category: Facilities, Campus Life 1. PURPOSE To standardise and manage parking and transport on the Curtin Bentley campus including that support both State Government and University objectives in a manner that encourages public transport use

  12. CALIFORNIA CENTER FOR INNOVATIVE TRANSPORTATION INSTITUTE OF TRANSPORTATION STUDIES

    E-Print Network [OSTI]

    CALIFORNIA CENTER FOR INNOVATIVE TRANSPORTATION INSTITUTE OF TRANSPORTATION STUDIES UNIVERSITY-ITS-CWP-2010-4 This work was performed by the California Center for Innovative Transportation, a research group at the University of California, Berkeley, in cooperation with the State of California Business, Transportation

  13. CALIFORNIA CENTER FOR INNOVATIVE TRANSPORTATION INSTITUTE OF TRANSPORTATION STUDIES

    E-Print Network [OSTI]

    CALIFORNIA CENTER FOR INNOVATIVE TRANSPORTATION INSTITUTE OF TRANSPORTATION STUDIES UNIVERSITY-ITS-CWP-2011-6 ISSN 1557-2269 The California Center for Innovative Transportation works with researchers that improve the efficiency, safety, and security of the transportation system. #12;#12;CALIFORNIA CENTER

  14. Structural analysis in support of the waterborne transport of radioactive materials

    SciTech Connect (OSTI)

    Ammerman, D.J.

    1996-08-01

    The safety of the transportation of radioactive materials by road and rail has been well studied and documented. However, the safety of waterborne transportation has received much less attention. Recent highly visible waterborne transportation campaigns have led to DOE and IAEA to focus attention on the safety of this transportation mode. In response, Sandia National Laboratories is conducting a program to establish a method to determine the safety of these shipments. As part of that program the mechanics involved in ship-to-ship collisions are being evaluated to determine the loadings imparted to radioactive material transportation packages during these collisions. This paper will report on the results of these evaluations.

  15. Some Ramsey Problems Involving Triangles -Computational Approach

    E-Print Network [OSTI]

    Radziszowski, Stanislaw P.

    Some Ramsey Problems Involving Triangles - Computational Approach Stanislaw P. Radziszowski at the Ramsey Theory Workshop DIMACS, Rutgers University, May 28, 2009 Definition 1. For graphs G and H, R(G, H graphs, and the multicolor Ramsey numbers R(G1, · · · , Gr). Computational problems for Ramsey numbers

  16. Fire Classifications Fires involving the ordinary

    E-Print Network [OSTI]

    Jia, Songtao

    , plastics, etc. Fires involving combustible or flammable liquids such as gasoline, kerosene, oils, grease to prevent fires. Personnel should neither create nor tolerate conditions that could cause or fuel a fire is protected by various devices such as smoke detectors, sprinkler systems, and manual fire alarm pull stations

  17. Environmental science and ecology involve studies

    E-Print Network [OSTI]

    Christensen, Dan

    Environmental science and ecology involve studies of the biosphere, hydro- sphere, and lithosphere in environmental science is conducted on spatial scales varying from a single algal cell to the Earth as a whole's environmental scientists require investigation by an interdisciplinary team, including members from several

  18. Guide to radiological accident considerations for siting and design of DOE nonreactor nuclear facilities

    SciTech Connect (OSTI)

    Elder, J.C.; Graf, J.M.; Dewart, J.M.; Buhl, T.E.; Wenzel, W.J.; Walker, L.J.; Stoker, A.K.

    1986-01-01

    This guide was prepared to provide the experienced safety analyst with accident analysis guidance in greater detail than is possible in Department of Energy (DOE) Orders. The guide addresses analysis of postulated serious accidents considered in the siting and selection of major design features of DOE nuclear facilities. Its scope has been limited to radiological accidents at nonreactor nuclear facilities. The analysis steps addressed in the guide lead to evaluation of radiological dose to exposed persons for comparison with siting guideline doses. Other possible consequences considered are environmental contamination, population dose, and public health effects. Choices of models and parameters leading to estimation of source terms, release fractions, reduction and removal factors, dispersion and dose factors are discussed. Although requirements for risk analysis have not been established, risk estimates are finding increased use in siting of major nuclear facilities, and are discussed in the guide. 3 figs., 9 tabs.

  19. Accident Generated Particulate Materials and Their Characteristics -- A Review of Background Information

    SciTech Connect (OSTI)

    Sutter, S. L.

    1982-05-01

    Safety assessments and environmental impact statements for nuclear fuel cycle facilities require an estimate of the amount of radioactive particulate material initially airborne (source term) during accidents. Pacific Northwest Laboratory (PNL) has surveyed the literature, gathering information on the amount and size of these particles that has been developed from limited experimental work, measurements made from operational accidents, and known aerosol behavior. Information useful for calculating both liquid and powder source terms is compiled in this report. Potential aerosol generating events discussed are spills, resuspension, aerodynamic entrainment, explosions and pressurized releases, comminution, and airborne chemical reactions. A discussion of liquid behavior in sprays, sparging, evaporation, and condensation as applied to accident situations is also included.

  20. RELAP5 Application to Accident Analysis of the NIST Research Reactor

    SciTech Connect (OSTI)

    Baek, J.; Cuadra Gascon, A.; Cheng, L.Y.; Diamond, D.

    2012-03-18

    Detailed safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The time-dependent analysis of the primary system is determined with a RELAP5 transient analysis model that includes the reactor vessel, the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. A post-processing of the simulation results has been conducted to evaluate minimum critical heat flux ratio (CHFR) using the Sudo-Kaminaga correlation. Evaluations are performed for the following accidents: (1) the control rod withdrawal startup accident and (2) the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur because of sufficient coolant flow through the fuel channels and the negative scram reactivity insertion.

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

    SciTech Connect (OSTI)

    Jae, M.; Apostolakis, G.E. (California Univ., Los Angeles, CA (United States). Dept. of Mechanical, Aerospace and Nuclear Engineering)

    1992-08-01

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

  2. Analysis of Reactivity Induced Accident for Control Rods Ejection with Loss of Cooling

    E-Print Network [OSTI]

    Saad, Hend Mohammed El Sayed; Wahab, Moustafa Aziz Abd El

    2013-01-01

    Understanding of the time-dependent behavior of the neutron population in nuclear reactor in response to either a planned or unplanned change in the reactor conditions, is a great importance to the safe and reliable operation of the reactor. In the present work, the point kinetics equations are solved numerically using stiffness confinement method (SCM). The solution is applied to the kinetics equations in the presence of different types of reactivities and is compared with different analytical solutions. This method is also used to analyze reactivity induced accidents in two reactors. The first reactor is fueled by uranium and the second is fueled by plutonium. This analysis presents the effect of negative temperature feedback with the addition positive reactivity of control rods to overcome the occurrence of control rod ejection accident and damaging of the reactor. Both power and temperature pulse following the reactivity- initiated accidents are calculated. The results are compared with previous works and...

  3. Analysis of Reactivity Induced Accident for Control Rods Ejection with Loss of Cooling

    E-Print Network [OSTI]

    Hend Mohammed El Sayed Saad; Hesham Mohammed Mohammed Mansour; Moustafa Aziz Abd El Wahab

    2013-06-05

    Understanding of the time-dependent behavior of the neutron population in nuclear reactor in response to either a planned or unplanned change in the reactor conditions, is a great importance to the safe and reliable operation of the reactor. In the present work, the point kinetics equations are solved numerically using stiffness confinement method (SCM). The solution is applied to the kinetics equations in the presence of different types of reactivities and is compared with different analytical solutions. This method is also used to analyze reactivity induced accidents in two reactors. The first reactor is fueled by uranium and the second is fueled by plutonium. This analysis presents the effect of negative temperature feedback with the addition positive reactivity of control rods to overcome the occurrence of control rod ejection accident and damaging of the reactor. Both power and temperature pulse following the reactivity- initiated accidents are calculated. The results are compared with previous works and satisfactory agreement is found.

  4. Superheated-steam test of ethylene propylene rubber cables using a simultaneous aging and accident environment

    SciTech Connect (OSTI)

    Bennett, P.R.; St. Clair, S.D.; Gilmore, T.W.

    1986-06-01

    The superheated-steam test exposed different ethylene propylene rubber (EPR) cables and insulation specimens to simultaneous aging and a 21-day simultaneous accident environment. In addition, some insulation specimens were exposed to five different aging conditions prior to the 21-day simultaneous accident simulation. The purpose of this superheated-steam test (a follow-on to the saturated-steam tests (NUREG/CR-3538)) was to: (1) examine electrical degradation of different configurations of EPR cables; (2) investigate differences between using superheated-steam or saturated-steam at the start of an accident simulation; (3) determine whether the aging technique used in the saturated-steam test induced artificial degradation; and (4) identify the constituents in EPR that affect moisture absorption.

  5. Transport moments beyond the leading order

    E-Print Network [OSTI]

    Gregory Berkolaiko; Jack Kuipers

    2011-07-11

    For chaotic cavities with scattering leads attached, transport properties can be approximated in terms of the classical trajectories which enter and exit the system. With a semiclassical treatment involving fine correlations between such trajectories we develop a diagrammatic technique to calculate the moments of various transport quantities. Namely, we find the moments of the transmission and reflection eigenvalues for systems with and without time reversal symmetry. We also derive related quantities involving an energy dependence: the moments of the Wigner delay times and the density of states of chaotic Andreev billiards, where we find that the gap in the density persists when subleading corrections are included. Finally, we show how to adapt our techniques to non-linear statistics by calculating the correlation between transport moments. In each setting, the answer for the $n$-th moment is obtained for arbitrary $n$ (in the form of a moment generating function) and for up to the three leading orders in terms of the inverse channel number. Our results suggest patterns which should hold for further corrections and by matching with the low order moments available from random matrix theory we derive likely higher order generating functions.

  6. Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

    SciTech Connect (OSTI)

    Carbajo, Juan; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Schmidt, Rodney Cannon; Thomas, Justin; Wei, Tom; Sofu, Tanju; Ludewig, Hans; Tobita, Yoshiharu; Ohshima, Hiroyuki; Serre, Frederic

    2011-06-01

    This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the experienced user-base and the experimental validation base was decaying away quickly.

  7. TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

    SciTech Connect (OSTI)

    Joy L. Rempe; Darrell L. Knudson

    2013-03-01

    The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation process that could be implemented in upcoming Fukushima Daiichi recovery efforts.

  8. TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

    SciTech Connect (OSTI)

    Joy L. Rempe; Darrell L. Knudson

    2014-05-01

    The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation process that could be implemented in upcoming Fukushima Daiichi recovery efforts.

  9. Integrating regional strategic transportation planning and supply chain management : along the path to sustainability

    E-Print Network [OSTI]

    Sgouridis, Sgouris P

    2005-01-01

    A systems perspective for regional strategic transportation planning (RSTP) for freight movements involves an understanding of Supply Chain Management (SCM). This thesis argues that private sector freight shippers and ...

  10. High compliance all-terrain transport and heavy cargo hybrid bicycle

    E-Print Network [OSTI]

    Soto-Fernández, Orlando

    2005-01-01

    A design project was carried out which involved the design, manufacturing, and assembly of a hybrid bicycle. The bicycle was required to operate between two modes, one that permitted fast transport of the operator from one ...

  11. Spent Fuel Transportation Package Response to the Baltimore Tunnel Fire Scenario

    SciTech Connect (OSTI)

    Adkins, Harold E.; Cuta, Judith M.; Koeppel, Brian J.; Guzman, Anthony D.; Bajwa, Christopher S.

    2006-11-15

    On July 18, 2001, a freight train carrying hazardous (non-nuclear) materials derailed and caught fire while passing through the Howard Street railroad tunnel in downtown Baltimore, Maryland. The United States Nuclear Regulatory Commission (USNRC), one of the agencies responsible for ensuring the safe transportation of radioactive materials in the United States, undertook an investigation of the train derailment and fire to determine the possible regulatory implications of this particular event for the transportation of spent nuclear fuel by railroad. Shortly after the accident occurred, the USNRC met with the National Transportation Safety Board (NTSB, the U.S. agency responsible for determining the cause of transportation accidents), to discuss the details of the accident and the ensuing fire. Following these discussions, the USNRC assembled a team of experts from the National Institute of Standards and Technology (NIST), the Center for Nuclear Waste Regulatory Analyses (CNWRA), and Pacific Northwest National Laboratory (PNNL) to determine the thermal conditions that existed in the Howard Street tunnel fire and analyze the effects of this fire on various spent fuel transportation package designs. The Fire Dynamics Simulator (FDS) code, developed by NIST, was used to determine the thermal environment present in the Howard Street tunnel during the fire. The FDS results were used as boundary conditions in the COBRA-SFS and ANSYS® computer codes to evaluate the thermal performance of different package designs. The staff concluded that larger transportation packages resembling the HOLTEC Model No. HI STAR 100 and TransNuclear Model No. TN-68 would withstand a fire with thermal conditions similar to those that existed in the Baltimore tunnel fire event with only minor damage to peripheral components. This is due to their sizable thermal inertia and design specifications in compliance with currently imposed regulatory requirements. The staff also concluded that some components of smaller transportation packages resembling the NAC Model No. LWT, despite placement within an ISO container, could degrade. USNRC staff evaluated the radiological consequences of the package responses to the Baltimore tunnel fire. Though components in some packages heated up beyond their service temperatures, the staff determined that there would be no significant dose as a result of the fire for any of these and similar packages.

  12. An ISP-27 accident scenario for analysis of Krsko Nuclear Power Plant SBLOCA

    SciTech Connect (OSTI)

    Petelin, S.; Mavko, B.; Gortnar, O.; Parzer, I.

    1994-12-31

    The reactor safety analysis group of Jozef Stefan Institute (IJS) has participated in analyses of International Standard Problem 27 (ISP-27), which was based on test 9.1 b performed at the BETHSY experimental facility (France). In addition, we realized the ISP-27 transient scenario in the analysis of a small-break loss-of-coolant accident (SBLOCA) for Krsko nuclear power plant (NPP). The objective was to evaluate the effectiveness of the ISP-27 proposed accident management procedure for a real NPP and to compare the physical phenomena known from experimental background with the phenomena predicted by simulation of a real plant transient.

  13. Analysis of hydrogen mitigation for degraded core accidents in the Sequoyah Nuclear Power Plant

    SciTech Connect (OSTI)

    Berman, M.; Sherman, M.P.; Cummings, J.C.; Baer, M.R.; Griffiths, S.K.

    1981-04-01

    The report presents the results of a scoping investigation to ascertain the effectiveness and practicability of three hydrogen control measures for the Sequoyah Nuclear Power Plant--deliberate ignition, water fogging, and Halon addition after accident initiation. The authors conclude that no one of these hydrogen control measures alone is clearly superior to the other under all accident conditions. Advantages and disadvantages were identified for all control measures. In addition to providing a basic discussion of how each measure works to mitigate or control hydrogen combustion, we have answered specific questions posed by the U. S. Nuclear Regulatory Commission.

  14. LOCA with consequential or delayed LOOP accidents: Unique issues, plant vulnerability, and CDF contributions

    SciTech Connect (OSTI)

    Martinez-Guridi, G.; Samanta, P.; Chu, L.; Yang, J.

    1998-08-01

    A loss-of-coolant accident (LOCA) can cause a loss-of-offsite power (LOOP) wherein the LOOP is usually delayed by few seconds or longer. Such an accident is called LOCA with consequential LOOP, or LOCA with delayed LOOP (here, abbreviated as LOCA/LOOP). This paper analyzes the unique conditions that are associated with a LOCA/LOOP, presents a model, and quantifies its contribution to core damage frequency (CDF). The results show that the CDF contribution can be a dominant contributor to risk for certain plant designs, although boiling water reactors (BWRs) are less vulnerable than pressurized water reactors (PWRs).

  15. Technical Advisory Team (TAT) report on the rocket sled test accident of October 9, 2008.

    SciTech Connect (OSTI)

    Stofleth, Jerome H.; Dinallo, Michael Anthony; Medina, Anthony J.

    2009-01-01

    This report summarizes probable causes and contributing factors that led to a rocket motor initiating prematurely while employees were preparing instrumentation for an AIII rocket sled test at SNL/NM, resulting in a Type-B Accident. Originally prepared by the Technical Advisory Team that provided technical assistance to the NNSA's Accident Investigation Board, the report includes analyses of several proposed causes and concludes that the most probable source of power for premature initiation of the rocket motor was the independent battery contained in the HiCap recorder package. The report includes data, evidence, and proposed scenarios to substantiate the analyses.

  16. NREL: Transportation Research - Transportation Deployment Support

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration wouldMass map shines lightGeospatial ToolkitSMARTSWorking WithSuccess StoriesTransportation

  17. NREL: Transportation Research - Transportation Secure Data Center

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration wouldMass map shines lightGeospatial ToolkitSMARTSWorking WithSuccessTransportation Secure Data

  18. NREL: Transportation Research - Transportation and Hydrogen Newsletter:

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration wouldMass map shines lightGeospatial ToolkitSMARTSWorking WithSuccessTransportation Secure

  19. NREL: Transportation Research - Transportation and Hydrogen Newsletter:

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration wouldMass map shines lightGeospatial ToolkitSMARTSWorking WithSuccessTransportation

  20. Thermal and Electrical Transport in Oxide Heterostructures

    E-Print Network [OSTI]

    Ravichandran, Jayakanth

    2011-01-01

    of thermal conductivity . . . . . . . . . . . . . . . .4.4 Thermal transport in2.3.2 Thermal transport . . . . . . . . . . . . . . . .