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Note: This page contains sample records for the topic "accident conditions view" from the National Library of EnergyBeta (NLEBeta).
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1

Instrument Performance Under Severe Accident Conditions: Ways to Acquire Information From Instrumentation Affected by an Accident  

Science Conference Proceedings (OSTI)

Under accident conditions, information is needed for diagnosing plant status and confirming plant responses to mitigative actions. This makes it important to understand how instruments behave in severe accident environments and to find ways to obtain information from the instruments under conditions that can be more severe than their design bases.

1993-12-01T23:59:59.000Z

2

Accidents  

NLE Websites -- All DOE Office Websites (Extended Search)

Health Risks » Accidents Health Risks » Accidents DUF6 Health Risks line line Accidents Storage Conversion Manufacturing Disposal Transportation Accidents A discussion of accidents involving depleted UF6 storage cylinders, including possible health effects, accident risk, and accident history. Potential Health Effects from Cylinder Accidents Accidents involving depleted UF6 storage cylinders are a concern because they could result in an uncontrolled release of UF6 to the environment, which could potentially affect the health of workers and members of the public living downwind of the accident site. Accidental release of UF6 from storage cylinders or during processing activities could result in injuries or fatalities. The most immediate hazard after a release would be from inhalation of hydrogen fluoride (HF), a highly corrosive gas formed when

3

Hypothetical accident conditions thermal analysis of the 5320 package  

Science Conference Proceedings (OSTI)

An axisymmetric model of the 5320 package was created to perform hypothetical accident conditions (HAC) thermal calculations. The analyses assume the 5320 package contains 359 grams of plutonium-238 (203 Watts) in the form of an oxide powder at a minimum density of 2.4 g/cc or at a maximum density of 11.2 g/cc. The solution from a non-solar 100 F ambient steady-state analysis was used as the initial conditions for the fire transient. A 30 minute 1,475 F fire transient followed by cooling via natural convection and thermal radiation to a 100 F non-solar environment was analyzed to determine peak component temperatures and vessel pressures. The 5320 package was considered to be horizontally suspended within the fire during the entire transient.

Hensel, S.J.; Gromada, R.J.

1995-12-31T23:59:59.000Z

4

Hydrogen-control systems for severe LWR accident conditions - a state-of-technology report  

Science Conference Proceedings (OSTI)

This report reviews the current state of technology regarding hydrogen safety issues in light water reactor plants. Topics considered in this report relate to control systems and include combustion prevention, controlled combustion, minimization of combustion effects, combination of control concepts, and post-accident disposal. A companion report addresses hydrogen generation, distribution, and combustion. The objectives of the study were to identify the key safety issues related to hydrogen produced under severe accident conditions, to describe the state of technology for each issue, and to point out ongoing programs aimed at resolving the open issues.

Hilliard, R.K.; Postma, A.K.; Jeppson, D.W.

1983-03-01T23:59:59.000Z

5

PRESSURE INTEGRITY OF 3013 CONTAINER UNDER POSTULATED ACCIDENT CONDITIONS  

Science Conference Proceedings (OSTI)

A series of tests was carried out to determine the threshold for deflagration-to-detonation transition (DDT), structural loading, and structural response of the Department of Energy 3013 storage systems for the case of an accidental explosion of evolved gas within the storage containers. Three experimental fixtures were used to examine the various issues and three mixtures consisting of either stoichiometric hydrogen-oxygen, stoichiometric hydrogen-oxygen with added nitrogen, or stoichiometric hydrogen-oxygen with an added nitrogen-helium mixture were tested. Tests were carried out as a function of initial pressure from 1 to 3.5 bar and initial temperature from room temperature to 150 C. The elevated temperature tests resulted in a slight increase in the threshold pressure for DDT. The elevated temperature tests were performed to ensure the test results were bounding. Because the change was not significant the elevated temperature data are not presented in the paper. The explosions were initiated with either a small spark or a hot surface. Based on the results of these tests under the conditions investigated, it can be concluded that DDT of a stoichiometric hydrogen-oxygen mixture (and mixtures diluted with nitrogen and helium) within the 3013 containment system does not pose a threat to the structural integrity of the outer container.

Rawls, G.

2010-02-01T23:59:59.000Z

6

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

Science Conference Proceedings (OSTI)

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

Sebrell, W.

1983-07-01T23:59:59.000Z

7

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

SciTech Connect

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.

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

2012-10-01T23:59:59.000Z

8

Experiments to Address Lower Plenum Response Under Severe Accident Conditions: Volume 1: Technical Report  

Science Conference Proceedings (OSTI)

This report describes a set of experiments that were performed to address reactor pressure vessel lower plenum response under severe accident conditions. High temperature (2400 degrees K) debris was used to study the response of BWR and PWR lower head instrument tube penetrations of a BWR drain line and of molten debris quenching in a sub-cooled water pool. The importance of water as a heat sink within the instrument tube penetrations and drain line was quantified. Furthermore, the axial penetration of m...

1994-06-07T23:59:59.000Z

9

Experimental results from containment piping bellows subjected to severe accident conditions: Results from bellows tested in corroded conditions. Volume 2  

Science Conference Proceedings (OSTI)

Bellows are an integral part of the containment pressure boundary in nuclear power plants. They are used at piping penetrations to allow relative movement between piping and the containment wall, while minimizing the load imposed on the piping and wall. Piping bellows are primarily used in steel containments; however, they have received limited use in some concrete (reinforced and prestressed) containments. In a severe accident they may be subjected to pressure and temperature conditions that exceed the design values, along with a combination of axial and lateral deflections. A test program to determine the leak-tight capacity of containment penetration bellows is being conducted at Sandia National Laboratories under the sponsorship of the US Nuclear Regulatory Commission. Several different bellows geometries, representative of actual containment bellows, have been subjected to extreme deflections along with pressure and temperature loads. The bellows geometries and loading conditions are described along with the testing apparatus and procedures. A total of nineteen bellows have been tested. Thirteen bellows were tested in ``like-new`` condition (results reported in Volume 1), and six were tested in a corroded condition. The tests showed that bellows in ``like-new`` condition are capable of withstanding relatively large deformations, up to, or near, the point of full compression or elongation, before developing leakage, while those in a corroded condition did not perform as well, depending on the amount of corrosion. The corroded bellows test program and results are presented in this report.

Lambert, L.D.; Parks, M.B. [Sandia National Labs., Albuquerque, NM (United States)

1995-10-01T23:59:59.000Z

10

Experiments to Address Lower Plenum Response Under Severe Accident Conditions: Volume 2: Data Report, Part 1: Tests 1-6  

Science Conference Proceedings (OSTI)

This report describes a set of experiments that were performed to address reactor pressure vessel lower plenum response under severe accident conditions. High temperature (2400 degrees K) debris was used to study the response of BWR and PWR lower head instrument tube penetrations of a BWR drain line, and of molten debris quenching in a sub-cooled water pool. The importance of water as a heat sink within the instrument tube penetrations and drain line was quantified. Furthermore, the axial penetration of ...

1994-06-07T23:59:59.000Z

11

Experiments to Address Lower Plenum Response Under Severe Accident Conditions: Volume 2: Data Report, Part 2: Tests 7-10  

Science Conference Proceedings (OSTI)

This report describes a set of experiments that were performed to address reactor pressure vessel lower plenum response under severe accident conditions. High temperature (2400 degrees K) debris was used to study the response of BWR and PWR lower head instrument tube penetrations of a BWR drain line, and of molten debris quenching in a sub-cooled water pool. The importance of water as a heat sink within the instrument tube penetrations and drain line was quantified. Furthermore, the axial penetration of ...

1994-06-07T23:59:59.000Z

12

Revisiting Insights from Three Mile Island Unit 2 Postaccident Examinations and Evaluations in View of the Fukushima Daiichi Accident  

Science Conference Proceedings (OSTI)

The Three Mile Island Unit 2 (TMI-2) accident, which occurred on March 28, 1979, led industry and regulators to enhance strategies to protect against severe accidents in commercial nuclear power plants. Investigations in the years after the accident concluded that at least 45% of the core had melted and that nearly 19 tonnes of the core material had relocated to the lower head. Postaccident examinations indicate that about half of that material formed a solid layer near the lower head and above it was a layer of fragmented rubble. As discussed in this paper, numerous insights related to pressurized water reactor accident progression were gained from postaccident evaluations of debris, reactor pressure vessel (RPV) specimens, and nozzles taken from the RPV. In addition, information gleaned from TMI-2 specimen evaluations and available data from plant instrumentation were used to improve severe accident simulation models that form the technical basis for reactor safety evaluations. Finally, the TMI-2 accident led the nuclear community to dedicate considerable effort toward understanding severe accident phenomenology as well as the potential for containment failure. Because available data suggest that significant amounts of fuel heated to temperatures near melting, the events at Fukushima Daiichi Units 1, 2, and 3 offer an unexpected opportunity to gain similar understanding about boiling water reactor accident progression. To increase the international benefit from such an endeavor, we recommend that an international effort be initiated to (a) prioritize data needs; (b) identify techniques, samples, and sample evaluations needed to address each information need; and (c) help finance acquisition of the required data and conduct of the analyses.

Joy Rempe; Mitchell Farmer; Michael Corradini; Larry Ott; Randall Gauntt; Dana Powers

2012-11-01T23:59:59.000Z

13

Thermodynamic analysis of spent pyrochemical salts in the stored condition and in viable accident scenarios  

Science Conference Proceedings (OSTI)

This study involves examining ``spent`` electrorefining (ER) salts in the form present after usage (as stored), and then after exposure to water in a proposed accident scenario. Additionally, the equilibrium composition of the salt after extended exposure to air was also calculated by computer modeling and those results are also presented herein. It should be noted that these salts are extremely similar to spent MSE salts from the Rocky Flats MSE campaigns using NaCl-KCl- MgCl{sub 2}.

Axler, K.M.

1994-03-01T23:59:59.000Z

14

Computational Assessment of the GT-MHR Graphite Core Support Structural Integrity in Air-Ingress Accident Condition  

Science Conference Proceedings (OSTI)

The objective of this project was to perform stress analysis for graphite support structures of the General Atomics 600 MWth GT-MHR prismatic core design using ABAQUS (ver. 6.75) to assess their structural integrity in air-ingress accident conditions where the structure weakens over time due to oxidation damages. The graphite support structures of prismatic type GT-MHR was analyzed based on the change of temperature, burn-off and corrosion depth during the accident period predicted by GAMMA, a multi-dimensional gas multi-component mixture analysis code developed in the Republic of Korea (ROK)/United States (US) International Nuclear Engineering Research Initiative (I-NERI) project. Both the loading and thermal stresses were analyzed, but the thermal stress was not significant, leaving the loading stress to be the major factor. The mechanical strengths are exceeded between 11 to 11.5 days after loss-of-coolant-accident (LOCA), corresponding to 5.5 to 6 days after the start of natural convection.

Jong B. Lim; Eung S. Kim; Chang H. Oh; Richard R. Schultz; David A. Petti

2008-10-01T23:59:59.000Z

15

Study of Air Ingress Across the Duct During the Accident Conditions  

Science Conference Proceedings (OSTI)

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

Hassan, Yassin

2013-05-06T23:59:59.000Z

16

Analysis of containment performance and radiological consequences under severe accident conditions for the Advanced Neutron Source Reactor at the Oak Ridge National Laboratory  

SciTech Connect

A severe accident study was conducted to evaluate conservatively scoped source terms and radiological consequences to support the Advanced Neutron Source (ANS) Conceptual Safety Analysis Report (CSAR). Three different types of severe accident scenarios were postulated with a view of evaluating conservatively scoped source terms. The first scenario evaluates maximum possible steaming loads and associated radionuclide transport, whereas the next scenario is geared towards evaluating conservative containment loads from releases of radionuclide vapors and aerosols with associated generation of combustible gases. The third scenario follows the prescriptions given by the 10 CFR 100 guidelines. It was included in the CSAR for demonstrating site-suitability characteristics of the ANS. Various containment configurations are considered for the study of thermal-hydraulic and radiological behaviors of the ANS containment. Severe accident mitigative design features such as the use of rupture disks were accounted for. This report describes the postulated severe accident scenarios, methodology for analysis, modeling assumptions, modeling of several severe accident phenomena, and evaluation of the resulting source term and radiological consequences.

Kim, S.H.; Taleyarkhan, R.P.

1994-01-01T23:59:59.000Z

17

Experimental results from containment piping bellows subjected to severe accident conditions. Volume 1, Results from bellows tested in `like-new` conditions  

Science Conference Proceedings (OSTI)

Bellows are an integral part of the containment pressure boundary in nuclear power plants. They are used at piping penetrations to allow relative movement between piping and the containment wall, while minimizing the load imposed on the piping and wall. Piping bellows are primarily used in steel containments; however, they have received limited use in some concrete (reinforced and prestressed) containments. In a severe accident they may be subjected to pressure and temperature conditions that exceed the design values, along with a combination of axial and lateral deflections. A test program to determine the leak-tight capacity of containment penetration bellows is being conducted under the sponsorship of the US Nuclear Regulatory Commission at Sandia National Laboratories. Several different bellows geometries, representative of actual containment bellows, have been subjected to extreme deflections along with pressure and temperature loads. The bellows geometries and loading conditions are described along with the testing apparatus and procedures. A total of thirteen bellows have been tested, all in the `like-new` condition. (Additional tests are planned of bellows that have been subjected to corrosion.) The tests showed that bellows are capable of withstanding relatively large deformations, up to, or near, the point of full compression or elongation, before developing leakage. The test data is presented and discussed.

Lambert, L.D.; Parks, M.B. [Sandia National Labs., Albuquerque, NM (United States)

1994-09-01T23:59:59.000Z

18

TRUMP-BD: A computer code for the analysis of nuclear fuel assemblies under severe accident conditions  

Science Conference Proceedings (OSTI)

TRUMP-BD (Boil Down) is an extension of the TRUMP (Edwards 1972) computer program for the analysis of nuclear fuel assemblies under severe accident conditions. This extension allows prediction of the heat transfer rates, metal-water oxidation rates, fission product release rates, steam generation and consumption rates, and temperature distributions for nuclear fuel assemblies under core uncovery conditions. The heat transfer processes include conduction in solid structures, convection across fluid-solid boundaries, and radiation between interacting surfaces. Metal-water reaction kinetics are modeled with empirical relationships to predict the oxidation rates of steam-exposed Zircaloy and uranium metal. The metal-water oxidation models are parabolic in form with an Arrhenius temperature dependence. Uranium oxidation begins when fuel cladding failure occurs; Zircaloy oxidation occurs continuously at temperatures above 13000{degree}F when metal and steam are available. From the metal-water reactions, the hydrogen generation rate, total hydrogen release, and temporal and spatial distribution of oxide formations are computed. Consumption of steam from the oxidation reactions and the effect of hydrogen on the coolant properties is modeled for independent coolant flow channels. Fission product release from exposed uranium metal Zircaloy-clad fuel is modeled using empirical time and temperature relationships that consider the release to be subject to oxidation and volitization/diffusion ( bake-out'') release mechanisms. Release of the volatile species of iodine (I), tellurium (Te), cesium (Ce), ruthenium (Ru), strontium (Sr), zirconium (Zr), cerium (Cr), and barium (Ba) from uranium metal fuel may be modeled.

Lombardo, N.J.; Marseille, T.J.; White, M.D.; Lowery, P.S.

1990-06-01T23:59:59.000Z

19

LabVIEW with fuzzy logic controller simulation panel for condition monitoring of oil and dry type transformer  

E-Print Network (OSTI)

AbstractCondition monitoring of electrical power equipment has attracted considerable attention for many years. The aim of this paper is to use Labview with Fuzzy Logic controller to build a simulation system to diagnose transformer faults and monitor its condition. The front panel of the system was designed using LabVIEW to enable computer to act as customer-designed instrument. The dissolved gas-in-oil analysis (DGA) method was used as technique for oil type transformer diagnosis; meanwhile terminal voltages and currents analysis method was used for dry type transformer. Fuzzy Logic was used as expert system that assesses all information keyed in at the front panel to diagnose and predict the condition of the transformer. The outcome of the Fuzzy Logic interpretation will be displayed at front panel of LabVIEW to show the user the conditions of the transformer at any time. KeywordsLabVIEW, Fuzzy Logic, condition monitoring, oil transformer, dry transformer, DGA, terminal values.

N. A. Muhamad; S. A. M. Ali

2006-01-01T23:59:59.000Z

20

Severe Accident Studies  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Severe Accident Studies Severe Accident Studies Christopher S. Bajwa Division of Spent Fuel Storage and Transportation Office of Nuclear Material Safety and Safeguards USNRC 2012 U.S. DOE National Transportation Stakeholders Forum (NTSF) May 15 - 17, 2012 Knoxville, TN * Going The Distance? - The Safe Transport of Spent Nuclear Fuel and High-Level Radioactive Waste in the United States * Released February 9, 2006 * Conclusions: * NRC safety regulations are adequate to ensure package containment effectiveness over a wide range of transport conditions, including most credible accident conditions. * The radiological risks are well understood and are generally low, with the possible exception of risks from releases in extreme accidents involving long duration, fully engulfing fires.

Note: This page contains sample records for the topic "accident conditions view" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

Licensing topical report: the measurement and modelling of time-dependent fission product release from failed HTGR fuel particles under accident conditions  

Science Conference Proceedings (OSTI)

The release of fission products from failed fuel particles was measured under simulated accident (core heatup) conditions. A generic model and specific model parameters that describe delayed fission product release from the kernels of failed HTGR fuel particles were developed from the experimental results. The release of fission products was measured from laser-failed BISO ThO/sub 2/ and highly enriched (HEU) TRISO UC/sub 2/ particles that had been irradiated to a range of kernel burnups. The burnups were 0.25, 1.4, and 15.7% FIMA for ThO/sub 2/ particles and 23.5 and 74% FIMA for UC/sub 2/ particles. The fission products measured were nuclides of xenon, iodine, krypton, tellurium, and cesium.

Myers, B.F.; Morrissey, R.E.

1980-04-01T23:59:59.000Z

22

Nuclear Reactor Severe Accident Experiments  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Reactor Severe Accident Experiments Nuclear Reactor Severe Accident Experiments Capabilities Engineering Experimentation Reactor Safety Testing and Analysis Overview Nuclear Reactor Severe Accident Experiments MAX NSTF SNAKE Aerosol Experiments System Components Laser Applications Robots Applications Other Facilities Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Nuclear Reactor Severe Accident Experiments 1 2 3 4 5 6 7 We perform experiments simulating reactor core melt phenomena in which molten core debris ("corium") erodes the concrete floor of a containment building. This occurred during the Fukushima nuclear power plant accident though the extent of concrete damage is yet unknown. This video shows the top view of a churning molten pool of uranium oxide at 2000°C (3600°F) seen during an experiment at Argonne. Corium behaves much like lava.

23

Nuclear Reactor Accidents  

NLE Websites -- All DOE Office Websites (Extended Search)

Reactor Accidents The accidents at the Three Mile Island (TMI) and Chernobyl nuclear reactors have triggered particularly intense concern about radiation hazards. The TMI accident,...

24

Accident Investigation Handbook  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

SENSI NOT MEAS UREMENT TIVE D DOE-HDBK-1 1208-2012 July 2012 DOE E HA ANDBOOK K Ac ccide ent and d Op pera ational Sa afety y An naly ysis Volume e I: Ac ccide ent A Analy ysis Tec chniq ques U.S. Depar rtmen nt of En nergy Was shingto on, D.C C. 205 85 DOE-HDBK-1208-2012 INTRODUCTION - HANDBOOK APPLICATION AND SCOPE Accident Investigations (AI) and Operational Safety Reviews (OSR) are valuable for evaluating technical issues, safety management systems and human performance and environmental conditions to prevent accidents, through a process of continuous organizational learning. This Handbook brings together the strengths of the experiences gained in conducting Department of Energy (DOE) accident investigations over the past many years. That experience encourages us

25

Stress in accident and post-accident management at Chernobyl ?  

E-Print Network (OSTI)

Abstract. The effects of the Chernobyl nuclear accident on the psychology of the affected population have been much discussed. The psychological dimension has been advanced as a factor explaining the emergence, from 1990 onwards, of a post-accident crisis in the main CIS countries affected. This article presents the conclusions of a series of European studies, which focused on the consequences of the Chernobyl accident. These studies show that the psychological and social effects associated with the post-accident situation arise from the interdependency of a number of complex factors exerting a deleterious effect on the population. We shall first attempt to characterise the stress phenomena observed among the population affected by the accident. Secondly, we will be presenting an analysis of the various factors that have contributed to the emerging psychological and social features of population reaction to the accident and in post-accident phases, while not neglecting the effects of the pre-accident situation on the target population. Thirdly, we shall devote some initial consideration to the conditions that might be conducive to better management of postaccident stress. In conclusion, we shall emphasise the need to restore confidence among the population generally. 1.

Gilles Heriard Dubreuil

1996-01-01T23:59:59.000Z

26

EMERGENCY RESPONSE TO A TRANSPORTATION ACCIDENT INVOLVING RADIOACTIVE MATERIAL  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Emer Emer Emer Emer Emer Emergency Response to a T gency Response to a T gency Response to a T gency Response to a T gency Response to a Transportation ransportation ransportation ransportation ransportation Accident Involving Radioactive Material Accident Involving Radioactive Material Accident Involving Radioactive Material Accident Involving Radioactive Material Accident Involving Radioactive Material DISCLAIMER DISCLAIMER DISCLAIMER DISCLAIMER DISCLAIMER Viewing this video and completing the enclosed printed study material do not by themselves provide sufficient skills to safely engage in or perform duties related to emergency response to a transportation accident involving radioactive material. Meeting that goal is beyond the scope of this video and requires either additional

27

Systematics of Reconstructed Process Facility Criticality Accidents  

SciTech Connect

The systematics of the characteristics of twenty-one criticality accidents occurring in nuclear processing facilities of the Russian Federation, the United States, and the United Kingdom are examined. By systematics the authors mean the degree of consistency or agreement between the factual parameters reported for the accidents and the experimentally known conditions for criticality. The twenty-one reported process criticality accidents are not sufficiently well described to justify attempting detailed neutronic modeling. However, results of classic hand calculations confirm the credibility of the reported accident conditions.

Pruvost, N.L.; McLaughlin, T.P.; Monahan, S.P.

1999-09-19T23:59:59.000Z

28

Accident Investigation Handbook  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Improvement (HPI). The recommended techniques apply equally well to DOE Federal-led accident investigations conducted under DOE Order (O) 225.1B, Accident Investigations,...

29

Combining neural methods and knowledge-based methods in accident management  

Science Conference Proceedings (OSTI)

Accident management became a popular research issue in the early 1990s. Computerized decision support was studied from many points of view. Early fault detection and information visualization are important key issues in accident management also today. ...

Miki Sirola, Jaakko Talonen

2012-01-01T23:59:59.000Z

30

DECONTAMINATION DRESSDOWN AT A TRANSPORTATION ACCIDENT INVOLVING RADIOACTIVE MATERIAL  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Video User' s Guide Video User' s Guide DECONTAMINATION DRESSDOWN AT A TRANSPORTATION ACCIDENT INVOLVING RADIOACTIVE MATERIAL DISCLAIMER Viewing this video and completing the enclosed printed study material do not by themselves provide sufficient skills to safely engage in or perform duties related to emergency response to a transportation accident involving radioactive material. Meeting that goal is beyond

31

Next-generation nuclear fuel withstands high-temperature accident...  

NLE Websites -- All DOE Office Websites (Extended Search)

(more than 200 degrees Celsius greater than postulated accident conditions) most fission products remained inside the fuel particles, which each boast their own primary...

32

TMI-2 accident: core heat-up analysis  

SciTech Connect

This report summarizes NSAC study of reactor core thermal conditions during the accident at Three Mile Island, Unit 2. The study focuses primarily on the time period from core uncovery (approximately 113 minutes after turbine trip) through the initiation of sustained high pressure injection (after 202 minutes). The transient analysis is based upon established sequences of events; plant data; post-accident measurements; interpretation or indirect use of instrument responses to accident conditions.

Ardron, K.H.; Cain, D.G.

1981-01-01T23:59:59.000Z

33

Nuclear criticality accidents  

SciTech Connect

Criticality occurs when a sufficient quantity of fissionable material is accumulated, and it results in the liberation of nuclear energy. All process accidents have involved plutonium or highly enriched uranium, as have most of the critical experiment accidents. Slightly enriched uranium systems require much larger quantities of material to achieve criticality. An appreciation of criticality accidents should be based on an understanding of factors that influence criticality, which are discussed in this article. 11 references.

Smith, D.R. (Los Alamos National Laboratory, New Mexico (Unites States))

1991-10-01T23:59:59.000Z

34

Severe Accident Studies | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Severe Accident Studies Severe Accident Studies Powerpoint discussing studies and conclusions on transportation accidents and safety. Severe Accident Studies More Documents &...

35

Substantiation of Thermodynamic Criteria of Explosion Safety in Process of Severe Accidents in Pressure Vessel Reactors  

E-Print Network (OSTI)

The paper represents original development of thermodynamic criteria of occurrence conditions of steam-gas explosions in the process of severe accidents. The received results can be used for modelling of processes of severe accidents in pressure vessel reactors.

Skalozubov, V I; Jarovoj, S S; Kochnyeva, V Yu

2012-01-01T23:59:59.000Z

36

Substantiation of Thermodynamic Criteria of Explosion Safety in Process of Severe Accidents in Pressure Vessel Reactors  

E-Print Network (OSTI)

The paper represents original development of thermodynamic criteria of occurrence conditions of steam-gas explosions in the process of severe accidents. The received results can be used for modelling of processes of severe accidents in pressure vessel reactors.

V. I. Skalozubov; V. N. Vashchenko; S. S. Jarovoj; V. Yu. Kochnyeva

2012-03-27T23:59:59.000Z

37

Accident resistant transport container  

DOE Patents (OSTI)

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.

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

1980-01-01T23:59:59.000Z

38

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

SciTech Connect

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.

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

1990-06-01T23:59:59.000Z

39

Microsoft Word - Unrelated Accident  

NLE Websites -- All DOE Office Websites (Extended Search)

For Immediate Release 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 involved an 18-wheeler is not related to Waste Isolation Pilot Plant (WIPP) transuranic waste shipments. Involved in the accident was a load of new, unused 55-gallon drums manufactured in Carlsbad that was en route to Richland, Washington. The Waste Isolation Pilot Plant is a U.S. Department of Energy facility designed to safely isolate defense-related transuranic waste from people and the environment. Waste temporarily stored at sites around the country is shipped to WIPP and permanently disposed in rooms mined out of an ancient salt formation 2,150 feet below the surface. WIPP, which began waste

40

Computerized Accident Incident Reporting System  

Energy.gov (U.S. Department of Energy (DOE))

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...

Note: This page contains sample records for the topic "accident conditions view" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

April 2013 Most Viewed Documents for Fission And Nuclear Technologies |  

Office of Scientific and Technical Information (OSTI)

April 2013 Most Viewed Documents for Fission And Nuclear Technologies April 2013 Most Viewed Documents for Fission And Nuclear Technologies Behavior of spent nuclear fuel in water pool storage Johnson, A.B. Jr. (null) 298 Estimation of gas leak rates through very small orifices and channels. [From sealed PuO/sub 2/ containers under accident conditions] Bomelburg, H.J. (null) 292 Graphite design handbook Ho, F.H. (1988) 216 System Definition and Analysis: Power Plant Design and Layout NONE (1996) 123 Flow-induced vibration of circular cylindrical structures Chen, S.S. (1985) 116 Stress analysis and evaluation of a rectangular pressure vessel. [For equipment for sampling Hanford tank radwaste] Rezvani, M.A.; Ziada, H.H. (Westinghouse Hanford Co., Richland, WA (United States)); Shurrab, M.S. (Westinghouse Savannah River Co., Aiken, SC (United

42

September 2013 Most Viewed Documents for Fission And Nuclear...  

Office of Scientific and Technical Information (OSTI)

Fission And Nuclear Technologies Estimation of gas leak rates through very small orifices and channels. From sealed PuOsub 2 containers under accident conditions Bomelburg,...

43

Barriers to Switching Accidents  

Science Conference Proceedings (OSTI)

The EPRI Switching Safety & Reliability Project Steering Committee sponsored development of a self-study based training program for personnel who perform switching. Some of the earlier EPRI Switching Safety & Reliability research projects that focused on the causes of switching errors, highlighted a need to reduce the 'complacency' that tends to develop as switching activities are performed over and over again and become 'routine.' Most switching accidents or incidents involve personnel who were trained ...

2005-12-22T23:59:59.000Z

44

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

Science Conference Proceedings (OSTI)

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.

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

2010-03-01T23:59:59.000Z

45

Severe accident analysis using dynamic accident progression event trees.  

E-Print Network (OSTI)

??In present, the development and analysis of Accident Progression Event Trees (APETs) are performed in a manner that is computationally time consuming, difficult to reproduce (more)

Hakobyan, Aram P

2006-01-01T23:59:59.000Z

46

Modular Accident Analysis Program (MAAP5) Applications Assessment  

Science Conference Proceedings (OSTI)

The Modular Accident Analysis Program (MAAP) is widely used throughout North America, Europe, and the Far East to analyze plant responses over a broad spectrum of potential accident conditions. The use of MAAP continues to increase because its representation of integral plant response and short run times make this program ideal for supporting engineering evaluations. With greater use, however, the level of detail to be represented within the reactor core, reactor coolant system (RCS), and containment has...

2005-12-08T23:59:59.000Z

47

Study on drywell cooler applicability to severe accident management  

SciTech Connect

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

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

2004-07-01T23:59:59.000Z

48

APS Guideline for Accident Investigations  

NLE Websites -- All DOE Office Websites (Extended Search)

occurring in CATXSDs facilities at the APS. Definitions Accident: an unexpected event that produces personal injury, illness, or death; damage to or loss of property or...

49

Accurate accident reconstruction in VANET  

Science Conference Proceedings (OSTI)

We propose a forensic VANET application to aid an accurate accident reconstruction. Our application provides a new source of objective real-time data impossible to collect using existing methods. By leveraging inter-vehicle communications, we compile ... Keywords: EDR, VANET, accident reconstruction, in-vehicle applications

Yuliya Kopylova; Csilla Farkas; Wenyuan Xu

2011-07-01T23:59:59.000Z

50

Less than severe worst case accidents  

Science Conference Proceedings (OSTI)

Many systems can provide tremendous benefit if operating correctly, produce only an inconvenience if they fail to operate, but have extreme consequences if they are only partially disabled such that they operate erratically or prematurely. In order to assure safety, systems are often tested against the most severe environments and accidents that are considered possible to ensure either safe operation or safe failure. However, it is often the less severe environments which result in the ``worst case accident`` since these are the conditions in which part of the system may be exposed or rendered unpredictable prior to total system failure. Some examples of less severe mechanical, thermal, and electrical environments which may actually be worst case are described as cautions for others in industries with high consequence operations or products.

Sanders, G.A.

1996-08-01T23:59:59.000Z

51

EPR Severe Accident Threats and Mitigation  

SciTech Connect

Despite the extremely low EPR core melt frequency, an improved defence-in-depth approach is applied in order to comply with the EPR safety target: no stringent countermeasures should be necessary outside the immediate plant vicinity like evacuation, relocation or food control other than the first harvest in case of a severe accident. Design provisions eliminate energetic events and maintain the containment integrity and leak-tightness during the entire course of the accident. Based on scenarios that cover a broad range of physical phenomena and which provide a sound envelope of boundary conditions associated with each containment challenge, a selection of representative loads has been done, for which mitigation measures have to cope with. This paper presents the main critical threats and the approach used to mitigate those threats. (authors)

Azarian, G. [Framatome ANP SAS, Tour Areva, Place de la Coupole 92084 Paris la Defense (France); Kursawe, H.M.; Nie, M.; Fischer, M.; Eyink, J. [Framatome ANP GmbH, Freyeslebenstrasse, 1, D-91058 Erlangen (Germany); Stoudt, R.H. [Framatome ANP Inc. - 3315 Old Forest Rd, Lynchburgh, VA 24501 (United States)

2004-07-01T23:59:59.000Z

52

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

Science Conference Proceedings (OSTI)

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.

Su'ud, Zaki; Anshari, Rio [Nuclear and Biophysics Research Group, Dept. of Physics, Bandung Institute of Technology, Jl.Ganesha 10, Bandung, 40132 (Indonesia)

2012-06-06T23:59:59.000Z

53

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

SciTech Connect

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

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

2006-07-01T23:59:59.000Z

54

Most Viewed Documents - Fission and Nuclear Technologies | OSTI, US Dept of  

Office of Scientific and Technical Information (OSTI)

Most Viewed Documents - Fission and Nuclear Technologies Most Viewed Documents - Fission and Nuclear Technologies Metals design handbook Betts, W.S. (1988) Estimation of gas leak rates through very small orifices and channels. [From sealed PuO/sub 2/ containers under accident conditions] Bomelburg, H.J. () Graphite design handbook Ho, F.H. (1988) Motor-operated valve (MOV) actuator motor and gearbox testing DeWall, K.; Watkins, J.C.; Bramwell, D. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)] (1997) Environmental Aspects, Objectives and Targets Identification Process R. Green (2002) Flow-induced vibration of circular cylindrical structures Chen, S.S. (1985) System Definition and Analysis: Power Plant Design and Layout NONE (1996) Materials and design bases issues in ASME Code Case N-47 Huddleston, R.L.; Swindeman, R.W. (Oak Ridge National Lab., TN (United

55

View / Download  

Science Conference Proceedings (OSTI)

thetic Coal-Petcoke Slags (Al2O3-CaO-. FeO-SiO2-V2O3) under Simulated Gasifi - cation Conditions, published in Energy and Fuels. 2012 Class of Fellows.

56

A CANDU Severe Accident Analysis  

Science Conference Proceedings (OSTI)

As interest in severe accident studies has increased in the last years, we have developed a set of simple models to analyze severe accidents for CANDU reactors that should be integrated in the EU codes. The CANDU600 reactor uses natural uranium fuel and heavy water (D2O) as both moderator and coolant, with the moderator and coolant in separate systems. We chose to analyze accident development for a LOCA with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperatures inside a pressure tube reaches 10000 deg C, a contact between pressure tube and calandria tube occurs and the residual heat is transferred to the moderator. Due to the lack of cooling, the moderator eventually begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) will be uncovered, then will disintegrate and fall down to the calandria vessel bottom. After all the quantity of moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which normally surrounds the calandria vessel. The phenomena described above are modelled, analyzed and compared with the existing data. The results are encouraging. (authors)

Negut, Gheorghe; Catana, Alexandru [Institute for Nuclear Research, 1, Compului Str., Mioveni, PO Box 78, 0300 Pitesti (Romania); Prisecaru, Ilie [University Politehnica Bucharest (Romania)

2006-07-01T23:59:59.000Z

57

Naval Spent Fuel Rail Shipment Accident Exercise Objectives ...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

58

Fast Transient And Spatially Non-Homogenous Accident Analysis Of Two-Dimensional Cylindrical Nuclear Reactor  

Science Conference Proceedings (OSTI)

The research about fast transient and spatially non-homogenous nuclear reactor accident analysis of two-dimensional nuclear reactor has been done. This research is about prediction of reactor behavior is during accident. In the present study, space-time diffusion equation is solved by using direct methods which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference discretization method is solved by using iterative methods ADI (Alternating Direct Implicit). The indication of accident is decreasing macroscopic absorption cross-section that results large external reactivity. The power reactor has a peak value before reactor has new balance condition. Changing of temperature reactor produce a negative Doppler feedback reactivity. The reactivity will reduce excess positive reactivity. Temperature reactor during accident is still in below fuel melting point which is in secure condition.

Yulianti, Yanti [Dept. of Physics, Universitas Lampung (UNILA), Jl. Sumantri Brojonegor No.1 Bandar Lampung (Indonesia); Dept. of Physics, Institut Teknologi Bandung (ITB), Jl. Ganesha 10 Bandung (Indonesia); Su'ud, Zaki; Waris, Abdul; Khotimah, S. N. [Dept. of Physics, Institut Teknologi Bandung (ITB), Jl. Ganesha 10 Bandung (Indonesia); Shafii, M. Ali [Dept. of Physics, Institut Teknologi Bandung (ITB), Jl. Ganesha 10 Bandung (Indonesia); Dept. of Physics, Universitas Andalas (UNAND), Kampus Limau Manis, Padang, Sumatera Barat (Indonesia)

2010-12-23T23:59:59.000Z

59

Severe Accident Management Guidance Technical Basis Report  

Science Conference Proceedings (OSTI)

Guidance to aid operating crews in responding to a severe core damage accident was first developed as a response to the 1979 accident at Three Mile Island Unit 2. This guidance encompasses those actions that could be considered to arrest the progression of a core damage accident or to limit the extent of resulting releases of fission products. The original guidance was developed in a logical manner, starting with compiling the best information regarding severe-accident phenomena available at that ...

2012-10-31T23:59:59.000Z

60

Chernobyl Nuclear Accident | National Nuclear Security Administration  

NLE Websites -- All DOE Office Websites (Extended Search)

Chernobyl Nuclear Accident | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response...

Note: This page contains sample records for the topic "accident conditions view" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

Evolvable neural networks ensembles for accidents diagnosis  

Science Conference Proceedings (OSTI)

Prediction and diagnosis of nuclear accidents is one of the most important tasks for nuclear safety. Since accurate diagnosis of nuclear accident is a very important issue for avoidance of disastrous outcomes, it is more desirable to make a decision ... Keywords: ensembles, neuroevolution, nuclear accidents

Hany Sallam; Carlo S. Regazzoni; Ihab Talkhan; Amir Atiya

2008-07-01T23:59:59.000Z

62

Evironmental health policy in ukraine after the Chernobyl accident  

SciTech Connect

The 1986 accident at the Chernobyl nuclear power plant in Ukraine produced severe environmental health problems. This paper reports on the environmental health conditions in Ukraine after the accident and the health policy approaches employed to respond to the environmental conditions and health problems. Crisis conditions and a period of rapid change in Ukraine contributed to the difficulties of developing and implementing policy to address serious environmental health problems. Despite these difficulties, Ukraine is taking effective action. The paper describes the primary environmental health problem areas and the efforts taken to solve them. The effect of intense public fear of radiation on policymaking is described. The paper discusses the ability of public fear to distort health policy towards certain problems, leaving problems of greater importance with fewer resources. 35 refs., 1 fig.

Page, G.W.; Bobyleva, O.A.; Naboka, M.V. [and others

1995-09-01T23:59:59.000Z

63

View-Augmented Abstractions  

Science Conference Proceedings (OSTI)

This paper introduces view-augmented abstractions, which specialize an underlying numeric domain to focus on a particular expression or set of expressions. A view-augmented abstraction adds a set of materialized views to the original domain. View augmentation ... Keywords: Numeric abstract domains, abstract-interpretation precision, view maintenance

Matt Elder; Denis Gopan; Thomas Reps

2010-10-01T23:59:59.000Z

64

Thermal-Hydraulic Modeling of the Primary Coolant System of Light Water Reactors During Severely Degraded Core Accidents  

Science Conference Proceedings (OSTI)

The transport of fission-product vapors and aerosols that would be released from an LWR primary system in postulated severe accidents depends on the prevalent thermal-hydraulic conditions. The analytic models developed in this study are incorporated in the PSAAC modular computer program, which can help predict more realistic estimates of accident consequences.

1984-07-01T23:59:59.000Z

65

Fission converter heat removal and safety under accident conditions  

Science Conference Proceedings (OSTI)

The design and safety of the heat removal system of the Massachusetts Institute of Technology (MIT) design for a fission converter-based epithermal beam is discussed in this paper. Plate-type reactor fuel elements, used in the MIT research reactor (MITR-II), are also used for the fission converter. This fission converter-based beam provides epithermal neutron fluxes at the patient position in excess of 10{sup 10} n/cm{sup 2}s with very low contamination of fast neutrons and gamma rays.

Sutharshan, B.; Todreas, N.E.; Harling, O.K. [Massachusetts Inst. of Technology, Cambridge, MA (United States)

1996-12-31T23:59:59.000Z

66

Traffic Conditions and Truck Accidents on Urban Freeways  

E-Print Network (OSTI)

Highway Safety. University of Kentucky, College ofwas conducted at the University of Kentucky in an effort to

Golob, Thomas F; Regan, Amelia C

2004-01-01T23:59:59.000Z

67

Computerized Accident/Incident Reporting System  

NLE Websites -- All DOE Office Websites (Extended Search)

Accident Recordkeeping and Reporting Accident Recordkeeping and Reporting Accident/Incident Recordkeeping and Reporting CAIRS logo 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. Injury and Illness Dashboard The Dashboard provides an alternate interface to CAIRS information. The initial release of the Dashboard allows analysis of composite DOE-wide information and summary information by Program Office, and site. Additional data feature are under development. CAIRS Registration Form CAIRS is a Government computer system and, as such, has security requirements that must be followed. Access to the

68

Recommendations for Analyzing Accidents Under NEPA  

Energy.gov (U.S. Department of Energy (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.

69

Accident Tolerant Fuels for Light Water Reactors  

Science Conference Proceedings (OSTI)

Presentation Title, Accident Tolerant Fuels for Light Water Reactors. Author(s), Steven J. Zinkle, Kurt A. Terrani, Lance L. Snead. On-Site Speaker (Planned)...

70

ORISE: REAC/TS Radiation Accident Registries  

NLE Websites -- All DOE Office Websites (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...

71

Accident Investigation Report Plutonium Contamination in the...  

NLE Websites -- All DOE Office Websites (Extended Search)

Accident Investigation Report Plutonium Contamination in the Zero Power Physics Reactor Facility at the Idaho National Laboratory, November 8, 2011 January 2012 Disclaimer...

72

Accident management for indian pressurized heavy water reactors  

Science Conference Proceedings (OSTI)

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

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

2006-07-01T23:59:59.000Z

73

Analysis of Kuosheng Station Blackout Accident Using MELCOR 1.8.4  

Science Conference Proceedings (OSTI)

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

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

2000-11-15T23:59:59.000Z

74

Mitigation of Severe Accident Consequences Using Inherent Safety Principles  

Science Conference Proceedings (OSTI)

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

R. A. Wigeland; J. E. Cahalan

2009-12-01T23:59:59.000Z

75

The Hartford Life and Accident Insurance  

E-Print Network (OSTI)

The Hartford Life and Accident Insurance Company Group Numbers Basic Group Term Life AD&D-677984 Life and Accident Insurance Company. (Referred to as The Hartford or Hartford.) General information industry. Europ Assist has been helping customers in times of crisis for more than 46 years. They have

76

Does Daylight Savings Time Affect Traffic Accidents?  

E-Print Network (OSTI)

This paper studies the effect of changes in accident pattern due to Daylight Savings Time (DST). The extension of the DST in 2007 provides a natural experiment to determine whether the number of traffic accidents is affected by shifts in hours of daylight using the year as control group. Using data on traffic accidents in Texas based on crash reports provided by the Texas Transportation Institute, and a difference in differences technique, this study creates a regression model to determine how significant this factor is in affecting traffic accident patterns as observed in the data. Results show that DST has no statistically significant effect on traffic accidents of all categories including (but not limited to) highway, non-highway, and accidents, accidents with injuries and no injuries, and accidents by drivers of all age-groups. This implies that the federal governments policy of DST (and its extension) has no costs incurred by a rise in motor vehicle crashes when it gets dark early.

Deen, Sophia 1988-

2012-05-01T23:59:59.000Z

77

Accident states simulation: process fluids release  

Science Conference Proceedings (OSTI)

Seveso II Directive imposes for high hazardous plants quantitative risk evaluation of the major accident. In a general context the risk is defined as product between frequency and consequences of accident state. There are five steps in quantitative risk ... Keywords: hazard, hydrogen sulphide, mathematical model, release, risk, safety system, simulation

Cornelia Croitoru; Mihai Anghel; Floarea Pop; Ioan Stefanescu; Gheorghe Titescu; Mihai Patrascu; Ervin Watzlawek; Dorin Cheresdi

2008-08-01T23:59:59.000Z

78

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

Science Conference Proceedings (OSTI)

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).

NONE

1991-09-01T23:59:59.000Z

79

User_ViewRecords  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

User Records User Records © 2011 SuccessFactors, Inc. - 1 - SuccessFactors Learning Confidential. All rights reserved. Job Aid: Viewing User Records Purpose The purpose of this job aid is to guide users through the step-by-step process of viewing their records. Each task demonstrates viewing of different records. Task A. View To-Do List Enter the web address (URL) of the user application into your browser Address field and press the Enter key. Enter your user ID in the User ID textbox. Enter your password in the Password textbox. Click Sign In. View To-Do List (filter, view) 7 Steps Task A View Completed Work 8 Steps Task B View Curriculum Status and Details 11 Steps Task C 3 3 1 2 2 1 SuccessFactors Learning v 6.4 User Job Aid Viewing User Records

80

Commercial SNF Accident Release Fractions  

Science Conference Proceedings (OSTI)

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

J. Schulz

2004-11-05T23:59:59.000Z

Note: This page contains sample records for the topic "accident conditions view" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

The Fukushima Daiichi Accident Study Information Portal  

SciTech Connect

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.

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

2012-11-01T23:59:59.000Z

82

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Web Based Course: SAF-230DE, Accident Investigation Overview Promotional Video Web Based Course: SAF-230DE, Accident Investigation Overview Promotional Video September 20, 2013 -...

83

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

NLE Websites -- All DOE Office Websites (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...

84

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

85

Insights into the behavior of nuclear power plant containments during severe accidents  

SciTech Connect

The containment building surrounding a nuclear reactor offers the last barrier to the release of radioactive materials from a severe accident into the environment. The loading environment of the containment under severe accident conditions may include much greater than design pressures and temperatures. Investigations into the performance of containments subject to ultimate or failure pressure and temperature conditions have been performed over the last several years through a program administered by the Nuclear Regulatory Commission (NRC). These NRC sponsored investigations are subsequently discussed. Reviewed are the results of large scale experiments on reinforced concrete, prestressed concrete, and steel containment models pressurized to failure. In conjunction with these major tests, the results of separate effect testing on many of the critical containment components; that is, aged and unaged seals, a personnel air lock and electrical penetration assemblies subjected to elevated temperature and pressure have been performed. An objective of the NRC program is to gain an understanding of the behavior of typical existing and planned containment designs subject to postulated severe accident conditions. This understanding has led to the development of experimentally verified analytical tools that can be applied to accurately predict their ultimate capacities useful in developing severe accident mitigation schemes. Finally, speculation on the response of containments subjected to severe accident conditions is presented.

Horschel, D.S.; Ludwigsen, J.S.; Parks, M.B.; Lambert, L.D. [Sandia National Labs., Albuquerque, NM (United States); Dameron, R.A.; Rashid, Y.R. [ANATECH Research Corp., San Diego, CA (United States)

1993-06-01T23:59:59.000Z

86

Transactional process views  

Science Conference Proceedings (OSTI)

To enable effective interorganisational collaborations, process providers have to disclose relevant parts of their local business processes in public process views. A public process view has to be consistent with the underlying private process. Local ...

Rik Eshuis; Jochem Vonk; Paul Grefen

2011-10-01T23:59:59.000Z

87

A POTENTIAL APPLICATION OF UNCERTAINTY ANALYSIS TO DOE-STD-3009-94 ACCIDENT ANALYSIS  

Science Conference Proceedings (OSTI)

The objective of this paper is to assess proposed transuranic waste accident analysis guidance and recent software improvements in a Windows-OS version of MACCS2 that allows the inputting of parameter uncertainty. With this guidance and code capability, there is the potential to perform a quantitative uncertainty assessment of unmitigated accident releases with respect to the 25 rem Evaluation Guideline (EG) of DOE-STD-3009-94 CN3 (STD-3009). Historically, the classification of safety systems in a U.S. Department of Energy (DOE) nuclear facility's safety basis has involved how subject matter experts qualitatively view uncertainty in the STD-3009 Appendix A accident analysis methodology. Specifically, whether consequence uncertainty could be larger than previously evaluated so the site-specific accident consequences may challenge the EG. This paper assesses whether a potential uncertainty capability for MACCS2 could provide a stronger technical basis as to when the consequences from a design basis accident (DBA) truly challenges the 25 rem EG.

Palmrose, D E; Yang, J M

2007-05-10T23:59:59.000Z

88

Analysis of PWR RCS Injection Strategy During Severe Accident  

Science Conference Proceedings (OSTI)

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

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

2004-05-15T23:59:59.000Z

89

Markov Model of Severe Accident Progression and Management  

SciTech Connect

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

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

2012-06-25T23:59:59.000Z

90

Site restoration: Estimation of attributable costs from plutonium-dispersal accidents  

SciTech Connect

A nuclear weapons accident is an extremely unlikely event due to the extensive care taken in operations. However, under some hypothetical accident conditions, plutonium might be dispersed to the environment. This would result in costs being incurred by the government to remediate the site and compensate for losses. This study is a multi-disciplinary evaluation of the potential scope of the post-accident response that includes technical factors, current and proposed legal requirements and constraints, as well as social/political factors that could influence decision making. The study provides parameters that can be used to assess economic costs for accidents postulated to occur in urban areas, Midwest farmland, Western rangeland, and forest. Per-area remediation costs have been estimated, using industry-standard methods, for both expedited and extended remediation. Expedited remediation costs have been evaluated for highways, airports, and urban areas. Extended remediation costs have been evaluated for all land uses except highways and airports. The inclusion of cost estimates in risk assessments, together with the conventional estimation of doses and health effects, allows a fuller understanding of the post-accident environment. The insights obtained can be used to minimize economic risks by evaluation of operational and design alternatives, and through development of improved capabilities for accident response.

Chanin, D.I.; Murfin, W.B. [Technadyne Engineering Consultants, Inc., Albuquerque, NM (United States)

1996-05-01T23:59:59.000Z

91

Rail transportation risk and accident severity: A statistical analysis of variables in FRA's accident/incident data base  

Science Conference Proceedings (OSTI)

The Federal Railroad Administration (US DOT) maintains a file of carrier-reported railroad accidents and incidents that meet stipulated threshold criteria for damage cost and/or casualties. A thoroughly-cleaned five-year time series of this data base was subjected to unbiased statistical procedures to discover (a) important causative variables in severe (high damage cost) accidents and (b) other key relationships between objective accident conditions and frequencies. Just under 6000 records, each representing a single event involving rail freight shipments moving on mainline track, were subjected to statistical frequency analysis, then included in the construction of classification and regression trees as described by Breimann et al. (1984). Variables related to damage cost defined the initial splits,'' or branchings of the tree. An interesting implication of the results of this analysis with respect to transportation of hazardous wastes by rail is that movements should be avoided when ambient temperatures are extreme (significantly 80{degrees}F), but that there should be no a priori bias against shipping wastes in longer train consists. 2 refs., 2 figs., 12 tabs.

Saricks, C.L. (Argonne National Lab., IL (USA). Energy Systems Div.); Janssen, I. (Argonne National Lab., IL (USA). Biological and Medical Research Div.)

1991-01-01T23:59:59.000Z

92

Decontamination Dressdown at a Transportation Accident Involving  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Decontamination Dressdown at a Transportation Accident Involving Decontamination Dressdown at a Transportation Accident Involving Radioactive Material Decontamination Dressdown at a Transportation Accident Involving Radioactive Material 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 reviewing those points with students. The Student Handout should be distributed to students after the video is shown and the instructor should use the Guide to facilitate a discussion on how the decontamination dressdown process is implemented. During this discussion, the instructor can present various scenarios, each of which would discuss decontamination at the accident scene. The purpose of this discussion would be to cover how responders

93

A systems approach to food accident analysis  

E-Print Network (OSTI)

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

Helferich, John D

2011-01-01T23:59:59.000Z

94

Chernobyl accident: A comprehensive risk assessment  

SciTech Connect

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.

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

1999-11-01T23:59:59.000Z

95

A SUMMARY OF INDUSTRIAL ACCIDENTS IN USAEC FACILITIES  

SciTech Connect

The summary includes descriptions of serious accidents for l959 and 1960, AEC industrial injury frequency rates, criticality accidents, radiation exposures, accidents involving radioactive materials in AEC activities during 1959 and 1960, and accidents involving fatalities in AEC activities during l959 and 1960. (B.O.G.)

1961-12-01T23:59:59.000Z

96

TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident  

SciTech Connect

The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensors survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation process that could be implemented in upcoming Fukushima Daiichi recovery efforts.

Joy L. Rempe; Darrell L. Knudson

2013-03-01T23:59:59.000Z

97

Three Mile Island accident and post-accident recovery: what did we learn  

SciTech Connect

A description of the accident at Three Mile Island-2 reactor is presented. Activities related to the cleanup and decontamination of the reactor are described.

Collins, E.D.

1982-01-01T23:59:59.000Z

98

Low level waste shipment accident lessons learned  

SciTech Connect

On October 1, 1994 a shipment of low-level waste from the Fernald Environmental Management Project, Fernald, Ohio, was involved in an accident near Rolla, Missouri. The accident did not result in the release of any radioactive material. The accident did generate important lessons learned primarily in the areas of driver and emergency response communications. The shipment was comprised of an International Standards Organization (ISO) container on a standard flatbed trailer. The accident caused the low-level waste package to separate from the trailer and come to rest on its top in the median. The impact of the container with the pavement and median inflicted relatively minor damage to the container. The damage was not substantial enough to cause failure of container integrity. The success of the package is attributable to the container design and the packaging procedures used at the Fernald Environmental Management Project for low-level waste shipments. Although the container survived the initial wreck, is was nearly breached when the first responders attempted to open the ISO container. Even though the container was clearly marked and the shipment documentation was technically correct, this information did not identify that the ISO container was the primary containment for the waste. The lessons learned from this accident have DOE complex wide applicability. This paper is intended to describe the accident, subsequent emergency response operations, and the lessons learned from this incident.

Rast, D.M.; Rowe, J.G.; Reichel, C.W.

1995-02-01T23:59:59.000Z

99

An analysis of evacuation options for nuclear accidents  

Science Conference Proceedings (OSTI)

In this report we consider the threat posed by the accidental release of radionuclides from a nuclear power plant. The objective is to establish relationships between radiation dose and the cost of evacuation under a wide variety of conditions. The dose can almost always be reduced by evacuating the population from a larger area. However, extending the evacuation zone outward will cause evacuation costs to increase. The purpose of this analysis was to provide the Environmental Protection Agency (EPA) a data base for evaluating whether implementation costs and risks averted could be used to justify evacuation at lower doses. The procedures used and results of these analyses are being made available as background information for use by others. We develop cost/dose relationships for 54 scenarios that are based upon the severity of the reactor accident, meteorological conditions during the release of radionuclides into the environment, and the angular width of the evacuation zone. The 54 scenarios are derived from combinations of three accident severity levels, six meteorological conditions and evacuation zone widths of 70{degree}, 90{degree}, and 180{degree}.

Tawil, J.J.; Strenge, D.L.; Schultz, R.W. [Battelle Memorial Inst., Richland, WA (United States)

1987-11-01T23:59:59.000Z

100

April 2013 Most Viewed Documents for Fossil Fuels | OSTI, US...  

Office of Scientific and Technical Information (OSTI)

Viewed Documents for Fossil Fuels EXPERIMENTAL AND THEORETICAL DETERMINATION OF HEAVY OIL VISCOSITY UNDER RESERVOIR CONDITIONS Dr. Jorge Gabitto; Maria Barrufet (2003) 208 Fluid...

Note: This page contains sample records for the topic "accident conditions view" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


101

SAF-230DE - Web Based Course: Accident Investigation Overview | Department  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

SAF-230DE - Web Based Course: Accident Investigation Overview SAF-230DE - Web Based Course: Accident Investigation Overview SAF-230DE - Web Based Course: Accident Investigation Overview September 18, 2013 - 10:52am Addthis SAF-230DE - Web Based Course: Accident Investigation Overview The Office of Health Safety and Security (HSS) National Training Center (NTC) in collaboration with the HSS Accident Investigation Program (HS-24) has developed and released a course that provides an overview of the fundamentals of accident investigation. This course is intended to meet the every five year refresher training requirement for DOE Federal Accident Investigators under DOE Order 225.1B "Accident Investigations", and serves as an orientation to other DOE Federal Accident Investigation Board Members who need a basic knowledge of

102

Severe Accident Management Guidance Technical Basis Report: Volumes 1 and 2  

Science Conference Proceedings (OSTI)

Severe accident management guidance encompasses actions that would be taken to recover from a damaged core condition and to prevent or mitigate the release of fission products. This report provides the technical basis for developing such guidance by the nuclear steam supply system owners groups.

1993-04-01T23:59:59.000Z

103

Severe accident testing of a personnel airlock  

Science Conference Proceedings (OSTI)

Sandia National Laboratories (Sandia) is investigating the leakage potential of mechanical penetrations as part of a research program on containment integrity under severe accident loads for the US Nuclear Regulatory Commission (NRC). Barnes et al. (1984) and Shackelford et al. (1985) identified leakage from personnel airlocks as an important failure mode of containments subject to severe accident loads. However, these studies were based on relatively simple analysis methods. The complex structural interaction between the door, gasket, and bulkhead in personnel airlocks makes analytical evaluation of leakage difficult. In order to provide data to validate methods for evaluating the leakage potential, a full-size personnel airlock was subject to simulated severe accident loads consisting of pressure and temperature up to 300 psig and 800/degree/F. The test was conducted at Chicago Bridge and Iron under contract to Sandia. Julien and Peters (1989) provide a detailed report on the test program. 6 refs., 5 figs.

Clauss, D.B.; Parks, M.B.; Julien, J.T.; Peters, S.W.

1989-01-01T23:59:59.000Z

104

Assessment of CRBR core disruptive accident energetics  

Science Conference Proceedings (OSTI)

The results of an independent assessment of core disruptive accident energetics for the Clinch River Breeder Reactor are presented in this document. This assessment was performed for the Nuclear Regulatory Commission under the direction of the CRBR Program Office within the Office of Nuclear Reactor Regulation. It considered in detail the accident behavior for three accident initiators that are representative of three different classes of events; unprotected loss of flow, unprotected reactivity insertion, and protected loss of heat sink. The primary system's energetics accommodation capability was realistically, yet conservatively, determined in terms of core events. This accommodation capability was found to be equivalent to an isentropic work potential for expansion to one atmosphere of 2550 MJ or a ramp rate of about 200 $/s applied to a classical two-phase disassembly.

Theofanous, T.G.; Bell, C.R.

1984-03-01T23:59:59.000Z

105

A Review of Criticality Accidents 2000 Revision  

SciTech Connect

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.

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

2000-05-01T23:59:59.000Z

106

Study of Safety Auditors' Views on the Use of BIM for Safety in Hong Kong  

Science Conference Proceedings (OSTI)

Traditionally site safety is a concern for the Architectural, Engineering & Construction AEC industry. In view of government of Hong Kong initiating a number of large scale AEC projects and a relative high number of serious accidents continue to ... Keywords: Architectural Engineering & Construction AEC, Building Information Model BIM, Hong Kong, Safety Management System SMS, Sustainable Development SD

Allen Wan, Andrew Platten, Tim Briggs

2013-01-01T23:59:59.000Z

107

LESSONS LEARNED FROM A RECENT LASER ACCIDENT  

SciTech Connect

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.

Woods, Michael; /SLAC

2011-01-26T23:59:59.000Z

108

Final safety analysis report for the Galileo Mission: Volume 2, Book 2: Accident model document: Appendices  

Science Conference Proceedings (OSTI)

This section of the Accident Model Document (AMD) presents the appendices which describe the various analyses that have been conducted for use in the Galileo Final Safety Analysis Report II, Volume II. Included in these appendices are the approaches, techniques, conditions and assumptions used in the development of the analytical models plus the detailed results of the analyses. Also included in these appendices are summaries of the accidents and their associated probabilities and environment models taken from the Shuttle Data Book (NSTS-08116), plus summaries of the several segments of the recent GPHS safety test program. The information presented in these appendices is used in Section 3.0 of the AMD to develop the Failure/Abort Sequence Trees (FASTs) and to determine the fuel releases (source terms) resulting from the potential Space Shuttle/IUS accidents throughout the missions.

Not Available

1988-12-15T23:59:59.000Z

109

Analysis of Reactivity Induced Accident for Control Rods Ejection with Loss of Cooling  

E-Print Network (OSTI)

Understanding of the time-dependent behavior of the neutron population in nuclear reactor in response to either a planned or unplanned change in the reactor conditions, is a great importance to the safe and reliable operation of the reactor. In the present work, the point kinetics equations are solved numerically using stiffness confinement method (SCM). The solution is applied to the kinetics equations in the presence of different types of reactivities and is compared with different analytical solutions. This method is also used to analyze reactivity induced accidents in two reactors. The first reactor is fueled by uranium and the second is fueled by plutonium. This analysis presents the effect of negative temperature feedback with the addition positive reactivity of control rods to overcome the occurrence of control rod ejection accident and damaging of the reactor. Both power and temperature pulse following the reactivity- initiated accidents are calculated. The results are compared with previous works and...

Saad, Hend Mohammed El Sayed; Wahab, Moustafa Aziz Abd El

2013-01-01T23:59:59.000Z

110

Analysis of Reactivity Induced Accident for Control Rods Ejection with Loss of Cooling  

E-Print Network (OSTI)

Understanding of the time-dependent behavior of the neutron population in nuclear reactor in response to either a planned or unplanned change in the reactor conditions, is a great importance to the safe and reliable operation of the reactor. In the present work, the point kinetics equations are solved numerically using stiffness confinement method (SCM). The solution is applied to the kinetics equations in the presence of different types of reactivities and is compared with different analytical solutions. This method is also used to analyze reactivity induced accidents in two reactors. The first reactor is fueled by uranium and the second is fueled by plutonium. This analysis presents the effect of negative temperature feedback with the addition positive reactivity of control rods to overcome the occurrence of control rod ejection accident and damaging of the reactor. Both power and temperature pulse following the reactivity- initiated accidents are calculated. The results are compared with previous works and satisfactory agreement is found.

Hend Mohammed El Sayed Saad; Hesham Mohammed Mohammed Mansour; Moustafa Aziz Abd El Wahab

2013-06-05T23:59:59.000Z

111

Markov Model of Accident Progression at Fukushima Daiichi  

DOE Green Energy (OSTI)

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

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

2012-11-11T23:59:59.000Z

112

Assessment of ICARE/CATHARE V1 Severe Accident Code  

SciTech Connect

The ICARE/CATHARE code system has been developed by the French 'Institut de Radioprotection et de Surete Nucleaire' (IRSN) in the last decade for the detailed evaluation of Severe Accident (SA) consequences in a primary system. It is composed of the coupling of the core degradation IRSN code ICARE2 and of the thermal-hydraulics French code CATHARE2. It has been extensively used to support the level 2 Probabilistic Safety Assessment (PSA-2) of the 900 MWe PWR. This paper presents the synthesis of the ICARE/CATHARE V1 assessment which was conducted in the frame of the 'International ICARE/CATHARE Users' Club', under the management of IRSN. The ICARE/CATHARE V1 validation matrix is composed of more than 60 experiments, distributed in few thermal-hydraulics non-regression tests (to handle the front end phase of a severe accident), numerous Separate-Effect Tests, about 30 Integral Tests covering both the early and the late degradation phases, as well as a 'circuit' experiment including hydraulics loops. Finally, the simulation of the TMI-2 accident was also added to assess the code against real conditions. This validation task was aimed at assessing the ICARE/CATHARE V1 capabilities (including the stand-alone ICARE2 V3mod1 version) and also at proposing recommendations for an optimal use of this version ('Users' Guidelines'). Thus, with a correct account for the recommended guidelines, it appeared that the last ICARE/CATHARE V1 version could be reasonably used to perform best-estimate reactor studies up to a large corium slumping into the lower head. (authors)

Chatelard, Patrick; Fleurot, Joelle; Marchand, Olivier; Drai, Patrick [IRSN, 31, avenue de la Division Leclerc, BP 17 - 92262 Fontenay-aux-Roses Cedex (France)

2006-07-01T23:59:59.000Z

113

Taking the long view  

NLE Websites -- All DOE Office Websites (Extended Search)

Taking the long view Taking the long view Taking the long view on environmental stewardship A newly articulated mission for environmental stewardship at the Laboratory can be summed up in a simple phrase: clean up the past, control current operations, and create a sustainable future. March 20, 2012 Los Alamos Aerial Aerial view of a canyon in Los Alamos, New Mexico. Contact Environmental Communication & Public Outreach P.O. Box 1663 MS M996 Los Alamos, NM 87545 (505) 667-0216 Email "The future viability of the Lab hinges on demonstrating to public that we protect human health and the environment." Environmental stewardship strategy looks 50 years into the future As a way of integrating environmental protection activities into a comprehensive strategy, Kevin Smith, manager of the U.S. Department of

114

A View from Elsewhere  

E-Print Network (OSTI)

Arcadia Project Review - A View From Elsewhere Tony Hirst Department of Communication and Systems, The Open University a.j.hirst@open.ac.uk Summary This report provides an overview of activities carried out during Michaelmas Term...

Hirst, Tony

2009-01-01T23:59:59.000Z

115

Accident Investigation of the Fall Injury at the Savannah River...  

NLE Websites -- All DOE Office Websites (Extended Search)

U.S. Department of Energy Office of Environmental Management Accident Investigation Report Fall Injury Accident at the Savannah River Site on July 1, 2011 August 8, 2011 Disclaimer...

116

DOE-ID FOIA Type A Accident Investigation Board Report - July 28, 1998  

NLE Websites -- All DOE Office Websites (Extended Search)

Electronic Reading Room Documents Electronic Reading Room Documents Freedom of Information Act (FOIA) More on FOIA/Privacy Act Electronic Request Form Frequently Ask Questions Frequently Requested Documents Guides Helpful Links Points of Contact and Privacy Act Advisory Privacy Act Reading Room DOE-ID Public Reading Room Research Library You are here: DOE-ID Home > FOIA > Type A Accident Investigation Board Report - July 28, 1998 Type A Accident Investigation Board Report - July 28, 1998 Fatality and Multiple Injuries Resulting From Release of Carbon Dioxide at Building 648, Test Reactor Area Idaho National Engineering and Environmental Laboratory Files are provided in Adobe Acrobat format. If you experience problems reading these files please download the latest Adobe Acrobat reader at http://get.adobe.com/reader/ Download Acrobat Reader . Adobe Acrobat file Reader will allow you to view, navigate, and print these PDF files on all major computing platforms.

117

Technical evaluation: 300 Area steam line valve accident  

SciTech Connect

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

Not Available

1993-08-01T23:59:59.000Z

118

Uncertainty Assessments in Severe Nuclear Accident Scenarios  

Science Conference Proceedings (OSTI)

Managing uncertainties in industrial systems is a daily challenge to ensure improved design, robust operation, accountable performance and responsive risk control. This paper aims to illustrate the different depth analyses that the uncertainty software ... Keywords: Monte Carlo simulation, nuclear power plant, sensitivity analysis, severe accident, uncertainty

Bertrand Iooss; Fabrice Gaudier; Michel Marques; Bertrand Spindler; Bruno Tourniaire

2009-09-01T23:59:59.000Z

119

Blasting practices and explosives accidents in Utah coal mines  

SciTech Connect

Practices in use in Utah are commended and accidents incident to blasting are reviewed with suggestions as to future avoidance.

Parker, D.J.

1935-01-01T23:59:59.000Z

120

RECENT LASER ACCIDENTS AT DEPARTMENT OF ENERGY LABORATORIES  

SciTech Connect

Recent laser accidents and incidents at research laboratories across the Department of Energy complex are reviewed in this paper. Factors that contributed to the accidents are examined. Conclusions drawn from the accident reports are summarized and compared. Control measures that could have been implemented to prevent the accidents will be summarized and compared. Recommendations for improving laser safety programs are outlined and progress toward achieving them are summarized.

ODOM, CONNON R. [Los Alamos National Laboratory

2007-02-02T23:59:59.000Z

Note: This page contains sample records for the topic "accident conditions view" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

Accident Investigation and Materials Failure Analysis at the ...  

Science Conference Proceedings (OSTI)

Both are independent federal agencies charged with investigating transportation accidents in all modes, including aviation, railroad, highway, marine, pipeline,...

122

DOE O 225.1B, Accident Investigations  

Directives, Delegations, and Requirements

This Order prescribes organizational responsibilities, authorities, and requirements for conducting investigations of certain accidents occurring at DOE sites, ...

2011-03-04T23:59:59.000Z

123

Advanced sodium fast reactor accident source terms : research needs.  

Science Conference Proceedings (OSTI)

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

Powers, Dana Auburn; Clement, Bernard [IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France] IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France; Denning, Richard [Ohio State University, Columbus, OH] Ohio State University, Columbus, OH; Ohno, Shuji [Japan Atomic Energy Agency, Ibaraki, Japan] Japan Atomic Energy Agency, Ibaraki, Japan; Zeyen, Roland [Institute for Energy Petten, Saint-Paul-lez-Durance, France] Institute for Energy Petten, Saint-Paul-lez-Durance, France

2010-09-01T23:59:59.000Z

124

Advanced sodium fast reactor accident source terms : research needs.  

SciTech Connect

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

Powers, Dana Auburn; Clement, Bernard [IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France] IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France; Denning, Richard [Ohio State University, Columbus, OH] Ohio State University, Columbus, OH; Ohno, Shuji [Japan Atomic Energy Agency, Ibaraki, Japan] Japan Atomic Energy Agency, Ibaraki, Japan; Zeyen, Roland [Institute for Energy Petten, Saint-Paul-lez-Durance, France] Institute for Energy Petten, Saint-Paul-lez-Durance, France

2010-09-01T23:59:59.000Z

125

Retrospection of Chernobyl nuclear accident for decision analysis concerning remedial actions in Ukraine  

SciTech Connect

It is considered the efficacy of decisions concerning remedial actions when of-site radiological monitoring in the early and (or) in the intermediate phases was absent or was not informative. There are examples of such situations in the former Soviet Union where many people have been exposed: releases of radioactive materials from 'Krasnoyarsk-26' into Enisey River, releases of radioactive materials from 'Chelabinsk-65' (the Kishtim accident), nuclear tests at the Semipalatinsk Test Site, the Chernobyl nuclear accident etc. If monitoring in the early and (or) in the intermediate phases is absent the decisions concerning remedial actions are usually developed on the base of permanent monitoring. However decisions of this kind may be essentially erroneous. For these cases it is proposed to make retrospection of radiological data of the early and intermediate phases of nuclear accident and to project decisions concerning remedial actions on the base of both retrospective data and permanent monitoring data. In this Report the indicated problem is considered by the example of the Chernobyl accident for Ukraine. Their of-site radiological monitoring in the early and intermediate phases was unsatisfactory. In particular, the pasture-cow-milk monitoring had not been made. All official decisions concerning dose estimations had been made on the base of measurements of {sup 137}Cs in body (40 measurements in 135 days and 55 measurements in 229 days after the Chernobyl accident). For the retrospection of radiological data of the Chernobyl accident dynamic model has been developed. This model has structure similar to the structure of Pathway model and Farmland model. Parameters of the developed model have been identified for agricultural conditions of Russia and Ukraine. By means of this model dynamics of 20 radionuclides in pathways and dynamics of doses have been estimated for the early, intermediate and late phases of the Chernobyl accident. The main results are following: - During the first year after the Chernobyl accident 75-93% of Commitment Effective Dose had been formed; - During the first year after the Chernobyl accident 85-90% of damage from radiation exposure had been formed. During the next 50 years (the late phase of accident) only 10-15% of damage from radiation exposure will have been formed; - Remedial actions (agricultural remedial actions as most effective) in Ukraine are intended for reduction of the damage from consumption of production which is contaminated in the late phase of accident. I.e. agricultural remedial actions have been intended for minimization only 10 % of the total damage from radiation exposure; - Medical countermeasures can minimize radiation exposure damage by an order of magnitude greater than agricultural countermeasures. - Thus, retrospection of nuclear accident has essentially changed type of remedial actions and has given a chance to increase effectiveness of spending by an order of magnitude. This example illustrates that in order to optimize remedial actions it is required to use data of retrospection of nuclear accidents in all cases when monitoring in the early and (or) intermediate phases is unsatisfactory. (author)

Georgievskiy, Vladimir [Russian Research Center 'Kurchatov Insitute', Kurchatov Sq., 1, 123182 Moscow (Russian Federation)

2007-07-01T23:59:59.000Z

126

Evaluation of LLNL's Nuclear Accident Dosimeters at the CALIBAN Reactor September 2010  

Science Conference Proceedings (OSTI)

The Lawrence Livermore National Laboratory uses neutron activation elements in a Panasonic TLD holder as a personnel nuclear accident dosimeter (PNAD). The LLNL PNAD has periodically been tested using a Cf-252 neutron source, however until 2009, it was more than 25 years since the PNAD has been tested against a source of neutrons that arise from a reactor generated neutron spectrum that simulates a criticality. In October 2009, LLNL participated in an intercomparison of nuclear accident dosimeters at the CEA Valduc Silene reactor (Hickman, et.al. 2010). In September 2010, LLNL participated in a second intercomparison of nuclear accident dosimeters at CEA Valduc. The reactor generated neutron irradiations for the 2010 exercise were performed at the Caliban reactor. The Caliban results are described in this report. The procedure for measuring the nuclear accident dosimeters in the event of an accident has a solid foundation based on many experimental results and comparisons. The entire process, from receiving the activated NADs to collecting and storing them after counting was executed successfully in a field based operation. Under normal conditions at LLNL, detectors are ready and available 24/7 to perform the necessary measurement of nuclear accident components. Likewise LLNL maintains processing laboratories that are separated from the areas where measurements occur, but contained within the same facility for easy movement from processing area to measurement area. In the event of a loss of LLNL permanent facilities, the Caliban and previous Silene exercises have demonstrated that LLNL can establish field operations that will very good nuclear accident dosimetry results. There are still several aspects of LLNL's nuclear accident dosimetry program that have not been tested or confirmed. For instance, LLNL's method for using of biological samples (blood and hair) has not been verified since the method was first developed in the 1980's. Because LLNL and the other DOE participants were limited in what they were allowed to do at the Caliban and Silene exercises and testing of various elements of the nuclear accident dosimetry programs cannot always be performed as guests at other sites, it has become evident that DOE needs its own capability to test nuclear accident dosimeters. Angular dependence determination and correction factors for NADs desperately need testing as well as more evaluation regarding the correct determination of gamma doses. It will be critical to properly design any testing facility so that the necessary experiments can be performed by DOE laboratories as well as guest laboratories. Alternate methods of dose assessment such as using various metals commonly found in pockets and clothing have yet to be evaluated. The DOE is planning to utilize the Godiva or Flattop reactor for testing nuclear accident dosimeters. LLNL has been assigned the primary operational authority for such testing. Proper testing of nuclear accident dosimeters will require highly specific characterization of the pulse fields. Just as important as the characterization of the pulsed fields will be the design of facilities used to process the NADs. Appropriate facilities will be needed to allow for early access to dosimeters to test and develop quick sorting techniques. These facilities will need appropriate laboratory preparation space and an area for measurements. Finally, such a facility will allow greater numbers of LLNL and DOE laboratory personnel to train on the processing and interpretation of nuclear accident dosimeters and results. Until this facility is fully operational for test purposes, DOE laboratories may need to continue periodic testing as guests of other reactor facilities such as Silene and Caliban.

Hickman, D P; Wysong, A R; Heinrichs, D P; Wong, C T; Merritt, M J; Topper, J D; Gressmann, F A; Madden, D J

2011-06-21T23:59:59.000Z

127

Prarie View RDF  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

PRAIRIE VIEW RDF PRAIRIE VIEW RDF 2 Prairie View RDF  Located at JAAP (approx. 40 miles southwest of Chicago), 223 acres on 455 Acre Parcel  Will County Owner; Waste Management, Operator  Maximum 23-Year Life WM/Will County Methane to Energy Plant  Landfill Contract Signed w/WM in 1997 w/Gas-to- Energy Plant Clause  County Retains Gas Rights &WM Installs Gas Collection System  WM owns Methane to Energy Plant &Tax Credits 3 CONTRACT PHASE  DOE Grant Applied 6/09  County Board Approves DOE EECBG Strategy 11/09  1 Million DOE Funds to Methane to Energy Plant  Schiff Harden Hired to Negotiate Gas to Energy Contract  February 2010 County Board Authorizes Contract Execution 4 5 METHANE TO ENERGY PLANT DETAILS  Waste Management Required To

128

REAC/TS Radiation Accident Registry: An Overview  

Science Conference Proceedings (OSTI)

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

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

2012-12-12T23:59:59.000Z

129

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

NLE Websites -- All DOE Office Websites (Extended Search)

Monthly Accident Statistics Monthly Accident Statistics Latest Accident Statistics Accident Statistics (through December 2013) Archived Accident Statistics 2013 Through November Through October Through September Through August Through July Through June Through May Through April Through March Through February Through January 2012 Through December Through November Through October Through September Through August Through July Through June Through May Through February Through January 2011 Through December Through November Through October Through September Through August Through July Through June Through May Through April Through March Through February Through January 2010 Through December Through November Through October Through September Through August Through July Through June Through May Through April Through March Through February

130

Trends status: Post-accident fission product chemistry  

DOE Green Energy (OSTI)

It is important to understand and model the chemical and physical behavior of vapor iodine species in containment environments for the following reasons: This behavior can contribute significantly to severe accident source terms; the development of accident mitigation or management strategies (e.g., an effective filter system); for long-term clean-up and recovery following an accident; regulatory requirements (e.g., spray or pool additives); and design basis accidents (i.e., steam generator tube rupture). This document discusses the Oak Ridge National Laboratory ''Post-Accident'' Chemistry Program.

Kress, T.S.; Beahm, E.C.; Shockley, W.C.; Weber, C.F.

1988-01-01T23:59:59.000Z

131

Accident, Maryland: Energy Resources | Open Energy Information  

Open Energy Info (EERE)

Accident, Maryland: Energy Resources Accident, Maryland: Energy Resources Jump to: navigation, search Equivalent URI DBpedia Coordinates 39.628696°, -79.319759° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":39.628696,"lon":-79.319759,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

132

Characterization of a nuclear accident dosimeter  

E-Print Network (OSTI)

The 23rd nuclear accident dosimetry intercomparison was held during the week of June 12-16, 1995 at Los Alamos National Laboratory. This report presents the results of this event, referred to as NAD 23, as related to the performance of Sandia National Laboratories' (SNL) personal nuclear accident dosimeter (PNAD). Two separate critical assemblies, SHEBA and Godiva, were used to generate seven separate neutron spectra for use in dose comparisons. SNL's PNAD measured absorbed doses that were within +16 to +26 percent of the reference doses. In addition, a preliminary investigation was undertaken to determine the feasibility of using the data obtained from an irradiated PNAD to correct for body orientation. This portion of the experiment was performed with a TRIGA reactor at the Nuclear Science Center at Texas A&M University.

Burrows, Ronald Allen

1995-01-01T23:59:59.000Z

133

Characterization of a nuclear accident dosimeter  

Science Conference Proceedings (OSTI)

The 23rd nuclear accident dosimetry intercomparison was held during the week of June 12--16, 1995 at Los Alamos National Laboratory. This report presents the results of this event, referred to as NAD 23, as related to the performance of Sandia National Laboratories (SNL) personal nuclear accident dosimeter (PNAD). Two separate critical assemblies, SHEBA and Godiva, were used to generate seven separate neutron spectra for use in dose comparisons. SNL`s PNAD measured absorbed doses that were within +16 to +26% of the reference doses. In addition, a preliminary investigation was undertaken to determine the feasibility of using the data obtained from an irradiated PNAD to correct for body orientation. This portion of the experiment was performed with a TRIGA reactor at the Nuclear Science Center at Texas A and M University.

Burrows, R.A.

1995-12-01T23:59:59.000Z

134

US Department of Energy Chernobyl accident bibliography  

SciTech Connect

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.

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

1992-04-01T23:59:59.000Z

135

An overview of severe accident modeling and analysis work for the ANS reactor conceptual safety analysis report  

Science Conference Proceedings (OSTI)

ORNL`s Advanced Neutron Source (ANS) will be a new user facility for all kinds of neutron research, centered around a research reactor of unprecedented neutron beam flux. A defense-in-depth philosophy has been adopted. In response to this commitment, ANS Project management has initiated severe accident analysis and related technology development efforts early-on in the design phase itself. Early consideration of severe accident issues will aid in designing a sufficiently robust containment for retention and controlled release of radionuclides in the event of such an accident. It will also provide a means for satisfying on- and off-site regulatory requirements and provide containment response and source term analyses for level-2 and -3 Probabilistic Risk Analyses (PRAs) that will be produced. Moreover, it will provide the best possible understanding of the ANS under severe accident conditions, and consequently provide insights for the development of strategies and design philosophies for accident management, mitigation, and emergency preparedness. This paper presents a perspective overview of the severe accident modeling and analysis work for the ANS Conceptual Safety Analysis Report (CSAR)

Taleyarkhan, R.P.

1992-12-31T23:59:59.000Z

136

Analysis of Kuosheng Large-Break Loss-of-Coolant Accident with MELCOR 1.8.4  

SciTech Connect

The MELCOR code, developed by Sandia National Laboratories, is capable of simulating the severe accident phenomena of light water reactor nuclear power plants (NPPs). A specific large-break loss-of-coolant accident (LOCA) for Kuosheng NPP is simulated with the use of the MELCOR 1.8.4 code. This accident is induced by a double-ended guillotine break of one of the recirculation pipes concurrent with complete failure of the emergency core cooling system. The MELCOR input deck for the Kuosheng NPP is established based on the design data of the Kuosheng NPP and the MELCOR users' guides. The initial steady-state conditions are generated with a developed self-initialization algorithm. The effect of the MELCOR 1.8.4-provided initialization process is demonstrated. The main severe accident phenomena and the corresponding fission product released fractions associated with the large-break LOCA sequences are simulated. The MELCOR 1.8.4 predicts a longer time interval between the core collapse and vessel failure and a higher source term. This MELCOR 1.8.4 input deck will be applied to the probabilistic risk assessment, the severe accident analysis, and the severe accident management study of the Kuosheng NPP in the near future.

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

2000-09-15T23:59:59.000Z

137

Fukushima Daiichi Accident -- Technical Causal Factor Analysis  

Science Conference Proceedings (OSTI)

On March 11, 2011, the Fukushima Daiichi nuclear power plant experienced a seismic event and subsequent tsunami. The accident and the ensuing mitigation and recovery activities occurred over several days, involved a number of incidents, and might provide several opportunities for lessons learned. The objective of this report is to determine the fundamental causative factors for the loss of critical systems at the Fukushima Daiichi reactors that resulted in core damage and subsequent radioactive release. ...

2012-03-27T23:59:59.000Z

138

ParaView at NERSC  

NLE Websites -- All DOE Office Websites (Extended Search)

ParaView ParaView ParaView Introduction ParaView is an open-source, multi-platform data analysis and visualization application. ParaView users can quickly build visualizations to analyze their data using qualitative and quantitative techniques. The data exploration can be done interactively in 3D or programmatically using ParaView's batch processing capabilities. ParaView was developed to analyze extremely large datasets using distributed memory computing resources. It can be run on supercomputers to analyze datasets of terascale as well as on laptops for smaller data. Remote Visualization with ParaView ParaView is a client-server application. The ParaView client (or simply paraview) will run on your desktop while the server will run at the remote supercomputing site. The following describes the steps you will take to

139

Stereoscopic optical viewing system  

DOE Patents (OSTI)

An improved optical system which provides the operator with a stereoscopic viewing field and depth of vision, particularly suitable for use in various machines such as electron or laser beam welding and drilling machines. The system features two separate but independently controlled optical viewing assemblies from the eyepiece to a spot directly above the working surface. Each optical assembly comprises a combination of eye pieces, turning prisms, telephoto lenses for providing magnification, achromatic imaging relay lenses and final stage pentagonal turning prisms. Adjustment for variations in distance from the turning prisms to the workpiece, necessitated by varying part sizes and configurations and by the operator's visual accuity, is provided separately for each optical assembly by means of separate manual controls at the operator console or within easy reach of the operator.

Tallman, C.S.

1986-05-02T23:59:59.000Z

140

A review of the technical issues of air ingression during severe reactor accidents  

Science Conference Proceedings (OSTI)

Severe reactor accident scenarios involving air ingression into the reactor coolant system are described. Evidence from modem reactor accident analyses and from the accident at Three Mile Island show residual fuel will be present in the core region when air ingression is possible. This residual fuel can interact with the air. Exploratory calculations with the MELCOR code of station blackout accidents during shutdown conditions and during operations are used to examine clad oxidation by air and ruthenium release from fuel in air. Extensive ruthenium release is predicted when air ingression rates exceed about 10 moles/s. Past studies of air interactions with irradiated reactor fuel are reviewed. Effects air ingression may have on fission product release, transport, deposition and revaporization are discussed. Perhaps the most important effects of air ingression are expected to be enhanced release of ruthenium from the fuel and the formation of copious amounts of aerosol from uranium oxide vapors. Revaporization of iodine and tellurium retained in the reactor coolant system might be expected.

Powers, D.A.; Kmetyk, L.N.; Schmidt, R.C.

1994-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "accident conditions view" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

Safety Valve Performance Considerations During High-Pressure Station Black-Out Severe Accidents  

Science Conference Proceedings (OSTI)

An assessment of the operating history and test performance of pressurizer safety valves (PSVs) and main steam safety valves (MSSVs) has led to new conclusions on their expected performance during high-pressure station blackout (SBO) severe accident conditions. This report updates conclusions documented in Volume I, focusing on thermal-hydraulic considerations surrounding the reactor coolant system response to an SBO and valve lifts during an SBO event. The report also reconsiders PSV and MSSV tests and ...

1998-01-02T23:59:59.000Z

142

A framework for the assessment of severe accident management strategies  

SciTech Connect

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.

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

1993-09-01T23:59:59.000Z

143

states apps list view | Data.gov  

NLE Websites -- All DOE Office Websites (Extended Search)

states apps list view states apps list view States Data Apps Challenges Policies States You are here Data.gov » Communities » States List View Showing 1 - 5 of 5 results. Resources sort ascending Type Last Updated On TrashBackwards Trash Backwards will connect you with current recycling information, Do It Yourself reuse and upcycle tutorials that are easy to follow, reliable articles and infographics, upcycled products and durable alternatives to common single-use items, and links to non-profits and other groups who will benefit greatly from the stuff you're done with. Currently includes Massachusetts and Washington. Apps 02/27/2013 MoDOT's Traveler Information Map App With the launch of MoDOT's new Traveler Information Map mobile app, Missouri travelers can get the latest information on road conditions, work zones, flooding and incidents on their iPhones and Androids. Mobile 05/22/2013

144

Dynamic View Management System for Query Prediction to View Materialization  

Science Conference Proceedings (OSTI)

On-Line Analytical Processing OLAP systems based on data warehouses are the main systems for managerial decision making and must have a quick response time. Several algorithms have been presented to select the proper set of data and elicit suitable structured ... Keywords: Data Warehousing, Dynamic View Materialization, OLAP, Probabilistic Reasoning Approaches, View Prediction, View Selection

Negin Daneshpour; Ahmad Abdollahzadeh Barfourosh

2011-04-01T23:59:59.000Z

145

Preliminary dose assessment of the Chernobyl accident  

Science Conference Proceedings (OSTI)

From the major accident at Unit 4 of the Chernobyl nuclear power station, a plume of airborne radioactive fission products was initially carried northwesterly toward Poland, thence toward Scandinavia and into Central Europe. Reports of the levels of radioactivity in a variety of media and of external radiation levels were collected in the Department of Energy's Emergency Operations Center and compiled into a data bank. Portions of these and other data which were obtained directly from published and official reports were utilized to make a preliminary assessment of the extent and magnitude of the external dose to individuals downwind from Chernobyl. Radioactive /sup 131/I was the predominant fission product. The time of arrival of the plume and the maximum concentrations of /sup 131/I in air, vegetation and milk and the maximum reported depositions and external radiation levels have been tabulated country by country. A large amount of the total activity in the release was apparently carried to a significant elevation. The data suggest that in areas where rainfall occurred, deposition levels were from ten to one-hundred times those observed in nearby ''dry'' locations. Sufficient spectral data were obtained to establish average release fractions and to establish a reference spectra of the other nuclides in the release. Preliminary calculations indicated that the collective dose equivalent to the population in Scandinavia and Central Europe during the first year after the Chernobyl accident would be about 8 x 10/sup 6/ person-rem. From the Soviet report, it appears that a first year population dose of about 2 x 10/sup 7/ person-rem (2 x 10/sup 5/ Sv) will be received by the population who were downwind of Chernobyl within the U.S.S.R. during the accident and its subsequent releases over the following week. 32 refs., 14 figs., 20 tabs.

Hull, A.P.

1987-01-01T23:59:59.000Z

146

Sec. Herrington Leads Delegation in Response to Chernobyl Accident...  

NLE Websites -- All DOE Office Websites (Extended Search)

Sec. Herrington Leads Delegation in Response to Chernobyl Accident | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering...

147

Median Light Rail Crossing: Accident Causation And Countermeasures  

E-Print Network (OSTI)

Integration of Light Rail Transit Into City Streets. TCRPInfluencing Safety at Highway-Rail Grade Crossings. InK. , W. Hucke and W. Berg. Rail Highway Crossing Accident

Coifman, Benjamin; Bertini, Robert L.

1997-01-01T23:59:59.000Z

148

HEALTH EFFECTS OF THE NUCLEAR ACCIDENT AT THREE MILE ISLAND  

E-Print Network (OSTI)

In) Symposium on Nuclear Reactor Safety: Perspective. Ahealth effects of the nuclear reactor accident at Three Mile50-mile radius of the nuclear reactor site, approximately

Fabrikant, J.I.

2010-01-01T23:59:59.000Z

149

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

SciTech Connect

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

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

2006-07-01T23:59:59.000Z

150

The View from HQ  

National Nuclear Security Administration (NNSA)

A publication of the Office of Advanced Simulation & Computing, NNSA Defense Programs A publication of the Office of Advanced Simulation & Computing, NNSA Defense Programs NA-ASC-500-07-Issue 3 May 2007 The View from HQ by Dimitri Kusnezov I have been spending much of my time these days thinking about science, technology and engineering and the role of the laboratories and how that will be reflected in the Complex of the future. This is on my mind for two reasons: one is my responsibility to

151

Informative Views and Active Recognition  

E-Print Network (OSTI)

Informative Views and Active Recognition Tal Arbel, Frank P. Ferrie, and Peter Whaite TR-CIM-94-7897 Email: cim@cim.mcgill.ca #12;Informative Views and Active Recognition Tal Arbel, Frank P. Ferrie

Dudek, Gregory

152

Protective laser beam viewing device  

SciTech Connect

A protective laser beam viewing system or device including a camera selectively sensitive to laser light wavelengths and a viewing screen receiving images from the laser sensitive camera. According to a preferred embodiment of the invention, the camera is worn on the head of the user or incorporated into a goggle-type viewing display so that it is always aimed at the area of viewing interest to the user and the viewing screen is incorporated into a video display worn as goggles over the eyes of the user.

Neil, George R.; Jordan, Kevin Carl

2012-12-18T23:59:59.000Z

153

The View from HQ  

National Nuclear Security Administration (NNSA)

  NA-ASC-500-07 Issue 2 January 2007 The View from HQ Sitting in airports and planes is risky beyond the obvious dangers now in the news. Uninter- rupted time to think may lead to new ideas. Instinct instructs us that when we hear Wash- ington has some new ideas, the result must be bad. After all, ideas suggest change, which is inherently disruptive. Today the notion of predictivity is on my mind as I am leaving the V&V 2007 meeting in Los Alamos. Predictivity is on my short list of overused, ill-defined words. Washington main- tains a full lexicon of such words-a fair number of which find their way into common usage.

154

False color viewing device  

DOE Patents (OSTI)

This invention consists of a viewing device for observing objects in near-infrared false-color comprising a pair of goggles with one or more filters in the apertures, and pads that engage the face for blocking stray light from the sides so that all light reaching, the user`s eyes come through the filters. The filters attenuate most visible light and pass near-infrared (having wavelengths longer than approximately 700 nm) and a small amount of blue-green and blue-violet (having wavelengths in the 500 to 520 nm and shorter than 435 nm, respectively). The goggles are useful for looking at vegetation to identify different species and for determining the health of the vegetation, and to detect some forms of camouflage.

Kronberg, J.W.

1991-05-08T23:59:59.000Z

155

False color viewing device  

DOE Patents (OSTI)

A viewing device for observing objects in near-infrared false-color comprising a pair of goggles with one or more filters in the apertures, and pads that engage the face for blocking stray light from the sides so that all light reaching the user's eyes come through the filters. The filters attenuate most visible light and pass near-infrared (having wavelengths longer than approximately 700 nm) and a small amount of blue-green and blue-violet (having wavelengths in the 500 to 520 nm and shorter than 435 nm, respectively). The goggles are useful for looking at vegetation to identify different species and for determining the health of the vegetation, and to detect some forms of camouflage.

Kronberg, James W. (108 Independent Blvd., Aiken, SC 29801)

1992-01-01T23:59:59.000Z

156

False color viewing device  

DOE Patents (OSTI)

A viewing device for observing objects in near-infrared false-color comprising a pair of goggles with one or more filters in the apertures, and pads that engage the face for blocking stray light from the sides so that all light reaching the user's eyes come through the filters. The filters attenuate most visible light and pass near-infrared (having wavelengths longer than approximately 700 nm) and a small amount of blue-green and blue-violet (having wavelengths in the 500 to 520 nm and shorter than 435 nm, respectively). The goggles are useful for looking at vegetation to identify different species and for determining the health of the vegetation, and to detect some forms of camouflage. 7 figs.

Kronberg, J.W.

1992-10-20T23:59:59.000Z

157

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

Energy.gov (U.S. Department of Energy (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.

158

Variable selection and ranking for analyzing automobile traffic accident data  

Science Conference Proceedings (OSTI)

Variable ranking and feature selection are important concepts in data mining and machine learning. This paper introduces a new variable ranking technique named Sum Max Gain Ratio (SMGR). The new technique is evaluated within the domain of traffic accident ... Keywords: decision tree, traffic accident data, variable and feature selection, variable ranking

Huanjing Wang; Allen Parrish; Randy K. Smith; Susan Vrbsky

2005-03-01T23:59:59.000Z

159

Evaluation of Accident Frequencies at the Canister Storage Bldg (CSB)  

DOE Green Energy (OSTI)

By using simple frequency calculations and fault tree logic, an evaluation of the design basis accident frequencies at the Canister Storage Building has been performed. The following are the design basis accidents: Mechanical damage of MCO; Gaseous release from the MCO; MCO internal hydrogen deflagration; MCO external hydrogen deflagration; Thermal runaway reactions inside the MCO; and Violation of design temperature criteria.

POWERS, T.B.

2000-03-20T23:59:59.000Z

160

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation  

Science Conference Proceedings (OSTI)

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.

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

2000-03-23T23:59:59.000Z

Note: This page contains sample records for the topic "accident conditions view" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

Assessment of Existing Plant Instrumentation for Severe Accident Management  

Science Conference Proceedings (OSTI)

During an accident, information would be needed for diagnosing a plant's status and confirming its response to mitigative actions. It is important to determine the information necessary for severe accident management and to ensure that this information could be derived from plant instrumentation.

1993-12-01T23:59:59.000Z

162

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

Science Conference Proceedings (OSTI)

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.

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

2007-11-01T23:59:59.000Z

163

A flammability and combustion model for integrated accident analysis. [Advanced light water reactors  

DOE Green Energy (OSTI)

A model for flammability characteristics and combustion of hydrogen and carbon monoxide mixtures is presented for application to severe accident analysis of Advanced Light Water Reactors (ALWR's). Flammability of general mixtures for thermodynamic conditions anticipated during a severe accident is quantified with a new correlation technique applied to data for several fuel and inertant mixtures and using accepted methods for combining these data. Combustion behavior is quantified by a mechanistic model consisting of a continuity and momentum balance for the burned gases, and considering an uncertainty parameter to match the idealized process to experiment. Benchmarks against experiment demonstrate the validity of this approach for a single recommended value of the flame flux multiplier parameter. The models presented here are equally applicable to analysis of current LWR's. 21 refs., 16 figs., 6 tabs.

Plys, M.G.; Astleford, R.D.; Epstein, M. (Fauske and Associates, Inc., Burr Ridge, IL (USA))

1988-01-01T23:59:59.000Z

164

Naval Spent Fuel Rail Shipment Accident Exercise Objectives  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

NAVAL SPENT FUEL RAIL SHIPMENT NAVAL SPENT FUEL RAIL SHIPMENT ACCIDENT EXERCISE OBJECTIVES * Familiarize stakeholders with the Naval spent fuel ACCIDENT EXERCISE OBJECTIVES Familiarize stakeholders with the Naval spent fuel shipping container characteristics and shipping practices * Gain understanding of how the NNPP escorts who accompany the spent fuel shipments will interact with civilian emergency services representatives g y p * Allow civilian emergency services agencies the opportunity to evaluate their response to a pp y p simulated accident * Gain understanding of how the communications links that would be activated in an accident involving a Naval spent fuel shipment would work 1 NTSF May 11 ACCIDENT EXERCISE TYPICAL TIMELINE * Conceptual/Organizational Meeting - April 6 E R T i d it t t d TYPICAL TIMELINE

165

MELCOR accident analysis for ARIES-ACT  

Science Conference Proceedings (OSTI)

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.

Paul W. Humrickhouse; Brad J. Merrill

2012-08-01T23:59:59.000Z

166

Angular dependence of a simple accident dosimeter  

SciTech Connect

A simple dosimeter made of a sulfur tablet, bare and cadmium covered indium foils and a cadmium covered copper foil has been modeled using MCNP5. Studies of the model without phantom or other confounding factors have shown that the cross sections and fluence-to-dose factors generated by the Monte Carlo method agree with those generated by analytic expressions for the high energy component. The threshold cross sections for the detectors on a phantom were calculated. The resulting doses assigned agree well with exposures made to three critical assemblies. In this study the angular dependence on a phantom is studied and compared with measurements taken on the GODIVA reactor. The dosimeter positions on the phantom are facing the source, on the back and the side. In previous papers the modeling of a simple dosimeter made of a sulfur tablet, bare and cadmium covered indium foils and a cadmium covered copper foil has been modeled using MCNP5. The conclusion made was that most of the neutron dose from criticality assemblies results from the high energy neutron fluences determined by the sulfur and indium detectors. The results using doses measured from the GODIVA, SHEBA, and bare and lead shielded SILENE reactors confirmed this. The angular dependence of an accident dosemeter is of interest in evaluating the exposure of personnel. To investigate this effect accident dosemeters were placed on a phantom and exposed to the GODIVA reactor at phantom orientations of 0{sup o}, 45{sup o}, 90{sup o}, 135{sup o}, and 180{sup o} to the assembly center line.

Devine, R. T. (Robert T.); Romero, L. L. (Leonard L.); Olsher, R. H. (Richard H.)

2004-01-01T23:59:59.000Z

167

Accident Performance of Light Water Reactor Cladding Materials  

DOE Green Energy (OSTI)

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

Nelson, Andrew T. [Los Alamos National Laboratory

2012-07-24T23:59:59.000Z

168

Investigation of Strategies for Mitigating Radiological Releases in Severe Accidents  

Science Conference Proceedings (OSTI)

The Fukushima Dai-ichi accident highlights the need to reduce the magnitude of radioactive fission product releases from BWR Mark I and II containments following beyond-design-basis events. There is no evidence that this accident has a long-term effect on public health and safety; however, the Fukushima Dai-ichi accident did result in widespread contamination of surrounding areas, both on-site and off-site. This report assesses various strategies that can be used to maintain BWR Mark I and II ...

2012-09-24T23:59:59.000Z

169

Severe accident sequences analyzed for a two-loop PWR  

Science Conference Proceedings (OSTI)

Different severe accident sequences have been analyzed for a two-loop Westinghouse pressurized water reactor (PWR) using the MELCOR code, version 1.8.4. The purpose of this study was to calculate source terms and the timing of events for severe accident sequences at this type of PWR to be used in the HAS-CAL code .The results calculated by MELCOR have been compared to results from the individual plant examination (IPE) of the Kewaunee nuclear power plant, also a two-loop Westinghouse PWR. The results of the Kewaunee IPE were obtained with the severe accident code MAAP.

Carbajo, J.J. [Oak Ridge National Lab., TN (United States)

1997-12-01T23:59:59.000Z

170

An evaluation of the leakage potential of a personnel airlock subject to severe accident loads  

SciTech Connect

A systematic investigation of the performance of light water reactor containment buildings subject to severe accident loads must include the consideration of leakage between the sealing surfaces of penetrations. As part of its work on containment integrity for the US Nuclear Regulatory Commission (USNRC), Sandia National Laboratories is developing test validated methods for predicting leakage from mechanical penetrations. The primary emphasis has been on large diameter operable penetrations, such as equipment hatches, personnel airlocks, and drywell heads. Several studies conducted for the USNRC have identified leakage from personnel airlocks as a potentially significant failure mode of containment buildings subject to severe accident loads, including Barnes et al. (1984) and Shackelford et al. (1985). Barnes et al (1984) conducted finite element analyses to predict separation of the sealing surfaces on the door and bulkhead, but they did not consider elevated temperature effects and they did not take credit for the performance of the seal material in calculating leak areas. To the author's knowledge, personnel airlock designs with flat bulkhead/door assemblies and seals have never been tested under severe accident conditions, i.e., elevated temperatures and pressure. This paper will describe preliminary analyses and plans for testing a full-size personnel airlock.

Clauss, D.B.

1987-01-01T23:59:59.000Z

171

Informative Views and Sequential Recognition  

E-Print Network (OSTI)

Informative Views and Sequential Recognition Tal Arbel and Frank P. Ferrie TR-CIM-95-10 Nov. 1995 3 Telephone: 514 398-6319 Telex: 05 268510 FAX: 514 398-7348 Email: cim@cim.mcgill.ca #12;Informative Views

Dudek, Gregory

172

Criteria for calculating the efficiency of HEPA filters during and after design basis accidents  

SciTech Connect

We have reviewed the literature on the performance of high efficiency particulate air (HEPA) filters under normal and abnormal conditions to establish criteria for calculating the efficiency of HEPA filters in a DOE nonreactor nuclear facility during and after a Design Basis Accident (DBA). The literature review included the performance of new filters and parameters that may cause deterioration in the filter performance such as filter age, radiation, corrosive chemicals, seismic and rough handling, high temperature, moisture, particle clogging, high air flow and pressure pulses. The deterioration of the filter efficiency depends on the exposure parameters; in severe exposure conditions the filter will be structurally damaged and have a residual efficiency of 0%. Despite the many studies on HEPA filter performance under adverse conditions, there are large gaps and limitations in the data that introduce significant error in the estimates of HEPA filter efficiencies under DBA conditions. Because of this limitation, conservative values of filter efficiency were chosen when there was insufficient data.

Bergman, W. [Lawrence Livermore National Lab., CA (United States); First, M.W. [Harvard School of Public Health, Boston, MA (United States); Anderson, W.L. [Consultant, LaPlata, MD (United States); Gilbert, H. [Consultant, McLean, VA (United States); Jacox, J.W. [Consultant, Columbus, OH (United States)

1994-12-01T23:59:59.000Z

173

MELCOR Simulation of the TMI-2 Severe Accident and Initial Recovery Phases  

Science Conference Proceedings (OSTI)

MELCOR has become the preferred code package within the Swiss nuclear community for severe accident analysis of nuclear power plants, on account of its integrated systems-level approach and validation against experiments and more detailed codes. The present work extends previous MELCOR analysis at PSI from when a site emergency was declared, 18000 s, through to 70000 s, a point where recovery actions were initiated that eventually proved sufficient to restore the reactor to a safe and stable state. It arises out of a programme to assess MELCOR independently using empirical data consistent with the recommendations of the OECD/CSNI validation matrix for core degradation codes. It is the first successful attempt to simulate the whole plant sequence through to the recovery phase. The calculations were performed with code version 1.8.5RD, starting with the model for phases 1 to 4 reported at ICAPP-05. This was extended with a representation of the fission product release and transport pathways, and of the containment, as well as for the extended time period analysed, the so-called phase 5. Reference was made to original sources to obtain the appropriate time-dependent boundary conditions. This paper compares the results of the calculations with observed and deduced data for major accident signatures such as primary system pressure, hot leg temperatures; liquid levels in the vessel and in the pressurizer, and fission product release. The results show that the code can give a credible account of the accident, when reasonable assumptions are made regarding the input where uncertainties exist. The analysis therefore supports the use of the MELCOR-based strategy for severe accident plant transient analysis in Switzerland. Finally, observations are made regarding recent improvements in the code, on which further assessment will concentrate. (authors)

Haste, T.; Birchley, J. [Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Cazzoli, E.; Vitazkova, J. [Cazzoli Consulting, Wiesenweg 14, 5415 Nussbaumen (Switzerland)

2006-07-01T23:59:59.000Z

174

Analysis of Severe Accident Management Strategy for a BWR-4 Nuclear Power Plant  

SciTech Connect

The Chinshan nuclear power plant (NPP) is a Mark-I boiling water reactor (BWR) NPP located in northern Taiwan. The Chinshan NPP severe accident management guidelines (SAMGs) were developed based on the BWR Owners Group Emergency Procedure Guidelines/Severe Accident Guidelines and were developed at the end of 2003. The MAAP4 code has been used as a tool to validate the SAMG strategies. The development process and characteristics of the Chinshan SAMGs are described. The T{sub 5}U{sub t}X{sub C} sequence, the highest core damage frequency in the probabilistic risk assessment insight of the Chinshan NPP, is cited as a reference case for SAMG validation. Not all safety injection systems are operated in the T{sub 5}U{sub t}X{sub C} sequence. The severe accident progression is simulated, and the entry condition of the SAMGs is described. Then, the T{sub 5}U{sub t}X{sub C} sequence is simulated with reactor pressure vessel (RPV) depressurization. Mitigation actions based on the Chinshan NPP SAMGs are then applied to demonstrate the effectiveness of the SAMGs. Sensitivity studies on RPV depressurization with the reactor water level and minimum RPV injection flow rate are also investigated in this study. Based on MAAP4 calculation and the default values of the parameters calculating the severe accident phenomena, the result shows that RPV depressurization before the reactor water level reaches one-fourth of the core water level can prevent the core from damage in the T{sub 5}U{sub t}X{sub C} sequence. The flow rate of two control rod drive pumps is enough to maintain the reactor water level above the top of active fuel and cool down the core in the T{sub 5}U{sub t}X{sub C} sequence without operator action.

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

2005-12-15T23:59:59.000Z

175

Cladding embrittlement during postulated loss-of-coolant accidents.  

DOE Green Energy (OSTI)

The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200 C, ring compression tests were performed to determine post-quench ductility at {le} 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. Among other findings, embrittlement was found to be sensitive to fabrication processes--especially surface finish--but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueled-and-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur.

Billone, M.; Yan, Y.; Burtseva, T.; Daum, R.; Nuclear Engineering Division

2008-07-31T23:59:59.000Z

176

Failsafe : living with man-made disaster and accident  

E-Print Network (OSTI)

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

Higgins, Saoirse, 1966-

2004-01-01T23:59:59.000Z

177

Environment/Health/Safety/Security (EHSS): Report an Accident...  

NLE Websites -- All DOE Office Websites (Extended Search)

Report an Accident or Incident car and foot The law and DOE require prompt notification of all work-related EHS incidentsaccidents. Report all such events immediately to your...

178

Emergency Response to a Transportation Accident Involving Radioactive  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Response to a Transportation Accident Involving Response to a Transportation Accident Involving Radioactive Material Emergency Response to a Transportation Accident Involving Radioactive Material 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 reviewing those points with students. The Student Handout should be distributed to students after the video is shown and the instructor should use the Guide to facilitate a discussion on each response disciplines' activities or duties at the scene. During this discussion, the instructor can present response scenarios, each of which would have a different discipline arriving first at the accident scene. The purpose of this discussion

179

Accidents, engineering and history at NASA: 1967-2003  

E-Print Network (OSTI)

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

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

2009-01-01T23:59:59.000Z

180

FAQ 30-Have there been accidents involving uranium hexafluoride...  

NLE Websites -- All DOE Office Websites (Extended Search)

UF6 was released, which reacted with steam from the process and created HF and uranyl fluoride. This accident resulted in two deaths from HF inhalation and three individuals...

Note: This page contains sample records for the topic "accident conditions view" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


181

HEALTH EFFECTS OF THE NUCLEAR ACCIDENT AT THREE MILE ISLAND  

E-Print Network (OSTI)

within 50 miles of the nuclear power plant was estimated tothe radiation from the nuclear power plant accident. From anand the Peach Bottom nuclear power plants, like the general

Fabrikant, J.I.

2010-01-01T23:59:59.000Z

182

Advanced Steels for Accident Tolerant Fuel Cladding in Commercial ...  

Science Conference Proceedings (OSTI)

... (depending on the LWR system and accident scenario) while maintaining or ... Analysis of the Fragmentation of AlON and Three MgAl2O4 Spinels under...

183

Structural evaluation of electrosleeved tubes under severe accident transients.  

Science Conference Proceedings (OSTI)

A flow stress model was developed for predicting failure of Electrosleeved PWR steam generator tubing under severe accident transients. The Electrosleeve, which is nanocrystalline pure nickel, loses its strength at temperatures greater than 400 C during severe accidents because of grain growth. A grain growth model and the Hall-Petch relationship were used to calculate the loss of flow stress as a function of time and temperature during the accident. Available tensile test data as well as high temperature failure tests on notched Electrosleeved tube specimens were used to derive the basic parameters of the failure model. The model was used to predict the failure temperatures of Electrosleeved tubes with axial cracks in the parent tube during postulated severe accident transients.

Majumdar, S.

1999-11-12T23:59:59.000Z

184

Geometry features measurement of traffic accident for reconstruction based on close-range photogrammetry  

Science Conference Proceedings (OSTI)

This paper studies the feasibility of investigating a traffic accident and offering initial data for traffic accident reconstruction (TAR) using a photogrammetric technique. Compared with the conventional roller tape applied by the traffic police of ... Keywords: Accident reconstruction, Close-range photogrammetry, Direct linear transformation, Traffic accident scene, Vehicle deformation

Xinguang Du; Xianlong Jin; Xiaoyun Zhang; Jie Shen; Xinyi Hou

2009-07-01T23:59:59.000Z

185

Evaluation of Accident Frequencies at the Canister Storage Bldg (CSB)  

DOE Green Energy (OSTI)

By using the fault tree logic, an evaluation of the design basis accident frequencies at the Canister Storage Building has been performed. The evaluation demonstrates that due to low frequency of occurrences, the following design basis accidents are considered not credible (annual frequency of less than 10{sup -6}): Rearrangement of multidster overpack (MCO) internals; Gaseous release from the MCO; MCO internal hydrogen explosion; MCO external hydrogen explosion; Thermal runaway reactions inside the MCO; and Violation of design temperature criteria.

LIU, Y.J.

1999-09-02T23:59:59.000Z

186

Fallen conductor accidents: The challenge to improve safety  

SciTech Connect

What is the worst nightmare of an electric utility manager or engineer Many respond that it is an electrocution resulting from a fallen conductor accident. Few subjects in the operation of an electric utility are more emotional and sobering than this. Traditionally, a utility could do little to prevent such accidents, but some answers from research are emerging, calling for a new look at this old problem.

Aucoin, B.M.; Russell, B.D.

1992-02-01T23:59:59.000Z

187

Review of cladding-coolant interactions during LWR accident transients  

Science Conference Proceedings (OSTI)

Some of the coolant-cladding interactions that can take place during the design basis loss-of-coolant accident and the Three Mile Island loss-of-coolant accident are analyzed. The physical manifestations of the interactions are quite similar, but the time sequences involved can cause very different end results. These results are described and a listing is given of the main research programs that are involved in coolant-cladding interaction research.

Hobson, D.O.

1980-01-01T23:59:59.000Z

188

Trees as Filters of Radioactive Fallout from the Chernobyl Accident  

E-Print Network (OSTI)

This paper is a copy of an unpublished study of the filtering effect of red maple trees (acer rubrum) on fission product fallout near Binghamton, NY, USA following the 1986 Chernobyl accident. The conclusions of this work may offer some insight into what is happening in the forests exposed to fallout from the Fukushima Daiichi Nuclear Plant accident. This posting is in memory of Noel K. Yeh.

Brownridge, James D

2011-01-01T23:59:59.000Z

189

Evaluation of accident frequencies at the canister storage building  

DOE Green Energy (OSTI)

By using the fault tree logic, an evaluation of the design basis accident frequencies at the Canister Storage Building has been performed. The evaluation demonstrates that due to low frequency of occurrences, the following design basis accidents are considered not credible (annual frequency of less than 10{sup -6}): Rearrangement of multi-canister overpack (MCO) internals; Gaseous release from the MCO; MCO internal hydrogen explosion; MCO external hydrogen explosion; Thermal runaway reactions inside the MCO; and Violation of design temperature criteria.

LIU, Y.J.

1999-05-13T23:59:59.000Z

190

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Adequacy of DOE Natural Adequacy of DOE Natural Phenomena Hazards Performance Goals from an Accident Analysis Perspective Jeff Kimball Defense Nuclear Facilities Safety Board Staff Department of Energy NPH Conference October 26, 2011 The views expressed are solely those of the author and no official support or endorsement of this presentation by the Defense Nuclear Facilities Safety Board or the federal government is intended or should be inferred. 1 OBJECTIVE: Assess whether the DOE NPH performance goal concept as used in the Documented Safety Analysis process is adequate or needs additional guidance Background * ANS Standard 2.26 and the concept of Seismic Design Categories (SDC) and Limit States (LS) * ASCE Standard 43-05 and the concept of Design Categories

191

Industry approach to seismic severe accident policy implementation  

Science Conference Proceedings (OSTI)

The Nuclear Regulatory Commission (NRC) issued a severe reactor accident policy for existing plants on August 8, 1985 which describes the formal basis by which the NRC intends to resolve issues related to potential severe reactor accidents. Examination of plant-specific vulnerabilities due to seismic and other externally initiated events was considered on a later schedule and is addressed in Supplement 4 of the NRC Generic Letter No. 88-20 and a NRC guidance document, NUREG-1407, issued in June 1991. This report was prepared to provide a coherent and effective approach for seismic severe accident review which meets the intent of Generic Letter No. 88-20, Supplement 4. The recommendations in this report provide guidance on plant review types and review implementations which is consistent with the limited-scope'' intent of systematic evaluations as described in the NRC's Severe Accident Policy Statement. In addition, to assist in implementing cost-effective modifications that reduce vulnerabilities, this report also presents specific guidelines for identification and treatment of vulnerabilities that may be used as a basis for defining closure of earthquake-related severe-accident issues. This report provides procedural instructions and guidance to support resolution of earthquake-related severe accident issues. More detailed background and technical justifications for the methods are documented elsewhere, and are referenced throughout this report as appropriate.

Reed, J.W. (Benjamin (Jack R.) and Associates, Inc., Mountain View, CA (United States)); O'Hara, T.F.; Jacobson, J.P. (Yankee Atomic Electric Co., Bolton, MA (United States)); Sewell, R.T.; Cornell, C.A. (Risk Engineering, Inc., Golden, CO (United States)); Buttemer, D.R. (Pickard, Lowe and Garrick, Inc., Encinitas, CA (United States)); Schmidt, W.R.; Freed, D.A. (MPR Associates, Inc., Washington, D

1991-11-01T23:59:59.000Z

192

Development of hydrogeological modelling approaches for assessment of consequences of hazardous accidents at nuclear power plants  

SciTech Connect

This paper introduces some modeling approaches for predicting the influence of hazardous accidents at nuclear reactors on groundwater quality. Possible pathways for radioactive releases from nuclear power plants were considered to conceptualize boundary conditions for solving the subsurface radionuclides transport problems. Some approaches to incorporate physical-and-chemical interactions into transport simulators have been developed. The hydrogeological forecasts were based on numerical and semi-analytical scale-dependent models. They have been applied to assess the possible impact of the nuclear power plants designed in Russia on groundwater reservoirs.

Rumynin, V.G.; Mironenko, V.A.; Konosavsky, P.K.; Pereverzeva, S.A. [St. Petersburg Mining Inst. (Russian Federation)

1994-07-01T23:59:59.000Z

193

Significant factors in rail freight accidents: A statistical analysis of predictive and severity indices in the FRA accident/incident data base  

Science Conference Proceedings (OSTI)

The Federal Railroad Association maintains a file of carrier-reported accidents and incidents that meet threshold criteria for damage cost and/or casualties. Using a five year period from this data base, an investigation was conducted into the relationship between quantifiable risk factors and accident frequency and severity. Specific objectives were to identify key variables in accidents, formulate a model to predict future accidents, and assess the relative importance of these variables from the perspective of routing and shipping decision making. The temporal factors YEAR and MONTH were found to be significant predictors of risk; accident severity was greatest for accidents caused by track and roadbed defects. Train speed was an indicator of accident severity; track class and training tonnage were inversely proportional to accident severity. Investigation of the data base is continuing, with a final report expected by late summer. 15 refs., 1 fig., 10 tabs.

Lee, Tze-San; Saricks, C.L.

1991-01-01T23:59:59.000Z

194

Chisolm View | Open Energy Information  

Open Energy Info (EERE)

Chisolm View Chisolm View Jump to: navigation, search Name Chisolm View Facility Chisolm View Sector Wind energy Facility Type Commercial Scale Wind Facility Status In Service Owner GE Energy Financial Service / Enel Green Power North America Developer TradeWind Energy Energy Purchaser Alabama Power Company Location Hunter OK Coordinates 36.59527057°, -97.54501104° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":36.59527057,"lon":-97.54501104,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

195

Evolving Views of Pectin Biosynthesis  

NLE Websites -- All DOE Office Websites (Extended Search)

BIOSYNTHESIS DURING PRIMARY AND SECONDARY WALL FORMATION Multiple lines of evidence have led to a new view of primary wall architecture and the role of pectin therein. Dick-P...

196

Risk View Software Functional Specification  

Science Conference Proceedings (OSTI)

This report defines the functional requirements for a new Risk View software product to be developed as part of the Electric Power Research Institute's (EPRI's) Operations and Maintenance Excellence (OMX) initiative. plant information sources.

2010-09-30T23:59:59.000Z

197

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Evaluation of severe accident risk during mid-loop operations. Volume 6, Part 2: Appendices  

SciTech Connect

The objectives are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation. In phase 2, mid-loop operation was selected as the plant configuration to be analyzed. Volume 1 summarizes the results of the study. The scope of the level-1 study includes plant damage state analyses, and uncertainty analysis. The internal event analysis is documented in Volume 2. The internal fire and internal flood analysis are documented in Volumes 3 and 4, respectively. A separate study on seismic analysis, documented in Volume 5, was performed for the NRC by Future Resources Associated, Inc. A phased approach was used in the level 2/3 PRA program, however both phases addressed the risk from only mid-loop operation. The first phase of the level 2/3 PRA was initiated in late 1991 and consisted of an Abridged Risk Study. This study was completed in May 1992 and was focused on accident progression and consequences, conditional on core damage. Phase 2 is a more detailed study in which an evaluation of risk during mid-loop operation was performed. The results of the phase 2 level 2/3 study are the subject of this volume of NUREG/CR-6144, Volume 6. This report, Volume 6, Part 2, consists of five appendices containing supporting information for: the PDS (plant damage state) analysis; the accident progression analysis; the source term analysis; the consequence analysis; and the Melcor analysis. 73 figs., 21 tabs.

Jo, J.; Lin, C.C.; Neymotin, L.; Mubayi, V. [Brookhaven National Lab., Upton, NY (United States)

1995-05-01T23:59:59.000Z

198

PNNL Results from 2010 CALIBAN Criticality Accident Dosimeter Intercomparison Exercise  

SciTech Connect

This document reports the results of the Hanford personnel nuclear accident dosimeter (PNAD) and fixed nuclear accident dosimeter (FNAD) during a criticality accident dosimeter intercomparison exercise at the CEA Valduc Center on September 20-23, 2010. Pacific Northwest National Laboratory (PNNL) participated in a criticality accident dosimeter intercomparison exercise at the Commissariat a Energie Atomique (CEA) Valduc Center near Dijon, France on September 20-23, 2010. The intercomparison exercise was funded by the U.S. Department of Energy, Nuclear Criticality Safety Program, with Lawrence Livermore National Laboratory as the lead Laboratory. PNNL was one of six invited DOE Laboratory participants. The other participating Laboratories were: Lawrence Livermore National Laboratory (LLNL), Los Alamos National Laboratory (LANL), Savannah River Site (SRS), the Y-12 National Security Complex at Oak Ridge, and Sandia National Laboratory (SNL). The goals of PNNL's participation in the intercomparison exercise were to test and validate the procedures and algorithm currently used for the Hanford personnel nuclear accident dosimeters (PNADs) on the metallic reactor, CALIBAN, to test exposures to PNADs from the side and from behind a phantom, and to test PNADs that were taken from a historical batch of Hanford PNADs that had varying degrees of degradation of the bare indium foil. Similar testing of the PNADs was done on the Valduc SILENE test reactor in 2009 (Hill and Conrady, 2010). The CALIBAN results are reported here.

Hill, Robin L.; Conrady, Matthew M.

2011-10-28T23:59:59.000Z

199

Severe accidents in spent fuel pools in support of generic safety, Issue 82  

SciTech Connect

This investigation provides an assessment of the likelihood and consequences of a severe accident in a spent fuel storage pool - the complete draining of the pool. Potential mechanisms and conditions for failure of the spent fuel, and the subsequent release of the fission products, are identified. Two older PWR and BWR spent fuel storage pool designs are considered based on a preliminary screening study which tried to identify vulnerabilities. Internal and external events and accidents are assessed. Conditions which could lead to failure of the spent fuel Zircaloy cladding as a result of cladding rupture or as a result of a self-sustaining oxidation reaction are presented. Propagation of a cladding fire to older stored fuel assemblies is evaluated. Spent fuel pool fission product inventory is estimated and the releases and consequences for the various cladding scenarios are provided. Possible preventive or mitigative measures are qualitatively evaluated. The uncertainties in the risk estimate are large, and areas where additional evaluations are needed to reduce uncertainty are identified.

Sailor, V.L.; Perkins, K.R.; Weeks, J.R.; Connell, H.R.

1987-07-01T23:59:59.000Z

200

Analysis of Severe Accident Scenarios and Proposals for Safety Improvements for ADS Transmuters with Dedicated Fuel  

SciTech Connect

So-called dedicated fuels will be utilized to obtain maximum transmutation and incineration rates of minor actinides (MAs) in accelerator-driven systems (ADSs). These fuels are characterized by a high-MA content and the lack of the classical fertile materials such as {sup 238}U or {sup 232}Th. Dedicated fuels still have to be developed; however, programs are under way for their fabrication, irradiation, and testing. In Europe, mainly the oxide route is investigated and developed. A dedicated core will contain multiple 'critical' fuel masses, resulting in a certain recriticality potential under core degradation conditions. The use of dedicated fuels may also lead to strong deterioration of the safety parameters of the reactor core, such as, e.g., the void worth, Doppler or the kinetics quantities, neutron generation time, and {beta}{sub eff}. Critical reactors with this kind of fuel might encounter safety problems, especially under severe accident conditions. For ADSs, it is assumed that because of the subcriticality of the system, the poor safety features of such fuels could be coped with. Analyses reveal some safety problems for ADSs with dedicated fuels. Additional inherent and passive safety measures are proposed to achieve the required safety level. A safety strategy along the lines of a defense approach is presented where these measures can be integrated. The ultimate goal of these measures is to eliminate any mechanistic severe accident scenario and the potential for energetics.

Maschek, Werner [Forschungszentrum Karlsruhe Institute for Nuclear and Energy Technologies (Germany); Rineiski, Andrei [Forschungszentrum Karlsruhe Institute for Nuclear and Energy Technologies (Germany); Flad, Michael [Forschungszentrum Karlsruhe Institute for Nuclear and Energy Technologies (Germany); Morita, Koji [Kyushu University Institute of Environmental Systems (Japan); Coste, Pierre [Commissariat a l'Energie Atomique CE Grenoble (France)

2003-02-15T23:59:59.000Z

Note: This page contains sample records for the topic "accident conditions view" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

An application of probabilistic safety assessment methods to model aircraft systems and accidents  

DOE Green Energy (OSTI)

A case study modeling the thrust reverser system (TRS) in the context of the fatal accident of a Boeing 767 is presented to illustrate the application of Probabilistic Safety Assessment methods. A simplified risk model consisting of an event tree with supporting fault trees was developed to represent the progression of the accident, taking into account the interaction between the TRS and the operating crew during the accident, and the findings of the accident investigation. A feasible sequence of events leading to the fatal accident was identified. Several insights about the TRS and the accident were obtained by applying PSA methods. Changes proposed for the TRS also are discussed.

Martinez-Guridi, G.; Hall, R.E.; Fullwood, R.R.

1998-08-01T23:59:59.000Z

202

List View | Data.gov  

NLE Websites -- All DOE Office Websites (Extended Search)

List View List View Safety Data/Tools Apps Challenges Resources Blogs Let's Talk Safety You are here Data.gov » Communities » Safety List View Interactive applications that visually display large datasets provide a portal to explore data and make discoveries. Federal agencies collect information on energy production, use, natural resources, and energy infrastructure logistics and this data can be used to create calculators, interactive maps, and other applications that leverage this data. These applications provide user communities the ability to highlight the energy issues that are occurring within their communities, aid businesses plan and analyze their proposed projects, and provide a baseline for analyzing how energy resources can be most optimally and efficiently used. This page

203

Sec. Herrington Leads Delegation in Response to Chernobyl Accident |  

National Nuclear Security Administration (NNSA)

Sec. Herrington Leads Delegation in Response to Chernobyl Accident | Sec. Herrington Leads Delegation in Response to Chernobyl Accident | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > About Us > Our History > NNSA Timeline > Sec. Herrington Leads Delegation in Response to ... Sec. Herrington Leads Delegation in Response to Chernobyl Accident

204

Accident Investigation at the Idaho National Laboratory Engineering  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

205

A Device for Search of Gamma-Radiation Intensive Sources at the Radiation Accident Condition  

Science Conference Proceedings (OSTI)

The procedure designed for measuring angular distributions of gamma radiation and for search of gamma radiation intensive sources is described. It is based on application of the original multidetector device ShD-1, for measuring an angular distribution in a complete solid angle (4 pi). The calibration results and data on the angular distributions of intensity of gamma radiation at the roof of Chornobyl NPP ''Shelter'' are presented.

Batiy, Valeriy; Klyuchnykov, A; Kochnev, N; Rudko, Vladimir; shcherbin, vladimir; Yegorov, V; Schmieman, Eric A.

2005-08-08T23:59:59.000Z

206

Lighting for remote viewing systems  

SciTech Connect

Scenes viewed by television do not provide the same channels of information for judgment of distances as scenes viewed directly, since television eliminates or degrades several depth perception cues. However, it may be possible to improve depth perception of televised scenes by enhancing the information available through depth cues that are available from lighting. A literature survey and expert opinions were integrated to design a remote lighting arrangement which could enhance depth perception of operators performing remote handling operations. This paper describes the lighting arrangement and discusses some of its advantages and disadvantages. 10 references, 2 figures.

Draper, J.V.

1984-01-01T23:59:59.000Z

207

Lighting for remote viewing systems  

SciTech Connect

Scenes viewed by television do not provide the same channels of information for judgement of distances as scenes viewed directly, since television eliminates or degrades several depth perception cues. However, it may be possible to improve depth perception of televised scenes by enhancing the information available through depth cues that are available from lighting. A literature survey and expert opinions were integrated to design a remote lighting arrangement which could enhance depth perception of operators performing remote handling operations. This paper describes the lighting arrangement and discusses some of its advantages and disadvantages. 10 references, 2 figures.

Draper, J.V.

1984-01-01T23:59:59.000Z

208

Large Scale Verification of External RPV Cooling in Case of Severe Accident  

SciTech Connect

The design of the SWR 1000 - developed by Framatome ANP, consists of components, to flood the exterior of the reactor pressure vessel (RPV) in the case of a hypothetical core melt accident. FANP performed tests to demonstrate that there are significant safety margins against occurring of a critical heat flux (CHF). For this purpose different pretests have been performed in a water / air operated test facility. Within the first pretests the global flow conditions around the RPV have been investigated by measuring the local void fractions with impedance and fiber optical probes. In addition the local water velocities have been measured with a Laser Doppler Anemometer. In further pretests a section model has been implemented in the test facility and the geometry has been modified until the flow conditions in the section model and the global model have been similar. Scaling procedures proved, that the water/ air tests of the section model could be transferred to a water /steam operated heated 1:1 scaled model. Such a model has been manufactured and integrated in the BENSON test-rig - a highly flexible water/steam separate effect test-facility operated by Framatome ANP. The model has been equipped with heating wires pressed in slots in the surface of the model, which represented the RPV- wall. The distance between the slots has been chosen in such a way that the decay heat flux profile, which would have to be removed in a core melt accident, could be simulated. In addition more than 300 thermocouples have been installed on the heated surface to measure the wall temperature and observe whether a CHF has occurred. During the tests no CHF occurred - the tests have been limited only by the test set-up. The tests have been performed at ambient conditions, whereas the pressure under accident conditions would be significantly above these conditions. The critical heat flux increases, with increasing pressure. Considering this effect the proven margin against occurring of a CHF is about four. As the measurements have been limited by the test set-up and as the literature shows that CHF-values of inclined and curved comparable structures are much higher, than the tested heat fluxes it can be expected that the real safety margins against occurring of a CHF are much higher than four. (author)

Schmidt, H. [Framatome ANP GmbH, P.O. Box 3220, D91050 Erlangen (Germany)

2004-07-01T23:59:59.000Z

209

Comparison of fusedose and MACCS2 accident dose codes  

Science Conference Proceedings (OSTI)

The purpose of this paper is to present and document the differences discovered when comparing the two accident dose codes FUSEDOSE and MACCS2. Each code`s methodology is first discussed. With this background, the important comparison parameters are discussed and the resulting differences are presented. It is not the purpose of this paper to draw conclusions as to which code is more reliable but, it is hoped that the data presented will help in deciding upon further actions to be taken, if at all, to improve accident dose calculations. 7 refs., 1 fig., 1 tab.

Sevigny, L.M. [Univ. of California, Berkeley, CA (United States)

1996-12-31T23:59:59.000Z

210

In-vessel Zircaloy oxidation/hydrogen generation behavior during severe accidents  

DOE Green Energy (OSTI)

In-vessel Zircaloy oxidation and hydrogen generation data from various US Nuclear Regulatory Commission severe-fuel damage test programs are presented and compared, where the effects of Zircaloy melting, bundle reconfiguration, and bundle quenching by reflooding are assessed for common findings. The experiments evaluated include fuel bundles incorporating fresh and previously irradiated fuel rods, as well as control rods. Findings indicate that the extent of bundle oxidation is largely controlled by steam supply conditions and that high rates of hydrogen generation continued after melt formation and relocation. Likewise, no retardation of hydrogen generation was noted for experiments which incorporated control rods. Metallographic findings indicate extensive oxidation of once-molten Zircaloy bearing test debris. Such test results indicate no apparent limitations to Zircaloy oxidation for fuel bundles subjected to severe-accident coolant-boiloff conditions. 46 refs., 22 figs., 12 tabs.

Cronenberg, A.W. (Science and Engineering Associates, Inc., Albuquerque, NM (USA))

1990-09-01T23:59:59.000Z

211

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E-Print Network (OSTI)

NSAs have to guarantee that no private information from ...... tive to complementary cell suppression, manuscript, Energy Information Admin- istration

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and algorithms in this area, as well as several application areas where ...... solver reached assignment for proof timeout) found of SAT. Gap. 269. 51. 2. 20. 664.

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1.2 Notation and organization of the paper ... This paper is organized as follows. ..... i is the i-th eigenvalue of C. Then, letting R := QT RQ and H := QT HQ, we.

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May 22, 2003 ... although the corresponding proximity function does not have a ..... Suppose that the statement of the lemma does not hold, i.e., ...... 397413.

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Naive exploration of this neighborhood is computationally ..... Table 2: Problem Sizes for Production Planning Test Instances. In both Table 1 and Table 2, the...

216

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E-Print Network (OSTI)

Note to PractitionersIn practice, production scheduling is a crucial task at ..... the search when the exploration is close to the root, and thus makes cutoff less...

217

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Mar 1, 2007 ... Some of the most successful methods are: Tabu search in [Gho03], a combination of Tabu ... equality constraints are dualized as before, but they also enter the ... term it is possible to avoid fast increase of the Lagrangian dual...

218

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E-Print Network (OSTI)

of image reconstruction from projections in [26], significant and valuable ac- ... ods, which makes them successful in real-world applications, is computa-. 4...

219

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[23]T. S. Motzkin. The arithmetic-geometric inequality. In Inequalities ( Proceedings of Symposium Wright. Patterson Air Force Base, Ohio, 1965), pages 205224.

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an astutely selected small working set does not significantly increase the total number of iterations required to solve the problem, and sometimes even reduces it...

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Introduction. Consider the linear programming problem in the standard form [5, 14] minimize ... row index, say i, has to be determined such that ?j = min {bl/al,j ...

222

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As is known to us, Cognition process is the instinct learning ability of human being, ..... What's more, to alleviate the undesirable effects of estimation error in the...

223

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M? denotes the solar mass. which mimic the corresponding natural processes, ..... Gravitational radiation. In S. W. Hawking and W. Israel, editors,. 300 Years of...

224

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Science Conference Proceedings (OSTI)

Energy Conversion Photovoltaic, Concentrating Solar Power, and Thermoelectric ... Green Technologies for Materials Manufacturing and Processing IV.

225

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Science Conference Proceedings (OSTI)

Feb 15, 2011 ... cost for low power plant ... Fitness for small electricity grids, reduced .... Need a factory to make the price attractive, need an attractive price to.

226

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tioning, clearly identify the utility function that characterizes their attitude toward risk. The use of expected ...... with largest fixed interest rate. Alternatively, our...

227

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to the performance of the exact method of [1]. They are very comparable to those of [8], although have a higher rate of increase with dimension d and/or number...

228

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breakpoint at ?, and given by the two lines L1 and L2. Clearly, if ?L(v) ..... overhead which would be, for example, incurred at every node of a branch- and-

229

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Science Conference Proceedings (OSTI)

availability-manufacturing, processing, fabrication joining, uniformity, balance of properties (what properties are important limiting), qualification, codes, safety...

230

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E-Print Network (OSTI)

Oct 30, 2002 ... Turing machine model, our analysis yields an O(n. 3.5 .... analysis of Vavasis and Ye's algorithm is based on the notion of crossover event,.

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May 12, 2005 ... Our algorithm is a result of the combination of stochastic ap- proximation ideas ... 2EDF R&D ... 3EDF R&D ..... Let us state the main result of this section: ..... IEEE Transactions on Systems, Science and Cybernetics, 5:307. 14.

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1France Tlcom R&D, 38-40 rue du gl Leclerc, 92130 Issy-les-Moulineaux, France ... The results obtained outperform many methods based on earlier literature. ...... Linear programming under uncertainty, Management Science, 1 (

233

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Sep 22, 2006 ... minimizing the cost of the thermal power production if the hydro-power plant cannot supply the demand completely. Dynamic programming is a...

234

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Science Conference Proceedings (OSTI)

Jul 13, 2012 ... Current emphasis: visualization, data mining and applications in oil and ..... Internships. Agreement Enhancements. Discount on indirect costs.

235

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Jan 27, 2011 ... Cost Horizons and Certainty Equivalents: An Approach to Stochastic Programming of Heating Oil. Management Science, 4(3):253. 263, 1958.

236

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Optimization Technical Report. ISyE Department, University of Wisconsin- Madison. Oktay Gnlk1 Jeff Linderoth2. Perspective Reformulations of Mixed Integer.

237

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Science Conference Proceedings (OSTI)

Red Herring Dec 2001. Nanotechnology is the understanding and control of matter at dimensions of roughly 1 to. 100 nanometers, where unique phenomena .

238

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E-Print Network (OSTI)

We tested these formu- ... and finding radiation treatment plans in the presence of organ motion. ...... We report the number instances for which the heuristic.

239

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E-Print Network (OSTI)

Jul 18, 2008 ... Let us define the energy product and corresponding energy norm with respect to an ...... 3.2 Shape optimization in 3D for EIT and DC resistivity.

240

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E-Print Network (OSTI)

2003. The Solver Manuals. GAMS Development Corporation, Washington, DC, USA. Gilbert, J. 2007a. Organization of the Modulopt collection of optimization...

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erwin@gams.com. GAMS Development Corp. Washington DC. 19th June 2001. Abstract. The `Progressive Party Problem' [9] has long been considered a prob-.

242

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Oct 6, 2004... and Systems Engineering, Georgia Institute of Technology, Atlanta, .... The methods described in this paper reduce the energy function (1) at...

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digital communication networks. .... bound. IEEE Transactions on Communications, 52:632642, 2004. ... uscript, Georgia Institute of Technology, 2007.

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It models optimal placement of communications relay nodes in the pres- ence of obstacles. This problem ...... tion Technology (CENIIT). The authors would like to

245

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AbstractWe consider the classical problem of estimating ... (a) to provide an estimate fn of f such that the first 2n+1 .... course at a higher computational cost.

246

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G as a road system, T as petrol stations and a driver who wants to go from u to v. Then B(u, v) is ..... [9] (gas distribution) provide instances where almost all...

247

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Aug 30, 2004 ... (where s?1 denotes the vector of reciprocals of the components of s), so that necessarily x = s?1 > 0, ... be written in the more symmetrical form.

248

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The colon cancer dataset contains 22 normal and 40 colon cancer tissues ..... P. Wolfe, An algorithm for quadratic programming, Naval Research Logistics.

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Oct 14, 2009 ... where A?(y) = ?i yiA(i) maps Rm to Sn, C is the cone of copositive ...... [14] Lovsz, L. (1979), On the Shannon capacity of a graph, IEEE...

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each component is non-zero precisely once in every cycle of n iterations. Thus, ... and these may not inherit the overall rate of the combined pair. Similarly, in.

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[8] J. Eckstein and M. Fukushima, Some reformulation and applications of the alternating ... [10] M. Fukushima, Application of the alternating direction method of

252

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Feb 5, 2002 ... environment (profile) of an amino acid in terms of its contact shells. ..... that in order to use the MaxF heuristic we need to be able to load all.

253

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However, numerical simulations allow us to conjecture ... property of the optimal weights a?0,...,a?n and also an illustrative example of ... Let us define the weighted sum of exponential functions for a ? RN+1 and h ...... be two azimuth angles. ... [2] A. Labeyrie, Resolved imaging of extra-solar planets with future 10-

254

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trajectories, dissipative dynamical systems, Lyapunov analysis, weak .... the link with some classical results concerning semi-groups of contractions generated.

255

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borhood and a curve is built by linking all the points. The coordinates of ..... Learning algorithms could also be used to automatically guide the search through...

256

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E-Print Network (OSTI)

by learning mathematical principles, and subsequent implementation of mathematical ... area under a curve as a sum of areas of rectangles as the width of each...

257

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E-Print Network (OSTI)

Sep 18, 2009 ... telecommunication industry [49] which can be reformulated as a two-stage ...... problem under equilibrium constraints in electricity spot market...

258

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Science Conference Proceedings (OSTI)

Jun 6, 2012 ... fossil fuels and chart a new course on energy in this country, we are condemning future generations to global catastrophe. Senator Obama...

259

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E-Print Network (OSTI)

Oct 19, 2010 ... In aircraft routing, maintenance events replenish aircrafts' ability to fly. In .... with several different label data structure and treatment strategies.

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of an extension of the Fundamental Theorem of Linear Programming, and proved , in ..... with ?v for v = [0,..., 0, 1] ? Rn (it does not matter that Nowak drops the last row and ...... Handbook of Combinatorial Optimization (supp. Vol. A), 174.

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and Boyd (1996), Todd (2001), the SDP handbook edited by Wolkowicz et al. ..... cut polytope CHULL(G) does not admit a polynomial time separation oracle, but ...... problem. In cutting plane algorithms it is of fundamental importance that re-.

262

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E-Print Network (OSTI)

s k C bakc D s k C bas a.s k/c s k C bas .s k/c D s k C as .s k/ D as; the claim follows. Lemma A.2 (Chernoff bounds, see Mitzenmacher and Upfal [17]). Let X1;:

263

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In order to increase the search capability of MBH, a population framework has been proposed ... The key idea is to avoid new individuals to enter the population if someone similar (in a ..... We measure the efficiency in terms of number of (two-.

264

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E-Print Network (OSTI)

These plugins have two essential callbacks that are called .... assigned to opened facilities so that the total sum of connection distances is minimized and the...

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Feb 28, 2012 ... assumed to be essential and the only decision criterion is the travel distance. The portion ... On the other side, customers choose the facilities to.

266

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E-Print Network (OSTI)

Rep., Argonne National Laboratory, Argonne,. Illinois, USA, 1992. [2] A. Borzi and K. Kunisch, A globalization strategy for the multigrid solution of elliptic optimal...

267

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E-Print Network (OSTI)

one additional order of magnitude) by finding the best axis-parallel ellipsoid, which is however too expensive to be done (outside the root node) within our...

268

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E-Print Network (OSTI)

does not apply here because the objective function of LFMP may be unbounded ...... Handbook of Global Optimization, Kluwer Academic Publishers, 495-608.

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E-Print Network (OSTI)

Department of Applied Mathematics, Beijing Jiaotong University, Beijing ... (SDP), plays a fundamental role in mathematical programming, see, e.g., [21, 2, 7]. ...... [ 21] H. Wolkowicz, R. Saigal, L. Vandenberghe: Handbook of Semidefinite...

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E-Print Network (OSTI)

Feb 22, 2002 ... This problem is of fundamental importance in mathematics and physical sciences ... ?Department of Mathematics and Statistics, University of Guelph, Ontario N1G 2W1, Canada. Email: ...... Handbook of Semidefinite Program-.

271

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Science Conference Proceedings (OSTI)

Aug 6, 2010 ... (doi:10.1007/s00134-010-1979-1) contains supplementary material, which is .... The visual inspection of the distribution of the EQ-5D scores...

272

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Science Conference Proceedings (OSTI)

Feb 16, 1998 ... temperatures and quenched by injecting helium gas. All processes ..... retrofitted with mushroom shaped TiB2-G cathode elements and op- erated for ..... same time improving the formability of alloys and reducing the natural aging rate. .... of many magnesium die castings in production vehicles. In order for...

273

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E-Print Network (OSTI)

Jul 3, 2007 ... some cases, give conjectures in a fully automated way). Considering that ..... The input of the program is the definition of a given problem written in a text file. .... For instance, one can get statistics ... To determine S, GraPHedron does not need to access the graphs. ..... on Graph Theory (Novi Sad, 1983), pp.

274

Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents  

DOE Green Energy (OSTI)

Preliminary designs are described for models of hydrogen and oxygen uptake in fuel rod cladding during severe accidents. Calculation of the uptake involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the clad-ding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental results are presented that show a rapid uptake of hydrogen in the event of dissolution of the oxide layer and a rapid release of hydrogen in the event of cracking of the oxide layer. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert's law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for hydrogen and oxygen uptake and cladding embrittlement into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the proposed models on the calculated behavior of fuel rods in severe accident conditions. This report is a revision and reissue of the report entitled; "Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents."

Siefken, Larry James

1999-02-01T23:59:59.000Z

275

Census and viewing of organisms  

NLE Websites -- All DOE Office Websites (Extended Search)

Census and viewing of organisms Census and viewing of organisms Name: m hariaczyi Status: N/A Age: N/A Location: N/A Country: N/A Date: Around 1993 Question: How many organisms exist in the world today? What is the most powerful microscope that could be used for viewing organism? Replies: The most powerful microscope is called an electron microscope, which can be used for viewing entire organisms, although few organisms are small enough to see all of them at high magnifications allowed by this microscope. So most often its used to look at fixed sections of organisms. Since the electron microscope only works in a vacuum, with no air, you cannot look at live organisms. To do that, probably the most powerful microscope is called a Nomarski, or in technical terms, a "differential interference contrast" microscope. This is a modification of a normal light microscope that allows better contrast in living tissue. It is not any more powerful than a light microscope, and is much less powerful than an electron microscope, but it allows you to see living things much better.

276

Bayesian networks: A teacher's view  

Science Conference Proceedings (OSTI)

Teachers viewing Bayesian network-based proficiency estimates from a classroom full of students face a different problem from a tutor looking at one student at a time. Fortunately, individual proficiency estimates can be aggregated into classroom and ... Keywords: Aggregation, Bayesian networks, Computer graphics, Probabilities

Russell G. Almond; Valerie J. Shute; Jody S. Underwood; Juan-Diego Zapata-Rivera

2009-03-01T23:59:59.000Z

277

On the computation of relational view complements  

Science Conference Proceedings (OSTI)

Views as a means to describe parts of a given data collection play an important role in many database applications. In dynamic environments where data is updated, not only information provided by views, but also information provided by data sources yet ... Keywords: Relational algebra, data warehouses, minimal complements, self-maintainability, view complements, views

Jens Lechtenbrger; Gottfried Vossen

2003-06-01T23:59:59.000Z

278

Three dimensional effects in analysis of PWR steam line break accident  

E-Print Network (OSTI)

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

Tsai, Chon-Kwo

279

Report on the Scope of the Accident Investigation of the Tristan...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Report on the Scope of the Accident Investigation of the Tristan Fire at the DOE Brookhaven National Laboratory, IG-0386 Report on the Scope of the Accident Investigation of the...

280

Normal accidents: Data quality problems in ERP-enabled manufacturing  

Science Conference Proceedings (OSTI)

The efficient operation of Enterprise Resource Planning (ERP) systems largely depends on data quality. ERP can improve data quality and information sharing within an organization. It can also pose challenges to data quality. While it is well known that ... Keywords: Data quality, ERP, complexity, enterprise resource planning, normal accident, tight coupling

Lan Cao, Hongwei Zhu

2013-05-01T23:59:59.000Z

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281

Hanford Waste Tank Bump Accident and Consequence Analysis  

Science Conference Proceedings (OSTI)

This report provides a new evaluation of the Hanford tank bump accident analysis and consequences for incorporation into the Authorization Basis. The analysis scope is for the safe storage of waste in its current configuration in single-shell and double-shell tanks.

BRATZEL, D.R.

2000-06-20T23:59:59.000Z

282

Criticality accident alarm system at the Fernald Environmental Management Project  

SciTech Connect

The purpose of this paper is to give a description of the Criticality Accident Alarm System (CAAS) presently installed at the Fernald Environmental Management Project (FEMP) for monitoring areas requiring criticality controls, and some of the concerns associated with the operation of this system. The system at the FEMP is known as the Radiation Detection Alarm (RDA) System.

Marble, R.C.; Brown, T.D.; Wooldridge, J.C.

1994-06-01T23:59:59.000Z

283

THE ANALYSIS OF FATAL ACCIDENTS IN INDIAN D. Sengupta1  

E-Print Network (OSTI)

THE ANALYSIS OF FATAL ACCIDENTS IN INDIAN COAL MINES A. Mandal D. Sengupta1 Indian Statistical of Indian coal mines from April 1989 to March 1998. It is found that Indian mines have considerably higher over 600,000 miners and other workers. Safety in the Indian coal mines is therefore a very important

Mandal, Abhyuday

284

A SUMMARY OF INDUSTRIAL ACCIDENTS IN USAEC FACILITIES  

SciTech Connect

The accident experience of the AEC contractor operation for 1959 and 1960 is reported. Incidents involving radio active materials are described. A table of inadvertent criticality was included to supplement other tables. A tabulation of exposure records at values from 0 to 15 r is given. (M.C.G.)

1962-10-31T23:59:59.000Z

285

Test Data for USEPR Severe Accident Code Validation  

SciTech Connect

This document identifies data that can be used for assessing various models embodied in severe accident analysis codes. Phenomena considered in this document, which were limited to those anticipated to be of interest in assessing severe accidents in the USEPR developed by AREVA, include: Fuel Heatup and Melt Progression Reactor Coolant System (RCS) Thermal Hydraulics In-Vessel Molten Pool Formation and Heat Transfer Fuel/Coolant Interactions during Relocation Debris Heat Loads to the Vessel Vessel Failure Molten Core Concrete Interaction (MCCI) and Reactor Cavity Plug Failure Melt Spreading and Coolability Hydrogen Control Each section of this report discusses one phenomenon of interest to the USEPR. Within each section, an effort is made to describe the phenomenon and identify what data are available modeling it. As noted in this document, models in US accident analysis codes (MAAP, MELCOR, and SCDAP/RELAP5) differ. Where possible, this report identifies previous assessments that illustrate the impact of modeling differences on predicting various phenomena. Finally, recommendations regarding the status of data available for modeling USEPR severe accident phenomena are summarized.

J. L. Rempe

2007-05-01T23:59:59.000Z

286

Getting to necessary and sufficient-developing accident scenarios for risk assessment  

Science Conference Proceedings (OSTI)

This paper presents a simple, systematic approach for developing accident scenarios using generic accident types. Result is a necessary and sufficient set of accident scenarios that can be used to establish the safety envelope for a facility or operation. Us of this approach along with the methodology of SAND95-0320 will yield more consistent accident analyses between facilities and provide a sound basis for allocating limited risk reduction resources.

Mahn, J.A.

1996-05-01T23:59:59.000Z

287

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

DOE Green Energy (OSTI)

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

RITTMANN, P.D.

1999-10-07T23:59:59.000Z

288

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

DOE Green Energy (OSTI)

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

PIEPHO, M.G.

2000-01-10T23:59:59.000Z

289

A SUMMARY OF INDUSTRIAL ACCIDENTS IN USAEC FACILITIES, 1961-1962  

SciTech Connect

Information is presented on accidents andd incidents occurring during 1961 and 1962 in plants owned and operated by the AEC. Revised reporting requirements established by the AEC in April 1962 are outlined. Data are summarized on radiation exposure of AEC contractor personnel, accidents involving radioactive materials, andd accidents involving fatalities. (C.H.)

1964-10-31T23:59:59.000Z

290

Radionuclide release calculations for selected severe accident scenarios  

Science Conference Proceedings (OSTI)

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.

Denning, R.S.; Leonard, M.T.; Cybulskis, P.; Lee, K.W.; Kelly, R.F.; Jordan, H.; Schumacher, P.M.; Curtis, L.A. (Battelle Columbus Div., OH (USA))

1990-08-01T23:59:59.000Z

291

Views of the solar system  

SciTech Connect

Views of the Solar System has been created as an educational tour of the solar system. It contains images and information about the Sun, planets, moons, asteroids and comets found within the solar system. The image processing for many of the images was done by the author. This tour uses hypertext to allow space travel by simply clicking on a desired planet. This causes information and images about the planet to appear on screen. While on a planet page, hyperlinks travel to pages about the moons and other relevant available resources. Unusual terms are linked to and defined in the Glossary page. Statistical information of the planets and satellites can be browsed through lists sorted by name, radius and distance. History of Space Exploration contains information about rocket history, early astronauts, space missions, spacecraft and detailed chronology tables of space exploration. The Table of Contents page has links to all of the various pages within Views Of the Solar System.

Hamilton, C.

1995-02-01T23:59:59.000Z

292

Alternative Fuels Data Center: Natural Gas Safety after a Traffic Accident  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

Natural Gas Safety Natural Gas Safety after a Traffic Accident to someone by E-mail Share Alternative Fuels Data Center: Natural Gas Safety after a Traffic Accident on Facebook Tweet about Alternative Fuels Data Center: Natural Gas Safety after a Traffic Accident on Twitter Bookmark Alternative Fuels Data Center: Natural Gas Safety after a Traffic Accident on Google Bookmark Alternative Fuels Data Center: Natural Gas Safety after a Traffic Accident on Delicious Rank Alternative Fuels Data Center: Natural Gas Safety after a Traffic Accident on Digg Find More places to share Alternative Fuels Data Center: Natural Gas Safety after a Traffic Accident on AddThis.com... More in this section... Natural Gas Basics Benefits & Considerations Stations Vehicles Availability Conversions Emissions

293

Portsmouth Site Plant Surpasses Five Years Without Lost-Time Accident |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Plant Surpasses Five Years Without Lost-Time Plant Surpasses Five Years Without Lost-Time Accident Portsmouth Site Plant Surpasses Five Years Without Lost-Time Accident November 26, 2013 - 12:00pm Addthis BWCS employees from all departments of the DUF6 project at the Portsmouth site come together to mark five years without a lost-time accident. BWCS employees from all departments of the DUF6 project at the Portsmouth site come together to mark five years without a lost-time accident. Russ Hall, environment, safety and health supervisor, changes the DUF6 project sign to mark five years without a lost-time accident. Russ Hall, environment, safety and health supervisor, changes the DUF6 project sign to mark five years without a lost-time accident. BWCS employees from all departments of the DUF6 project at the Portsmouth site come together to mark five years without a lost-time accident.

294

Order Module--DOE Order 225.1B, ACCIDENT INVESTIGATIONS | Department of  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Order 225.1B, ACCIDENT INVESTIGATIONS Order 225.1B, ACCIDENT INVESTIGATIONS Order Module--DOE Order 225.1B, ACCIDENT INVESTIGATIONS DOE O 225.1B prescribes organizational responsibilities, authorities, and requirements for conducting investigations of certain accidents occurring at DOE sites, facilities, areas, operations, and activities. The purpose of the accident investigation is to understand and identify the causes that contributed to the accident so those deficiencies can be addressed and corrected. This, in turn, is intended to prevent recurrence and promote improved environmental protection and safety and health of DOE employees, contractors, and the public. Moreover, accident investigations are used to promote the values and concepts of a learning organization. The department's integrated safety management (ISM) feedback and improvement

295

September 2013 Most Viewed Documents for Fission And Nuclear Technologies |  

Office of Scientific and Technical Information (OSTI)

Fission And Nuclear Technologies Fission And Nuclear Technologies Estimation of gas leak rates through very small orifices and channels. [From sealed PuO/sub 2/ containers under accident conditions] Bomelburg, H.J. (1977) 133 Stress analysis and evaluation of a rectangular pressure vessel. [For equipment for sampling Hanford tank radwaste] Rezvani, M.A.; Ziada, H.H. (Westinghouse Hanford Co., Richland, WA (United States)); Shurrab, M.S. (Westinghouse Savannah River Co., Aiken, SC (United States)) (1992) 78 Graphite design handbook Ho, F.H. (1988) 76 Feed-pump hydraulic performance and design improvement, Phase I: research program design. Final report Brown, W.H.; Gopalakrishnan, S.; Fehlau, R.; Thompson, W.E.; Wilson, D.G. (1982) 69 Flow-induced vibration of circular cylindrical structures Chen, S.S. (1985)

296

Volume II - Accident and Operational Safety Analysis Handbook  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

208-2012 208-2012 July 2012 DOE HANDBOOK Accident and Operational Safety Analysis Volume II: Operational Safety Analysis Techniques U.S. Department of Energy Washington, D.C. 20585 NOT MEASUREMENT SENSITIVE DOE-HDBK-1208-2012 i ACKNOWLEDGEMENTS This Department of Energy (DOE) Accident and Operational Safety Analysis Handbook was prepared under the sponsorship of the DOE Office of Health Safety and Security (HSS), Office of Corporate Safety Programs, and the Energy Facility Contractors Operating Group (EFCOG), Industrial Hygiene and Safety Sub-group of the Environmental Health and Safety Working Group. The preparers would like to gratefully acknowledge the authors whose works are referenced in this document, and the individuals who provided valuable technical insights and/or specific

297

Review of ARAC's involvement in the Titan II missile accident  

SciTech Connect

The Atmospheric Release Advisory Capability (ARAC) response to the Titan II accident near Damascus, Arkansas on 19 September 1980 entailed 12 personnel for periods ranging from 2 to 12 hours. The first call was a NEST Standby alert at 0415L (PCT), followed by a request for dispersal calculations at 0615L, personnel callout at 0630L, crude estimates of plausible source term scenarios at 0845-0900L, first model calculations at 1130L and final model calculations at 1500L. While several new firsts were recorded for ARAC, demonstrating expanded capabilities for NEST-type responses, time lines were very long, essential information was very scant to non-existent, and useful communication of final calculations to the accident site impossible. A detailed chronology is found in Appendix A and a list of acronyms and abbreviations is contained in Appendix B.

Sullivan, T.J.

1980-10-01T23:59:59.000Z

298

Enhanced Accident Tolerant LWR Fuels National Metrics Workshop Report  

SciTech Connect

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

Lori Braase

2013-01-01T23:59:59.000Z

299

Logical queries over views: Decidability and expressiveness  

Science Conference Proceedings (OSTI)

We study the problem of deciding the satisfiability of first-order logic queries over views, with our aim to delimit the boundary between the decidable and the undecidable fragments of this language. Views currently occupy a central place in database ... Keywords: Lwenheim class, Satisfiability, conjunctive query, containment, database query, database view, decidability, first-order logic, monadic logic, ontology reasoning, unary logic, unary view

James Bailey; Guozhu Dong; Anthony WIDJAJA To

2010-01-01T23:59:59.000Z

300

Process hazards analysis (PrHA) program, bridging accident analyses and operational safety  

SciTech Connect

Recently the Final Safety Analysis Report (FSAR) for the Plutonium Facility at Los Alamos National Laboratory, Technical Area 55 (TA-55) was revised and submitted to the US. Department of Energy (DOE). As a part of this effort, over seventy Process Hazards Analyses (PrHAs) were written and/or revised over the six years prior to the FSAR revision. TA-55 is a research, development, and production nuclear facility that primarily supports US. defense and space programs. Nuclear fuels and material research; material recovery, refining and analyses; and the casting, machining and fabrication of plutonium components are some of the activities conducted at TA-35. These operations involve a wide variety of industrial, chemical and nuclear hazards. Operational personnel along with safety analysts work as a team to prepare the PrHA. PrHAs describe the process; identi fy the hazards; and analyze hazards including determining hazard scenarios, their likelihood, and consequences. In addition, the interaction of the process to facility systems, structures and operational specific protective features are part of the PrHA. This information is rolled-up to determine bounding accidents and mitigating systems and structures. Further detailed accident analysis is performed for the bounding accidents and included in the FSAR. The FSAR is part of the Documented Safety Analysis (DSA) that defines the safety envelope for all facility operations in order to protect the worker, the public, and the environment. The DSA is in compliance with the US. Code of Federal Regulations, 10 CFR 830, Nuclear Safety Management and is approved by DOE. The DSA sets forth the bounding conditions necessary for the safe operation for the facility and is essentially a 'license to operate.' Safely of day-to-day operations is based on Hazard Control Plans (HCPs). Hazards are initially identified in the PrI-IA for the specific operation and act as input to the HCP. Specific protective features important to worker safety are incorporated so the worker can readily identify the safety parameters of the their work. System safety tools such as Preliminary Hazard Analysis, What-If Analysis, Hazard and Operability Analysis as well as other techniques as necessary provide the groundwork for both determining bounding conditions for facility safety, operational safety, and day-to-clay worker safety.

Richardson, J. A. (Jeanne A.); McKernan, S. A. (Stuart A.); Vigil, M. J. (Michael J.)

2003-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "accident conditions view" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


301

AQUATIC ASSESSMENT OF THE CHERNOBYL NUCLEAR ACCIDENT AND ITS REMEDIATION  

Science Conference Proceedings (OSTI)

This modeling study evaluated aquatic environment affected by the Chernobyl nuclear accident and the effectiveness of remediation efforts. Study results indicate that radionuclide concentrations in the Pripyat and Dnieper rivers were well above the drinking water limits immediately after the Chernobyl accident, but have decreased significantly in subsequent years due to flashing, burying, and decay. Because high concentrations of 90Sr and 137Cs, the major radionuclides affecting human health through aquatic pathways, are associated with flooding, an earthen dike was constructed along the Pripyat River in its most contaminated floodplain. The dike was successful in reducing the 90Sr influx to the river by half. A 100-m-high movable dome called the New Safe Confinement is planned to cover the Chernobyl Shelter (formally called the sarcophagus) that was erected shortly after the accident. The NSC will reduce radionuclide contamination further in these rivers and nearby groundwater; however, even if the Chernobyl Shelter collapses before the NSC is built, the resulting peak concentrations of 90Sr and 137Cs in the Dnieper River would still be below the drinking water limits.

Onishi, Yasuo; Kivva, Sergey L.; Zheleznyak, Mark J.; Voitsekhovitch, Oleg V.

2007-11-01T23:59:59.000Z

302

Cold Vacuum Drying facility design basis accident analysis documentation  

SciTech Connect

This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report (FSAR), ''Cold Vacuum Drying Facility 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 FSAR. The calculations in this document address the design basis accidents (DBAs) selected for analysis in HNF-3553, ''Spent Nuclear Fuel Project Final Safety Analysis Report'', Annex B, ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' The objective is to determine the quantity of radioactive particulate available for release at any point during processing at the Cold Vacuum Drying Facility (CVDF) and to use that quantity to determine the amount of radioactive material released during the DBAs. The radioactive material released is used to determine dose consequences to receptors at four locations, and the dose consequences are compared with the appropriate evaluation guidelines and release limits to ascertain the need for preventive and mitigative controls.

CROWE, R.D.

2000-08-08T23:59:59.000Z

303

Internally deposited fallout from the Chernobyl reactor accident  

SciTech Connect

Measurements of fallout radioactivity were made in the thyroid region, abdomen, whole body, or urine of 96 persons who were in eastern Europe at the time of the Chernobyl reactor accident or who went there shortly afterward. The most frequently encountered radionuclides were /sup 131/I, /sup 134,137/Cs, and /sup 103/Ru//sup 103/Rh. The median /sup 131/I activity in the thyroids of 42 subjects in whom radioiodine was detected and who were in Europe when the accident began was projected as 42 nCi the day the accident began. The median total body activity of /sup 134/Cs in 40 subjects in which it was detected was 1.7 nCi upon arrival in the US. For 51 subjects with detectable /sup 137/Cs burdens, the total body activity was 4.6 nCi. The risk of fatal thyroid cancer is less than 3 x 10/sup -6/ for nearly all subjects in this series. The risk of fatal cancer from /sup 134,137/Cs for subjects with cesium exposures similar to the ones observed by us, but who remained in Europe, is estimated as 1.4 x 10/sup -6/ to 4.2 x 10/sup -5/ with 95% of the risk attributable to /sup 137/Cs. 5 refs., 4 tabs.

Schlenker, R.A.; Oltman, B.G.; Lucas, H.F.

1987-01-01T23:59:59.000Z

304

ANS-8. 23: Criticality accident emergency planning and response  

SciTech Connect

A study group has been formed under the auspices of ANS-8 to examine the need for a standard on nuclear criticality accident emergency planning and response. This standard would be ANS-8.23. ANSI/ANS-8.19-1984, Administrative Practices for Nuclear Criticality Safety, provides some guidance on the subject in Section 10 titled -- Planned Response to Nuclear Criticality Accidents. However, the study group has formed a consensus that Section 10 is inadequate in that technical guidance in addition to administrative guidance is needed. The group believes that a new standard which specifically addresses emergency planning and response to a perceived criticality accident is needed. Plans for underway to request the study group be designated a writing group to create a draft of such a new standard. The proposed standard will divide responsibility between management and technical staff. Generally, management will be charged with providing the necessary elements of emergency planning such as a criticality detection and alarm system, training, safe evacuation routes and assembly areas, a system for timely accountability of personnel, and an effective emergency response organization. The technical staff, on the other hand, will be made responsible for establishing specific items such as safe and clearly posted evacuation evacuation routes and dose criteria for personnel assembly areas. The key to the question of responsibilities is that management must provide the resources for the technical staff to establish the elements of an emergency response effort.

Pruvost, N.L.

1991-06-24T23:59:59.000Z

305

Enhanced Accident Tolerant Fuels for LWRS - A Preliminary Systems Analysis  

Science Conference Proceedings (OSTI)

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

Gilles Youinou; R. Sonat Sen

2013-09-01T23:59:59.000Z

306

Health effects of the nuclear accident at Three Mile Island  

SciTech Connect

Between March 28 and April 15, 1979 the collective dose resulting from the radioactivity released to the population living within a 50-mile radius of the Three Mile Island nuclear plant was about 2000 person-rems, less than 1% of the annual natural background level. The average dose to a person living within 5 miles of the nuclear plant was less than 10% of annual background radiation. The maximum estimated radiation dose received by any one individual in the general population (excluding the nuclear plant workers) during the accident was 70 mrem. The doses received by the general population as a result of the accident were so small that there will be no detectable additional cases of cancer, developmental abnormalities, or genetic ill-health. Three Three Mile Island nuclear workers received radiation doses of about 3 to 4 rem, exceeding maximum permissible quarterly dose of 3 rem. The major health effect of the accident at Three Mile Island was that of a pronounced demoralizing effect on the general population in the Three Mile Island area, including teenagers and mothers of preschool children and the nuclear plant workers. However, this effect proved transient in all groups studied except the nuclear workers.

Fabrikant, J.I.

1980-05-01T23:59:59.000Z

307

TotalView Parallel Debugger at NERSC  

NLE Websites -- All DOE Office Websites (Extended Search)

Totalview Totalview Totalview Description TotalView from Rogue Wave Software is a parallel debugging tool that can be run with up to 512 processors. It provides both X Windows-based Graphical User Interface (GUI) and command line interface (CLI) environments for debugging. The performance of the GUI can be greatly improved if used in conjunction with free NX software. The TotalView documentation web page is a good resource for learning more about some of the advanced TotalView features. Accessing Totalview at NERSC To use TotalView at NERSC, first load the TotalView modulefile to set the correct environment settings with the following command: % module load totalview Compiling Code to Run with TotalView In order to use TotalView, code must be compiled with the -g option. We

308

Adaptation of user views to business requirements: towards adaptive views models  

Science Conference Proceedings (OSTI)

This article describes an approach to model user views in service-oriented groupware systems in order to support the adaptation of views in collaborative projects environments. The issue is to adapt the visualization to the users' business requirements ... Keywords: AEC (architecture engineering construction), CSCW, information visualization, model driven engineering, user view, user view model

Conrad Boton; Sylvain Kubicki; Gilles Halin

2010-09-01T23:59:59.000Z

309

A Regulator's View of Cogeneration  

E-Print Network (OSTI)

The Pennsylvania Public Utility Commission regulates essentially all types of public utilities and has the authority to investigate issues of public interest. To establish a point of reference, Pennsylvania's utilities contribute about 5 percent of the total national electric generation. In view of the energy requirements of Pennsylvania's industry and the impact of increasing energy costs on employment the Commission directed its technical staff to investigate the potential for industrial cogeneration and a pricing formula consistent with the electric utilities' costs. The Commission's technical staff has completed proposed regulations to implement the provisions of the Public Utility Regulatory Policies Act (PURPA) Section 210 concerning small power producers. The regulations incorporate suggestions from both potential producers and utilities. Staff has devised a strategy for utility purchases of energy and capacity which should be of interest to regulators in other jurisdictions, encourage potential cogenerators and satisfy utilities.

Shanaman, S. M.

1982-01-01T23:59:59.000Z

310

SCDAP/RELAP5/MOD 3.1 code manual: MATPRO, A library of materials properties for Light-Water-Reactor accident analysis. Volume 4  

Science Conference Proceedings (OSTI)

The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light -- water-reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume, Volume IV, describes the material properties correlations and computer subroutines (MATPRO) used by SCDAP/RELAP5. formulation of the materials properties are generally semi-empirical in nature. The materials property subroutines contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, cadmium, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, fill gas mixtures, carbon steel, and tungsten. This document also contains descriptions of the reaction and solution rate models needed to analyze a reactor accident.

Hagrman, D.T. [ed.; Allison, C.M.; Berna, G.A. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)] [and others

1995-06-01T23:59:59.000Z

311

DOE Order Self Study Modules - DOE O 225.1B, Accident Investigation  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

225.1B 225.1B ACCIDENT INVESTIGATIONS DOE O 225.1B Accident Investigations Familiar Level June 2011 1 June 2011 DOE ORDER O 225.1B ACCIDENT INVESTIGATIONS FAMILIAR LEVEL OBJECTIVES Given the familiar level of this module and the resources listed below, you will be able to: 1. State the purpose of implementing U.S. Department of Energy (DOE) O 225.1B. 2. Discuss the responsibilities of the heads of field elements for accident investigations. 3. Discuss the responsibilities of the appointing official in an accident investigation. 4. Discuss the responsibilities of the Accident Investigation Board Chairperson. 5. Discuss the criteria identified in appendix A of DOE O 225.1B. Note: If you think that you can complete the practice at the end of this level without

312

Descriptions of selected accidents that have occurred at nuclear reactor facilities  

SciTech Connect

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

Bertini, H.W.

1980-04-01T23:59:59.000Z

313

Descriptions of selected accidents that have occurred at nuclear reactor facilities  

SciTech Connect

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

Bertini, H.W.

1980-04-01T23:59:59.000Z

314

MACCS usage at Rocky Flats Plant for consequence analysis of postulated accidents  

Science Conference Proceedings (OSTI)

The MELCOR Accident Consequence Code System (MACCS) has been applied to the radiological consequence assessment of potential accidents from a non-reactor nuclear facility. MACCS has been used in a variety of applications to evaluate radiological dose and health effects to the public from postulated plutonium releases and from postulated criticalities. These applications were conducted to support deterministic and probabilistic accident analyses for safety analyses for safety analysis reports, radiological sabotage studies, and other regulatory requests.

Foppe, T.L.; Peterson, V.L.

1993-10-01T23:59:59.000Z

315

MAAP5 Simulation of Accidents at Fukushima Dai-ichi Units 1, 2, and 3  

Science Conference Proceedings (OSTI)

The original MAAP4 code functional design specification (circa 1989) was defined to address the full extent of degraded core accidents with the potential for reflooding of a badly damaged core. It was intended to support probabilistic risk assessment (PRA) and severe accident management guideline (SAMG) applications that previously were limited by the relatively rudimentary design for MAAP3.0B, the predecessor code.The accidents at Fukushima Dai-ichi Units 1, 2, and 3 prompted a ...

2013-02-23T23:59:59.000Z

316

Safety evaluation of MHTGR licensing basis accident scenarios  

SciTech Connect

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

Kroeger, P.G.

1989-04-01T23:59:59.000Z

317

Accident analysis and safety review of DOE Category B reactors  

SciTech Connect

DOE is employing the principle of comparability with the NRC requirements to guide its safety program. Since the safety record of research reactors licensed by the NRC has been established and accepted, the comparison of DOE Orders applicable to DOE research reactors with the NRC regulations applicable to research reactors would identify strengths and weaknesses of the DOE Orders. The comparison was made in 14 general topics of safety which are labeled Areas of Safety Concerns. This paper focuses on the Area of accident analysis and safety review and presents recommendations in these areas. 12 refs.

Kimura, C.Y.

1990-08-07T23:59:59.000Z

318

Advance plant severe accident/thermal hydraulic issues for ACRS  

DOE Green Energy (OSTI)

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

Kress, T.S.

1994-09-01T23:59:59.000Z

319

Compendium of information on hydrogen behavior during LWR accidents  

DOE Green Energy (OSTI)

A manual has been written which attempts to present, in a simple and understandable way, information concerning the generation, transport, detection and combustion of hydrogen which might occur during serious accidents in light water reactors. More than a thousand documents were surveyed over a three month period. Of these, several hundred documents were extensively reviewed. The manual summarizes the results of this review. The manual is divided into four major sections covering hydrogen generation, detection, combustion, and existing schemes for mitigating the effects of combustion.

Berman, M.

1980-01-01T23:59:59.000Z

320

Determination of influence factors and accident rates for the Armored Tractor/Safe Secure Trailer  

Science Conference Proceedings (OSTI)

Operating environments, such as road type, road location, and time of day, play an important role in the observed accident rates of heavy trucks used in general commerce. These same factors influence the accident rate of the Armored Tractor/Safe Secure Trailer (AT/SST) used by the Department of Energy to transport hazardous cargos within the continental United States. This report discusses the development of accident rate influence factors. These factors, based on heavy trucks used in general commerce, are used to modify the observed overall AT/SST accident rate to account for the different operating environments.

Phillips, J.S.; Clauss, D.B. [Sandia National Labs., Albuquerque, NM (United States); Blower, D.F. [Univ. of Michigan Transportation Research Institute, Ann Arbor, MI (United States). Center for National Truck Statistics

1994-04-01T23:59:59.000Z

Note: This page contains sample records for the topic "accident conditions view" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


321

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

SciTech Connect

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.

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.

1992-11-01T23:59:59.000Z

322

Facts and Lessons of the Fukushima Nuclear Accident and Safety Improvement  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Facts and Lessons of the Fukushima Nuclear Accident and Safety Facts and Lessons of the Fukushima Nuclear Accident and Safety Improvement - The Operator Viewpoints Facts and Lessons of the Fukushima Nuclear Accident and Safety Improvement - The Operator Viewpoints September 19, 2012 Presenter: Akira Kawano, General Manager, Nuclear International Relations and Strategy Group, Nuclear Power and Plant Siting Administrative Department, Tokyo Electric Power Company Topics Covered: How Tsunami Struck Fukushima Sites Tsunami Height Estimation How we responded in the Recovery Process Safety Improvement and Further Enhancement of Nuclear Safety Facts and Lessons of the Fukushima Nuclear Accident and Safety Improvement - The Operator Viewpoints More Documents & Publications January2005 NNSANews Meeting Materials: June 15, 2011

323

Order Module--DOE Order 225.1B, ACCIDENT INVESTIGATIONS  

Energy.gov (U.S. Department of Energy (DOE))

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

324

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

Science Conference Proceedings (OSTI)

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

Burgherr, P.; Hirschberg, S. [Paul Scherrer Institute, Villigen (Switzerland)

2008-07-01T23:59:59.000Z

325

Severe Accident Phenomenology Analyses and Fission Gas Release in Advanced Nuclear Reactors.  

E-Print Network (OSTI)

??The aim of this work is to contribute to qualify a model in order to simulate the progression of a severe accident (SA), evaluating the (more)

MAZZINI, GUIDO

2012-01-01T23:59:59.000Z

326

A SCOPING STUDY: Development of Probabilistic Risk Assessment Models for Reactivity Insertion Accidents During Shutdown In U.S. Commercial Light Water Reactors  

SciTech Connect

This report documents the scoping study of developing generic simplified fuel damage risk models for quantitative analysis from inadvertent reactivity insertion events during shutdown (SD) in light water pressurized and boiling water reactors. In the past, nuclear fuel reactivity accidents have been analyzed both mainly deterministically and probabilistically for at-power and SD operations of nuclear power plants (NPPs). Since then, many NPPs had power up-rates and longer refueling intervals, which resulted in fuel configurations that may potentially respond differently (in an undesirable way) to reactivity accidents. Also, as shown in a recent event, several inadvertent operator actions caused potential nuclear fuel reactivity insertion accident during SD operations. The set inadvertent operator actions are likely to be plant- and operation-state specific and could lead to accident sequences. This study is an outcome of the concern which arose after the inadvertent withdrawal of control rods at Dresden Unit 3 in 2008 due to operator actions in the plant inadvertently three control rods were withdrawn from the reactor without knowledge of the main control room operator. The purpose of this Standardized Plant Analysis Risk (SPAR) Model development project is to develop simplified SPAR Models that can be used by staff analysts to perform risk analyses of operating events and/or conditions occurring during SD operation. These types of accident scenarios are dominated by the operator actions, (e.g., misalignment of valves, failure to follow procedures and errors of commissions). Human error probabilities specific to this model were assessed using the methodology developed for SPAR model human error evaluations. The event trees, fault trees, basic event data and data sources for the model are provided in the report. The end state is defined as the reactor becomes critical. The scoping study includes a brief literature search/review of historical events, developments of a small set of comprehensive event trees and fault trees and recommendation for future work.

S. Khericha

2011-06-01T23:59:59.000Z

327

Accident Fault Trees for Defense Waste Processing Facility  

Science Conference Proceedings (OSTI)

The purpose of this report is to document fault tree analyses which have been completed for the Defense Waste Processing Facility (DWPF) safety analysis. Logic models for equipment failures and human error combinations that could lead to flammable gas explosions in various process tanks, or failure of critical support systems were developed for internal initiating events and for earthquakes. These fault trees provide frequency estimates for support systems failures and accidents that could lead to radioactive and hazardous chemical releases both on-site and off-site. Top event frequency results from these fault trees will be used in further APET analyses to calculate accident risk associated with DWPF facility operations. This report lists and explains important underlying assumptions, provides references for failure data sources, and briefly describes the fault tree method used. Specific commitments from DWPF to provide new procedural/administrative controls or system design changes are listed in the ''Facility Commitments'' section. The purpose of the ''Assumptions'' section is to clarify the basis for fault tree modeling, and is not necessarily a list of items required to be protected by Technical Safety Requirements (TSRs).

Sarrack, A.G.

1999-06-22T23:59:59.000Z

328

Accident sequence precursor events with age-related contributors  

SciTech Connect

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

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

1995-12-31T23:59:59.000Z

329

PERSPECTIVES ON A DOE CONSEQUENCE INPUTS FOR ACCIDENT ANALYSIS APPLICATIONS  

Science Conference Proceedings (OSTI)

Department of Energy (DOE) accident analysis for establishing the required control sets for nuclear facility safety applies a series of simplifying, reasonably conservative assumptions regarding inputs and methodologies for quantifying dose consequences. Most of the analytical practices are conservative, have a technical basis, and are based on regulatory precedent. However, others are judgmental and based on older understanding of phenomenology. The latter type of practices can be found in modeling hypothetical releases into the atmosphere and the subsequent exposure. Often the judgments applied are not based on current technical understanding but on work that has been superseded. The objective of this paper is to review the technical basis for the major inputs and assumptions in the quantification of consequence estimates supporting DOE accident analysis, and to identify those that could be reassessed in light of current understanding of atmospheric dispersion and radiological exposure. Inputs and assumptions of interest include: Meteorological data basis; Breathing rate; and Inhalation dose conversion factor. A simple dose calculation is provided to show the relative difference achieved by improving the technical bases.

(NOEMAIL), K; Jonathan Lowrie, J; David Thoman (NOEMAIL), D; Austin Keller (NOEMAIL), A

2008-07-30T23:59:59.000Z

330

Worldwide health effects of the Fukushima Daiichi nuclear accident  

E-Print Network (OSTI)

This study quantifies worldwide health effects of the Fukushima Daiichi nuclear accident on 11 March 2011. Effects are quantified with a 3-D global atmospheric model driven by emission estimates and evaluated against daily worldwide Comprehensive Nuclear-Test-Ban Treaty Organization (CTBTO) measurements and observed deposition rates. Inhalation exposure, ground-level external exposure, and atmospheric external exposure pathways of radioactive iodine-131, cesium-137, and cesium-134 released from Fukushima are accounted for using a linear no-threshold (LNT) model of human exposure. Exposure due to ingestion of contaminated food and water is estimated by extrapolation. We estimate an additional 130 (151100) cancer-related mortalities and 180 (241800) cancer-related morbidities incorporating uncertainties associated with the exposuredose and doseresponse models used in the study. We also discuss the LNT models uncertainty at low doses. Sensitivities to emission rates, gas to particulate I-131 partitioning, and the mandatory evacuation radius around the plant are also explored, and may increase upper bound mortalities and morbidities in the ranges above to 1300 and 2500, respectively. Radiation exposure to workers at the plant is projected to result in 2 to 12 morbidities. An additional 600 mortalities have been reported due to non-radiological causes such as mandatory evacuations. Lastly, a hypothetical accident at the Diablo Canyon Power Plant in

John E. Ten Hoeve A; Mark Z. Jacobson B

2012-01-01T23:59:59.000Z

331

Analyzing the BWR rod drop accident in high-burnup cores  

SciTech Connect

This study was undertaken for the US Nuclear Regulatory Commission to determine the fuel enthalpy during a rod drop accident (RDA) for cores with high burnup fuel. The calculations were done with the RAMONA-4B code which models the core with 3-dimensional neutron kinetics and multiple parallel coolant channels. The calculations were done with a model for a BWR/4 with fuel bundles having burnups up to 30 GWd/t and also with a model with bundle burnups to 60 GWd/t. This paper also discusses potential sources of uncertainty in calculations with high burnup fuel. One source is the ``rim`` effect which is the extra large peaking of the power distribution at the surface of the pellet. This increases the uncertainty in reactor physics and heat conduction models that assume that the energy deposition has a less peaked spatial distribution. Two other sources of uncertainty are the result of the delayed neutron fraction decreasing with burnup and the positive moderator temperature feedback increasing with burnup. Since these effects tend to increase the severity of the event, an RDA calculation for high burnup fuel will underpredict the fuel enthalpy if the effects are not properly taken into account. Other sources of uncertainty that are important come from the initial conditions chosen for the RDA. This includes the initial control rod pattern as well as the initial thermal-hydraulic conditions.

Diamond, D.J.; Neymotin, L.; Kohut, P.

1995-08-01T23:59:59.000Z

332

Answering XML queries using materialized views revisited  

Science Conference Proceedings (OSTI)

Answering queries using views is a well-established technique in databases. In this context, two outstanding problems can be formulated. The first one consists in deciding whether a query can be answered exclusively using one or multiple materialized ... Keywords: XML, materialized views, xpath query evaluation

Xiaoying Wu; Dimitri Theodoratos; Wendy Hui Wang

2009-11-01T23:59:59.000Z

333

Query and update through XML views  

Science Conference Proceedings (OSTI)

XML has become a standard medium for data exchange, and XML views are frequently used as an interface to relational database and XML data. There have been a considerable number of studies on building and querying XML views, while updating related topics ...

Gao Cong

2007-10-01T23:59:59.000Z

334

Designing presentations for on-demand viewing  

Science Conference Proceedings (OSTI)

Increasingly often, presentations are given before a live audience, while simultaneously being viewed remotely and recorded for subsequent viewing on-demand over the Web. How should video presentations be designed for web access? How is video accessed ... Keywords: digital library, streaming media, video on-demand

Liwei He; Jonathan Grudin; Anoop Gupta

2000-12-01T23:59:59.000Z

335

April 2013 Most Viewed Documents for Energy Storage, Conversion, And  

Office of Scientific and Technical Information (OSTI)

April 2013 Most Viewed Documents for Energy Storage, Conversion, And April 2013 Most Viewed Documents for Energy Storage, Conversion, And Utilization Seventh Edition Fuel Cell Handbook NETL (2004) 628 Continuously variable transmissions: theory and practice Beachley, N.H.; Frank, A.A. (null) 205 A study of lead-acid battery efficiency near top-of-charge and the impact on PV system design Stevens, J.W.; Corey, G.P. (1996) 173 Energy Saving Potentials and Air Quality Benefits of Urban HeatIslandMitigation Akbari, Hashem (2005) 153 Building a secondary containment system Broder, M.F. (1994) 144 An Improved Method of Manufacturing Corrugated Boxes: Lateral Corrugator Frank C. Murray Ph.D.; , Roman Popil Ph.D.; Michael Shaepe (formerly with IPST, now at Cargill. Inc) (2008) 141 Ammonia usage in vapor compression for refrigeration and air-conditioning in the United States

336

Most Viewed Documents - Environmental Sciences | OSTI, US Dept of Energy,  

Office of Scientific and Technical Information (OSTI)

Most Viewed Documents - Environmental Sciences Most Viewed Documents - Environmental Sciences Separation of heavy metals: Removal from industrial wastewaters and contaminated soil Peters, R.W.; Shem, L. (1993) Special Report on Emissions Scenarios : a special report of Working Group III of the Intergovernmental Panel on Climate Change Nakicenovic, Nebojsa; Alcamo, Joseph; Davis, Gerald; et al. (2000) CARBON DIOXIDE (REDUCTION) FUJITA,E. (2000) Ecological Screening Values for Surface Water, Sediment, and Soil Friday, G. P. (1999) Mitigation options for accidental releases of hazardous gases Fthenakis, V.M. (1995) Atmospheric Chemistry and Greenhouse Gases Ehhalt, D.; Prather, M.; Dentener, F.; et al. (2001) Ammonia usage in vapor compression for refrigeration and air-conditioning in the United States Fairchild, P.D.; Baxter, V.D. (1995)

337

User_Sup_ViewEmpMatrix  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

an Employee Matrix (Supervisor) an Employee Matrix (Supervisor) © 2011 SuccessFactors, Inc. - 1 - SuccessFactors Learning Confidential. All rights reserved. Job Aid: Viewing an Employee Matrix (Supervisor) Purpose The purpose of this job aid is to guide supervisor users through the step-by-step process of viewing an employee matrix within SuccessFactors Learning. Task A. View an Employee Matrix From the Home page, click the My Employees tab. Click the Employee Matrix supervisor link. Click the Change Measures expand arrow ( ) to select criteria for comparison. View an Employee Matrix 8 Steps Task A 3 3 1 1 2 2 SuccessFactors Learning v 6.4 User Job Aid Viewing an Employee Matrix (Supervisor) © 2011 SuccessFactors, Inc. - 2 - SuccessFactors Learning Select a measure for the

338

Mountain View Grand | Open Energy Information  

Open Energy Info (EERE)

Grand Grand Jump to: navigation, search Name Mountain View Grand Facility Mountain View Grand Sector Wind energy Facility Type Small Scale Wind Facility Status In Service Owner Mountain View Grand Developer Sustainable Energy Developments Energy Purchaser Mountain View Grand Location Mountain View Grand Resort & Spa NH Coordinates 44.397987°, -71.590306° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":44.397987,"lon":-71.590306,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

339

U-008: Symantec Data Loss Prevention Bugs in KeyView Filter Lets Remote  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

08: Symantec Data Loss Prevention Bugs in KeyView Filter Lets 08: Symantec Data Loss Prevention Bugs in KeyView Filter Lets Remote Users Deny Service U-008: Symantec Data Loss Prevention Bugs in KeyView Filter Lets Remote Users Deny Service October 11, 2011 - 8:00am Addthis PROBLEM: Symantec Data Loss Prevention Bugs in KeyView Filter Lets Remote Users Deny Service PLATFORM: Symantec Data Loss Prevention Enforce/Detection Servers for Windows 10.x, 11.x ABSTRACT: A remote user can create a file that, when processed by the target filter, will cause partial denial of service conditions. reference LINKS: Symantec Security Advisory SYM11-013 SecurityTracker Alert ID: 1026157 IMPACT ASSESSMENT: Medium Discussion: Multiple vulnerabilities were reported in Symantec Data Loss Prevention. A remote user can cause denial of service conditions on the target system.A

340

Reactivity Initiated Accident Test Series Test RIA 1-2 Experiment Operating Specification  

SciTech Connect

This document describes the experiment operating specifications for the Reactivity Initiated Accident (RIA) Test RIA 1-2 to be conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. The RIA Series I research objectives are to determine fuel failure thresholds, modes and consequences as functions of enthalpy insertion, irradiation history, and fuel design. Coolant conditions of pressure, temperature, and flow rate that are typical of hot-startup conditions in commercial boiling water reactors {BWRs) will be used. The second test in Series I, Test RIA 1-2, will be comprised of four individual rods, each surrounded by a separate flow shroud. The four rods will be preirradiated. The specific objectives of the test are to: (1) characterize the response of preirradiated fuel rods during a RIA event conducted at BWR hot-startup conditions and (2) evaluate the effect of internal rod pressure on preirradiated fuel rod transient response. The test sequence will begin with steady state power operation to condition the fuel (pellet cracking and relocation) and determine the fuel rod power calibration. The loop will then be cooled down, the test train removed from the in-pile tube, and the cobalt flux wires that are mounted on each flow shroud will be replaced. The transient fuel rod energy deposition for the Test RIA 1-2 rods will be chosen from the fuel rod response vs. energy deposition observed in the first three phases of the RIA Scoping Test and the first test of Series J, Test RIA 1-1. The design of the test fuel rods, test assembly, and instrumentation associated with Test RIA 1-2 are described. The planned experiment conduct for the test is described. The data recording and reduction requirements are provided. The posttest operations support and the postirradiation examination requirements associated with Test RIA 1-2 are described.

1978-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "accident conditions view" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

Reactivity Initiated Accident Test Series Test RIA 1-1 Experiment Operating Specification  

SciTech Connect

This document describes the experiment operating specifications for the Reactivity Initiated Accident (RIA) Test RIA 1-1 to be conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. The RIA Series I research objectives are to determine fuel failure thresholds, modes and consequences as functions of enthalpy insertion, irradiation history, and fuel design. Coolant conditions of pressure, temperature, and flow rate that are typical of hot-startup conditions in commercial boiling water reactors (BWRs) will be used. The first test in Series I, Test RIA 1-1, will be comprised of four individual rods, each surrounded by a separate flow shroud. Two rods will be preirradiated and two rods will be unirradiated. The specific objectives of the test are to: (1) characterize the response of unirradiated and preirradiated fuel rods during a RIA event conducted at BWR hot-startup conditions and (2) evaluate test instrumentation response during an RIA. The test sequence will begin with steady state power operation to condition the fuel (pellet cracking and relocation) and determine the fuel rod power calibration. The loop will then be cooled down, the test train removed from the in-pile tube, and one of the unirradiated rods will be removed for fission product analysis and replaced with an identical unirradiated rod. The transient fuel rod energy deposition for Test RIA 1-1 will be chosen from the fuel rod response vs. energy deposition data observed in the first three phases of the RIA Scoping Test. It is anticipated that a fuel pellet surface energy deposition of about 1100 J/g will be required to ensure cladding failure of all four rods. The design of the test fuel rods, test assembly, and instrumentation associated with Test RIA 1-1 are described. The experiment conduct for the test is described. The data recording and reduction requirements are provided. The posttest support and the postirradiation examination requirements associated with Test RIA 1-1 are described.

1978-08-01T23:59:59.000Z

342

Analysis of traffic accident severity using Decision Rules via Decision Trees  

Science Conference Proceedings (OSTI)

A Decision Tree (DT) is a potential method for studying traffic accident severity. One of its main advantages is that Decision Rules (DRs) can be extracted from its structure. And these DRs can be used to identify safety problems and establish certain ... Keywords: Decision Rules, Decision Trees, Road safety, Severity, Traffic accident

JoaquN AbellN, Griselda LPez, Juan De OA

2013-11-01T23:59:59.000Z

343

Recommendations for Analyzing Accidents under the National Environmental Policy Act (July 2002)  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

National National Environmental Policy Act RECOMMENDATIONS for ANALYZING ACCIDENTS under the NATIONAL ENVIRONMENTAL POLICY ACT N E P A July 2002 U.S. Department of Energy Environment, Safety and Health Office of NEPA Policy and Compliance Recommendations for Analyzing Accidents under NEPA Contents 1.0 Introduction ............................................................................................................................... 1 1.1 Definition............................................................................................................................ 1 1.2 Purpose.............................................................................................................................. 1 1.3 Sliding Scale ......................................................................................................................

344

SD-GIS-based temporal-spatial simulation of water quality in sudden water pollution accidents  

Science Conference Proceedings (OSTI)

System dynamics (SD) is well suited for studying dynamic nonlinear complex systems. In this paper, SD is applied to a rapid-onset water pollution accident using a 1-D water quality model and a conceptual GIS-SD framework is constructed to simulate the ... Keywords: System dynamics, Temporal-spatial simulation, Water pollution accidents

Bo Zhang; Yu Qin; Mingxiang Huang; Qiang Sun; Shun Li; Liqiang Wang; Chaohui Yu

2011-07-01T23:59:59.000Z

345

A novel approach for traffic accidents sanitary resource allocation based on multi-objective genetic algorithms  

Science Conference Proceedings (OSTI)

The development of communication technologies integrated in vehicles allows creating new protocols and applications to improve assistance in traffic accidents. Combining this technology with intelligent systems will permit to automate most of the decisions ... Keywords: Multi-objective genetic algorithms, Resource allocation, Traffic accidents assistance

Manuel Fogue; Piedad Garrido; Francisco J. Martinez; Juan-Carlos Cano; Carlos T. Calafate; Pietro Manzoni

2013-01-01T23:59:59.000Z

346

Search and View a Service Request Search and View a Service Request,vlr  

E-Print Network (OSTI)

FAMIS Search and View a Service Request Search and View a Service Request,vlr 10/6/2012 © 2012 Northwestern University 1 Search and View a Service Request This training guide will show you how to search completion of this guide you will be able to: · Sign in to FAMIS Self Service · Search for service requests

Shull, Kenneth R.

347

Detailed Analysis of In-Vessel Melt Progression in the Loss of Coolant Accident of OPR1000  

Science Conference Proceedings (OSTI)

An in-vessel severe accident progression has been analyzed to generate the basic data for an evaluation of the in-vessel severe accident management strategies and to identify the thermal hydraulic condition of the reactor vessel and the damage state of the in-vessel materials at a reactor vessel failure by using the SCDAP/RELAP5/MOD3.3 computer code during the Loss Of Coolant Accident (LOCA) without the Safety Injection (SI) of the OPR (Optimized Pressurize Reactor) 1000. Best estimate calculation of the small break LOCAs of 1.35 inch and 2 inch, the medium break LOCAs of 3 inch and a 4.28 inch, and a large break LOCA of 9.8 inch without the SI have been performed from a transient initiation to a reactor vessel failure. The SCDAP/RELAP5/MOD3.3 results have shown that in all the transients, approximately 30-40 % of the core material was melted and relocated to the lower plenum of the reactor vessel at the time of a reactor vessel failure. In the small and large break LOCAs, the reactor vessel failed at an early time of approximately 70-110 minutes after the transients were initiated. Since the Safety Injection Tanks (SITs) were actuated effectively in the medium break LOCAs, the reactor vessel failed at a later time of approximately 200-400 minutes after the transients were initiated. At the time of a reactor vessel failure, approximately 45-55 % of the fuel rod cladding was oxidized in the small and medium break LOCAs. However, approximately 20 % of the fuel rod cladding was oxidized because of a coolant loss through the break in the large break LOCA of the OPR1000. (authors)

Park, R.J.; Kim, S.B.; Kim, H.D. [Korea Atomic Energy Research Institute, 150 Dukjin-dong, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

2006-07-01T23:59:59.000Z

348

Application of accident progression event tree technology to the Savannah River Site Defense Waste Processing Facility SAR analysis  

SciTech Connect

The Accident Analysis in the Safety Analysis Report (SAR) for the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) has recently undergone an upgrade. Non-reactor SARs at SRS (and other Department of Energy (DOE) sites) use probabilistic techniques to assess the frequency of accidents at their facilities. This paper describes the application of an extension of the Accident Progression Event Tree (APET) approach to accidents at the SRS DWPF. The APET technique allows an integrated model of the facility risk to be developed, where previous probabilistic accident analyses have been limited to the quantification of the frequency and consequences of individual accident scenarios treated independently. Use of an APET allows a more structured approach, incorporating both the treatment of initiators that are common to more than one accident, and of accident progression at the facility.

Brandyberry, M.D.; Baker, W.H.; Wittman, R.S. [Westinghouse Savannah River Co., Aiken, SC (United States); Amos, C.N. [Science Applications International Corp., Albuquerque, NM (United States)

1993-12-31T23:59:59.000Z

349

The Accident Sequence Precursor program: Methods improvements and current results  

Science Conference Proceedings (OSTI)

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

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

1987-01-01T23:59:59.000Z

350

Down syndrome clusters in Germany after the Chernobyl accident  

Science Conference Proceedings (OSTI)

In two independent studies using different approaches and covering West Berlin and Bavaria, respectively, highly significant temporal clusters of Down syndrome were found. Both sharp increases occurred in areas receiving relatively low Chernobyl fallout and concomitant radiation exposures. Only for the Berlin cluster was fallout present at the time of the affected meiosis, whereas the Nuremberg cluster preceded the radioactive contamination by 1 month. Hypotheses on possible causal relationships are compared. Radiation from the Chernobyl accident is an unlikely factor, because the associated cumulative dose was so low in comparison with natural background. Microdosimetric considerations would indicate that fewer than 1 in 200 oocyte nuclei would have experienced an ionizing event from Chernobyl radioactivity. Given the lack of understanding of what causes Down syndrome, other than factors associated with increased maternal age, additional research into environmental and infectious risk factors is warranted. 23 refs., 4 figs., 2 tabs.

Burkart, W.; Grosche, B.; Schoetzau, A. [Institute for Radiation Hygiene, Oberschleissheim (Germany)

1997-03-01T23:59:59.000Z

351

The role of chemical reactions in the Chernobyl accident  

SciTech Connect

It is shown that chemical reactions played an essential role in the Chernobyl accident at all of its stages. It is important that the reactor before the explosion was at maximal xenon poisoning, and its reactivity, apparently, was not destroyed by the explosion. The reactivity release due to decay of Xe-235 on the second day after the explosion led to a reactor power of 80-110 MW. Owing to this power, the chemical reactions of reduction of uranium, plutonium, and other metals at a temperature of about 2000 Degree-Sign C occurred in the core. The yield of fission products thus sharply increased. Uranium and other metals flew down in the bottom water communications and rooms. After reduction of the uranium and its separation from the graphite, the chain reaction stopped, the temperature of the core decreased, and the activity yield stopped.

Grishanin, E. I., E-mail: egrishanin@orexovo.net [Russian Research Center Kurchatov Institute (Russian Federation)

2010-12-15T23:59:59.000Z

352

Probabilistic Accident Consequence Uncertainty - A Joint CEC/USNRC Study  

Science Conference Proceedings (OSTI)

The joint USNRC/CEC consequence uncertainty study was chartered after the development of two new probabilistic accident consequence codes, MACCS in the U.S. and COSYMA in Europe. Both the USNRC and CEC had a vested interest in expanding the knowledge base of the uncertainty associated with consequence modeling, and teamed up to co-sponsor a consequence uncertainty study. The information acquired from the study was expected to provide understanding of the strengths and weaknesses of current models as well as a basis for direction of future research. This paper looks at the elicitation process implemented in the joint study and discusses some of the uncertainty distributions provided by eight panels of experts from the U.S. and Europe that were convened to provide responses to the elicitation. The phenomenological areas addressed by the expert panels include atmospheric dispersion and deposition, deposited material and external doses, food chain, early health effects, late health effects and internal dosimetry.

Gregory, Julie J.; Harper, Frederick T.

1999-07-28T23:59:59.000Z

353

MACCS2: An improved code for assessing nuclear accident consequences  

Science Conference Proceedings (OSTI)

The MACCS computer code was developed to predict probabilistic assessments of the consequences from severe accidents at nuclear power plants.For DOE applications and sensitivity studies of emergency response actions at nuclear power plants, MACCS2 represents a significant improvement in modeling flexibility over MACCS 1.5. This increased flexibility is obtained with an approximate doubling of the code`s run time and memory requirements. The software can be adapted to most computers. An executable is included in the code package for 386/486 IBM-compatible personal computers with 8 megabytes of random access memory (RAM). MACCS2 is being benchmarked against the RSAC-5 code developed by INEL. A new set of code documentation is being prepared that describes the use of the code, the models implemented, and the code benchmarking. Current plans are to have the code package (including source code) available to the public at the end of fiscal year 1994.

Chanin, D.I.; Banjac, V.; Miller, L.A. [Sandia National Lab., Albuquerque, NM (United States)

1994-12-31T23:59:59.000Z

354

Corium Physical Properties for Severe Accident R and D  

SciTech Connect

Corium is a mixture formed - in the hypothetical case of a severe accident - of molten core and products from the decomposition of the internal structures, the vessel and the concrete. Before any calculation, any modelling, any experimental interpretation can be made, it is necessary to estimate the corium physical properties. Corium being a multicomponent mixture, special attention has been given at CEA to the mixing laws for multi-phase, multi-constituent mixtures. Compared to the previous database, considerable progress in characterizing of corium constituents physical properties has been achieved. The thermo-physical data of corium constituents for pure components and mixtures and the different physical laws that are recommended by experts are now implemented in the CORPRO (Corium Properties) database. (authors)

Journeau, C.; Piluso, P.; Frolov, K.N. [CEA Cadarache, Severe Accident Mastering Laboratory (DEN/DTN/STRI/LMA), 13108 St Paul lez Durance (France)

2004-07-01T23:59:59.000Z

355

The issue of stress state during mechanical tests to assess cladding performance during a reactivity-initiated accident (RIA)  

E-Print Network (OSTI)

of Materials Properties for Light Water Reactor Accident Analysis", NUREG/CR-6150, INEL-96.0422, chapter 4

Motta, Arthur T.

356

Preliminary results of the PWR low power and shutdown accident frequencies program: Coarse screening analysis for Surry  

Science Conference Proceedings (OSTI)

This document presents the preliminary internal events Level 1 results (including fire and flood) obtained as a result of a coarse screening analysis on the low power and shutdown accident frequencies of the Surry Nuclear Power Plant. The work was performed by Brookhaven National Laboratory (BNL) for the Nuclear Regulatory Commission Office of Nuclear Regulatory Research (RES). This coarse screening analysis was performed in support of the NRC staff's follow-up actions subsequent to the March 20, 1990 Vogtle incident with the objective of providing high-level qualitative insights within a relatively short time frame. It is the first phase of a major study that will ultimately produce estimates on the core damage frequency of a pressurized water reactor (PWR) during low power and shutdown conditions. Phase 2 of the study will be guided by the Phase 1 results in order to concentrate the effort on the various plant operational states, the dominant accident sequences, and pertinent data items according to their importance to core damage frequency and risk.

Chu, T.L.; Musicki, Z.; Luckas, W.; Wong, S.M.; Neymotin, L.; Diamond, D.J.; Hsu, C.J.; Bozoki, G.; Kohut, P.; Fitzpatrick, R. (Brookhaven National Lab., Upton, NY (United States)); Siu, N. (Massachusetts Inst. of Tech., Cambridge, MA (United States))

1991-01-01T23:59:59.000Z

357

Preliminary results of the PWR low power and shutdown accident frequencies program: Coarse screening analysis for Surry  

Science Conference Proceedings (OSTI)

This document presents the preliminary internal events Level 1 results (including fire and flood) obtained as a result of a coarse screening analysis on the low power and shutdown accident frequencies of the Surry Nuclear Power Plant. The work was performed by Brookhaven National Laboratory (BNL) for the Nuclear Regulatory Commission Office of Nuclear Regulatory Research (RES). This coarse screening analysis was performed in support of the NRC staff`s follow-up actions subsequent to the March 20, 1990 Vogtle incident with the objective of providing high-level qualitative insights within a relatively short time frame. It is the first phase of a major study that will ultimately produce estimates on the core damage frequency of a pressurized water reactor (PWR) during low power and shutdown conditions. Phase 2 of the study will be guided by the Phase 1 results in order to concentrate the effort on the various plant operational states, the dominant accident sequences, and pertinent data items according to their importance to core damage frequency and risk.

Chu, T.L.; Musicki, Z.; Luckas, W.; Wong, S.M.; Neymotin, L.; Diamond, D.J.; Hsu, C.J.; Bozoki, G.; Kohut, P.; Fitzpatrick, R. [Brookhaven National Lab., Upton, NY (United States); Siu, N. [Massachusetts Inst. of Tech., Cambridge, MA (United States)

1991-12-31T23:59:59.000Z

358

Ground control failures. A pictorial view of case studies  

Science Conference Proceedings (OSTI)

The book shows, in pictorial views, many forms and/or stages of types of failures in mines, for instance, cutter, roof falls, and cribs. In each case, the year of occurrence is stated in the beginning so that the environment or technological background under which it occurred are reflected. The narrative than begins with the mining and geological conditions, followed by a description of the ground control problems and recommended solutions and results, if any. The sections cover failure of pillars, roof falls, longwall, roof bolting, multiple-seam mining, floor heave, longwall, flooding and weathering of coal, old workings, and shortwall and thin-seam plow longwall.

Peng, S.S.

2007-07-01T23:59:59.000Z

359

Berkeley Lab View -- March 28, 2008  

NLE Websites -- All DOE Office Websites (Extended Search)

March 28th, 2008 Search the View Archive March 28th, 2008 Search the View Archive State of the Lab: New Initiatives, Construction Daniel Chemla (1940-2008): A Remembrance of His Career The View is Going Green DOE Excellence Award to Foundry Project Team Berkeley Lab View Here Comes BELLA: The BErkeley Lab Laser Acceleration Project Berkeley Lab Science Roundup State of the Lab: New Initiatives, Construction By Lynn Yarris image Photo by Roy Kaltschmidt, CSO Free electron lasers with attosecond capabilities, a high-energy electron accelerator less than a meter in length, the arrival of NERSC-6 and the departure of GELCO-4 were some of the highlights of Berkeley Lab Director Steve Chu's State-of-the-Lab address, which he delivered at the Building 50 Auditorium during the noon hour on March 10, with simulcast to the

360

Lake View Geothermal Facility | Open Energy Information  

Open Energy Info (EERE)

Page Page Edit with form History Facebook icon Twitter icon » Lake View Geothermal Facility Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Lake View Geothermal Facility General Information Name Lake View Geothermal Facility Facility Lake View Sector Geothermal energy Location Information Location The Geysers, California Coordinates 38.823527148671°, -122.78173327446° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":38.823527148671,"lon":-122.78173327446,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

Note: This page contains sample records for the topic "accident conditions view" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

Incorporating video into Google Mobile Street View  

E-Print Network (OSTI)

Mobile Street View is a compelling application but suffers from significant latency problems, especially in limited bandwidth circumstances. Currently, the application uses static images to display street level information. ...

Wright, Christina (Christina E.)

2010-01-01T23:59:59.000Z

362

A common-view disciplined oscillator  

SciTech Connect

This paper describes a common-view disciplined oscillator (CVDO) that locks to a reference time scale through the use of common-view global positioning system (GPS) satellite measurements. The CVDO employs a proportional-integral-derivative controller that obtains near real-time common-view GPS measurements from the internet and provides steering corrections to a local oscillator. A CVDO can be locked to any time scale that makes real-time common-view data available and can serve as a high-accuracy, self-calibrating frequency and time standard. Measurement results are presented where a CVDO is locked to UTC(NIST), the coordinated universal time scale maintained at the National Institute of Standards and Technology in Boulder, Colorado.

Lombardi, Michael A. [Time and Frequency Division, National Institute of Standards and Technology (NIST), Boulder, Colorado 80305 (United States); Dahlen, Aaron P. [Loran Support Unit, United States Coast Guard (USCG), Wildwood, New Jersey 08260 (United States)

2010-05-15T23:59:59.000Z

363

A view-sequential 3D display  

E-Print Network (OSTI)

This thesis outlines the various techniques for creating electronic 3D displays and analyzes their commercial potential. The thesis argues for the use of view-sequential techniques in the design of 3D displays based on ...

Cossairt, Oliver S. (Oliver Strider), 1978-

2003-01-01T23:59:59.000Z

364

The new option view of investment  

E-Print Network (OSTI)

This paper provides a simple introduction to the new option view of investment. We explain the shortcomings of the orthodox theory, and then outline the basic ideas behind the option framework. Several industry examples ...

Dixit, Avinash K.

1995-01-01T23:59:59.000Z

365

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

SciTech Connect

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

Zhegang Ma

2013-09-01T23:59:59.000Z

366

Hazards and accident analyses, an integrated approach, for the Plutonium Facility at Los Alamos National Laboratory  

Science Conference Proceedings (OSTI)

This paper describes an integrated approach to perform hazards and accident analyses for the Plutonium Facility at Los Alamos National Laboratory. A comprehensive hazards analysis methodology was developed that extends the scope of the preliminary/process hazard analysis methods described in the AIChE Guidelines for Hazard Evaluations. Results fro the semi-quantitative approach constitute a full spectrum of hazards. For each accident scenario identified, there is a binning assigned for the event likelihood and consequence severity. In addition, each accident scenario is analyzed for four possible sectors (workers, on-site personnel, public, and environment). A screening process was developed to link the hazard analysis to the accident analysis. Specifically the 840 accident scenarios were screened down to about 15 accident scenarios for a more through deterministic analysis to define the operational safety envelope. The mechanics of the screening process in the selection of final scenarios for each representative accident category, i.e., fire, explosion, criticality, and spill, is described.

Pan, P.Y.; Goen, L.K.; Letellier, B.C.; Sasser, M.K.

1995-07-01T23:59:59.000Z

367

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

SciTech Connect

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.

Youngen, G.

1988-10-01T23:59:59.000Z

368

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

Science Conference Proceedings (OSTI)

In the ambit of the Severe Accident Network of Excellence Project (SARNET), funded by the European Union, 6. FISA (Fission Safety) Programme, one of the main tasks is the development and validation of the European Accident Source Term Evaluation Code (ASTEC Code). One of the reference codes used to compare ASTEC results, coming from experimental and Reactor Plant applications, is MELCOR. ENEA is a SARNET member and also an ASTEC and MELCOR user. During the first 18 months of this project, we performed a series of MELCOR and ASTEC calculations referring to a French PWR 900 MWe and to the accident sequence of 'Loss of Steam Generator (SG) Feedwater' (known as H2 sequence in the French classification). H2 is an accident sequence substantially equivalent to a Station Blackout scenario, like a TMLB accident, with the only difference that in H2 sequence the scram is forced to occur with a delay of 28 seconds. The main events during the accident sequence are a loss of normal and auxiliary SG feedwater (0 s), followed by a scram when the water level in SG is equal or less than 0.7 m (after 28 seconds). There is also a main coolant pumps trip when {delta}Tsat < 10 deg. C, a total opening of the three relief valves when Tric (core maximal outlet temperature) is above 603 K (330 deg. C) and accumulators isolation when primary pressure goes below 1.5 MPa (15 bar). Among many other points, it is worth noting that this was the first time that a MELCOR 1.8.5 input deck was available for a French PWR 900. The main ENEA effort in this period was devoted to prepare the MELCOR input deck using the code version v.1.8.5 (build QZ Oct 2000 with the latest patch 185003 Oct 2001). The input deck, completely new, was prepared taking into account structure, data and same conditions as those found inside ASTEC input decks. The main goal of the work presented in this paper is to put in evidence where and when MELCOR provides good enough results and why, in some cases mainly referring to its specific models (candling, corium pool behaviour, etc.) they were less good. A future work will be the preparation of an input deck for the new MELCOR 1.8.6. and to perform a code-to-code comparison with ASTEC v1.2 rev. 1. (author)

De Rosa, Felice [ENEA, Italian National Agency for New Technologies, Energy and the Environment (Italy)

2006-07-01T23:59:59.000Z

369

THERMAL ANALYSIS OF A 9975 PACKAGE IN A FACILITY FIRE ACCIDENT  

SciTech Connect

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.

Gupta, N.

2011-02-14T23:59:59.000Z

370

TRACE/PARCS Core Modeling of a BWR/5 for Accident Analysis of ATWS Events  

Science Conference Proceedings (OSTI)

The TRACE/PARCS computational package [1, 2] isdesigned to be applicable to the analysis of light water reactor operational transients and accidents where the coupling between the neutron kinetics (PARCS) and the thermal-hydraulics and thermal-mechanics (TRACE) is important. TRACE/PARCS has been assessed for itsapplicability to anticipated transients without scram(ATWS) [3]. The challenge, addressed in this study, is to develop a sufficiently rigorous input model that would be acceptable for use in ATWS analysis. Two types of ATWS events were of interest, a turbine trip and a closure of main steam isolation valves (MSIVs). In the first type, initiated by turbine trip, the concern is that the core will become unstable and large power oscillations will occur. In the second type,initiated by MSIV closure,, the concern is the amount of energy being placed into containment and the resulting emergency depressurization. Two separate TRACE/PARCS models of a BWR/5 were developed to analyze these ATWS events at MELLLA+ (maximum extended load line limit plus)operating conditions. One model [4] was used for analysis of ATWS events leading to instability (ATWS-I);the other [5] for ATWS events leading to emergency depressurization (ATWS-ED). Both models included a large portion of the nuclear steam supply system and controls, and a detailed core model, presented henceforth.

Cuadra A.; Baek J.; Cheng, L.; Aronson, A.; Diamond, D.; Yarsky, P.

2013-11-10T23:59:59.000Z

371

August 2003, Columbia Accident Investigation Report Volume I. Chapter 5-8  

NLE Websites -- All DOE Office Websites (Extended Search)

9 7 9 7 R e p o r t V o l u m e I A u g u s t 2 0 0 3 Part Two Why The Accident Occurred Many accident investigations do not go far enough. They identify the technical cause of the accident, and then connect it to a variant of "operator error" - the line worker who forgot to insert the bolt, the engineer who miscalculated the stress, or the manager who made the wrong decision. But this is sel-

372

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

1, 1995 1, 1995 IG-1 INFORMATION: Report on "Audit of the Department of Energy's Transportation Accident Resistant Container Program" The Secretary BACKGROUND: The U.S. Department of Energy (Department) has ultimate responsibility for the safety of all nuclear explosives and weapons operations conducted by the Department and its contractors. The Department also has joint responsibility for the safety of nuclear weapons in the custody of the Armed Services. Since the 1970s, the Department has designed, developed, and produced accident resistant containers to promote safety when transporting certain types of nuclear weapons by air. DISCUSSION: After successfully developing and modifying accident resistant containers for

373

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

Science Conference Proceedings (OSTI)

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

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

2006-07-01T23:59:59.000Z

374

Statistical description of heavy truck accidents on representative segments of interstate highway  

SciTech Connect

Any quantitative analysis of the risk of transportation accidents requires the use of many different statistical distributions. Included among these are the types of accidents which occur and the severity of these when they do occur. Several previous studies have derived this type of information for truck traffic over U. S. highways in general; these data are not necessarily applicable for the anticipated LMFBR spent fuel cask routes. This report presents data for highway segments representative of the specific LMFBR cask routes which are anticipated. These data are based upon a detailed record-by-record review of filed reports for accidents which occurred along the specified route segments.

Hartman, W.F.; Davidson, C.A.; Foley, J.T.

1977-01-01T23:59:59.000Z

375

Energy Use in Chinese Buildings: Views of an Outsider Looking...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Energy Use in Chinese Buildings: Views of an Outsider Looking In Energy Use in Chinese Buildings: Views of an Outsider Looking In Energy Use in Chinese Buildings: Views of an...

376

Semiclassical energy conditions  

E-Print Network (OSTI)

We present and develop several nonlinear energy conditions suitable for use in the semiclassical regime. In particular, we consider the recently formulated "flux energy condition" (FEC), and the novel "trace-of-square" (TOSEC) and "determinant" (DETEC) energy conditions. As we shall show, these nonlinear energy conditions behave much better than the classical linear energy conditions in the presence of semiclassical quantum effects. Moreover, whereas the quantum extensions of these nonlinear energy conditions seem to be quite widely satisfied as one enters the quantum realm, analogous quantum extensions are generally not useful for the linear classical energy conditions.

Martin-Moruno, Prado

2013-01-01T23:59:59.000Z

377

Source terms for analysis of accidents at a high level waste repository  

SciTech Connect

This paper describes an approach to identifying source terms from possible accidents during the preclosure phase of a high-level nuclear waste repository. A review of the literature on repository safety analyses indicated that source term estimation is in a preliminary stage, largely based on judgement-based scoping analyses. The approach developed here was to partition the accident space into domains defined by certain threshold values of temperature and impact energy density which may arise in potential accidents and specify release fractions of various radionuclides, present in the waste form, in each domain. Along with a more quantitative understanding of accident phenomenology, this approach should help in achieving a clearer perspective on scenarios important to preclosure safety assessments of geologic repositories. 18 refs., 3 tabs.

Mubayi, V.; Davis, R.E.; Youngblood, R.

1989-01-01T23:59:59.000Z

378

Human error and general aviation accidents: A comprehensive, fine-grained analysis using HFACS  

E-Print Network (OSTI)

The Human Factors Analysis and Classification System (HFACS) is a theoretically based tool for investigating and analyzing human error associated with accidents and incidents. Previous research performed at both at the University of Illinois and the Civil Aerospace Medical Institute (CAMI) have been highly successful and have shown that HFACS can be reliably used to analyze the underlying human causes of both commercial and general aviation (GA) accidents. these analyses have helped identify general trends in the types of human factors issues and aircrew errors that have contributed to civil aviation accidents. The next step is to identify the exact nature of the human errors identified. The purpose of this research effort, therefore, was to address these questions by performing a fine-grained HFACS analysis of the individual human causal factors associated with GA accidents and to assist in the generation of intervention programs. This report details those findings and offers an approach for developing interventions to address them.

Douglas A. Wiegmann; Albert Boquet; Cristy Detwiler; Kali Holcomb; Troy Faaborg; Douglas A. Wiegmann, Ph.D., Ph.D.; Albert Boquet, Ph.D.; Cristy Detwiler; Kali Holcomb; Troy Faaborg

2005-01-01T23:59:59.000Z

379

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

Science Conference Proceedings (OSTI)

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

WILLIAMS, J.C.

2003-11-15T23:59:59.000Z

380

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

SciTech Connect

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

Budnitz, R.J.; Lambert, H.E.; Apostolakis, G. [and others] and others

1998-04-01T23:59:59.000Z

Note: This page contains sample records for the topic "accident conditions view" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

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

E-Print Network (OSTI)

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

Yurko, Joseph Paul

2010-01-01T23:59:59.000Z

382

K Basins floor sludge retrieval system knockout pot basket fuel burn accident  

SciTech Connect

The K Basins Sludge Retrieval System Preliminary Hazard Analysis Report (HNF-2676) identified and categorized a series of potential accidents associated with K Basins Sludge Retrieval System design and operation. The fuel burn accident was of concern with respect to the potential release of contamination resulting from a runaway chemical reaction of the uranium fuel in a knockout pot basket suspended in the air. The unmitigated radiological dose to an offsite receptor from this fuel burn accident is calculated to be much less than the offsite risk evaluation guidelines for anticipated events. However, because of potential radiation exposure to the facility worker, this accident is precluded with a safety significant lifting device that will prevent the monorail hoist from lifting the knockout pot basket out of the K Basin water pool.

HUNT, J.W.

1998-11-11T23:59:59.000Z

383

Meteorological Factors Influencing the Radioactive Deposition in Finland after the Chernobyl Accident  

Science Conference Proceedings (OSTI)

After the accident at the Chernobyl nuclear plant on 26 April 1986, much of Europe was affected by radioactive pollution. The first releases were transported toward Scandinavia, where most of the fallout was attributable to wet deposition. This ...

Timo Puhakka; Kirsti Jylh; Pirkko Saarikivi; Jarmo Koistinen; Janne Koivukoski

1990-09-01T23:59:59.000Z

384

Lab Breakthrough: Desiccant Enhanced Evaporative Air Conditioning |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Desiccant Enhanced Evaporative Air Conditioning Desiccant Enhanced Evaporative Air Conditioning Lab Breakthrough: Desiccant Enhanced Evaporative Air Conditioning May 29, 2012 - 5:22pm Addthis This breakthrough combines desiccant materials, which remove moisture from the air using heat, and advanced evaporative technologies to develop a cooling unit that uses 90 percent less electricity and up to 80 percent less total energy than traditional air conditioning. This solution, called the desiccant enhanced evaporative air conditioner (DEVAP), also controls humidity more effectively to improve the comfort of people in buildings. View the entire Lab Breakthrough playlist. What are the key facts? Recent materials advances and liquid desiccant advances to design the compact and cost-effective DEVAP system. DEVAP uses 90 percent less electricity and up to 80 percent less

385

Lab Breakthrough: Desiccant Enhanced Evaporative Air Conditioning |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Lab Breakthrough: Desiccant Enhanced Evaporative Air Conditioning Lab Breakthrough: Desiccant Enhanced Evaporative Air Conditioning Lab Breakthrough: Desiccant Enhanced Evaporative Air Conditioning May 29, 2012 - 5:22pm Addthis This breakthrough combines desiccant materials, which remove moisture from the air using heat, and advanced evaporative technologies to develop a cooling unit that uses 90 percent less electricity and up to 80 percent less total energy than traditional air conditioning. This solution, called the desiccant enhanced evaporative air conditioner (DEVAP), also controls humidity more effectively to improve the comfort of people in buildings. View the entire Lab Breakthrough playlist. What are the key facts? Recent materials advances and liquid desiccant advances to design the compact and cost-effective DEVAP system.

386

Most Viewed Documents - Fission and Nuclear Technologies | OSTI...  

Office of Scientific and Technical Information (OSTI)

design bases issues in ASME Code Case N-47 Huddleston, R.L.; Swindeman, R.W. (Oak Ridge National Lab., TN (United States)) (1993) Severe accidents in spent fuel pools in support...

387

Criteria for calculating the efficiency of deep-pleated HEPA filters with aluminum separators during and after design basis accidents  

SciTech Connect

The authors have reviewed the literature on the performance of high efficiency particulate air (HEPA) filters under normal and abnormal conditions to establish criteria for calculating the efficiency of HEPA filters in a DOE nonreactor nuclear facility during and after a Design Basis Accident (DBA). This study is only applicable to the standard deep-pleated HEPA filter with aluminum separators as specified in ASME N509. The literature review included the performance of new filters and parameters that may cause deterioration in the filter performance such as filter age, radiation, corrosive chemicals, seismic and rough handling, high temperature, moisture, particle clogging, high air flow and pressure pulses. The deterioration of the filter efficiency depends on the exposure parameters; in severe exposure conditions the filter will be structurally damaged and have a residual efficiency of 0%. Despite the many studies on HEPA filter performance under adverse conditions, there are large gaps and limitations in the data that introduce significant error in the estimates of HEPA filter efficiencies under DBA conditions. Because of this limitation, conservative values of filter efficiency were chosen when there was insufficient data.

Bergman, W. [Lawrence Livermore National Lab., CA (United States); First, M.W. [Harvard School of Public Health, Boston, MA (United States); Anderson, W.L. [Anderson (W.L.), LaPlata, MD (United States); Gilbert, H. [Gilbert (H.), McLean, VA (United States); Jacox, J.W. [Jacox (J.W.), Columbus, OH (United States)

1995-02-01T23:59:59.000Z

388

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

SciTech Connect

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.

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

2012-09-30T23:59:59.000Z

389

Valley View Wind Farm | Open Energy Information  

Open Energy Info (EERE)

Wind Farm Wind Farm Jump to: navigation, search Name Valley View Wind Farm Facility Valley View Sector Wind energy Facility Type Commercial Scale Wind Facility Status In Service Owner Juhl Wind Developer Valley View Transmission Energy Purchaser Xcel Energy Location Outside Chandler MN Coordinates 43.905808°, -96.020508° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":43.905808,"lon":-96.020508,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

390

Business Apps List View | Data.gov  

NLE Websites -- All DOE Office Websites (Extended Search)

Apps List View Apps List View BusinessUSA Data/Tools Apps Challenges Let's Talk BusinessUSA You are here Data.gov » Communities » BusinessUSA List View Showing 1 - 12 of 12 results. Resources sort ascending Type Last Updated On Traveling Entrepreneur Find green government opportunities for small businesses. The app based on your current location retrieves all the government programs and then allows you to share your favorite program on Facebook. Mobile 02/01/2012 Small Business Toolkit Small Business Toolbox provides a way to easily keep up-to-date on government programs, grants, awards, solicitations, and small business programs. Find the closest SBA district office, easily gathering license and permit information. Mobile 01/31/2012 SBIR.gov Awards Search This app allows users to search the SBIR awards database by agency, keyword, and year.

391

Bibliography for nuclear criticality accident experience, alarm systems, and emergency management  

SciTech Connect

The characteristics, detection, and emergency management of nuclear criticality accidents outside reactors has been an important component of criticality safety for as long as the need for this specialized safety discipline has been recognized. The general interest and importance of such topics receives special emphasis because of the potentially lethal, albeit highly localized, effects of criticality accidents and because of heightened public and regulatory concerns for any undesirable event in nuclear and radiological fields. This bibliography lists references which are potentially applicable to or interesting for criticality alarm, detection, and warning systems; criticality accident emergency management; and their associated programs. The lists are annotated to assist bibliography users in identifying applicable: industry and regulatory guidance and requirements, with historical development information and comments; criticality accident characteristics, consequences, experiences, and responses; hazard-, risk-, or safety-analysis criteria; CAS design and qualification criteria; CAS calibration, maintenance, repair, and testing criteria; experiences of CAS designers and maintainers; criticality accident emergency management (planning, preparedness, response, and recovery) requirements and guidance; criticality accident emergency management experience, plans, and techniques; methods and tools for analysis; and additional bibliographies.

Putman, V.L.

1995-09-01T23:59:59.000Z

392

Application of RELAP/SCDAPSIM and COCOSYS Codes for Severe Accident Analysis in RBMK-1500 Reactor  

Science Conference Proceedings (OSTI)

Regardless low probability of occurrence the severe accident phenomena are investigated for all types of nuclear reactors in the world because the consequences of such accident could be catastrophic. Most of research is performed for the prevailing vessel-type light water reactors like PWRs and BWRs. Less research is performed for the channel-type reactors like CANDUs and RBMKs as they are operated just in a few countries. Up to now the phenomena that could occur in case of a severe accident in RBMK reactors were not analysed in detail and little literature is available on this topic. The paper presents one of the first integrated analyses of severe accident in RBMK-1500 reactor. RELAP/SCDAPSIM code is used to simulate the phenomena in the reactor core and reactor cooling system and COCOSYS code is used to simulate the confinement phenomena during the same accident scenario. The performed analysis provided information regarding code acceptability for the severe accident analysis in RBMK reactor and assessment of the timing of the key events, i.e. core uncover, fuel cladding rupture, etc, and provided assessment regarding hydrogen distribution in confinement. (authors)

Urbonavicius, E.; Uspuras, E.; Rimkevicius, S.; Kaliatka, A. [Lithuanian Energy Institute, Breslaujos g. 3, LT-44403 Kaunas (Lithuania)

2006-07-01T23:59:59.000Z

393

Accident investigation of the electrical shock incident at the PG and E PVUSA site Davis, California  

DOE Green Energy (OSTI)

This report summarizes the findings of the Accident Investigation Team (Team) assembled in response to a request from Pacific Gas and Electric Company (PG and E) to the US Department of Energy (DOE) to understand the events surrounding the electric shock of a worker at the PVUSA site in Davis, California and to provide recommendations to prevent such events from recurring. The report gives complete details on the sequence of events surrounding the accident and identifies 27 facts related to accident itself. Four technical deficiencies in the electrical systems which require further investigation were identified. The Team believes that the root cause of this accident was related to the absence of a proactive organizational entity responsible for overall health and safety on the site. Two contributing factors were identified. First, the prototype nature and associated operational difficulties of the electrical inverter resulted in large maintenance demands. Second, several of the injured employee`s co-workers noted that he occasionally failed to use appropriate personal protective equipment, but they never reported this practice to management. The direct cause of this accident was the failure of the injured employee to wear appropriate personal protective equipment (i.e., rubber gloves). Based on the review of the facts established in this investigation, five recommendations are presented to the funding agencies to reduce the possibility of future accidents at the PVUSA site.

Jacobson, L.; Moskowitz, P.D.; Garrett, J.O.; Tyler, R.

1992-02-01T23:59:59.000Z

394

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

Science Conference Proceedings (OSTI)

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

NONE

1999-04-01T23:59:59.000Z

395

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

Science Conference Proceedings (OSTI)

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

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

1982-11-01T23:59:59.000Z

396

Air Conditioning and lungs  

NLE Websites -- All DOE Office Websites (Extended Search)

Air Conditioning and lungs Name: freeman Status: NA Age: NA Location: NA Country: NA Date: Around 1993 Question: What affect does air conditioning have upon the lungs of the...

397

Field Guide: Inspection of Transmission Line Grounding Systems (Optimized for Electronic Viewing)  

Science Conference Proceedings (OSTI)

This EPRI report, one in a series of practical guides designed as reference aids for utility personnel working in the field, visually catalogs the various condition issues that commonly affect transmission line grounding systems. It presents photographs and short written descriptions of the conditions and lists associated causes, failure modes, and impacts. This field guide has been optimized for viewing on electronic devices. For a standard PDF or copy of this product printed on high-quality paper ...

2011-12-09T23:59:59.000Z

398

Evaluating Alternative "Countermeasures" Against Food Contamination Resulting From Nuclear Accidents  

E-Print Network (OSTI)

Nuclear accidents such as Chernobyl have far reaching impacts on ecological systems. Likewise they have major implications for agricultural systems, since crops and livestock can become contaminated and rendered unfit for human consumption. A range of `countermeasures' exists however, which can mitigate these impacts and allow food products to be saved. The CESER project has been concerned with the development of a system to assess the environmental side-effects of such countermeasures. Estimates of the economic costs of these environmental side-effects have been made for a number of case study sites in the UK, using environmental models and an original contingent valuation study. Estimates of farm level (private) costs are also included. 1. Professor, Department of Economics, University of Glasgow, Glasgow G12 8RT 2. Senior Lecturer, Department of Environmental Science, University of Stirling, Stirling FK9 4LA 3. former Research Associate, Department of Environmental Science, University of Stirling 4. former Research Associate, Department. of Environmental Science, University of Stirling. Manuscript date: 4/10/2000. Acknowledgements An earlier version of this paper was presented to the Agricultural Economics Society conference, Belfast, March 1999. The CESER project was funded by the European Union's Fourth Framework, Nuclear Fission Safety Programme (DGXII). Several academic institutions from across Europe participated in this project, including the University of Stirling, University of Bremen, Finnish Environment Institute, North-Trondelag College (Norway) and University of Salzburg. David Aitchison and Bill Jamieson of the cartography unit in the Department of Environmental Science, University of Stirling created the images used in the contingent valuation su...

Nick Hanley; Carol A. Salt; Mike Wilson; Meara Culligan-Dunsmore

2000-01-01T23:59:59.000Z

399

Sharing the viewing experience through second screens  

Science Conference Proceedings (OSTI)

Despite the ever expanding forms of digital entertainment and the emergence of consumer recording facilities, allowing viewers to time shift their TV viewing habits, there are still certain TV shows and events that create an audience desire to be part ... Keywords: interactive, mobile, narrative, performance, second screen, shared experience, television, twitter

Mark Lochrie; Paul Coulton

2012-07-01T23:59:59.000Z

400

Book review A Clear View of  

E-Print Network (OSTI)

Magazine R1 Book review A Clear View of the Cell Cycle James E. Ferrell, Jr. The Cell Cycle." It was a timely book, coming right on the heels of the discovery that cyclin-dependent kinases (Cdks) drive cell book. Enter "The Cell Cycle: Principles of Control", by David Morgan, just published by New Science

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401

Conditional belief types  

E-Print Network (OSTI)

We study type spaces where a players type at a state is a conditional probability on the space. We axiomatize these type spaces using conditional belief operators, and examine three additional axioms of increasing strength. First, introspection, which requires the agent to be unconditionally certain of her beliefs. Second, echo, according to which the unconditional beliefs implied by the condition must be held given the condition. Third, determination, which says that the conditional beliefs are the unconditional beliefs that are conditionally certain. The echo axiom implies that conditioning on an event is the same as conditioning on the event being certain, which formalizes the standard informal interpretation of conditioning in probability theory. The echo axiom also implies that the conditional probability given an event is a prior of the unconditional probability. The game-theoretic application of our model, which we treat in the context of an example, sheds light on a number of basic issues in the analysis of extensive form games. Type spaces are closely related to the sphere models of counterfactual conditionals and to models of hypothetical knowledge, and we discuss these relationships in detail.

Alfredo Di; Tillio Joseph; Y. Halpern; Dov Samet

2013-01-01T23:59:59.000Z

402

Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1: Evaluation of severe accident risks for plant operational state 5 during a refueling outage. Supporting MELCOR calculations, Volume 6, Part 2  

SciTech Connect

To gain a better understanding of the risk significance of low power and shutdown modes of operation, the Office of Nuclear Regulatory Research at the NRC established programs to investigate the likelihood and severity of postulated accidents that could occur during low power and shutdown (LP&S) modes of operation at commercial nuclear power plants. To investigate the likelihood of severe core damage accidents during off power conditions, probabilistic risk assessments (PRAs) were performed for two nuclear plants: Unit 1 of the Grand Gulf Nuclear Station, which is a BWR-6 Mark III boiling water reactor (BWR), and Unit 1 of the Surry Power Station, which is a three-loop, subatmospheric, pressurized water reactor (PWR). The analysis of the BWR was conducted at Sandia National Laboratories while the analysis of the PWR was performed at Brookhaven National Laboratory. This multi-volume report presents and discusses the results of the BWR analysis. The subject of this part presents the deterministic code calculations, performed with the MELCOR code, that were used to support the development and quantification of the PRA models. The background for the work documented in this report is summarized, including how deterministic codes are used in PRAS, why the MELCOR code is used, what the capabilities and features of MELCOR are, and how the code has been used by others in the past. Brief descriptions of the Grand Gulf plant and its configuration during LP&S operation and of the MELCOR input model developed for the Grand Gulf plant in its LP&S configuration are given.

Kmetyk, L.N.; Brown, T.D. [Sandia National Labs., Albuquerque, NM (United States)

1995-03-01T23:59:59.000Z

403

Multi-component gas transport in CANDU fuel rods during severe accidents.  

DOE Green Energy (OSTI)

The multi-component transport of steam, hydrogen and stable fission gas in the fuel-to-clad gap of defective CANDU fuel rods, during severe accident conditions, is investigated. Based on a general Stefan-Maxwell treatment this work considers how incoming steam will diffuse into a breached rod against a counter-current flow of non-condensable fission gases and out-flowing hydrogen that is produced from the internal reaction of steam with the Zircaloy cladding or urania. The ability of the oxidized clad to act as a physical barrier to either hydrogen or oxygen diffusion was further investigated in the current work with a molecular-dynamics approach, with the interactions between atoms represented by a Modified Embedded Atom Method. During the initial Zircaloy oxidation phase in the CRL experiments, the model was able to predict the reduced fission product release kinetics as well as the timing for the completion of the clad-oxidation process. In this simulation, the model (with an effective gap size of 20 {micro}m) was able to successfully predict whether singlesided or double-sided oxidation had occurred in accordance with the metallographic examination. However, in order to account for the observed release kinetics after the completion of clad oxidation, it was necessary to assume a greater atmospheric exchange due to possible cracking of the brittle oxide layer. With the assumption of cracking (by assuming a reduced path length for gas transport), the model was successfully able to reproduce the fission product release kinetics and the final fuel stoichiometry as determined from end-of-test weight gain measurements. This analysis particularly shows that local hydrogen production (from the internal fuel oxidation process) will result in a reduced local oxygen potential in the fuel-to-clad gap compared to that which occurs in the bulk coolant.

Szpunar, B; Lewis, B. J.; Arimescu, V. I.; Dickson, R. S.; Dickson, L. W.; Baskes, M. I. (Michael I.)

2001-01-01T23:59:59.000Z

404

The view-selection problem has an exponential-time lower bound for conjunctive queries and views  

Science Conference Proceedings (OSTI)

The view-selection problem is to design and materialize a set of views over a database schema, such that the choice of views minimizes the cost of evaluating the selected workload of queries, and the combined size of the materialized views does not exceed ...

Rada Chirkova

2002-06-01T23:59:59.000Z

405

MODELING OF 2LIBH4 PLUS MGH2 HYDROGEN STORAGE SYSTEM ACCIDENT SCENARIOS USING EMPIRICAL AND THEORETICAL THERMODYNAMICS  

DOE Green Energy (OSTI)

It is important to understand and quantify the potential risk resulting from accidental environmental exposure of condensed phase hydrogen storage materials under differing environmental exposure scenarios. This paper describes a modeling and experimental study with the aim of predicting consequences of the accidental release of 2LiBH{sub 4}+MgH{sub 2} from hydrogen storage systems. The methodology and results developed in this work are directly applicable to any solid hydride material and/or accident scenario using appropriate boundary conditions and empirical data. The ability to predict hydride behavior for hypothesized accident scenarios facilitates an assessment of the of risk associated with the utilization of a particular hydride. To this end, an idealized finite volume model was developed to represent the behavior of dispersed hydride from a breached system. Semiempirical thermodynamic calculations and substantiating calorimetric experiments were performed in order to quantify the energy released, energy release rates and to quantify the reaction products resulting from water and air exposure of a lithium borohydride and magnesium hydride combination. The hydrides, LiBH{sub 4} and MgH{sub 2}, were studied individually in the as-received form and in the 2:1 'destabilized' mixture. Liquid water hydrolysis reactions were performed in a Calvet calorimeter equipped with a mixing cell using neutral water. Water vapor and oxygen gas phase reactivity measurements were performed at varying relative humidities and temperatures by modifying the calorimeter and utilizing a gas circulating flow cell apparatus. The results of these calorimetric measurements were compared with standardized United Nations (UN) based test results for air and water reactivity and used to develop quantitative kinetic expressions for hydrolysis and air oxidation in these systems. Thermodynamic parameters obtained from these tests were then inputted into a computational fluid dynamics model to predict both the hydrogen generation rates and concentrations along with localized temperature distributions. The results of these numerical simulations can be used to predict ignition events and the resultant conclusions will be discussed.

James, C; David Tamburello, D; Joshua Gray, J; Kyle Brinkman, K; Bruce Hardy, B; Donald Anton, D

2009-04-01T23:59:59.000Z

406

SILENE Benchmark Critical Experiments for Criticality Accident Alarm Systems  

SciTech Connect

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

Miller, Thomas Martin [ORNL; Reynolds, Kevin H. [Y-12 National Security Complex

2011-01-01T23:59:59.000Z

407

The lived experience of post-traumatic stress disorder as described by motor vehicle accident victims in Jordan.  

E-Print Network (OSTI)

??Aim: To explore the lived experience of post-traumatic stress disorder (PTSD) as described by individuals who have been involved in a motor vehicle accident (MVA) (more)

Al-Kofahy, Lilibeth

2011-01-01T23:59:59.000Z

408

Thermo-fluid-dynamics analysis of the unit 3 Fukushima Daiichi Accident with the RELAP5\\SCDAP code.  

E-Print Network (OSTI)

??The aim of this thesis is the analysis of the physical phenomena involved in the nuclear accident at Fukushima NPP. This study has been articulated (more)

VENTURI, FRANCESCO LINO

2012-01-01T23:59:59.000Z

409

Experimental and Analytical Simulation of MFCI (Molten Fuel Coolant Interaction) during CDA (Core Disruptive Accident) in Sodium Cooled Fast Reactor.  

E-Print Network (OSTI)

??With increasing demand for understanding Severe Accident Scenario in Sodium Cooled Fast Reactors, there is an urgent need of enhancing numerical and experimental simulation techniques. (more)

Natarajan, Venkataraman

2011-01-01T23:59:59.000Z

410

Terms and Conditions  

NLE Websites -- All DOE Office Websites (Extended Search)

Terms and Conditions Terms and Conditions Terms and Conditions As a premier national research and development laboratory, LANL seeks to do business with qualified companies that offer value and high quality products and services. Contact Small Business Office (505) 667-4419 Email Use information below as guideline to doing business An "Appendix SFA-1" contains FAR and DEAR Clauses that are incorporated by reference into a particular subcontract. "Exhibit A General Conditions" are the general terms and conditions applicable to a particular subcontract. Note: The contents of the SFA-1 and Exhibit A (below) are not the only terms and conditions that will be in a LANS subcontract but represent the terms that generally do not change in a particular type of procurement. The

411

Documentation for RISKIN: A risk integration code for MACCS (MELCOR Accident Consequence Code System) output  

Science Conference Proceedings (OSTI)

This document has been prepared as a user's guide for the computer program RISKIN developed at Sandia National Laboratories. The RISKIN code generates integrated risk tables and the weighted mean risk associated with a user-selected set of consequences from up to five output files generated by the MELCOR Accident Consequence Code System (MACCS). Each MACCS output file can summarize the health and economic consequences resulting from up to 60 distinct severe accident source terms. Since the accident frequency associated with these source terms is not included as a MACCS input parameter a postprocessor is required to derived results that must incorporate accident frequency. The RISKIN code is such a postprocessor. RISKIN will search the MACCS output files for the mean and peak consequence values and the complementary cumulative distributive function (CCDF) tables for each requested consequence. Once obtained, RISKIN combines this data with accident frequency data to produce frequency weighted results. A postprocessor provides RISKIN an interface to the proprietary DISSPLA plot package. The RISKIN code has been written using ANSI Standard FORTRAN 77 to maximize its portability.

Rollstin, J.A. (GRAM, Inc., Albuquerque, NM (USA)); Hong, Kou-John (Argonne National Lab., IL (USA))

1990-11-01T23:59:59.000Z

412

Testing and analysis of structural integrity of electrosleeved tubes under severe accident transients  

Science Conference Proceedings (OSTI)

The structural integrity of flawed steam generator tubing with Electrosleeves{trademark} under simulated severe accident transients was analyzed by analytical models that used available material properties data and results from high-temperature tests conducted on Electrosleeved tubes. The Electrosleeve material is almost pure Ni and derives its strength and other useful properties from its nanocrystalline microstructure, which is stable at reactor operating temperatures. However, it undergoes rapid grain growth, at the high temperatures expected during severe accidents, resulting in a loss of strength and a corresponding decrease in flow stress. The magnitude of this decrease depends on the time-temperature history during the accident. Failure tests were conducted at ANL and FTI on internally pressurized Electrosleeved tubes with 80% and 100% throughwall machined axial notches in tie parent tubes that were subjected to simulated severe accident temperature transients. The test results, together with the analytical model, were used to estimate the unaged flow stress curve of the Electrosleeved material at high temperatures. Failure temperatures for Electrosleeved tubes with throughwall and part-throughwall axial cracks of various lengths in the parent tubes were calculated for a postulated severe accident transient.

Majumdar, S.

1999-12-10T23:59:59.000Z

413

consumer apps list view | Data.gov  

NLE Websites -- All DOE Office Websites (Extended Search)

consumer apps list view consumer apps list view Consumer Data Apps Challenges Resources About Blogs Let's Talk Feedback Consumer You are here Data.gov » Communities » Consumer Smart Disclosure Apps This page highlights apps and websites that use Smart Disclosure-style data to empower consumers to make better informed choices. The purpose is to illustrate the kinds of innovative apps-web- and mobile-based-that Smart Disclosure can fuel. The galleries below include both government-produced apps and apps created by innovators outside government that have won Federal challenges. Showing 1 - 19 of 19 results. Resources sort ascending Type Last Updated On Smart Traveler Smart Traveler, the official State Department app for U.S. travelers, invites you to see the world with easy access to frequently updated official country information, travel alerts, travel warnings, maps, U.S. embassy locations, and more. Mobile 09/14/2012

414

SOAJ Search : Main View : Deep Federated Search  

Office of Scientific and Technical Information (OSTI)

SOAJ Search SOAJ Search Search Powered By Deep Web Technologies New Search Preferences Powered by Deep Web Technologies HOME ABOUT ADVANCED SEARCH CONTACT US HELP Science Open Access Journals (SOAJ) Science Open Access Journals Main View This view is used for searching all possible sources. Additional Information Keyword: Title: Additional Information Author: Fields to Match: All Any Field(s) Additional Information Date Range: Beginning Date Range Pick Year 2013 2012 2011 2010 2009 2008 2007 2006 2005 2004 2003 2002 2001 2000 1999 1998 1997 1996 1995 1994 1993 1992 1991 1990 toEnding Date Range Pick Year 2013 2012 2011 2010 2009 2008 2007 2006 2005 2004 2003 2002 2001 2000 1999 1998 1997 1996 1995 1994 1993 1992 1991 1990 DWT Logo Search Clear All Help Simple Search Select All

415

Reasearch apps list view | Data.gov  

NLE Websites -- All DOE Office Websites (Extended Search)

Reasearch apps list view Reasearch apps list view Research Menu Data/Tools Apps Resources Let's Talk Research Alpha You are here Data.gov » Communities » Research Applications Showcase Showing 1 - 6 of 6 results. Resources sort ascending Type Last Updated On Wireless Spectrum Research & Development Senior Steering Group's Testbed Information Portal The Wireless Spectrum Research & Development Senior Steering Group's Testbed Information Portal provides information about spectrum testing facilities to government, academic, and industry researchers in need of such facilities. Apps 11/20/2012 The R&D Dashboard The R&D Dashboard beta website provides an initial look at U.S. Federal investments in science and research from two agencies: the National Institutes of Health (NIH) and the National Science Foundation (NSF) from years 2000-2009. The R&D Dashboard will expand in a future iteration to include ALL Federal research and development spending and expanded information on outputs. Apps 11/20/2012

416

Movements in air conditioning.  

E-Print Network (OSTI)

??Movements in Air Conditioning is a collection of poems that explores the obstacles inherent in creating a new sense of home in a country that (more)

Hitt, Robert D. (Robert David)

2013-01-01T23:59:59.000Z

417

Operating Conditions Opportunity  

NLE Websites -- All DOE Office Websites (Extended Search)

Components in Aggressive Operating Conditions Opportunity Research is active on the patent-pending technology, titled "3-Dimensional Functionally Gradient Coatings for...

418

An nual En ergy Re view 2001  

Gasoline and Diesel Fuel Update (EIA)

An nual En ergy Re view 2001 An nual En ergy Re view 2001 The An nual En ergy Re view (AER) pres ents the En ergy In for ma tion Ad min - is tra tion's his tor i cal en ergy sta tis tics. For many se ries, sta tis tics are given for ev ery year from 1949 through 2001. The sta tis tics, ex pressed in ei ther phys i cal units or Brit ish ther mal units, cover all ma jor en ergy ac tiv i ties, in - clud ing con sump tion, pro duc tion, trade, stocks, and prices, for all ma jor en - ergy com mod i ties, in clud ing fos sil fu els, elec tric ity, and re new able en ergy sources. Pub li ca tion of this re port is re quired un der Pub lic Law 95-91 (De part ment of En ergy Or ga ni za tion Act), Sec tion 205(c), and is in keep ing with re spon - si bil i ties given to the En ergy In for ma tion Ad min is tra tion un der Sec tion

419

Assessment of the potential for ferrocyanide propagating reaction accidents  

Science Conference Proceedings (OSTI)

This report contains safety criteria for the storage of ferrocyanide bearing waste sludges in Hanford underground waste storage tanks. In addition, the tank wastes are categorized with this criteria into SAFE, CONDITIONALLY SAFE, and UNSAFE categories based on available historical records and sample information. Fourteen (14) tanks are classified as CONDITIONALLY SAFE, while four (4) C-Farm tanks are categorized as SAFE. This report therefore provides a technical basis to resolve the ferrocyanide safety issue for these four tanks and supports their removal from the Watch List. The 14 CONDITIONALLY SAFE tanks will be re-evaluated in a future revision to this report as representative sample data becomes available. It is anticipated that the 14 tanks will be re-categorized as SAFE at that time.

Meacham, J.E.; Cash, R.J.; Dickinson, D.R. [and others

1996-01-01T23:59:59.000Z

420

Simplified Space Conditioning  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Simplified Space Conditioning Simplified Space Conditioning Duncan Prahl, RA IBACOS, Inc. Building America Technical Update April 29, 2013 Simplified Space Conditioning Rethinking HVAC Design * Traditional Method - Assume envelope losses dictate the load - Room by room load analysis - Pick Equipment and distribute to meet the load in each room * New Method - Consider how the occupants live in the building - Seriously consider internal gains in both heating and cooling - Consider ventilation strategy - Design system Simplified Space Conditioning If you are: * A production builder * Participating in "above code" programs * Following ACCA Manual RS or ASHRAE 55 * Need to prove "delivering heat to each habitable room" * Concerned about litigation * Play it safe, Use Manual J, S & D and condition every

Note: This page contains sample records for the topic "accident conditions view" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


421

U.S. Department of Energy Accident Resistant SiC Clad Nuclear Fuel  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

U.S. Department of Energy Accident Resistant SiC Clad Nuclear Fuel U.S. Department of Energy Accident Resistant SiC Clad Nuclear Fuel Development U.S. Department of Energy Accident Resistant SiC Clad Nuclear Fuel Development A significant effort is being placed on silicon carbide ceramic matrix composite (SiC CMC) nuclear fuel cladding by Light Water Reactor Sustainability (LWRS) Advanced Light Water Reactor Nuclear Fuels Pathway. The intent of this work is to invest in a high-risk, high-reward technology that can be introduced in a relatively short time. The LWRS goal is to demonstrate successful advanced fuels technology that suitable for commercial development to support nuclear relicensing. Ceramic matrix composites are an established non-nuclear technology that utilizes ceramic fibers embedded in a ceramic matrix. A thin interfacial layer between the

422

WIPP Workers Reach Two Million Man-Hours Without a Lost-Time Accident  

NLE Websites -- All DOE Office Websites (Extended Search)

Workers Reach Two Million Man-Hours Workers Reach Two Million Man-Hours Without a Lost-Time Accident CARLSBAD, N.M., February 22, 2001 - Workers at the U.S. Department of Energy's (DOE) Waste Isolation Pilot Plant (WIPP) reached a safety milestone Feb. 19 by working two million man-hours without a lost-time accident. According to the National Safety Council, facilities with the same industry code as WIPP lose an average of 20.6 workdays (or 164.8 man-hours) a year to accidents. "Safety is at the core of all WIPP operations," said Dr. Inés Triay, Manager of DOE's Carlsbad Field Office. "We are particularly pleased that WIPP workers reached the two million mark during the time in which they mined a new panel and increased shift work." "To make safety a number one priority means more than creating a safe

423

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

Science Conference Proceedings (OSTI)

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

Abdullah, Ade Gafar [Physics Dept. Institut Teknologi Bandung (ITB), Jl. Ganesha 10 Bandung (Indonesia); Electrical Engineering Dept. Universitas Pendidikan Indonesia (UPI), Jl. Dr. Setiabudhi 229 Bandung (Indonesia); Su'ud, Zaki; Kurniadi, Rizal; Kurniasih, Neny [Physics Dept. Institut Teknologi Bandung (ITB), Jl. Ganesha 10 Bandung (Indonesia); Yulianti, Yanti [Physics Dept. Institut Teknologi Bandung (ITB), Jl. Ganesha 10 Bandung (Indonesia); Physics Dept. Universitas Lampung (UNILA), Jl. Sumantri Brojonegoro 1 Bandar Lampung (Indonesia)

2010-12-23T23:59:59.000Z

424

Slide Rule for Rapid Response Estimation of Radiological Dose from Criticality Accidents  

SciTech Connect

This paper describes a functional slide rule that provides a readily usable ?in-hand? method for estimating nuclear criticality accident information from sliding graphs, thereby permitting (1) the rapid estimation of pertinent criticality accident information without laborious or sophisticated calculations in a nuclear criticality emergency situation, (2) the appraisal of potential fission yields and external personnel radiation exposures for facility safety analyses, and (3) a technical basis for emergency preparedness and training programs at nonreactor nuclear facilities. The slide rule permits the estimation of neutron and gamma dose rates and integrated doses based upon estimated fission yields, distance from the fission source, and time-after criticality accidents for five different critical systems. Another sliding graph permits the estimation of critical solution fission yields based upon fissile material concentration, critical vessel geometry, and solution addition rate. Another graph provides neutron and gamma dose-reduction factors for water, steel, and concrete shields.

Broadhead, B.L.; Childs, R.L.; Hopper, C.M.; Parks, C.V.

1999-09-20T23:59:59.000Z

425

RELAP5 Application to Accident Analysis of the NIST Research Reactor  

Science Conference Proceedings (OSTI)

Detailed safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The time-dependent analysis of the primary system is determined with a RELAP5 transient analysis model that includes the reactor vessel, the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. A post-processing of the simulation results has been conducted to evaluate minimum critical heat flux ratio (CHFR) using the Sudo-Kaminaga correlation. Evaluations are performed for the following accidents: (1) the control rod withdrawal startup accident and (2) the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur because of sufficient coolant flow through the fuel channels and the negative scram reactivity insertion.

Baek, J.; Cuadra Gascon, A.; Cheng, L.Y.; Diamond, D.

2012-03-18T23:59:59.000Z

426

Estimates of the financial consequences of nuclear-power-reactor accidents  

SciTech Connect

This report develops preliminary techniques for estimating the financial consequences of potential nuclear power reactor accidents. Offsite cost estimates are based on CRAC2 calculations. Costs are assigned to health effects as well as property damage. Onsite costs are estimated for worker health effects, replacement power, and cleanup costs. Several classes of costs are not included, such as indirect costs, socio-economic costs, and health care costs. Present value discounting is explained and then used to calculate the life cycle cost of the risks of potential reactor accidents. Results of the financial consequence estimates for 156 reactor-site combinations are summarized, and detailed estimates are provided in an appendix. The results indicate that, in general, onsite costs dominate the consequences of potential accidents.

Strip, D.R.

1982-09-01T23:59:59.000Z

427

Evaluation of severe accident risks, Peach Bottom, Unit 2: Main report  

Science Conference Proceedings (OSTI)

In support of the Nuclear Regulatory Commission's (NRC's) assessment of the risk from severe accidents at commercial nuclear power plants in the US reported NUREG-1150, the Severe Accident Risk Reduction Program (SARRP) has completed a revised calculation of the risk to the general public from severe accidents at the Peach Bottom Atomic Power Station, Unit 2. This power plant, located in southeastern Pennsylvania, is operated by the Philadelphia Electric Company. The emphasis in this risk analysis was not on determining a so-called'' point estimate of risk. Rather, it was to determine the distribution of risk, and to discover the uncertainties that account for the breadth of this distribution. Off-site risk initiated by events both internal and external to the power station were assessed. 39 refs., 174 figs., 133 tabs.

Payne, A.C.; Breeding, R.J.; Jow, H.N.; Shiver, A.W. (Sandia National Labs., Albuquerque, NM (USA)); Helton, J.C. (Arizona State Univ., Tempe, AZ (USA)); Smith, L.N. (Science Applications International Corp., Albuquerque, NM (USA))

1990-12-01T23:59:59.000Z

428

Management of Ultimate Risk of Nuclear Power Plants by Source Terms - Lessons Learned from the Chernobyl Accident  

Science Conference Proceedings (OSTI)

The term 'ultimate risk' is used here to describe the probabilities and radiological consequences that should be incorporated in siting, containment design and accident management of nuclear power plants for hypothetical accidents. It is closely related with the source terms specified in siting criteria which assures an adequate separation of radioactive inventories of the plants from the public, in the event of a hypothetical and severe accident situation. The author would like to point out that current source terms which are based on the information from the Windscale accident (1957) through TID-14844 are very outdated and do not incorporate lessons learned from either the Three Miles Island (TMI, 1979) nor Chernobyl accident (1986), two of the most severe accidents ever experienced. As a result of the observations of benign radionuclides released at TMI, the technical community in the US felt that a more realistic evaluation of severe reactor accident source terms was necessary. In this background, the 'source term research project' was organized in 1984 to respond to these challenges. Unfortunately, soon after the time of the final report from this project was released, the Chernobyl accident occurred. Due to the enormous consequences induced by then accident, the one time optimistic perspectives in establishing a more realistic source term were completely shattered. The Chernobyl accident, with its human death toll and dispersion of a large part of the fission fragments inventories into the environment, created a significant degradation in the public's acceptance of nuclear energy throughout the world. In spite of this, nuclear communities have been prudent in responding to the public's anxiety towards the ultimate safety of nuclear plants, since there still remained many unknown points revolving around the mechanism of the Chernobyl accident. In order to resolve some of these mysteries, the author has performed a scoping study of the dispersion and deposition mechanisms of fuel particles and fission fragments during the initial phase of the Chernobyl accident. Through this study, it is now possible to generally reconstruct the radiological consequences by using a dispersion calculation technique, combined with the meteorological data at the time of the accident and land contamination densities of {sup 137}Cs measured and reported around the Chernobyl area. Although it is challenging to incorporate lessons learned from the Chernobyl accident into the source term issues, the author has already developed an example of safety goals by incorporating the radiological consequences of the accident. The example provides safety goals by specifying source term releases in a graded approach in combination with probabilities, i.e. risks. The author believes that the future source term specification should be directly linked with safety goals. (author)

Genn Saji [Ex-Secretariate of Nuclear Safety Commission of Japan (Japan)

2006-07-01T23:59:59.000Z

429

Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.  

SciTech Connect

This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the experienced user-base and the experimental validation base was decaying away quickly.

Carbajo, Juan (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin, Madison, WI); Schmidt, Rodney Cannon; Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Ludewig, Hans (Brookhaven National Laboratory, Upton, NY); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Serre, Frederic (Centre d'%C3%94etudes nucl%C3%94eaires de Cadarache %3CU%2B2013%3E CEA, France)

2011-06-01T23:59:59.000Z

430

Study on severe accident fuel dispersion behavior in the Advanced Neutron Source reactor at Oak Ridge National Laboratory  

SciTech Connect

Core flow blockage events are a leading contributor to core damage initiation risk in the Advanced Neutron Source (ANS) reactor. During such an accident, insufficient cooling of the fuel could result in core heatup and melting under full coolant flow condition. Coolant inertia forces acting on the melt surface would likely break up the melt into small particles. Under thermal-hydraulic conditions of ANS coolant channel, micro-fine melt particles are expected. Heat transfer between melt particle and coolant, which affects particle breakup, was studied. The study indicates that the thermal effect on melt fragmentation seems to be negligible because the time corresponding to the breakup due to hydrodynamic forces is much shorter than the time for the melt surface to solidify. The study included modeling and analyses to predict transient behavior and transport of debris particles throughout the coolant system. The transient model accounts for the surface forces acting on the particle that results from the pressure variation on the surface, inertia, virtual mass, viscous force due to relative motion of particle in the coolant, gravitation, and resistance due to inhomogenous coolant velocity radially across piping due to possible turbulent coolant motions. Results indicate that debris particles would reside longest in heat exchangers because of lower coolant velocity there. Also core debris tends to move together upon melting and entrainment.

Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V. [Oak Ridge National Lab., TN (United States); Xiang, J.Y. [Wabash Coll., Crawfordsville, IN (United States)

1995-12-31T23:59:59.000Z

431

Study on severe accident fuel dispersion behavior in the Advanced Neutron Source reactor at Oak Ridge National Laboratory  

SciTech Connect

Core flow blockage events have been identified as a leading contributor to core damage initiation risk in the Advanced Neutron Source (ANS) reactor. During such an accident, insufficient cooling of the fuel in a few adjacent blocked coolant channels out of several hundred channels, could also result in core heatup and melting under full coolant flow condition in other coolant channels. Coolant inertia forces acting on the melt surface would likely break up the melt into small particles. Under thermal-hydraulic conditions of ANS coolant channel, micro-fine melt particles are expected. Heat transfer between melt particle and coolant, which affects the particle breakup characteristics, was studied. The study indicates that the thermal effect on melt fragmentation seems to be negligible because the time corresponding to the breakup due to hydrodynamic forces is much shorter than the time for the melt surface to solidify. The study included modeling and analyses to predict transient behavior and transport of debris particles throughout the coolant system. The transient model accounts for the surface forces acting on the particle that result from the pressure variation on the surface, inertia, virtual mass, viscous force due to the relative motion of the particle in the coolant, gravitation, and resistance due to inhomogeneous coolant velocity radially across piping due to expected turbulent coolant motions. The results indicate that debris particles would reside longest in the heat exchangers because of lower coolant velocity there. Also they are entrained and move together in a cloud.

Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V.

1995-09-01T23:59:59.000Z

432

3D View Inside the Skeleton with X-ray Microscopy: Imaging Bone at the  

NLE Websites -- All DOE Office Websites (Extended Search)

3D View Inside the Skeleton with X-ray Microscopy: Imaging Bone 3D View Inside the Skeleton with X-ray Microscopy: Imaging Bone at the Nanoscale Scientists studying osteoporosis and other skeletal diseases are interested in the 3D structure of bone and its responses to conditions such as weightlessness, radiation (of particular interest to astronauts) and vitamin D deficiency. The current gold standard, micro-computed tomography (micro-CT), provides 3D images of trabeculae, the small interior struts of bone tissue, and electron microscopy can provide nanometer resolution of thin tissue slices. Hard X-ray transmission microscopy has provided the first 3D view of bone structure within individual trabeculae on the nanoscale. figure 1 Figure 1 Micro-CT (left) shows trabecular structure inside of bone. Transmission X-ray microscopy (TXM; center and right) can reveal localized details of osteocyte lacunae and their processes.

433

Importance of emergency response actions to reactor accidents within a probabilistic consequence assessment model  

Science Conference Proceedings (OSTI)

An uncertainty and sensitivity analysis of early health consequences of severe accidents at nuclear power plants as a function of the emergency response parameters has been performed using a probabilistic consequence assessment code. The importance of various emergency response parameters in predicting the consequences for a range of accident source terms was determined through training a neural network algorithm which relates the sensitivity of the output to various choices of the input. Extensions of this approach should be helpful to planners in prioritizing the emergency responses at nuclear power plants.

Mubayi, V.; Neymotin, L.

1997-03-01T23:59:59.000Z

434

Technical Advisory Team (TAT) report on the rocket sled test accident of October 9, 2008.  

DOE Green Energy (OSTI)

This report summarizes probable causes and contributing factors that led to a rocket motor initiating prematurely while employees were preparing instrumentation for an AIII rocket sled test at SNL/NM, resulting in a Type-B Accident. Originally prepared by the Technical Advisory Team that provided technical assistance to the NNSA's Accident Investigation Board, the report includes analyses of several proposed causes and concludes that the most probable source of power for premature initiation of the rocket motor was the independent battery contained in the HiCap recorder package. The report includes data, evidence, and proposed scenarios to substantiate the analyses.

Stofleth, Jerome H.; Dinallo, Michael Anthony; Medina, Anthony J.

2009-01-01T23:59:59.000Z

435

Cost Sensitive Conditional Planning  

E-Print Network (OSTI)

While POMDPs provide a general platform for conditional planning under a wide range of quality metrics they have limited scalability. On the other hand, uniform probability conditional planners scale very well, but many lack the ability to optimize plan quality metrics. We present an innovation to planning graph based heuristics that helps uniform probability conditional planners both scale and generate high quality plans when using actions with non uniform costs. We make empirical comparisons with two state of the art planners to show the benefit of our techniques.

Daniel Bryce; Subbarao Kambhampati

2005-01-01T23:59:59.000Z

436

OpenEI:Neutral point of view | Open Energy Information  

Open Energy Info (EERE)

given to minor points of view. Neutral point of view is a principal core policy of the OpenEI platform. For further information, please refer to our model for this policy,...

437

JOBAID-VIEWING AN EMPLOYEE MATRIX (SUPERVISOR) | Department of...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

VIEWING AN EMPLOYEE MATRIX (SUPERVISOR) JOBAID-VIEWING AN EMPLOYEE MATRIX (SUPERVISOR) The purpose of this job aid is to guide supervisor users through the step-by-step process of...

438

Simulation of an ammonia plant accident using rigorous heterogeneous models: Effect of shift converters disturbances on the methanator  

Science Conference Proceedings (OSTI)

Disturbance introduced into the shift converters section of the ammonia production line may lead to problems in the ammonia production line which manifest themselves in other units of the production line. A real accident that took place in an ammonia ... Keywords: Accident, Ammonia, Catalytic reactors, Heterogeneous models, Modelling, Simulation

F. M. Alhabdan; S. S. E. H. Elnashaie

1995-02-01T23:59:59.000Z

439

Ballast Accidents Analysis and Evaluation of Urban Rail Transit Based on Method of Causality Analysis and Faulty Tree Analysis  

Science Conference Proceedings (OSTI)

Ballast casualty often incurs severe sequence once takes place, such as abnormal operation,, personnel injury or even death accident , especially for lines below grade. Causality Analysis and Fault Tree analysis method is applied to research of personnel ... Keywords: ballast accident, causality analysis, faulty tree analysis, urban rail transit

Jing He; Zhi-gang Liu

2009-04-01T23:59:59.000Z

440

ARM - Measurement - Surface condition  

NLE Websites -- All DOE Office Websites (Extended Search)

condition condition ARM Data Discovery Browse Data Comments? We would love to hear from you! Send us a note below or call us at 1-888-ARM-DATA. Send Measurement : Surface condition State of the surface, including vegetation, land use, surface type, roughness, and such; often provided in model output. Categories Surface Properties Instruments The above measurement is considered scientifically relevant for the following instruments. Refer to the datastream (netcdf) file headers of each instrument for a list of all available measurements, including those recorded for diagnostic or quality assurance purposes. ARM Instruments NAV : Navigational Location and Attitude SURFLOG : SGP Surface Conditions Observations by Site Technicians S-TABLE : Stabilized Platform MET : Surface Meteorological Instrumentation

Note: This page contains sample records for the topic "accident conditions view" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


441

Terms and Conditions  

NLE Websites -- All DOE Office Websites (Extended Search)

Terms and Conditions Terms and Conditions R&D Overview 100G Testbed Testbed Description Testbed Results Current Testbed Research Proposal Process Terms and Conditions Virtual Circuits (OSCARS) Performance (perfSONAR) Tools Development Green Networking Authentication & Trust Federation (ATF) Partnerships Contact Us Technical Assistance: 1 800-33-ESnet (Inside the US) 1 800-333-7638 (Inside the US) 1 510-486-7600 (Globally) 1 510-486-7607 (Globally) Report Network Problems: trouble@es.net Provide Web Site Feedback: info@es.net Terms and Conditions Researchers must provide ESnet copies of any articles, presentations, and publications based on testbed research for posting on the ESnet Testbed web site. All publications based on work conducted on the testbed must include the following statement:

442

Anemometry in Icing Conditions  

Science Conference Proceedings (OSTI)

The accuracy of wind measurements in icing conditions is discussed, and wind tunnel calibrations as well as field comparisons are presented for three heated anemometers that use different measuring principles. It is pointed out that ice-free ...

Lasse Makkonen; Pertti Lehtonen; Lauri Helle

2001-09-01T23:59:59.000Z

443

Calculation of hydrogen and oxygen uptake in fuel rod cladding during severe accidents using the integral diffusion method -- Preliminary design report  

DOE Green Energy (OSTI)

Preliminary designs are described for models of hydrogen and oxygen uptake in fuel rod cladding during severe accidents. Calculation of the uptake involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the cladding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental results are presented that show a rapid uptake of hydrogen in the event of dissolution of the oxide layer and a rapid release of hydrogen in the event of cracking of the oxide layer. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert's law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for hydrogen and oxygen uptake and cladding embrittlement into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the proposed models on the calculated behavior of fuel rods in severe accident conditions. This report is a revision and reissue of the report entitled; ``Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents.''

Siefken, L.J.

1999-02-01T23:59:59.000Z

444

Exponential Conditional Volatility Models  

E-Print Network (OSTI)

- relation function (ACF) is less straightforward than it is for a GARCHmodel, analytic expressions can be obtained and these expressions are more general. Speci?cally, formulae for the ACF of the (absolute values of ) the observations raised to any power can... proposes an exponential link function for the conditional mean in gamma and Weibull distributions. As well as setting out the conditions for the asymptotic theory to be valid, expressions for moments, ACFs and multi-step forecasts are derived. Leverage...

Harvey, Andrew

2010-08-24T23:59:59.000Z

445

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

E-Print Network (OSTI)

One of the primary reasons of the escalating rates of injuries and fatalities in the construction industry is the ever so complex, dynamic and continually changing nature of construction work. Use of cranes has become imperative to overcome technical challenges, which has lead to escalation of danger on a construction site. Data from OSHA show that crane accidents have increased rapidly from 2000 to 2004. By analyzing the characteristics of all the crane accident inspections, we can better understand the significance of the many variables involved in a crane accident. For this research, data were collected from the U.S. Department of Labor website via the OSHA database. The data encompass crane accident inspections for all the states. The data were divided into categories with respect to accident types, construction operations, degree of accident, fault, contributing factors, crane types, victims occupation, organs affected and load. Descriptive analysis was performed to compliment the previous studies, the only difference being that both fatal and non-fatal accidents have been considered. Multinomial regression has been applied to derive probability models and correlation between different accident types and the factors involved for each crane accident type. A log likelihood test as well as chi-square test was performed to validate the models. The results show that electrocution, crane tip over and crushed during assembly/disassembly have more probability of occurrence than other accident types. Load is not a significant factor for the crane accidents, and manual fault is more probable a cause for crane accident than is technical fault. Construction operations identified in the research were found to be significant for all the crane accident types. Mobile crawler crane, mobile truck crane and tower crane were found to be more susceptible. These probability models are limited as far as the inculcation of unforeseen variables in construction accidents are concerned. In fact, these models utilize the past to portray the future, and therefore significant change in the variables involved is required to be added to attain correct and expedient results.

Bains, Amrit Anoop Singh

2010-08-01T23:59:59.000Z

446

Comparison Between Sodium Nitrite & Sodium Hydroxide Spray Accident  

SciTech Connect

The purpose of this analysis is to compare the consequences of an 8 molar NaNO2 spray leak to the Tank Farm Final Safety Analysis Report (FSAR) evaluation of sprays of up to 19 molar (50%) NaOH. Four conditions were evaluated. These are: a spray during transfers from a one-inch pipe, a spray resulting from a truck tank Crack, a spray resulting from a truck tank rupture, and a spray in the 204-AR Waste Unloading Facility.

WILLIAMS, J.C.; HEY, B.E.

2001-11-07T23:59:59.000Z

447

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

DOE Green Energy (OSTI)

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.

Goldman, M. (California Univ., Davis, CA (USA)); Nelson, R.C. (EG and G Idaho, Inc., Idaho Falls, ID (USA)); Bollinger, L. (Air Force Inspection and Safety Center, Kirtland AFB, NM (USA)); Hoover, M.D. (Lovelace Biomedical and Environmental Research Inst., Albuquerque, NM (USA). Inhalation Toxicology Research Inst.); Templeton, W. (Pacific Northwest Lab., Richland, WA (USA)); Anspaugh, L. (Lawren

1990-11-02T23:59:59.000Z

448

Views and Patterns in E-Commerce Application Design  

Science Conference Proceedings (OSTI)

Separation of concerns is a well-established principle in software engineering that supports reuse by hiding complexity through abstraction mechanisms. The Abstract Design Views model was created with reuse in mind and allows the designer to apply separation ... Keywords: abstract design views, design, design patterns, object-oriented relationships, reuse, separation of concerns, software engineering, views

Marcus E. Markiewicz; Carlos J. P. Lucena; Paulo S. C. Alencar; Donald D. Cowan

2002-06-01T23:59:59.000Z

449

An Analysis on the Characteristics of Boiling Liquid Expanding Vapor Explosion Accidents in Marine Transportation  

Science Conference Proceedings (OSTI)

BLEVE is a kind of disaster that may cause serious consequences in the process of maritime transportation of liquefied petroleum gas, liquefied natural gas. To analyze the accident characteristics of both the external environment and the internal causes ... Keywords: BLEVE, boiler, characteristics analysis, liquefied gas storage tank

Sining Chen; Yinquan Duo; Lijun Wei

2010-01-01T23:59:59.000Z

450

Topical Report on Reactivity Initiated Accident: Bases for RIA Fuel and Core Coolability Criteria  

Science Conference Proceedings (OSTI)

Revised acceptance criteria have been developed for the response of light water reactor (LWR) fuel under reactivity initiated accidents (RIA). Development of these revisions is part of an industry effort to extend burnup levels beyond currently licensed limits. The revised criteria are proposed for use in licensing burnup extensions or new fuel designs.

2002-05-22T23:59:59.000Z

451

Computer simulation of femur fractures in the case of car accidents  

Science Conference Proceedings (OSTI)

According to a study carried out by one of the authors at the County Clinical Emergency Hospital in Arad, over 40% of the accidents resulting in femur fracture take place on roads and highways. This work briefly presents these results, and on the basis ... Keywords: biomechanics, femur, fractures, model, simulation

Cris Precup; Antoanela Naaji; Csongor Toth; Arpad Toth

2008-06-01T23:59:59.000Z

452

A data assimilation methodology for the plume phase of a nuclear accident  

Science Conference Proceedings (OSTI)

In the aftermath of the Chernobyl nuclear accident and following more than ten years of research and development, the Real time On-line DecisiOn Support system (RODOS) offers a wide range of alternatives to dealing in an effective and efficient fashion ...

R. O. Puch; P. Astrup; J. Q. Smith; H. P. Wynn; C. Turcanu; C. Rojas-Palma

2002-01-01T23:59:59.000Z

453

Incorporation of phenomenological uncertainties in probabilistic safety analysis - application to LMFBR core disruptive accident energetics  

SciTech Connect

This report describes a method for quantifying frequency and consequence uncertainty distribution associated with core disruptive accidents (CDAs). The method was developed to estimate the frequency and magnitude of energy impacting the reactor vessel head of the Clinch River Breeder Plant (CRBRP) given the occurrence of hypothetical CDAs. The methodology is illustrated using the CRBR example.

Najafi, B.; Theofanous, T.G.; Rumble, E.T.; Atefi, B.

1984-08-01T23:59:59.000Z

454

An expert system for strategic control of accidents and insurers' risks in building construction projects  

Science Conference Proceedings (OSTI)

Building construction projects appear to have higher accident rates. Contractors procure workers' compensation insurance (WCI) to transfer these risks to insurance companies. The commitment of insurers under WCI is extremely broad; there are no exclusions ... Keywords: Buildings, Expert system, Fuzzy logic, Occupational health and safety, Singapore, Workers' compensation insurance

Kamardeen Imriyas

2009-03-01T23:59:59.000Z

455

VEACON: A Vehicular Accident Ontology designed to improve safety on the roads  

Science Conference Proceedings (OSTI)

Vehicles are nowadays provided with a variety of new sensors capable of gathering information about themselves and from their surroundings. In a near future, these vehicles will also be capable of sharing all the harvested information, with the surrounding ... Keywords: Intelligent Transportation Systems (ITS), Ontologies, VANETs, Vehicular Networks, Vehicular accidents

Javier Barrachina; Piedad Garrido; Manuel Fogue; Francisco J. Martinez; Juan-Carlos Cano; Carlos T. Calafate; Pietro Manzoni

2012-11-01T23:59:59.000Z

456

View-Dependent Simplification of Arbitrary Polygonal Environments  

E-Print Network (OSTI)

, the dominantcontributorto the totalcostof nuclear electricity is the powerplant construction cost. Thefueland operating's worst nuclear accident occurred in 1986at Reactor Number 4 of the four- unit Chernobyl nuclear power109 7 Nuclear Power and Its Fuel Cycle No technological system more dramatically illustrates

Whitton, Mary C.

457

ECWEBTermsandConditions.doc  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

US DEPARTMENT OF ENERGY US DEPARTMENT OF ENERGY ELECTRONIC COMMERCE PROCEDURES, TERMS AND CONDITIONS Version 1.0 May 1, 1998 PROCEDURES What follows prescribes the general procedures and policies to be followed when Electronic Data Interchange (EDI) is used for transmitting requests for quote, quotations, purchase orders, or other business information in lieu of creating one or more paper documents normally associated with conducting business with the Government. See the Terms and Conditions section below for a list of the standard FAR (Federal Acquisition Regulation) and DEAR (Department of Energy Acquisition Regulation) clauses and provisions mandated to be included in Government contracting actions. The Terms and Conditions together with any clauses specified in the specific contract transaction,

458

The relationship between truck accidents and geometric design of road sections: Poisson versus negative binomial regressions  

SciTech Connect

This paper evaluates the performance of Poisson and negative binomial (NB) regression models in establishing the relationship between truck accidents and geometric design of road sections. Three types of models are considered. Poisson regression, zero-inflated Poisson (ZIP) regression, and NB regression. Maximum likelihood (ML) method is used to estimate the unknown parameters of these models. Two other feasible estimators for estimating the dispersion parameter in the NB regression model are also examined: a moment estimator and a regression-based estimator. These models and estimators are evaluated based on their (1) estimated regression parameters, (2) overall goodness-of-fit, (3) estimated relative frequency of truck accident involvements across road sections, (4) sensitivity to the inclusion of short mad sections, and (5) estimated total number of truck accident involvements. Data from the highway Safety Information System (HSIS) are employed to examine the performance of these models in developing such relationships. The evaluation results suggest that the NB regression model estimated using the moment and regression-based methods should be used with caution. Also, under the ML method, the estimated regression parameters from all three models are quite consistent and no particular model outperforms the other two models in terms of the estimated relative frequencies of truck accident involvements across road sections. It is recommended that the Poisson regression model be used as an initial model for developing the relationship. If the overdispersion of accident data is found to be moderate or high, both the NB and ZIP regression model could be explored. Overall, the ZIP regression model appears to be a serious candidate model when data exhibit excess zeros due, e.g., to underreporting.

Miaou, Shaw-Pin

1993-07-01T23:59:59.000Z

459

Evaluation of the 17 June 1997 Criticality Accident at Arzamas-16  

Science Conference Proceedings (OSTI)

On June 17, 1997, a critically accident occurred at Arzamas-16, which resulted in the death (within three days) of A. N. Zakharov, a Russian scientist with 20 years' experience conducting multiassembly experiments. In this case, the multiplying assembly was a fast metal system consisting of a {sup 235}U (90% enriched) core and a copper reflector. According to the Russian press, ''Zakharov misjudged the degree of criticality of the breeding system and committed several gross violations of regulations.'' As we see it, there were three major causes of this accident. First, the experiment was flawed by Zakharov's misreading of the appropriate size of the assembly, which he took from a notebook that described the old experiment he was attempting to repeat. Second, he disregarded the appropriate procedures and safety regulations. Third, these two mistakes were compounded by an improperly set audible alarm system and Zakharov's unsafe use of the table. We also discuss our reconstruction of the accident based on information given by the Russians to US scientists and information culled from Russian newspaper and magazine articles. We also describe our thoughts on the behavior of the assembly following the accident and the radiation dose level Zakharov may have received. These levels match values we have lately obtained from translations of Russian news articles. This accident clearly points out the penalty for weak administrative control of work with multiplying systems. Criticality experimentation requires formality of operation. The experimenter, his peers, and a trained safety person need to document that they understand the experiment and how it will be conducted. Knowing that the experiment was successfully run several decades ago does not justify bypassing a safety evaluation.

Morris Klein

1999-04-01T23:59:59.000Z

460

Uncertainty and sensitivity analysis of food pathway results with the MACCS Reactor Accident Consequence Model  

Science Conference Proceedings (OSTI)

Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis and stepwise regression analysis are used in an investigation with the MACCS model of the food pathways associated with a severe accident at a nuclear power station. The primary purpose of this study is to provide guidance on the variables to be considered in future review work to reduce the uncertainty in the important variables used in the calculation of reactor accident consequences. The effects of 87 imprecisely-known input variables on the following reactor accident consequences are studied: crop growing season dose, crop long-term dose, milk growing season dose, total food pathways dose, total ingestion pathways dose, total long-term pathways dose, area dependent cost, crop disposal cost, milk disposal cost, condemnation area, crop disposal area and milk disposal area. When the predicted variables are considered collectively, the following input variables were found to be the dominant contributors to uncertainty: fraction of cesium deposition on grain fields that is retained on plant surfaces and transferred directly to grain, maximum allowable ground concentrations of Cs-137 and Sr-90 for production of crops, ground concentrations of Cs-134, Cs-137 and I-131 at which the disposal of milk will be initiated due to accidents that occur during the growing season, ground concentrations of Cs-134, I-131 and Sr-90 at which the disposal of crops will be initiated due to accidents that occur during the growing season, rate of depletion of Cs-137 and Sr-90 from the root zone, transfer of Sr-90 from soil to legumes, transfer of Cs-137 from soil to pasture, transfer of cesium from animal feed to meat, and the transfer of cesium, iodine and strontium from animal feed to milk.

Helton, J.C. [Arizona State Univ., Tempe, AZ (United States); Johnson, J.D.; Rollstin, J.A. [GRAM, Inc., Albuquerque, NM (United States); Shiver, A.W.; Sprung, J.L. [Sandia National Labs., Albuquerque, NM (United States)

1995-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "accident conditions view" from the National Library of EnergyBeta (NLEBeta).
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461

Mineralogy under extreme conditions  

Science Conference Proceedings (OSTI)

We have performed measurements of minerals based on the synchrotron source for single crystal and powder X-ray diffraction, inelastic scattering, spectroscopy and radiography by using diamond anvil cells. We investigated the properties of iron (Fe), iron-magnesium oxides (Fe, Mg)O, silica(SiO{sub 2}), iron-magnesium silicates (Fe, Mg)SiO{sub 3} under simulated high pressure-high temperature extreme conditions of the Earth's crust, upper mantle, low mantle, core-mantle boundary, outer core, and inner core. The results provide a new window on the investigation of the mineral properties at Earth's conditions.

Shu, Jinfu (CIW)

2012-02-07T23:59:59.000Z

462

Field Guide: Visual Inspection of Steel Structures (Optimized for Electronic Viewing)  

Science Conference Proceedings (OSTI)

This EPRI report, one in a series of practical guides designed as reference aids for utility personnel working in the field, visually catalogs the various condition issues that commonly affect transmission line steel structures. The scope includes steel poles, lattice structures, connections, foundations, weathering steel, and coatings.This field guide has been optimized for viewing on electronic devices. For a copy of this product printed on high-quality paper and ring-bound in a ...

2011-12-09T23:59:59.000Z

463

Mountain View IV | Open Energy Information  

Open Energy Info (EERE)

IV IV Facility Mountain View IV Sector Wind energy Facility Type Commercial Scale Wind Facility Status In Service Owner AES Wind Generation Developer AES Wind Generation Energy Purchaser Southern California Edison Co Location White Water CA Coordinates 33.95475187°, -116.7015839° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":33.95475187,"lon":-116.7015839,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

464

Consumer Views on Transportation and Energy  

DOE Green Energy (OSTI)

This report has been assembled to provide the Office of Energy Efficiency and Renewable Energy (EERE) with an idea of how the American public views various transportation, energy, and environmental issues. An issue that still needs attention from EERE is the finding that the public tends to lack information about hybrid vehicles, hydrogen, and alternative fuels for passenger vehicles. Also, the public seems to want fuel-efficiency improvements and cleaner fuels, but is not very willing to pay for these benefits. The public also says that it supports initiatives to promote energy conservation over increased production and that it is willing to make changes such as driving less in an effort to reduce oil consumption.

Steiner, E.

2003-08-01T23:59:59.000Z

465

Accident source terms for light-water nuclear power plants using high-burnup or MOX fuel.  

Science Conference Proceedings (OSTI)

Representative accident source terms patterned after the NUREG-1465 Source Term have been developed for high burnup fuel in BWRs and PWRs and for MOX fuel in a PWR with an ice-condenser containment. These source terms have been derived using nonparametric order statistics to develop distributions for the timing of radionuclide release during four accident phases and for release fractions of nine chemical classes of radionuclides as calculated with the MELCOR 1.8.5 accident analysis computer code. The accident phases are those defined in the NUREG-1465 Source Term - gap release, in-vessel release, ex-vessel release, and late in-vessel release. Important differences among the accident source terms derived here and the NUREG-1465 Source Term are not attributable to either fuel burnup or use of MOX fuel. Rather, differences among the source terms are due predominantly to improved understanding of the physics of core meltdown accidents. Heat losses from the degrading reactor core prolong the process of in-vessel release of radionuclides. Improved understanding of the chemistries of tellurium and cesium under reactor accidents changes the predicted behavior characteristics of these radioactive elements relative to what was assumed in the derivation of the NUREG-1465 Source Term. An additional radionuclide chemical class has been defined to account for release of cesium as cesium molybdate which enhances molybdenum release relative to other metallic fission products.

Salay, Michael (U.S. Nuclear Regulatory Commission, Washington, D.C.); Gauntt, Randall O.; Lee, Richard Y. (U.S. Nuclear Regulatory Commission, Washington, D.C.); Powers, Dana Auburn; Leonard, Mark Thomas

2011-01-01T23:59:59.000Z

466

Metal-fueled HWR (heavy water reactors) severe accident issues: Differences and similarities to commercial LWRs (light water reactors)  

DOE Green Energy (OSTI)

Differences and similarities in severe accident progression and phenomena between commercial Light Water Reactors (LWR) and metal-fueled isotopic production Heavy Water Reactors (HWR) are described. It is very important to distinguish between accident progression in the two systems because each reactor type behaves in a unique manner to a fuel melting accident. Some of the lessons learned as a result of the extensive commercial severe accident research are not applicable to metal-fueled heavy water reactors. A direct application of severe accident phenomena developed from oxide-fueled LWRs to metal-fueled HWRs may lead to large errors or substantial uncertainties. In general, the application of severe accident LWR concepts to HWRs should be done with the intent to define the relevant issues, define differences, and determine areas of overlap. This paper describes the relevant differences between LWR and metal-fueled HWR severe accident phenomena. Also included in the paper is a description of the phenomena that govern the source term in HWRs, the areas where research is needed to resolve major uncertainties, and areas in which LWR technology can be directly applied with few modifications.

Ellison, P.G.; Hyder, M.L.; Monson, P.R. (Westinghouse Savannah River Co., Aiken, SC (USA)); Coryell, E.W. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

1990-01-01T23:59:59.000Z

467

Core structure heat-up and material relocation in a BWR short-term station blackout accident  

Science Conference Proceedings (OSTI)

This paper presents an analytical and numerical analysis which evaluates the core-structure heat-up and subsequent relocation of molten core materials during a NWR short-term station blackout accident with ADS. A simplified one-dimensional approach coupled with bounding arguments is first presented to establish an estimate of the temperature differences within a BWR assembly at the point when structural material first begins to melt. This analysis leads to the conclusions that the control blade will be the first structure to melt and that at this point in time, overall temperature differences across the canister-blade region will not be more than 200 K. Next, a three-dimensional heat-transfer model of the canister-blade region within the core is presented that uses a diffusion approximation for the radiation heat transfer. This is compared to the one-dimensional analysis to establish its compatibility. Finally, the extension of the three-dimensional model to include melt relocation using a porous media type approximation is described. The results of this analysis suggest that under these conditions significant amounts of material will relocate to the core plate region and refreeze, potentially forming a significant blockage. The results also indicate that a large amount of lateral spreading of the melted blade and canister material into the fuel rod regions will occur during the melt progression process. 22 refs., 18 figs., 1 tab.

Schmidt, R.C.; Dosanjh, S.S.

1990-01-01T23:59:59.000Z

468

Light-Weight Radioisotope Heater Unit final safety analysis report (LWRHU-FSAR): Volume 2: Accident Model Document (AMD)  

Science Conference Proceedings (OSTI)

The purpose of this volume of the LWRHU SAR, the Accident Model Document (AMD), are to: Identify all malfunctions, both singular and multiple, which can occur during the complete mission profile that could lead to release outside the clad of the radioisotopic material contained therein; Provide estimates of occurrence probabilities associated with these various accidents; Evaluate the response of the LWRHU (or its components) to the resultant accident environments; and Associate the potential event history with test data or analysis to determine the potential interaction of the released radionuclides with the biosphere.