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Sample records for transportation accident involving

  1. DECONTAMINATION DRESSDOWN AT A TRANSPORTATION ACCIDENT INVOLVING...

    Office of Environmental Management (EM)

    Video User' s Guide DECONTAMINATION DRESSDOWN AT A TRANSPORTATION ACCIDENT INVOLVING ... related to emergency response to a transportation accident involving radioactive material. ...

  2. EMERGENCY RESPONSE TO A TRANSPORTATION ACCIDENT INVOLVING RADIOACTIVE...

    Office of Environmental Management (EM)

    ransportation ransportation ransportation ransportation Accident Involving Radioactive Material Accident Involving Radioactive Material Accident Involving Radioactive ...

  3. Accident resistant transport container

    DOE Patents [OSTI]

    Andersen, John A.; Cole, James K.

    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.

  4. Accident resistant transport container

    DOE Patents [OSTI]

    Anderson, J.A.; Cole, K.K.

    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.

  5. A statistical description of the types and severities of accidents involving tractor semi-trailers

    SciTech Connect (OSTI)

    Clauss, D.B.; Wilson, R.K.; Blower, D.F.; Campbell, K.L.

    1994-06-01

    This report provides a statistical description of the types and severities of tractor semi-trailer accidents involving at least one fatality. The data were developed for use in risk assessments of hazardous materials transportation. Several accident databases were reviewed to determine their suitability to the task. The TIFA (Trucks Involved in Fatal Accidents) database created at the University of Michigan Transportation Research Institute was extensively utilized. Supplementary data on collision and fire severity, which was not available in the TIFA database, were obtained by reviewing police reports for selected TIFA accidents. The results are described in terms of frequencies of different accident types and cumulative distribution functions for the peak contact velocity, rollover skid distance, fire temperature, fire size, fire separation, and fire duration.

  6. Health Physics Code System for Evaluating Accidents Involving Radioactive Materials.

    Energy Science and Technology Software Center (OSTI)

    2014-10-01

    Version 03 The HOTSPOT Health Physics codes were created to provide Health Physics personnel with a fast, field-portable calculational tool for evaluating accidents involving radioactive materials. HOTSPOT codes provide a first-order approximation of the radiation effects associated with the atmospheric release of radioactive materials. The developer's website is: http://www.llnl.gov/nhi/hotspot/. Four general programs, PLUME, EXPLOSION, FIRE, and RESUSPENSION, calculate a downwind assessment following the release of radioactive material resulting from a continuous or puff release, explosivemore » release, fuel fire, or an area contamination event. Additional programs deal specifically with the release of plutonium, uranium, and tritium to expedite an initial assessment of accidents involving nuclear weapons. The FIDLER program can calibrate radiation survey instruments for ground survey measurements and initial screening of personnel for possible plutonium uptake in the lung. The HOTSPOT codes are fast, portable, easy to use, and fully documented in electronic help files. HOTSPOT supports color high resolution monitors and printers for concentration plots and contours. The codes have been extensively used by the DOS community since 1985. Tables and graphical output can be directed to the computer screen, printer, or a disk file. The graphical output consists of dose and ground contamination as a function of plume centerline downwind distance, and radiation dose and ground contamination contours. Users have the option of displaying scenario text on the plots. HOTSPOT 3.0.1 fixes three significant Windows 7 issues: � Executable installed properly under "Program Files/HotSpot 3.0". Installation package now smaller: removed dependency on older Windows DLL files which previously needed to \\ � Forms now properly scale based on DPI instead of font for users who change their screen resolution to something other than 100%. This is a more common feature in Windows 7

  7. Site Specific Analyses of a Spent Nuclear Fuel Transportation Accident

    SciTech Connect (OSTI)

    Biwer, B. M.; Chen, S. Y.

    2003-02-24

    The number of spent nuclear fuel (SNF) shipments is expected to increase significantly during the time period that the United States' inventory of SNF is sent to a final disposal site. Prior work estimated that the highest accident risks of a SNF shipping campaign to the proposed geologic repository at Yucca Mountain were in the corridor states, such as Illinois. The largest potential human health impacts would be expected to occur in areas with high population densities such as urban settings. Thus, our current study examined the human health impacts from the most plausible severe SNF transportation accidents in the Chicago metropolitan area. The RISKIND 2.0 program was used to model site-specific data for an area where the largest impacts might occur. The results have shown that the radiological human health consequences of a severe SNF rail transportation accident on average might be similar to one year of exposure to natural background radiation for those persons living a nd working in the most affected areas downwind of the actual accident location. For maximally exposed individuals, an exposure similar to about two years of exposure to natural background radiation was estimated. In addition to the accident probabilities being very low (approximately 1 chance in 10,000 or less during the entire shipping campaign), the actual human health impacts are expected to be lower if any of the accidents considered did occur, because the results are dependent on the specific location and weather conditions, such as wind speed and direction, that were selected to maximize the results. Also, comparison of the results of longer duration accident scenarios against U.S. Environmental Protection Agency guidelines was made to demonstrate the usefulness of this site-specific analysis for emergency planning purposes.

  8. 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.

  9. Probabilistic assessment of spent fuel shipping cask response to severe transportation accident conditions. Report summary

    SciTech Connect (OSTI)

    Fischer, L.E.; Kimura, C.Y.; Witte, M.C.

    1985-01-01

    The licensing of commercial nuclear spent shipping casks in the United States is regulated by 10CFR71. In order to be licensed, casks must be designed not to fail under hypothetical test conditions specified in Appendix B of this regulation. Questions have been raised about the suitability of these tests in simulating actual transportation accident conditions. Our study addresses the adequacy of current regulations by comparing real-world accident conditions with regulatory test specifications using more complete accident statistics and more sophisticated structural analyses than have been used in studies to date. Our objective is to evaluate the protection provided by current regulations against severe accident conditions for commercial spent nuclear fuel casks that are transported by truck or rail. The complete spectrum of truck and rail accidents will be reviewed in order to determine the frequency (or infrequency) of cask failures during transportation accidents. 3 references, 1 figure.

  10. 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.

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

    SciTech Connect (OSTI)

    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.

  12. 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...

  13. Emergency Response to a Transportation Accident Involving Radioactive Material

    Office of Energy Efficiency and Renewable Energy (EERE)

    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...

  14. A Computer Code To Analyze The Gas-Phase Transport of Fission Products In Reactor Cooling System Under Severe Accidents.

    Energy Science and Technology Software Center (OSTI)

    1990-12-06

    Version 00 HORN calculates the transport of volatile fission products in a dry primary cooling circuit under severe accidents of water reactors.

  15. 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.

  16. Effects of molybdenum and silver on iodine transport in primary circuit on severe nuclear accidents

    SciTech Connect (OSTI)

    Kalilainen, J.; Rantanen, P.; Karkela, T.; Lipponen, M.; Auvinen, A.; Jokiniemi, J.

    2012-07-01

    This experimental study was a continuation of the study conducted at VTT to investigate the effects of reactions on primary circuit surfaces to transport of gaseous and aerosol phase iodine during the hypothetical severe nuclear accident. Cesium iodide was used as a precursor in every experiment. In the experiments it was observed that the hydrogen in the atmosphere decreased the fraction of released gaseous iodine. As the temperature was lowered, less iodine was released, but the fraction of gaseous iodine from the overall released iodine was increased. As molybdenum trioxide was introduced to the precursor, the fraction of gaseous iodine from the overall released iodine was increased significantly. Also, Mo decreased the transport of Cs and caused significant depositions to the reaction furnace. Addition of silver to the CsI precursor at 650 deg. C decreased the release of iodine as well as the fraction of gaseous iodine. At 400 deg. C, Ag + CsI as well as Ag + MoO{sub 3} + CsI precursor significantly increased the release of gaseous iodine, where almost no aerosol particles were released. With B{sub 2}O{sub 3} + CsI precursor it was observed that in the atmosphere without H{sub 2}O, the released iodine was mostly in gaseous form. (authors)

  17. DOE Partnerships with States, Tribes and Other Federal Programs Help Responders Prepare for Challenges Involving Transport of Radioactive Materials

    SciTech Connect (OSTI)

    Marsha Keister

    2001-02-01

    DOE Partnerships with States, Tribes and Other Federal Programs Help Responders Prepare for Challenges Involving Transport of Radioactive Materials Implementing adequate institutional programs and validating preparedness for emergency response to radiological transportation incidents along or near U.S. Department of Energy (DOE) shipping corridors poses unique challenges to transportation operations management. Delayed or insufficient attention to State and Tribal preparedness needs may significantly impact the transportation operations schedule and budget. The DOE Transportation Emergency Preparedness Program (TEPP) has successfully used a cooperative planning process to develop strong partnerships with States, Tribes, Federal agencies and other national programs to support responder preparedness across the United States. DOE TEPP has found that building solid partnerships with key emergency response agencies ensures responders have access to the planning, training, technical expertise and assistance necessary to safely, efficiently and effectively respond to a radiological transportation accident. Through the efforts of TEPP over the past fifteen years, partnerships have resulted in States and Tribal Nations either using significant portions of the TEPP planning resources in their programs and/or adopting the Modular Emergency Response Radiological Transportation Training (MERRTT) program into their hazardous material training curriculums to prepare their fire departments, law enforcement, hazardous materials response teams, emergency management officials, public information officers and emergency medical technicians for responding to transportation incidents involving radioactive materials. In addition, through strong partnerships with Federal Agencies and other national programs TEPP provided technical expertise to support a variety of radiological response initiatives and assisted several programs with integration of the nationally recognized MERRTT program

  18. Potential health risks from postulated accidents involving the Pu-238 RTG (radioisotope thermoelectric generator) on the Ulysses solar exploration mission

    SciTech Connect (OSTI)

    Goldman, M. ); Nelson, R.C. ); Bollinger, L. ); Hoover, M.D. . Inhalation Toxicology Research Inst.); Templeton, W. ); Anspaugh, L. (Lawren

    1990-11-02

    Potential radiation impacts from launch of the Ulysses solar exploration experiment were evaluated using eight postulated accident scenarios. Lifetime individual dose estimates rarely exceeded 1 mrem. Most of the potential health effects would come from inhalation exposures immediately after an accident, rather than from ingestion of contaminated food or water, or from inhalation of resuspended plutonium from contaminated ground. For local Florida accidents (that is, during the first minute after launch), an average source term accident was estimated to cause a total added cancer risk of up to 0.2 deaths. For accidents at later times after launch, a worldwide cancer risk of up to three cases was calculated (with a four in a million probability). Upper bound estimates were calculated to be about 10 times higher. 83 refs.

  19. Potential health risks from postulated accidents involving the Pu-238 RTG on the Ulysses solar exploration mission

    SciTech Connect (OSTI)

    Goldman, M. ); Nelson, R.C. ); Bollinger, L. ); Hoover, M.D. ); Templeton, W. ); Anspaugh, L. )

    1991-01-01

    Potential radiation impacts from launch of the Ulysses solar exploration experiment were evaluated using eight postulated accident scenarios. Lifetime individual dose estimates rarely exceeded 1 mrem. Most of the potential health effects would come from inhalation exposures immediately after an accident, rather than from ingestion of contaminated food or water, or from inhalation of resuspended plutonium from contaminated ground. For local Florida accidents (that is, during the first minute after launch), an average source term accident was estimated to cause a total added cancer risk of up to 0.2 deaths. For accidents at later times after launch, a worldwide cancer risk of up to three cases was calculated (with a four in a million probability). Upper bound estimates were calculated to be about 10 times higher.

  20. Microsoft Word - Unrelated Accident

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

    For Immediate Release Truck Accident Did Not Involve WIPP Shipment CARLSBAD, N.M., October 1, 2009 - A Wednesday night truck accident north of Albuquerque on Highway 165 that ...

  1. Severe Accident Studies | Department of Energy

    Office of Environmental Management (EM)

    accidents and safety. PDF icon Severe Accident Studies More Documents & Publications Spent Fuel Transportation Risk Assessment DOE-STD-101-92 EIS-0218-SA-07: Supplement Analysis

  2. Accident motivates scholarship recipient

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

    Accident Response Group NNSA's Accident Response Group (ARG) provides technical guidance and responds to U.S. nuclear weapons accidents. ARG_Logo The team assists in assessing weapons damage and risk, and in developing and implementing procedures for safe weapon recovery, packaging, transportation, and disposal of damaged weapons. The ARG headquarters is located in Albuquerque, New Mexico and is supported by Lawrence Livermore National Laboratory, Los Alamos National Laboratory, Sandia National

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

    SciTech Connect (OSTI)

    Kang, Qinjin; Lichtner, Peter C; Viswanathan, Hari S; Abdel-fattah, Amr I

    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.

  4. 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

  5. LPG land transportation and storage safety. Final report

    SciTech Connect (OSTI)

    Not Available

    1981-09-01

    This report contains an analytical examination of fatal accidents involving liquefied petroleum gas (LPG) releases during transportation and/or transportation related storage. Principal emphasis was on accidents during the nine-year period 1971 to 1979. Fatalities to members of the general public (i.e., those at the scene of the accident through coincidence or curiosity) were of special interest. Transportation accidents involving railroad tank cars, trucks, and pipelines were examined as were accidents at storage facilities, including loading and unloading at such facilities. The main sources of the necessary historical accident data were the accident reports submitted to the Department of Transportation by LPG carriers, National Transportation Safety Board accident reports, articles in the National Fire Protection Association journals, other literature, and personal interviews with firemen, company personnel, and others with knowledge of certain accidents. The data indicate that, on the average, releases of LPG during transportation and intermediate storage cause approximately six fatalities per year to members of the general public. The individual risk is about 1 death per 37,000,000 persons; about the same as the risk of a person on the ground being killed by an airplane crash, and much less than the risk of death by lightning, tornadoes, or dam failures.

  6. LPG land transportation and storage safety. Final report

    SciTech Connect (OSTI)

    Martinsen, W.E.; Cavin, W.D.

    1981-09-01

    This report contains an analytical examination of fatal accidents involving liquefied petroleum gas (LPG) releases during transportation and/or transportation related storage. Principal emphasis was on accidents during the nine-year period 1971 through 1979. Fatalities to members of the general public (i.e., those at the scene of the accident through coincidence or curiosity) were of special interest. Transportation accidents involving railroad tank cars, trucks, and pipelines were examined as were accidents at storage facilities, including loading and unloading at such facilities. The main sources of the necessary historical accident data were the accident reports submitted to the Department of Transportation by LPG carriers, National Transportation Safety Board accident reports, articles in the National Fire Protection Association journals, other literature, and personal interviews with firemen, company personnel, and others with knowledge of certain accidents. The data indicate that, on the average, releases of LPG during transportation and intermediate storage cause approximately six fatalities per year to members of the general public. The individual risk is about 1 death per 37,000,000 persons; about the same as the risk of a person on the ground being killed by an airplane crash, and much less than the risk of death by lightning, tornadoes, or dam failures.

  7. 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.

  8. 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

  9. 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

  10. COMMERCIAL SNF ACCIDENT RELEASE FRACTIONS

    SciTech Connect (OSTI)

    S.O. Bader

    1999-10-18

    The purpose of this design analysis is to specify and document the total and respirable fractions for radioactive materials that are released from an accident event at the Monitored Geologic Repository (MGR) involving commercial spent nuclear fuel (CSNF) in a dry environment. The total and respirable release fractions will be used to support the preclosure licensing basis for the MGR. The total release fraction is defined as the fraction of total CSNF 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. The 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. Potential accidents may involve waste forms that are characterized as either bare (unconfined) fuel assemblies or confined fuel assemblies. The confined CSNF assemblies at the MGR are contained in shipping casks, canisters, or disposal containers (waste packages). In contrast to the bare fuel assemblies, the container that confines the fuel assemblies has the potential of providing an additional barrier for diminishing the total release fraction should the fuel rod cladding breach during an accident. However, this analysis will not take credit for this additional bamer and will establish only the total release fractions for bare unconfined CSNF assemblies, which may however be

  11. Transportation safety training

    SciTech Connect (OSTI)

    Jones, E.

    1990-01-01

    Over the past 25 years extensive federal legislation involving the handling and transport of hazardous materials/waste has been passed that has resulted in numerous overlapping regulations administered and enforced by different federal agencies. The handling and transport of hazardous materials/waste involves a significant number of workers who are subject to a varying degree of risk should an accident occur during handling or transport. Effective transportation training can help workers address these risks and mitigate them, and at the same time enable ORNL to comply with the federal regulations concerning the transport of hazardous materials/waste. This presentation will outline how the Environmental and Health Protection Division's Technical Resources and Training Section at the Oak Ridge National Laboratory, working with transportation and waste disposal personnel, have developed and implemented a comprehensive transportation safety training program to meet the needs of our workers while satisfying appropriate federal regulations. 8 refs., 3 tabs.

  12. 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.

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

    SciTech Connect (OSTI)

    Farmer, M. T.

    2015-11-01

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

  14. 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.

  15. DOE TMD transportation training module 14 transportation of explosives

    SciTech Connect (OSTI)

    Griffith, R.L. Jr.

    1994-07-01

    The Department of Energy Transportation Management Division has developed training module 14, entitled {open_quotes}Transportation of Explosives{close_quotes} to compliment the basic {open_quotes}core ten{close_quotes} training modules of the Hazardous Materials Modular Training Program. The purpose of this training module is to increase awareness of the Department of Transportation (DOT) requirements concerning the packaging and transportation of explosives. Topics covered in module 14 include the classification of explosives, approval and registration of explosives, packaging requirements, hazard communication requirements, separation and segregation compatibility requirements, loading and unloading operations, as well as safety measures required in the event of a vehicle accident involving explosives.

  16. 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.

  17. 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

  18. 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.

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

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

    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 ...

  20. Accident Investigation Handbook

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

    Accident Investigations (AI) and Operational Safety Reviews (OSR) are valuable for ... in conducting Department of Energy (DOE) accident investigations over the past many years. ...

  1. 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.

  2. Assessment of the risk of transporting propane by truck and train

    SciTech Connect (OSTI)

    Geffen, C.A.

    1980-03-01

    The risk of shipping propane is discussed and the risk assessment methodology is summarized. The risk assessment model has been constructed as a series of separate analysis steps to allow the risk to be readily reevaluated as additional data becomes available or as postulated system characteristics change. The transportation system and accident environment, the responses of the shipping system to forces in transportation accidents, and release sequences are 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 comparison with other reports in this series. Based on the information presented, accidents involving tank truck shipments of propane will be expected to occur at a rate of 320 every year; accidents involving bobtails would be expected at a rate of 250 every year. Train accidents involving propane shipments would be expected to occur at a rate of about 60 every year. A release of any amount of material from propane trucks, under both normal transportation and transport accident conditions, is to be expected at a rate of about 110 per year. Releases from propane rail tank cars would occur about 40 times a year. However, only those releases that occur during a transportation accident or involve a major tank defect will include sufficient propane to present the potential for danger to the public. These significant releases can be expected at the lower rate of about fourteen events per year for truck transport and about one event every two years for rail tank car transport. The estimated number of public fatalities resulting from these significant releases in 1985 is fifteen. About eleven fatalities per year result from tank truck operation, and approximately half a death per year stems from the movement of propane in rail tank cars.

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

    Office of Environmental Management (EM)

    7, 2012, Construction Accident - Structural Steel Collapse at The Over pack Storage ... Accident Investigation of the June 17, 2012, Construction Accident - Structural Steel ...

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

    Office of Environmental Management (EM)

    an Accident Investigation Board to investigate an underground mine fire involving a salt haul truck occurred at DOE's WIPP near Carlsbad, New Mexico. The Board began the...

  5. Radionuclide release calculations for selected severe accident scenarios

    SciTech Connect (OSTI)

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

    1990-08-01

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

  6. transportation

    National Nuclear Security Administration (NNSA)

    security missions undertaken by the U.S. government.

    Pantex Plant's Calvin Nelson honored as Analyst of the Year for Transportation Security http:nnsa.energy.gov...

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

    SciTech Connect (OSTI)

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

    1996-01-12

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

  8. Accident Response Group | National Nuclear Security Administration | (NNSA)

    National Nuclear Security Administration (NNSA)

    Accident Response Group NNSA's Accident Response Group (ARG) provides technical guidance and responds to U.S. nuclear weapons accidents. ARG_Logo The team assists in assessing weapons damage and risk, and in developing and implementing procedures for safe weapon recovery, packaging, transportation, and disposal of damaged weapons. The ARG headquarters is located in Albuquerque, New Mexico and is supported by Lawrence Livermore National Laboratory, Los Alamos National Laboratory, Sandia National

  9. Transportation training: Focusing on movement of hazardous substances and wastes

    SciTech Connect (OSTI)

    Jones, E.; Moreland, W.M.

    1988-01-01

    Over the past 25 years extensive federal legislation involving the handling and transport of hazardous materials/waste has been passed that has resulted in numerous overlapping regulations administered and enforced by different federal agencies. The handling and transport of hazardous materials/waste involves a significant number of workers who are subject to a varying degree of risk should an accident occur during handling or transport. Effective transportation training can help workers address these risks and mitigate them, and at the same time enable ORNL to comply with the federal regulations concerning the transport of hazardous materials/waste. This presentation will outline how the Environmental and Health Protection Division's Technical Resources and Training Program at the Oak Ridge National Laboratory, working with transportation and waste disposal personnel, are developing and implementing a comprehensive transportation safety training program to meet the needs of our workers while satisfying appropriate federal regulations. 8 refs., 5 figs., 3 tabs.

  10. HTGR severe accident sequence analysis

    SciTech Connect (OSTI)

    Harrington, R.M.; Ball, S.J.; Kornegay, F.C.

    1982-01-01

    Thermal-hydraulic, fission product transport, and atmospheric dispersion calculations are presented for hypothetical severe accident release paths at the Fort St. Vrain (FSV) high temperature gas cooled reactor (HTGR). Off-site radiation exposures are calculated for assumed release of 100% of the 24 hour post-shutdown core xenon and krypton inventory and 5.5% of the iodine inventory. The results show conditions under which dose avoidance measures would be desirable and demonstrate the importance of specific release characteristics such as effective release height. 7 tables.

  11. Minimising the error in eigenvalue calculations involving the Boltzmann transport equation using goal-based adaptivity on unstructured meshes

    SciTech Connect (OSTI)

    Goffin, Mark A.; Baker, Christopher M.J.; Buchan, Andrew G.; Pain, Christopher C.; Eaton, Matthew D.; Smith, Paul N.

    2013-06-01

    This article presents a method for goal-based anisotropic adaptive methods for the finite element method applied to the Boltzmann transport equation. The neutron multiplication factor, k{sub eff}, is used as the goal of the adaptive procedure. The anisotropic adaptive algorithm requires error measures for k{sub eff} with directional dependence. General error estimators are derived for any given functional of the flux and applied to k{sub eff} to acquire the driving force for the adaptive procedure. The error estimators require the solution of an appropriately formed dual equation. Forward and dual error indicators are calculated by weighting the Hessian of each solution with the dual and forward residual respectively. The Hessian is used as an approximation of the interpolation error in the solution which gives rise to the directional dependence. The two indicators are combined to form a single error metric that is used to adapt the finite element mesh. The residual is approximated using a novel technique arising from the sub-grid scale finite element discretisation. Two adaptive routes are demonstrated: (i) a single mesh is used to solve all energy groups, and (ii) a different mesh is used to solve each energy group. The second method aims to capture the benefit from representing the flux from each energy group on a specifically optimised mesh. The k{sub eff} goal-based adaptive method was applied to three examples which illustrate the superior accuracy in criticality problems that can be obtained.

  12. 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.

  13. Chernobyl accident: A comprehensive risk assessment

    SciTech Connect (OSTI)

    Vargo, G.J.; Poyarkov, V.; Baryakhtar, V.; Kukhar, V.; Los, I.

    1999-01-01

    The authors, all of whom are Ukrainian and Russian scientists involved with Chernobyl nuclear power plant since the April 1986 accident, present a comprehensive review of the accident. In addition, they present a risk assessment of the remains of the destroyed reactor and its surrounding shelter, Chernobyl radioactive waste storage and disposal sites, and environmental contamination in the region. The authors explore such questions as the risks posed by a collapse of the shelter, radionuclide migration from storage and disposal facilities in the exclusion zone, and transfer from soil to vegetation and its potential regional impact. The answers to these questions provide a scientific basis for the development of countermeasures against the Chernobyl accident in particular and the mitigation of environmental radioactive contamination in general. They also provide an important basis for understanding the human health and ecological risks posed by the accident.

  14. WIPP Documents - Transportation

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

    Transportation

  15. Alisol B 23-acetate protects against ANIT-induced hepatotoxity and cholestasis, due to FXR-mediated regulation of transporters and enzymes involved in bile acid homeostasis

    SciTech Connect (OSTI)

    Meng, Qiang; Chen, Xin-li; Wang, Chang-yuan; Liu, Qi; Sun, Hui-jun; Sun, Peng-yuan; Huo, Xiao-kui; Liu, Zhi-hao; Yao, Ji-hong; Liu, Ke-xin

    2015-03-15

    Intrahepatic cholestasis is a clinical syndrome with systemic and intrahepatic accumulation of excessive toxic bile acids that ultimately cause hepatobiliary injury. Appropriate regulation of bile acids in hepatocytes is critically important for protection against liver injury. In the present study, we characterized the protective effect of alisol B 23-acetate (AB23A), a natural triterpenoid, on alpha-naphthylisothiocyanate (ANIT)-induced liver injury and intrahepatic cholestasis in mice and further elucidated the mechanisms in vivo and in vitro. AB23A treatment dose-dependently protected against liver injury induced by ANIT through reducing hepatic uptake and increasing efflux of bile acid via down-regulation of hepatic uptake transporters (Ntcp) and up-regulation of efflux transporter (Bsep, Mrp2 and Mdr2) expression. Furthermore, AB23A reduced bile acid synthesis through repressing Cyp7a1 and Cyp8b1, increased bile acid conjugation through inducing Bal, Baat and bile acid metabolism through an induction in gene expression of Sult2a1. We further demonstrate the involvement of farnesoid X receptor (FXR) in the hepatoprotective effect of AB23A. The changes in transporters and enzymes, as well as ameliorative liver histology in AB23A-treated mice were abrogated by FXR antagonist guggulsterone in vivo. In vitro evidences also directly demonstrated the effect of AB23A on FXR activation in a dose-dependent manner using luciferase reporter assay in HepG2 cells. In conclusion, AB23A produces protective effect against ANIT-induced hepatotoxity and cholestasis, due to FXR-mediated regulation of transporters and enzymes. - Highlights: • AB23A has at least three roles in protection against ANIT-induced liver injury. • AB23A decreases Ntcp, and increases Bsep, Mrp2 and Mdr2 expression. • AB23A represses Cyp7a1 and Cyp8b1 through inducing Shp and Fgf15 expression. • AB23A increases bile acid metabolism through inducing Sult2a1 expression. • FXR activation is involved

  16. 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.

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

    SciTech Connect (OSTI)

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

    1993-09-01

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

  18. 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

  19. 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

  20. Federally Led Accident Investigation Reports | Department of...

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

    Federally Led Accident Investigation Reports Federally Led Accident Investigation Reports Includes Pre-March 2011 Type A Reports June 1, 1999 Type A Accident Investigation Board...

  1. Chernobyl Nuclear Accident | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Chernobyl Nuclear Accident Chernobyl Nuclear Accident Chernobyl, Ukraine A catastrophic nuclear accident occurs at Chernobyl Reactor 4 in the then Soviet Republic of Ukraine

  2. Type B Accident Investigation Board Report Subcontractor Radioactive

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

    Release During Transportation Activities on May 14, 2004, Bechtel Jacobs Company LLC, Oak Ridge, Tennessee (Amended) | Department of Energy Subcontractor Radioactive Release During Transportation Activities on May 14, 2004, Bechtel Jacobs Company LLC, Oak Ridge, Tennessee (Amended) Type B Accident Investigation Board Report Subcontractor Radioactive Release During Transportation Activities on May 14, 2004, Bechtel Jacobs Company LLC, Oak Ridge, Tennessee (Amended) August 17, 2004 On Friday,

  3. Type B Accident Investigation Report of the October 28, 2004, Burn Injuries

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

    Sustained During an Office of Secure Transportation Joint Training Exercise at Fort Hunter-Liggett, CA | Department of Energy 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 of the October 28, 2004, Burn Injuries Sustained During an Office of Secure Transportation Joint Training Exercise at Fort Hunter-Liggett, CA February 1, 2005 TYPE B Accident Investigation

  4. Longitudinal review of state-level accident statistics for carriers of interstate freight

    SciTech Connect (OSTI)

    Saricks, C.; Kvitek, T.

    1994-03-01

    State-level accident rates by mode of freight transport have been developed and refined for application to the US Department of Energy`s (DOE`s) environmental mitigation program, which may involve large-quantity shipments of hazardous and mixed wastes from DOE facilities. These rates reflect multi-year data for interstate-registered highway earners, American Association of Railroads member carriers, and coastal and internal waterway barge traffic. Adjustments have been made to account for the share of highway combination-truck traffic actually attributable to interstate-registered carriers and for duplicate or otherwise inaccurate entries in the public-use accident data files used. State-to-state variation in rates is discussed, as is the stability of rates over time. Computed highway rates have been verified with actual carriers of high- and low-level nuclear materials, and the most recent truck accident data have been used, to ensure that the results are of the correct order of magnitude. Study conclusions suggest that DOE use the computed rates for the three modes until (1) improved estimation techniques for highway combination-truck miles by state become available; (2) continued evolution of the railroad industry significantly increases the consolidation of interstate rail traffic onto fewer high-capacity trunk lines; or (3) a large-scale off-site waste shipment campaign is imminent.

  5. Severe Accident Modeling

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

    Severe Accident Modeling - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Energy Defense Waste Management Programs Advanced

  6. 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.

  7. Accident Investigation Report- Fire Report

    Broader source: Energy.gov [DOE]

    On February 7, 2014, Deputy Assistant Secretary, Safety, Security, and Quality Programs Environmental Management, DOE, formally appointed an Accident Investigation Board to investigate an...

  8. Guidance for Radiation Accident Management

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

    Procedure Demonstration Introduction Radioactive materials are among the many kinds of hazardous substances emergency responders might have to deal with in an accident. It is ...

  9. No Recordable Accidents in 17,000 Waste Shipments Across 3.4 Million Miles

    Broader source: Energy.gov [DOE]

    WASHINGTON, D.C. – No U.S. Department of Transportation recordable accidents resulted from DOE’s nearly 17,000 radioactive, hazardous material and waste shipments across 3.4 million miles in fiscal 2015.

  10. 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.

  11. 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.

  12. WIPP transportation exercise to test emergency response capablities for Midland-Odessa

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

    Transportation Exercise to Test Emergency Response Capabilities for Midland-Odessa CARLSBAD, N.M., January 10, 2000 - Emergency response agencies from Midland and Odessa, Texas, will take part in a 1 p.m. (CST) training exercise Jan. 12 at the Ector County Coliseum. The graded exercise will help agencies determine whether emergency personnel are prepared to respond to a possible accident involving a shipment of transuranic radioactive waste headed for the U.S. Department of Energy's (DOE) Waste

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

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

    Additionally, HSS has developed: SAF-230 - 40 hour Accident Investigator, and the online SAF-230DE Accident Investigation Orientation, training courses. For more information see ...

  14. 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.

  15. 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.

  16. Revised accident source terms for light-water reactors

    SciTech Connect (OSTI)

    Soffer, L.

    1995-02-01

    This paper presents revised accident source terms for light-water reactors incorporating the severe accident research insights gained in this area over the last 15 years. Current LWR reactor accident source terms used for licensing date from 1962 and are contained in Regulatory Guides 1.3 and 1.4. These specify that 100% of the core inventory of noble gases and 25% of the iodine fission products are assumed to be instantaneously available for release from the containment. The chemical form of the iodine fission products is also assumed to be predominantly elemental iodine. These assumptions have strongly affected present nuclear air cleaning requirements by emphasizing rapid actuation of spray systems and filtration systems optimized to retain elemental iodine. A proposed revision of reactor accident source terms and some im implications for nuclear air cleaning requirements was presented at the 22nd DOE/NRC Nuclear Air Cleaning Conference. A draft report was issued by the NRC for comment in July 1992. Extensive comments were received, with the most significant comments involving (a) release fractions for both volatile and non-volatile species in the early in-vessel release phase, (b) gap release fractions of the noble gases, iodine and cesium, and (c) the timing and duration for the release phases. The final source term report is expected to be issued in late 1994. Although the revised source terms are intended primarily for future plants, current nuclear power plants may request use of revised accident source term insights as well in licensing. This paper emphasizes additional information obtained since the 22nd Conference, including studies on fission product removal mechanisms, results obtained from improved severe accident code calculations and resolution of major comments, and their impact upon the revised accident source terms. Revised accident source terms for both BWRS and PWRS are presented.

  17. Transportation Data Archiving

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

    Transportation Data Archiving This email address is being protected from spambots. You need JavaScript enabled to view it. - TRACC Director Background Urban and regional transportation planning and operations applications, (e.g. traffic modeling) require a large volume of accurate traffic-related data for a wide range of conditions. Significant real-time data on traffic volumes, highway construction, accidents, weather, airline flights, commuter and rail schedules, etc., are recorded each day by

  18. Source terms for plutonium aerosolization from nuclear weapon accidents

    SciTech Connect (OSTI)

    Stephens, D.R.

    1995-07-01

    The source term literature was reviewed to estimate aerosolized and respirable release fractions for accidents involving plutonium in high-explosive (HE) detonation and in fuel fires. For HE detonation, all estimates are based on the total amount of Pu. For fuel fires, all estimates are based on the amount of Pu oxidized. I based my estimates for HE detonation primarily upon the results from the Roller Coaster experiment. For hydrocarbon fuel fire oxidation of plutonium, I based lower bound values on laboratory experiments which represent accident scenarios with very little turbulence and updraft of a fire. Expected values for aerosolization were obtained from the Vixen A field tests, which represent a realistic case for modest turbulence and updraft, and for respirable fractions from some laboratory experiments involving large samples of Pu. Upper bound estimates for credible accidents are based on experiments involving combustion of molten plutonium droplets. In May of 1991 the DOE Pilot Safety Study Program established a group of experts to estimate the fractions of plutonium which would be aerosolized and respirable for certain nuclear weapon accident scenarios.

  19. 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.

  20. 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.

  1. 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...

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

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

    Type B Accident Investigation on the February 17, 2004, Personal Injury Accident, Bettis Atomic Power Laboratory Type B Accident Investigation of the Arc Flash at Brookhaven ...

  3. Type B Accident Investigation on the February 17, 2004, Personal...

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

    on the February 17, 2004, Personal Injury Accident, Bettis Atomic Power Laboratory Type B Accident Investigation on the February 17, 2004, Personal Injury Accident, Bettis Atomic ...

  4. Naval Spent Fuel Rail Shipment Accident Exercise Objectives ...

    Office of Environmental Management (EM)

    Naval Spent Fuel Rail Shipment Accident Exercise Objectives Naval Spent Fuel Rail Shipment Accident Exercise Objectives PDF icon Naval Spent Fuel Rail Shipment Accident Exercise ...

  5. 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)

  6. ORISE: REAC/TS Radiation Accident Registries

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

    Accident Registries The Radiation Emergency Assistance Center/Training Site (REAC/TS) at the Oak Ridge Institute for Science and Education (ORISE) maintains a number of radiation accident registries that provide medical professionals with up-to-date radiation accident information. Information for these accident registries is gathered from many sources, including the World Health Organization, International Atomic Energy Agency, U.S. Nuclear Regulatory Commission, state radiological health

  7. Type B Accident Investigation, Subcontractor Employee Personal...

    Office of Environmental Management (EM)

    Investigation, Subcontractor Employee Personal Protective Equipment Ignition Incident on ... Type B Accident Investigation, Subcontractor Employee Personal Protective Equipment ...

  8. 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.

  9. Trends in state-level freight accident rates: An enhancement of risk factor development for RADTRAN

    SciTech Connect (OSTI)

    Saricks, C.; Kvitek, T.

    1991-01-01

    Under the Nuclear Waste Policy Act, the Department of Energy's Office of Civilian Radioactive Waste Management (OCRWM) is concerned with understanding and managing risk as it applies to the shipment of spent commercial nuclear reactor fuel. Understanding risk in relation to mode and geography may provide opportunities to minimize radiological and non-radiological risks of transportation. To enhance such an understanding, a set of state-or waterway-specific accident, fatality, and injury rates (expressed as rates per shipment kilometer) by transportation mode and highway administrative class was developed, using publicly-available data bases. Adjustments made to accommodate miscoded or incomplete information in accident data are described, as well as the procedures for estimating state-level flow data. Results indicate that the shipping conditions under which spent fuel is likely to be transported should be less subject to accidents than the average'' shipment within mode. 10 refs., 3 tabs.

  10. 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).

  11. 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.

  12. Community emergency response to nuclear power plant accidents: A selected and partially annotated bibliography

    SciTech Connect (OSTI)

    Youngen, G.

    1988-10-01

    The role of responding to emergencies at nuclear power plants is often considered the responsibility of the personnel onsite. This is true for most, if not all, of the incidents that may happen during the course of the plant`s operating lifetime. There is however, the possibility of a major accident occurring at anytime. Major nuclear accidents at Chernobyl and Three Mile Island have taught their respective countries and communities a significant lesson in local emergency preparedness and response. Through these accidents, the rest of the world can also learn a great deal about planning, preparing and responding to the emergencies unique to nuclear power. This bibliography contains books, journal articles, conference papers and government reports on emergency response to nuclear power plant accidents. It does not contain citations for ``onsite`` response or planning, nor does it cover the areas of radiation releases from transportation accidents. The compiler has attempted to bring together a sampling of the world`s collective written experience on dealing with nuclear reactor accidents on the sate, local and community levels. Since the accidents at Three Mile Island and Chernobyl, that written experience has grown enormously.

  13. Accident analysis of heavy water cooled thorium breeder reactor

    SciTech Connect (OSTI)

    Yulianti, Yanti; Su’ud, Zaki; Takaki, Naoyuki

    2015-04-16

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  14. Calculation notes that support accident scenario and consequence development for the subsurface leak remaining subsurface accident

    SciTech Connect (OSTI)

    Ryan, G.W., Westinghouse Hanford

    1996-07-12

    This document supports the development and presentation of the following accident scenario in the TWRS Final Safety Analysis Report: Subsurface Leak Remaining Subsurface. The calculations needed to quantify the risk associated with this accident scenario are included within.

  15. Calculation notes that support accident scenario and consequence development for the subsurface leak remaining subsurface accident

    SciTech Connect (OSTI)

    Ryan, G.W., Westinghouse Hanford

    1996-09-19

    This document supports the development and presentation of the following accident scenario in the TWRS Final Safety Analysis Report: Subsurface Leak Remaining Subsurface. The calculations needed to quantify the risk associated with this accident scenario are included within.

  16. ORISE: REAC/TS Radiation Accident Registries

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

    Accident Registries The Radiation Emergency Assistance CenterTraining Site (REACTS) at the Oak Ridge Institute for Science and Education (ORISE) maintains a number of radiation ...

  17. Recommendations for Analyzing Accidents Under NEPA

    Broader source: Energy.gov [DOE]

    This DOE guidance clarifies and supplements "Recommendations for the Preparation of Environmental Assessments and Environmental Impact Statements." It focuses on principles of accident analyses under NEPA.

  18. MELCOR Accident Consequence Code System (MACCS)

    SciTech Connect (OSTI)

    Chanin, D.I. ); Sprung, J.L.; Ritchie, L.T.; Jow, Hong-Nian )

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previous CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. This document, Volume 1, the Users's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems.

  19. Audit of the Department of Energy's Transportation Accident Resistant...

    Office of Environmental Management (EM)

    ... - Department and Department of Defense (DOD) technical manuals related to the ... and correspondence of the Department, DOD, and Air Force concerning the container program. ...

  20. 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.

  1. 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.

  2. 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

  3. SAF-230DE- Web Based Course: Accident Investigation Overview

    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.

  4. 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.

  5. 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.

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

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

    Processing Facility on October 6, 2009 Type B Accident Investigation Board Report of the ... This report documents the results of the Type B Accident Investigation Board (Board) ...

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

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

    Type B Accident Investigation Board Report on the October 15, 2001, Grout Injection ... Type B Accident Investigation Board Report on the October 15, 2001, Grout Injection ...

  8. Accident analysis of heavy water cooled thorium breeder reactor...

    Office of Scientific and Technical Information (OSTI)

    Accident analysis of heavy water cooled thorium breeder reactor Citation Details In-Document Search Title: Accident analysis of heavy water cooled thorium breeder reactor ...

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

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

    (BPA) Chief Safety Officer, a Level I Accident Investigation was convened to investigate an accident that resulted in the fatality of a Wilson Construction Company Crew Foreman. ...

  10. 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 ...

  11. Accident Tolerant Fuels for LWRs: A Perspective (Journal Article...

    Office of Scientific and Technical Information (OSTI)

    Journal Article: Accident Tolerant Fuels for LWRs: A Perspective Citation Details In-Document Search Title: Accident Tolerant Fuels for LWRs: A Perspective Authors: Zinkle, Steven ...

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

    Energy Savers [EERE]

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

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

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

    Upon arrival, an incoming radiological survey was performed. PDF icon Type B Accident ... Preliminary Notice of Violation, Bechtel Jacobs Company, LLC - EA-2005-04 Type B Accident ...

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

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

    Accident Investigation of the February 5, 2014, Underground Salt Haul Truck Fire at the Waste Isolation Pilot Plant, Carlsbad NM March 26, 2014 Accident Investigation of the ...

  15. 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 ...

  16. Los Alamos National Laboratory Accident Investigation Board Corrective...

    Office of Environmental Management (EM)

    Accident Investigation Board Corrective Action Plan Update Los Alamos National Laboratory Accident Investigation Board Corrective Action Plan Update Topic: Status of the Corrective ...

  17. Type B Accident Investigation of the July 14, 2005, Americium...

    Energy Savers [EERE]

    14, 2005, Americium Contamination Accident at the Sigma Facility, Los Alamos National Laboratory Type B Accident Investigation of the July 14, 2005, Americium Contamination ...

  18. Accident Investigation Report Phase II | Department of Energy

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

    On March 4, 2014, an Accident Investigation Board (the Board) was appointed by Matthew ... appointed an Accident Investigation Board to complete the investigation (Phase 2). ...

  19. Volume II - Accident and Operational Safety Analysis Handbook

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

    208-2012 July 2012 DOE HANDBOOK Accident and Operational Safety Analysis Volume II: ... This Department of Energy (DOE) Accident and Operational Safety Analysis Handbook ...

  20. 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 ...

  1. 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

  2. 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.

  3. Psychophysiological and other factors affecting human performance in accident prevention and investigation. [Comparison of aviation with other industries

    SciTech Connect (OSTI)

    Klinestiver, L.R.

    1980-01-01

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

  4. Accident Investigation of the 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, Idaho

    Broader source: Energy.gov [DOE]

    This report documents the Naval Reactors investigation into the collapse ofa partially-erected spent fuel storage building, Overpack Storage Expansion #2 (OSE2), at the Naval Reactors Facility. The Accident Investigation Board inspected the scene, collected physical and photographic evidence, interviewed involved personnel, and reviewed relevant documents to determine the key causes of the accident. Based on the information gathered during the investigation, the Board identified several engineering and safety deficiencies that need to be addressed to prevent recurrence.

  5. 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`.

  6. 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.

  7. 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.

  8. DOE Accident Prevention and Investigation Program | Department of Energy

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

    DOE Accident Prevention and Investigation Program DOE Accident Prevention and Investigation Program The Department of Energy (DOE) Accident Prevention and Investigation Program serves as a key DOE corporate safety resource for promoting accident PREVENTION through exchange of lessons learned and information 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 indicators

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

    SciTech Connect (OSTI)

    Sdouz, Gert

    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

  10. Passive decay heat removal by natural air convection after severe accidents

    SciTech Connect (OSTI)

    Erbacher, F.J.; Neitzel, H.J.; Cheng, X.

    1995-09-01

    The composite containment proposed by the Research Center Karlsruhe and the Technical University Karlsruhe is to cope with severe accidents. It pursues the goal to restrict the consequences of core meltdown accidents to the reactor plant. One essential of this new containment concept is its potential to remove the decay heat by natural air convection and thermal radiation in a passive way. To investigate the coolability of such a passive cooling system and the physical phenomena involved, experimental investigations are carried out at the PASCO test facility. Additionally, numerical calculations are performed by using different codes. A satisfying agreement between experimental data and numerical results is obtained.

  11. Estimating vehicle roadside encroachment frequency using accident prediction models

    SciTech Connect (OSTI)

    Miaou, S.-P.

    1996-07-01

    The existing data to support the development of roadside encroachment- based accident models are extremely limited and largely outdated. Under the sponsorship of the Federal Highway Administration and Transportation Research Board, several roadside safety projects have attempted to address this issue by providing rather comprehensive data collection plans and conducting pilot data collection efforts. It is clear from the results of these studies that the required field data collection efforts will be expensive. Furthermore, the validity of any field collected encroachment data may be questionable because of the technical difficulty to distinguish intentional from unintentional encroachments. This paper proposes an alternative method for estimating the basic roadside encroachment data without actually field collecting them. The method is developed by exploring the probabilistic relationships between a roadside encroachment event and a run-off-the-road event With some mild assumptions, the method is capable of providing a wide range of basic encroachment data from conventional accident prediction models. To illustrate the concept and use of such a method, some basic encroachment data are estimated for rural two-lane undivided roads. In addition, the estimated encroachment data are compared with the existing collected data. The illustration shows that the method described in this paper can be a viable approach to estimating basic encroachment data without actually collecting them which can be very costly.

  12. Code System To Analyze Radiological Impact From Radwaste Transportation.

    Energy Science and Technology Software Center (OSTI)

    1988-05-01

    Version 00 RADSHIP-2 is a computer code system used to analyze the environmental impact of radwaste transportation in Taiwan. The specific transport scheme including the land transport by truck and sea transport by ship or barge were considered in the analysis for normal transport and transport accident conditions. The code combines meteorological, population, health physics, transportation, packaging and material factors and has the capability to obtain the results of the expected annual population radiation exposure,more » the expected number of annual latent cancer fatalities and the annual probability of a given number of early fatalities.« less

  13. Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

    SciTech Connect (OSTI)

    Farmer, Mitchell T.; Bunt, R.; Corradini, M.; Ellison, Paul B.; Francis, M.; Gabor, John D.; Gauntt, R.; Henry, C.; Linthicum, R.; Luangdilok, W.; Lutz, R.; Paik, C.; Plys, M.; Rabiti, Cristian; Rempe, J.; Robb, K.; Wachowiak, R.

    2015-01-31

    The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy’s (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).

  14. Probative Investigation of the Thermal Stability of Wastes Involved in

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

    February 2014 Waste Isolation Pilot Plant (WIPP) Waste Drum Breach Event | Department of Energy Probative Investigation of the Thermal Stability of Wastes Involved in February 2014 Waste Isolation Pilot Plant (WIPP) Waste Drum Breach Event Probative Investigation of the Thermal Stability of Wastes Involved in February 2014 Waste Isolation Pilot Plant (WIPP) Waste Drum Breach Event This document was used to determine facts and conditions during the Department of Energy Accident Investigation

  15. 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 2000C 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.

  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. Type A Accident Report of the June 26, 2009 Vehicle Fatality at Lawrence

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

    Livermore National Laboratory | Department of Energy Report of the June 26, 2009 Vehicle Fatality at Lawrence Livermore National Laboratory Type A Accident Report of the June 26, 2009 Vehicle Fatality at Lawrence Livermore National Laboratory October 1, 2009 On June 26, 2009, a Lawrence Livermore National Security (LLNS) employee was in the process of transporting six boxes containing personal property to his new office in preparation for a routine transfer to another position within the

  18. 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.

  19. Type B Accident Investigation Board Report of the April 23, 1997...

    Office of Environmental Management (EM)

    April 23, 1997, Helicopter Accident at Raton Pass, Raton Pass, Colorado Type B Accident Investigation Board Report of the April 23, 1997, Helicopter Accident at Raton Pass, Raton ...

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

    Energy Savers [EERE]

    Los Alamos National Laboratory Type B Accident Investigation on the August 5, 2003, ... The Accident Investigation Board concluded that the direct cause of the accident was the ...

  1. 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.

  2. 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.

  3. Nuclear Facilities Fire Accident Model

    Energy Science and Technology Software Center (OSTI)

    1999-09-01

    4. NATURE OF PROBLEM SOLVED FIRAC predicts fire-induced flows, thermal and material transport, and radioactive and nonradioactive source terms in a ventilation system. It is designed to predict the radioactive and nonradioactive source terms that lead to gas dynamic, material transport, and heat transfer transients. FIRAC's capabilities are directed toward nuclear fuel cycle facilities and the primary release pathway, the ventilation system. However, it is applicable to other facilities and can be used to modelmore » other airflow pathways within a structure. The basic material transport capability of FIRAC includes estimates of entrainment, convection, deposition, and filtration of material. The interrelated effects of filter plugging, heat transfer, and gas dynamics are also simulated. A ventilation system model includes elements such as filters, dampers, ducts, and blowers connected at nodal points to form networks. A zone-type compartment fire model is incorporated to simulate fire-induced transients within a facility. 5. METHOD OF SOLUTION FIRAC solves one-dimensional, lumped-parameter, compressible flow equations by an implicit numerical scheme. The lumped-parameter method is the basic formulation that describes the gas dynamics system. No spatial distribution of parameters is considered in this approach, but an effect of spatial distribution can be approximated by noding. Network theory, using the lumped parameter method, includes a number of system elements, called branches, joined at certain points, called nodes. Ventilation system components that exhibit flow resistance and inertia, such as dampers, ducts, valves, and filters, and those that exhibit flow potential, such as blowers, are located within the branches of the system. The connection points of branches are nodes for components that have finite volumes, such as rooms, gloveboxes, and plenums, and for boundaries where the volume is practically infinite. All internal nodes, therefore, possess some

  4. 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. ); Medford, G.T. )

    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.

  5. 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.

  6. LESSONS LEARNED FROM A RECENT LASER ACCIDENT

    SciTech Connect (OSTI)

    Woods, Michael; /SLAC

    2011-01-26

    A graduate student received a laser eye injury from a femtosecond Ti:sapphire laser beam while adjusting a polarizing beam splitter optic. The direct causes for the accident included failure to follow safe alignment practices and failure to wear the required laser eyewear protection. Underlying root causes included inadequate on-the-job training and supervision, inadequate adherence to requirements, and inadequate appreciation for dimly visible beams outside the range of 400-700nm. This paper describes how the accident occurred, discusses causes and lessons learned, and describes corrective actions being taken.

  7. Accident consequence calculations for project W-058 safetyanalysis

    SciTech Connect (OSTI)

    Van Keuren, J.C.

    1997-06-10

    Accident consequence analyses have been performed for Project W-058, the Replacement Cross Site Transfer System. using the assumption and analysis techniques developed for the Tank Remediation Waste system Basis for Interim Operation. most potential accident involving the FISTS are bounded by the TWRS BIO analysis. However, the spray leak and pool leak scenarios require revised analyses since the RCSTS design utilizes larger diameter pipe and higher pressures than those analyzed in the TWRS BIO. Also the volume of diversion box and vent station are larger than that assumed for the valve pits in the TWRS BIO, which effects results of sprays or spills into the pits. the revised analysis for the spray leak is presented in Section 2, for the above ground spill in Section 3, for the presented in Section 2, for the above ground spill in Section 3, for the subsurface spill forming a pool in Section 4, and for the subsurface pool remaining subsurface in Section 5. The conclusion from these sections are summarized below.

  8. 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.

  9. Transportation Research

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

    transportation-research TRACC RESEARCH Computational Fluid Dynamics Computational Structural Mechanics Transportation Systems Modeling Transportation Research Current Research Overview The U.S. Department of Transportation (USDOT) has established its only high-performance computing and engineering analysis research facility at Argonne National Laboratory to provide applications support in key areas of applied research and development for the USDOT community. The Transportation Research and

  10. Commitment to Public Involvement

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

    Create a Sustainable Future Commitment to Public Involvement Commitment to Public Involvement LANL is committed to our neighbors August 1, 2013 Lab Director McMillan talks with...

  11. 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.

  12. Accident Investigation Board (AIB) findings about the drum breach...

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

    Accident Investigation Board findings Accident Investigation Board (AIB) findings about the drum breach at WIPP WHEN: Apr 23, 2015 5:30 PM - 7:00 PM WHERE: Fuller Lodge 2132 ...

  13. DOE - NNSA/NFO -- News & Views Accident Trap

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

    Accident Traps Four Workers 1,800 Feet below Ground Photo - rescue from 1,800 feet below ... Thousands of workers have completed millions of accident-free hours at this heavy industry ...

  14. Y-12's 1958 nuclear criticality accident and increased safety...

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

    accident and increased safety - 1958 brought accidents, more safety The first X-ray machine was brought to Y-12 in February, 1949. It was a 1,000 KV system installed in Building...

  15. ANS severe accident program overview & planning document

    SciTech Connect (OSTI)

    Taleyarkhan, R.P.

    1995-09-01

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

  16. 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.

  17. Computerized Accident Incident Reporting System | Department of Energy

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

    Computerized Accident Incident Reporting System Computerized Accident Incident Reporting System CAIRS Database The Computerized Accident/Incident Reporting System is a database used to collect and analyze DOE and DOE contractor reports of injuries, illnesses, and other accidents that occur during DOE operations. CAIRS is a Government computer system and, as such, has security requirements that must be followed. Access to the database is open to DOE and DOE contractors. Additional information

  18. Radiation Transport

    SciTech Connect (OSTI)

    Urbatsch, Todd James

    2015-06-15

    We present an overview of radiation transport, covering terminology, blackbody raditation, opacities, Boltzmann transport theory, approximations to the transport equation. Next we introduce several transport methods. We present a section on Caseology, observing transport boundary layers. We briefly broach topics of software development, including verification and validation, and we close with a section on high energy-density experiments that highlight and support radiation transport.

  19. Severe Accident Test Station Design Document

    SciTech Connect (OSTI)

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

    2015-09-01

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

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

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

    Phase II Accident Investigation Board Briefing Phase II Accident Investigation Board Briefing Topic: Ted Wyka DOE, Provided a Brief on the Findings in the WIPP Accident Investigation. Information Provided included the Judgments of NEED and Causes That Contributed to the Incident. AIB Brief - May 20, 2015 (1.98

  1. Calculation notes that support accident scenario and consequence development for the steam intrusion from interfacing systems accident

    SciTech Connect (OSTI)

    Ryan, G.W., Westinghouse Hanford

    1996-07-25

    This document supports the development and presentation of the following accident scenario in the TWRS Final Safety Analysis Report: Steam Intrusion from Interfacing Systems. The calculations needed to quantify the risk associated with this accident scenario are included within.

  2. 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.

  3. Chamber transport

    SciTech Connect (OSTI)

    OLSON,CRAIG L.

    2000-05-17

    Heavy ion beam transport through the containment chamber plays a crucial role in all heavy ion fusion (HIF) scenarios. Here, several parameters are used to characterize the operating space for HIF beams; transport modes are assessed in relation to evolving target/accelerator requirements; results of recent relevant experiments and simulations of HIF transport are summarized; and relevant instabilities are reviewed. All transport options still exist, including (1) vacuum ballistic transport, (2) neutralized ballistic transport, and (3) channel-like transport. Presently, the European HIF program favors vacuum ballistic transport, while the US HIF program favors neutralized ballistic transport with channel-like transport as an alternate approach. Further transport research is needed to clearly guide selection of the most attractive, integrated HIF system.

  4. 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.

  5. 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

  6. A probabilistic risk assessment of the LLNL Plutonium facility`s evaluation basis fire operational accident

    SciTech Connect (OSTI)

    Brumburgh, G.

    1994-08-31

    The Lawrence Livermore National Laboratory (LLNL) Plutonium Facility conducts numerous involving plutonium to include device fabrication, development of fabrication techniques, metallurgy research, and laser isotope separation. A Safety Analysis Report (SAR) for the building 332 Plutonium Facility was completed rational safety and acceptable risk to employees, the public, government property, and the environment. This paper outlines the PRA analysis of the Evaluation Basis Fire (EDF) operational accident. The EBF postulates the worst-case programmatic impact event for the Plutonium Facility.

  7. 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.

  8. 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.

  9. 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.

  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.

    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. Development of Onsite Transportation Safety Documents for Nevada Test Site

    SciTech Connect (OSTI)

    Frank Hand, Willard Thomas, Frank Sciacca, Manny Negrete, Susan Kelley

    2008-05-08

    Department of Energy (DOE) Orders require each DOE site to develop onsite transportation safety documents (OTSDs). The Nevada Test Site approach divided all onsite transfers into two groups with each group covered by a standalone OTSD identified as Non-Nuclear and Nuclear. The Non-Nuclear transfers involve all radioactive hazardous material in less than Hazard Category (HC)-3 quantities and all chemically hazardous materials. The Nuclear transfers involve all radioactive material equal to or greater than HC-3 quantities and radioactive material mated with high explosives regardless of quantity. Both OTSDs comply with DOE O 460.1B requirements. The Nuclear OTSD also complies with DOE O 461.1A requirements and includes a DOE-STD-3009 approach to hazard analysis (HA) and accident analysis as needed. All Nuclear OTSD proposed transfers were determined to be non-equivalent and a methodology was developed to determine if equivalent safety to a fully compliant Department of Transportation (DOT) transfer was achieved. For each HA scenario, three hypothetical transfers were evaluated: a DOT-compliant, uncontrolled, and controlled transfer. Equivalent safety is demonstrated when the risk level for each controlled transfer is equal to or less than the corresponding DOT-compliant transfer risk level. In this comparison the typical DOE-STD-3009 risk matrix was modified to reflect transportation requirements. Design basis conditions (DBCs) were developed for each non-equivalent transfer. Initial DBCs were based solely upon the amount of material present. Route-, transfer-, and site-specific conditions were evaluated and the initial DBCs revised as needed. Final DBCs were evaluated for each transfers packaging and its contents.

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

    SciTech Connect (OSTI)

    Morris, Robert Noel

    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

  13. Severe Accident Scoping Simulations of Accident Tolerant Fuel Concepts for BWRs

    SciTech Connect (OSTI)

    Robb, Kevin R.

    2015-08-01

    Accident-tolerant fuels (ATFs) are fuels and/or cladding that, in comparison with the standard uranium dioxide Zircaloy system, can tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operations [1]. It is important to note that the currently used uranium dioxide Zircaloy fuel system tolerates design basis accidents (and anticipated operational occurrences and normal operation) as prescribed by the US Nuclear Regulatory Commission. Previously, preliminary simulations of the plant response have been performed under a range of accident scenarios using various ATF cladding concepts and fully ceramic microencapsulated fuel. Design basis loss of coolant accidents (LOCAs) and station blackout (SBO) severe accidents were analyzed at Oak Ridge National Laboratory (ORNL) for boiling water reactors (BWRs) [2]. Researchers have investigated the effects of thermal conductivity on design basis accidents [3], investigated silicon carbide (SiC) cladding [4], as well as the effects of ATF concepts on the late stage accident progression [5]. These preliminary analyses were performed to provide initial insight into the possible improvements that ATF concepts could provide and to identify issues with respect to modeling ATF concepts. More recently, preliminary analyses for a range of ATF concepts have been evaluated internationally for LOCA and severe accident scenarios for the Chinese CPR1000 [6] and the South Korean OPR-1000 [7] pressurized water reactors (PWRs). In addition to these scoping studies, a common methodology and set of performance metrics were developed to compare and support prioritizing ATF concepts [8]. A proposed ATF concept is based on iron-chromium-aluminum alloys (FeCrAl) [9]. With respect to enhancing accident tolerance, FeCrAl alloys have substantially slower oxidation kinetics compared to the zirconium alloys typically employed. During a severe accident, Fe

  14. Fallout: The experiences of a medical team in the care of a Marshallese population accidently exposed to fallout radiation

    SciTech Connect (OSTI)

    Conard, R.A.

    1992-09-01

    This report presents an historical account of the experiences of the Brookhaven Medical Team in the examination and treatment of the Marshallese people following their accidental exposure to radioactive fallout in 1954. This is the first time that a population has been heavily exposed to radioactive fallout, and even though this was a tragic mishap, the medical findings have provided valuable information for other accidents involving fallout such as the recent reactor accident at Chernobyl. Noteworthy has been the unexpected importance of radioactive iodine in the fallout in producing thyroid abnormalities.

  15. 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.

  16. Beam Transport

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

    Beam Transport A simplified drawing of the beam transport system from the linac to Target-1 (Lujan Center), Target-2 (Blue Room) and Target-4 is shown below. In usual operation ...

  17. 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.

  18. WIPP Transportation

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

    Transuranic Waste Transportation Container Documents Documents related to transuranic waste containers and packages. CBFO Tribal Program Information about WIPP shipments across tribal lands. Transportation Centralized Procurement Program - The Centralized Procurement Program provides a common method to procure standard items used in the packaging and handling of transuranic wasted destined for WIPP. Transuranic Waste Transportation Routes - A map showing transuranic waste generator sites and

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

    SciTech Connect (OSTI)

    Hammersley, R.J.

    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.

  20. A review of post-accident mitigative measures affecting transport and isolation of radionuclides released from the Chernobyl accident

    SciTech Connect (OSTI)

    Waters, R.; Gibson, D.; Bugai, D.; Shalsky, A.; Dgepo, S.; Voitsekhovitch, O.

    1994-09-01

    This paper summarizes the results of eight years of mitigative measures to radioactive contamination within the 30 kilometer exclusion zone surrounding the Chernobyl Nuclear Power Plant. We hope to demonstrate that effectiveness of mitigative measures depends not only on proper application of technology but also on selection of projects offering significant risk reduction potential. In a limited national economy, environmental mitigation projects must maximize risk reduction and cost effectiveness or risk losing funding to more pressing social issues.

  1. Oregon Department of Transportation | Open Energy Information

    Open Energy Info (EERE)

    services; transportation safety programs; driver and vehicle licensing; and motor carrier regulation. ODOT is actively involved in developing Oregon's system of...

  2. Federally Led Accident Investigation Reports | Department of Energy

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

    Federally Led Accident Investigation Reports Federally Led Accident Investigation Reports Includes Pre-March 2011 Type A Reports April 21, 2015 Accident Investigations of the February 14, 2014, Radiological Release at the Waste Isolation Pilot Plant, Carlsbad, NM On Friday, February 14, 2014, there was a release of radioactive material in the underground repository at the Department of Energy (DOE) Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico. Following initial visual inspections

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

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

    Management (Published by REAC/TS) The Medical Basis for Radiation-Accident Preparedness: Medical Management Proceedings of the Fifth International REAC/TS Symposium on the Medical Basis for Radiation-Accident Preparedness and the Biodosimetry Workshop As part of its mission to provide continuing education for personnel responsible for treating radiation injuries, REAC/TS hosted the Fifth International REAC/TS Symposium on the Medical Basis for Radiation-Accident Preparedness symposium and

  4. Recommendations for Analyzing Accidents Under NEPA (DOE, 2002)

    Broader source: Energy.gov [DOE]

    This DOE guidance clarifies and supplements "Recommendations for the Preparation of Environmental Assessments and Environmental Impact Statements." It focuses on principles of accident analyses under NEPA.

  5. Accident Investigation of the December 11, 2013, Integrated Device...

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

    Sandia National Laboratory Site 9920, Albuquerque, NM Accident Investigation of the December 11, 2013, Integrated Device Fireset and Detonator Accidental Discharge at the Sandia ...

  6. Code System to Model LWR Meltdown Accident Response.

    Energy Science and Technology Software Center (OSTI)

    2001-04-25

    MARCH2 describes the response of water cooled reactors to severe accidents, including consideration of the primary coolant system as well as the containment.

  7. Sandia Energy - Waste Isolation Pilot Plant Accident Investigation...

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

    Home Energy Nuclear Energy News News & Events Research & Capabilities Systems Analysis Materials Science Computational Modeling & Simulation Waste Isolation Pilot Plant Accident...

  8. Neutronic Analysis of Candidate Accident-tolerant Cladding Concepts...

    Office of Scientific and Technical Information (OSTI)

    Concepts in Light Water Reactors Citation Details In-Document Search Title: Neutronic Analysis of Candidate Accident-tolerant Cladding Concepts in Light Water Reactors Authors: ...

  9. Neutronic Analysis of Candidate Accident-Tolerant Cladding Concepts...

    Office of Scientific and Technical Information (OSTI)

    in Pressurized Water Reactors Citation Details In-Document Search Title: Neutronic Analysis of Candidate Accident-Tolerant Cladding Concepts in Pressurized Water Reactors ...

  10. 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...

  11. Accident Investigation Reports - Type B | Department of Energy

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

    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). October...

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

    Energy Savers [EERE]

    IMPROVEMENT OF DESIGN CODES TO ACCOUNT FOR ACCIDENT THERMAL EFFECTS ON SEISMIC PERFORMANCE ... PROJECT OBJECTIVES (CONT'D) Develop design guidelines and recommendations for ...

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

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

    Independent Oversight Follow-up Review, Savannah River National Laboratory - January 2012 Type B Accident Investigation of the Arc Flash at Brookhaven National Laboratory, April ...

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

    Energy Savers [EERE]

    Arc Flash at Brookhaven National Laboratory, April 14, 2006 Type B Accident Investigation of the Arc Flash at Brookhaven National Laboratory, April 14, 2006 February 10, 2006 An ...

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

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

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

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

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

    PDF icon Accident Investigation of the August 21, 2012, Contamination Incident at the Los Alamos Neutron Science Center at the Los Alamos National Laboratory More Documents & ...

  17. 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...

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

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

    barrier analysis, change analysis, and event and causal factor analysis. PDF icon Type B Accident Investigation Board Report for the January 11, 2006, Personal Injury During ...

  19. Accident Investigation Reports - Type B | Department of Energy

    Office of Environmental Management (EM)

    September 1, 2010 Type B Accident Investigation Board Report Employee Puncture Wound at ... investigation of the June 14, 2010, employee puncture wound at the Department of ...

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

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

    Employee Foot Injury on December 17, 2003, at the East Tennessee Technology Park Building K-31 Type B Accident Investigation Board Report BNFL, Inc. Employee Foot Injury on ...

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

    Energy Savers [EERE]

    LLC, Employee Fall Injury on July 1, 2009, At The 336 Building, Hanford Site, Washington Type B Accident Investigation At Washington Closure Hanford, LLC, Employee Fall Injury ...

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

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

    Company Employee Fatality near Patrick's Knob Radio Station, Bonneville Power Administration Accident Investigation of the October 1, 2013, Tice Electric Company Employee Fatality ...

  3. Type B Accident Investigation Board Report Employee Puncture...

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

    investigation of the June 14, 2010, employee puncture wound at the Department of ... TYPE B ACCIDENT INVESTIGATION BOARD REPORT EMPLOYEE PUNCTURE WOUND AT THE F-TRU WASTE ...

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

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

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

  5. Accident Investigation of the June 1, 2013, Stairway Fall Resulting...

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

    in a Federal Employee Fatality at DOE Headquarters Germantown, Maryland Accident Investigation of the June 1, 2013, Stairway Fall Resulting in a Federal Employee Fatality at DOE ...

  6. Type B Accident Investigation of the August 22, 2000, Injury...

    Office of Environmental Management (EM)

    Chemical Reaction at the Portsmouth Gaseous Diffusion Plant, X-701B Site Type B Accident Investigation of the August 22, 2000, Injury Resulting From Violent Exothermic Chemical ...

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

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

    Report, May 8, 2004, Exothermic Metal Reactor Event During Sodium Transfer Activities, ... Type B Accident Investigation Board Report, May 8, 2004, Exothermic Metal Reactor Event ...

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

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

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

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

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

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

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

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

    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 ...

  11. 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.

  12. 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)

  13. Bounding Radionuclide Inventory and Accident Consequence Calculation for the 1L Target

    SciTech Connect (OSTI)

    Kelsey, Charles T. IV

    2011-01-01

    A bounding radionuclide inventory for the tungsten of the Los Alamos Neutron Science Center (LANSCE) IL Target is calculated. Based on the bounding inventory, the dose resulting from the maximum credible incident (MCI) is calculated for the maximally exposed offsite individual (MEOl). The design basis accident involves tungsten target oxidation following a loss of cooling accident. Also calculated for the bounding radionuclide inventory is the ratio to the LANSCE inventory threshold for purposes of inventory control as described in the target inventory control policy. A bounding radionuclide inventory calculation for the lL Target was completed using the MCNPX and CINDER'90 codes. Continuous beam delivery at 200 {micro}A to 2500 mA{center_dot}h was assumed. The total calculated activity following this irradiation period is 205,000 Ci. The dose to the MEOI from the MCI is 213 mrem for the bounding inventory. The LANSCE inventory control threshold ratio is 132.

  14. Accident Investigation at the Idaho National Laboratory Engineering Demonstration Facility, February 2013

    Office of Energy Efficiency and Renewable Energy (EERE)

    On Monday, February 12, 2013, a principal investigator at the Idaho National Laboratory (INL) Engineering Demonstration Facility (IEDF) was testing the system configuration of experimental process involving liquid sodium carbonate. An unanticipated event occurred that resulted in the ejection of the 900° C liquid sodium carbonate from the system. The ejected liquid came into contact with the principal investigator and caused multiple second and third degree burn injuries to approximately 10 percent of his body. The Office of Health, Safety and Security (HSS) Site Lead for the Idaho Site shadowed the accident investigation team assembled by the contractor in an effort to independently verify that a rigorous, thorough, and unbiased investigation was taking place, and to maintain awareness of the events surrounding the accident

  15. 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

  16. Type A Accident Investigation of the July 15, 2004, Hanford 200 East Area

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

    Fall Fatality | Department of Energy July 15, 2004, Hanford 200 East Area Fall Fatality Type A Accident Investigation of the July 15, 2004, Hanford 200 East Area Fall Fatality August 1, 2004 On the morning of Thursday, July 15, 2004, a non-government contractor employee of All Mobile Transporting & Repairs (AMTR) was found motionless at the bottom of a ladder with a serious head injury. It was believed the employee had been standing on the ladder, using a battery-powered drill, to remove

  17. 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.

  18. 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.

  19. Type B Accident Investigation of the March 20, 2003, Stair Installation Accident at Building 752, Sandia National Laboratories

    Broader source: Energy.gov [DOE]

    This report is an independent product of the Type B Accident Investigation Board appointed by Karen L. Boardman, Manager, Sandia Site Office (SSO), National Nuclear Security Administration (NNSA).

  20. Calculation notes that support accident scenario and consequence development for the steam intrusion from interfacing systems accident

    SciTech Connect (OSTI)

    Van Vleet, R.J.; Ryan, G.W.; Crowe, R.D.; Lindberg, S.E., Fluor Daniel Hanford

    1997-03-04

    This document supports the development and presentation of the following accident scenario in the TWRS Final Safety Analysis Report (FSAR): Steam Intrusion From Interfacing Systems. The calculations needed to quantify the risk associated with this accident scenario are included in the following sections to aid in the understanding of this accident scenario. Information validation forms citing assumptions that were approved for use specifically in this analysis are included in Appendix A. Copies of these forms are also on file with TWRS Project Files. Calculations performed in this document, in general, are expressed in traditional (English) units to aid understanding of the accident scenario and related parameters.

  1. Type B Accident Investigation Board Report of the April 23, 1997, Helicopter Accident at Raton Pass, Raton Pass, Colorado

    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.

  2. 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.

  3. Greening Transportation

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

    Transportation Goal 2: Greening Transportation LANL supports and encourages employees to reduce their personal greenhouse gas emissions by offering various commuting and work schedule options. Our goal is to reduce emissions related to employee travel and commuting to and from work by 13 percent. Energy Conservation» Efficient Water Use & Management» High Performance Sustainable Buildings» Greening Transportation» Green Purchasing & Green Technology» Pollution Prevention» Science

  4. 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.

  5. 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.

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

    SciTech Connect (OSTI)

    CROWE, R.D.

    1999-09-09

    This document provides 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.

  7. 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.

  8. 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.

  9. Canister storage building design basis accident analysis documentation

    SciTech Connect (OSTI)

    KOPELIC, S.D.

    1999-02-25

    This document provides 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.

  10. Accident response group (ARG) containers for recovery of damaged warheads

    SciTech Connect (OSTI)

    York, A.R. II; Hoffman, J.P.

    1993-09-01

    This report provides an overview of the containers that are currently stored at Pantex and available for use in response to an accident or for use in any other application where a sealed containment vessel and accident resistant overpack may be needed.

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

    Office of Energy Efficiency and Renewable Energy (EERE)

    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.

  12. GPHS-RTG launch accident analysis for Galileo and Ulysses

    SciTech Connect (OSTI)

    Bradshaw, C.T. )

    1991-01-01

    This paper presents the safety program conducted to determine the response of the General Purpose Heat Source (GPHS) Radioisotope Thermoelectric Generator (RTG) to potential launch accidents of the Space Shuttle for the Galileo and Ulysses missions. The National Aeronautics and Space Administration (NASA) provided definition of the Shuttle potential accidents and characterized the environments. The Launch Accident Scenario Evaluation Program (LASEP) was developed by GE to analyze the RTG response to these accidents. RTG detailed response to Solid Rocket Booster (SRB) fragment impacts, as well as to other types of impact, was obtained from an extensive series of hydrocode analyses. A comprehensive test program was conducted also to determine RTG response to the accident environments. The hydrocode response analyses coupled with the test data base provided the broad range response capability which was implemented in LASEP.

  13. 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.

  14. 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.

  15. 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.

  16. Transportation Energy

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

    Energy Home/Transportation Energy CRF_climatechange Permalink Gallery Understanding Hazardous Combustion Byproducts Reduces Factors Impacting Climate Change CRF, Global Climate & Energy, News, News & Events, Transportation Energy Understanding Hazardous Combustion Byproducts Reduces Factors Impacting Climate Change By Micheal Padilla Researchers at Sandia's Combustion Research Facility are developing the understanding necessary to build cleaner combustion technologies that will in turn

  17. Commitment to Public Involvement

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

    Create a Sustainable Future » Commitment to Public Involvement Commitment to Public Involvement LANL is committed to our neighbors August 1, 2013 Lab Director McMillan talks with Lab historian Ellen McGhee during Lab tour Lab Director McMillan talks with Lab historian Ellen McGhee during Lab tour RELATED IMAGES http://farm4.staticflickr.com/3815/9442115459_390e5bc841_t.jpg Enlarge http://farm8.staticflickr.com/7111/7650961650_ed1571174f_t.jpg Enlarge

  18. Reactor safety study. An assessment of accident risks in U. S...

    Office of Scientific and Technical Information (OSTI)

    An assessment of accident risks in U. S. commercial nuclear power plants. Executive ... An assessment of accident risks in U. S. commercial nuclear power plants. Executive ...

  19. Type A Accident Investigation Board Report on the April 19, 1999...

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

    Type A Accident Investigation Board Report on the April 19, 1999, Special Agent Fatality ... responsibility for conducting a Type A accident investigation to the AL Manager on April ...

  20. Type B Accident Investigation of the July 31, 2006, Fall from...

    Office of Environmental Management (EM)

    31, 2006, Fall from Ladder Accident at the Lawrence Livermore National Laboratory, Livermore, California Type B Accident Investigation of the July 31, 2006, Fall from Ladder ...

  1. Simulation of a beyond design-basis-accident with RELAP5/MOD3.1

    SciTech Connect (OSTI)

    Banati, J.

    1995-09-01

    This paper summarizes the results of analyses, parametric and sensitivity studies, performed using the RELAP5/MOD3.1 computer code for the 4th IAEA Standard Problem Exercise (SPE-4). The test, conducted on the PMK-2 facility in Budapest, involved simulation of a Small Break Loss Of Coolant Accident (SBLOCA) with a 7.4% break in the cold leg of a VVER-440 type pressurized water reactor. According to the scenario, the unavailability of the high pressure injection system led to a beyond design basis accident. For prevention of core damage, secondary side bleed-and-feed accident management measures were applied. A brief description of the PMK-2 integral type test facility is presented, together with the profile and some key phenomenological aspects of this particular experiment. Emphasis is placed on the ability of the code to predict the main trends observed in the test and thus, an assessment is given for the code capabilities to represent the system transient.

  2. 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.

  3. 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.

  4. Effects of spent fuel types on offsite consequences of hypothetical accidents

    SciTech Connect (OSTI)

    Courtney, J. C.; Dwight, C. C.; Lehto, M. A.

    2000-02-18

    Argonne National Laboratory (ANL) conducts experimental work on the development of waste forms suitable for several types of spent fuel at its facility on the Idaho National Engineering and Environmental Laboratory (INEEL) located 48 km West of Idaho Falls, ID. The objective of this paper is to compare the offsite radiological consequences of hypothetical accidents involving the various types of spent nuclear fuel handled in nonreactor nuclear facilities. The highest offsite total effective dose equivalents (TEDEs) are estimated at a receptor located about 5 km SSE of ANL facilities. Criticality safety considerations limit the amount of enriched uranium and plutonium that could be at risk in any given scenario. Heat generated by decay of fission products and actinides does not limit the masses of spent fuel within any given operation because the minimum time elapsed since fissions occurred in any form is at least five years. At cooling times of this magnitude, fewer than ten radionuclides account for 99% of the projected TEDE at offsite receptors for any credible accident. Elimination of all but the most important nuclides allows rapid assessments of offsite doses with little loss of accuracy. Since the ARF (airborne release fraction), RF (respirable fraction), LPF (leak path fraction) and atmospheric dilution factor ({chi}/Q) can vary by orders of magnitude, it is not productive to consider nuclides that contribute less than a few percent of the total dose. Therefore, only {sup 134}Cs, {sup 137}Cs-{sup 137m}Ba, and the actinides significantly influence the offsite radiological consequences of severe accidents. Even using highly conservative assumptions in estimating radiological consequences, they remain well below current Department of Energy guidelines for highly unlikely accidents.

  5. 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

  6. 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

  7. TRANSPORTATION OPTIONS

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

    TRANSPORTATION OPTIONS The Pittsburgh Airport Marriott provides complimentary shuttle service. The hotel asks all guests arriving at the Pittsburgh International Airport to collect luggage in the baggage claim area of the airport and then call for the shuttle at 412-788- 8800. Let the Hotel Operator know that you have collected your luggage and have a reservation at the Marriott and need transportation from the airport. The Hotel Operator will instruct the guest which door to exit, which curb to

  8. Transportation | NREL

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

    Transportation NREL's transportation infrastructure and programs are designed to significantly reduce petroleum use campus-wide. This infographic shows NREL's FY2015 fleet performance and fleet vehicle history compared to baseline FY 2005 and FY 2014. Petroleum fuel use decreased 28% from 2014 and increased 17% from baseline 2005. Alternative fuel use increased 53% from 2014 and increased 127% from baseline 2005. In baseline 2005, the fleet used 6,521 gasoline gallon equivalent (GGE) of E-85, in

  9. Involvement and Communication Committee.

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

    Public Involvement and Communication Committee Chair 1 : Liz Mattson 4 , Member Vice Chair 1 : Shannon Cram, Member Committee Members 2 : Shelley Cimon Paige Knight Alissa Cordner Liz Mattson Shannon Cram Ken Niles Rob Davis Ed Pacheco Tom Galioto Dan Serres Floyd Hodges Jean Vanni Unofficial Committee Members or Other Interested Parties 3 : Earl Fordham Gerry Pollet Susan Leckband 4 Steve Hudson 4 Facilitator: Cathy McCague 4 Agency & Technical Support: Kris Holmes 4 DOE-RL Randy Bradbury 4

  10. Material selection for accident tolerant fuel cladding

    SciTech Connect (OSTI)

    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 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 α’ formation. The composition effects and critical limits to retaining protective scale formation at >1400°C are still being evaluated.