Sample records for railroad vessel bunkering

  1. The bunkering industry and its effect on shipping tanker operations

    E-Print Network [OSTI]

    Boutsikas, Angelos

    2004-01-01T23:59:59.000Z

    The bunkering industry provides the shipping industry with the fuel oil that the vessels consume. The quality of the fuel oil provided will ensure the safe operation of vessels. Shipping companies under their fuel oil ...

  2. Coal bunkers in underground mines

    SciTech Connect (OSTI)

    Polak, J.; Zegzulka, J. [VSB-Technical Univ., Ostrava (Czech Republic)

    1996-12-31T23:59:59.000Z

    In spite of the technical progress in the application of face technological equipment, the fluctuation of its output has been still considerable. A coal clearance system can be on one hand overloaded by production peaks and on the other hand its stoppages unfavorably influence production of faces. It has been proved that the most effective coal conveying system incorporates surge bunkers to eliminate the above mentioned problems. The surge bunkers have been used in the Czech mines since the middle of the sixties. There were 17 bunkers with an average capacity of 200 m{sup 3} in the biggest Czech coal mine basin OKD in 1967. Presently the number of bunkers has increased to 66 with a total capacity of 40,000 m{sup 3}. It represents the possibility of storing 56% of the daily OKD running of mine output. Two thirds of the number are gate bunkers with an average capacity of 540 m{sup 3} and the rest are skip ones with an average capacity of 740 m{sup 3}, situated at the shaft side.

  3. Residual Fuel Oil Sales for Vessel Bunkering Use

    U.S. Energy Information Administration (EIA) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov YouKizildere IRaghurajiConventionalMississippi"site.1 Relative Standard Errors forA2. For9,250 14,609 9,8515,257,810

  4. Distillate Fuel Oil Sales for Vessel Bunkering Use

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at1,066,688 760,877 951,322 1,381,127 1,710,513

  5. Bunker Fuel Market | OpenEI Community

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are beingZealand Jump to:EzfeedflagBiomassSustainable andBucoda, Washington: Energy(B2G) (SmartBullittBuncombeBunker

  6. Federal Railroad Administration

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

    mandated the study. Federal Railroad Administration DOT Reorganization DOT is in the process of looking at reorganizing the Department to try to achieve even greater...

  7. Enhancing Railroad Hazardous Materials Transportation Safety...

    Office of Environmental Management (EM)

    Enhancing Railroad Hazardous Materials Transportation Safety Rail Routing Enhancing Railroad Hazardous Materials Transportation Safety Rail Routing Presentation made by Kevin...

  8. Intermodal Railroad Daniel Brock

    E-Print Network [OSTI]

    Nagurney, Anna

    of Railroad: Shipments First rail completed in 1830 - Minerals, timber, and agriculture Boxcar 1830s to 1960 made on rail infrastructure are done by owning company. Highways are public and must be funded publicly Southern CSX #12;Who? Revenue by Intermodal BNSF:37% UP: 19% CSX: 14% NS: 20% Rail companies should

  9. Bunker Hill Village, Texas: Energy Resources | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are beingZealand Jump to:EzfeedflagBiomassSustainable andBucoda, Washington: Energy(B2G)Bunker Hill Village, Texas:

  10. Evaluating flexibility in railroad construction projects

    E-Print Network [OSTI]

    Oh, Choong Ryun, 1972-

    2005-01-01T23:59:59.000Z

    This thesis aims to valuate flexibilities in a large-scale railroad construction project. In general, a railroad construction project involves a large amount of flexibilities due to its long construction period and conflicts ...

  11. Railroad and locomotive technology roadmap.

    SciTech Connect (OSTI)

    Stodolsky, F.; Gaines, L.; Energy Systems

    2003-02-24T23:59:59.000Z

    Railroads are important to the U.S. economy. They transport freight efficiently, requiring less energy and emitting fewer pollutants than other modes of surface transportation. While the railroad industry has steadily improved its fuel efficiency--by 16% over the last decade--more can, and needs to, be done. The ability of locomotive manufacturers to conduct research into fuel efficiency and emissions reduction is limited by the small number of locomotives manufactured annually. Each year for the last five years, the two North American locomotive manufacturers--General Electric Transportation Systems and the Electro-Motive Division of General Motors--have together sold about 800 locomotives in the United States. With such a small number of units over which research costs can be spread, outside help is needed to investigate all possible ways to reduce fuel usage and emissions. Because fuel costs represent a significant portion of the total operating costs of a railroad, fuel efficiency has always been an important factor in the design of locomotives and in the operations of a railroad. However, fuel efficiency has recently become even more critical with the introduction of strict emission standards by the U.S. Environmental Protection Agency, to be implemented in stages (Tiers 0, 1, and 2) between 2000 and 2005. Some of the technologies that could be employed to meet the emission standards may negatively affect fuel economy--by as much as 10-15% when emissions are reduced to Tier 1 levels. Lowering fuel economy by that magnitude would have a serious impact on the cost to the consumer of goods shipped by rail, on the competitiveness of the railroad industry, and on this country's dependence on foreign oil. Clearly, a joint government/industry R&D program is needed to help catalyze the development of advanced technologies that will substantially reduce locomotive engine emissions while also improving train system energy efficiency. DOE convened an industry-government workshop in January 2001 to gauge industry interest. As a result, the railroads, their suppliers, and the federal government5 have embarked on a cooperative effort to further improve railroad fuel efficiency--by 25% between now and 2010 and by 50% by 2020, on an equivalent gallon per revenue ton-mile basis, while meeting emission standards, all in a cost-effective, safe manner. This effort aims to bring the collaborative approaches of other joint industry-government efforts, such as FreedomCAR and the 21st Century Truck partnership, to the problem of increasing rail fuel efficiency. Under these other programs, DOE's Office of FreedomCAR and Vehicle Technologies has supported research on technologies to reduce fuel use and air emissions by light- and heavy-duty vehicles. DOE plans to bring similar efforts to bear on improving locomotives. The Department of Transportation's Federal Railroad Administration will also be a major participant in this new effort, primarily by supporting research on railroad safety. Like FreedomCAR and the 21st Century Truck program, a joint industry-government research effort devoted to locomotives and railroad technology could be a 'win' for the public and a 'win' for industry. Industry's expertise and in-kind contributions, coupled with federal funding and the resources of the DOE's national laboratories, could make for an efficient, effective program with measurable energy efficiency targets and realistic deployment schedules. This document provides the necessary background for developing such a program. Potential R&D pathways to greatly improve the efficiency of freight transportation by rail, while meeting future emission standards in a cost-effective, safe manner, were developed jointly by an industry-government team as a result of DOE's January 2001 Workshop on Locomotive Emissions and System Efficiency and are presented here. The status of technology, technical targets, barriers, and technical approaches for engine, locomotive, rail systems, and advanced power plants and fuels are presented.

  12. SO, YOU WANT TO BUILD A RAILROAD?

    E-Print Network [OSTI]

    Barkan, Christopher P.L.

    Public/private benefit issues Competitive issues Political issues #12;Building A New Railroad is Tough Discount Rate · BNSF Cost Inflation (Except Fuel) · BNSF Fuel Price and Escalation Rate Year 2011 2012 2013 railroad ­ roughly 12 miles a year! At that rate, it would take over 200 years to build a transcontinental

  13. Union Pacific Railroad`s LNG locomotive test program

    SciTech Connect (OSTI)

    Grimaila, B.

    1995-12-31T23:59:59.000Z

    Union Pacific Railroad is testing LNG in six locomotives through 1997 to determine if the liquefied natural gas technology is right for them. Two of the six LNG test locomotives are switch, or yard, locomotives. These 1,350 horsepower locomotives are the industry`s first locomotives totally fueled by natural gas. They`re being tested in the yard in the Los Angeles area. The other four locomotives are long-haul locomotives fueled by two tenders. These units are duel-fueled, operating on a mixture of LNG and diesel and are being tested primarily on the Los Angeles to North Platte, Nebraska corridor. All the information concerning locomotive emissions, locomotive performance, maintenance requirements, the overall LNG system design and the economic feasibility of the project will be analyzed to determine if UPR should expand, or abandon, the LNG technology.

  14. Final Report 527 Economic Impact Analysis of Short Line Railroads

    E-Print Network [OSTI]

    Stephens, Jacqueline

    railroads. First described is the scope and presence of the eleven short line railroads currently operatingFinal Report 527 Economic Impact Analysis of Short Line Railroads by Jared J. Llorens, Ph.D. James's Catalog No. 4. Title and Subtitle Economic Impact Analysis of Short Line Railroads 5. Report Date October

  15. William W. Hay Railroad Engineering Seminar Transportation Energy

    E-Print Network [OSTI]

    Barkan, Christopher P.L.

    William W. Hay Railroad Engineering Seminar Transportation Energy Analysis and Modal Comparisons;Transportation Energy Analysis and Modal Comparisons William W Hay Railroad Engineering Seminar September 27 (Energy Efficiency and Air Emissions Intensity) 2. Intercity Passenger Rail Comparison (Air Emissions

  16. Locomotive Fuelling Problem (LFP) in Railroad Operations Bodhibrata Nag1

    E-Print Network [OSTI]

    Murty, Katta G.

    on electric traction to a large extent, US railroads rely mostly on diesel fuel (100% of US freight is hauled is the cost of fuel. This case study deals with minimizing the cost of fuel and the cost of contracting trucks About 75% of the world's railroads operate with diesel fuel. Even though European railroads rely

  17. Optimal Railroad Rail Grinding for Fatigue Mitigation

    E-Print Network [OSTI]

    Tangtragulwong, Potchara

    2012-02-14T23:59:59.000Z

    This dissertation aims to study the benefit of rail grinding on service life of railroad rails, focusing on failures due to rolling contact fatigue (RCF) at the rail head. Assuming a tangent rail with one-point contact at the running surface, a...

  18. Multispectral Machine Vision for Improved Undercarriage Inspection of Railroad Rolling Stock

    E-Print Network [OSTI]

    Ahuja, Narendra

    Multispectral Machine Vision for Improved Undercarriage Inspection of Railroad Rolling Stock Railroad Engineering Program - Department of Civil and Environmental Engineering *Computer Vision monitoring, safety, electrical, component 1. INTRODUCTION Current practices for inspection of railroad

  19. Train derailment and railroad barrier interaction simulation

    E-Print Network [OSTI]

    Zhang, Heping

    1990-01-01T23:59:59.000Z

    ; and (4) impact force between cars. The properties of these forces are discussed below: 11 1. Coupling force and coupling moment a) Coupling force: Figure 3. 4 shows the force-displacement relation which was used in the simulation for coupler action... or eliminate the risk of the passenger track intrusion by derailing cars on adjacent track is of concern to the railroad authorities. The purpose of this study is through computer simulation to investigate the behavior of the train derailment and to determine...

  20. Enhancing Railroad Hazardous Materials Transportation Safety

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011AT&T,OfficeEnd of Year 2010 SNFEnergySession0-02 -Railroad Hazardous g Materials

  1. Optimal Fueling Strategies for Locomotive Fleets in Railroad Networks

    E-Print Network [OSTI]

    Barkan, Christopher P.L.

    · Fuel (diesel) price influenced by: ­ Crude oil price ­ Refining ­ Distribution and marketing ­ Others 4 Price 3 · Railroad fuel consumption remains steady · Crude oil price sharply increases in recent years · Fuel-related expenditure is one of the biggest cost items in the railroad industry #12;Fuel Price

  2. Diesel Locomotive Fueling Problem (LFP) in Railroad Operations

    E-Print Network [OSTI]

    Murty, Katta G.

    Chapter 2 Diesel Locomotive Fueling Problem (LFP) in Railroad Operations Bodhibrata Nag Katta G their operating costs low. About 75% of transport by railroads in the world is based on diesel locomotives by diesel locomotives. One of the major compo- nents in the operating cost of diesel powered rail transport

  3. Predicting violations at gated active highway-railroad grade crossings

    E-Print Network [OSTI]

    Bean, Jonathan Alan

    1997-01-01T23:59:59.000Z

    -railroad grade crossing. It also was recommended that constant warning time devices be used whenever feasible with priority given to crossings with multiple tracks and/or average train speeds below 35 km/hr....

  4. An analysis of railroad accident dynamics in College Station, Texas 

    E-Print Network [OSTI]

    Brown, Hart Stefan

    1998-01-01T23:59:59.000Z

    The intersection between the Union Pacific Railroad (UP) Valley Junction line and Texas A&M University (TAMU) poses a significant potential or risk to the people who utilize the campus. The line carried over I 1,000 loads ...

  5. Assessing the level of service for shipments originating or terminating on short line railroads

    E-Print Network [OSTI]

    Alpert, Steven M

    2007-01-01T23:59:59.000Z

    This thesis measures railroad freight trip time and trip time reliability for freight rail shipments involving short lines in 2006. It is based on an underlying MIT study commissioned by members of the short line railroading ...

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

    SciTech Connect (OSTI)

    NONE

    1991-09-01T23:59:59.000Z

    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).

  7. Energy Conservation Projects to Benefit the Railroad Industry

    SciTech Connect (OSTI)

    Clifford Mirman; Promod Vohra

    2009-12-31T23:59:59.000Z

    The Energy Conservation Projects to benefit the railroad industry using the Norfolk Southern Company as a model for the railroad industry has five unique tasks which are in areas of importance within the rail industry, and specifically in the area of energy conservation. The NIU Engineering and Technology research team looked at five significant areas in which research and development work can provide unique solutions to the railroad industry in energy the conservation. (1) Alternate Fuels - An examination of various blends of bio-based diesel fuels for the railroad industry, using Norfolk Southern as a model for the industry. The team determined that bio-diesel fuel is a suitable alternative to using straight diesel fuel, however, the cost and availability across the country varies to a great extent. (2) Utilization of fuel cells for locomotive power systems - While the application of the fuel cell has been successfully demonstrated in the passenger car, this is a very advanced topic for the railroad industry. There are many safety and power issues that the research team examined. (3) Thermal and emission reduction for current large scale diesel engines - The current locomotive system generates large amount of heat through engine cooling and heat dissipation when the traction motors are used to decelerate the train. The research team evaluated thermal management systems to efficiently deal with large thermal loads developed by the operating engines. (4) Use of Composite and Exotic Replacement Materials - Research team redesigned various components using new materials, coatings, and processes to provide the needed protection. Through design, analysis, and testing, new parts that can withstand the hostile environments were developed. (5) Tribology Applications - Identification of tribology issues in the Railroad industry which play a significant role in the improvement of energy usage. Research team analyzed and developed solutions which resulted in friction modification to improve energy efficiency.

  8. Occupational safety regulation and economic efficiency: the railroad industry

    SciTech Connect (OSTI)

    French, M.T.

    1986-01-01T23:59:59.000Z

    This thesis develops a cost-benefit model for the railroad industry to estimate the efficient level of disabling injuries across class 1 carriers. The neoclassical cost function and the hedonic wage function serve as the estimating equations for the cost and benefit estimation, respectively. The cost function is fitted to a sample of thirty-five class 1 railroads in 1980. The hedonic wage function is fitted to a sample of 481 railroad industry employees for the same year. The efficient level of 1980 disabling injuries for each railroad is determined by setting the marginal cost (MC) of injury reduction equal to its marginal benefit (MB). In addition, the cost savings accruing to efficient regulatory reform are calculated for each firm as well as the entire industry. A second-best cost-minimizing regulatory policy is also introduced. The cost-minimizing level of injuries and resulting cost savings for individual firms and the total industry are estimated. The results support the prior hypothesis that MC exceeds MB for the vast majority of firms. Substantial cost savings are possible under reformed regulatory policy.

  9. Fatigue tests of under-strenghth timber railroad bridge stringers

    E-Print Network [OSTI]

    Borchers, Christopher Ryan

    2002-01-01T23:59:59.000Z

    , experimentally, the fatigue behavior of under-strength timber railroad bridge stringers. The sample group that was tested included nineteen southern pine solid-sawn stringers that were 178 mm (7 in.) wide by 356 mm (14 in.) deep by 4.8 m (16 ft) long...

  10. Paper No. 01-0238 Railroads in Greece

    E-Print Network [OSTI]

    Prevedouros, Panos D.

    Paper No. 01-0238 Railroads in Greece: History, Characteristics and Forecasts John A. Paravantis Piraeus, Greece (jparav@unipi.gr) Panos D. Prevedouros, Ph.D.1 Department of Civil Engineering, University;2 Abstract Greece was relatively tardy in implementing a national rail network. The Greek Railways

  11. PROLONGING THE USEFUL LIFE OF RAILROAD TIMBER BRIDGES

    E-Print Network [OSTI]

    and maintenance, categorized into preventive, early remedial, and major are discussed. The paper concludes on maintenance practices which will prolong the useful life of existing railroad bridges. This paper outlines the various factors which play a role in the deterioration of timber bridges and the maintenance required

  12. Superfund Record of Decision (EPA Region 10): Bunker Hill Mining and Metallurgical Complex, Shoshone County, ID. (First remedial action), August 1991

    SciTech Connect (OSTI)

    Not Available

    1991-08-30T23:59:59.000Z

    The Bunker Hill Mining and Metallurgical Complex site is a 21 square-mile area centered around an inactive industrial mining and smelting site, and includes the cities of Kellogg, Smelterville, Wardner, Pinehurst, and Page, in Shoshone County, Idaho. The inactive industrial complex includes the Bunker Hill mine and mill, a lead smelter, a zinc smelter and a phosphoric acid fertilizer plant, all totalling several hundred acres. Initially, most of the solid and liquid residue from the complex was discharged into the river. When the river flooded, these materials were deposited onto the valley floor, and have leached into onsite soil and ground water. The selected remedial action for the site includes soil sampling; excavating contaminated soil and sod exceeding 1,000 mg/kg lead on approximately 1,800 residential properties, and replacing it with clean soil and sod; disposing of the contaminated soil and sod at an onsite repository; and capping the repository.

  13. Evaluation of passive railroad grade crossing warning devices

    E-Print Network [OSTI]

    Lillo, Sarah Marie

    1992-01-01T23:59:59.000Z

    CONTROL AT PASSIVELY CONTROLLED CROSSINGS . MOTORIST UNDERSTANDING OF RAILROAD WARNING DEVICES EVALUATION TECHNIQUES FOR TRAFFIC CONTROL DEVICES MEASURES OF DRIVER PERFORMANCE EFFECTIVENESS OF ICON SIGNS 12 16 19 III STUDY DESIGN SIGN SYSTEMS... a longer stopping distance, and 3 to 7 percent believed that a heavy truck required as much or more distance to stop (14, 15, 16, 17). The ramifications of this lack of knowledge about train stopping distance are multiplied when one considers...

  14. The potential for LNG as a railroad fuel in the U.S.

    SciTech Connect (OSTI)

    Fritz, S.G.

    2000-01-01T23:59:59.000Z

    Freight railroad operations in the US represent a substantial opportunity for liquefied natural gas (LNG) to displace diesel fuel. With the promise of achieving an overwhelming economic advantage over diesel fuel, this paper presents some discussion to the question, ``Why is the application of LNG for railroad use in the US moving so slowly?'' A brief overview of the freight railroad operations in the US is given, along with a summary of several railroad LNG demonstration projects. US Environmental Protection Agency and California Air Resources Board exhaust emission regulations may cause the railroad industry to move from small-scale LNG demonstration projects to using LNG as a primary freight railroad transportation fuel in selected regions or route-specific applications.

  15. Vacuum Vessel Remote Handling

    E-Print Network [OSTI]

    and Remote Handling 4 Vacuum vessel functions · Plasma vacuum environment · Primary tritium confinement, incl ports 65 tonnes - Weight of torus shielding 100 tonnes · Coolant - Normal Operation Water, Handling 12 Vessel octant subassembly fab. (3) · Octant-to-octant splice joint requires double wall weld

  16. Neutrino Factory Target Vessel

    E-Print Network [OSTI]

    McDonald, Kirk

    by UT-Battelle for the U.S. Department of Energy Target Vessel Update 26 June 2012 Cooling Channel in both walls for draining · Downstream end can be shortened, assuming the window cooling is adequate #12;11 Managed by UT-Battelle for the U.S. Department of Energy Target Vessel Update 26 June 2012 Remote Handling

  17. Joint Sandhouse Gang Meeting & Transportation Center Seminar........ "Recent Railroad Research at Northwestern"

    E-Print Network [OSTI]

    Bustamante, Fabián E.

    of transportation -- with particular emphasis on the trucking and railroad industries. David Kosnik David Kosnik monitoring, monitoring data management and visualization, and acoustic emission testing, to help bring

  18. 16 TAC 3.30 - Memorandum of Understanding between the Railroad...

    Open Energy Info (EERE)

    of Understanding between the Railroad Commission and the Commission on Environmental Quality Jump to: navigation, search OpenEI Reference LibraryAdd to library Memorandum: 16 TAC...

  19. Dual shell pressure balanced vessel

    DOE Patents [OSTI]

    Fassbender, Alexander G. (West Richland, WA)

    1992-01-01T23:59:59.000Z

    A dual-wall pressure balanced vessel for processing high viscosity slurries at high temperatures and pressures having an outer pressure vessel and an inner vessel with an annular space between the vessels pressurized at a pressure slightly less than or equivalent to the pressure within the inner vessel.

  20. HEART AND BLOOD VESSELS CARDIOVASCULARCARDIOVASCULAR

    E-Print Network [OSTI]

    Cochran-Stafira, D. Liane

    HEART AND BLOOD VESSELS CARDIOVASCULARCARDIOVASCULAR SYSTEMSYSTEM SYSTEM COMPONENTS · Heart pumps blood though blood vessels where exchanges can take place with the interstitial fluid (between cells) · Heart and blood vessels regulate blood flow according to the needs of the body

  1. Reactor vessel support system

    DOE Patents [OSTI]

    Golden, Martin P. (Trafford, PA); Holley, John C. (McKeesport, PA)

    1982-01-01T23:59:59.000Z

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  2. James Pon Memorial Award Scholarship Opportunity 2014 Foundation to Commemorate Chinese Railroad Workers in Canada

    E-Print Network [OSTI]

    Barthelat, Francois

    Workers in Canada The Foundation to Commemorate the Chinese Railroad Workers in Canada (FCCRWC and awareness of the contribution to Canada by the Chinese Railroad Workers who participated in the construction of the first transcontinental railway (the CPR) that united Canada geographically and politically

  3. Reactor vessel annealing system

    DOE Patents [OSTI]

    Miller, Phillip E. (Greensburg, PA); Katz, Leonoard R. (Pittsburgh, PA); Nath, Raymond J. (Murrysville, PA); Blaushild, Ronald M. (Export, PA); Tatch, Michael D. (Randolph, NJ); Kordalski, Frank J. (White Oak, PA); Wykstra, Donald T. (Pittsburgh, PA); Kavalkovich, William M. (Monroeville, PA)

    1991-01-01T23:59:59.000Z

    A system for annealing a vessel (14) in situ by heating the vessel (14) to a defined temperature, composed of: an electrically operated heater assembly (10) insertable into the vessel (14) for heating the vessel (14) to the defined temperature; temperature monitoring components positioned relative to the heater assembly (10) for monitoring the temperature of the vessel (14); a controllable electric power supply unit (32-60) for supplying electric power required by the heater assembly (10); a control unit (80-86) for controlling the power supplied by the power supply unit (32-60); a first vehicle (2) containing the power supply unit (32-60); a second vehicle (4) containing the control unit (80-86); power conductors (18,22) connectable between the power supply unit (32-60) and the heater unit (10) for delivering the power supplied by the power supply unit (32-60) to the heater assembly (10); signal conductors (20,24) connectable between the temperature monitoring components and the control unit (80-86) for delivering temperature indicating signals from the temperature monitoring components to the control unit (80-86); and control conductors (8) connectable between the control unit (80-86) and the power supply unit (32-60) for delivering to the power supply unit (32-60) control signals for controlling the level of power supplied by the power supply unit (32-60) to the heater assembly (10).

  4. Neutrino Factory Mercury Vessel

    E-Print Network [OSTI]

    McDonald, Kirk

    Neutrino Factory Mercury Vessel: Initial Cooling Calculations V. Graves Target Studies Nov 15, 2012 #12;2 Managed by UT-Battelle for the U.S. Department of Energy Cooling Calculations 15 Nov 2012 Target · Separates functionality, provides double mercury containment, simplifies design and remote handling · Each

  5. Analysis of improvements in system efficiency and safety at highway-railroad-pedestrian grade crossings

    E-Print Network [OSTI]

    Tydlacka, Jonathan Michael

    2013-02-22T23:59:59.000Z

    The purpose of this project was to perform micro-simulation analyses on intersections near Highway-Railroad Grade Crossings to determine if controlling mean train speed and train speed variability would improve safety and reduce delays. The first...

  6. Crew resource management training's effect on railroad crews' perceptions of task interdependence and teamwork

    E-Print Network [OSTI]

    Kyte, Tobin Bruce

    2008-10-10T23:59:59.000Z

    (Helmreich & Foushee, 1993), healthcare (Cooper, Newbower, & Kitz, 1984), military (Anderson & Sandza, 1987), oil and gas (Cullen, 1990), and railroad (FRA, 1999) industries have also been caused by breakdowns in coordination (assertiveness, 3...

  7. Preemption strategy for traffic signals at intersections near highway-railroad grade crossings

    E-Print Network [OSTI]

    Cho, Hanseon

    2004-09-30T23:59:59.000Z

    Because the operational characteristics of signalized intersections near highway-railroad grade crossings (IHRGCs) are different from those of signalized intersections located elsewhere in the traffic system, standard operational strategies do...

  8. An evaluation of modifications to the advance warning sign at passive highway-railroad grade crossings

    E-Print Network [OSTI]

    Keller, Bridgette DeWees

    1995-01-01T23:59:59.000Z

    , driver behavior at highway-railroad grade crossings, and effectiveness of traffic control devices that have supplemented the standard signing system with strobe lights. The objective of this research was to determine if the addition of a strobe light...

  9. Fire modeling for Building 221-T - T Plant Canyon Deck and Railroad Tunnel

    SciTech Connect (OSTI)

    Oar, D.L.

    1994-09-29T23:59:59.000Z

    This report was prepared by Hughes Associates, Inc. to document the results of fire models for building 221-T Canyon Deck and Railroad Tunnel. Backup data is contained in document No. WHC-SD-CP-ANAL-010, Rev. 0.

  10. Communication and Interpretation of Results of Railroad Hazardous Materials Transportation Route Risk Analyses

    E-Print Network [OSTI]

    Barkan, Christopher P.L.

    practices to reduce the likelihood or severity of accidents (1), improved training of personnel (2 Risk Analyses Athaphon Kawprasert (Corresponding Author) Graduate Research Assistant Railroad Engineering Program Department of Civil and Environmental Engineering University of Illinois at Urbana

  11. A field evaluation of experimental passive railroad-highway grade crossing warning systems

    E-Print Network [OSTI]

    Beitler, Michael Morgan

    1993-01-01T23:59:59.000Z

    A FIELD EVALUATION OF EKPERIMENTAL PASSIVE RAILROAD-HIGHWAY GRADE CROSSING WARNING SYSTEMS A Thesis MICHAEL MORGAN BEITLER Submitted to the Office of Graduate Studies of Texas A&M University in partial fulfillment of the requirements... for the degree of MASTER OF SCIENCE August 1993 Major Subject: Civil Engineering )~~I ABSTRACT A Field Evaluation of Experimental Passive Railroad- Highway Grade Crossing Warning Systems. (August 1993) Michael Morgan Beitler, B. S. , University...

  12. Railroad competition and rate making: the implications for industrial coal prices

    SciTech Connect (OSTI)

    none,

    1981-09-01T23:59:59.000Z

    As the railroad industry, which is the primary transportation mode for coal, enters the 1980s, two factors are significantly altering the way in which railroads do business, particularly in relationship to shippers who must rely solely on rail. The first factor, railroad deregulation, is embodied in the provisions of the Staggers Rail Act of 1980 (PL96-448). The second factor affecting rail transportation for industrial coal users is the trend toward rail mergers and consolidations, in some cases due to financial problems. This trend, coupled with the willingness of government to allow major bankrupt carriers such as the Rock Island and the Milwaukee Road to be liquidated in whole or in part, is resulting in fewer railraods with increased market power. Discussed in this report are: the impacts of railroad deregulation on industrial coal users; the effects of railroad consolidations on the bargaining power of coal producers, receivers and railroads; and finally, strategy and tactics for industrial coal users to maximize their leverage weight in obtaining efficient coal transportation at reasonable rates. (DMC)

  13. Railroad electrification in America's future: an assessment of prospects and impacts. Final report

    SciTech Connect (OSTI)

    White, R.K.; Yabroff, I.W.; Dickson, E.M.; Zink, R.A.; Gray, M.E.; Moon, A.E.

    1980-01-01T23:59:59.000Z

    Such considerations as the level of traffic, the relative financial health of individual railroads, the capacity of the associated supply and engineering/construction industries, and the logical connecting points at classifying yards, as well as the national interest value of creating a continuous system, continental in scope, were used to construct a scenario for railroad electrification that closely approximates how an electrification program might be implemented. For the economic reasons cited, much of the US railroad system would remain conventionally powered. This scenario provides for an electrified network involving 14 mainlines operated by 10 companies that could transport much of the nation's rail-borne freight. Five years of planning and engineering work would be required for each link before construction could begin. With 1000 miles or less of electrified route per year, 14 years would be needed to construct the 9000-mile network of our scenario. (The scenario constructed runs from 1980 to 1998.) The analysis was aided with the construction of the SRI Railroad Industry Model. Basically a model of industry operations and finances, the model produces income statements and balance sheets at yearly intervals. Railroad energy costs, railroad freight levels, maintenance costs, purchases and leases of rooling stock, electrification facility investments, future inflation, rate setting practices, annual depreciation, taxes, and profits were calculated.

  14. High pressure storage vessel

    DOE Patents [OSTI]

    Liu, Qiang

    2013-08-27T23:59:59.000Z

    Disclosed herein is a composite pressure vessel with a liner having a polar boss and a blind boss a shell is formed around the liner via one or more filament wrappings continuously disposed around at least a substantial portion of the liner assembly combined the liner and filament wrapping have a support profile. To reduce susceptible to rupture a locally disposed filament fiber is added.

  15. Reactor pressure vessel nozzle

    DOE Patents [OSTI]

    Challberg, R.C.; Upton, H.A.

    1994-10-04T23:59:59.000Z

    A nozzle for joining a pool of water to a nuclear reactor pressure vessel includes a tubular body having a proximal end joinable to the pressure vessel and a distal end joinable in flow communication with the pool. The body includes a flow passage therethrough having in serial flow communication a first port at the distal end, a throat spaced axially from the first port, a conical channel extending axially from the throat, and a second port at the proximal end which is joinable in flow communication with the pressure vessel. The inner diameter of the flow passage decreases from the first port to the throat and then increases along the conical channel to the second port. In this way, the conical channel acts as a diverging channel or diffuser in the forward flow direction from the first port to the second port for recovering pressure due to the flow restriction provided by the throat. In the backflow direction from the second port to the first port, the conical channel is a converging channel and with the abrupt increase in flow area from the throat to the first port collectively increase resistance to flow therethrough. 2 figs.

  16. Amendment 80 vessel replacement 1 Implementation and of Amendment 80 Vessel Replacement Provisions

    E-Print Network [OSTI]

    Amendment 80 vessel replacement 1 Implementation and of Amendment 80 Vessel Replacement Provisions vessels to use non-qualifying vessels in the sector, thus allowing replacement of a lost qualifying vessel of the CRP ambiguous as to whether replacement of qualifying vessels with non-qualifying vessels

  17. Vessel structural support system

    DOE Patents [OSTI]

    Jenko, James X. (N. Versailles, PA); Ott, Howard L. (Kiski Twp., Allegheny County, PA); Wilson, Robert M. (Plum Boro, PA); Wepfer, Robert M. (Murrysville, PA)

    1992-01-01T23:59:59.000Z

    Vessel structural support system for laterally and vertically supporting a vessel, such as a nuclear steam generator having an exterior bottom surface and a side surface thereon. The system includes a bracket connected to the bottom surface. A support column is pivotally connected to the bracket for vertically supporting the steam generator. The system also includes a base pad assembly connected pivotally to the support column for supporting the support column and the steam generator. The base pad assembly, which is capable of being brought to a level position by turning leveling nuts, is anchored to a floor. The system further includes a male key member attached to the side surface of the steam generator and a female stop member attached to an adjacent wall. The male key member and the female stop member coact to laterally support the steam generator. Moreover, the system includes a snubber assembly connected to the side surface of the steam generator and also attached to the adjacent wall for dampening lateral movement of the steam generator. In addition, the system includes a restraining member of "flat" attached to the side surface of the steam generator and a bumper attached to the adjacent wall. The flat and the bumper coact to further laterally support the steam generator.

  18. Ferrocarriles nacionales de Mexico: the existing rate structure for grain/soybean shipments and likely effect on freight rates as a result of railroad privatization in Mexico 

    E-Print Network [OSTI]

    Neyer, David McAlister

    1994-01-01T23:59:59.000Z

    The government owned railway monopoly in Mexico, Ferrocarriles Nacionales de Mexico (FNM), is undergoing a series of reforms designed to create a market orientated railroad company. Railroad freight rates are being deregulated ...

  19. Coal gasification vessel

    DOE Patents [OSTI]

    Loo, Billy W. (Oakland, CA)

    1982-01-01T23:59:59.000Z

    A vessel system (10) comprises an outer shell (14) of carbon fibers held in a binder, a coolant circulation mechanism (16) and control mechanism (42) and an inner shell (46) comprised of a refractory material and is of light weight and capable of withstanding the extreme temperature and pressure environment of, for example, a coal gasification process. The control mechanism (42) can be computer controlled and can be used to monitor and modulate the coolant which is provided through the circulation mechanism (16) for cooling and protecting the carbon fiber and outer shell (14). The control mechanism (42) is also used to locate any isolated hot spots which may occur through the local disintegration of the inner refractory shell (46).

  20. LNG as a fuel for railroads: Assessment of technology status and economics. Topical report, June-September 1992

    SciTech Connect (OSTI)

    Pera, C.J.; Moyer, C.B.

    1993-01-06T23:59:59.000Z

    The objective of the research was to investigate the feasibility of liquefied natural gas (LNG) as a fuel for railroads. The investigation included assessment of the status of relevant technologies (i.e., LNG-fueled locomotive engines, tender cars, refueling equipment), a review of current demonstration projects, and an analytical evaluation of LNG railroad economics.

  1. Reactor vessel support system. [LMFBR

    DOE Patents [OSTI]

    Golden, M.P.; Holley, J.C.

    1980-05-09T23:59:59.000Z

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  2. CHF Enhancement by Vessel Coating for External Reactor Vessel Cooling

    SciTech Connect (OSTI)

    Fan-Bill Cheung; Joy L. Rempe

    2004-06-01T23:59:59.000Z

    In-vessel retention (IVR) is a key severe accident management (SAM) strategy that has been adopted by some operating nuclear power plants and advanced light water reactors (ALWRs). One viable means for IVR is the method of external reactor vessel cooling (ERVC) by flooding of the reactor cavity during a severe accident. As part of a joint Korean – United States International Nuclear Energy Research Initiative (K-INERI), an experimental study has been conducted to investigate the viability of using an appropriate vessel coating to enhance the critical heat flux (CHF) limits during ERVC. Toward this end, transient quenching and steady-state boiling experiments were performed in the SBLB (Subscale Boundary Layer Boiling) facility at Penn State using test vessels with micro-porous aluminum coatings. Local boiling curves and CHF limits were obtained in these experiments. When compared to the corresponding data without coatings, substantial enhancement in the local CHF limits for the case with surface coatings was observed. Results of the steady state boiling experiments showed that micro-porous aluminum coatings were very durable. Even after many cycles of steady state boiling, the vessel coatings remained rather intact, with no apparent changes in color or structure. Moreover, the heat transfer performance of the coatings was found to be highly desirable with an appreciable CHF enhancement in all locations on the vessel outer surface but with very little effect of aging.

  3. RAILROAD STRATEGY FOR CRUDE OIL TRANSPORT: Considering Public Policy and Pipeline Competition

    E-Print Network [OSTI]

    Entekhabi, Dara

    of crude oil transportation by rail from the Alberta oil sands has been slower than the growth, stakeholders, and objectives for the bitumen transport system from the Alberta oil sands, and reviews: Existing railroad network: Some products shipped: Fig. 1: Predicted oil sands production growth

  4. EFFECT OF COAL DUST ONEFFECT OF COAL DUST ON RAILROAD BALLAST STRENGTHRAILROAD BALLAST STRENGTH

    E-Print Network [OSTI]

    Barkan, Christopher P.L.

    1 EFFECT OF COAL DUST ONEFFECT OF COAL DUST ON RAILROAD BALLAST STRENGTHRAILROAD BALLAST STRENGTH for Laboratory StudyFouling Mechanism / Need for Laboratory Study Mechanical Properties of Coal DustMechanical Properties of Coal Dust Grain Size AnalysisGrain Size Analysis AtterbergAtterberg LimitsLimits Specific

  5. A technique for classifying and evaluating railroad-highway grade crossings

    E-Print Network [OSTI]

    Newton, Thomas Monroe

    1974-01-01T23:59:59.000Z

    . Track View of Research Annex Boring Site 50 33. Locations for Test Holes at Research Annex Boring Site 52 34. Goodyear Rubber Crossing Installation at FM Highway 1960 and Rock Island Railroad 57 35. Cross Section Through Rubber Panel Crossing 57 36...

  6. Toward a State of Good Repair for Railroad Bridges through Structural Health Monitoring

    E-Print Network [OSTI]

    Bustamante, Fabián E.

    Toward a State of Good Repair for Railroad Bridges through Structural Health Monitoring NUTC of this research is to add a new tool to the asset management toolbox: structural health monitoring, particularly is Structural Health Monitoring (SHM)? SHM is process of intelligently collecting data from sensors on critical

  7. Sawadisavi et al. 09-1369 1 Machine-Vision Inspection of Railroad Track

    E-Print Network [OSTI]

    Barkan, Christopher P.L.

    Assistant Computer Vision and Robotics Laboratory University of Illinois at Urbana-Champaign 1614 Beckman. Sawadisavi Graduate Research Assistant Railroad Engineering Program Department of Civil and Environmental and Robotics Laboratory University of Illinois at Urbana-Champaign 3021 Beckman Institute for Advanced Science

  8. A Simulation Tool for Combined Rail-Road Transport in Intermodal Terminals

    E-Print Network [OSTI]

    Gambardella, Luca Maria

    A Simulation Tool for Combined Rail-Road Transport in Intermodal Terminals Andrea E. Rizzoli. A simulation tool to model the flow of Intermodal Terminal Units (ITUs) among intermodal terminals is presented. This terminal simulator tool is part of the DGVII EC-funded PLATFORM project. The terminal model is composed

  9. A Nanofluidic Railroad Switch for DNA Robert Riehn,*,,, Robert H. Austin, and James C. Sturm

    E-Print Network [OSTI]

    A Nanofluidic Railroad Switch for DNA Robert Riehn,*,,,§ Robert H. Austin, and James C. Sturm of electric fields in such structures, and the impact of ac fields on DNA in a complex nanofluidic environment by a 200 µm long and 120 µm wide nanofluidic region. (b) Enlarged schematic of the nanofluidic region

  10. Level indicator for pressure vessels

    DOE Patents [OSTI]

    Not Available

    1982-04-28T23:59:59.000Z

    A liquid-level monitor for tracking the level of a coal slurry in a high-pressure vessel including a toroidal-shaped float with magnetically permeable bands thereon disposed within the vessel, two pairs of magnetic-field generators and detectors disposed outside the vessel adjacent the top and bottom thereof and magnetically coupled to the magnetically permeable bands on the float, and signal-processing circuitry for combining signals from the top and bottom detectors for generating a monotonically increasing analog control signal which is a function of liquid level. The control signal may be utilized to operate high-pressure control valves associated with processes in which the high-pressure vessel is used.

  11. Tailoring Topology Optimization to Composite Pressure Vessel Design with Simultaneous

    E-Print Network [OSTI]

    Paulino, Glaucio H.

    ;Introduction ­ CNG Pressure Vessels Compressed Natural Gas (CNG) Pressure Vessels CNG Cargo Containment System

  12. Current Trends and Future Challenges in the Freight Railroad Industry: Balancing Private Industry Interests and the Public Welfare 

    E-Print Network [OSTI]

    Allen, Sarah; Kelson, Kendra; Migl, Hayden; Schmidt, Rodney; Shoemaker, David; Thomson, Heather

    2008-01-01T23:59:59.000Z

    Commission (ICC). The Staggers Act is based on the premise that while the railroad industry once constituted a monopoly, requiring ICC regulation , this is no longer true (Keeler). It is held that competition is generally adequate to constrain any... shippers and to limit rates from falling below variable costs. Maximum rates are identified through specified revenue to variable cost ratios but apply only if the railroad is earning an "adequate" return on investment (Keeler). In addition...

  13. Report on audit of the use of Hanford site railroad system

    SciTech Connect (OSTI)

    NONE

    1997-03-20T23:59:59.000Z

    The DOE in partnership with its contractors, strives to operate, maintain, and use existing physical assets to meet mission needs. Field offices have first line responsibility to ensure stewardship of the assets. Part of stewardship responsibility involves ensuring assets are fully evaluated for use in meeting the DOE`s mission. One of Richland`s oldest, yet well-maintained multi-million dollar physical assets, is the Hanford railroad system. This system, constructed in the 1940s, includes tracks, railcars, and locomotives and provides sitewide transportation. Because of the availability of this asset, the objective of the audit was to determine if Richland ensured that Hanford contractors were fully utilizing the railroad system.

  14. A microsimulation analysis of highway intersections near highway-railroad grade crossings

    E-Print Network [OSTI]

    Tydlacka, Jonathan Michael

    2004-11-15T23:59:59.000Z

    A MICROSIMULATION ANALYSIS OF HIGHWAY INTERSECTIONS NEAR HIGHWAY-RAILROAD GRADE CROSSINGS A Thesis by JONATHAN MICHAEL TYDLACKA Submitted to the Office of Graduate Studies of Texas A&M University in partial... A Thesis by JONATHAN MICHAEL TYDLACKA Submitted to Texas A&M University in partial fulfillment of the requirements for the degree of MASTER OF SCIENCE Approved as to style and content by...

  15. Preparedness of hazardous materials emergencies in railyards: Guidance for railroads and adjacent communities

    SciTech Connect (OSTI)

    Not Available

    1991-09-01T23:59:59.000Z

    Railroads are a key part of the distribution system for hazardous materials and, thus, much hazardous material passes through railyards en route to intermediate or final consumers. While the vast majority of these materials are shipped without incident, both the number of shipments and the nature of the materials themselves dictate that railyards and surrounding communities be prepared to respond quickly and effectively to emergencies. This report contains information on 11 emergency preparedness functions and 150 guidance recommendations.

  16. Light Sources on the Nylon Vessels' Surfaces

    E-Print Network [OSTI]

    Chapter 7 Light Sources on the Nylon Vessels' Surfaces The nylon vessels are justifiably the most the IV. A set of light diffusers has been placed on pre-defined points of both vessels. These are attached to the tip of an optical fiber that carries light from a source outside the WT (LED 184 #12

  17. The Planning, Licensing, Modifications, and Use of a Russian Vessel for Shipping Spent Nuclear Fuel by Sea in Support of the DOE RRRFR Program

    SciTech Connect (OSTI)

    Michael Tyacke; Dr. Igor Bolshinsky; Wlodzimierz Tomczak; Sergey Naletov; Oleg Pichugin

    2001-10-01T23:59:59.000Z

    The Russian Research Reactor Fuel Return (RRRFR) Program, under the U.S. Department of Energy’s Global Threat Reduction Initiative, began returning Russian-supplied high-enriched uranium (HEU) spent nuclear fuel (SNF), stored at Russian-designed research reactors throughout the world, to Russia in January 2006. During the first years of making HEU SNF shipments, it became clear that the modes of transportation needed to be expanded from highway and railroad to include sea and air to meet the extremely aggressive commitment of completing the first series of shipments by the end of 2010. The first shipment using sea transport was made in October 2008 and used a non-Russian flagged vessel. The Russian government reluctantly allowed a one-time use of the foreign-owned vessel into their highly secured seaport, with the understanding that any future shipments would be made using a vessel owned and operated by a Russian company. ASPOL-Baltic of St. Petersburg, Russia, owns and operates a small fleet of vessels and has a history of shipping nuclear materials. ASPOL-Baltic’s vessels were licensed for shipping nuclear materials; however, they were not licensed to transport SNF materials. After a thorough review of ASPOL Baltic’s capabilities and detailed negotiations, it was agreed that a contract would be let with ASPOL-Baltic to license and refit their MCL Trader vessel for hauling SNF in support of the RRRFR Program. This effort was funded through a contract between the RRRFR Program, Idaho National Laboratory, and Radioactive Waste Management Plant of Swierk, Poland. This paper discusses planning, Russian and international maritime regulations and requirements, Russian authorities’ reviews and approvals, licensing, design, and modifications made to the vessel in preparation for SNF shipments. A brief summary of actual shipments using this vessel, experiences, and lessons learned also are described.

  18. Market effects of environmental regulation: coal, railroads, and the 1990 Clean Air Act

    SciTech Connect (OSTI)

    Busse, M.R.; Keohane, N.O. [University of California Berkeley, Berkeley, CA (United States)

    2007-01-01T23:59:59.000Z

    Many environmental regulations encourage the use of 'clean' inputs. When the suppliers of such an input have market power, environmental regulation will affect not only the quantity of the input used but also its price. We investigate the effect of the Title IV emissions trading program for sulfur dioxide on the market for low-sulfur coal. We find that the two railroads transporting coal were able to price discriminate on the basis of environmental regulation and geographic location. Delivered prices rose for plants in the trading program relative to other plants, and by more at plants near a low-sulfur coal source.

  19. Reactor pressure vessel vented head

    DOE Patents [OSTI]

    Sawabe, James K. (San Jose, CA)

    1994-01-11T23:59:59.000Z

    A head for closing a nuclear reactor pressure vessel shell includes an arcuate dome having an integral head flange which includes a mating surface for sealingly mating with the shell upon assembly therewith. The head flange includes an internal passage extending therethrough with a first port being disposed on the head mating surface. A vent line includes a proximal end disposed in flow communication with the head internal passage, and a distal end disposed in flow communication with the inside of the dome for channeling a fluid therethrough. The vent line is fixedly joined to the dome and is carried therewith when the head is assembled to and disassembled from the shell.

  20. Tow Vessel | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov YouKizildere IRaghuraji Agro Industries PvtStratosolarTharaldsonInformationTorpedo SpecialityVessel Jump to:

  1. INTERLINE 5. 0 -- An expanded railroad routing model: Program description, methodology, and revised user's manual

    SciTech Connect (OSTI)

    Johnson, P.E.; Joy, D.S. (Oak Ridge National Lab., TN (United States)); Clarke, D.B.; Jacobi, J.M. (Tennessee Univ., Knoxville, TN (United States). Transportation Center)

    1993-03-01T23:59:59.000Z

    A rail routine model, INTERLINE, has been developed at the Oak Ridge National Laboratory to investigate potential routes for transporting radioactive materials. In Version 5.0, the INTERLINE routing algorithms have been enhanced to include the ability to predict alternative routes, barge routes, and population statistics for any route. The INTERLINE railroad network is essentially a computerized rail atlas describing the US railroad system. All rail lines, with the exception of industrial spurs, are included in the network. Inland waterways and deep water routes along with their interchange points with the US railroadsystem are also included. The network contains over 15,000 rail and barge segments (links) and over 13,000 stations, interchange points, ports, and other locations (nodes). The INTERLINE model has been converted to operate on an IBM-compatible personal computer. At least a 286 computer with a hard disk containing approximately 6 MB of free space is recommended. Enhanced program performance will be obtained by using arandom-access memory drive on a 386 or 486 computer.

  2. Reactor pressure vessel vented head

    DOE Patents [OSTI]

    Sawabe, J.K.

    1994-01-11T23:59:59.000Z

    A head for closing a nuclear reactor pressure vessel shell includes an arcuate dome having an integral head flange which includes a mating surface for sealingly mating with the shell upon assembly therewith. The head flange includes an internal passage extending therethrough with a first port being disposed on the head mating surface. A vent line includes a proximal end disposed in flow communication with the head internal passage, and a distal end disposed in flow communication with the inside of the dome for channeling a fluid therethrough. The vent line is fixedly joined to the dome and is carried therewith when the head is assembled to and disassembled from the shell. 6 figures.

  3. Evaluation of in-vessel corium retention through external reactor vessel cooling for integral reactor

    SciTech Connect (OSTI)

    Park, R. J.; Lee, J. R.; Kim, S. B.; Jin, Y.; Kim, H. Y. [Korea Atomic Energy Research Inst., 1045 Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

    2012-07-01T23:59:59.000Z

    In-vessel corium retention through external reactor vessel cooling (IVR-ERVC) for a small integral reactor has been evaluated to determine the thermal margin for the prevention of a reactor vessel failure. A thermal load analysis from the corium pool to the outer reactor vessel wall in the lower plenum of the reactor vessel has been performed to determine the heat flux distribution. The critical heat flux (CHF) on the outer reactor vessel wall has been determined to fix the maximum heat removal rate through the external coolant between the outer reactor vessel and the insulation of the reactor vessel. Finally, the thermal margin has been evaluated by comparison of the thermal load with the maximum heat removal rate of the CHF on the outer reactor vessel wall. The maximum heat flux from the corium pool to the outer reactor vessel is estimated at approximately 0.25 MW/m{sup 2} in the metallic layer because of the focusing effect. The CHF of the outer reactor vessel is approximately 1.1 MW/m{sup 2} because of a two phase natural circulation mass flow. Since the thermal margin for the IVR-ERVC is sufficient, the reactor vessel integrity is maintained during a severe accident of a small integral reactor. (authors)

  4. Fabrication of Separator Demonstration Facility process vessel

    SciTech Connect (OSTI)

    Oberst, E.F.

    1985-01-15T23:59:59.000Z

    The process vessel system is the central element in the Separator Development Facility (SDF). It houses the two major process components, i.e., the laser-beam folding optics and the separators pods. This major subsystem is the critical-path procurement for the SDF project. Details of the vaious parts of the process vessel are given.

  5. Foam vessel for cryogenic fluid storage

    DOE Patents [OSTI]

    Spear, Jonathan D (San Francisco, CA)

    2011-07-05T23:59:59.000Z

    Cryogenic storage and separator vessels made of polyolefin foams are disclosed, as are methods of storing and separating cryogenic fluids and fluid mixtures using these vessels. In one embodiment, the polyolefin foams may be cross-linked, closed-cell polyethylene foams with a density of from about 2 pounds per cubic foot to a density of about 4 pounds per cubic foot.

  6. Application for Amendment 80 Vessel Replacement Page 1 of 6

    E-Print Network [OSTI]

    Application for Amendment 80 Vessel Replacement Page 1 of 6 Revised: 12/23/2013 OMB Control No. 0648-0565 Expiration Date: 01/31/2016 APPLICATION FOR AMENDMENT 80 VESSEL REPLACEMENT United States OF THE AMENDMENT 80 VESSEL BEING REPLACED 1. Vessel Name: 2. ADF&G Vessel Registration No.: 3. USCG Documentation

  7. Nuclear reactor vessel fuel thermal insulating barrier

    DOE Patents [OSTI]

    Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

    2013-03-19T23:59:59.000Z

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

  8. Reactor pressure vessel. Status report

    SciTech Connect (OSTI)

    Elliot, B.J.; Hackett, E.M.; Lee, A.D. [and others

    1996-10-01T23:59:59.000Z

    This report describes the issues raised as a result of the staffs review of Generic Letter (GL) 92-01, Revision 1, responses and plant-specific reactor pressure vessel (RPV) assessments and the actions taken or work in progress to address these issues. In addition, the report describes actions taken by the staff and the nuclear industry to develop a thermal annealing process for use at U.S. commercial nuclear power plants. This process is intended to be used as a means of mitigating the effects of neutron radiation on the fracture toughness of RPV materials. The Nuclear Regulatory Commission (NRC) issued GL 92-01, Revision 1, Supplement 1, to obtain information needed to assess compliance with regulatory requirements and licensee commitments regarding RPV integrity. GL 92-01, Revision 1, Supplement 1, was issued as a result of generic issues that were raised in the NRC staff`s reviews of licensee responses to GL 92-01, Revision 1, and plant-specific RPV evaluations. In particular, an integrated review of all data submitted in response to GL 92-01, Revision 1, indicated that licensees may not have considered all relevant data in their RPV assessments. This report is representative of submittals to and evaluations by the staff as of September 30, 1996. An update of this report will be issued at a later date.

  9. Lightweight bladder lined pressure vessels

    DOE Patents [OSTI]

    Mitlitsky, F.; Myers, B.; Magnotta, F.

    1998-08-25T23:59:59.000Z

    A lightweight, low permeability liner is described for graphite epoxy composite compressed gas storage vessels. The liner is composed of polymers that may or may not be coated with a thin layer of a low permeability material, such as silver, gold, or aluminum, deposited on a thin polymeric layer or substrate which is formed into a closed bladder using tori spherical or near tori spherical end caps, with or without bosses therein, about which a high strength to weight material, such as graphite epoxy composite shell, is formed to withstand the storage pressure forces. The polymeric substrate may be laminated on one or both sides with additional layers of polymeric film. The liner may be formed to a desired configuration using a dissolvable mandrel or by inflation techniques and the edges of the film sealed by heat sealing. The liner may be utilized in most any type of gas storage system, and is particularly applicable for hydrogen, gas mixtures, and oxygen used for vehicles, fuel cells or regenerative fuel cell applications, high altitude solar powered aircraft, hybrid energy storage/propulsion systems, and lunar/Mars space applications, and other applications requiring high cycle life. 19 figs.

  10. Lightweight bladder lined pressure vessels

    DOE Patents [OSTI]

    Mitlitsky, Fred (1125 Canton Ave., Livermore, CA 94550); Myers, Blake (4650 Almond Cir., Livermore, CA 94550); Magnotta, Frank (1206 Bacon Way, Lafayette, CA 94549)

    1998-01-01T23:59:59.000Z

    A lightweight, low permeability liner for graphite epoxy composite compressed gas storage vessels. The liner is composed of polymers that may or may not be coated with a thin layer of a low permeability material, such as silver, gold, or aluminum, deposited on a thin polymeric layer or substrate which is formed into a closed bladder using torispherical or near torispherical end caps, with or without bosses therein, about which a high strength to weight material, such as graphite epoxy composite shell, is formed to withstand the storage pressure forces. The polymeric substrate may be laminated on one or both sides with additional layers of polymeric film. The liner may be formed to a desired configuration using a dissolvable mandrel or by inflation techniques and the edges of the film seamed by heat sealing. The liner may be utilized in most any type of gas storage system, and is particularly applicable for hydrogen, gas mixtures, and oxygen used for vehicles, fuel cells or regenerative fuel cell applications, high altitude solar powered aircraft, hybrid energy storage/propulsion systems, and lunar/Mars space applications, and other applications requiring high cycle life.

  11. Glossary Glossary

    U.S. Energy Information Administration (EIA) Indexed Site

    F-77), and Bunker C fuel oil. Residual fuel oil is used for the production of electric power, space heating, vessel bunkering, and various industrial purposes. Retailer: A firm...

  12. Glossary

    U.S. Energy Information Administration (EIA) Indexed Site

    F-77), and Bunker C fuel oil. Residual fuel oil is used for the production of electric power, space heating, vessel bunkering, and various industrial purposes. Retailer: A firm...

  13. Thermal wake/vessel detection technique

    DOE Patents [OSTI]

    Roskovensky, John K. (Albuquerque, NM); Nandy, Prabal (Albuquerque, NM); Post, Brian N (Albuquerque, NM)

    2012-01-10T23:59:59.000Z

    A computer-automated method for detecting a vessel in water based on an image of a portion of Earth includes generating a thermal anomaly mask. The thermal anomaly mask flags each pixel of the image initially deemed to be a wake pixel based on a comparison of a thermal value of each pixel against other thermal values of other pixels localized about each pixel. Contiguous pixels flagged by the thermal anomaly mask are grouped into pixel clusters. A shape of each of the pixel clusters is analyzed to determine whether each of the pixel clusters represents a possible vessel detection event. The possible vessel detection events are represented visually within the image.

  14. Multi-modal Transportation > Highway Transportation > Trucking > Railroad transportation > Public transit > Rural transportation > Rural transit > Freight pipeline transportation > Airport planning and development > Airport maintenance > Bicycle and pedes

    E-Print Network [OSTI]

    Multi-modal Transportation > Highway Transportation > Trucking > Railroad transportation > Public transit > Rural transportation > Rural transit > Freight pipeline transportation > Airport planning and development > Airport maintenance > Bicycle and pedestrian > Ports and waterways >>> Transportation operat

  15. Multi-modal Transportation > Highway Transportation > Trucking > Railroad transportation > Public transit > Rural transportation > Rural transit > Freig pipeline transportation > Airport planning and development > Airport maintenance > Bicycle and pedestr

    E-Print Network [OSTI]

    Multi-modal Transportation > Highway Transportation > Trucking > Railroad transportation > Public transit > Rural transportation > Rural transit > Freig pipeline transportation > Airport planning and development > Airport maintenance > Bicycle and pedestrian > Ports and waterways >>> Transportation ope

  16. http://tti.tamu.edu Multi-modal Transportation > Highway Transportation > Trucking > Railroad transportation > Public transit > Rural transportation > Rural transit > Freight

    E-Print Network [OSTI]

    http://tti.tamu.edu Multi-modal Transportation > Highway Transportation > Trucking > Railroad transportation > Public transit > Rural transportation > Rural transit > Freight pipeline transportation >>> Transportation operat > Freight traffic > Commodities > Travel time > Travel demand > http

  17. Ferrocarriles nacionales de Mexico: the existing rate structure for grain/soybean shipments and likely effect on freight rates as a result of railroad privatization in Mexico

    E-Print Network [OSTI]

    Neyer, David McAlister

    1994-01-01T23:59:59.000Z

    's existing rate structure and, based on collected FNM Waybill data, regression models were estimated. Projected costs in the privatized railroad system were estimated to provide a lower bound estimate of rates. An upper bound estimate of rates was based...

  18. artificial blood vessel: Topics by E-print Network

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

    Websites Summary: HEART AND BLOOD VESSELS CARDIOVASCULARCARDIOVASCULAR SYSTEMSYSTEM SYSTEM COMPONENTS Heart pumps blood though blood vessels where exchanges can take place...

  19. Automatic Lung Vessel Segmentation via Stacked Multiscale Feature Learning

    E-Print Network [OSTI]

    Toronto, University of

    Automatic Lung Vessel Segmentation via Stacked Multiscale Feature Learning Ryan Kiros, Karteek We introduce a representation learning approach to segmenting vessels in the lungs. Our algorithm

  20. Ion transport membrane module and vessel system

    DOE Patents [OSTI]

    Stein, VanEric Edward (Allentown, PA); Carolan, Michael Francis (Allentown, PA); Chen, Christopher M. (Allentown, PA); Armstrong, Phillip Andrew (Orefield, PA); Wahle, Harold W. (North Canton, OH); Ohrn, Theodore R. (Alliance, OH); Kneidel, Kurt E. (Alliance, OH); Rackers, Keith Gerard (Louisville, OH); Blake, James Erik (Uniontown, OH); Nataraj, Shankar (Allentown, PA); van Doorn, Rene Hendrik Elias (Obersulm-Willsbach, DE); Wilson, Merrill Anderson (West Jordan, UT)

    2008-02-26T23:59:59.000Z

    An ion transport membrane system comprising (a) a pressure vessel having an interior, an exterior, an inlet, and an outlet; (b) a plurality of planar ion transport membrane modules disposed in the interior of the pressure vessel and arranged in series, each membrane module comprising mixed metal oxide ceramic material and having an interior region and an exterior region, wherein any inlet and any outlet of the pressure vessel are in flow communication with exterior regions of the membrane modules; and (c) one or more gas manifolds in flow communication with interior regions of the membrane modules and with the exterior of the pressure vessel.The ion transport membrane system may be utilized in a gas separation device to recover oxygen from an oxygen-containing gas or as an oxidation reactor to oxidize compounds in a feed gas stream by oxygen permeated through the mixed metal oxide ceramic material of the membrane modules.

  1. Ion transport membrane module and vessel system

    DOE Patents [OSTI]

    Stein, VanEric Edward (Allentown, PA); Carolan, Michael Francis (Allentown, PA); Chen, Christopher M. (Allentown, PA); Armstrong, Phillip Andrew (Orefield, PA); Wahle, Harold W. (North Canton, OH); Ohrn, Theodore R. (Alliance, OH); Kneidel, Kurt E. (Alliance, OH); Rackers, Keith Gerard (Louisville, OH); Blake, James Erik (Uniontown, OH); Nataraj, Shankar (Allentown, PA); Van Doorn, Rene Hendrik Elias (Obersulm-Willsbach, DE); Wilson, Merrill Anderson (West Jordan, UT)

    2012-02-14T23:59:59.000Z

    An ion transport membrane system comprising (a) a pressure vessel having an interior, an exterior, an inlet, and an outlet; (b) a plurality of planar ion transport membrane modules disposed in the interior of the pressure vessel and arranged in series, each membrane module comprising mixed metal oxide ceramic material and having an interior region and an exterior region, wherein any inlet and any outlet of the pressure vessel are in flow communication with exterior regions of the membrane modules; and (c) one or more gas manifolds in flow communication with interior regions of the membrane modules and with the exterior of the pressure vessel. The ion transport membrane system may be utilized in a gas separation device to recover oxygen from an oxygen-containing gas or as an oxidation reactor to oxidize compounds in a feed gas stream by oxygen permeated through the mixed metal oxide ceramic material of the membrane modules.

  2. Future characteristics of Offshore Support Vessels

    E-Print Network [OSTI]

    Rose, Robin Sebastian Koske

    2011-01-01T23:59:59.000Z

    The objective of this thesis is to examine trends in Offshore Support Vessel (OSV) design and determine the future characteristics of OSVs based on industry insight and supply chain models. Specifically, this thesis focuses ...

  3. Neutron shielding panels for reactor pressure vessels

    DOE Patents [OSTI]

    Singleton, Norman R. (Murrysville, PA)

    2011-11-22T23:59:59.000Z

    In a nuclear reactor neutron panels varying in thickness in the circumferential direction are disposed at spaced circumferential locations around the reactor core so that the greatest radial thickness is at the point of highest fluence with lesser thicknesses at adjacent locations where the fluence level is lower. The neutron panels are disposed between the core barrel and the interior of the reactor vessel to maintain radiation exposure to the vessel within acceptable limits.

  4. PURE NIOBIUM AS A PRESSURE VESSEL MATERIAL

    SciTech Connect (OSTI)

    Peterson, T. J.; Carter, H. F.; Foley, M. H.; Klebaner, A. L.; Nicol, T. H.; Page, T. M.; Theilacker, J. C.; Wands, R. H.; Wong-Squires, M. L.; Wu, G. [Fermi National Accelerator Laboratory, Batavia, Illinois 60510 (United States)

    2010-04-09T23:59:59.000Z

    Physics laboratories around the world are developing niobium superconducting radio frequency (SRF) cavities for use in particle accelerators. These SRF cavities are typically cooled to low temperatures by direct contact with a liquid helium bath, resulting in at least part of the helium container being made from pure niobium. In the U.S., the Code of Federal Regulations allows national laboratories to follow national consensus pressure vessel rules or use of alternative rules which provide a level of safety greater than or equal to that afforded by ASME Boiler and Pressure Vessel Code. Thus, while used for its superconducting properties, niobium ends up also being treated as a material for pressure vessels. This report summarizes what we have learned about the use of niobium as a pressure vessel material, with a focus on issues for compliance with pressure vessel codes. We present results of a literature search for mechanical properties and tests results, as well as a review of ASME pressure vessel code requirements and issues.

  5. Current Trends and Future Challenges in the Freight Railroad Industry: Balancing Private Industry Interests and the Public Welfare

    E-Print Network [OSTI]

    Allen, Sarah; Kelson, Kendra; Migl, Hayden; Schmidt, Rodney; Shoemaker, David; Thomson, Heather

    2008-01-01T23:59:59.000Z

    ?dimensional?impact?on?the?public?welfare,?local,?state,?and?federal?government?entities?play?a? significant?role?in?ensuring?the?industry?operates?efficiently?and?safely.?In?October?2006,?the?Government? Accountability?Office?(GAO)?released?a?report?on...? ? Current?Trends?and?Future?Challenges?in? the?Freight?Railroad?Industry Balancing?Private?Industry?Interests?and?the?Public?Welfare? ? ? ? Sarah?Allen? Kendra?Kelson? Hayden?Migl? Rodney?Schmidt? David?Shoemaker? Heather?Thomson? ? ? A?Report...

  6. Vulnerability of Xylem Vessels to Cavitation in Sugar Maple. Scaling from Individual Vessels to

    E-Print Network [OSTI]

    Melcher, Peter

    nega- tive pressures (Dixon and Joly, 1895; Briggs, 1950) allows plants to power the movement of water to withstand tension-induced cavitation is typ- ically inferred from "vulnerability curves" generatedVulnerability of Xylem Vessels to Cavitation in Sugar Maple. Scaling from Individual Vessels

  7. EDS V25 containment vessel explosive qualification test report.

    SciTech Connect (OSTI)

    Rudolphi, John Joseph

    2012-04-01T23:59:59.000Z

    The V25 containment vessel was procured by the Project Manager, Non-Stockpile Chemical Materiel (PMNSCM) as a replacement vessel for use on the P2 Explosive Destruction Systems. It is the first EDS vessel to be fabricated under Code Case 2564 of the ASME Boiler and Pressure Vessel Code, which provides rules for the design of impulsively loaded vessels. The explosive rating for the vessel based on the Code Case is nine (9) pounds TNT-equivalent for up to 637 detonations. This limit is an increase from the 4.8 pounds TNT-equivalency rating for previous vessels. This report describes the explosive qualification tests that were performed in the vessel as part of the process for qualifying the vessel for explosive use. The tests consisted of a 11.25 pound TNT equivalent bare charge detonation followed by a 9 pound TNT equivalent detonation.

  8. Reactor pressure vessel with forged nozzles

    DOE Patents [OSTI]

    Desai, Dilip R. (Fremont, CA)

    1993-01-01T23:59:59.000Z

    Inlet nozzles for a gravity-driven cooling system (GDCS) are forged with a cylindrical reactor pressure vessel (RPV) section to which a support skirt for the RPV is attached. The forging provides enhanced RPV integrity around the nozzle and substantial reduction of in-service inspection costs by eliminating GDCS nozzle-to-RPV welds.

  9. An Analysis of Stream Culvert Fish Passage on the Navy Railroad Line between Bremerton and Shelton, Washington

    SciTech Connect (OSTI)

    May, Christopher W.; Miller, Martin C.; Southard, John A.

    2004-10-25T23:59:59.000Z

    The Navy railroad service line runs between Shelton, Bremerton, and Silverdale, and is used by the Navy to transfer freight to its facilities. It is also used by commercial clients to ship service items and bulk cargo for municipalities along portions of the route. Culverts of various size and construction convey streams and stormwater runoff under the railroad line. These allow transfer of water and, in some cases allow for passage of juvenile and adult salmon into waters upstream of the culverts. As part of this project, 21 culverts along a 34-mile reach (Shelton to Bremerton) of this railroad were surveyed to evaluate their function and ability to allow salmon to utilize the streams. The culverts and attached watersheds were evaluated using criteria developed by the Washington Department of Fish and Wildlife to assign a Priority Index (PI) to barriers present on each fish-bearing stream. The PI is a relative numeric rating indicator, assigned using consistent criteria related to the degree of potential habitat gained by removing barriers and improving the function of the watershed. Of the 21 culverts evaluated, five were found to be complete fish-passage barriers and six were found to be partial barriers, primarily to juvenile salmon. Three of these culverts had PI ratings above 10 and five others had ratings between 7 and 10. Corrective action can be taken based on any PI rating, but the WDFW normally assigns lower priority to projects with PI scores lower than 15. Several of the stream and culverts had previously been evaluated for structural integrity and function and have been scheduled for repair. A narrative indicating the condition of the culvert has been prepared as well as a table indicating the PI scores and a summary of recommendations for action for each culvert.

  10. Large-scale dynamic observation planning for unmanned surface vessels

    E-Print Network [OSTI]

    Miller, John V. (John Vaala)

    2007-01-01T23:59:59.000Z

    With recent advances in research and technology, autonomous surface vessel capabilities have steadily increased. These autonomous surface vessel technologies enable missions and tasks to be performed without the direction ...

  11. aftercastle masted vessel with aftercastle is found on a Spanish

    E-Print Network [OSTI]

    masted vessel with aftercastle is found on a Spanish ations it would have any idea of crusader ships aces for the new tack as large as the crusader vessels (

  12. Study Reveals Challenges and Opportunities Related to Vessels...

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

    Reveals Challenges and Opportunities Related to Vessels for U.S. Offshore Wind Study Reveals Challenges and Opportunities Related to Vessels for U.S. Offshore Wind October 1, 2013...

  13. Structural loading of cross deck connections for trimaran vessels

    E-Print Network [OSTI]

    Rhoads, Jason L

    2004-01-01T23:59:59.000Z

    This work investigates the fundamental relationships of wave loading on cross deck structures for trimaran vessels. In contrast with a monohull ship, trimaran vessels experience several possible structural loading cases ...

  14. RETINAL BLOOD VESSEL SEGMENTATION USING GEODESIC VOTING METHODS Youssef Rouchdy

    E-Print Network [OSTI]

    Cohen, Laurent

    RETINAL BLOOD VESSEL SEGMENTATION USING GEODESIC VOTING METHODS Youssef Rouchdy and Laurent D to segment retinal blood vessels are presented. Many authors have used minimal cost paths, or similarly on the use of a set of such geodesic paths to extract retinal blood vessels, using minimal interaction

  15. Modeling Torsion of Blood Vessels in Surgical Simulation and Planning

    E-Print Network [OSTI]

    Leow, Wee Kheng

    Modeling Torsion of Blood Vessels in Surgical Simulation and Planning Hao LI a,1 , Wee Kheng LEOW a hybrid approach for modeling torsion of blood vessels that undergo deformation and joining. The proposed model takes 3D mesh of the blood vessel as input. It first fits a generalized cylinder to extract

  16. INTERLINE 5.0 -- An expanded railroad routing model: Program description, methodology, and revised user`s manual

    SciTech Connect (OSTI)

    Johnson, P.E.; Joy, D.S. [Oak Ridge National Lab., TN (United States); Clarke, D.B.; Jacobi, J.M. [Tennessee Univ., Knoxville, TN (United States). Transportation Center

    1993-03-01T23:59:59.000Z

    A rail routine model, INTERLINE, has been developed at the Oak Ridge National Laboratory to investigate potential routes for transporting radioactive materials. In Version 5.0, the INTERLINE routing algorithms have been enhanced to include the ability to predict alternative routes, barge routes, and population statistics for any route. The INTERLINE railroad network is essentially a computerized rail atlas describing the US railroad system. All rail lines, with the exception of industrial spurs, are included in the network. Inland waterways and deep water routes along with their interchange points with the US railroadsystem are also included. The network contains over 15,000 rail and barge segments (links) and over 13,000 stations, interchange points, ports, and other locations (nodes). The INTERLINE model has been converted to operate on an IBM-compatible personal computer. At least a 286 computer with a hard disk containing approximately 6 MB of free space is recommended. Enhanced program performance will be obtained by using arandom-access memory drive on a 386 or 486 computer.

  17. From Cold War to cold vessels

    SciTech Connect (OSTI)

    Melrath, C.

    1996-09-01T23:59:59.000Z

    This article describes a former Soviet weapons plant which is converted to produce cryogenic vessels and other peaceful cylinders. In 1995, Byelocorp Scientific Inc. (BSI), a New York-based firm that specializes in transferring technologies developed in the former Soviet Union, began converting a huge military defense plant in Kazakhstan into civilian-industrial use. The nearly 750,000-square-foot factory in Almaty, the capital of the former Soviet republic, was previously used to manufacture torpedo shells and ballistic rocket casings. The old defense plant, which was known as Gidromash, will now manufacture cylinders of a kinder, gentler variety--cryogenic vessels. The Kazakhstan operation is being managed jointly with Supco Srl., an Italian manufacturing, engineering, and construction company. With financing from the US Department of Defense, BSI, Supco, and the Kazakhstan government, a new joint venture called Byelkamit (a combination of Byelocorp, Kazakhstan, America, and Italy) was established.

  18. Photoacoustic removal of occlusions from blood vessels

    DOE Patents [OSTI]

    Visuri, Steven R. (Livermore, CA); Da Silva, Luiz B. (Danville, CA); Celliers, Peter M. (Berkeley, CA); London, Richard A. (Orinda, CA); Maitland, IV, Duncan J. (Lafayette, CA); Esch, Victor C. (San Francisco, CA)

    2002-01-01T23:59:59.000Z

    Partial or total occlusions of fluid passages within the human body are removed by positioning an array of optical fibers in the passage and directing treatment radiation pulses along the fibers, one at a time, to generate a shock wave and hydrodynamics flows that strike and emulsify the occlusions. A preferred application is the removal of blood clots (thrombin and embolic) from small cerebral vessels to reverse the effects of an ischemic stroke. The operating parameters and techniques are chosen to minimize the amount of heating of the fragile cerebral vessel walls occurring during this photo acoustic treatment. One such technique is the optical monitoring of the existence of hydrodynamics flow generating vapor bubbles when they are expected to occur and stopping the heat generating pulses propagated along an optical fiber that is not generating such bubbles.

  19. Confinement Vessel Assay System: Calibration and Certification Report

    SciTech Connect (OSTI)

    Frame, Katherine C. [Los Alamos National Laboratory; Bourne, Mark M. [Los Alamos National Laboratory; Crooks, William J. [Los Alamos National Laboratory; Evans, Louise [Los Alamos National Laboratory; Gomez, Cipriano [Retired CMR-OPS: OPERATIONS; Mayo, Douglas R. [Los Alamos National Laboratory; Miko, David K. [Los Alamos National Laboratory; Salazar, William R. [Los Alamos National Laboratory; Stange, Sy [Los Alamos National Laboratory; Vigil, Georgiana M. [Los Alamos National Laboratory

    2012-07-17T23:59:59.000Z

    Los Alamos National Laboratory has a number of spherical confinement vessels (CVs) remaining from tests involving nuclear materials. These vessels have an inner diameter of 6 feet with 1 to 2 inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the vessels. The Confinement Vessel Assay System (CVAS) was developed to measure the amount of SNM in CVs before and after cleanout. Prior to cleanout, the system will be used to perform a verification measurement of each vessel. After cleanout, the system will be used to perform safeguards-quality assays of {le} 100-g {sup 239}Pu equivalent in a vessel for safeguards termination. The system was calibrated in three different mass regions (low, medium, and high) to cover the entire plutonium mass range that will be assayed. The low mass calibration and medium mass calibration were verified for material positioned in the center of an empty vessel. The systematic uncertainty due to position bias was estimated using an MCNPX model to simulate the response of the system to material localized at various points along the inner surface of the vessel. The background component due to cosmic ray spallation was determined by performing measurements of an empty vessel and comparing to measurements in the same location with no vessel present. The CVAS has been tested and calibrated in preparation for verification and safeguards measurements of CVs before and after cleanout.

  20. Improved ultrasonic testing of railroad rail for transverse discontinuities in the rail head using higher order Rayleigh (M{sub 21}) waves

    SciTech Connect (OSTI)

    Grewal, D.S. [Texas A and M Univ., College Station, TX (United States)

    1996-09-01T23:59:59.000Z

    Railroads are a very efficient and economical way of moving bulk commodities over large distances. However, like any other mode of transportation, this mode has its fair share of problems. It can be seen from the data (AAR, 1993) that over a period of time the average carload and trainload tonnage has gone up substantially. The average length of haul has also increased. However, the amount of new rail laid has actually decreased significantly. This presents a challenging opportunity to research better and more efficient ways of keeping the current railroad rail track in service and to find better ways of testing those rails to pre-empt any causes for accidents. An important factor to be considered here is the testing of the railroad rail for the amount of cold work in the top layer and the presence of microcracks in that layer. This article describes a surface guided acoustic mode with several inherent characteristics that make it particularly useful for detecting subsurface transverse flaws in railroad rail. These flaws have been previously in the blind spot of other ultrasonic NDT techniques.

  1. Autonomous Radiation Monitoring of Small Vessels

    SciTech Connect (OSTI)

    Fabris, Lorenzo [ORNL; Hornback, Donald Eric [ORNL

    2010-01-01T23:59:59.000Z

    Small private vessels are one avenue by which nuclear materials may be smuggled across international borders. While one can contemplate using the terrestrial approach of radiation portal monitors on the navigable waterways that lead to many ports, these systems are ill-suited to the problem. They require vehicles to pass at slow speeds between two closely-spaced radiation sensors, relying on the uniformity of vehicle sizes to space the detectors, and on proximity to link an individual vehicle to its radiation signature. In contrast to roadways where lanes segregate vehicles, and motion is well controlled by inspection booths; channels, inlets, and rivers present chaotic traffic patterns populated by vessels of all sizes. We have developed a unique solution to this problem based on our portal-less portal monitor instrument that is designed to handle free-flowing traffic on roadways with up to five-traffic lanes. The instrument uses a combination of visible-light and gamma-ray imaging to acquire and link radiation images to individual vehicles. It was recently tested in a maritime setting. In this paper we present the instrument, how it functions, and the results of the recent tests.

  2. Application of Computational Physics: Blood Vessel Constrictions and Medical Infuses

    E-Print Network [OSTI]

    Suprijadi; Mohamad Rendi; Petrus Subekti; Sparisoma Viridi

    2013-12-14T23:59:59.000Z

    Application of computation in many fields are growing fast in last two decades. Increasing on computation performance helps researchers to understand natural phenomena in many fields of science and technology including in life sciences. Computational fluid dynamic is one of numerical methods which is very popular used to describe those phenomena. In this paper we propose moving particle semi-implicit (MPS) and molecular dynamics (MD) to describe different phenomena in blood vessel. The effect of increasing the blood pressure on vessel wall will be calculate using MD methods, while the two fluid blending dynamics will be discussed using MPS. Result from the first phenomenon shows that around 80% of constriction on blood vessel make blood vessel increase and will start to leak on vessel wall, while from the second phenomenon the result shows the visualization of two fluids mixture (drugs and blood) influenced by ratio of drugs debit to blood debit. Keywords: molecular dynamic, blood vessel, fluid dynamic, moving particle semi implicit.

  3. Application of Computational Physics: Blood Vessel Constrictions and Medical Infuses

    E-Print Network [OSTI]

    Suprijadi,; Subekti, Petrus; Viridi, Sparisoma

    2013-01-01T23:59:59.000Z

    Application of computation in many fields are growing fast in last two decades. Increasing on computation performance helps researchers to understand natural phenomena in many fields of science and technology including in life sciences. Computational fluid dynamic is one of numerical methods which is very popular used to describe those phenomena. In this paper we propose moving particle semi-implicit (MPS) and molecular dynamics (MD) to describe different phenomena in blood vessel. The effect of increasing the blood pressure on vessel wall will be calculate using MD methods, while the two fluid blending dynamics will be discussed using MPS. Result from the first phenomenon shows that around 80% of constriction on blood vessel make blood vessel increase and will start to leak on vessel wall, while from the second phenomenon the result shows the visualization of two fluids mixture (drugs and blood) influenced by ratio of drugs debit to blood debit. Keywords: molecular dynamic, blood vessel, fluid dynamic, movin...

  4. alternative reactor vessel: Topics by E-print Network

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

    FOR ADDED ESCORT van Dorp, Johan Ren 111 FIRE Vacuum Vessel Design and Analysis Plasma Physics and Fusion Websites Summary: pressure, coolant pressure - EM loads on...

  5. Forum Agenda: International Hydrogen Fuel and Pressure Vessel...

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

    Hydrogen Fuel and Pressure Vessel Forum 2010 Proceedings Workshop Agenda: Compressed Natural Gas and Hydrogen Fuels, Lesssons Learned for the Safe Deployment of Vehicles...

  6. asme pressure vessels: Topics by E-print Network

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

    Program & Sensors Hydrogen Delivery Composite Overwrapped Pressure Vessels (COPVs) Pipeline for Off-Board Hydrogen: L-C Atmosphere: 1500 psi H2, ambient pressure Air...

  7. asme pressure vessel: Topics by E-print Network

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

    Program & Sensors Hydrogen Delivery Composite Overwrapped Pressure Vessels (COPVs) Pipeline for Off-Board Hydrogen: L-C Atmosphere: 1500 psi H2, ambient pressure Air...

  8. alloy pressure vessels: Topics by E-print Network

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

    Program & Sensors Hydrogen Delivery Composite Overwrapped Pressure Vessels (COPVs) Pipeline for Off-Board Hydrogen: L-C Atmosphere: 1500 psi H2, ambient pressure Air...

  9. alloy pressure vessel: Topics by E-print Network

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

    Program & Sensors Hydrogen Delivery Composite Overwrapped Pressure Vessels (COPVs) Pipeline for Off-Board Hydrogen: L-C Atmosphere: 1500 psi H2, ambient pressure Air...

  10. aged pressure vessel: Topics by E-print Network

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

    Program & Sensors Hydrogen Delivery Composite Overwrapped Pressure Vessels (COPVs) Pipeline for Off-Board Hydrogen: L-C Atmosphere: 1500 psi H2, ambient pressure Air...

  11. Webinar: Material Characterization of Storage Vessels for Fuel Cell Forklifts

    Broader source: Energy.gov [DOE]

    Video recording of the webinar titled, Material Characterization of Storage Vessels for Fuel Cell Forklifts, originally presented on August 14, 2012.

  12. Method and device for supporting blood vessels during anastomosis

    DOE Patents [OSTI]

    Doss, J.D.

    1985-05-20T23:59:59.000Z

    A device and method for preventing first and second severed blood vessels from collapsing during attachment to each other. The device comprises a dissolvable non-toxic stent that is sufficiently rigid to prevent the blood vessels from collapsing during anastomosis. The stent can be hollow or have passages to permit blood flow before it dissolves. A single stent can be inserted with an end in each of the two blood vessels or separate stents can be inserted into each blood vessel. The stent may include a therapeutically effective amount of a drug which is slowly released into the blood stream as the stent dissolves. 12 figs.

  13. Water Quality Impacts of Bunker Silos

    E-Print Network [OSTI]

    Balser, Teri C.

    and water) as well as feed particles and soil transported by flow. #12;Management and Disposal Options 1 Engineering Department UW ­ Madison with assistance from Larry D. Geohring Biological and Environmental Engineering Department Cornell University Area Soil and Soil & Water Meetings November 28 ­ December 7, 2006

  14. Bunker Fuel | OpenEI Community

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are beingZealand Jump to:EzfeedflagBiomassSustainable andBucoda, Washington: Energy(B2G)

  15. Heat-transfer coefficients in agitated vessels. Sensible heat models

    SciTech Connect (OSTI)

    Kumpinsky, E. [Ashland Chemical Co., Columbus, OH (United States). Research and Development Dept.

    1995-12-01T23:59:59.000Z

    Transient models for sensible heat were developed to assess the thermal performance of agitated vessels with coils and jackets. Performance is quantified with the computation of heat-transfer coefficients by introducing vessel heating and cooling data into model equations. Of the two model categories studied, differential and macroscopic, the latter is preferred due to mathematical simplicity and lower sensitivity to experimental data variability.

  16. Austenite Grain Growth in a Nuclear Pressure Vessel Steel

    E-Print Network [OSTI]

    Cambridge, University of

    . Cogswellb , H. K. D. H. Bhadeshiaa aDepartment of Materials Science and Metallurgy, University of Cambridge vessels, partly because the qualifica- tion of such materials requires an enormous amount of time-consuming work. The reactor pressure vessels (RPV) in particular have demanding requirements for tensile strength

  17. Coalesced Martensite in Pressure Vessel Steels Hector Pous-Romero

    E-Print Network [OSTI]

    Cambridge, University of

    Coalesced Martensite in Pressure Vessel Steels Hector Pous-Romero Department of Materials Science.ac.uk Harry Bhadeshia Department of Materials Science & Metallurgy University of Cambridge Cambridge RPV Reactor pressure vessels. SEM Scanning electron microscopy. HAZ Heat affected zone. Bs Bainite

  18. Simultaneous Irradiation and Imaging of Blood Vessels During Pulsed

    E-Print Network [OSTI]

    Barton, Jennifer K.

    energy produced hemorrhage. In larger vessels, coagula often were attached to the superficial vessel wall; port wine stains INTRODUCTION Previous studies examining the effect of la- ser irradiation on cutaneous preparation. The short pulse duration illus- trated an extreme; energy was deposited quickly Contract grant

  19. Commercial marine vessel contributions to emission inventories. Final report

    SciTech Connect (OSTI)

    Not Available

    1991-10-07T23:59:59.000Z

    The Clean Air Act Amendments of 1990 require the US Environmental Protection Agency (EPA) to conduct a survey of emissions from combustion engines associates with non-road vehicles and stationary sources. Among the emission source categories under scrutiny of the EPA are commercial marine vessels. This group of sources includes revenue vessels operated on US ports and waterways in such diverse pursuits as international and domestic trade, port and ship service, offshore and coastal industry, and passenger transport. For the purposes of the study, EPA is assessing commercial marine vessel operations at selected ports around the country which are characterized by a high level of commercial marine vessel activity. Booz-Allen has been retained by the EPA to assist in developing emission inventories from marine vessels for up to six ports, based on vessel arrival/departure data, are believed to exhibit high levels of marine generated emissions. Booz-Allen developed a listing of the top 20 major ports in terms of total vessel activity (as measured by annual tonnage of cargo and annual vessel calls).

  20. Optimal Short-Range Routing of Vessels in a Seaway

    E-Print Network [OSTI]

    Smith, Robert L.

    Optimal Short-Range Routing of Vessels in a Seaway Irina S. Dolinskaya¹ Miltiadis Kotinis² Michael Industrial and Operations Engineering 1205 Beal Avenue Ann Arbor, Michigan 48109 ²Old Dominion University Short-Range Routing of Vessels in a Seaway Dolinskaya, I. S.1 , Kotinis, M.2 , Parsons, M. G.3

  1. Nuclear reactor construction with bottom supported reactor vessel

    DOE Patents [OSTI]

    Sharbaugh, John E. (Bullskin Township, Fayette County, PA)

    1987-01-01T23:59:59.000Z

    An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment structure base mat so as to insulate the reactor vessel bottom end wall from the containment structure base mat and allow the reactor vessel bottom end wall to freely expand as it heats up while providing continuous support thereof. Further, a deck is supported upon the side wall of the containment structure above the top open end of the reactor vessel, and a plurality of serially connected extendible and retractable annular bellows extend between the deck and the top open end of the reactor vessel and flexibly and sealably interconnect the reactor vessel at its top end to the deck. An annular guide ring is disposed on the containment structure and extends between its side wall and the top open end of the reactor vessel for providing lateral support of the reactor vessel top open end by limiting imposition of lateral loads on the annular bellows by the occurrence of a lateral seismic event.

  2. Welding the AT-400A Containment Vessel

    SciTech Connect (OSTI)

    Brandon, E.

    1998-11-01T23:59:59.000Z

    Early in 1994, the Department of Energy assigned Sandia National Laboratories the responsibility for designing and providing the welding system for the girth weld for the AT-400A containment vessel. (The AT-400A container is employed for the shipment and long-term storage of the nuclear weapon pits being returned from the nation's nuclear arsenal.) Mason Hanger Corporation's Pantex Plant was chosen to be the production facility. The project was successfully completed by providing and implementing a turnkey welding system and qualified welding procedure at the Pantex Plant. The welding system was transferred to Pantex and a pilot lot of 20 AT-400A containers with W48 pits was welded in August 1997. This document is intended to bring together the AT-400A welding system and product (girth weld) requirements and the activities conducted to meet those requirements. This document alone is not a complete compilation of the welding development activities but is meant to be a summary to be used with the applicable references.

  3. Radiation effects on reactor pressure vessel supports

    SciTech Connect (OSTI)

    Johnson, R.E. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering Technology; Lipinski, R.E. [Idaho National Engineering Lab., Rockville, MD (United States)

    1996-05-01T23:59:59.000Z

    The purpose of this report is to present the findings from the work done in accordance with the Task Action Plan developed to resolve the Nuclear Regulatory Commission (NRC) Generic Safety Issue No. 15, (GSI-15). GSI-15 was established to evaluate the potential for low-temperature, low-flux-level neutron irradiation to embrittle reactor pressure vessel (RPV) supports to the point of compromising plant safety. An evaluation of surveillance samples from the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) had suggested that some materials used for RPV supports in pressurized-water reactors could exhibit higher than expected embrittlement rates. However, further tests designed to evaluate the applicability of the HFIR data to reactor RPV supports under operating conditions led to the conclusion that RPV supports could be evaluated using traditional method. It was found that the unique HFIR radiation environment allowed the gamma radiation to contribute significantly to the embrittlement. The shielding provided by the thick steel RPV shell ensures that degradation of RPV supports from gamma irradiation is improbable or minimal. The findings reported herein were used, in part, as the basis for technical resolution of the issue.

  4. Reactor Pressure Vessel Head Packaging & Disposal

    SciTech Connect (OSTI)

    Wheeler, D. M.; Posivak, E.; Freitag, A.; Geddes, B.

    2003-02-26T23:59:59.000Z

    Reactor Pressure Vessel (RPV) Head replacements have come to the forefront due to erosion/corrosion and wastage problems resulting from the susceptibility of the RPV Head alloy steel material to water/boric acid corrosion from reactor coolant leakage through the various RPV Head penetrations. A case in point is the recent Davis-Besse RPV Head project, where detailed inspections in early 2002 revealed significant wastage of head material adjacent to one of the Control Rod Drive Mechanism (CRDM) nozzles. In lieu of making ASME weld repairs to the damaged head, Davis-Besse made the decision to replace the RPV Head. The decision was made on the basis that the required weld repair would be too extensive and almost impractical. This paper presents the packaging, transport, and disposal considerations for the damaged Davis-Besse RPV Head. It addresses the requirements necessary to meet Davis Besse needs, as well as the regulatory criteria, for shipping and burial of the head. It focuses on the radiological characterization, shipping/disposal package design, site preparation and packaging, and the transportation and emergency response plans that were developed for the Davis-Besse RPV Head project.

  5. Start-up control system and vessel for LMFBR

    DOE Patents [OSTI]

    Durrant, Oliver W. (Akron, OH); Kakarala, Chandrasekhara R. (Clinton, OH); Mandel, Sheldon W. (Galesburg, IL)

    1987-01-01T23:59:59.000Z

    A reflux condensing start-up system includes a steam generator, a start-up vessel connected parallel to the steam generator, a main steam line connecting steam outlets of the steam generator and start-up vessel to a steam turbine, a condenser connected to an outlet of the turbine and a feedwater return line connected between the condenser and inlets of the steam generator and start-up vessel. The start-up vessel has one or more heaters at the bottom thereof for heating feedwater which is supplied over a start-up line to the start-up vessel. Steam is thus generated to pressurize the steam generator before the steam generator is supplied with a heat transfer medium, for example liquid sodium, in the case of a liquid metal fast breeder reactor. The start-up vessel includes upper and lower bulbs with a smaller diameter mid-section to act as water and steam reservoirs. The start-up vessel can thus be used not only in a start-up operation but as a mixing tank, a water storage tank and a level control at low loads for controlling feedwater flow.

  6. Start-up control system and vessel for LMFBR

    DOE Patents [OSTI]

    Durrant, Oliver W. (Akron, OH); Kakarala, Chandrasekhara R. (Clinton, OH); Mandel, Sheldon W. (Galesburg, IL)

    1987-01-01T23:59:59.000Z

    A reflux condensing start-up system comprises a steam generator, a start-up vessel connected parallel to the steam generator, a main steam line connecting steam outlets of the steam generator and start-up vessel to a steam turbine, a condenser connected to an outlet of the turbine and a feedwater return line connected between the condenser and inlets of the steam generator and start-up vessel. The start-up vessel has one or more heaters at the bottom thereof for heating feedwater which is supplied over a start-up line to the start-up vessel. Steam is thus generated to pressurize the steam generator before the steam generator is supplied with a heat transfer medium, for example liquid sodium, in the case of a liquid metal fast breeder reactor. The start-up vessel includes upper and lower bulbs with a smaller diameter mid-section to act as water and steam reservoirs. The start-up vessel can thus be used not only in a start-up operation but as a mixing tank, a water storage tank and a level control at low loads for controlling feedwater flow.

  7. Reactor Vessel and Reactor Vessel Internals Segmentation at Zion Nuclear Power Station - 13230

    SciTech Connect (OSTI)

    Cooke, Conrad; Spann, Holger [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)] [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)

    2013-07-01T23:59:59.000Z

    Zion Nuclear Power Station (ZNPS) is a dual-unit Pressurized Water Reactor (PWR) nuclear power plant located on the Lake Michigan shoreline, in the city of Zion, Illinois approximately 64 km (40 miles) north of Chicago, Illinois and 67 km (42 miles) south of Milwaukee, Wisconsin. Each PWR is of the Westinghouse design and had a generation capacity of 1040 MW. Exelon Corporation operated both reactors with the first unit starting production of power in 1973 and the second unit coming on line in 1974. The operation of both reactors ceased in 1996/1997. In 2010 the Nuclear Regulatory Commission approved the transfer of Exelon Corporation's license to ZionSolutions, the Long Term Stewardship subsidiary of EnergySolutions responsible for the decommissioning of ZNPS. In October 2010, ZionSolutions awarded Siempelkamp Nuclear Services, Inc. (SNS) the contract to plan, segment, remove, and package both reactor vessels and their respective internals. This presentation discusses the tools employed by SNS to remove and segment the Reactor Vessel Internals (RVI) and Reactor Vessels (RV) and conveys the recent progress. SNS's mechanical segmentation tooling includes the C-HORCE (Circumferential Hydraulically Operated Cutting Equipment), BMT (Bolt Milling Tool), FaST (Former Attachment Severing Tool) and the VRS (Volume Reduction Station). Thermal segmentation of the reactor vessels will be accomplished using an Oxygen- Propane cutting system. The tools for internals segmentation were designed by SNS using their experience from other successful reactor and large component decommissioning and demolition (D and D) projects in the US. All of the designs allow for the mechanical segmentation of the internals remotely in the water-filled reactor cavities. The C-HORCE is designed to saw seven circumferential cuts through the Core Barrel and Thermal Shield walls with individual thicknesses up to 100 mm (4 inches). The BMT is designed to remove the bolts that fasten the Baffle Plates to the Baffle Former Plates. The FaST is designed to remove the Baffle Former Plates from the Core Barrel. The VRS further volume reduces segmented components using multiple configurations of the 38i and horizontal reciprocating saws. After the successful removal and volume reduction of the Internals, the RV will be segmented using a 'First in the US' thermal cutting process through a co-operative effort with Siempelkamp NIS Ingenieurgesellschaft mbH using their experience at the Stade NPP and Karlsruhe in Germany. SNS mobilized in the fall of 2011 to commence execution of the project in order to complete the RVI segmentation, removal and packaging activities for the first unit (Unit 2) by end of the 2012/beginning 2013 and then mobilize to the second unit, Unit 1. Parallel to the completion of the segmentation of the reactor vessel internals at Unit 1, SNS will segment the Unit 2 pressure vessel and at completion move to Unit 1. (authors)

  8. PERFORMANCE OF A CONTAINMENT VESSEL CLOSURE FOR RADIOACTIVE GAS CONTENTS

    SciTech Connect (OSTI)

    Blanton, P.; Eberl, K.

    2010-07-09T23:59:59.000Z

    This paper presents a summary of the design and testing of the containment vessel closure for the Bulk Tritium Shipping Package (BTSP). This package is a replacement for a package that has been used to ship tritium in a variety of content configurations and forms since the early 1970s. The containment vessel closure incorporates features specifically designed for the containment of tritium when subjected to the normal and hypothetical conditions required of Type B radioactive material shipping Packages. The paper discusses functional performance of the containment vessel closure of the BTSP prototype packages and separate testing that evaluated the performance of the metallic C-Rings used in a mock BTSP closure.

  9. Float level switch for a nuclear power plant containment vessel

    DOE Patents [OSTI]

    Powell, J.G.

    1993-11-16T23:59:59.000Z

    This invention is a float level switch used to sense rise or drop in water level in a containment vessel of a nuclear power plant during a loss of coolant accident. The essential components of the device are a guide tube, a reed switch inside the guide tube, a float containing a magnetic portion that activates a reed switch, and metal-sheathed, ceramic-insulated conductors connecting the reed switch to a monitoring system outside the containment vessel. Special materials and special sealing techniques prevent failure of components and allow the float level switch to be connected to a monitoring system outside the containment vessel. 1 figures.

  10. Processing and analysis techniques involving in-vessel material generation

    DOE Patents [OSTI]

    Schabron, John F. (Laramie, WY); Rovani, Jr., Joseph F. (Laramie, WY)

    2011-01-25T23:59:59.000Z

    In at least one embodiment, the inventive technology relates to in-vessel generation of a material from a solution of interest as part of a processing and/or analysis operation. Preferred embodiments of the in-vessel material generation (e.g., in-vessel solid material generation) include precipitation; in certain embodiments, analysis and/or processing of the solution of interest may include dissolution of the material, perhaps as part of a successive dissolution protocol using solvents of increasing ability to dissolve. Applications include, but are by no means limited to estimation of a coking onset and solution (e.g., oil) fractionating.

  11. Float level switch for a nuclear power plant containment vessel

    DOE Patents [OSTI]

    Powell, James G. (Clifton Park, NY)

    1993-01-01T23:59:59.000Z

    This invention is a float level switch used to sense rise or drop in water level in a containment vessel of a nuclear power plant during a loss of coolant accident. The essential components of the device are a guide tube, a reed switch inside the guide tube, a float containing a magnetic portion that activates a reed switch, and metal-sheathed, ceramic-insulated conductors connecting the reed switch to a monitoring system outside the containment vessel. Special materials and special sealing techniques prevent failure of components and allow the float level switch to be connected to a monitoring system outside the containment vessel.

  12. Processing and analysis techniques involving in-vessel material generation

    DOE Patents [OSTI]

    Schabron, John F. (Laramie, WY); Rovani, Jr., Joseph F. (Laramie, WY)

    2012-09-25T23:59:59.000Z

    In at least one embodiment, the inventive technology relates to in-vessel generation of a material from a solution of interest as part of a processing and/or analysis operation. Preferred embodiments of the in-vessel material generation (e.g., in-vessel solid material generation) include precipitation; in certain embodiments, analysis and/or processing of the solution of interest may include dissolution of the material, perhaps as part of a successive dissolution protocol using solvents of increasing ability to dissolve. Applications include, but are by no means limited to estimation of a coking onset and solution (e.g., oil) fractionating.

  13. PISCES FY11 Research Highlight Tritium accumulation within the ITER vessel is expected to be dominated

    E-Print Network [OSTI]

    PISCES FY11 Research Highlight Tritium accumulation within the ITER vessel is expected vessel. Another possible technique to mitigate tritium accumulation in these codeposited surfaces

  14. Subsea Kick Detection on Floating Vessels: A Parametric Study

    E-Print Network [OSTI]

    Collette, Eric Peter

    2013-07-22T23:59:59.000Z

    SUBSEA KICK DETECTION ON FLOATING VESSELS: A PARAMETRIC STUDY A Thesis by ERIC P. COLLETTE Submitted to the Office of Graduate Studies of Texas A&M University in partial fulfillment of the requirements for the degree...

  15. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel...

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

    This document presents the results of one of these workshops, the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPVs). These workshops made a substantial effort to...

  16. 2005 ASME Pressure Vessels and Piping Conference Denver, Colorado, USA

    E-Print Network [OSTI]

    Özer, Mutlu

    1 DRAFT 2005 ASME Pressure Vessels and Piping Conference Denver, Colorado, USA July 17-21, 2005 subjected to lateral earthquake loads. The results are verified with different codes (e.g. Eurocode8, API

  17. DESIGN OF THE ITER IN-VESSEL COILS

    SciTech Connect (OSTI)

    Neumeyer, C; Bryant, L; Chrzanowski, J; Feder, R; Gomez, M; Heitzenroeder, P; Kalish, M; Lipski, A; Mardenfeld, M; Simmons, R; Titus, P; Zatz, I; Daly, E; Martin, A; Nakahira, M; Pillsbury, R; Feng, J; Bohm, T; Sawan, M; Stone, H; Griffiths, I

    2010-11-27T23:59:59.000Z

    The ITER project is considering the inclusion of two sets of in-vessel coils, one to mitigate the effect of Edge Localized Modes (ELMs) and another to provide vertical stabilization (VS). The in-vessel location (behind the blanket shield modules, mounted to the vacuum vessel inner wall) presents special challenges in terms of nuclear radiation (~3000 MGy) and temperature (100oC vessel during operations, 200oC during bakeout). Mineral insulated conductors are well suited to this environment but are not commercially available in the large cross section required. An R&D program is underway to demonstrate the production of mineral insulated (MgO or Spinel) hollow copper conductor with stainless steel jacketing needed for these coils. A preliminary design based on this conductor technology has been developed and is presented herein.

  18. aging blood vessels: Topics by E-print Network

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

    Japan is one 302 VESSEL TRAFFIC RISK ASSESSMENT (VTRA) 2010 Engineering Websites Summary: Oil Loss Dr. J. Rene van Dorp and Dr. Jason R.W Merrick 12132013 1 GW-VCU December 2013...

  19. abdominal blood vessels: Topics by E-print Network

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

    Terri 311 VESSEL TRAFFIC RISK ASSESSMENT (VTRA) 2010 Engineering Websites Summary: Oil Loss Dr. J. Rene van Dorp and Dr. Jason R.W Merrick 12132013 1 GW-VCU December 2013...

  20. abnormal blood vessels: Topics by E-print Network

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

    Japan is one 327 VESSEL TRAFFIC RISK ASSESSMENT (VTRA) 2010 Engineering Websites Summary: Oil Loss Dr. J. Rene van Dorp and Dr. Jason R.W Merrick 12132013 1 GW-VCU December 2013...

  1. alters blood vessel: Topics by E-print Network

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

    G Tien, Joe 308 VESSEL TRAFFIC RISK ASSESSMENT (VTRA) 2010 Engineering Websites Summary: Oil Loss Dr. J. Rene van Dorp and Dr. Jason R.W Merrick 12132013 1 GW-VCU December 2013...

  2. Simple program calculates partial liquid volumes in vessels

    SciTech Connect (OSTI)

    Koch, P.

    1992-04-13T23:59:59.000Z

    This paper reports on a simple calculator program which solves problems of partial liquid volumes for a variety of storage and process vessels, including inclined cylindrical vessels and those with conical heads. Engineers in the oil refining and chemical industries are often confronted with the problem of estimating partial liquid volumes in storage tanks or process vessels. Cistern, the calculator program presented here, allows fast and accurate resolution of problems for a wide range of vessels without user intervention, other than inputting the problem data. Running the program requires no mathematical skills. Cistern is written for Hewlett-Packard HP 41CV or HP 41CX programmable calculators (or HP 41C with extended memory modules).

  3. Hydrodynamic evaluation of high-speed semi-SWATH vessels

    E-Print Network [OSTI]

    Guttenplan, Adam (Adam David)

    2007-01-01T23:59:59.000Z

    High-speed semi-displacement vessels have enjoyed rapid development and widespread use over the past 25 years. Concurrent with their growth as viable commercial and naval platforms, has been the advancement of three-dimensional ...

  4. A cog-like vessel from the Netherlands

    E-Print Network [OSTI]

    Van de Moortel, Aleydis Maria P. A.

    1987-01-01T23:59:59.000Z

    , more than thirty iconographic representations, mostly medieval city seals, have been discovered. 4 They show that cogs were compact and tubby vessels with a sharply built lower hull, combining a large cargo capacity with good sailing qualities.... The broad central part of the vessel immedistelv suaaested it had been a merchantman, but no trace of cargo was found. The onlv contents were some fraaments of bricks and ceramics. a Few iron serape' some smail cattle bones, and, under the ceiling...

  5. Using SA508/533 for the HTGR Vessel Material

    SciTech Connect (OSTI)

    Larry Demick

    2012-06-01T23:59:59.000Z

    This paper examines the influence of High Temperature Gas-cooled Reactor (HTGR) module power rating and normal operating temperatures on the use of SA508/533 material for the HTGR vessel system with emphasis on the calculated times at elevated temperatures approaching or exceeding ASME Code Service Limits (Levels B&C) to which the reactor pressure vessel could be exposed during postulated pressurized and depressurized conduction cooldown events over its design lifetime.

  6. Streamlined Approach for Environmental Restoration (SAFER) Plan for Corrective Action Unit 539: Area 25 and Area 26 Railroad Tracks, Nevada Test Site, Nevada, Revision 0

    SciTech Connect (OSTI)

    Mark Krauss

    2010-06-01T23:59:59.000Z

    This Streamlined Approach for Environmental Restoration (SAFER) Plan addresses the actions needed to achieve closure for Corrective Action Unit (CAU) 539, Areas 25 and 26 Railroad Tracks, as identified in the Federal Facility Agreement and Consent Order (FFACO). A modification to the FFACOwas approved in May 2010 to transfer the two Railroad Tracks corrective action sites (CASs) from CAU 114 into CAU539. The two CASs are located in Areas 25 and 26 of the Nevada Test Site: • 25-99-21, Area 25 Railroad Tracks • 26-99-05, Area 26 Railroad Tracks This plan provides the methodology for field activities needed to gather the necessary information for closing the two CASs. There is sufficient information and process knowledge from historical documentation and investigations of similar sites regarding the expected nature and extent of potential contaminants to recommend closure of the CAU 539 Railroad Tracks CASs using the SAFER process. Additional information will be obtained by conducting a field investigation before selecting the appropriate corrective action for each CAS. The results of the field investigation should support a defensible recommendation that no further corrective action is necessary. If it is determined that complete clean closure cannot be accomplished during the SAFER, then a hold point will have been reached and the Nevada Division of Environmental Protection (NDEP) will be consulted to determine whether the remaining contamination will be closed under the alternative corrective action of closure in place with use restrictions. This will be presented in a closure report that will be prepared and submitted to the NDEP for review and approval. The sites will be investigated based on the data quality objectives (DQOs) developed on December 14, 2009, by representatives of U.S.Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office; Navarro Nevada Environmental Services, LLC (NNES); and National Security Technologies, LLC. The DQO process has been used to identify and define the type, amount, and quality of data needed to determine and implement appropriate corrective actions for each Railroad Tracks CAS in CAU 539. The following text summarizes the SAFER activities that will support the closure of CAU 539: • Perform site preparation activities (e.g., utilities clearances, radiological surveys). • Collect in situ dose measurements. • Collect environmental samples from designated target populations (e.g., lead bricks) to confirm or disprove the presence of contaminants of concern (COCs) as necessary to supplement existing information. • If no COCs are present at a CAS, establish no further action as the corrective action. • If COCs exist, collect environmental samples from designated target populations (e.g., clean soil adjacent to contaminated soil) and submit for laboratory analyses to define the extent of COC contamination. If a COC is present at a CAS, NNES will consult NDEP to determine the path forward, then either: • Establish clean closure as the corrective action. The material to be remediated will be removed, disposed of as waste, and verification samples will be collected from remaining soil, or • Establish closure in place as the corrective action and implement the appropriate use restrictions.

  7. Standard practice for evaluation of surveillance capsules from light-water moderated nuclear power reactor vessels

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2010-01-01T23:59:59.000Z

    Standard practice for evaluation of surveillance capsules from light-water moderated nuclear power reactor vessels

  8. Predictive Simulation of Bidirectional Glenn Shunt Using a Hybrid Blood Vessel Model

    E-Print Network [OSTI]

    Leow, Wee Kheng

    Predictive Simulation of Bidirectional Glenn Shunt Using a Hybrid Blood Vessel Model Hao Li1 to model the deformation of blood vessels. The hybrid blood vessel model consists of a reference Cosserat rod and a surface mesh. The reference Cosserat rod models the blood vessel's global bending

  9. Photoacoustic spectroscopy sample array vessel and photoacoustic spectroscopy method for using the same

    DOE Patents [OSTI]

    Amonette, James E.; Autrey, S. Thomas; Foster-Mills, Nancy S.; Green, David

    2005-03-29T23:59:59.000Z

    Methods and apparatus for analysis of multiple samples by photoacoustic spectroscopy are disclosed. Particularly, a photoacoustic spectroscopy sample array vessel including a vessel body having multiple sample cells connected thereto is disclosed. At least one acoustic detector is acoustically coupled with the vessel body. Methods for analyzing the multiple samples in the sample array vessels using photoacoustic spectroscopy are provided.

  10. Fuel efficient train operations: A preliminary investigation with the locomotive data acquisition package on the Union Pacific Railroad. Interim report Jun-Nov 80

    SciTech Connect (OSTI)

    Larsen, K.W.

    1981-12-01T23:59:59.000Z

    The Locomotive Data Acquisition Package (LDAP) is a research data acquisition system designed for use on board diesel locomotives. Between June 1980 and November 1980, a fuel efficient train operations experiment was conducted in cooperation with the Union Pacific Railroad using the LDAP. The experiment served both as a checkout of the LDAP unit, and as a preliminary study of the relationship between train handling and fuel consumption. The LDAP unit was operated on a Union Pacific Locomotive for 127 days during which data were collected for 53,936 miles of locomotive operation.

  11. Investigation of vessel exterior air cooling for a HLMC reactor

    SciTech Connect (OSTI)

    Sienicki, J. J.; Spencer, B. W.

    2000-01-13T23:59:59.000Z

    The Secure Transportable Autonomous Reactor (STAR) concept under development at Argonne National Laboratory provides a small (300 MWt) reactor module for steam supply that incorporates design features to attain proliferation resistance, heightened passive safety, and improved cost competitiveness through extreme simplification. Examples are the achievement of 100%+ natural circulation heat removal from the low power density/low pressure drop ultra-long lifetime core and utilization of lead-bismuth eutectic (LBE) coolant enabling elimination of main coolant pumps as well as the need for an intermediate heat transport circuit. It is required to provide a passive means of removing decay heat and effecting reactor cooldown in the event that the normal steam generator heat sink, including its normal shutdown heat removal mode, is postulated to be unavailable. In the present approach, denoted as the Reactor Exterior Cooling System (RECS), passive decay heat removal is provided by cooling the outside of the containment/guard vessel with air. RECS is similar to the Reactor Vessel Auxiliary Cooling System (RVACS) incorporated into the PRISM design. However, to enhance the heat removal, RECS incorporates fins on the containment vessel exterior to enhance heat transfer to air as well as removable steel venetian conductors that provide a conduction heat transfer path across the reactor vessel-containment vessel gap to enhance heat transfer between the vessels. The objective of the present work is to investigate the effectiveness of air cooling in removing heat from the vessel and limiting the coolant temperature increase following a sudden complete loss of the steam generator heat sink.

  12. Investigation of vessel exterior air cooling for an HLMC reactor

    SciTech Connect (OSTI)

    Sienicki, J.J.; Spencer, B.W.

    2000-07-01T23:59:59.000Z

    The secure transportable autonomous reactor (STAR) concept under development at Argonne National Laboratory provides a small [300-MW(thermal)] reactor module for steam supply that incorporates design features to attain proliferation resistance, heightened passive safety, and improved cost competitiveness through extreme simplification. Examples are the achievement of 100% + natural-circulation heat removal from the low-power-density/low-pressure-drop ultralong lifetime core and utilization of lead-bismuth eutectic (LBE) coolant enabling elimination of main coolant pumps as well as the need for an intermediate heat transport circuit. It is required to provide a passive means of removing decay heat and effecting reactor cooldown in the event that the normal steam generator heat sink, including its normal shutdown heat removal mode, is postulated to be unavailable. In the present approach, denoted as the reactor exterior cooling system (RECS), passive decay heat removal is provided by cooling the outside of the containment/guard vessel with air. RECS is similar to the reactor vessel auxiliary cooling system (RVACS) incorporated into the PRISM design. However, to enhance the heat removal, RECS incorporates fins on the containment vessel exterior to enhance heat transfer to air as well as removable steel venetian conductors that provide a conduction heat transfer path across the reactor vessel-containment vessel gap to enhance heat transfer between the vessels. The objective of the present work is to investigate the effectiveness of air cooling in removing heat from the vessel and limiting the coolant temperature increase following a sudden complete loss of the steam generator heat sink.

  13. Design Considerations For Blast Loads In Pressure Vessels.

    SciTech Connect (OSTI)

    Rodriguez, E. A. (Edward A.); Nickell, Robert E.; Pepin, J. E. (Jason E.)

    2007-01-01T23:59:59.000Z

    Los Alamos National Laboratory (LANL), under the auspices of the U.S. Department of Energy (DOE) and the National Nuclear Security Administration (NNSA), conducts confined detonation experiments utilizing large, spherical, steel pressure vessels to contain the reaction products and hazardous materials from high-explosive (HE) events. Structural design and analysis considerations include: (a) Blast loading phase (i.e., impulsive loading); (b) Dynamic structural response; (c) Fragment (i.e., shrapnel) generation and penetration; (d) Ductile and non-ductile fracture; and (e) Design Criteria to ASME Code Sec. VIII, Div. 3, Impulsively Loaded Vessels. These vessels are designed for one-time-use only, efficiently utilizing the significant plastic energy absorption capability of ductile vessel materials. Alternatively, vessels may be designed for multiple-detonation events, in which case the material response is restricted to elastic or near-elastic range. Code of Federal Regulations, Title 10 Part 50 provides requirements for commercial nuclear reactor licensing; specifically dealing with accidental combustible gases in containment structures that might cause extreme loadings. The design philosophy contained herein may be applied to extreme loading events postulated to occur in nuclear reactor and non-nuclear systems or containments.

  14. Sterilization of fermentation vessels by ethanol/water mixtures

    DOE Patents [OSTI]

    Wyman, C.E.

    1999-02-09T23:59:59.000Z

    A method is described for sterilizing process fermentation vessels with a concentrated alcohol and water mixture integrated in a fuel alcohol or other alcohol production facility. Hot, concentrated alcohol is drawn from a distillation or other purification stage and sprayed into the empty fermentation vessels. This sterilizing alcohol/water mixture should be of a sufficient concentration, preferably higher than 12% alcohol by volume, to be toxic to undesirable microorganisms. Following sterilization, this sterilizing alcohol/water mixture can be recovered back into the same distillation or other purification stage from which it was withdrawn. The process of this invention has its best application in, but is not limited to, batch fermentation processes, wherein the fermentation vessels must be emptied, cleaned, and sterilized following completion of each batch fermentation process. 2 figs.

  15. Sterilization of fermentation vessels by ethanol/water mixtures

    DOE Patents [OSTI]

    Wyman, Charles E. (Lakewood, CO)

    1999-02-09T23:59:59.000Z

    A method for sterilizing process fermentation vessels with a concentrated alcohol and water mixture integrated in a fuel alcohol or other alcohol production facility. Hot, concentrated alcohol is drawn from a distillation or other purification stage and sprayed into the empty fermentation vessels. This sterilizing alcohol/water mixture should be of a sufficient concentration, preferably higher than 12% alcohol by volume, to be toxic to undesirable microorganisms. Following sterilization, this sterilizing alcohol/water mixture can be recovered back into the same distillation or other purification stage from which it was withdrawn. The process of this invention has its best application in, but is not limited to, batch fermentation processes, wherein the fermentation vessels must be emptied, cleaned, and sterilized following completion of each batch fermentation process.

  16. Report of the terawatt laser pressure vessel committee

    SciTech Connect (OSTI)

    Woodle, M.H.; Beauman, R.; Czajkowski, C.; Dickinson, T.; Lynch, D.; Pogorelsky, I.; Skjaritka, J.

    2000-09-25T23:59:59.000Z

    In 1995 the ATF project sent out an RFP for a CO2 Laser System having a TeraWatt output. Eight foreign and US firms responded. The Proposal Evaluation Panel on the second round selected Optoel, a Russian firm based in St. Petersburg, on the basis of the technical criteria and cost. Prior to the award, BNL representatives including the principal scientist, cognizant engineer and a QA representative visited the Optoel facilities to assess the company's capability to do the job. The contract required Optoel to provide a x-ray preionized high pressure amplifier that included: a high pressure cell, x-ray tube, internal optics and a HV pulse forming network for the main discharge and preionizer. The high-pressure cell consists of a stainless steel pressure vessel with various ports and windows that is filled with a gas mixture operating at 10 atmospheres. In accordance with BNL Standard ESH 1.4.1 ''Pressurized Systems For Experimental Use'', the pressure vessel design criteria is required to comply with the ASME Boiler and Pressure Vessel Code In 1996 a Preliminary Design Review was held at BNL. The vendor was requested to furnish drawings so that we could confirm that the design met the above criteria. The vendor furnished drawings did not have all dimensions necessary to completely analyze the cell. Never the less, we performed an analysis on as much of the vessel as we could with the available information. The calculations concluded that there were twelve areas of concern that had to be addressed to assure that the pressure vessel complied with the requirements of the ASME code. This information was forwarded to the vendor with the understanding that they would resolve these concerns as they continued with the vessel design and fabrication. The assembled amplifier pressure vessel was later hydro tested to 220 psi (15 Atm) as well as pneumatically to 181 psi (12.5 Atm) at the fabricator's Russian facility and was witnessed by a BNL engineer. The unit was shipped to the US and installed at the ATF. As part of the commissioning of the device the amplifier pressure vessel was disassembled several times at which time it became apparent that the vendor had not addressed 7 of the 12 issues previously identified. Closer examination of the vessel revealed some additional concerns including quality of workmanship. Although not required by the contract, the vendor furnished radiographs of a number of pressure vessel welds. A review of the Russian X-rays revealed radiographs of both poor and unreadable quality. However, a number of internal weld imperfections could be observed. All welds in question were excavated and then visually and dye penetrant inspected. These additional inspections confirmed that the weld techniques used to make some of these original welds were substandard. The applicable BNL standard, ESH 1.4.1, addresses the problem of pressure vessel non-compliance by having a committee appointed by the Department Chairman review the design and provide engineering solutions to assure equivalent safety. On January 24, 2000 Dr. M. Hart, the NSLS Chairman, appointed this committee with this charge. This report details the engineering investigations, deliberations, solutions and calculations which were developed by members of this committee to determine that with repairs, new components, appropriate NDE, and lowering the design pressure, the vessel can be considered safe to use.

  17. Retrospective dosimetry analyses of reactor vessel cladding samples

    SciTech Connect (OSTI)

    Greenwood, L. R.; Soderquist, C. Z. [Battelle Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Fero, A. H. [Westinghouse Electric Company, Cranberry Twp., PA 16066 (United States)

    2011-07-01T23:59:59.000Z

    Reactor pressure vessel cladding samples for Ringhals Units 3 and 4 in Sweden were analyzed using retrospective reactor dosimetry techniques. The objective was to provide the best estimates of the neutron fluence for comparison with neutron transport calculations. A total of 51 stainless steel samples consisting of chips weighing approximately 100 to 200 mg were removed from selected locations around the pressure vessel and were sent to Pacific Northwest National Laboratory for analysis. The samples were fully characterized and analyzed for radioactive isotopes, with special interest in the presence of Nb-93m. The RPV cladding retrospective dosimetry results will be combined with a re-evaluation of the surveillance capsule dosimetry and with ex-vessel neutron dosimetry results to form a comprehensive 3D comparison of measurements to calculations performed with 3D deterministic transport code. (authors)

  18. Closure Report for Corrective Action Unit 539: Areas 25 and 26 Railroad Tracks Nevada National Security Site, Nevada with ROTC-1, Revision 0

    SciTech Connect (OSTI)

    Mark Kauss

    2011-06-01T23:59:59.000Z

    This Closure Report (CR) presents information supporting the closure of Corrective Action Unit (CAU) 539: Areas 25 and 26 Railroad Tracks, Nevada National Security Site, Nevada. This CR complies with the requirements of the Federal Facility Agreement and Consent Order (FFACO) that was agreed to by the State of Nevada; U.S. Department of Energy (DOE), Environmental Management; U.S. Department of Defense; and DOE, Legacy Management. The corrective action sites (CASs) within CAU 539 are located within Areas 25 and 26 of the Nevada National Security Site. Corrective Action Unit 539 comprises the following CASs: • 25-99-21, Area 25 Railroad Tracks • 26-99-05, Area 26 Railroad Tracks The purpose of this CR is to provide documentation supporting the completed corrective actions and provide data confirming that the closure objectives for CASs within CAU 539 were met. To achieve this, the following actions were performed: • Reviewed documentation on historical and current site conditions, including the concentration and extent of contamination. • Conducted radiological walkover surveys of railroad tracks in both Areas 25 and 26. • Collected ballast and soil samples and calculated internal dose estimates for radiological releases. • Collected in situ thermoluminescent dosimeter measurements and calculated external dose estimates for radiological releases. • Removed lead bricks as potential source material (PSM) and collected verification samples. • Implemented corrective actions as necessary to protect human health and the environment. • Properly disposed of corrective action and investigation wastes. • Implemented an FFACO use restriction (UR) for radiological contamination at CAS 25-99-21. The approved UR form and map are provided in Appendix F and will be filed in the DOE, National Nuclear Security Administration Nevada Site Office (NNSA/NSO), Facility Information Management System; the FFACO database; and the NNSA/NSO CAU/CAS files. From November 29, 2010, through May 2, 2011, closure activities were performed as set forth in the Streamlined Approach for Environmental Restoration (SAFER) Plan for Corrective Action Unit 539: Areas 25 and 26 Railroad Tracks, Nevada Test Site, Nevada. The purposes of the activities as defined during the data quality objectives process were as follows: • Determine whether contaminants of concern (COCs) are present. • If COCs are present, determine their nature and extent, implement appropriate corrective actions, and properly dispose of wastes. Analytes detected during the closure activities were evaluated against final action levels (FALs) to determine COCs for CAU 539. Assessment of the data generated from closure activities revealed the following: • At CAS 26-99-05, the total effective dose for radiological releases did not exceed the FAL of 25 millirem per Industrial Area year. Potential source material in the form of lead bricks was found at three locations. A corrective action of clean closure was implemented at these locations, and verification samples indicated that no further action is necessary. • At CAS 25-99-21, the total effective dose for radiological releases exceeds the FAL of 25 millirem per Industrial Area year. Potential source material in the form of lead bricks was found at eight locations. A corrective action was implemented by removing the lead bricks and soil above FALs at these locations, and verification samples indicated that no further action is necessary. Pieces of debris with high radioactivity were identified as PSM and remain within the CAS boundary. A corrective action of closure in place with a UR was implemented at this CAS because closure activities showed evidence of remaining soil contamination and radioactive PSM. Future land use will be restricted from surface and intrusive activities. Closure activities generated waste streams consisting of industrial solid waste, recyclable materials, low-level radioactive waste, and mixed low-level radioactive waste. Wastes were disposed of in the appropriate onsite landfills. The NNSA/NSO prov

  19. Ex-vessel demand by size for the Gulf shrimp

    E-Print Network [OSTI]

    Chui, Margaret Kam-Too

    1980-01-01T23:59:59.000Z

    EX-VESSEL DEMAND BY SIZE FOR THE GULF SHRIMP A Thesis by MARGARET RAM-TOO CHUI Submitted to the Graduate College of Texas A&M University in partial fulfillment of the requirement for the degree of MASTER OF SCIENCE August 1980 Major... Subject: Agricultural Economics EX-VESSEL DEMAND BY SIZE FOR SHRIMP IN THE GULF OF MEXICO A Thesis by MARGARET KAM-TOO CHUI Approved as to style and content by: ai an of Committee) (Hea f ep tment) (Member) (Member) August 1980 ABSTRACT Ex...

  20. Simulation of Diffusive Lithium Evaporation Onto the NSTX Vessel Walls

    SciTech Connect (OSTI)

    Stotler, D. P.; Skinner, C. H.; Blanchard, W. R.; Krstic, P. S.; Kugel, H. W.; Schneider, H.; Zakharov, L. E.

    2010-12-09T23:59:59.000Z

    A model for simulating the diffusive evaporation of lithium into a helium filled NSTX vacuum vessel is described and validated against an initial set of deposition experiments. The DEGAS 2 based model consists of a three-dimensional representation of the vacuum vessel, the elastic scattering process, and a kinetic description of the evaporated atoms. Additional assumptions are required to account for deuterium out-gassing during the validation experiments. The model agrees with the data over a range of pressures to within the estimated uncertainties. Suggestions are made for more discriminating experiments that will lead to an improved model.

  1. MELCOR ex-vessel LOCA simulations for ITER{sup +}

    SciTech Connect (OSTI)

    Gaeta, M.J.; Merrill, B.J. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Bartels, H.W. [ITER San Diego Joint Work Site, La Jolla, CA (United States)] [and others

    1995-11-01T23:59:59.000Z

    Ex-vessel Loss-of-Coolant-Accident (LOCA) simulations for the International Thermonuclear Experimental Reactor (ITER) were performed using the MELCOR code. The main goals of this work were to estimate the ultimate pressurization of the heat transport system (HTS) vault in order to gauge the potential for stack releases and to estimate the total amount of hydrogen generated during a design basis ex-vessel LOCA. Simulation results indicated that the amount of hydrogen produced in each transient was below the flammability limit for the plasma chamber. In addition, only moderate pressurization of the HTS vault indicated a very small potential for releases through the stack.

  2. Lightweight cryogenic-compatible pressure vessels for vehicular fuel storage

    DOE Patents [OSTI]

    Aceves, Salvador; Berry, Gene; Weisberg, Andrew H.

    2004-03-23T23:59:59.000Z

    A lightweight, cryogenic-compatible pressure vessel for flexibly storing cryogenic liquid fuels or compressed gas fuels at cryogenic or ambient temperatures. The pressure vessel has an inner pressure container enclosing a fuel storage volume, an outer container surrounding the inner pressure container to form an evacuated space therebetween, and a thermal insulator surrounding the inner pressure container in the evacuated space to inhibit heat transfer. Additionally, vacuum loss from fuel permeation is substantially inhibited in the evacuated space by, for example, lining the container liner with a layer of fuel-impermeable material, capturing the permeated fuel in the evacuated space, or purging the permeated fuel from the evacuated space.

  3. Ion transport membrane module and vessel system with directed internal gas flow

    DOE Patents [OSTI]

    Holmes, Michael Jerome (Thompson, ND); Ohrn, Theodore R. (Alliance, OH); Chen, Christopher Ming-Poh (Allentown, PA)

    2010-02-09T23:59:59.000Z

    An ion transport membrane system comprising (a) a pressure vessel having an interior, an inlet adapted to introduce gas into the interior of the vessel, an outlet adapted to withdraw gas from the interior of the vessel, and an axis; (b) a plurality of planar ion transport membrane modules disposed in the interior of the pressure vessel and arranged in series, each membrane module comprising mixed metal oxide ceramic material and having an interior region and an exterior region; and (c) one or more gas flow control partitions disposed in the interior of the pressure vessel and adapted to change a direction of gas flow within the vessel.

  4. Federal Railroad Administration

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

    accident. FRA is forgoing an official position on the use of dedicated trains pending the process of evaluating the results of the study and the required costbenefit analysis in...

  5. Federal Railroad Administration

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

    very similar to this bill in regard to the core provisions Core provisions: Review and reform the Federal hours of service requirements Establish a new risk reduction program...

  6. Inexpensive delivery of compressed hydrogen with advanced vessel technology

    E-Print Network [OSTI]

    of flexible refueling (compressed/cryogenic H2/(L)H2) #12;The PVT properties of H2 drive storage and delivery) Explore station demand from 70 kg H2/day to 1000 kg H2/day · Real hydrogen thermodynamic and PVT diagram and vessel characteristics to minimize delivery cost · Hydrogen and material properties Increased

  7. Response of a vessel to waves at zero ship speed

    E-Print Network [OSTI]

    Response of a vessel to waves at zero ship speed: preliminary full scale experiments By: Kim Klaka of experiment were conducted ­ free roll decay tests and irregular wave tests. An inclining test was also with and without the mainsail hoisted, in very light winds. The irregular wave tests were conducted again in very

  8. International Hydrogen Fuel and Pressure Vessel Forum 2010 Beijing, China

    E-Print Network [OSTI]

    challenges in harmonizing test protocols and requirements for compressed natural gas (CNG), hydrogen, and CNGInternational Hydrogen Fuel and Pressure Vessel Forum 2010 Beijing, China September 27-29, 2010 Background The China Association for Hydrogen Energy, the Engineering Research Center of High Pressure

  9. Modelling the Induced Magnetic Signature of Naval Vessels

    E-Print Network [OSTI]

    Low, Robert

    vessels stealth is an important design feature. With recent advances in electromagnetic sensor technology with the magnetic signature resulting from the magnetisation of the ferromagnetic material of the ship, under is constructed from non-magnetic materials, but arises from the combined e#11;ect of the individual items

  10. Large Blood Vessels 1.1 Introduction --The Cardiovascular System

    E-Print Network [OSTI]

    Luo, Xiaoyu

    Chapter 1 Large Blood Vessels 1.1 Introduction -- The Cardiovascular System The heart is a pump that circulates blood to the lungs for oxygenation (pul- monary circulation) and then throughout the systemic arterial system with a total cycle time of about one minute. From the left ventricle of the heart, blood

  11. PublicationsmailagreementNo.40014024 the VeSSeL WILL

    E-Print Network [OSTI]

    Pedersen, Tom

    fuel. The hybrid system will provide energy for low-speed maneuvering and stationPublicationsmailagreementNo.40014024 THE 1st the VeSSeL WILL Be the WORLD'S FIRSt PLUG-IN hYBRID's first plug-in hybrid "green ship" powered by electricity, hydrogen fuel cells and low- emission diesel

  12. Vessel Cold-Ironing Using a Barge Mounted PEM Fuel Cell: Project...

    Energy Savers [EERE]

    Vessel Cold-Ironing Using a Barge Mounted PEM Fuel Cell: Project Scoping and Feasibility Vessel Cold-Ironing Using a Barge Mounted PEM Fuel Cell: Project Scoping and Feasibility...

  13. Pulmonary Hypertension and Computed Tomography Measurement of Small Pulmonary Vessels in

    E-Print Network [OSTI]

    Pulmonary Hypertension and Computed Tomography Measurement of Small Pulmonary Vessels in Severe alteration of small pulmonary vessels is one of the characteristic features of pulmonary hypertension in chronic obstruc- tive pulmonary disease. The in vivo relationship between pulmonary hypertension

  14. E-Print Network 3.0 - axicell vacuum vessel Sample Search Results

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

    T. Brown, H-M Fan, G. Jones... 12;13 July 2002 Snowmass Review: FIRE Vacuum Vessel and Remote Handling 2 Presentation Outline... Vacuum Vessel - Design requirements - Design...

  15. LOW ALLOY STEELS FOR THICK WALL PRESSURE VESSELS Yearly Report for Period Oct. 1, 1976 to Sept. 30, 1977.

    E-Print Network [OSTI]

    Horn, R.M.

    2011-01-01T23:59:59.000Z

    Vessel Fabrication Under ASME Code Current Pressure Vessel Sc a t i o n under the ASME code current s t e e l s , and (VESSEL FABRICATION UNDER ASME CODE Interactions with Babcock

  16. A Survey of Pressure Vessel Code Compliance for Superconducting RF Cryomodules

    SciTech Connect (OSTI)

    Peterson, Thomas; Klebaner, Arkadiy; Nicol, Tom; Theilacker, Jay; /Fermilab; Hayano, Hitoshi; Kako, Eiji; Nakai, Hirotaka; Yamamoto, Akira; /KEK, Tsukuba; Jensch, Kay; Matheisen, Axel; /DESY; Mammosser, John; /Jefferson Lab

    2011-06-07T23:59:59.000Z

    Superconducting radio frequency (SRF) cavities made from niobium and cooled with liquid helium are becoming key components of many particle accelerators. The helium vessels surrounding the RF cavities, portions of the niobium cavities themselves, and also possibly the vacuum vessels containing these assemblies, generally fall under the scope of local and national pressure vessel codes. In the U.S., Department of Energy rules require national laboratories to follow national consensus pressure vessel standards or to show ''a level of safety greater than or equal to'' that of the applicable standard. Thus, while used for its superconducting properties, niobium ends up being treated as a low-temperature pressure vessel material. Niobium material is not a code listed material and therefore requires the designer to understand the mechanical properties for material used in each pressure vessel fabrication; compliance with pressure vessel codes therefore becomes a problem. This report summarizes the approaches that various institutions have taken in order to bring superconducting RF cryomodules into compliance with pressure vessel codes. In Japan, Germany, and the U.S., institutions building superconducting RF cavities integrated in helium vessels or procuring them from vendors have had to deal with pressure vessel requirements being applied to SRF vessels, including the niobium and niobium-titanium components of the vessels. While niobium is not an approved pressure vessel material, data from tests of material samples provide information to set allowable stresses. By means of procedures which include adherence to code welding procedures, maintaining material and fabrication records, and detailed analyses of peak stresses in the vessels, or treatment of the vacuum vessel as the pressure boundary, research laboratories around the world have found methods to demonstrate and document a level of safety equivalent to the applicable pressure vessel codes.

  17. RIS-M-2186 INTERPRETATIOM OF STRAIN HBASUREMEMTS ON NUCLEAR PRESSURE VESSELS

    E-Print Network [OSTI]

    RISÃ?-M- 2186 INTERPRETATIOM OF STRAIN HBASUREMEMTS ON NUCLEAR PRESSURE VESSELS Svend Ib Andersen Preben Engbzk Abstract. Selected results from strain measurements on 4 nuclear pressure vessels, EXPERIMENTAL DATA, GRAPHS, MECHANICAL TESTS, PERFORMANCE TESTING, PRESSURE VESSELS, tMR TYPE REACTORS, STEELS

  18. NEUTRON DAMAGE IN REACTOR PRESSURE-VESSEL STEEL EXAMINED WITH POSITRON ANNIHILATION LIFETIME SPECTROSCOPY

    E-Print Network [OSTI]

    Motta, Arthur T.

    NEUTRON DAMAGE IN REACTOR PRESSURE-VESSEL STEEL EXAMINED WITH POSITRON ANNIHILATION LIFETIME-vessel steels. We irradiated samples ofASTM A508 nuclear reactor pressure-vessel steel to fast neutron 17 2 (PALS) to study the effects of neutron damage in the steels on positron lifetimes. Non

  19. Photoacoustic spectroscopy sample array vessels and photoacoustic spectroscopy methods for using the same

    DOE Patents [OSTI]

    Amonette, James E.; Autrey, S. Thomas; Foster-Mills, Nancy S.

    2006-02-14T23:59:59.000Z

    Methods and apparatus for simultaneous or sequential, rapid analysis of multiple samples by photoacoustic spectroscopy are disclosed. Particularly, a photoacoustic spectroscopy sample array vessel including a vessel body having multiple sample cells connected thereto is disclosed. At least one acoustic detector is acoustically positioned near the sample cells. Methods for analyzing the multiple samples in the sample array vessels using photoacoustic spectroscopy are provided.

  20. PRESSURIZATION OF CONTAINMENT VESSELS FROM PLUTONIUM OXIDE CONTENTS

    SciTech Connect (OSTI)

    Hensel, S.

    2012-03-27T23:59:59.000Z

    Transportation and storage of plutonium oxide is typically done using a convenience container to hold the oxide powder which is then placed inside a containment vessel. Intermediate containers which act as uncredited confinement barriers may also be used. The containment vessel is subject to an internal pressure due to several sources including; (1) plutonium oxide provides a heat source which raises the temperature of the gas space, (2) helium generation due to alpha decay of the plutonium, (3) hydrogen generation due to radiolysis of the water which has been adsorbed onto the plutonium oxide, and (4) degradation of plastic bags which may be used to bag out the convenience can from a glove box. The contributions of these sources are evaluated in a reasonably conservative manner.

  1. Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database

    SciTech Connect (OSTI)

    Wang, Jy-An John [ORNL

    2010-08-01T23:59:59.000Z

    Materials behaviors caused by neutron irradiation under fission and/or fusion environments can be little understood without practical examination. Easily accessible material information system with large material database using effective computers is necessary for design of nuclear materials and analyses or simulations of the phenomena. The developed Embrittlement Data Base (EDB) at ORNL is this comprehensive collection of data. EDB database contains power reactor pressure vessel surveillance data, the material test reactor data, foreign reactor data (through bilateral agreements authorized by NRC), and the fracture toughness data. The lessons learned from building EDB program and the associated database management activity regarding Material Database Design Methodology, Architecture and the Embedded QA Protocol are described in this report. The development of IAEA International Database on Reactor Pressure Vessel Materials (IDRPVM) and the comparison of EDB database and IAEA IDRPVM database are provided in the report. The recommended database QA protocol and database infrastructure are also stated in the report.

  2. Superheat effects on localized vessel breach enlargement during corium ejection

    SciTech Connect (OSTI)

    Sienicki, J.J.; Spencer, B.W.

    1986-01-01T23:59:59.000Z

    The evaluation of the consequences of hypothetical severe accident sequences in light water reactors includes those sequences in which molten corium is postulated to melt through the reactor pressure vessel (RPV) lower head and enter the region beneath the RPV. An important issue is the mode by which the lower head is breached and molten corium introduced into the reactor cavity (PWR) or pedestal (BWR). Reported here are the results of an investigation into the dependency of ablation-induced enlargement on the initial corium temperature, or more specifically, the initial corium superheat (i.e., excess temperature above the freezing temperature). A model is introduced here to predict the vessel erosion and is employed to scope the effects of variations in the superheat.

  3. IMPACT OF NUCLEAR MATERIAL DISSOLUTION ON VESSEL CORROSION

    SciTech Connect (OSTI)

    Mickalonis, J.; Dunn, K.; Clifton, B.

    2012-10-01T23:59:59.000Z

    Different nuclear materials require different processing conditions. In order to maximize the dissolver vessel lifetime, corrosion testing was conducted for a range of chemistries and temperature used in fuel dissolution. Compositional ranges of elements regularly in the dissolver were evaluated for corrosion of 304L, the material of construction. Corrosion rates of AISI Type 304 stainless steel coupons, both welded and non-welded coupons, were calculated from measured weight losses and post-test concentrations of soluble Fe, Cr and Ni.

  4. Dynamic response of containment vessels to blast loading

    SciTech Connect (OSTI)

    Karpp, R.R.; Duffey, T.A.; Neal, T.R.; Warnes, R.H.; Thompson, J.D.

    1982-01-01T23:59:59.000Z

    The dynamic response of steel, spherical containment vessels loaded by internal explosive blast was studied by experiments, computations, and analysis. Instrumentation used in the experiments consisted of strain and pressure gauges and a velocity interferometer. Data were used to rank the blast wave mitigating properties of several filler materials and to develop a scaling law relating strain, filler material, and explosive energy or explosive mass.

  5. PARTICLE TRANSPORTATION AND DEPOSITION IN HOT GAS FILTER VESSELS - A COMPUTATIONAL AND EXPERIMENTAL MODELING APPROACH

    SciTech Connect (OSTI)

    Goodarz Ahmadi

    2002-07-01T23:59:59.000Z

    In this project, a computational modeling approach for analyzing flow and ash transport and deposition in filter vessels was developed. An Eulerian-Lagrangian formulation for studying hot-gas filtration process was established. The approach uses an Eulerian analysis of gas flows in the filter vessel, and makes use of the Lagrangian trajectory analysis for the particle transport and deposition. Particular attention was given to the Siemens-Westinghouse filter vessel at Power System Development Facility in Wilsonville in Alabama. Details of hot-gas flow in this tangential flow filter vessel are evaluated. The simulation results show that the rapidly rotation flow in the spacing between the shroud and the vessel refractory acts as cyclone that leads to the removal of a large fraction of the larger particles from the gas stream. Several alternate designs for the filter vessel are considered. These include a vessel with a short shroud, a filter vessel with no shroud and a vessel with a deflector plate. The hot-gas flow and particle transport and deposition in various vessels are evaluated. The deposition patterns in various vessels are compared. It is shown that certain filter vessel designs allow for the large particles to remain suspended in the gas stream and to deposit on the filters. The presence of the larger particles in the filter cake leads to lower mechanical strength thus allowing for the back-pulse process to more easily remove the filter cake. A laboratory-scale filter vessel for testing the cold flow condition was designed and fabricated. A laser-based flow visualization technique is used and the gas flow condition in the laboratory-scale vessel was experimental studied. A computer model for the experimental vessel was also developed and the gas flow and particle transport patterns are evaluated.

  6. Fracture toughness test results of thermal aged reactor vessel materials

    SciTech Connect (OSTI)

    DeVan, M.J.; Lowe, A.L. Jr. [B and W Nuclear Technologies Inc., Lynchburg, VA (United States). Nuclear Engineering Dept.; Hall, J.B. [Babcock and Wilcox Co., Alliance, OH (United States)

    1996-12-31T23:59:59.000Z

    Thermal-aged surveillance materials consisting of Sa-533, Grade B, Class 1 plate material; SA-508, Class 2 forging material; and 2 Mn-Mo-Ni/Linde 80 weld metals were removed from two commercial reactor pressure vessels. The material from the first reactor vessel received a thermal exposure of approximately 103,000 hours at 282 C, while the material from the second reactor vessel received a thermal exposure of approximately 93,000 hours at 282 C. Tensile and 1/2 T compact fracture toughness specimens were fabricated from these materials and tested. In addition, to examine the effects of annealing, selected thermal-aged and unaged specimens were annealed at 454 C (850 F) and tested. Varying responses in the fracture toughness properties were observed for all materials after exposure to the thermal-aging temperature. The base metal plate had an observed decrease in J-values after its respective aging exposure, while no significant difference in the J-values were observed for the Linde 80 weld metals. No significant difference was seen in the J-data for the aged/annealed materials, but because of the small number of test specimens available, no conclusion could be determined for the response to annealing.

  7. Start-up control system and vessel for lmfbr

    SciTech Connect (OSTI)

    Durrant, O.W.; Kakarala, C.R.; Mandel, S.W.

    1987-04-07T23:59:59.000Z

    This patent describes a start-up vessel for a start-up system suitable for liquid metal fast breeder reactors comprising: a lower bulb defining a lower space; an upper bulb defining an upper space; a mid-section of a cross-sectional diameter less than that of the lower and upper bulbs, defining a mid-space and connected between the upper and lower bulbs; heating means associated with the lower bulb for heating water in the lower space; at least one inlet conduit connection connected to the vessel for admitting feed water to the lower space to be heated by the heating means to produce steam; at least one outlet conduit connection connected to the vessel for discharging steam; an auxiliary feed water line connected through the upper bulb having at least one nozzle at the end thereof for spraying feed water into the upper space; and a main steam inlet connection connected to the upper bulb for heating the auxiliary feed water to produce steam.

  8. Method and apparatus for detecting irregularities on or in the wall of a vessel

    DOE Patents [OSTI]

    Bowling, Michael Keith (Blackborough Cullompton, GB)

    2000-09-12T23:59:59.000Z

    A method of detecting irregularities on or in the wall of a vessel by detecting localized spatial temperature differentials on the wall surface, comprising scanning the vessel surface with a thermal imaging camera and recording the position of the or each region for which the thermal image from the camera is indicative of such a temperature differential across the region. The spatial temperature differential may be formed by bacterial growth on the vessel surface; alternatively, it may be the result of defects in the vessel wall such as thin regions or pin holes or cracks. The detection of leaks through the vessel wall may be enhanced by applying a pressure differential or a temperature differential across the vessel wall; the testing for leaks may be performed with the vessel full or empty, and from the inside or the outside.

  9. Dual shell pressure balanced reactor vessel. Final project report

    SciTech Connect (OSTI)

    Robertus, R.J.; Fassbender, A.G.

    1994-10-01T23:59:59.000Z

    The Department of Energy`s Office of Energy Research (OER) has previously provided support for the development of several chemical processes, including supercritical water oxidation, liquefaction, and aqueous hazardous waste destruction, where chemical and phase transformations are conducted at high pressure and temperature. These and many other commercial processes require a pressure vessel capable of operating in a corrosive environment where safety and economy are important requirements. Pacific Northwest Laboratory (PNL) engineers have recently developed and patented (U.S. patent 5,167,930 December 1, 1992) a concept for a novel Dual Shell Pressure Balanced Vessel (DSPBV) which could solve a number of these problems. The technology could be immediately useful in continuing commercialization of an R&D 100 award-winning technology, Sludge-to-oil Reactor System (STORS), originally developed through funding by OER. Innotek Corporation is a small business that would be one logical end-user of the DSPBV reactor technology. Innotek is working with several major U.S. engineering firms to evaluate the potential of this technology in the disposal of wastes from sewage treatment plants. PNL entered into a CRADA with Innotek to build a bench-scale demonstration reactor and test the system to advance the economic feasibility of a variety of high pressure chemical processes. Hydrothermal processing of corrosive substances on a large scale can now be made significantly safer and more economical through use of the DSPBV. Hydrothermal chemical reactions such as wet-air oxidation and supercritical water oxidation occur in a highly corrosive environment inside a pressure vessel. Average corrosion rates from 23 to 80 miles per year have been reported by Rice (1994) and Latanision (1993).

  10. Vessel eddy current measurement for the National Spherical Torus Experiment

    SciTech Connect (OSTI)

    Gates, D.A.; Menard, J.E.; Marsala, R.J. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, New Jersey 08543 (United States)

    2004-12-01T23:59:59.000Z

    A simple analog circuit that measures the National Spherical Torus Experiment (NSTX) axisymmetric eddy current distribution has been designed and constructed. It is based on simple circuit model of the NSTX vacuum vessel that was calibrated using a special axisymmetric eddy current code which was written so that accuracy was maintained in the vicinity of the current filaments [J. Menard, J. Fusion Tech. (to be published)]. The measurement and the model have been benchmarked against data from numerous vacuum shots and they are in excellent agreement. This is an important measurement that helps give more accurate equilibrium reconstructions.

  11. Heat-transfer coefficients in agitated vessels. Latent heat models

    SciTech Connect (OSTI)

    Kumpinsky, E. [Ashland Chemical Co., Columbus, OH (United States)] [Ashland Chemical Co., Columbus, OH (United States)

    1996-03-01T23:59:59.000Z

    Latent heat models were developed to calculate heat-transfer coefficients in agitated vessels for two cases: (1) heating with a condensable fluid flowing through coils and jackets; (2) vacuum reflux cooling with an overhead condenser. In either case the mathematical treatment, based on macroscopic balances, requires no iterative schemes. In addition to providing heat-transfer coefficients, the models predict flow rates of service fluid through the coils and jackets, estimate the percentage of heat transfer due to latent heat, and compute reflux rates.

  12. Facing the transition. [Retrofitting vessels for burning coal

    SciTech Connect (OSTI)

    Not Available

    1981-08-01T23:59:59.000Z

    Historically, environmental regulations prohibiting black smoke in port and marine disposal of ashes caused many coal-burning vessels in the Great Lakes shipping industry to convert to oil-burning systems. With a return to coal-burning plants on-board, these problems and others are being addressed. Improvements are being made in stack emission control. The need for monitoring devices is discussed. Mechanisms are described which will help control dust, heat, noise and ash. To reduce the need for excessive stockpiles of various grades of coal, equipment is being designed which will burn a range of coals available in many ports. (JMT)

  13. Gamma ray-induced embrittlement of pressure vessel alloys

    SciTech Connect (OSTI)

    Alexander, D.E.; Rehn, L.E. [Argonne National Lab., IL (United States); Farrell, K.; Stoller, R.E. [Oak Ridge National Lab., TN (United States)

    1994-11-01T23:59:59.000Z

    High-energy gamma rays emitted from the core of a nuclear reactor produce displacement damage in the reactor pressure vessel (RPV). The contribution of gamma damage to RPV embrittlement has in the past been largely ignored. However, in certain reactor designs the gamma flux at the RPV is sufficiently large that its contribution to displacement damage can be substantial. For example, gamma rays have been implicated in the accelerated RPV embrittlement observed in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. In the present study, mechanical property changes induced by 10-MeV electron irradiation of a model Fe alloy and an RPV alloy of interest to the HFIR were examined. Mini-tensile specimens were irradiated with high-energy electrons to reproduce damage characteristic of the Compton recoil-electrons induced by gamma bombardment. Substantial increases in yield and ultimate stress were observed in the alloys after irradiation to doses up to 5.3x10{sup {minus}3} dpa at temperatures ({approximately}50{degrees}C) characteristic of the HFIR pressure vessel. These measured increases were similar to those previously obtained following neutron irradiation, despite the highly disparate nature of the damage generated during electron and neutron irradiation.

  14. Impacts of reducing shipboard NOx? and SOx? emissions on vessel performance

    E-Print Network [OSTI]

    Caputo, Ronald J., Jr. (Ronald Joseph)

    2010-01-01T23:59:59.000Z

    The international maritime community has been experiencing tremendous pressures from environmental organizations to reduce the emissions footprint of their vessels. In the last decade, air emissions, including nitrogen ...

  15. a533b pressure vessel: Topics by E-print Network

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

    Program & Sensors Hydrogen Delivery Composite Overwrapped Pressure Vessels (COPVs) Pipeline for Off-Board Hydrogen: L-C Atmosphere: 1500 psi H2, ambient pressure Air...

  16. Nuclear reactor having a polyhedral primary shield and removable vessel insulation

    DOE Patents [OSTI]

    Ekeroth, D.E.; Orr, R.

    1993-12-07T23:59:59.000Z

    A nuclear reactor is provided having a generally cylindrical reactor vessel disposed within an opening in a primary shield. The opening in the primary shield is defined by a plurality of generally planar side walls forming a generally polyhedral-shaped opening. The reactor vessel is supported within the opening in the primary shield by reactor vessel supports which are in communication and aligned with central portions of some of the side walls. The reactor vessel is connected to the central portions of the reactor vessel supports. A thermal insulation polyhedron formed from a plurality of slidably insertable and removable generally planar insulation panels substantially surrounds at least a portion of the reactor vessel and is disposed between the reactor vessel and the side walls of the primary shield. The shape of the insulation polyhedron generally corresponds to the shape of the opening in the primary shield. Reactor monitoring instrumentation may be mounted in the corners of the opening in the primary shield between the side walls and the reactor vessel such that insulation is not disposed between the instrumentation and the reactor vessel. 5 figures.

  17. Nuclear reactor having a polyhedral primary shield and removable vessel insulation

    DOE Patents [OSTI]

    Ekeroth, Douglas E. (Delmont, PA); Orr, Richard (Pittsburgh, PA)

    1993-01-01T23:59:59.000Z

    A nuclear reactor is provided having a generally cylindrical reactor vessel disposed within an opening in a primary shield. The opening in the primary shield is defined by a plurality of generally planar side walls forming a generally polyhedral-shaped opening. The reactor vessel is supported within the opening in the primary shield by reactor vessel supports which are in communication and aligned with central portions of some of the side walls. The reactor vessel is connected to the central portions of the reactor vessel supports. A thermal insulation polyhedron formed from a plurality of slidably insertable and removable generally planar insulation panels substantially surrounds at least a portion of the reactor vessel and is disposed between the reactor vessel and the side walls of the primary shield. The shape of the insulation polyhedron generally corresponds to the shape of the opening in the primary shield. Reactor monitoring instrumentation may be mounted in the corners of the opening in the primary shield between the side walls and the reactor vessel such that insulation is not disposed between the instrumentation and the reactor vessel.

  18. R&D of Large Stationary Hydrogen/CNG/HCNG Storage Vessels

    Broader source: Energy.gov [DOE]

    These slides were presented at the International Hydrogen Fuel and Pressure Vessel Forum on September 27 – 29, 2010, in Beijing, China.

  19. Albatrosses Following Fishing Vessels: How Badly Hooked Are They on an Easy Meal?

    E-Print Network [OSTI]

    Resources, Fisheries Department, Stanley, Falkland Islands, 4 Eco-Ethology Research Unit, ISPA, Lisboa effort was spent near ships. Nevertheless, a few individuals repeatedly visited fishing vessels, which

  20. Evaluation of hydrogen pressure vessels using slow strain rate testing and fracture mechanics analysis

    SciTech Connect (OSTI)

    Murray, S.H. [National Aeronautics and Space Administration, Kennedy Space Center, FL (United States). Materials Science Div.; Desai, V.H. [Univ. of Central Florida, Orlando, FL (United States)

    1998-12-31T23:59:59.000Z

    A total of 108 seamless, forged pressure vessels, fabricated from ASTM A372 type IV (UNS K14508) and type V low alloy steel, are currently in 4,200 psi (29 MPa) gaseous hydrogen (GH{sub 2}) service at the Kennedy Space Center`s (KSC) Space Shuttle Launch Complex 39 (LC-39). The vessels were originally used in 6,000 psi (41 MPa) GH{sub 2} service during the Apollo program. NASA recently received a letter of warning from the manufacturer of the vessels stating that the subject vessels should be now be removed from GH{sub 2} service due to the fact that the ultimate tensile strength (UTS) of many of the vessels exceeds the maximum limit of 126 ksi (869 MPa) now imposed on A372 steel intended for GH{sub 2} service, and therefore are susceptible to hydrogen environment embrittlement. Due to the expense associated with vessel replacement, it was decided to determine by testing and analysis whether or not the vessels needed to be removed from GH{sub 2} service. Slow strain rate testing was performed under hydrogen charging conditions to determine the value of the threshold fracture toughness for sustained loading crack growth in GH{sub 2}, (K{sub H}) for the vessel material, this value was then used in a fracture mechanics safe-life analysis (a 20-year service life was modeled) that indicated the vessels are safe for continued use.

  1. X:\\L6046\\Data_Publication\\Pma\\current\\ventura\\pma.vp

    U.S. Energy Information Administration (EIA) Indexed Site

    No. 6 fuel oil includes Bunker C fuel oil and is used for the pro- duction of electric power, space heating, vessel bunk- ering, and various industrial purposes. Retailer: A...

  2. untitled

    U.S. Energy Information Administration (EIA) Indexed Site

    No. 6 fuel oil includes Bunker C fuel oil and is used for the pro- duction of electric power, space heating, vessel bunk- ering, and various industrial purposes. Retailer: A...

  3. The Disruption of Vessel-Spanning Bubbles with Sloped Fins in Flat-Bottom and 2:1 Elliptical-Bottom Vessels

    SciTech Connect (OSTI)

    Gauglitz, Phillip A.; Buchmiller, William C.; Jenks, Jeromy WJ; Chun, Jaehun; Russell, Renee L.; Schmidt, Andrew J.; Mastor, Michael M.

    2010-09-22T23:59:59.000Z

    Radioactive sludge was generated in the K-East Basin and K-West Basin fuel storage pools at the Hanford Site while irradiated uranium metal fuel elements from the N Reactor were being stored and packaged. The fuel has been removed from the K Basins, and currently, the sludge resides in the KW Basin in large underwater Engineered Containers. The first phase to the Sludge Treatment Project being led by CH2MHILL Plateau Remediation Company (CHPRC) is to retrieve and load the sludge into sludge transport and storage containers (STSCs) and transport the sludge to T Plant for interim storage. The STSCs will be stored inside T Plant cells that are equipped with secondary containment and leak-detection systems. The sludge is composed of a variety of particulate materials and water, including a fraction of reactive uranium metal particles that are a source of hydrogen gas. If a situation occurs where the reactive uranium metal particles settle out at the bottom of a container, previous studies have shown that a vessel-spanning gas layer above the uranium metal particles can develop and can push the overlying layer of sludge upward. The major concern, in addition to the general concern associated with the retention and release of a flammable gas such as hydrogen, is that if a vessel-spanning bubble (VSB) forms in an STSC, it may drive the overlying sludge material to the vents at the top of the container. Then it may be released from the container into the cell’s secondary containment system at T Plant. A previous study demonstrated that sloped walls on vessels, both cylindrical coned-shaped vessels and rectangular vessels with rounded ends, provided an effective approach for disrupting a VSB by creating a release path for gas as a VSB began to rise. Based on the success of sloped-wall vessels, a similar concept is investigated here where a sloped fin is placed inside the vessel to create a release path for gas. A key potential advantage of using a sloped fin compared to a vessel with a sloped wall is that a small fin decreases the volume of a vessel available for sludge storage by a very small fraction compared to a cone-shaped vessel. The purpose of this study is to quantify the capability of sloped fins to disrupt VSBs and to conduct sufficient tests to estimate the performance of fins in full-scale STSCs. Experiments were conducted with a range of fin shapes to determine what slope and width were sufficient to disrupt VSBs. Additional tests were conducted to demonstrate how the fin performance scales with the sludge layer thickness and the sludge strength, density, and vessel diameter based on the gravity yield parameter, which is a dimensionless ratio of the force necessary to yield the sludge to its weight.( ) Further experiments evaluated the difference between vessels with flat and 2:1 elliptical bottoms and a number of different simulants, including the KW container sludge simulant (complete), which was developed to match actual K-Basin sludge. Testing was conducted in 5-in., 10-in., and 23-in.-diameter vessels to quantify how fin performance is impacted by the size of the test vessel. The most significant results for these scale-up tests are the trend in how behavior changes with vessel size and the results from the 23-in. vessel. The key objective in evaluating fin performance is to determine the conditions that minimize the volume of a VSB when disruption occurs because this reduces the potential for material inside the STSC from being released through vents.

  4. Bull. U. S. F. C. 1890. Fashine Vessels of the Pacific Coast. (To lace page 13.) PLATEV. %-THE FISHING VESSELS AND BOATS OF THE PACIFIC*COAST."

    E-Print Network [OSTI]

    Bull. U. S. F. C. 1890. Fashine Vessels of the Pacific Coast. (To lace page 13.) PLATEV. #12;%-THE iiitenciod for publication as n part of a report on tho fisheries of the Pacific Coast of the TJnited Stattes

  5. Effect of the Young modulus variability on the mechanical behaviour of a nuclear containment vessel

    E-Print Network [OSTI]

    Effect of the Young modulus variability on the mechanical behaviour of a nuclear containment vessel on the mechanical behaviour of a nuclear containment vessel in case of a loss of cooling agent accident and under values are observed. Preprint submitted to Nuclear Engineering and Design April 19, 2010 hal-00542640

  6. Integrated Design, Operation and Control of Batch Extractive Distillation with a Middle Vessel

    E-Print Network [OSTI]

    Skogestad, Sigurd

    Integrated Design, Operation and Control of Batch Extractive Distillation with a Middle Vessel E. K distillation for separating homogeneous minimum-boiling azeotropic mixtures, where the extractive agent and a control structure for the batch extractive middle vessel distillation is proposed. In extractive

  7. Integrated Design, Operation and Control of Batch Extractive Distillation with a Middle Vessel

    E-Print Network [OSTI]

    Skogestad, Sigurd

    Integrated Design, Operation and Control of Batch Extractive Distillation with a Middle Vessel E. K distillation for separating homogeneous minimum­boiling azeotropic mixtures, where the extractive agent and a control structure for the batch extractive middle vessel distillation is proposed. In extractive

  8. Tissue-Engineered Vascular Grafts as In Vitro Blood Vessel Mimics for the Evaluation of Endothelialization

    E-Print Network [OSTI]

    Barton, Jennifer K.

    -dimensional in vitro blood vessel mimic (BVM) would be ideal for device testing before animal or clinicalTissue-Engineered Vascular Grafts as In Vitro Blood Vessel Mimics for the Evaluation nuclear staining and optical coherence tomography (OCT). En face and cross-sectional evaluation

  9. Modelling the Electron Beam Welding of Nuclear Reactor Pressure Vessel Steel

    E-Print Network [OSTI]

    Cambridge, University of

    Modelling the Electron Beam Welding of Nuclear Reactor Pressure Vessel Steel Christopher J. Duffy fabrication of thick-section steel for critical components such as reactor pressure vessels. Electron beam weld tests performed by Rolls-Royce and The Welding Institute of SA 508 Grade 3 and SA 508 Grade 4N

  10. D0 Silicon Upgrade: Gas Helium Storage Tank Pressure Vessel Engineering Note

    SciTech Connect (OSTI)

    Rucinski, Russ; /Fermilab

    1996-11-11T23:59:59.000Z

    This is to certify that Beaird Industries, Inc. has done a white metal blast per SSPC-SP5 as required per specifications on the vessel internal. Following the blast, a black light inspection was performed by Beaird Quality Control personnel to assure that all debris, grease, etc. was removed and interior was clean prior to closing vessel for helium test.

  11. Southwest Research Institute (SwRI) designs, analyzes, and fabricates pressure vessels

    E-Print Network [OSTI]

    Chapman, Clark R.

    vessels using: n ASME B&PV Code, Section VIII, Division 1 n ASME B&PV Code, Section VIII, Division 2 n ASME B&PV Code, Section VIII, Division 3 n ASME Pressure Vessels for Human Occupancy n American Bureau for the Design, Fabrication, and Erection of Structural Steel for Buildings" n Fabrication n ASME B&PV Code

  12. Sustainable What Federal permits are required for charter/party vessels?

    E-Print Network [OSTI]

    the fish is lying on its size (see Figure 1 on page 2). For black sea bass, the total length measurement flounder, scup, and black sea bass are among the most popular recreationally caught fish along the Atlantic types of summer flounder, scup, and black sea bass vessel permits-- one for vessels for hire (charter

  13. Conceptual Design of a Reactor Pressure Vessel and its Internals for a HPLWR

    SciTech Connect (OSTI)

    Fischer, Kai [EnBW Kraftwerke AG, Kernkraftwerk Philippsburg, Rheinschanzinsel D-76661 Philippsburg (Germany); Starflinger, Joerg; Schulenberg, Thomas [Forschungszentrum Karlsruhe GmbH, Institute for Nuclear and Energy Technologies P.O. Box 3640, D-76021 Karlsruhe (Germany)

    2006-07-01T23:59:59.000Z

    A design for the Reactor Pressure Vessel (RPV) and its internals for a HPLWR (High Performance Light Water Reactor) is presented. The RPV has been dimensioned using the pressure vessel code for nuclear power plants in Germany. In order to use conventional vessel materials such as 20 MnMoNi 5 5 (United States: SA 508), the vessel inner wall has to be kept only in contact with coolant at inlet temperature. Therefore, the hot coolant pipe connection from the steam plenum to the outlet is separated from the RPV inner wall using a thermal sleeve. The core inside the vessel rests on a support plate which is connected to the core barrel. The steam plenum is fixed on top of the core using support brackets which are attached to the adjustable steam outlet pipes. This way, the steam plenum rests on the outlet flanges of the lower vessel, while the core barrel is suspended at the closure head flange of the vessel to control thermal expansions between the internals and the RPV and to minimize thermal stresses. Both, inlet and outlet mass flows are separated via C-ring seals to prevent mixing. The control rod guides in the upper plenum are also suspended at the vessel flange and aligned inside the core barrel using centering pins. (authors)

  14. PPPL-3458 PPPL-3458 Visual Tritium Imaging Of In-Vessel Surfaces

    E-Print Network [OSTI]

    PPPL-3458 PPPL-3458 UC-70 Visual Tritium Imaging Of In-Vessel Surfaces by C. A. Gentile, S. J: http://www.ntis.gov/ordering.htm #12;1 Visual Tritium Imaging Of In-Vessel Surfaces C. A. Gentile, S. J Energy Research Institute, Tritium Engineering Laboratory, Tokai, Ibaraki 319-1195, Japan Abstract

  15. Blood Vessel Normalization in the Hamster Oral Cancer Model for Experimental Cancer Therapy Studies

    SciTech Connect (OSTI)

    Ana J. Molinari; Romina F. Aromando; Maria E. Itoiz; Marcela A. Garabalino; Andrea Monti Hughes; Elisa M. Heber; Emiliano C. C. Pozzi; David W. Nigg; Veronica A. Trivillin; Amanda E. Schwint

    2012-07-01T23:59:59.000Z

    Normalization of tumor blood vessels improves drug and oxygen delivery to cancer cells. The aim of this study was to develop a technique to normalize blood vessels in the hamster cheek pouch model of oral cancer. Materials and Methods: Tumor-bearing hamsters were treated with thalidomide and were compared with controls. Results: Twenty eight hours after treatment with thalidomide, the blood vessels of premalignant tissue observable in vivo became narrower and less tortuous than those of controls; Evans Blue Dye extravasation in tumor was significantly reduced (indicating a reduction in aberrant tumor vascular hyperpermeability that compromises blood flow), and tumor blood vessel morphology in histological sections, labeled for Factor VIII, revealed a significant reduction in compressive forces. These findings indicated blood vessel normalization with a window of 48 h. Conclusion: The technique developed herein has rendered the hamster oral cancer model amenable to research, with the potential benefit of vascular normalization in head and neck cancer therapy.

  16. Creep of A508/533 Pressure Vessel Steel

    SciTech Connect (OSTI)

    Richard Wright

    2014-08-01T23:59:59.000Z

    ABSTRACT Evaluation of potential Reactor Pressure Vessel (RPV) steels has been carried out as part of the pre-conceptual Very High Temperature Reactor (VHTR) design studies. These design studies have generally focused on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Initially, three candidate materials were identified by this process: conventional light water reactor (LWR) RPV steels A508 and A533, 2¼Cr-1Mo in the annealed condition, and Grade 91 steel. The low strength of 2¼Cr-1Mo at elevated temperature has eliminated this steel from serious consideration as the VHTR RPV candidate material. Discussions with the very few vendors that can potentially produce large forgings for nuclear pressure vessels indicate a strong preference for conventional LWR steels. This preference is based in part on extensive experience with forging these steels for nuclear components. It is also based on the inability to cast large ingots of the Grade 91 steel due to segregation during ingot solidification, thus restricting the possible mass of forging components and increasing the amount of welding required for completion of the RPV. Grade 91 steel is also prone to weld cracking and must be post-weld heat treated to ensure adequate high-temperature strength. There are also questions about the ability to produce, and very importantly, verify the through thickness properties of thick sections of Grade 91 material. The availability of large components, ease of fabrication, and nuclear service experience with the A508 and A533 steels strongly favor their use in the RPV for the VHTR. Lowering the gas outlet temperature for the VHTR to 750°C from 950 to 1000°C, proposed in early concept studies, further strengthens the justification for this material selection. This steel is allowed in the ASME Boiler and Pressure Vessel Code for nuclear service up to 371°C (700°F); certain excursions above that temperature are allowed by Code Case N-499-2 (now incorporated as an appendix to Section III Division 5 of the Code). This Code Case was developed with a rather sparse data set and focused primarily on rolled plate material (A533 specification). Confirmatory tests of creep behavior of both A508 and A533 are described here that are designed to extend the database in order to build higher confidence in ensuring the structural integrity of the VHTR RPV during off-normal conditions. A number of creep-rupture tests were carried out at temperatures above the 371°C (700°F) Code limit; longer term tests designed to evaluate minimum creep behavior are ongoing. A limited amount of rupture testing was also carried out on welded material. All of the rupture data from the current experiments is compared to historical values from the testing carried out to develop Code Case N-499-2. It is shown that the A508/533 basemetal tested here fits well with the rupture behavior reported from the historical testing. The presence of weldments significantly reduces the time to rupture. The primary purpose of this report is to summarize and record the experimental results in a single document.

  17. Lucrative Opportunities in Asia Pacific to Help Global Bunker...

    Open Energy Info (EERE)

    Energy Concerns to Push Global Market to Grow at 8.1% CAGR from 2013 to 2019 Oil Shale Market is Estimated to Reach USD 7,400.70 Million by 2022 more Group members (32)...

  18. The readers point vessel: hull analysis of an eighteenth century merchant sloop excavated in St. Ann's Bay, Jamaica

    E-Print Network [OSTI]

    Cook, Gregory D.

    1997-01-01T23:59:59.000Z

    's Bay, Jamaica in 1994. Excavators removed overburden and the ballast pile, recovering over 600 artifacts associated with the vessel-After exposing well-preserved hull remains, divers recorded the ship's structure. The vessel is preserved from the base...

  19. Investigation of downward facing critical heat flux with water-based nanofluids for In-Vessel Retention applications

    E-Print Network [OSTI]

    DeWitt, Gregory L

    2011-01-01T23:59:59.000Z

    In-Vessel Retention ("IVR") is a severe accident management strategy that is power limiting to the Westinghouse AP1000 due to critical heat flux ("CHF") at the outer surface of the reactor vessel. Increasing the CHF level ...

  20. Distribution of Hydrogen Isotopes, Carbon and Beryllium on In-Vessel Surfaces in the Various JET Divertors

    E-Print Network [OSTI]

    Distribution of Hydrogen Isotopes, Carbon and Beryllium on In-Vessel Surfaces in the Various JET Divertors

  1. Theoretical and Experimental Simulation of Accident Scenarios of the JET Cryogenic Components Part I: The JET In-vessel Cryopump

    E-Print Network [OSTI]

    Theoretical and Experimental Simulation of Accident Scenarios of the JET Cryogenic Components Part I: The JET In-vessel Cryopump

  2. Method for forming a bladder for fluid storage vessels

    DOE Patents [OSTI]

    Mitlitsky, Fred (Livermore, CA); Myers, Blake (Livermore, CA); Magnotta, Frank (Lafayette, CA)

    2000-01-01T23:59:59.000Z

    A lightweight, low permeability liner for graphite epoxy composite compressed gas storage vessels. The liner is composed of polymers that may or may not be coated with a thin layer of a low permeability material, such as silver, gold, or aluminum, deposited on a thin polymeric layer or substrate which is formed into a closed bladder using torispherical or near torispherical end caps, with or without bosses therein, about which a high strength to weight material, such as graphite epoxy composite shell, is formed to withstand the storage pressure forces. The polymeric substrate may be laminated on one or both sides with additional layers of polymeric film. The liner may be formed to a desired configuration using a dissolvable mandrel or by inflation techniques and the edges of the film seamed by heat sealing. The liner may be utilized in most any type of gas storage system, and is particularly applicable for hydrogen, gas mixtures, and oxygen used for vehicles, fuel cells or regenerative fuel cell applications, high altitude solar powered aircraft, hybrid energy storage/propulsion systems, and lunar/Mars space applications, and other applications requiring high cycle life.

  3. Immersive Volume Rendering of Blood Vessels Gregory Long, Han Suk Kim, Alison Marsden, Yuri Bazilevs, Jurgen P. Schulze

    E-Print Network [OSTI]

    Schulze, Jürgen P.

    Immersive Volume Rendering of Blood Vessels Gregory Long, Han Suk Kim, Alison Marsden, Yuri ABSTRACT In this paper, we present a novel method of visualizing flow in blood vessels. Our approach reads to the sparse structure of blood vessels, we utilize an octree to efficiently store the resampled data

  4. High pressure ejection of melt from a reactor pressure vessel. The discharge phase. Revision 7

    SciTech Connect (OSTI)

    Pilch, M.; Tarbell, W.M.

    1985-09-01T23:59:59.000Z

    Recent probabilistic risk-assessment studies identified potential accident sequences in which reactor vessel failure occurs while the primary system is at elevated pressure. The phenomenology of the discharge phase is reviewed here. We propose an improved model for hole ablation following vessel failure, and we compare the model with experiment data. Gas blowthrough is identified as a mechanism that allows steam to escape through the vessel breach before melt ejection is complete. Gas blowthrough leads to pneumatic atomization of the remaining melt before significant depressurization of the primary system occurs.

  5. R&D of Large Stationary Hydrogen/CNG/HCNG Storage Vessels

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

    Stationary HydrogenCNGHCNG Storage Vessels September 28, 2010 Add hydrogen to natural gas makes it burn more cleanly (notably reducing smog-causing NO X by 50%). HCNG...

  6. Bronchopulmonary Dysplasia, Idiopathic Pulmonary Arterial Hypertension, and Wave Modeling in Stented Vessels

    E-Print Network [OSTI]

    Peters, Andrew

    2011-08-04T23:59:59.000Z

    arterial hypertension (PAH), to identify the hemodynamic attributes which could be altered to ameliorate the progression of these diseases. We then simulated blood flow through five, simple finite element vessel models to determine the effects of stents...

  7. Potential market for LNG-fueled marine vessels in the United States

    E-Print Network [OSTI]

    Brett, Bridget C

    2008-01-01T23:59:59.000Z

    The growing global concern over ship emissions in recent years has driven policy change at the international level toward more stringent vessel emissions standards. The policy change has also been an impetus for innovation ...

  8. Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

    DOE Patents [OSTI]

    Schreiber, R.B.; Fero, A.H.; Sejvar, J.

    1997-12-16T23:59:59.000Z

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor. 8 figs.

  9. Analyses of a steel containment vessel with an outer contact structure under severe internal overpressurization conditions

    SciTech Connect (OSTI)

    Porter, V.L.

    1993-12-31T23:59:59.000Z

    Many Mark-I and Mark-II BWR plants are designed with a steel vessel as the primary containment. Typically, the steel containment vessel (SCV) is enclosed within a reinforced concrete shield building with only a small gap (50--90mm) separating the two structures. This paper describes finite element analyses performed to evaluate the effects of contact and friction between a steel containment vessel and an outer contact structure when the containment vessel is subjected to large internal pressures. These computations were motivated by a joint program on containment integrity involving the Nuclear Power Engineering Corporation (NUPEC) of Japan, the US Nuclear Regulatory Commission (NRC), and Sandia National Laboratories for testing model containments.

  10. A model for determining the fate of hazardous constituents in waste during in-vessel composting

    E-Print Network [OSTI]

    Bollineni, Prasanthi

    1994-01-01T23:59:59.000Z

    compound undergoes when subjected to composting. The purpose of this thesis is to define these processes and develop a model for determining the fate of organic compounds in waste during in-vessel composting Volatilization and biodegradation are found...

  11. Blood vessel detection in retinal images and its application in diabetic retinopathy screening

    E-Print Network [OSTI]

    Zhang, Ming

    2009-05-15T23:59:59.000Z

    transform (RCT) algorithm, which converts the intensity information in spatial domain to a high dimensional radial contrast domain. Different feature descriptors are designed to improve the speed, sensitivity, and expandability of the vessel detection system...

  12. ITER vacuum vessel design (D201 subtask 1.3 and subtask 3). Final report

    SciTech Connect (OSTI)

    NONE

    1996-08-01T23:59:59.000Z

    ITER Task No. D201, Vacuum Vessel Design (Subtask 1.3 and Subtask 3), was initiated to propose and evaluate local vacuum vessel reinforcement alternatives in proximity to the Neutral Beam, Radial Mid-Plane, Top, and Divertor Ports. These areas were reported to be highly stressed regions based on the results of preliminary stress analyses performed by the USHT (US Home Team) and the ITER Joint Central Team (JCT) at the Garching JWS (Joint Work Site). Initial design activities focused on the divertor port region which was reported to experience the highest stress intensities. Existing stress analysis models and results were reviewed with the USHT stress analysts to obtain an overall understanding of the vessel response to the various applied loads. These reviews indicated that the reported stress intensities in the divertor port region were significantly affected by the loads applied to the vessel in adjacent regions.

  13. Vessel Cold-Ironing Using a Barge Mounted PEM Fuel Cell: Project...

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

    Unlimited Release Printed February 2013 Vessel Cold-Ironing Using a Barge Mounted PEM Fuel Cell: Project Scoping and Feasibility Joseph W. Pratt and Aaron P. Harris Prepared by...

  14. E-Print Network 3.0 - advanced in-vessel retention Sample Search...

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

    Technologies... reactors. 12;Engineering Peer Review June 5-7, 2001 19 FIRE In-Vessel Remote Handling System Mi Transfer... Cask Articulated Boom Boom End-Effector Midplane...

  15. Cost reduction of polar class vessels : structural optimization that includes production factors

    E-Print Network [OSTI]

    Normore, Stephen S. (Stephen Selwyn)

    2013-01-01T23:59:59.000Z

    The design of ship structures was normally optimized to reduce construction material and maintain adequate strength while adhering to a given classification society's rules. In the case of Polar Class vessels, where weight ...

  16. Metallic Pressure Vessels Failures M. Mosnier, B. Daudonnet, J. Renard and G. Mavrothalassitis

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    to store or to transport gas or pressurized liquid (such as LPG or LNG), to dry, or as steam boiler... etc of thé vessel is usually achieved with thé help of handbooks, that sometimes overestimate effects

  17. Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

    DOE Patents [OSTI]

    Schreiber, Roger B. (Penn Twp., PA); Fero, Arnold H. (New Kensington, PA); Sejvar, James (Murrysville, PA)

    1997-01-01T23:59:59.000Z

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor.

  18. Painting blood vessels and atherosclerotic plaques with an adhesive drug depot

    E-Print Network [OSTI]

    Kastrup, Christian J.

    The treatment of diseased vasculature remains challenging, in part because of the difficulty in implanting drug-eluting devices without subjecting vessels to damaging mechanical forces. Implanting materials using adhesive ...

  19. Hydrogen degradation and microstructural effects of the near-threshold fatigue resistance of pressure vessel steels

    E-Print Network [OSTI]

    Fuquen-Molano, Rosendo

    1982-01-01T23:59:59.000Z

    Safety of pressure vessels for applications such as coal conversion reactors requires understanding of the mechanism of environmentally-induced crack propagation and the mechanism by which process-induced microstructures ...

  20. A critical contraction frequency in lymphatic vessels: transition to a state of partial summation

    E-Print Network [OSTI]

    Meisner, Joshua Keith

    2009-06-02T23:59:59.000Z

    , 57)). Therefore, the relaxation rate of ventricles is high, and a calcium plateau creates an extended refractory period in the cardiac action potential to minimize summation. In contrast, blood vessels, which must regulate blood flow through... lymphatic vessels possess a refractory period that prevents tetanus, the effective refractory period is less 3 than total contraction time, ending at ~50% relaxation (28). Taken together, no one has demonstrated a mechanism that prevents the summation...

  1. Transient PVT measurements and model predictions for vessel heat transfer. Part II.

    SciTech Connect (OSTI)

    Felver, Todd G.; Paradiso, Nicholas Joseph; Winters, William S., Jr.; Evans, Gregory Herbert; Rice, Steven F.

    2010-07-01T23:59:59.000Z

    Part I of this report focused on the acquisition and presentation of transient PVT data sets that can be used to validate gas transfer models. Here in Part II we focus primarily on describing models and validating these models using the data sets. Our models are intended to describe the high speed transport of compressible gases in arbitrary arrangements of vessels, tubing, valving and flow branches. Our models fall into three categories: (1) network flow models in which flow paths are modeled as one-dimensional flow and vessels are modeled as single control volumes, (2) CFD (Computational Fluid Dynamics) models in which flow in and between vessels is modeled in three dimensions and (3) coupled network/CFD models in which vessels are modeled using CFD and flows between vessels are modeled using a network flow code. In our work we utilized NETFLOW as our network flow code and FUEGO for our CFD code. Since network flow models lack three-dimensional resolution, correlations for heat transfer and tube frictional pressure drop are required to resolve important physics not being captured by the model. Here we describe how vessel heat transfer correlations were improved using the data and present direct model-data comparisons for all tests documented in Part I. Our results show that our network flow models have been substantially improved. The CFD modeling presented here describes the complex nature of vessel heat transfer and for the first time demonstrates that flow and heat transfer in vessels can be modeled directly without the need for correlations.

  2. William W. Hay Railroad Engineering Seminar Freight Railroad Energy

    E-Print Network [OSTI]

    Barkan, Christopher P.L.

    fuel & alternatives (Btu's per gallon) Biodiesel, Fischer-Tropsch syn fuel & DME Liquefied natural.2008 Billion US gallons = 27.258 Billion liters of diesel fuel equivalent * Ton-mile statistics from US DOT (importance v difficulties) North American freight RRs (defining characteristics) Energy density of diesel

  3. Automated segmentation of the pulmonary arteries in low-dose CT by vessel tracking

    E-Print Network [OSTI]

    Wala, Jeremiah; Lee, Jaesung; Jirapatnakul, Artit; Biancardi, Alberto; Reeves, Anthony

    2011-01-01T23:59:59.000Z

    We present a fully automated method for top-down segmentation of the pulmonary arterial tree in low-dose thoracic CT images. The main basal pulmonary arteries are identified near the lung hilum by searching for candidate vessels adjacent to known airways, identified by our previously reported airway segmentation method. Model cylinders are iteratively fit to the vessels to track them into the lungs. Vessel bifurcations are detected by measuring the rate of change of vessel radii, and child vessels are segmented by initiating new trackers at bifurcation points. Validation is accomplished using our novel sparse surface (SS) evaluation metric. The SS metric was designed to quantify the magnitude of the segmentation error per vessel while significantly decreasing the manual marking burden for the human user. A total of 210 arteries and 205 veins were manually marked across seven test cases. 134/210 arteries were correctly segmented, with a specificity for arteries of 90%, and average segmentation error of 0.15 mm...

  4. Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2010-01-01T23:59:59.000Z

    1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

  5. CFD Analyses of Damaged Fuel Inside a Cleaning Vessel

    SciTech Connect (OSTI)

    Legradi, Gabor; Boros, Ildiko; Aszodi, Attila [Budapest University of Technology and Economics, Muegyetem rkp. 3-9. H-1111 Budapest (Hungary)

    2006-07-01T23:59:59.000Z

    On 10-11 of April, 2003, a serious incident occurred in a special fuel assembly cleaning tank, which was installed into the service shaft of the 2. unit of the Paks NPP in Hungary. During this incident, most of the 30 fuel assemblies put into the cleaning tank have seriously damaged. In the Institute of Nuclear Techniques of the Budapest University of Technology and Economics several CFD investigations were performed concerning the course of the incident, the post incidental conditions and the recovery work. The main reason of the incident can be originated from the defective design of the cleaning tank which resulted in the insufficient cooling of the system in a special operational mode. Our investigation performed with a complex 3D CFX model clearly showed how could as strong temperature stratification develop inside the cleaning tank that it was able to block the coolant flow through the fuel assemblies. After the blocking of the flow, the coolant turned into boiling and the assemblies became uncovered. The temperature of the surfaces of the fuel assemblies went above 1000 deg. C. With the aid of the radiative heat transfer model of the CFX-5.6 code, the surface temperatures were analyzed. When the cleaning instrument got opened the fuel assemblies suffered a serious thermal shock and the assemblies highly damaged. The post-incident thermo-hydraulic state inside the cleaning vessel was investigated with a rather complex CFX model. The uncertainties were decreased by a wide parameter study. The recovery work is planned to be started in the close future. The operators of the damaged fuel removing equipments will work standing on a platform which will be placed into the service shaft just above the surface of the coolant of decreased level. Protecting the workers against unnecessary personal doses is a very important task. In this situation, while the coolant is important part of the biological shielding, it is also a source of radiation due to the considerable amount of radioactive contamination dispersed into it. Therefore, the 3D distribution of the contamination in the service shaft was estimated for different operational and incidental scenarios with a wide parameter study made by a 3D CFX model. This comprehensive work performed with several models and calculations is tersely outlined according to the limited extent of the paper. (authors)

  6. Structural integrity assessment of type 201LN stainless steel cryogenic pressure vessels

    SciTech Connect (OSTI)

    Rana, M.D.; Zawierucha, R. [Praxair, Inc., Tonawanda, NY (United States)

    1995-12-01T23:59:59.000Z

    The ASME Boiler and Pressure Vessel Code Committee approved the Code Case 2123 in 1992 which allows the use of Type 201LN stainless steel in the construction of ASME Section VIII, Division 1 and Division 2 pressure vessels for -320{degrees}F applications. Type 201LN stainless steel is a nitrogen strengthened modified version of ASTM A240, Type 201 stainless steel with a restricted chemistry. The Code allowable design stresses for Type 201LN for Division 1 vessels are approximately 27% higher than Type 304 stainless steel and equal to that of the 5 Ni and 9 Ni steels. This paper discusses the important features of the Code Case 2123 and the structural integrity assessment of Type 201LN stainless steel cryogenic vessels. Tensile, Charpy-V-notch and fracture properties have been obtained on several heats of this steel including weldments. A linear-elastic fracture mechanics analysis has been conducted to assess the expected fracture mode and the fracture-critical crack sizes. The results have been compared with Type 304 stainless steel, 5 Ni and 9 Ni steel vessels.

  7. Ex-Vessel Core Melt Modeling Comparison between MELTSPREAD-CORQUENCH and MELCOR 2.1

    SciTech Connect (OSTI)

    Robb, Kevin R [ORNL; Farmer, Mitchell [Argonne National Laboratory (ANL); Francis, Matthew W [ORNL

    2014-03-01T23:59:59.000Z

    System-level code analyses by both United States and international researchers predict major core melting, bottom head failure, and corium-concrete interaction for Fukushima Daiichi Unit 1 (1F1). Although system codes such as MELCOR and MAAP are capable of capturing a wide range of accident phenomena, they currently do not contain detailed models for evaluating some ex-vessel core melt behavior. However, specialized codes containing more detailed modeling are available for melt spreading such as MELTSPREAD as well as long-term molten corium-concrete interaction (MCCI) and debris coolability such as CORQUENCH. In a preceding study, Enhanced Ex-Vessel Analysis for Fukushima Daiichi Unit 1: Melt Spreading and Core-Concrete Interaction Analyses with MELTSPREAD and CORQUENCH, the MELTSPREAD-CORQUENCH codes predicted the 1F1 core melt readily cooled in contrast to predictions by MELCOR. The user community has taken notice and is in the process of updating their systems codes; specifically MAAP and MELCOR, to improve and reduce conservatism in their ex-vessel core melt models. This report investigates why the MELCOR v2.1 code, compared to the MELTSPREAD and CORQUENCH 3.03 codes, yield differing predictions of ex-vessel melt progression. To accomplish this, the differences in the treatment of the ex-vessel melt with respect to melt spreading and long-term coolability are examined. The differences in modeling approaches are summarized, and a comparison of example code predictions is provided.

  8. Growing consumption of petroleum products worldwide has resulted in the proliferation of vessels carrying oil, chemicals, and gases

    E-Print Network [OSTI]

    Neimark, Alexander V.

    Growing consumption of petroleum products worldwide has resulted in the proliferation of vessels carrying oil, chemicals, and gases into our harbors. Meeting our society's surging demand for commodities

  9. Modular Inspection System for a Complete IN-Service Examination of Nuclear Reactor Pressure Vessel, Including Beltline Region

    SciTech Connect (OSTI)

    David H. Bothell

    2000-04-30T23:59:59.000Z

    Final Report for a DOE Phase II Contract Describing the design and fabrication of a reactor inspection modular rover prototype for reactor vessel inspection.

  10. Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2003-01-01T23:59:59.000Z

    1.1 This guide covers the general procedures to be considered for conducting an in-service thermal anneal of a light-water moderated nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. The improvement in mechanical properties generally is assessed using Charpy V-notch impact test results, or alternatively, fracture toughness test results or inferred toughness property changes from tensile, hardness, indentation, or other miniature specimen testing (1). 1.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods. Certain inherent limiting factors must be considered in developing an annealing procedure. These factors include system-design limitations; physical constrain...

  11. RELAP5/MOD2 split reactor vessel model and steamline break analysis

    SciTech Connect (OSTI)

    Petelin, S.; Mavko, B.; Gortnar, O. (Univ. of Ljubljana, (Slovenia))

    1993-04-01T23:59:59.000Z

    A split reactor vessel model for the RELAP5/ MOD2 computer code is developed in an attempt to realize more realistic predictions of asymmetrical transients in a two-loop nuclear power plant. Based on this split reactor model, coolant mixing processes within the reactor vessel are examined. This study evaluates the model improvements in terms of thermal-hydraulic simulations of the reactor core inlet fluid condition and the consequent core behavior. Furthermore, the split reactor vessel model is introduced into an integral RELAP5/MOD2 power plant model, and a steamline break analysis is performed to determine the influence of the boron concentration in the boron injection tank on accident consequences.

  12. In-Vessel Coil Material Failure Rate Estimates for ITER Design Use

    SciTech Connect (OSTI)

    L. C. Cadwallader

    2013-01-01T23:59:59.000Z

    The ITER international project design teams are working to produce an engineering design for construction of this large tokamak fusion experiment. One of the design issues is ensuring proper control of the fusion plasma. In-vessel magnet coils may be needed for plasma control, especially the control of edge localized modes (ELMs) and plasma vertical stabilization (VS). These coils will be lifetime components that reside inside the ITER vacuum vessel behind the blanket modules. As such, their reliability is an important design issue since access will be time consuming if any type of repair were necessary. The following chapters give the research results and estimates of failure rates for the coil conductor and jacket materials to be used for the in-vessel coils. Copper and CuCrZr conductors, and stainless steel and Inconel jackets are examined.

  13. Reactor pressure vessel head vents and methods of using the same

    DOE Patents [OSTI]

    Gels, John L; Keck, David J; Deaver, Gerald A

    2014-10-28T23:59:59.000Z

    Internal head vents are usable in nuclear reactors and include piping inside of the reactor pressure vessel with a vent in the reactor upper head. Piping extends downward from the upper head and passes outside of the reactor to permit the gas to escape or be forcibly vented outside of the reactor without external piping on the upper head. The piping may include upper and lowers section that removably mate where the upper head joins to the reactor pressure vessel. The removable mating may include a compressible bellows and corresponding funnel. The piping is fabricated of nuclear-reactor-safe materials, including carbon steel, stainless steel, and/or a Ni--Cr--Fe alloy. Methods install an internal head vent in a nuclear reactor by securing piping to an internal surface of an upper head of the nuclear reactor and/or securing piping to an internal surface of a reactor pressure vessel.

  14. The chemistry of tributyl phosphate at elevated temperatures in the Plutonium Finishing Plant Process Vessels

    SciTech Connect (OSTI)

    Barney, G.S.; Cooper, T.D.

    1994-06-01T23:59:59.000Z

    Potentially violent chemical reactions of the tributyl phosphate solvent used by the Plutonium Finishing Plant at the Hanford Site were investigated. There is a small probability that a significant quantity of this solvent could be accidental transferred to heated process vessels and react there with nitric acid or plutonium nitrate also present in the solvent extraction process. The results of laboratory studies of the reactions show that exothermic oxidation of tributyl phosphate by either nitric acid or actinide nitrates is slow at temperatures expected in the heated vessels. Less than four percent of the tributyl phosphate will be oxidized in these vented vessels at temperatures between 125{degrees}C and 250{degrees}C because the oxidant will be lost from the vessels by vaporization or decomposition before the tributyl phosphate can be extensively oxidized. The net amounts of heat generated by oxidation with concentrated nitric acid and with thorium nitrate (a stand-in for plutonium nitrate) were determined to be about -150 and -220 joules per gram of tributyl phosphate initially present, respectively. This is not enough heat to cause violent reactions in the vessels. Pyrolysis of the tributyl phosphate occurred in these mixtures at temperatures of 110{degrees}C to 270{degrees}C and produced mainly 1-butene gas, water, and pyrophosphoric acid. Butene gas generation is slow at expected process vessel temperatures, but the rate is faster at higher temperatures. At 252{degrees}C the rate of butene gas generated was 0.33 g butene/min/g of tributyl phosphate present. The measured heat absorbed by the pyrolysis reaction was 228 J/g of tributyl phosphate initially present (or 14.5 kcal/mole of tributyl phosphate). Release of flammable butene gas into process areas where it could ignite appears to be the most serious safety consideration for the Plutonium Finishing Plant.

  15. Scaled Testing to Evaluate Pulse Jet Mixer Performance in Waste Treatment Plant Mixing Vessels

    SciTech Connect (OSTI)

    Fort, James A.; Meyer, Perry A.; Bamberger, Judith A.; Enderlin, Carl W.; Scott, Paul A.; Minette, Michael J.; Gauglitz, Phillip A.

    2010-03-07T23:59:59.000Z

    The Waste Treatment and Immobilization Plant (WTP) at Hanford is being designed and built to pre-treat and vitrify the waste in Hanford’s 177 underground waste storage tanks. Numerous process vessels will hold waste at various stages in the WTP. These vessels have pulse jet mixer (PJM) systems. A test program was developed to evaluate the adequacy of mixing system designs in the solids-containing vessels in the WTP. The program focused mainly on non-cohesive solids behavior. Specifically, the program addressed the effectiveness of the mixing systems to suspend settled solids off the vessel bottom, and distribute the solids vertically. Experiments were conducted at three scales using various particulate simulants. A range of solids loadings and operational parameters were evaluated, including jet velocity, pulse volume, and duty cycle. In place of actual PJMs, the tests used direct injection from tubes with suction at the top of the tank fluid. This gave better control over the discharge duration and duty cycle and simplified the facility requirements. The mixing system configurations represented in testing varied from 4 to 12 PJMs with various jet nozzle sizes. In this way the results collected could be applied to the broad range of WTP vessels with varying geometrical configurations and planned operating conditions. Data for “just-suspended velocity”, solids cloud height, and solids concentration vertical profile were collected, analyzed, and correlated. The correlations were successfully benchmarked against previous large-scale test results, then applied to the WTP vessels using reasonable assumptions of anticipated waste properties to evaluate adequacy of the existing mixing system designs.

  16. Protective interior wall and attach8ing means for a fusion reactor vacuum vessel

    DOE Patents [OSTI]

    Phelps, Richard D. (Greeley, CO); Upham, Gerald A. (Valley Center, CA); Anderson, Paul M. (San Diego, CA)

    1988-01-01T23:59:59.000Z

    An array of connected plates mounted on the inside wall of the vacuum vessel of a magnetic confinement reactor in order to provide a protective surface for energy deposition inside the vessel. All fasteners are concealed and protected beneath the plates, while the plates themselves share common mounting points. The entire array is installed with torqued nuts on threaded studs; provision also exists for thermal expansion by mounting each plate with two of its four mounts captured in an oversize grooved spool. A spool-washer mounting hardware allows one edge of a protective plate to be torqued while the other side remains loose, by simply inverting the spool-washer hardware.

  17. Ceramic vessel production, use and distribution in Northern Mesopotamia and Syria during the Middle Bronze Age II (c. 1800-1600 BC). A functional analysis of vessels from Tell Ahmar, North Syria. 

    E-Print Network [OSTI]

    Perini, Silvia

    2014-07-03T23:59:59.000Z

    and ceramic production at a local and regional level are further investigated. Since there is no one-to-one relation between vessel type and vessel function, the research adopts a multi-dimensional approach formed by the following hierarchical investigations...

  18. A Report on Policies and Practices of the U.S. Navy for Naming the Vessels of the Navy

    E-Print Network [OSTI]

    A Report on Policies and Practices of the U.S. Navy for Naming the Vessels of the Navy Prepared by: Department of the Navy 1000 Navy Pentagon Rm. 4E720 Washington, DC the Vessels of the Navy 1 Purpose Background Orthodox Traditionalists versus Pragmatic

  19. DEVELOPMENT OF ASME SECTION X CODE RULES FOR HIGH PRESSURE COMPOSITE HYDROGEN PRESSURE VESSELS WITH NON-LOAD SHARING LINERS

    SciTech Connect (OSTI)

    Rawls, G.; Newhouse, N.; Rana, M.; Shelley, B.; Gorman, M.

    2010-04-13T23:59:59.000Z

    The Boiler and Pressure Vessel Project Team on Hydrogen Tanks was formed in 2004 to develop Code rules to address the various needs that had been identified for the design and construction of up to 15000 psi hydrogen storage vessel. One of these needs was the development of Code rules for high pressure composite vessels with non-load sharing liners for stationary applications. In 2009, ASME approved new Appendix 8, for Section X Code which contains the rules for these vessels. These vessels are designated as Class III vessels with design pressure ranging from 20.7 MPa (3,000 ps)i to 103.4 MPa (15,000 psi) and maximum allowable outside liner diameter of 2.54 m (100 inches). The maximum design life of these vessels is limited to 20 years. Design, fabrication, and examination requirements have been specified, included Acoustic Emission testing at time of manufacture. The Code rules include the design qualification testing of prototype vessels. Qualification includes proof, expansion, burst, cyclic fatigue, creep, flaw, permeability, torque, penetration, and environmental testing.

  20. High-resolution imaging of vessels in the isolated rat brain M. F. Valverde Salzmann1

    E-Print Network [OSTI]

    for the distribution of vessels in the rat brain. Angiography is able to use the blood flow in the brain of the living into a test tube filled with fluorocarbon fluid for susceptibility matching. MR Images were acquired at a horizontal 16.4 T scanner with a homemade microstrip volume coil, using gradient echo sequences

  1. Scaling Theory for Pulsed Jet Mixed Vessels, Sparging, and Cyclic Feed Transport Systems for Slurries

    SciTech Connect (OSTI)

    Kuhn, William L.; Rector, David R.; Rassat, Scot D.; Enderlin, Carl W.; Minette, Michael J.; Bamberger, Judith A.; Josephson, Gary B.; Wells, Beric E.; Berglin, Eric J.

    2013-09-27T23:59:59.000Z

    This document is a previously unpublished work based on a draft report prepared by Pacific Northwest National Laboratory (PNNL) for the Hanford Waste Treatment and Immobilization Plant (WTP) in 2012. Work on the report stopped when WTP’s approach to testing changed. PNNL is issuing a modified version of the document a year later to preserve and disseminate the valuable technical work that was completed. This document establishes technical bases for evaluating the mixing performance of Waste Treatment Plant (WTP) pretreatment process tanks based on data from less-than-full-scale testing, relative to specified mixing requirements. The technical bases include the fluid mechanics affecting mixing for specified vessel configurations, operating parameters, and simulant properties. They address scaling vessel physical performance, simulant physical performance, and “scaling down” the operating conditions at full scale to define test conditions at reduced scale and “scaling up” the test results at reduced scale to predict the performance at full scale. Essentially, this document addresses the following questions: • Why and how can the mixing behaviors in a smaller vessel represent those in a larger vessel? • What information is needed to address the first question? • How should the information be used to predict mixing performance in WTP? The design of Large Scale Integrated Testing (LSIT) is being addressed in other, complementary documents.

  2. Proceedings of PVP2006-ICPVT-11 2006 ASME Pressure Vessels and Piping Division Conference

    E-Print Network [OSTI]

    Barr, Al

    Proceedings of PVP2006-ICPVT-11 2006 ASME Pressure Vessels and Piping Division Conference July 23 response leading to large deformations. Some issues in measurement technique and validation testing of scientific investigation. It is a hazard that is occasion- ally encountered in the chemical [1,2], nuclear [3

  3. HFIR Vessel Maximum Permissible Pressures for Operating Period 26 to 50 EFPY (100 MW)

    SciTech Connect (OSTI)

    Cheverton, R.D.; Inger, J.R.

    1999-01-01T23:59:59.000Z

    Extending the life of the HFIR pressure vessel from 26 to 50 EFPY (100 MW) requires an updated calculation of the maximum permissible pressure for a range in vessel operating temperatures (40-120 F). The maximum permissible pressure is calculated using the equal-potential method, which takes advantage of knowledge gained from periodic hydrostatic proof tests and uses the test conditions (pressure, temperature, and frequency) as input. The maximum permissible pressure decreases with increasing time between hydro tests but is increased each time a test is conducted. The minimum values that occur just prior to a test either increase or decrease with time, depending on the vessel temperature. The minimum value of these minimums is presently specified as the maximum permissible pressure. For three vessel temperatures of particular interest (80, 88, and 110 F) and a nominal time of 3.0 EFPY(100 MVV)between hydro tests, these pressures are 677, 753, and 850 psi. For the lowest temperature of interest (40 F), the maximum permissible pressure is 295 psi.

  4. An investigation of RVACS (reactor vessel auxiliary cooling system) design improvements

    SciTech Connect (OSTI)

    Tzanos, C.P.; Tessier, J.H.; Pedersen, D.R. (Argonne National Laboratory, IL (USA))

    1989-11-01T23:59:59.000Z

    One of the main safety features of the current liquid-metal reactor (LMR) designs is the utilization of decay heat removal systems that remove heat by natural convection. In the reactor vessel auxiliary cooling system (RVACS), decay heat is removed by naturally circulating air in the gap between the guard vessel and a baffle wall surrounding the guard vessel. The objective of this work was to determine the impact of a number of design parameters on the performance of the RVACS of a pool LMR. These parameters were (a) the stack height, (b) the size of the airflow gap, (c) the system pressure loss, (d) fins on the guard vessel or the baffle wall, and (e) roughness (in the form of repeated ribs) on the airflow channel walls. Reactor designs ranging from 400 to 3,500 MW(thermal) were considered. From the RVACS design parameters considered in this analysis, an optimized ribbed configuration gave the best improvement in RVACS performance. For a 3,500-MW(thermal) LMR, the peak sodium and cladding temperatures were reduced by 52 K.

  5. High Gain Observer for Backstepping Control of a MRI-guided Therapeutic Microrobot in Blood Vessels

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    -planned trajectory inherited from the model, with robustness concerns. This paper reports modeling and controlHigh Gain Observer for Backstepping Control of a MRI-guided Therapeutic Microrobot in Blood Vessels Laurent Arcese, Ali Cherry, Matthieu Fruchard, Antoine Ferreira Abstract-- This paper reports modeling

  6. Stingray / Skate / Angel Shark Species Description Observer code:________________________ Vessel Code: ________________ Trip ID: _______________

    E-Print Network [OSTI]

    spines How many? Pelvic fin lobes Dorsal fins Pairs rostral teeth Nearer to pelvic fin than tail tip Nearer to tail tip then pelvic fin None First dorsal fin (check one) VentralStingray / Skate / Angel Shark Species Description Observer code:________________________ Vessel

  7. NEWASH AND TECUMSETH: ANALYSIS OF TWO POST-WAR OF 1812 VESSELS ON THE GREAT LAKES

    E-Print Network [OSTI]

    Gordon, Leeanne E.

    2010-01-16T23:59:59.000Z

    In 1953 the tangled, skeletal remains of a ship were pulled from the small harbor of Penetanguishene, Ontario. Local historians had hoped to raise the hull of a War of 1812 veteran, but the vessel pulled from the depths did not meet the criteria...

  8. Designing A Pattern Stabilization Method Using Scleral Blood Vessels For Laser Eye Surgery

    E-Print Network [OSTI]

    Erdem, Erkut

    Designing A Pattern Stabilization Method Using Scleral Blood Vessels For Laser Eye Surgery Aydin,abc}@cs.hacettepe.edu.tr, hbcakmak@gmail.com Abstract-- In laser eye surgery, the accuracy of operation depends on coherent eye tracking and registration techniques. Main approach used in image processing based eye trackers

  9. Walking and Climbing Service Robots for Safety Inspection of Nuclear Reactor Pressure Vessels

    E-Print Network [OSTI]

    Chen, Sheng

    Walking and Climbing Service Robots for Safety Inspection of Nuclear Reactor Pressure Vessels B of Electronics and Computer Science, University of Southampton, Southampton, UK Abstract: Nuclear reactor and the usefulness of these robots for improving safety inspection of nuclear reactors in general are discussed

  10. A Liquid Xenon Ionization Chamber in an All-fluoropolymer Vessel

    SciTech Connect (OSTI)

    LePort, F.; Pocar, A.; Bartoszek, L.; DeVoe, R.; Fierlinger, P.; Flatt, B.; Gratta, G.; Green, M.; Montero Diez, M.; Neilson, R.; O'Sullivan, K.; Wodin, J.; Woisard, D.; Baussan, E.; Breidenbach, M.; Conley, R.; Fairbank, W., Jr.; Farine, J.; Hall, K.; Hallman, D.; Hargrove, C.; /Stanford U., Phys. Dept. /Applied Plastics Technology, Bristol/Neuchatel U. /SLAC /Colorado State U. /Laurentian U. /Carleton U. /Alabama U. /Moscow, ITEP; ,

    2007-02-26T23:59:59.000Z

    A novel technique has been developed to build vessels for liquid xenon ionization detectors entirely out of ultra-clean fluoropolymer. We describe the advantages in terms of low radioactivity contamination, provide some details of the construction techniques, and show the energy resolution achieved with a prototype all-fluoropolymer ionization detector.

  11. The strategic allocation of cyclically calling vessels for multi-terminal container operators

    E-Print Network [OSTI]

    Armbruster, Dieter

    The strategic allocation of cyclically calling vessels for multi-terminal container operators M.P.M. Hendriks , D. Armbruster, , M. Laumanns , E. Lefeber , J.T. Udding Abstract We consider a terminal operator, who provides his logistics services of container handling at multiple terminals within the same port

  12. First International Symposium on Fishing Vessel Energy Efficiency E-Fishing, Vigo, Spain, May 2010

    E-Print Network [OSTI]

    Lewandowski, Roger

    First International Symposium on Fishing Vessel Energy Efficiency E-Fishing, Vigo, Spain, May 2010 HydroPêche: a way to improve energy efficiency of fishing devices Grégory Germain 1 , Philippe Druault 2 should provide a substantial gain on the fuel consumed of actual fishing devices while maintaining

  13. DOE H2 Program Annual Review, 5-20-2003 Insulated Pressure Vessels for

    E-Print Network [OSTI]

    range. J. We are generating tank performance data. K. Testing BOP components. L. Low venting losses) car, km 0 1 2 3 4 5 hydrogenlosses,kg low-pressure LH2 tank MLVSI insulated pressure vessel fueled with LH2 LH2 80 K CH2 1998: thermodynamic analysis 1999: cryogenic cycling 2001: DOT/ISO Tests 2003

  14. J. Fluid Mech. (in press) 1 Shallow-water sloshing in vessels undergoing

    E-Print Network [OSTI]

    Bridges, Tom

    J. Fluid Mech. (in press) 1 Shallow-water sloshing in vessels undergoing prescribed rigid the predominant types of solution are the standing wave and travelling hydraulic jump. But in 3D shallow-dimensional hydraulic jumps and analytical methods are very effective for identifying parameter regimes for these basic

  15. ITER vacuum vessel fabrication plan and cost study (D 68) for the international thermonuclear experimental reactor

    SciTech Connect (OSTI)

    NONE

    1995-01-01T23:59:59.000Z

    ITER Task No. 8, Vacuum Vessel Fabrication Plan and Cost Study (D68), was initiated to assess ITER vacuum vessel fabrication, assembly, and cost. The industrial team of Raytheon Engineers & Constructors and Chicago Bridge & Iron (Raytheon/CB&I) reviewed the current vessel basis and prepared a manufacturing plan, assembly plan, and cost estimate commensurate with the present design. The guidance for the Raytheon/CB&I assessment activities was prepared by the ITER Garching Work Site. This guidance provided in the form of work descriptions, sketches, drawings, and costing guidelines for each of the presently identified vacuum vessel Work Breakdown Structure (WBS) elements was compiled in ITER Garching Joint Work Site Memo (Draft No. 9 - G 15 MD 01 94-17-05 W 1). A copy of this document is provided as Appendix 1 to this report. Additional information and clarifications required for the Raytheon/CB&I assessments were coordinated through the US Home Team (USHT) and its technical representative. Design details considered essential to the Task 8 assessments but not available from the ITER Joint Central Team (JCT) were generated by Raytheon/CB&I and documented accordingly.

  16. A Synthesis of Cost and Revenue Surveys for Gulf of Mexico Shrimp Vessels

    E-Print Network [OSTI]

    series cost data are not routinely collected for vessels operat ing in any of the U.S. southeast region of Published Cost and Revenue Data Differences underlying the cost and revenue surveys prevent direct compari in the survey data. For example, changes in the cost an

  17. Certification Testing and Demonstration of Insulated Pressure Vessels for Vehicular Hydrogen Storage

    E-Print Network [OSTI]

    characteristics as liquid hydrogen tanks (low weight and volume), with reduced energy consumption for liquefaction Ave., L-644, Livermore, CA 94551, USA, saceves@llnl.gov Abstract Insulated pressure vessels of electric vehicles to improve environmental quality and energy security, while providing the range

  18. 1 Copyright 2012 by ASME Proceedings of the ASME 2012 Pressure Vessels & Piping Division Conference

    E-Print Network [OSTI]

    Tijsseling, A.S.

    1 Copyright © 2012 by ASME Proceedings of the ASME 2012 Pressure Vessels & Piping Division.P. Andersens veg 7 N-7465 Trondheim Norway E-mail: bjoernar.svingen@rainpower.no Anton BERGANT Litostroj Power = 50 m; inner diameters D1 = 1 m and D2 = 0.2 m [1]. #12;2 Copyright © 2012 by ASME Figure 2. Technical

  19. Tritium Reduction and Control in the Vacuum Vessel During TFTR Outage and Decommissioning *

    E-Print Network [OSTI]

    Tritium Reduction and Control in the Vacuum Vessel During TFTR Outage and Decommissioning * W of the torus, a three tier system was developed for the outage in order to reduce and control the free tritium. The first phase of the program to reduce the free tritium consisted of direct flowthrough of room air

  20. Tritium Reduction and Control in the Vacuum Vessel During TFTR Outage and Decommissioning*

    E-Print Network [OSTI]

    Tritium Reduction and Control in the Vacuum Vessel During TFTR Outage and Decommissioning* W, a three tier system was developed for the outage in order to reduce and control the free tritium. The first phase of the program to reduce the free tritium consisted of direct flowthrough of room air

  1. CALCULATIONS OF TRITIUM FLOW BETWEEN THE BUFFER VESSEL UP TO THE FIRST VACUUM SYSTEM

    E-Print Network [OSTI]

    Sharipov, Felix

    CALCULATIONS OF TRITIUM FLOW BETWEEN THE BUFFER VESSEL UP TO THE FIRST VACUUM SYSTEM Felix Sharipov diff., Eq.(32) µ viscosity of tritium Pa s 1 Introduction The present work is a continuation of the previous report [1], where the preliminary results were obtained for the tritium flow through the source

  2. Mechanical Recanalization of Subacute Vessel Occlusion in Peripheral Arterial Disease with a Directional Atherectomy Catheter

    SciTech Connect (OSTI)

    Massmann, Alexander, E-mail: Alexander.Massmann@uks.eu; Katoh, Marcus [Saarland University Hospital, Department of Diagnostic and Interventional Radiology (Germany); Shayesteh-Kheslat, Roushanak [Saarland University Hospital, Department of General Surgery, Visceral, Vascular, and Pediatric Surgery (Germany); Buecker, Arno [Saarland University Hospital, Department of Diagnostic and Interventional Radiology (Germany)

    2012-10-15T23:59:59.000Z

    Purpose: To retrospectively examine the technical feasibility and safety of directional atherectomy for treatment of subacute infrainguinal arterial vessel occlusions. Methods: Five patients (one woman, four men, age range 51-81 years) with peripheral arterial disease who experienced sudden worsening of their peripheral arterial disease-related symptoms during the last 2-6 weeks underwent digital subtraction angiography, which revealed vessel occlusion in native popliteal artery (n = 4) and in-stent occlusion of the superficial femoral artery (n = 1). Subsequently, all patients were treated by atherectomy with the SilverHawk (ev3 Endovascular, USA) device. Results: The mean diameter of treated vessels was 5.1 {+-} 1.0 mm. The length of the occlusion ranged 2-14 cm. The primary technical success rate was 100%. One patient experienced a reocclusion during hospitalization due to heparin-induced thrombocytopenia. There were no further periprocedural complications, in particular no peripheral embolizations, until hospital discharge or during the follow-up period of 1 year. Conclusion: The recanalization of infrainguinal arterial vessel occlusions by atherectomy with the SilverHawk device is technically feasible and safe. In our limited retrospective study, it was associated with a high technical success rate and a low procedure-related complication rate.

  3. In-vessel Retention Strategy for High Power Reactors - K-INERI Final Report (includes SBLB Test Results for Task 3 on External Reactor Vessel Cooling (ERVC) Boiling Data and CHF Enhancement Correlations)

    SciTech Connect (OSTI)

    F. B. Cheung; J. Yang; M. B. Dizon; J. Rempe

    2005-01-01T23:59:59.000Z

    In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe PWR (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing LWRs. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a collaborative, three-year, U.S. - Korean International Nuclear Energy Research Initiative (INERI) project was completed in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) investigated the performance of ERVC and an in-vessel core catcher (IVCC) to determine if IVR is feasible for reactors up to 1500 MWe.

  4. ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING DEACTIVATION AND DECOMMISSIONING OF REACTOR VESSELS AT THE SAVANNAH RIVER SITE

    SciTech Connect (OSTI)

    Wiersma, B.; Serrato, M.; Langton, C.

    2010-11-10T23:59:59.000Z

    The R- and P-reactor vessels at the Savannah River Site (SRS) are being prepared for deactivation and decommissioning (D&D). D&D activities will consist primarily of physically isolating and stabilizing the reactor vessel by filling it with a grout material. The reactor vessels contain aluminum alloy materials, which pose a concern in that aluminum corrodes rapidly when it comes in contact with the alkaline grout. A product of the corrosion reaction is hydrogen gas and therefore potential flammability issues were assessed. A model was developed to calculate the hydrogen generation rate as the reactor is being filled with the grout material. Three options existed for the type of grout material for D&D of the reactor vessels. The grout formulation options included ceramicrete (pH 6-8), a calcium aluminate sulfate (CAS) based cement (pH 10), or Portland cement grout (pH 12.4). Corrosion data for aluminum in concrete were utilized as input for the model. The calculations considered such factors as the surface area of the aluminum components, the open cross-sectional area of the reactor vessel, the rate at which the grout is added to the reactor vessel, and temperature. Given the hydrogen generation rate, the hydrogen concentration in the vapor space of the reactor vessel above the grout was calculated. This concentration was compared to the lower flammability limit for hydrogen. The assessment concluded that either ceramicrete or the CAS grout may be used to safely grout the P-reactor vessel. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Portland cement grout, on the other hand, for the same range of process parameters did not provide a margin of safety against the accumulation of flammable gas in the reactor vessel during grouting operations in the P-reactor vessel. Therefore, it was recommended that this grout not be utilized for this task. On the other hand, the R-reactor vessel contained significantly less aluminum surface area that the P-reactor vessel based on current facility process knowledge, surface observations, and drawings. Therefore, a Portland cement grout may be considered for grouting operations as well as the other grout formulations.

  5. Weld Repair of a Stamped Pressure Vessel in a Radiologically Controlled Zone

    SciTech Connect (OSTI)

    Cannell, Gary L. [Fluor Enterprises, Inc.; Huth, Ralph J. [CH2MHill Plateau Remediation Company; Hallum, Randall T. [Fluor Government Group

    2013-08-26T23:59:59.000Z

    In September 2012 an ASME B&PVC Section VIII stamped pressure vessel located at the DOE Hanford Site Effluent Treatment Facility (ETF) developed a through-wall leak. The vessel, a steam/brine heat exchanger, operated in a radiologically controlled zone (by the CH2MHill PRC or CHPRC), had been in service for approximately 17 years. The heat exchanger is part of a single train evaporator process and its failure caused the entire system to be shut down, significantly impacting facility operations. This paper describes the activities associated with failure characterization, technical decision making/planning for repair by welding, logistical challenges associated with performing work in a radiologically controlled zone, performing the repair, and administrative considerations related to ASME code requirements.

  6. Use of Polycarbonate Vacuum Vessels in High-Temperature Fusion-Plasma Research

    SciTech Connect (OSTI)

    B. Berlinger, A. Brooks, H. Feder, J. Gumbas, T. Franckowiak and S.A. Cohen

    2012-09-27T23:59:59.000Z

    Magnetic fusion energy (MFE) research requires ultrahigh-vacuum (UHV) conditions, primarily to reduce plasma contamination by impurities. For radiofrequency (RF)-heated plasmas, a great benefit may accrue from a non-conducting vacuum vessel, allowing external RF antennas which avoids the complications and cost of internal antennas and high-voltage high-current feedthroughs. In this paper we describe these and other criteria, e.g., safety, availability, design flexibility, structural integrity, access, outgassing, transparency, and fabrication techniques that led to the selection and use of 25.4-cm OD, 1.6-cm wall polycarbonate pipe as the main vacuum vessel for an MFE research device whose plasmas are expected to reach keV energies for durations exceeding 0.1 s

  7. Standard Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E 706 (IH)

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2010-01-01T23:59:59.000Z

    1.1 This guide discusses test procedures that can be used in conjunction with, but not as alternatives to, those required by Practices E185 and E2215 for the surveillance of nuclear reactor vessels. The supplemental mechanical property tests outlined permit the acquisition of additional information on radiation-induced changes in fracture toughness, notch ductility, and yield strength properties of the reactor vessel steels. 1.2 This guide provides recommendations for the preparation of test specimens for irradiation, and identifies special precautions and requirements for reactor surveillance operations and postirradiation test planning. Guidance on data reduction and computational procedures is also given. Reference is made to other ASTM test methods for the physical conduct of specimen tests and for raw data acquisition.

  8. Calculational criticality analyses of 10- and 20-MW UF[sub 6] freezer/sublimer vessels

    SciTech Connect (OSTI)

    Jordan, W.C.

    1993-02-01T23:59:59.000Z

    Calculational criticality analyses have been performed for 10- and 20-MW UF[sub 6] freezer/sublimer vessels. The freezer/sublimers have been analyzed over a range of conditions that encompass normal operation and abnormal conditions. The effects of HF moderation of the UF[sub 6] in each vessel have been considered for uranium enriched between 2 and 5 wt % [sup 235]U. The results indicate that the nuclearly safe enrichments originally established for the operation of a 10-MW freezer/sublimer, based on a hydrogen-to-uranium moderation ratio of 0.33, are acceptable. If strict moderation control can be demonstrated for hydrogen-to-uranium moderation ratios that are less than 0.33, then the enrichment limits for the 10-MW freezer/sublimer may be increased slightly. The calculations performed also allow safe enrichment limits to be established for a 20-NM freezer/sublimer under moderation control.

  9. Calculational criticality analyses of 10- and 20-MW UF{sub 6} freezer/sublimer vessels

    SciTech Connect (OSTI)

    Jordan, W.C.

    1993-02-01T23:59:59.000Z

    Calculational criticality analyses have been performed for 10- and 20-MW UF{sub 6} freezer/sublimer vessels. The freezer/sublimers have been analyzed over a range of conditions that encompass normal operation and abnormal conditions. The effects of HF moderation of the UF{sub 6} in each vessel have been considered for uranium enriched between 2 and 5 wt % {sup 235}U. The results indicate that the nuclearly safe enrichments originally established for the operation of a 10-MW freezer/sublimer, based on a hydrogen-to-uranium moderation ratio of 0.33, are acceptable. If strict moderation control can be demonstrated for hydrogen-to-uranium moderation ratios that are less than 0.33, then the enrichment limits for the 10-MW freezer/sublimer may be increased slightly. The calculations performed also allow safe enrichment limits to be established for a 20-NM freezer/sublimer under moderation control.

  10. USING A CONTAINMENT VESSEL LIFTING APPARATUS FOR REMOTE OPERATIONS OF SHIPPING PACKAGES

    SciTech Connect (OSTI)

    Loftin, Bradley [Savannah River National Laboratory; Koenig, Richard [Savannah River National Laboratory

    2013-08-08T23:59:59.000Z

    The 9977 and the 9975 shipping packages are used in various nuclear facilities within the Department of Energy. These shipping packages are often loaded in designated areas with designs using overhead cranes or A-frames with lifting winches. However, there are cases where loading operations must be performed in remote locations where these facility infrastructures do not exist. For these locations, a lifting apparatus has been designed to lift the containment vessels partially out of the package for unloading operations to take place. Additionally, the apparatus allows for loading and closure of the containment vessel and subsequent pre-shipment testing. This paper will address the design of the apparatus and the challenges associated with the design, and it will describe the use of the apparatus.

  11. Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability

    DOE Patents [OSTI]

    Hunsbedt, A.; Boardman, C.E.

    1995-04-11T23:59:59.000Z

    A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor is disclosed. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo`s structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated. 5 figures.

  12. Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

    1995-01-01T23:59:59.000Z

    A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo's structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated.

  13. Vessel Cold-Ironing Using a Barge Mounted PEM Fuel Cell: Project Scoping and Feasibility.

    SciTech Connect (OSTI)

    Pratt, Joseph William; Harris, Aaron P

    2013-01-01T23:59:59.000Z

    A barge-mounted hydrogen-fueled proton exchange membrane (PEM) fuel cell system has the potential to reduce emissions and fossil fuel use of maritime vessels in and around ports. This study determines the technical feasibility of this concept and examines specific options on the U.S. West Coast for deployment practicality and potential for commercialization.The conceptual design of the system is found to be straightforward and technically feasible in several configurations corresponding to various power levels and run times.The most technically viable and commercially attractive deployment options were found to be powering container ships at berth at the Port of Tacoma and/or Seattle, powering tugs at anchorage near the Port of Oakland, and powering refrigerated containers on-board Hawaiian inter-island transport barges. Other attractive demonstration options were found at the Port of Seattle, the Suisun Bay Reserve Fleet, the California Maritime Academy, and an excursion vessel on the Ohio River.

  14. Analysis of hydrodynamic phenomena in simulant experiments investigating cavity interactions following postulated vessel meltthrough

    SciTech Connect (OSTI)

    Sienicki, J.J.; Spencer, B.W.

    1984-01-01T23:59:59.000Z

    An analysis of hydrodynamic phenomena in simulant experiments examining aspects of ex-vessel material interactions in a PWR reactor cavity following postulated core meltdown and localized breaching of the reactor vessel has been carried out. While previous analyses of the tests examined thresholds for the onset of sweepout of fluid from the cavity, the present analysis considers the progression of specific hydrodynamic phenomena involved in the dispersal process: crater formation due to gas jet impingement, radial wave motion and growth, entrainment and transport of liquid droplets, liquid layer formation due to droplet recombination, fluidization of liquid remaining in the cavity, removal of fluidized liquid droplets from the cavity, and the ultimate removal of the remaining liquid layer within the tunnel passageway. Phenomenological models which may be used to predict the phenomena are presented.

  15. Dosimetry assessments for the reactor pressure vessel and core barrel in UK PWR plant

    SciTech Connect (OSTI)

    Thornton, D.A.; Allen, D.A.; Huggon, A.P.; Picton, D.J.; Robinson, A.T.; Steadman, R.J. [Serco, Rutherford House, Quedgeley, Gloucester, Gl2 4NF (United Kingdom); Seren, T.; Lipponen, M.; Kekki, T. [VTT, Technical Research Centre of Finland, Otakaari 3 K, P.O. BOX 1000, Espoo, FI-02044 (Finland)

    2011-07-01T23:59:59.000Z

    Specimens for the Sizewell B reactor pressure vessel (RPV) inservice steels surveillance program are irradiated inside eight capsules located within the reactor pressure vessel and loaded prior to commissioning. The periodic removal of these capsules and testing of their contents provides material properties data at intervals during the lifetime of the plant. Neutron activation measurements and radiation transport calculations play an essential role in assessing the neutron exposure of the specimens and RPV. Following the most recent withdrawal, seven capsules have now been removed covering nine cycles of reactor operation. This paper summarizes the dosimetry results of the Sizewell B surveillance program obtained to date. In addition to an overview of the calculational methodology it includes a review of the measurements. Finally, it describes an extension of the methodology to provide dosimetry recommendations for the core barrel and briefly discusses the results that were obtained. (authors)

  16. Simulations of Blood Flow in Plain Cylindrical and Constricted Vessels with Single Cell Resolution

    E-Print Network [OSTI]

    Florian Janoschek; Federico Toschi; Jens Harting

    2011-05-31T23:59:59.000Z

    Understanding the physics of blood is challenging due to its nature as a suspension of soft particles and the fact that typical problems involve different scales. This is valid also for numerical investigations. In fact, many computational studies either neglect the existence of discrete cells or resolve relatively few cells very accurately. The authors recently developed a simple and highly efficient yet still particulate model with the aim to bridge the gap between currently applied methods. The present work focuses on its applicability to confined flows in vessels of diameters up to 100 micrometres. For hematocrit values below 30 percent, a dependence of the apparent viscosity on the vessel diameter in agreement with experimental literature data is found.

  17. Peristaltic Pumping of Blood Through Small Vessels of Varying Cross-section

    E-Print Network [OSTI]

    J. C. Misra; S. Maiti

    2012-01-30T23:59:59.000Z

    The paper is devoted to a study of the peristaltic motion of blood in the micro-circulatory system. The vessel is considered to be of varying cross-section. The progressive peristaltic waves are taken to be of sinusoidal nature. Blood is considered to be a Herschel-Bulkley fluid. Of particular concern here is to investigate the effects of amplitude ratio, mean pressure gradient, yield stress and the power law index on the velocity distribution, streamline pattern and wall shear stress. On the basis of the derived analytical expression, extensive numerical calculations have been made. The study reveals that velocity of blood and wall shear stress are appreciably affected due to the non-uniform geometry of blood vessels. They are also highly sensitive to the magnitude of the amplitude ratio and the value of the fluid index.

  18. Pressure vessel sliding support unit and system using the sliding support unit

    DOE Patents [OSTI]

    Breach, Michael R.; Keck, David J.; Deaver, Gerald A.

    2013-01-15T23:59:59.000Z

    Provided is a sliding support and a system using the sliding support unit. The sliding support unit may include a fulcrum capture configured to attach to a support flange, a fulcrum support configured to attach to the fulcrum capture, and a baseplate block configured to support the fulcrum support. The system using the sliding support unit may include a pressure vessel, a pedestal bracket, and a plurality of sliding support units.

  19. Analysis of the ANL Test Method for 6CVS Containment Vessels

    SciTech Connect (OSTI)

    Trapp, D.; Crow, G.

    2011-06-06T23:59:59.000Z

    In the fall of 2010, Argonne National Laboratory (ANL) contracted with vendors to design and build 6CVS containment vessels as part of their effort to ship Fuel Derived Mixed Fission Product material. The 6CVS design is based on the Savannah River National Laboratory's (SRNL) design for 9975 and 9977 six inch diameter containment vessels. The main difference between the designs is that the 6CVS credits the inner O-ring seal as the containment boundary while the SRNL design credits the outer O-ring seal. Since the leak test must be done with the inner O-ring in place, the containment vessel does not have a pathway for getting the helium into the vessel during the leak test. The leak testing contractor was not able to get acceptable leak rates with the specified O-ring, but they were able to pass the leak test with a slightly larger O-ring. ANL asked the SRNL to duplicate the leak test vendor's method to determine the cause of the high leak rates. The SRNL testing showed that the helium leak indications were caused by residual helium left within the 6CVS Closure Assembly by the leak test technique, and by helium permeation through the Viton O-ring seals. After SRNL completed their tests, the leak testing contractor was able to measure acceptable leak rates by using the slightly larger O-ring size, by purging helium from the lid threads, and by being very quick in getting the bell jar under a full vacuum. This paper describes the leak test vendor's test technique, and other techniques that could be have been used to successfully leak test the 6CVS's.

  20. Application of Negligible Creep Criteria to Candidate Materials for HTGR Pressure Vessels

    SciTech Connect (OSTI)

    Jetter, Robert I [Consultant; Sham, Sam [ORNL; Swindeman, Robert W [Consultant

    2011-01-01T23:59:59.000Z

    Two of the proposed High Temperature Gas Reactors (HTGRs) under consideration for a demonstration plant have the design object of avoiding creep effects in the reactor pressure vessel (RPV) during normal operation. This work addresses the criteria for negligible creep in Subsection NH, Division 1 of the ASME B&PV (Boiler and Pressure Vessel) Code, Section III, other international design codes and some currently suggested criteria modifications and their impact on permissible operating temperatures for various reactor pressure vessel materials. The goal of negligible creep could have different interpretations depending upon what failure modes are considered and associated criteria for avoiding the effects of creep. It is shown that for the materials of this study, consideration of localized damage due to cycling of peak stresses results in a lower temperature for negligible creep than consideration of the temperature at which the allowable stress is governed by creep properties. In assessing the effect of localized cyclic stresses it is also shown that consideration of cyclic softening is an important effect that results in a higher estimated temperature for the onset of significant creep effects than would be the case if the material were cyclically hardening. There are other considerations for the selection of vessel material besides avoiding creep effects. Of interest for this review are (1) the material s allowable stress level and impact on wall thickness (the goal being to minimize required wall thickness) and (2) ASME Code approval (inclusion as a permitted material in the relevant Section and Subsection of interest) to expedite regulatory review and approval. The application of negligible creep criteria to two of the candidate materials, SA533 and Mod 9Cr-1Mo (also referred to as Grade 91), and to a potential alternate, normalized and tempered 2 Cr-1Mo, is illustrated and the relative advantages and disadvantages of the materials are discussed.

  1. Decommissioning experience: One-piece removal and transport of a LWR pressure vessel and internals

    SciTech Connect (OSTI)

    Closs, J.W. [Northern States Power Co., Minneapolis, MN (United States)

    1993-12-31T23:59:59.000Z

    After a brief historical perspective, this document describes several key events which took place during the decommissioning of a commercial nuclear power plant. The scope of decommissioning work included: (a) the reactor building, the reactor vessel and the contents of the reactor building; (b) the fuel handling building and its contents; (c) the fuel transfer vault between the reactor building and the fuel handling building.

  2. Standard Practice for Determining NeutronExposures for Nuclear Reactor Vessel Support Structures

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2008-01-01T23:59:59.000Z

    1.1 This practice covers procedures for monitoring the neutron radiation exposures experienced by ferritic materials in nuclear reactor vessel support structures located in the vicinity of the active core. This practice includes guidelines for: 1.1.1 Selecting appropriate dosimetric sensor sets and their proper installation in reactor cavities. 1.1.2 Making appropriate neutronics calculations to predict neutron radiation exposures. 1.2 This practice is applicable to all pressurized water reactors whose vessel supports will experience a lifetime neutron fluence (E > 1 MeV) that exceeds 1 × 1017 neutrons/cm2 or 3.0 × 10?4 dpa. (See Terminology E 170.) 1.3 Exposure of vessel support structures by gamma radiation is not included in the scope of this practice, but see the brief discussion of this issue in 3.2. 1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and h...

  3. Fabrication Flaw Density and Distribution In Repairs to Reactor Pressure Vessel and Piping Welds

    SciTech Connect (OSTI)

    GJ Schuster, FA Simonen, SR Doctor

    2008-04-01T23:59:59.000Z

    The Pacific Northwest National Laboratory is developing a generalized fabrication flaw distribution for the population of nuclear reactor pressure vessels and for piping welds in U.S. operating reactors. The purpose of the generalized flaw distribution is to predict component-specific flaw densities. The estimates of fabrication flaws are intended for use in fracture mechanics structural integrity assessments. Structural integrity assessments, such as estimating the frequency of loss-of-coolant accidents, are performed by computer codes that require, as input, accurate estimates of flaw densities. Welds from four different reactor pressure vessels and a collection of archived pipes have been studied to develop empirical estimates of fabrication flaw densities. This report describes the fabrication flaw distribution and characterization in the repair weld metal of vessels and piping. This work indicates that large flaws occur in these repairs. These results show that repair flaws are complex in composition and sometimes include cracks on the ends of the repair cavities. Parametric analysis using an exponential fit is performed on the data. The relevance of construction records is established for describing fabrication processes and product forms. An analysis of these records shows there was a significant change in repair frequency over the years when these components were fabricated. A description of repair flaw morphology is provided with a discussion of fracture mechanics significance. Fabrication flaws in repairs are characterized using optimized-access, high-sensitivity nondestructive ultrasonic testing. Flaw characterizations are then validated by other nondestructive evaluation techniques and complemented by destructive testing.

  4. Predictive Reactor Pressure Vessel Steel Irradiation Embrittlement Models: Issues and Opportunities

    SciTech Connect (OSTI)

    Odette, George Robert [UCSB; Nanstad, Randy K [ORNL

    2009-01-01T23:59:59.000Z

    Nuclear plant life extension to 80 years will require accurate predictions of neutron irradiation-induced increases in the ductile-brittle transition temperature ( T) of reactor pressure vessel (RPV) steels at high fluence conditions that are far outside the existing database. Remarkable progress in mechanistic understanding of irradiation embrittlement has led to physically motivated T correlation models that provide excellent statistical fi ts to the existing surveillance database. However, an important challenge is developing advanced embrittlement models for low fl ux-high fl uence conditions pertinent to extended life. These new models must also provide better treatment of key variables and variable combinations and account for possible delayed formation of late blooming phases in low copper steels. Other issues include uncertainties in the compositions of actual vessel steels, methods to predict T attenuation away from the reactor core, verifi cation of the master curve method to directly measure the fracture toughness with small specimens and predicting T for vessel annealing remediation and re-irradiation cycles.

  5. A Multiscale Modeling Approach to Analyze Filament-Wound Composite Pressure Vessels

    SciTech Connect (OSTI)

    Nguyen, Ba Nghiep; Simmons, Kevin L.

    2013-07-22T23:59:59.000Z

    A multiscale modeling approach to analyze filament-wound composite pressure vessels is developed in this article. The approach, which extends the Nguyen et al. model [J. Comp. Mater. 43 (2009) 217] developed for discontinuous fiber composites to continuous fiber ones, spans three modeling scales. The microscale considers the unidirectional elastic fibers embedded in an elastic-plastic matrix obeying the Ramberg-Osgood relation and J2 deformation theory of plasticity. The mesoscale behavior representing the composite lamina is obtained through an incremental Mori-Tanaka type model and the Eshelby equivalent inclusion method [Proc. Roy. Soc. Lond. A241 (1957) 376]. The implementation of the micro-meso constitutive relations in the ABAQUS® finite element package (via user subroutines) allows the analysis of a filament-wound composite pressure vessel (macroscale) to be performed. Failure of the composite lamina is predicted by a criterion that accounts for the strengths of the fibers and of the matrix as well as of their interface. The developed approach is demonstrated in the analysis of a filament-wound pressure vessel to study the effect of the lamina thickness on the burst pressure. The predictions are favorably compared to the numerical and experimental results by Lifshitz and Dayan [Comp. Struct. 32 (1995) 313].

  6. The Louisiana State Museum Vessel: a historical and archaeological analysis of an American Civil War-era submersible boat

    E-Print Network [OSTI]

    Wills, Richard Keith

    2000-01-01T23:59:59.000Z

    During the spring of 1992, and again in the winter of 1993, seven graduate students from Texas A&M University's Nautical Archaeology Program participated in a project to document the Louisiana State Museum Vessel, an American Civil War...

  7. Avicennia germinans (black mangrove) vessel architecture is linked to chilling and salinity tolerance in the Gulf of Mexico

    E-Print Network [OSTI]

    Madrid, Eric N.; Armitage, Anna R.; Lopez-Portillo, Jorge

    2014-09-26T23:59:59.000Z

    Over the last several decades, the distribution of the black mangrove Avicennia germinans in the Gulf of Mexico has expanded, in part because it can survive the occasional freeze events and high soil salinities characteristic of the area. Vessel...

  8. Biomechanics of North Atlantic right whale bone : mandibular fracture as a fatal endpoint for blunt vessel-whale collision modeling

    E-Print Network [OSTI]

    Campbell-Malone, Regina P

    2007-01-01T23:59:59.000Z

    The North Atlantic right whale, Eubalaena glacialis, one of the most critically endangered whales in the world, is subject to high anthropogenic mortality. Vessel-whale collisions and entanglement in fishing gear were ...

  9. The Louisiana State Museum Vessel: a historical and archaeological analysis of an American Civil War-era submersible boat 

    E-Print Network [OSTI]

    Wills, Richard Keith

    2000-01-01T23:59:59.000Z

    During the spring of 1992, and again in the winter of 1993, seven graduate students from Texas A&M University's Nautical Archaeology Program participated in a project to document the Louisiana State Museum Vessel, an ...

  10. Seismic analysis of the Mirror Fusion Test Facility: soil structure interaction analyses of the Axicell vacuum vessel. Revision 1

    SciTech Connect (OSTI)

    Maslenikov, O.R.; Mraz, M.J.; Johnson, J.J.

    1986-03-01T23:59:59.000Z

    This report documents the seismic analyses performed by SMA for the MFTF-B Axicell vacuum vessel. In the course of this study we performed response spectrum analyses, CLASSI fixed-base analyses, and SSI analyses that included interaction effects between the vessel and vault. The response spectrum analysis served to benchmark certain modeling differences between the LLNL and SMA versions of the vessel model. The fixed-base analysis benchmarked the differences between analysis techniques. The SSI analyses provided our best estimate of vessel response to the postulated seismic excitation for the MFTF-B facility, and included consideration of uncertainties in soil properties by calculating response for a range of soil shear moduli. Our results are presented in this report as tables of comparisons of specific member forces from our analyses and the analyses performed by LLNL. Also presented are tables of maximum accelerations and relative displacements and plots of response spectra at various selected locations.

  11. Investigation of the use of nanofluids to enhance the In-Vessel Retention capabilities of Advanced Light Water Reactors

    E-Print Network [OSTI]

    Hannink, Ryan Christopher

    2007-01-01T23:59:59.000Z

    Nanofluids at very low concentrations experimentally exhibit a substantial increase in Critical Heat Flux (CHF) compared to water. The use of a nanofluid in the In-Vessel Retention (IVR) severe accident management strategy, ...

  12. Analysis of Mass Flow and Enhanced Mass Flow Methods of Flashing Refrigerant-22 from a Small Vessel

    E-Print Network [OSTI]

    Nutter, Darin Wayne

    The mass flow characteristics of flashing Refrigerant-22 from a small vessel were investigated. A flash boiling apparatus was designed and built. It was modeled after the flashing process encountered by the accumulator of air-source heat pump...

  13. Virtual Rogowskii loops for the evaluation of plasma and in-vessel eddy currents in the KSTAR tokamak

    SciTech Connect (OSTI)

    Tsaun, S.; Jhang, Hogun [RRC, Kurchatov Institute, Kurchartov Square 1, Moscow 123181 (Russian Federation); Korea Basic Science Institute, 52 Yeoeun-dong, Yusung-Gu, Daejon 305-333 (Korea, Republic of)

    2006-01-15T23:59:59.000Z

    A simple but effective method is proposed for the estimation of plasma and in-vessel eddy currents in the KSTAR device. It is based on the prescribed contour integrals of the tangential magnetic fields measured by magnetic probes. It is shown that the method can estimate the plasma current in 0.33% accuracy. It is also found that the total in-vessel eddy current can be determined with accuracy within a few kiloamperes regardless of plasma conditions.

  14. Structural integrity assessment of carbon and low-alloy steel pressure vessels using a simplified fracture mechanics procedure

    SciTech Connect (OSTI)

    Rana, M.D. (Praxair Inc., Tonawanda, NY (United States). Research and Development Dept.)

    1994-08-01T23:59:59.000Z

    This paper describes a simplified fracture analysis procedure which was developed by Pellini to quantify fracture critical-crack sizes and crack-arrest temperatures of carbon and low-alloy steel pressure vessels. Fracture analysis diagrams have been developed using the simplified analysis procedure for various grades of carbon and low-alloy steels used in the construction of ASME, Section VIII, Division 1 pressure vessels. Structural integrity assessments have been conducted from the analysis diagrams.

  15. Evidence for neutron irradiation-induced metallic precipitates in model alloys and pressure-vessel weld steel

    E-Print Network [OSTI]

    Motta, Arthur T.

    -vessel weld steel Stephen E. Cumblidge a , Arthur T. Motta a,*, Gary L. Catchen a , Gerhard Brauer b , Juurgen-irradiated model alloys (1 · 1023 n/m2 , E > 0:5 MeV) and 73W-weld steel (to 1.8 · 1023 n/m2 , E > 1 Me the pressure-vessel weld steel) showed evidence for both irradiation-induced metallic precipitation

  16. Manufacturing Cost Analysis of Novel Steel/Concrete Composite Vessel for Stationary Storage of High-Pressure Hydrogen

    SciTech Connect (OSTI)

    Feng, Zhili [ORNL; Zhang, Wei [ORNL; Wang, Jy-An John [ORNL; Ren, Fei [ORNL

    2012-09-01T23:59:59.000Z

    A novel, low-cost, high-pressure, steel/concrete composite vessel (SCCV) technology for stationary storage of compressed gaseous hydrogen (CGH2) is currently under development at Oak Ridge National Laboratory (ORNL) sponsored by DOE s Fuel Cell Technologies (FCT) Program. The SCCV technology uses commodity materials including structural steels and concretes for achieving cost, durability and safety requirements. In particular, the hydrogen embrittlement of high-strength low-alloy steels, a major safety and durability issue for current industry-standard pressure vessel technology, is mitigated through the use of a unique layered steel shell structure. This report presents the cost analysis results of the novel SCCV technology. A high-fidelity cost analysis tool is developed, based on a detailed, bottom-up approach which takes into account the material and labor costs involved in each of the vessel manufacturing steps. A thorough cost study is performed to understand the SCCV cost as a function of the key vessel design parameters, including hydrogen pressure, vessel dimensions, and load-carrying ratio. The major conclusions include: The SCCV technology can meet the technical/cost targets set forth by DOE s FCT Program for FY2015 and FY2020 for all three pressure levels (i.e., 160, 430 and 860 bar) relevant to the hydrogen production and delivery infrastructure. Further vessel cost reduction can benefit from the development of advanced vessel fabrication technologies such as the highly automated friction stir welding (FSW). The ORNL-patented multi-layer, multi-pass FSW can not only reduce the amount of labor needed for assembling and welding the layered steel vessel, but also make it possible to use even higher strength steels for further cost reductions and improvement of vessel structural integrity. It is noted the cost analysis results demonstrate the significant cost advantage attainable by the SCCV technology for different pressure levels when compared to the industry-standard pressure vessel technology. The real-world performance data of SCCV under actual operating conditions is imperative for this new technology to be adopted by the hydrogen industry for stationary storage of CGH2. Therefore, the key technology development effort in FY13 and subsequent years will be focused on the fabrication and testing of SCCV mock-ups. The static loading and fatigue data will be generated in rigorous testing of these mock-ups. Successful tests are crucial to enabling the near-term impact of the developed storage technology on the CGH2 storage market, a critical component of the hydrogen production and delivery infrastructure. In particular, the SCCV has high potential for widespread deployment in hydrogen fueling stations.

  17. CHARACTERIZATION OF RADIOACTIVITY IN THE REACTOR VESSEL OF THE HEAVY WATER COMPONENT TEST REACTOR

    SciTech Connect (OSTI)

    Vinson, Dennis

    2010-06-01T23:59:59.000Z

    The Heavy Water Component Test Reactor (HWCTR) facility is a pressurized heavy water reactor that was used to test candidate fuel designs for heavy water power reactors. The reactor operated at nominal power of 50 MW{sub th}. The reactor coolant loop operated at 1200 psig and 250 C. Two isolated test loop were designed into the reactor to provide special test conditions. Fig. 1 shows a cut-away view of the reactor. The two loops are contained in four inch diameter stainless steel piping. The HWCTR was operated for only a short duration, from March 1962 to December 1964 in order to test the viability of test fuel elements and other reactor components for use in a heavy water power reactor. The reactor achieved 13,882 MWd of total power while testing 36 different fuel assemblies. In the course of operation, HWCTR experienced the cladding failures of 10 separate test fuel assemblies. In each case, the cladding was breached with some release of fuel core material into the isolated test loop, causing fission product and actinide contamination in the main coolant loop and the liquid and boiling test loops. Despite the contribution of the contamination from the failed fuel, the primary source of radioactivity in the HWCTR vessel and internals is the activation products in the thermal shields, and to a lesser degree, activation products in the reactor vessel walls and liner. A detailed facility characterization report of the HWCTR facility was completed in 1996. Many of the inputs and assumptions in the 1996 characterization report were derived from the HWCTR decommissioning plan published in 1975. The current paper provides an updated assessment of the radioisotopic characteristics of the HWCTR vessel and internals to support decommissioning activities on the facility.

  18. Qualitative Reliability Issues for In-Vessel Solid and Liquid Wall Fusion Designs

    SciTech Connect (OSTI)

    Cadwallader, Lee Charles; Nygren, R. E.

    2001-10-01T23:59:59.000Z

    This paper presents the results of a study of the qualitative aspects of plasma facing component (PFC) reliability for actively cooled solid wall and liquid wall concepts for magnetic fusion reactor vessels. These two designs have been analyzed for component failure modes. The most important results of that study are given here. A brief discussion of reliability growth in design is included to illustrate how solid wall designs have begun as workable designs and have evolved over time to become more optimized designs with better longevity. The increase in tolerable heat fluxes shows the improvement. Liquid walls could also have reliability growth if the designs had similar development efforts.

  19. Vessel Eddy Current Measurement for the National Spherical Torus Experiment (NSTX)

    SciTech Connect (OSTI)

    D.A. Gates; J. Menard; R. Marsala

    2004-11-19T23:59:59.000Z

    A simple analog circuit that measures the NSTX axisymmetric eddy current distribution has been designed and constructed. It is based on simple circuit model of the NSTX vacuum vessel that was calibrated using a special axisymmetric eddy current code which was written so that accuracy was maintained in the vicinity of the current filaments. The measurement and the model have been benchmarked against data from numerous vacuum shots and they are in excellent agreement. This is an important measurement that helps give more accurate equilibrium reconstructions.

  20. Dye laser amplifier including a dye cell contained within a support vessel

    DOE Patents [OSTI]

    Davin, J.

    1992-12-01T23:59:59.000Z

    A large (high flow rate) dye laser amplifier in which a continuous replenished supply of dye is excited by a first light beam, specifically a copper vapor laser beam, in order to amplify the intensity of a second different light beam, specifically a dye beam, passing through the dye is disclosed herein. This amplifier includes a dye cell defining a dye chamber through which a continuous stream of dye is caused to pass at a flow rate of greater than 30 gallons/minute at a static pressure greater than 150 pounds/square inch and a specifically designed support vessel for containing the dye cell. 6 figs.

  1. Spreading of molten corium in Mk I geometry following vessel meltthrough

    SciTech Connect (OSTI)

    Sienicki, J.J.; Farmer, M.T.; Spencer, B.W.

    1988-01-01T23:59:59.000Z

    A one-dimensional, multicell, Eulerian computer code is under development to predict the gravity-driven spreading dynamics and thermal interactions of a molten corium layer flowing horizontally over a concrete substrate. The code is compared to recent experiments in which molten mixtures of iron and aluminum oxide flowed over concrete in the presence and absence of water. Results are presented from scoping calculations for the Mk I BWR system investigating the spreading-induced penetration immediately following the drainage of a predominantly oxide molten corium mixture from a localized breach in the reactor vessel. 12 refs., 7 figs.

  2. LNG Imports by Vessel into the U.S. Form | Department of Energy

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn't YourTransport(FactDepartment ofLetter Report:40PM toLED Lighting5-15Trade |Vessel into the U.S.

  3. The criteria of fracture in the case of the leak of pressure vessels

    SciTech Connect (OSTI)

    Habil; Ziliukas, A.

    1997-04-01T23:59:59.000Z

    In order to forecast the break of the high pressure vessels and the network of pipes in a nuclear reactor, according to the concept of leak before break of pressure vessels, it is necessary to analyze the conditions of project, production, and mounting quality as well as of exploitation. It is also necessary to evaluate the process of break by the help of the fracture criteria. In the Ignalina Nuclear Power Plant of, in Lithuania, the most important objects of investigation are: the highest pressure pipes, made of Japanese steel 19MN5 and having an anticorrosive austenitic: coal inside, the pipes of distribution, which arc made of 08X1810T steel. The steel of the network of pipes has a quality of plasticity: therefore the only criteria of fragile is impossible to apply to. The process of break would be best described by the universal criteria of elastic - plastic fracture. For this purpose the author offers the criterion of the double parameter.

  4. In-vessel coolability and retention of a core melt. Volume 2

    SciTech Connect (OSTI)

    Theofanous, T.G.; Liu, C.; Additon, S.; Angelini, S.; Kymaelaeinen, O.; Salmassi, T. [California Univ., Santa Barbara, CA (United States). Center for Risk Studies and Safety

    1996-10-01T23:59:59.000Z

    The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the Risk Oriented Accident Analysis Methodology (ROAAM), and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow the delineation of the failure boundaries. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is physically unreasonable. Use of this conclusion for any specific application is subject to verifying the required reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, including any applicable coatings. The AP600 is particularly favorable to in-vessel retention. Some ideas to enhance the assessment basis as well as performance in this respect, for applications to larger and/or higher power density reactors are also provided.

  5. Calculation of Eddy Currents In the CTH Vacuum Vessel and Coil Frame

    SciTech Connect (OSTI)

    A. Zolfaghari, A. Brooks, A. Michaels, J. Hanson, and G. Hartwell

    2012-09-25T23:59:59.000Z

    Knowledge of eddy currents in the vacuum vessel walls and nearby conducting support structures can significantly contribute to the accuracy of Magnetohydrodynamics (MHD) equilibrium reconstruction in toroidal plasmas. Moreover, the magnetic fields produced by the eddy currents could generate error fields that may give rise to islands at rational surfaces or cause field lines to become chaotic. In the Compact Toroidal Hybrid (CTH) device (R0 = 0.75 m, a = 0.29 m, B ? 0.7 T), the primary driver of the eddy currents during the plasma discharge is the changing flux of the ohmic heating transformer. Electromagnetic simulations are used to calculate eddy current paths and profile in the vacuum vessel and in the coil frame pieces with known time dependent currents in the ohmic heating coils. MAXWELL and SPARK codes were used for the Electromagnetic modeling and simulation. MAXWELL code was used for detailed 3D finite-element analysis of the eddy currents in the structures. SPARK code was used to calculate the eddy currents in the structures as modeled with shell/surface elements, with each element representing a current loop. In both cases current filaments representing the eddy currents were prepared for input into VMEC code for MHD equilibrium reconstruction of the plasma discharge. __________________________________________________

  6. Standard practice for examination of seamless, Gas-Filled, pressure vessels using acoustic emission

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2009-01-01T23:59:59.000Z

    1.1 This practice provides guidelines for acoustic emission (AE) examinations of seamless pressure vessels (tubes) of the type used for distribution or storage of industrial gases. 1.2 This practice requires pressurization to a level greater than normal use. Pressurization medium may be gas or liquid. 1.3 This practice does not apply to vessels in cryogenic service. 1.4 The AE measurements are used to detect and locate emission sources. Other nondestructive test (NDT) methods must be used to evaluate the significance of AE sources. Procedures for other NDT techniques are beyond the scope of this practice. See Note 1. Note 1—Shear wave, angle beam ultrasonic examination is commonly used to establish circumferential position and dimensions of flaws that produce AE. Time of Flight Diffraction (TOFD), ultrasonic examination is also commonly used for flaw sizing. 1.5 The values stated in inch-pound units are to be regarded as the standard. The values given in parentheses are for information only. 1.6 Thi...

  7. Generic BWR-4 degraded core in-vessel study. Status report

    SciTech Connect (OSTI)

    Not Available

    1984-11-01T23:59:59.000Z

    Original intent of this project was to produce a phenomenological study of the in-vessel degradation which occurs during the TQUX and TQUV sequences for a generic BWR-4 from the initiation of the FSAR Chapter 15 operational transient through core debris bed formation to the failure of the primary pressure boundary. Bounding calculations were to be performed for the two high pressure and low pressure non-LOCA scenarios to assess the uncertainties in the current state of knowledge regarding the source terms for containment integrity studies. Source terms as such were defined in terms of hydrogen generation, unreacted metal, and coolant inventroy, and in terms of the form, sequencing and mode of dispersal through the primary vessel boundary. Fission product release was not to be considered as part of this study. Premature termination of the project, however, led to the dicontinuation of work on an as is basis. Work on the in-core phase from the point of scram to core debris bed formation was largely completed. A preliminary scoping calculation on the debris bed phase had been initiated. This report documents the status of the study at termination.

  8. Spreading of molten corium in MK I geometry following vessel melt-through

    SciTech Connect (OSTI)

    Sienicki, J.J.; Farmer, M.T.; Spencer, B.W.

    1988-01-01T23:59:59.000Z

    For Mk I boiling water reactor severe-accident sequences in which molten corium is postulated to melt through the reactor pressure vessel (RPV) lower head, an important question concerns the relocation of the corium material that drains from the vessel. After filling the sump pits located in the pedestal concrete floor beneath the RPV, the molten corium that collects on the pedestal floor is generally free to flow through the doorway, which provides personnel access to the pedestal, and to spread out over the concrete floor in the annular region between the pedestal wall and the steel liner of the containment shell. A significant issue is whether the corium, after exiting the doorway, can spread under gravity all the way to the liner where thermal attack on the liner steel might be postulated to occur. A computer code (MELTSPREAD) has been developed to investigate the spreading dynamics and thermal interactions of a molten corium layer flowing horizontally over an ablating concrete substrate that may be initially covered with water. The principal objective is to predict, for specific conditions of corium composition, mass, and temperature, the lateral penetration of the corium that drains from a localized hole in the lower head immediately following RPV failure.

  9. Results of scoping tests in corium-water thermal interactions in ex-vessel geometry

    SciTech Connect (OSTI)

    Spencer, B.W.; McUmber, L.; Sehgal, B.R.; Sienicki, J.J.; Squarer, D.

    1983-01-01T23:59:59.000Z

    Results of scoping tests are reported which were performed in the ANL/EPRI Corium Ex-vessel interactions (COREXIT) Facility located at ANL. These tests are examining issues related to containment loading (i.e., steam generation, H/sub 2/ production, and debris dispersal) for hypothetical LWR accidents that are postulated to progress to the point of molten corium breaching the vessel bottom head and entering the reactor cavity. The geometry selected for these tests is a 1 : 30 linear scale of the Zion PWR containment design in which the cavity is connected to the containment volume by an open tunnel through which pass the in-core detector guide tubes. The effects of the corium-water mixing modes were investigated in the first two tests in the series. In test CWTI-1 the molten corium was ejected into water which filled the cavity mockup volume to one-half the passageway height. In CWTI-2, the molten corium was dropped atop the refractory base in the cavity mockup without the presence of water, and water was injected atop the corium melt immediately afterward as per accumulator discharge. These tests have shown significant differences in fuel fragmentation, steam generation rate, hydrogen production, and fuel dispersal. Particularly noteworthy was the significant amount of dispersal of both water and corium debris from the cavity mockup due to the initially rapid steam generation rate in CWTI-1.

  10. Results and analysis of reactor-material experiments on ex-vessel corium quench and dispersal

    SciTech Connect (OSTI)

    Spencer, B.W.; McUmber, L.M.; Sienicki, J.J.; Squarer, D.

    1984-01-01T23:59:59.000Z

    Results of reactor-material experiments and related analysis are described in which molten corium is injected into a mock-up of the reactor cavity region of a PWR. The experiments address ex-vessel interactions such as steam generation (for those cases in which water is present), water and corium dispersal from the cavity, hydrogen generation, direct atmosphere heating by dispersed corium, and debris characterization. Test results indicate efficiencies of steam generation by corium quench ranging up to 65%. Corium sweepout of up to 62% of the injected material was found for those conditions in which steam generation flowrate was augmented by vessel blowdown. The dispersed corium caused very little direct heating of the atmosphere for the configuration employing a trap at the exit of the cavity-to-containment pathway. Corium sweepout phenomena were modeled for high-pressure blowdown conditions, and the results applied to the full-size reactor system predict essentially complete sweepout of corium from the reactor cavity.

  11. Radiation damage characterization in reactor pressure vessel steels with nonlinear ultrasound

    SciTech Connect (OSTI)

    Matlack, K. H. [G.W. Woodruff School of Mechanical Engineering, Georgia Institute of Technology, Atlanta, GA 30332 (United States); Kim, J.-Y. [School of Civil and Environmental Engineering, Georgia Institute of Technology, Atlanta, GA 30332 (United States); Wall, J. J. [G.W. Woodruff School of Mechanical Engineering, Georgia Institute of Technology, Atlanta, GA 30332 and Nuclear Sector, The Electric Power Research Institute, Charlotte, NC 28262 (United States); Qu, J. [Department of Civil and Environmental Engineering, Northwestern University, Evanston, IL 60208 (United States); Jacobs, L. J. [G.W. Woodruff School of Mechanical Engineering, Georgia Institute of Technology, Atlanta, GA 30332 and School of Civil and Environmental Engineering, Georgia Institute of Technology, Atlanta, GA 30332 (United States)

    2014-02-18T23:59:59.000Z

    Nuclear generation currently accounts for roughly 20% of the US baseload power generation. Yet, many US nuclear plants are entering their first period of life extension and older plants are currently undergoing assessment of technical basis to operate beyond 60 years. This means that critical components, such as the reactor pressure vessel (RPV), will be exposed to higher levels of radiation than they were originally intended to withstand. Radiation damage in reactor pressure vessel steels causes microstructural changes such as vacancy clusters, precipitates, dislocations, and interstitial loops that leave the material in an embrittled state. The development of a nondestructive evaluation technique to characterize the effect of radiation exposure on the properties of the RPV would allow estimation of the remaining integrity of the RPV with time. Recent research has shown that nonlinear ultrasound is sensitive to radiation damage. The physical effect monitored by nonlinear ultrasonic techniques is the generation of higher harmonic frequencies in an initially monochromatic ultrasonic wave, arising from the interaction of the ultrasonic wave with microstructural features such as dislocations, precipitates, and their combinations. Current findings relating the measured acoustic nonlinearity parameter to increasing levels of neutron fluence for different representative RPV materials are presented.

  12. Comparison of attenuation coefficients for VVER-440 and VVER-1000 pressure vessels

    SciTech Connect (OSTI)

    Marek, M.; Rataj, J.; Vandlik, S. [Reactor Physics Dept., Research Centre Rez, Husinec 130, 25068 (Czech Republic)

    2011-07-01T23:59:59.000Z

    The paper summarizes the attenuation coefficient of the neutron fluence with E > 0.5 MeV through a reactor pressure vessel for vodo-vodyanoi energetichesky reactor (VVER) reactor types measured and/or calculated for mock-up experiments, as well as for operated nuclear power plant (NPP) units. The attenuation coefficient is possible to evaluate directly only by using the retro-dosimetry, based on a combination of the measured activities from the weld sample and concurrent ex-vessel measurement. The available neutron fluence attenuation coefficients (E > 0.5 MeV), calculated and measured at a mock-up experiment simulating the VVER-440-unit conditions, vary from 3.5 to 6.15. A similar situation is used for the calculations and mock-up experiment measurements for the VVER-1000 RPV, where the attenuation coefficient of the neutron fluence varies from 5.99 to 8.85. Because of the difference in calculations for the real units and the mock-up experiments, the necessity to design and perform calculation benchmarks both for VVER-440 and VVER-1000 would be meaningful if the calculation model is designed adequately to a given unit. (authors)

  13. Neutron flux estimations based on niobium impurities in reactor pressure vessel steel

    SciTech Connect (OSTI)

    Baers, L.B.; Hasanen, E.K. [Technical Research Centre of Finland, Espoo (Finland). Reactor Lab.

    1994-12-31T23:59:59.000Z

    The use of (ppm level) niobium impurities in reactor pressure vessel (RPV) steel for neutron flux estimations based on the reaction {sup 93}Nb (n,n{prime}) {sup 93m}Nb has been reported previously. The method has now been further investigated and refined. Small niobium fractions in RPV steel ({approx} ppm) and plating ({approx} 1%) materials have been separated by ion exchange chromatography in one to three steps. The measured Nb fractions in samples from some four pressure vessel (RPV) base materials were 1 to 3 ppm. The purification of tens of milligrams of RPV material provides sufficient amounts of niobium for mass determination with a highly sensitive (10{sup {minus}5} ppm) Inductively Coupled Plasma Mass Spectrometer (ICP-MS). The {sup 93m}Nb and small remaining {sup 54}Mn activities were measured with a Calibrated Liquid Scintillation Counter (LSC) based on dual label technique and almost 100% efficiency to {sup 93m}Nb. One purification is needed for plating materials ({approx}1% Nb) and two purifications of about one gram of steel with Nb impurities in order to resolve the needed activities ({approx}10 Bq {sup 93m}Nb/{mu}g Nb). The achieved accuracy of the measured specific {sup 93m}Nb activities was about {+-} 3% (1{sigma}) in irradiated RPV plating materials and about {+-} 4% for Nb ppm impurities.

  14. In-vessel coolability and retention of a core melt. Volume 1

    SciTech Connect (OSTI)

    Theofanous, T.G.; Liu, C.; Additon, S.; Angelini, S.; Kymaelaeinen, O.; Salmassi, T. [California Univ., Santa Barbara, CA (United States). Center for Risk Studies and Safety

    1996-10-01T23:59:59.000Z

    The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the Risk Oriented Accident Analysis Methodology (ROAAM), and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow the delineation of the failure boundaries. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is physically unreasonable. Use of this conclusion for any specific application is subject to verifying the required reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, including any applicable coatings. The AP600 is particularly favorable to in-vessel retention. Some ideas to enhance the assessment basis as well as performance in this respect, for applications to larger and/or higher power density reactors are also provided.

  15. Thermal Analysis to Calculate the Vessel Temperature and Stress in Alcator C-Mod Due to the Divertor Upgrade

    SciTech Connect (OSTI)

    Han Zhang, Peter H. Titus, Robert Ellis, Soren Harrison and Rui Vieira

    2012-08-29T23:59:59.000Z

    Alcator C-Mod is planning an upgrade to its outer divertor. The upgrade is intended to correct the existing outer divertor alignment with the plasma, and to operate at elevated temperatures. Higher temperature operation will allow study of edge physics behavior at reactor relevant temperatures. The outer divertor and tiles will be capable of operating at 600oC. Longer pulse length, together with the plasma and RF heat of 9MW, and the inclusion of heater elements within the outer divertor produces radiative energy which makes the sustained operation much more difficult than before. An ANSYS model based on ref. 1 was built for the global thermal analysis of C-Mod. It models the radiative surfaces inside the vessel and between the components, and also includes plasma energy deposition. Different geometries have been simulated and compared. Results show that steady state operation with the divertor at 600oC is possible with no damage to major vessel internal components. The differential temperature between inner divertor structure, or "girdle" and inner vessel wall is ~70oC. This differential temperature is limited by the capacity of the studs that hold the inner divertor backing plates to the vessel wall. At a 70oC temperature differential the stress on the studs is within allowable limits. The thermal model was then used for a stress pass to quantify vessel shell stresses where thermal gradients are significant.

  16. An Approach to Understanding Cohesive Slurry Settling, Mobilization, and Hydrogen Gas Retention in Pulsed Jet Mixed Vessels

    SciTech Connect (OSTI)

    Gauglitz, Phillip A.; Wells, Beric E.; Fort, James A.; Meyer, Perry A.

    2009-05-22T23:59:59.000Z

    The Hanford Waste Treatment and Immobilization Plant (WTP) is being designed and built to pretreat and vitrify a large portion of the waste in Hanford’s 177 underground waste storage tanks. Numerous process vessels will hold waste at various stages in the WTP. Some of these vessels have mixing-system requirements to maintain conditions where the accumulation of hydrogen gas stays below acceptable limits, and the mixing within the vessels is sufficient to release hydrogen gas under normal conditions and during off-normal events. Some of the WTP process streams are slurries of solid particles suspended in Newtonian fluids that behave as non-Newtonian slurries, such as Bingham yield-stress fluids. When these slurries are contained in the process vessels, the particles can settle and become progressively more concentrated toward the bottom of the vessels, depending on the effectiveness of the mixing system. One limiting behavior is a settled layer beneath a particle-free liquid layer. The settled layer, or any region with sufficiently high solids concentration, will exhibit non-Newtonian rheology where it is possible for the settled slurry to behave as a soft solid with a yield stress. In this report, these slurries are described as settling cohesive slurries.

  17. 1 Managed by UT-Battelle for the U.S. Department of Energy NF/IDS Hg Vessel Layout 30 Jun 09

    E-Print Network [OSTI]

    McDonald, Kirk

    1 Managed by UT-Battelle for the U.S. Department of Energy NF/IDS Hg Vessel Layout 30 Jun 09 Cryostat 2 Front Drain Mercury Vessel Concept Matthew F. Glisson Van Graves #12;2 Managed by UT-Battelle;3 Managed by UT-Battelle for the U.S. Department of Energy NF/IDS Hg Vessel Layout 30 Jun 09 Cross Section

  18. A guide for the ASME code for austenitic stainless steel containment vessels for high-level radioactive materials

    SciTech Connect (OSTI)

    Raske, D.T.

    1995-06-01T23:59:59.000Z

    The design and fabrication criteria recommended by the US Department of Energy (DOE) for high-level radioactive materials containment vessels used in packaging is found in Section III, Division 1, Subsection NB of the ASME Boiler and Pressure Vessel Code. This Code provides material, design, fabrication, examination, and testing specifications for nuclear power plant components. However, many of the requirements listed in the Code are not applicable to containment vessels made from austenitic stainless steel with austenitic or ferritic steel bolting. Most packaging designers, engineers, and fabricators are intimidated by the sheer volume of requirements contained in the Code; consequently, the Code is not always followed and many requirements that do apply are often overlooked during preparation of the Safety Analysis Report for Packaging (SARP) that constitutes the basis to evaluate the packaging for certification.

  19. Role of ex-vessel interactions in determining the severe reactor-accident source term for fission products. [PWR; BWR

    SciTech Connect (OSTI)

    Powers, D.A.; Brockmann, J.E.; Bradley, D.R.; Tarbell, W.W.

    1983-01-01T23:59:59.000Z

    The role fission-product release and aerosol generation outside the primary system can have in determining the severe reactor-accident source term is reviewed. Recent analytical and experimental studies of major causes of ex-vessel fission product release and aerosol generation are described. The ejection of molten-core debris from a pressurized-reactor vessel is shown to be a potentially large source of aerosols that has not been recognized in past severe-accident evaluations. A mechanistic model of fission-product release during core-debris interactions with concrete is discussed. Calculations with this model are compared to correlations of experimental data and previous estimates of ex-vessel fission-product release. Predictions with the mechanistic model agree quite well with the data correlations but do not agree at all well with estimates made in the past.

  20. ADDITIONAL STRESS AND FRACTURE MECHANICS ANALYSES OF PRESSURIZED WATER REACTOR PRESSURE VESSEL NOZZLES

    SciTech Connect (OSTI)

    Walter, Matthew [Structural Integrity Associates, Inc.; Yin, Shengjun [ORNL; Stevens, Gary [U.S. Nuclear Regulatory Commission; Sommerville, Daniel [Structural Integrity Associates, Inc.; Palm, Nathan [Westinghouse Electric Company, Cranberry Township, PA; Heinecke, Carol [Westinghouse Electric Company, Cranberry Township, PA

    2012-01-01T23:59:59.000Z

    In past years, the authors have undertaken various studies of nozzles in both boiling water reactors (BWRs) and pressurized water reactors (PWRs) located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Those studies described stress and fracture mechanics analyses performed to assess various RPV nozzle geometries, which were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-life (EOL) to require evaluation of embrittlement as part of the RPV analyses associated with pressure-temperature (P-T) limits. In this paper, additional stress and fracture analyses are summarized that were performed for additional PWR nozzles with the following objectives: To expand the population of PWR nozzle configurations evaluated, which was limited in the previous work to just two nozzles (one inlet and one outlet nozzle). To model and understand differences in stress results obtained for an internal pressure load case using a two-dimensional (2-D) axi-symmetric finite element model (FEM) vs. a three-dimensional (3-D) FEM for these PWR nozzles. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated. To investigate the applicability of previously recommended linear elastic fracture mechanics (LEFM) hand solutions for calculating the Mode I stress intensity factor for a postulated nozzle corner crack for pressure loading for these PWR nozzles. These analyses were performed to further expand earlier work completed to support potential revision and refinement of Title 10 to the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G, Fracture Toughness Requirements, and are intended to supplement similar evaluation of nozzles presented at the 2008, 2009, and 2011 Pressure Vessels and Piping (PVP) Conferences. This work is also relevant to the ongoing efforts of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, Working Group on Operating Plant Criteria (WGOPC) efforts to incorporate nozzle fracture mechanics solutions into a revision to ASME B&PV Code, Section XI, Nonmandatory Appendix G.

  1. RADIOACTIVE MATERIAL SHIPPING PACKAGINGS AND METAL TO METAL SEALS FOUND IN THE CLOSURES OF CONTAINMENT VESSELS INCORPORATING CONE SEAL CLOSURES

    SciTech Connect (OSTI)

    Loftin, B; Glenn Abramczyk, G; Allen Smith, A

    2007-06-06T23:59:59.000Z

    The containment vessels for the Model 9975 radioactive material shipping packaging employ a cone-seal closure. The possibility of a metal-to-metal seal forming between the mating conical surfaces, independent of the elastomer seals, has been raised. It was postulated that such an occurrence would compromise the containment vessel hydrostatic and leakage tests. The possibility of formation of such a seal has been investigated by testing and by structural and statistical analyses. The results of the testing and the statistical analysis demonstrate and procedural changes ensure that hydrostatic proof and annual leakage testing can be accomplished to the appropriate standards.

  2. TOPAZ: a computer code for modeling heat transfer and fluid flow in arbitrary networks of pipes, flow branches, and vessels

    SciTech Connect (OSTI)

    Winters, W.S.

    1984-01-01T23:59:59.000Z

    An overview of the computer code TOPAZ (Transient-One-Dimensional Pipe Flow Analyzer) is presented. TOPAZ models the flow of compressible and incompressible fluids through complex and arbitrary arrangements of pipes, valves, flow branches and vessels. Heat transfer to and from the fluid containment structures (i.e. vessel and pipe walls) can also be modeled. This document includes discussions of the fluid flow equations and containment heat conduction equations. The modeling philosophy, numerical integration technique, code architecture, and methods for generating the computational mesh are also discussed.

  3. EVALUATION OF TROQUE VS CLOSURE BOLT PRELOAD FOR A TYPICAL CONTAINMENT VESSEL UNDER SERVICE CONDITIONS

    SciTech Connect (OSTI)

    Smith, A.

    2010-02-16T23:59:59.000Z

    Radioactive material package containment vessels typically employ bolted closures of various configurations. Closure bolts must retain the lid of a package and must maintain required seal loads, while subjected to internal pressure, impact loads and vibration. The need for insuring that the specified preload is achieved in closure bolts for radioactive materials packagings has been a continual subject of concern for both designers and regulatory reviewers. The extensive literature on threaded fasteners provides sound guidance on design and torque specification for closure bolts. The literature also shows the uncertainty associated with use of torque to establish preload is typically between 10 and 35%. These studies have been performed under controlled, laboratory conditions. The ability to insure required preload in normal service is, consequently, an important question. The study described here investigated the relationship between indicated torque and resulting bolt load for a typical radioactive materials package closure using methods available under normal service conditions.

  4. Assemblies and methods for mitigating effects of reactor pressure vessel expansion

    DOE Patents [OSTI]

    Challberg, R.C.; Gou, P.F.; Chu, C.L.; Oliver, R.P.

    1999-07-27T23:59:59.000Z

    Support assemblies for allowing RPV radial expansion while simultaneously limiting horizontal, vertical, and azimuthal movement of the RPV within a nuclear reactor are described. In one embodiment, the support assembly includes a support block and a guide block. The support block includes a first portion and a second portion, and the first portion is rigidly coupled to the RPV adjacent the first portion. The guide block is rigidly coupled to a reactor pressure vessel support structure and includes a channel sized to receive the second portion of the support block. The second portion of the support block is positioned in the guide block channel to movably couple the guide block to the support block. 6 figs.

  5. Assemblies and methods for mitigating effects of reactor pressure vessel expansion

    DOE Patents [OSTI]

    Challberg, Roy C. (Livermore, CA); Gou, Perng-Fei (Saratoga, CA); Chu, Cherk Lam (San Jose, CA); Oliver, Robert P. (Topsham, ME)

    1999-01-01T23:59:59.000Z

    Support assemblies for allowing RPV radial expansion while simultaneously limiting horizontal, vertical, and azimuthal movement of the RPV within a nuclear reactor are described. In one embodiment, the support assembly includes a support block and a guide block. The support block includes a first portion and a second portion, and the first portion is rigidly coupled to the RPV adjacent the first portion. The guide block is rigidly coupled to a reactor pressure vessel support structure and includes a channel sized to receive the second portion of the support block. The second portion of the support block is positioned in the guide block channel to movably couple the guide block to the support block.

  6. Analysis of the pool critical assembly pressure vessel benchmark using pentran

    SciTech Connect (OSTI)

    Edgar, C. A.; Sjoden, G. E. [Nuclear and Radiological Engineering Program, George W. Woodruff School of Mechanical Engineering, Georgia Inst. of Technology, 770 State St, Atlanta, GA 30332-0745 (United States)

    2012-07-01T23:59:59.000Z

    The internationally circulated Pool Critical Assembly (PCA) Pressure Vessel Benchmark was analyzed using the PENTRAN Parallel Sn code system for the geometry, material, and source specifications as described in the PCA Benchmark documentation. This research focused on utilizing the BUGLE-96 cross section library and accompanying reaction rates, while examining both adaptive differencing on a coarse mesh basis as well as Directional Theta Weighted Sn differencing in order to compare the calculated PENTRAN results to measured data. The results show good comparison with the measured data as well as to the calculated results provided from TORT for the BUGLE-96 cross sections and reaction rates, which suggests PENTRAN is a viable and reliable code system for calculation of light water reactor neutron shielding and dosimetry calculations. (authors)

  7. The influence of metallurgical variables on the temperature dependence of irradiation hardening in pressure vessel steels

    SciTech Connect (OSTI)

    Odette, G.R.; Lucas, G.E.; Klingensmith, R.D. [Univ. of California, Santa Barbara, CA (United States). Dept. of Mechanical Engineering

    1996-12-31T23:59:59.000Z

    Yield stress elevations ({Delta}{sigma}{sub y}) in pressure vessel steels irradiated at intermediate flux and fluence systematically decreased with increasing temperature and decreasing copper and nickel content. Lower stress relief temperature also decreased {Delta}{sigma}{sub y} at bulk copper concentrations greater than about 0.3%. The dependence of {Delta}{sigma}{sub y} on irradiation temperature between 260 and 316 C increased with copper and nickel content and decreased with phosphorus content. When normalized by the average {Delta}{sigma}{sub y}, the fractional temperature dependence correlates with a simple empirical chemistry factor of copper and phosphorus. The correlation predicts data on the irradiation temperature dependence of {Delta}{sigma}{sub y} found in the literature within a standard error of about 0.3 MPa/{degree}C and is consistent with current understanding of hardening mechanisms. However, questions remain about the effects at very low flux and finer scale variations over smaller temperature intervals.

  8. A Unified Cohesive Zone Approach to Model Ductile Brittle Transition in Reactor Pressure Vessel Steels

    SciTech Connect (OSTI)

    Pritam Chakraborty; S. Bulent Biner

    2014-08-01T23:59:59.000Z

    In this study, a unified cohesive zone model has been proposed to predict, Ductile to Brittle Transition, DBT, in Reactor Pressure Vessel, RPV, steels. A general procedure is described to obtain the Cohesive Zone Model, CZM, parameters for the different temperatures and fracture probabilities. In order to establish the full master-curve, the procedure requires three calibration points with one at the upper-shelf for ductile fracture and two for the fracture probabilities, Pf, of 5% and 95% at the lower-shelf. In the current study, these calibrations were carried out by utilizing the experimental fracture toughness values and flow curves. After the calibration procedure, the simulations of fracture behavior (ranging from completely unstable to stable crack extension behavior) in one inch thick compact tension specimens at different temperatures yielded values that were comparable to the experimental fracture toughness values, indicating the viability of such unified modeling approach.

  9. Research and Development Roadmaps for Nondestructive Evaluation of Cables, Concrete, Reactor Pressure Vessels, and Piping Fatique

    SciTech Connect (OSTI)

    Clayton, Dwight A [ORNL] [ORNL; Bakhtiari, Sasan [Argonne National Laboratory (ANL)] [Argonne National Laboratory (ANL); Smith, Cyrus M [ORNL] [ORNL; Simmons, Kevin [Pacific Northwest National Laboratory (PNNL)] [Pacific Northwest National Laboratory (PNNL); Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL)] [Pacific Northwest National Laboratory (PNNL); Coble, Jamie [Pacific Northwest National Laboratory (PNNL)] [Pacific Northwest National Laboratory (PNNL); Brenchley, David [Pacific Northwest National Laboratory (PNNL)] [Pacific Northwest National Laboratory (PNNL); Meyer, Ryan [Pacific Northwest National Laboratory (PNNL)] [Pacific Northwest National Laboratory (PNNL)

    2013-01-01T23:59:59.000Z

    To address these research needs, the MAaD Pathway supported a series of workshops in the summer of 2012 for the purpose of developing R&D roadmaps for enhancing the use of Nondestructive Evaluation (NDE) technologies and methodologies for detecting aging and degradation of materials and predicting the remaining useful life. The workshops were conducted to assess requirements and technical gaps related to applications of NDE for cables, concrete, reactor pressure vessels (RPV), and piping fatigue for extended reactor life. An overview of the outcomes of the workshops is presented here. Details of the workshop outcomes and proposed R&D also are available in the R&D roadmap documents cited in the bibliography and are available on the LWRS Program website (http://www.inl.gov/lwrs).

  10. United States Department of Energy projects related to reactor pressure vessel annealing optimization

    SciTech Connect (OSTI)

    Rosinski, S.T.; Nakos, J.T.

    1993-09-01T23:59:59.000Z

    Light water reactor pressure vessel (RPV) material properties reduced by long-term exposure to neutron irradiation can be recovered through a thermal annealing treatment. This technique to extend RPV life, discussed in this report, provides a complementary approach to analytical methodologies to evaluate RPV integrity. RPV annealing has been successfully demonstrated in the former Soviet Union and on a limited basis by the US (military applications only). The process of demonstrating the technical feasibility of annealing commercial US RPVs is being pursued through a cooperative effort between the nuclear industry and the US Department of Energy (USDOE) Plant Lifetime Improvement (PLIM) Program. Presently, two projects are under way through the USDOE PLIM Program to demonstrate the technical feasibility of annealing commercial US RPVS, (1) annealing re-embrittlement data base development and (2) heat transfer boundary condition experiments.

  11. Reactor pressure vessel integrity research at the Oak Ridge National Laboratory

    SciTech Connect (OSTI)

    Corwin, W.R.; Pennell, W.E.; Pace, J.V.

    1995-12-31T23:59:59.000Z

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. The RPV is the only key safety-related component of the plant for which a duplicate or redundant backup system does not exist. It is therefore imperative to understand and be able to predict the integrity inherent in the RPV. For this reason, the U.S. Nuclear Regulatory Commission has established the related research programs at ORNL described herein to provide for the development and confirmation of the methods used for: (1) establishing the irradiation exposure conditions within the RPV in the Embrittlement Data Base and Dosimetry Evaluation Program, (2) assessing the effects of irradiation on the RPV materials in the Heavy-Section Steel Irradiation Program, and (3) developing overall structural and fracture analyses of RPVs in the Heavy-Section Steel Technology Program.

  12. REACTOR PRESSURE VESSEL ISSUES FOR THE LIGHT-WATER REACTOR SUSTAINABILITY PROGRAM

    SciTech Connect (OSTI)

    Nanstad, Randy K [ORNL; Odette, George Robert [UCSB

    2010-01-01T23:59:59.000Z

    The Light Water Reactor Sustainability Program Plan is a collaborative program between the U.S. Department of Energy and the private sector directed at extending the life of the present generation of nuclear power plants to enable operation to at least 80 years. The reactor pressure vessel (RPV) is one of the primary components requiring significant research to enable such long-term operation. There are significant issues that need to be addressed to reduce the uncertainties in regulatory application, such as, 1) high neutron fluence/long irradiation times, and flux effects, 2) material variability, 3) high-nickel materials, 4)specimen size effects and the fracture toughness master curve, etc. The first issue is the highest priority to obtain the data and mechanistic understanding to enable accurate, reliable embrittlement predictions at high fluences. This paper discusses the major issues associated with long-time operation of existing RPVs and the LWRSP plans to address those issues.

  13. THE DEVELOPMENT OF RADIATION EMBRITTLEMENT MODELS FOR U.S. POWER REACTOR PRESSURE VESSEL STEELS

    SciTech Connect (OSTI)

    Wang, Jy-An John [ORNL; Rao, Nageswara S [ORNL

    2006-01-01T23:59:59.000Z

    The information fusion technique is used to develop radiation embrittlement prediction models for reactor pressure vessel (RPV) steels from U.S. power reactors, including boiling water reactors and pressurized water reactors. The Charpy transition temperature-shift data is used as the primary index of RPV radiation embrittlement in this study. Six parameters {Cu, Ni, P, neutron fluence, irradiation time, and irradiation temperature {are used in the embrittlement prediction models. The results indicate that this new embrittlement predictor achieved reductions of about 49.5% and 52% in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The implications of dose-rate effect and irradiation temperature effects for the development of radiation embrittlement models are also discussed.

  14. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    SciTech Connect (OSTI)

    J. K. Wright; R. N. Wright

    2010-07-01T23:59:59.000Z

    The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750°C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

  15. Effects of irradiation on strength and toughness of commercial LWR vessel cladding

    SciTech Connect (OSTI)

    Haggag, F.M.; Corwin, W.R.; Alexander, D.J.; Nanstad, R.K.

    1987-01-01T23:59:59.000Z

    The potential for stainless steel cladding to improve the fracture behavior of an operating nuclear reactor pressure vessel, particularly during certain overcooling transients, may depend greatly on the properties of the irradiated cladding. Therefore, weld overlay cladding irradiated at temperatures and to fluences relevant to power reactor operation was examined. The cladding was applied to a pressure vessel steel plate by the three-wire series-arc commercial method. Cladding was applied in three layers to provide adequate thickness for the fabrication of test specimens. The three-wire series-arc procedure, developed by Combustion Engineering, Inc., Chattanooga, Tennessee, produced a highly controlled weld chemistry, microstructure, and fracture properties in all three layers of the weld. Charpy V-notch and tensile specimens were irradiated at 288/sup 0/C to fluence levels of 2 and 5 x 10/sup 19/ neutrons/cm/sup 2/ (>1 MeV). Postirradiation testing results show that, in the test temperature range from -125 to 288/sup 0/C, the yield strength increased by 8 to 30%, ductility insignificantly increased, while there was almost no change in ultimate tensile strength. All cladding exhibited ductile-to-brittle transition behavior during Charpy impact testing, due to the dominance of delta-ferrite failures at low temperatures. On the upper shelf, energy was reduced, due to irradiation exposure, 15 and 20%, while the lateral expansion was reduced 43 and 41%, at 2 and 5 x 10/sup 19/ neutrons/cm/sup 2/ (>1 MeV), respectively. In addition, radiation damage resulted in 13 and 28/sup 0/C shifts of the Charpy impact transition temperature at the 41-J level for the low and high fluences, respectively.

  16. Advanced Models of LWR Pressure Vessel Embrittlement for Low Flux-HighFluence Conditions

    SciTech Connect (OSTI)

    Odette, G. Robert; Yamamoto, Takuya

    2013-06-17T23:59:59.000Z

    Neutron embrittlement of reactor pressure vessels (RPVs) is an unresolved issue for light water reactor life extension, especially since transition temperature shifts (TTS) must be predicted for high 80-year fluence levels up to approximately 1,020 n/cm{sup 2}, far beyond the current surveillance database. Unfortunately, TTS may accelerate at high fluence, and may be further amplified by the formation of late blooming phases that result in severe embrittlement even in low-copper (Cu) steels. Embrittlement by this mechanism is a potentially significant degradation phenomenon that is not predicted by current regulatory models. This project will focus on accurately predicting transition temperature shifts at high fluence using advanced physically based, empirically validated and calibrated models. A major challenge is to develop models that can adjust test reactor data to account for flux effects. Since transition temperature shifts depend on synergistic combinations of many variables, flux-effects cannot be treated in isolation. The best current models systematically and significantly under-predict transition temperature at high fluence, although predominantly for irradiations at much higher flux than actual RPV service. This project will integrate surveillance, test reactor and mechanism data with advanced models to address a number of outstanding RPV embrittlement issues. The effort will include developing new databases and preliminary models of flux effects for irradiation conditions ranging from very low (e.g., boiling water reactor) to high (e.g., accelerated test reactor). The team will also develop a database and physical models to help predict the conditions for the formation of Mn-Ni-Si late blooming phases and to guide future efforts to fully resolve this issue. Researchers will carry out other tasks on a best-effort basis, including prediction of transition temperature shift attenuation through the vessel wall, remediation of embrittlement by annealing, and fracture toughness master curve issues.

  17. Development of the RCRV Design: Top Level Requirements The Top Level Requirements (TLR) for the Regional Class Research Vessel (RCRV) are

    E-Print Network [OSTI]

    underwater vehicles (AUVs) Deployment and recovery of unmanned aerial systems (UASs) and weather balloons. The vessel's Underwater Radiated Noise (URN) will be minimized through treatments and vibration dampening, monitoring, and servicing of remotely operated vehicles (ROVs) (appropriate for vessel size) and autonomous

  18. In-Vessel Retention Technology Development and Use for Advanced PWR Designs in the USA and Korea

    SciTech Connect (OSTI)

    T.G. Theofanous; S.J. Oh; J.H. Scobel

    2004-05-18T23:59:59.000Z

    In-Vessel Retention (IVR) of molten core debris by means of external reactor vessel flooding is a cornerstone of severe accident management for Westinghouse's AP600 (advanced passive light water reactor) design. The case for its effectiveness (made in previous work by the PI) has been thoroughly documented, reviewed as part of the licensing certification, and accepted by the US Nuclear Regulatory Commission. A successful IVR would terminate a severe accident, passively, with the core in a stable, coolable configuration (within the lower head), thus avoiding the largely uncertain accident evolution with the molten debris on the containment floor. This passive plant design has been upgraded by Westinghouse to the AP1000, a 1000 MWe plant very similar to the AP600. The severe accident management approach is very similar too, including In-Vessel Retention as the cornerstone feature, and initial evaluations indicated that this would be feasible at the higher power as well. A similar strategy is adopted in Korea for the APR1400 plant. The overall goal of this project is to provide experimental data and develop the necessary basic understanding so as to allow the robust extension of the AP600 In-Vessel Retention strategy for severe accident management to higher power reactors, and in particular, to the AP1000 advanced passive design.

  19. BLENDED CALCIUM ALUMINATE-CALCIUM SULFATE CEMENT-BASED GROUT FOR P-REACTOR VESSEL IN-SITU DECOMMISSIONING

    SciTech Connect (OSTI)

    Langton, C.; Stefanko, D.

    2011-03-10T23:59:59.000Z

    The objective of this report is to document laboratory testing of blended calcium aluminate - calcium hemihydrate grouts for P-Reactor vessel in-situ decommissioning. Blended calcium aluminate - calcium hemihydrate cement-based grout was identified as candidate material for filling (physically stabilizing) the 105-P Reactor vessel (RV) because it is less alkaline than portland cement-based grout which has a pH greater than 12.4. In addition, blended calcium aluminate - calcium hemihydrate cement compositions can be formulated such that the primary cementitious phase is a stable crystalline material. A less alkaline material (pH {<=} 10.5) was desired to address a potential materials compatibility issue caused by corrosion of aluminum metal in highly alkaline environments such as that encountered in portland cement grouts [Wiersma, 2009a and b, Wiersma, 2010, and Serrato and Langton, 2010]. Information concerning access points into the P-Reactor vessel and amount of aluminum metal in the vessel is provided elsewhere [Griffin, 2010, Stefanko, 2009 and Wiersma, 2009 and 2010, Bobbitt, 2010, respectively]. Radiolysis calculations are also provided in a separate document [Reyes-Jimenez, 2010].

  20. 1 Copyright 2010 by ASME Proceedings of the ASME 2010 Pressure Vessels & Piping Division / K-PVP Conference

    E-Print Network [OSTI]

    Tijsseling, A.S.

    1 Copyright © 2010 by ASME Proceedings of the ASME 2010 Pressure Vessels & Piping Division / K-6]. Developers and users of computational codes need full- scale data with which to compare their theoretical;2 Copyright © 2010 by ASME Here, a large-scale pipeline test rig at Deltares, Delft, The Netherlands has been

  1. Validation of vessel-based registration for correction of brain shift I. Reinertsen a,*, M. Descoteaux b,1

    E-Print Network [OSTI]

    Siddiqi, Kaleem

    Validation of vessel-based registration for correction of brain shift I. Reinertsen a,*, M April 2007 Abstract The displacement and deformation of brain tissue is a major source of error in image-guided neurosurgery systems. We have designed and implemented a method to detect and correct brain shift using pre

  2. Miniaturized reaction vessel system, method for performing site-specific biochemical reactions and affinity fractionation for use in DNA sequencing

    DOE Patents [OSTI]

    Mirzabekov, Andrei Darievich (Moscow, RU); Lysov, Yuri Petrovich (Moscow, RU); Dubley, Svetlana A. (Moscow, RU)

    2000-01-01T23:59:59.000Z

    A method for fractionating and sequencing DNA via affinity interaction is provided comprising contacting cleaved DNA to a first array of oligonucleotide molecules to facilitate hybridization between said cleaved DNA and the molecules; extracting the hybridized DNA from the molecules; contacting said extracted hybridized DNA with a second array of oligonucleotide molecules, wherein the oligonucleotide molecules in the second array have specified base sequences that are complementary to said extracted hybridized DNA; and attaching labeled DNA to the second array of oligonucleotide molecules, wherein the labeled re-hybridized DNA have sequences that are complementary to the oligomers. The invention further provides a method for performing multi-step conversions of the chemical structure of compounds comprising supplying an array of polyacrylamide vessels separated by hydrophobic surfaces; immobilizing a plurality of reactants, such as enzymes, in the vessels so that each vessel contains one reactant; contacting the compounds to each of the vessels in a predetermined sequence and for a sufficient time to convert the compounds to a desired state; and isolating the converted compounds from said array.

  3. Optimal Railroad Rail Grinding for Fatigue Mitigation 

    E-Print Network [OSTI]

    Tangtragulwong, Potchara

    2012-02-14T23:59:59.000Z

    or induction methods, 6 must be used to detect the subsurface cracks instead with some limitations on resolution and reliability. Focusing on subsurface crack nucleation in the head of a thermite welded rail, Fry et al. [12] determine an approximated...

  4. Field tests of timber railroad bridge piles

    E-Print Network [OSTI]

    Donovan, Kendra Ann

    2005-02-17T23:59:59.000Z

    with crane??????????? 29 2.8 Plan view of load cells and string pot arrangement beneath bent cap??.. 29 3.1 Plot of load (Crane B) vs. time for Pile 1?????????????. 30 3.2 Plot of load (Crane B) vs. time for Pile 2?????????????. 31 3.3 Plot... of load (Crane B) vs. time for Pile 3?????????????. 32 3.4 Plot of load (Crane B) vs. time for Pile 4?????????????..33 3.5 Plot of load (Crane B) vs. time for Pile 5?????????????. 34 3.6 Plot of load (Crane B) vs. time for Pile 6...

  5. SECURING AND PROTECTING AMERICA'S RAILROAD SYSTEM

    E-Print Network [OSTI]

    Nagurney, Anna

    , HARRISBURG JEREMY F. PLANT, PH. D SCHOOL OF PUBLIC AFFAIRS RICHARD R. YOUNG, PH. D SCHOOL OF BUSINESS ADMINISTRATION ©2007 The Pennsylvania State University, Harrisburg #12;2 TABLE OF CONTENTS EXECUTIVE SUMMARY 5

  6. Railroads and shippers clash over coal dust

    SciTech Connect (OSTI)

    Buchsbaum, L.

    2007-11-15T23:59:59.000Z

    In an effort to reduce coal spillage from railcars, mines in the Powder River Basin (PRB) now load coal with a loaf profile but, reportedly, beginning in 2008, Burlington Northern Santa Fe (BNSF) will announce guidelines requiring all PRB coal loads to be sprayed with a chemical surfactant. If this does not fix the problem, greater measures will be taken. At the time of going to press, the details of how this would be implemented and regulated were unresolved. 1 photo.

  7. ILLINOIS RAILROAD ENGINEERING A Quantitative Decision Support

    E-Print Network [OSTI]

    Barkan, Christopher P.L.

    between capital investment and delay cost Link (sub) properties Alternatives generator Capacity expansion alternatives (cost & additional capacity) Estimated future traffic (trains/day/OD) Budget Investment Selection: ­ Infrastructure ­ Equipment ­ Train dispatching, traffic mix ­ Human resources Network Capacity must be increased

  8. Distillate Fuel Oil Sales for Railroad Use

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40Coal Stocks at1,066,688 760,877 951,322 1,381,127 1,710,513 1,751,162

  9. Enhancing Railroad Hazardous Materials Transportation Safety | Department

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic Plan Departmentof EnergyPublic LawEnergyEnhanced Reduceof Energy

  10. Federal Railroad Administration | Department of Energy

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic Plan| Department of.pdf6-OPAMDepartment6Awards »Facilities |Federal

  11. Railroad Commission of Texas | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are beingZealand Jump to:Ezfeedflag JumpID-f < RAPID‎ | Roadmap Jump to:b <RGSRadium HotRail Splitter Wind

  12. Railroad Valley Geothermal Area | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are beingZealand Jump to:Ezfeedflag JumpID-f < RAPID‎ | Roadmap Jump to:b <RGSRadium HotRail Splitter

  13. Automated registration of multispectral MR vessel wall images of the carotid artery

    SciTech Connect (OSTI)

    Klooster, R. van 't; Staring, M.; Reiber, J. H. C.; Lelieveldt, B. P. F.; Geest, R. J. van der, E-mail: rvdgeest@lumc.nl [Department of Radiology, Division of Image Processing, Leiden University Medical Center, 2300 RC Leiden (Netherlands); Klein, S. [Department of Radiology and Department of Medical Informatics, Biomedical Imaging Group Rotterdam, Erasmus MC, Rotterdam 3015 GE (Netherlands)] [Department of Radiology and Department of Medical Informatics, Biomedical Imaging Group Rotterdam, Erasmus MC, Rotterdam 3015 GE (Netherlands); Kwee, R. M.; Kooi, M. E. [Department of Radiology, Cardiovascular Research Institute Maastricht, Maastricht University Medical Center, Maastricht 6202 AZ (Netherlands)] [Department of Radiology, Cardiovascular Research Institute Maastricht, Maastricht University Medical Center, Maastricht 6202 AZ (Netherlands)

    2013-12-15T23:59:59.000Z

    Purpose: Atherosclerosis is the primary cause of heart disease and stroke. The detailed assessment of atherosclerosis of the carotid artery requires high resolution imaging of the vessel wall using multiple MR sequences with different contrast weightings. These images allow manual or automated classification of plaque components inside the vessel wall. Automated classification requires all sequences to be in alignment, which is hampered by patient motion. In clinical practice, correction of this motion is performed manually. Previous studies applied automated image registration to correct for motion using only nondeformable transformation models and did not perform a detailed quantitative validation. The purpose of this study is to develop an automated accurate 3D registration method, and to extensively validate this method on a large set of patient data. In addition, the authors quantified patient motion during scanning to investigate the need for correction. Methods: MR imaging studies (1.5T, dedicated carotid surface coil, Philips) from 55 TIA/stroke patients with ipsilateral <70% carotid artery stenosis were randomly selected from a larger cohort. Five MR pulse sequences were acquired around the carotid bifurcation, each containing nine transverse slices: T1-weighted turbo field echo, time of flight, T2-weighted turbo spin-echo, and pre- and postcontrast T1-weighted turbo spin-echo images (T1W TSE). The images were manually segmented by delineating the lumen contour in each vessel wall sequence and were manually aligned by applying throughplane and inplane translations to the images. To find the optimal automatic image registration method, different masks, choice of the fixed image, different types of the mutual information image similarity metric, and transformation models including 3D deformable transformation models, were evaluated. Evaluation of the automatic registration results was performed by comparing the lumen segmentations of the fixed image and moving image after registration. Results: The average required manual translation per image slice was 1.33 mm. Translations were larger as the patient was longer inside the scanner. Manual alignment took 187.5 s per patient resulting in a mean surface distance of 0.271 ± 0.127 mm. After minimal user interaction to generate the mask in the fixed image, the remaining sequences are automatically registered with a computation time of 52.0 s per patient. The optimal registration strategy used a circular mask with a diameter of 10 mm, a 3D B-spline transformation model with a control point spacing of 15 mm, mutual information as image similarity metric, and the precontrast T1W TSE as fixed image. A mean surface distance of 0.288 ± 0.128 mm was obtained with these settings, which is very close to the accuracy of the manual alignment procedure. The exact registration parameters and software were made publicly available. Conclusions: An automated registration method was developed and optimized, only needing two mouse clicks to mark the start and end point of the artery. Validation on a large group of patients showed that automated image registration has similar accuracy as the manual alignment procedure, substantially reducing the amount of user interactions needed, and is multiple times faster. In conclusion, the authors believe that the proposed automated method can replace the current manual procedure, thereby reducing the time to analyze the images.

  14. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    SciTech Connect (OSTI)

    J. K. Wright; R. N. Wright

    2008-04-01T23:59:59.000Z

    The U.S. Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic, or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development Program is responsible for performing research and development on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2¼Cr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have concluded, however, that with adequate engineered cooling of the vessel, the A508/533 steels are also acceptable.

  15. Evolution of Nickel-Manganese-Silicon Dominated Phases in Highly Irradiated Reactor Pressure Vessel Steels

    SciTech Connect (OSTI)

    Peter B Wells; Yuan Wu; Tim Milot; G. Robert Odette; Takuya Yamamoto; Brandon Miller; James Cole

    2014-11-01T23:59:59.000Z

    Formation of a high density of Ni-Mn-Si nm-scale precipitates in irradiated reactor pressure vessel steels, both with and without Cu, could lead to severe embrittlement. Models long ago predicted that these precipitates, which are not treated in current embrittlement regulations, would emerge only at high fluence. However, the mechanisms and variables that control Ni-Mn- Si precipitate formation, and their detailed characteristics, have not been well understood. High flux irradiations of six steels with systematic variations in Cu and Ni were carried out at ˜ 295±5°C to high and very high neutron fluences of ˜ 1.3x1020 and 1.1x1021 n/cm2. Atom probe tomography (APT) shows that significant mole fractions of these precipitates form in the Cu bearing steels at ˜ 1.3x1020 n/cm2, while they are only beginning to develop in Cu-free steels. However, large mole fractions, far in excess of those found in previous studies, are observed at 1.1x1021 n/cm2 at all Cu levels. The precipitates diffract, and in one case are compositionally and structurally consistent with the Mn6Ni16Si7 G-phase. At the highest fluence, the large precipitate mole fractions primarily depend on the steel Ni content, rather than Cu, and lead to enormous strength increases up to about 700 MPa. The implications of these results to light water reactor life extension are discussed briefly.

  16. On-Site Oxy-Lance Size Reduction of South Texas Project Reactor Vessel Heads - 12324

    SciTech Connect (OSTI)

    Posivak, Edward [WMG, inc. (United States); Keeney, Gilbert; Wheeler, Dean [Shaw Group (United States)

    2012-07-01T23:59:59.000Z

    On-Site Oxy-Lance size reduction of mildly radioactive large components has been accomplished at other operating plants. On-Site Oxy-Lance size reduction of more radioactive components like Reactor Vessel Heads had previously been limited to decommissioning projects. Building on past decommissioning and site experience, subcontractors for South Texas Project Nuclear Operating Company (STPNOC) developed an innovative integrated system to control smoke, radioactive contamination, worker dose, and worker safety. STP's innovative, easy to use CEDM containment that provided oxy lance access, smoke control, and spatter/contamination control was the key to successful segmentation for cost-effective and ALARA packaging and transport for disposal. Relative to CEDM milling, STP oxy-lance segmentation saved approximately 40 person- REM accrued during 9,000 hours logged into the radiological controlled area (RCA) during more than 3,800 separate entries. Furthermore there were no personnel contamination events or respiratory uptakes of radioactive material during the course of the entire project. (authors)

  17. Lessons Learned Following the Successful Decommissioning of a Reaction Vessel Containing Lime Sludge and Technetium-99

    SciTech Connect (OSTI)

    Dawson, P. M.; Watson, D. D.; Hylko, J. M.

    2002-02-25T23:59:59.000Z

    This paper documents how WESKEM, LLC utilized available source term information, integrated safety management, and associated project controls to safely decommission a reaction vessel and repackage sludge containing various Resource Conservation and Recovery Act constituents and technetium-99 (Tc-99). The decommissioning activities were segmented into five separate stages, allowing the project team to control work related decisions based on their knowledge, experience, expertise, and field observations. The information and experience gained from each previous stage and rehearsals contributed to modifying subsequent entries, further emphasizing the importance of developing hold points and incorporating lessons learned. The hold points and lessons learned, such as performing detailed personal protective equipment (PPE) inspections during sizing and repackaging operations, and using foam-type piping insulation to prevent workers from cutting or puncturing their PPE on sharp edge s or small shards generated during sizing operations, minimized direct contact with the Tc-99. To prevent the spread of contamination, the decommissioning activities were performed inside a containment enclosure connected to negative air machines. After performing over 235 individual entries totaling over 285 project hours, only one first aid was recorded during this five-stage project.

  18. Head Loss Evaluation in a PWR Reactor Vessel Using CFD Analysis

    SciTech Connect (OSTI)

    Ji Hwan Jeong; Jong Pil Park [School of Mechanical Engineering, Pusan National University, Enesys Jangjeon-dong, Geumjeong-gu, Busan (Korea, Republic of); Byoung-Sub Han [Jangdae-dong, Yusong-gu, Daejeon (Korea, Republic of)

    2006-07-01T23:59:59.000Z

    Nuclear vendors and utilities perform lots of simulations and analyses in order to ensure the safe operation of nuclear power plants (NPPs). In general, the simulations are carried out using vendor-specific design codes and best-estimate system analysis codes and most of them were developed based on 1-dimensional lumped parameter models. During the past decade, however, computers, parallel computation methods, and 3-dimensional computational fluid dynamics (CFD) codes have been dramatically enhanced. It is believed to be beneficial to take advantage of advanced commercial CFD codes in safety analysis and design of NPPs. The present work aims to analyze the flow distribution in downcomer and lower plenum of Korean standard nuclear power plants (KSNPs) using STAR-CD. The lower plenum geometry of a PWR is very complicated since there are so many reactor internals, which hinders in CFD analysis for real reactor geometry up to now. The present work takes advantage of 3D CAD model so that real geometry of lower plenum is used. The results give a clear figure about flow fields in the reactor vessel, which is one of major safety concerns. The calculated pressure drop across downcomer and lower plenum appears to be in good agreement with the data in engineering calculation note. A algorithm which can evaluate head loss coefficient which is necessary for thermal-hydraulic system code running was suggested based on this CFD analysis results. (authors)

  19. Use of MCNP for characterization of reactor vessel internals waste from decommissioned nuclear reactors

    SciTech Connect (OSTI)

    Love, E.F.; Pauley, K.A.; Reid, B.D.

    1995-09-01T23:59:59.000Z

    This study describes the use of the Monte Carlo Neutron-Photon (MCNP) code for determining activation levels of irradiated reactor vessel internals hardware. The purpose of the analysis is to produce data for the Department of Energy`s Greater-Than-Class C Low-Level Radioactive Waste Program. An MCNP model was developed to analyze the Yankee Rowe reactor facility. The model incorporates reactor geometry, material compositions, and operating history data acquired from Yankee Atomic Electric Company. In addition to the base activation analysis, parametric studies were performed to determine the sensitivity of activation to specific parameters. A component sampling plan was also developed to validate the model results, although the plan was not implemented. The calculations for the Yankee Rowe reactor predict that only the core baffle and the core support plates will be activated to levels above the Class C limits. The parametric calculations show, however, that the large uncertainties in the material compositions could cause errors in the estimates that could also increase the estimated activation level of the core barrel to above the Class C limits. Extrapolation of the results to other reactor facilities indicates that in addition to the baffle and support plates, core barrels may also be activated to above Class C limits; however the classification will depend on the specific operating conditions of the reactor and the specific material compositions of the metal, as well as the use of allowable concentration averaging practices in packaging and classifying the waste.

  20. Application of micromechanical models of ductile fracture initiation to reactor pressure vessel materials

    SciTech Connect (OSTI)

    Chaouadi, R.; Walle, E. van; Fabry, A.; Velde, J. van de [SCK-CEN, Mol (Belgium); Meester, P. de [KUL, Heverlee (Belgium). Metals and Materials Science Dept.

    1996-12-31T23:59:59.000Z

    The aim of the current study is the application of local micromechanical models to predict crack initiation in ductile materials. Two reactor pressure vessel materials have been selected for this study: JRQ IAEA monitor base metal (A533B Cl.1) and Doel-IV weld material. Charpy impact tests have been performed in both un-irradiated and irradiated conditions. In addition to standard tensile tests, notched tensile specimens have been tested. The upper shelf energy of the weld material remains almost un-affected by irradiation, whereas a decrease of 20% is detected for the base metal. Accordingly, the tensile properties of the weld material do not reveal a clear irradiation effect on the yield and ultimate stresses, this in contrast to the base material flow properties. The tensile tests have been analyzed in terms of micromechanical models. A good correlation is found between the standard tests and the micromechanical models, that are able to predict the ductile damage evolution in these materials. Additional information on the ductility behavior of these materials is revealed by this micromechanical analysis.

  1. Modeling of irradiation embrittlement and annealing/recovery in pressure vessel steels

    SciTech Connect (OSTI)

    Lott, R.G.; Freyer, P.D. [Westinghouse Science and Technology Center, Pittsburgh, PA (United States)

    1996-12-31T23:59:59.000Z

    The results of reactor pressure vessel (RPV) annealing studies are interpreted in light of the current understanding of radiation embrittlement phenomena in RPV steels. An extensive RPV irradiation embrittlement and annealing database has been compiled and the data reveal that the majority of annealing studies completed to date have employed test reactor irradiated weldments. Although test reactor and power reactor irradiations result in similar embrittlement trends, subtle differences between these two damage states can become important in the interpretation of annealing results. Microstructural studies of irradiated steels suggest that there are several different irradiation-induced microstructural features that contribute to embrittlement. The amount of annealing recovery and the post-anneal re-embrittlement behavior of a steel are determined by the annealing response of these microstructural defects. The active embrittlement mechanisms are determined largely by the irradiation temperature and the material composition. Interpretation and thorough understanding of annealing results require a model that considers the underlying physical mechanisms of embrittlement. This paper presents a framework for the construction of a physically based mechanistic model of irradiation embrittlement and annealing behavior.

  2. Effects of thermal annealing and reirradiation on toughness of reactor pressure vessel steels

    SciTech Connect (OSTI)

    Nanstad, R.K.; Iskander, S.K.; Sokolov, M.A. [and others

    1996-12-31T23:59:59.000Z

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPV) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes recent experimental results from work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response, or {open_quotes}recovery,{close_quotes} of several irradiated RPV steels; it also includes recent results from both ORNL and the Russian Research Center-Kurchatov Institute (RRC-KI) on a cooperative program of irradiation, annealing and reirradiation of both U.S. and Russian RPV steels. The cooperative program was conducted under the auspices of Working Group 3, U.S./Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS). The materials investigated are an RPV plate and various submerged-arc welds, with tensile, Charpy impact toughness, and fracture toughness results variously determined. Experimental results are compared with applicable prediction guidelines, while observed differences in annealing responses and reirradiation rates are discussed.

  3. Pressure vessel embrittlement predictions based on a composite model of copper precipitation and point defect clustering

    SciTech Connect (OSTI)

    Stoller, R.E. [Oak Ridge National Lab., TN (United States). Metals and Ceramics Div.

    1996-12-31T23:59:59.000Z

    A theoretical model is used to investigate the relative importance of point defect clusters (PDC) and copper-rich precipitates in reactor pressure vessel (RPV) embrittlement and to examine the influence of a broad range of irradiation and material parameters on predicted yield strength changes. The results indicate that there are temperature and displacement rate regimes wherein either CRP or PDC can dominate the material`s response to irradiation, with both interstitial and vacancy type defects contributing to the PDC component. The different dependencies of the CRP and PDC on temperature and displacement rate indicate that simple data extrapolations could lead to poor predictions of RPV embrittlement. It is significant that the yield strength changes predicted by the composite PDC/CRP model exhibit very little dependence on displacement rate below about 10{sup {minus}9} dpa/s. If this result is confirmed, concerns about accelerated displacement rates in power reactor surveillance programs should be minimized. The sensitivity of the model to microstructural parameters highlights the need for more detailed microstructural characterization of RPV steels.

  4. Nondestructive characterization of embrittlement in reactor pressure vessel steels -- A feasibility study

    SciTech Connect (OSTI)

    McHenry, H.I.; Alers, G.A. [National Inst. of Standards and Technology, Boulder, CO (United States). Materials Reliability Div.

    1998-03-01T23:59:59.000Z

    The Nuclear Regulatory Commission recently initiated a study by NIST to assess the feasibility of using physical-property measurements for evaluating radiation embrittlement in reactor pressure vessel (RPV) steels. Ultrasonic and magnetic measurements provide the most promising approaches for nondestructive characterization of RPV steels because elastic waves and magnetic fields can sense the microstructural changes that embrittle materials. The microstructural changes of particular interest are copper precipitation hardening, which is the likely cause of radiation embrittlement in RPV steels, and the loss of dislocation mobility that is an attribute of the ductile-to-brittle transition. Measurements were made on a 1% copper steel, ASTM grade A710, in the annealed, peak-aged and overaged conditions, and on an RPV steel, ASTM grade A533B. Nonlinear ultrasonic and micromagnetic techniques were the most promising measures of precipitation hardening. Ultrasonic velocity measurements and the magnetic properties associated with hysteresis-loop measurements were not particularly sensitive to either precipitation hardening or the ductile-to-brittle transition. Measurements of internal friction using trapped ultrasonic resonance modes detected energy losses due to the motion of pinned dislocations; however, the ultrasonic attenuation associated with these measurements was small compared to the attenuation caused by beam spreading that would occur in conventional ultrasonic testing of RPVs.

  5. Creep failure of a reactor pressure vessel lower head under severe accident conditions

    SciTech Connect (OSTI)

    Pilch, M.M.; Ludwigsen, J.S.; Chu, T.Y. [Sandia National Labs., Albuquerque, NM (United States); Rashid, Y.R. [Anatech, San Diego, CA (United States)

    1998-08-01T23:59:59.000Z

    A severe accident in a nuclear power plant could result in the relocation of large quantities of molten core material onto the lower head of he reactor pressure vessel (RPV). In the absence of inherent cooling mechanisms, failure of the RPV ultimately becomes possible under the combined effects of system pressure and the thermal heat-up of the lower head. Sandia National Laboratories has performed seven experiments at 1:5th scale simulating creep failure of a RPV lower head. This paper describes a modeling program that complements the experimental program. Analyses have been performed using the general-purpose finite-element code ABAQUS-5.6. In order to make ABAQUS solve the specific problem at hand, a material constitutive model that utilizes temperature dependent properties has been developed and attached to ABAQUS-executable through its UMAT utility. Analyses of the LHF-1 experiment predict instability-type failure. Predicted strains are delayed relative to the observed strain histories. Parametric variations on either the yield stress, creep rate, or both (within the range of material property data) can bring predictions into agreement with experiment. The analysis indicates that it is necessary to conduct material property tests on the actual material used in the experimental program. The constitutive model employed in the present analyses is the subject of a separate publication.

  6. Effects of thermal annealing and reirradiation on toughness of reactor pressure vessel steels

    SciTech Connect (OSTI)

    Nanstad, R.K.; Iskander, S.K.; Sokolov, M.A. [Oak Ridge National Lab., TN (United States)] [and others

    1997-02-01T23:59:59.000Z

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPV) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes recent experimental results from work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response, or {open_quotes}recovery,{close_quotes} of several irradiated RPV steels; it also includes recent results from both ORNL and the Russian Research Center-Kurchatov Institute (RRC-KI) on a cooperative program of irradiation, annealing and reirradiation of both U.S. and Russian RPV steels. The cooperative program was conducted under the auspices of Working Group 3, U.S./Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS). The materials investigated are an RPV plate and various submerged-arc welds, with tensile, Charpy impact toughness, and fracture toughness results variously determined. Experimental results are compared with applicable prediction guidelines, while observed differences in annealing responses and reirradiation rates are discussed.

  7. A Modification of the Inner and Outer Core for Reactor Pressure Vessel Lifetime Extension

    SciTech Connect (OSTI)

    Seo, Bo Kyun [Hanyang University (Korea, Republic of); Kim, Jong Kyung [Hanyang University (Korea, Republic of); Shin, Chang Ho [Hanyang University (Korea, Republic of); Kwon, Tae Je [Nuclear Fuel Company (Korea, Republic of)

    2001-03-15T23:59:59.000Z

    The feasibility of nuclear power plant lifetime extension was examined by reducing the fast neutron fluence at the reactor pressure vessel (RPV) and relieving irradiation embrittlement of materials, and thus ensuring enough structural integrity beyond the design lifetime. Two fluence reduction options, peripheral assembly replacement and additional shield installation in the outer core structures, were applied to the Kori Unit-1 reactor, and the fluence reduction effect was carefully analyzed. For an accurate estimate of the neutron fluence at the RPV and a reasonable description of the modified peripheral assemblies, a full-scope explicit modeling of a Monte Carlo simulation was employed in all calculations throughout this study. The Kori Unit-1 cycle-16 core was modeled on a three-dimensional representation by using the MCNP4B code, and the fluence distribution was estimated at the inner wall beltline around the circumferential weld of the RPV. On the basis of fracture toughness requirements of the RPV, the two modified cases were predicted to have an additional life of 7 to 10 effective full-power years. Throughout the core nuclear characteristics analyses, it was confirmed that the critical peaking factors for safe reactor operation were satisfied with the design limits.

  8. International Atomic Energy Agency (IAEA) Coordinated Research Projects on Structural Integrity of Reactor Pressure Vessels

    SciTech Connect (OSTI)

    Server, W. L. [ATI Consulting, Pinehurst, NC; Nanstad, Randy K [ORNL

    2009-01-01T23:59:59.000Z

    The International Atomic Energy Agency (IAEA) has conducted a series of Coordinated Research Projects (CRPs) that have focused on irradiated reactor pressure vessel (RPV) steel fracture toughness properties and approaches for assuring structural integrity of RPVs throughout operating life. A series of nine CRPs have been sponsored by the IAEA, starting in the early 1970s, focused on neutron radiation effects on RPV steels. The purpose of the CRPs was to develop comparisons and correlations to test the uniformity of irradiated results through coordinated international research studies and data sharing. Consideration of dose rate effects, effects of alloying (nickel, manganese, silicon, etc.) and residual elements (eg., copper and phosphorus), and drop in upper shelf toughness are also important for assessing neutron embrittlement effects. The ultimate use of embrittlement understanding is assuring structural integrity of the RPV under current and future operation and accident conditions. Material fracture toughness is the key ingredient needed for this assessment, and many of the CRPs have focused on measurement and application of irradiated fracture toughness. This paper presents an overview of the progress made since the inception of the CRPs in the early 1970s. The chronology and importance of each CRP have been reviewed and put into context for continued and long-term safe operation of RPVs.

  9. BWR ex-vessel steam explosion analysis with MC3D code

    SciTech Connect (OSTI)

    Leskovar, M. [Josef Stefan Inst., Jamova cesta 39, 1001 Ljubljana (Slovenia)

    2012-07-01T23:59:59.000Z

    A steam explosion may occur, during a severe reactor accident, when the molten core comes into contact with the coolant water. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. To resolve the open issues in steam explosion understanding and modeling, the OECD program SERENA phase 2 was launched at the end of year 2007, focusing on reactor applications. To verify the progress made in the understanding and modeling of fuel coolant interaction key phenomena for reactor applications a reactor exercise has been performed. In this paper the BWR ex-vessel steam explosion study, which was carried out with the MC3D code in conditions of the SERENA reactor exercise for the BWR case, is presented and discussed. The premixing simulations were performed with two different jet breakup modeling approaches and the explosion was triggered also at the expected most challenging time. For the most challenging case, at the cavity wall the highest calculated pressure was {approx}20 MPa and the highest pressure impulse was {approx}90 kPa.s. (authors)

  10. TECHNICAL BASIS AND APPLICATION OF NEW RULES ON FRACTURE CONTROL OF HIGH PRESSURE HYDROGEN VESSEL IN ASME SECTION VIII, DIVISION 3 CODE

    SciTech Connect (OSTI)

    Rawls, G

    2007-04-30T23:59:59.000Z

    As a part of an ongoing activity to develop ASME Code rules for the hydrogen infrastructure, the ASME Boiler and Pressure Vessel Code Committee approved new fracture control rules for Section VIII, Division 3 vessels in 2006. These rules have been incorporated into new Article KD-10 in Division 3. The new rules require determining fatigue crack growth rate and fracture resistance properties of materials in high pressure hydrogen gas. Test methods have been specified to measure these fracture properties, which are required to be used in establishing the vessel fatigue life. An example has been given to demonstrate the application of these new rules.

  11. The toughness of irradiated pressure water reactor (PWR) vessel shell rings and the effect of segregation zones

    SciTech Connect (OSTI)

    Bethmont, M.; Frund, J.M. [Electricite de France, Moret-sur-Loing (France); Housin, B. [Framatome, Paris La Defense (France). Materials and Technology Dept.; Soulat, P. [Commissariat a l`Energie Atomique, Gif-sur-Yvette (France)

    1996-12-31T23:59:59.000Z

    To establish the integrity of pressure water reactor (PWR) vessels it is necessary to determine the toughness of A508Cl.3 steel at the end of its life, that is after thermal aging and irradiation embrittlement. In safety analyses the toughness can be deduced from a reference curve set forth in the code (ASME or RCC-M). The validity of the reference curve has been verified for several years for unirradiated French reactor vessels. Tests were performed on specimens taken from materials having heterogeneities in chemical composition. For most of the test results the reference curve is a lower bound. To solve te problem of determining the toughness of SA 508 Cl.3 steel after irradiation and in the presence of possible heterogeneities, the toughness results were gathered. The synthesis shows that the RCC-M code curve is conservative.

  12. FINAL REPORT - HYBRID-MIXING TESTS SUPPORTING THE CONCENTRATE RECEIPT VESSEL (CRV-VSL-00002A/2B) CONFIGURATION

    SciTech Connect (OSTI)

    GUERRERO, HECTORN.

    2004-09-01T23:59:59.000Z

    The Savannah River National Laboratory (SRNL) has performed scaled physical modeling of Pulse Jet Mixing Systems applicable to the Concentrate Receipt Vessel (CRV) of Hanford's Waste Treatment Plant (WTP) as part of the overall effort to validate pulse jet mixer (PJM) mixing in WTP vessels containing non-Newtonian fluids. The strategy developed by the Pulse Jet Mixing Task Team was to construct a quarter-scale model of the CRV, use a clear simulant to understand PJM mixing behavior, and down-select from a number of PJM configurations to a ''best design'' configuration. This ''best design'' would undergo final validation testing using a particulate simulant that has rheological properties closely similar to WTP waste streams. The scaled PJM mixing tests were to provide information on the operating parameters critical for the uniform movement (total mobilization) of these non-Newtonian slurries. Overall, 107 tests were performed during Phase I and Phase II testing.

  13. ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R- AND P-REACTOR VESSELS

    SciTech Connect (OSTI)

    Wiersma, B.

    2009-12-29T23:59:59.000Z

    The R- and P-reactor buildings were retired from service and are now being prepared for deactivation and decommissioning (D&D). D&D activities will consist primarily of immobilizing contaminated components and structures in a grout-like formulation. Aluminum corrodes very rapidly when it comes in contact with the alkaline grout materials and as a result produces hydrogen gas. To address this potential deflagration/explosion hazard, the Materials Science and Technology Directorate (MS&T) of the Savannah River National Laboratory (SRNL) has been requested to review and evaluate existing experimental and analytical studies of this issue to determine if any process constraints on the chemistry of the fill material and the fill operation are necessary. Various options exist for the type of grout material that may be used for D&D of the reactor vessels. The grout formulation options include ceramicrete (pH 6-8), low pH portland cement + silica fume grout (pH 10.4), or Portland cement grout (pH 12.5). The assessment concluded that either ceramicrete or the silica fume grout may be used to safely grout the P-reactor vessel. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Portland cement grout, on the other hand, for the same range of process parameters does not provide a significant margin of safety against the accumulation of flammable gas in the reactor vessel during grouting operations in the P-reactor vessel. It is recommended that this grout not be utilized for this task. The R-reactor vessel contains significantly less aluminum and thus a Portland cement grout may be considered as well. For example, if the grout fill rate is less than 1 inch/min and the grout temperature is maintained at 70 C or less, the risk of hydrogen accumulation in the R-reactor vessel is very low for the Portland cement. Alternatively, if the grout fill rate is less than 0.5 inch/min and the grout is maintained at a temperature of 80 C, the risk will again be very low. Although these calculations are conservative, there are some measures that may be taken to further minimize the potential for hydrogen evolution. (1) Minimize the temperature of the grout as much as practical. Lower temperatures will mean lower hydrogen generation rates. For P-reactor, grout temperatures less than 100 C should provide an adequate safety margin for the pH 8 and pH 10.4 grout formulations. For R-reactor, grout temperatures less than 70 C or 80 C will provide an adequate safety margin for the Portland cement. The other grout formulations are also viable options for R-reactor. (2) Minimize the grout fill rate as much as practical. Lowering the fill rate takes advantage of passivation of the aluminum components and hence lower hydrogen generation rates. For P-reactor, fill rates that are less than 2 inches/min for the ceramicrete and the silica fume grouts will reduce the chance of significant hydrogen accumulation. For R-reactor, fill rates less than 1 inch/min will again minimize the risk of hydrogen accumulation. (3) Ventilate the building as much as practical (e.g., leave doors open) to further disperse hydrogen. The volumetric hydrogen generation rates in the P-reactor vessel, however, are low for the pH 8 and pH 10.4 grout, (i.e., less than 0.32 ft{sup 3}/min). If further walk-down inspections of the reactor vessels suggest an increase in the actual areal density of aluminum, the calculations should be re-visited.

  14. Combination of Vessel-Targeting Agents and Fractionated Radiation Therapy: The Role of the SDF-1/CXCR4 Pathway

    SciTech Connect (OSTI)

    Chen, Fang-Hsin; Fu, Sheng-Yung [Department of Biomedical Engineering and Environmental Sciences, National Tsing Hua University, Taiwan (China)] [Department of Biomedical Engineering and Environmental Sciences, National Tsing Hua University, Taiwan (China); Yang, Ying-Chieh [Department of Radiation Oncology, National Taiwan University Hospital Hsin-Chu Branch, Taiwan (China)] [Department of Radiation Oncology, National Taiwan University Hospital Hsin-Chu Branch, Taiwan (China); Wang, Chun-Chieh [Department of Radiation Oncology, Chang Gung Memorial Hospital-LinKou, Taiwan (China) [Department of Radiation Oncology, Chang Gung Memorial Hospital-LinKou, Taiwan (China); Department of Medical Imaging and Radiological Science, Chang Gung University, Taiwan (China); Chiang, Chi-Shiun, E-mail: cschiang@mx.nthu.edu.tw [Department of Biomedical Engineering and Environmental Sciences, National Tsing Hua University, Taiwan (China)] [Department of Biomedical Engineering and Environmental Sciences, National Tsing Hua University, Taiwan (China); Hong, Ji-Hong, E-mail: jihong@adm.cgmh.org.tw [Department of Radiation Oncology, Chang Gung Memorial Hospital-LinKou, Taiwan (China) [Department of Radiation Oncology, Chang Gung Memorial Hospital-LinKou, Taiwan (China); Department of Medical Imaging and Radiological Science, Chang Gung University, Taiwan (China)

    2013-07-15T23:59:59.000Z

    Purpose: To investigate vascular responses during fractionated radiation therapy (F-RT) and the effects of targeting pericytes or bone marrow-derived cells (BMDCs) on the efficacy of F-RT. Methods and Materials: Murine prostate TRAMP-C1 tumors were grown in control mice or mice transplanted with green fluorescent protein-tagged bone marrow (GFP-BM), and irradiated with 60 Gy in 15 fractions. Mice were also treated with gefitinib (an epidermal growth factor receptor inhibitor) or AMD3100 (a CXCR4 antagonist) to examine the effects of combination treatment. The responses of tumor vasculatures to these treatments and changes of tumor microenvironment were assessed. Results: After F-RT, the tumor microvascular density (MVD) was reduced; however, the surviving vessels were dilated, incorporated with GFP-positive cells, tightly adhered to pericytes, and well perfused with Hoechst 33342, suggesting a more mature structure formed primarily via vasculogenesis. Although the gefitinib+F-RT combination affected the vascular structure by dissociating pericytes from the vascular wall, it did not further delay tumor growth. These tumors had higher MVD and better vascular perfusion function, leading to less hypoxia and tumor necrosis. By contrast, the AMD3100+F-RT combination significantly enhanced tumor growth delay more than F-RT alone, and these tumors had lower MVD and poorer vascular perfusion function, resulting in increased hypoxia. These tumor vessels were rarely covered by pericytes and free of GFP-positive cells. Conclusions: Vasculogenesis is a major mechanism for tumor vessel survival during F-RT. Complex interactions occur between vessel-targeting agents and F-RT, and a synergistic effect may not always exist. To enhance F-RT, using CXCR4 inhibitor to block BM cell influx and the vasculogenesis process is a better strategy than targeting pericytes by epidermal growth factor receptor inhibitor.

  15. THE IMPACT OF OZONE ON THE LOWER FLAMMABLE LIMIT OF HYDROGEN IN VESSELS CONTAINING SAVANNAH RIVER SITE HIGH LEVEL WASTE

    SciTech Connect (OSTI)

    Sherburne, Carol [Savannah River Site (SRS), Aiken, SC (United States). Savannah River Remediation, LLC; Osterberg, Paul [Fauske and Associates, LLC, Burr Ridge, IL (United States); Johnson, Tom [Fauske and Associates, LLC, Burr Ridge, IL (United States); Frawely, Thomas [Fauske and Associates, LLC, Burr Ridge, IL (United States)

    2013-01-23T23:59:59.000Z

    The Savannah River Site, in conjunction with AREVA Federal services, has designed a process to treat dissolved radioactive waste solids with ozone. It is known that in this radioactive waste process, radionuclides radiolytically break down water into gaseous hydrogen and oxygen, which presents a well defined flammability hazard. Flammability limits have been established for both ozone and hydrogen separately; however, there is little information on mixtures of hydrogen and ozone. Therefore, testing was designed to provide critical flammability information necessary to support safety related considerations for the development of ozone treatment and potential scale-up to the commercial level. Since information was lacking on flammability issues at low levels of hydrogen and ozone, a testing program was developed to focus on filling this portion of the information gap. A 2-L vessel was used to conduct flammability tests at atmospheric pressure and temperature using a fuse wire ignition source at 1 percent ozone intervals spanning from no ozone to the Lower Flammable Limit (LFL) of ozone in the vessel, determined as 8.4%(v/v) ozone. An ozone generator and ozone detector were used to generate and measure the ozone concentration within the vessel in situ, since ozone decomposes rapidly on standing. The lower flammability limit of hydrogen in an ozone-oxygen mixture was found to decrease from the LFL of hydrogen in air, determined as 4.2 % (v/v) in this vessel. From the results of this testing, Savannah River was able to develop safety procedures and operating parameters to effectively minimize the formation of a flammable atmosphere.

  16. Initial Evaluation of the Heat-Affected Zone, Local Embrittlement Phenomenon as it Applies to Nuclear Reactor Vessels

    SciTech Connect (OSTI)

    McCabe, D.E.

    1999-09-01T23:59:59.000Z

    The objective of this project was to determine if the local brittle zone (LBZ) problem, encountered in the testing of the heat-affected zone (HAZ) part of welds in offshore platform construction, can also be found in reactor pressure vessel (RPV) welds. Both structures have multipass welds and grain coarsening along the fusion line. Literature was obtained that described the metallurgical evidence and the type of research work performed on offshore structure welds.

  17. Prediction of failure behavior of a welded pressure vessel containing flaws during a hydrogen-charged burst test

    SciTech Connect (OSTI)

    Bhuyan, G.S. [Powertech Labs. Inc., Surrey, British Columbia (Canada); Sperling, E.J. [Amoco Corp., Naperville, IL (United States); Shen, G. [CANMET, Ottawa, Ontario (Canada). Metals Technology Labs.; Yin, H. [Mobil Research and Development Corp., Farmers Branch, TX (United States); Rana, M.D. [Praxair, Inc., Tonawanda, NY (United States)

    1996-12-01T23:59:59.000Z

    An industry-government collaborative program was carried out with an aim to promoting the acceptance of fracture mechanics based fitness-for-service assessment methodology for a service-damaged pressure vessel. A collaborative round robin exercise was carried out to predict the fracture behavior of a vessel containing hydrogen damage, fabrication related lack-of-fusion defects, an artificially induced fatigue crack and a localized thinned area. The fracture assessment procedures used include the US ASME Material Property Council`s PREFIS Program based on the British Standard (BS) Published Document (PD) 6493, ASME Section XI and The Central Electricity Generating Board (CEGB) R6 approach; The welding Institute (TWI) CRACKWISE program (based on BS PD6493 Level 2 approach), a variant of the R6 approach, J-tearing instability approaches, various J-estimation schemes, LEFM approach and simplified stress analysis. Assessments were compared with the results obtained from a hydrogen charged burst test of the vessel. Predictions, based on the J-tearing approach, compared well with the actual burst test results. Actual burst pressure was about five times the operating pressure.

  18. Prediction of failure behavior of a welded pressure vessel containing flaws during a hydrogen-charged burst test

    SciTech Connect (OSTI)

    Bhuyan, G.S. [Powertech Labs Inc., Surrey, British Columbia (Canada); Sperling, E.J. [BP-Amoco, Calgary, Alberta (Canada); Shen, G. [CANMET, Ottawa, Ontario (Canada). Metals Technology Labs.; Yin, H. [Mobil Technology Co., Dallas, TX (United States); Rana, M.D. [Praxair, Inc., Tonawanda, NY (United States)

    1999-08-01T23:59:59.000Z

    An industry-government collaborative program was carried out with an aim to promoting the acceptance of fracture mechanics-based fitness-for-service assessment methodology for a service-damaged pressure vessel. A collaborative round robin exercise was carried out to predict the fracture behavior of a vessel containing hydrogen damage, fabrication-related lack-of-fusion defects, an artificially induced fatigue crack, and a localized thinned area. The fracture assessment procedures used include the US ASME Material Property Council`s PREFIS Program based on the British Standard (BS) Published Document (PD) 6493, ASME Section XI and The Central Electricity Generating Board (CEGB) R6 approach, The Welding Institute (TWI) CRACKWISE program (based on BS PD6493 Level 2 approach), a variant of the R6 approach, J-tearing instability approaches, various J-estimation schemes, LEFM approach, and simplified stress analysis. Assessments were compared with the results obtained from a hydrogen-charged burst test of the vessel. Predictions, based on the J-tearing approach, compared well with the actual burst test results. Actual burst pressure was about five times the operating pressure.

  19. Peer review of the Three Mile Island Unit 2 Vessel Investigation Project metallurgical examinations

    SciTech Connect (OSTI)

    Bohl, R.W.; Gaydos, R.G.; Vander Voort, G.F.; Diercks, D.R. [Argonne National Lab., IL (United States)

    1994-07-01T23:59:59.000Z

    Fifteen samples recovered from the lower head of the Three Mile Island (TMI) Unit 2 nuclear reactor pressure vessel were subjected to detailed metallurgical examinations by the Idaho National Engineering Laboratory (INEL), with supporting work carried out by Argonne National Laboratory (ANL) and several of the European participants. These examinations determined that a portion of the lower head, a so-called elliptical ``hot spot`` measuring {approx}0.8 {times} 1 m, reached temperatures as high as 1100{degrees}C during the accident and cooled from these temperatures at {approx}10--100{degrees}C/min. The remainder of the lower head was found to have remained below the ferrite-toaustenite transformation temperature of 727{degrees}C during the accident. Because of the significance of these results and their importance to the overall analysis of the TMI accident, a panel of three outside peer reviewers, Dr. Robert W. Bohl, Mr. Richard G. Gaydos, and Mr. George F. Vander Voort, was formed to conduct an independent review of the metallurgical analyses. After a thorough review of the previous analyses and examination of photo-micrographs and actual lower head specimens, the panel determined that the conclusions resulting from the INEL study were fundamentally correct. In particular, the panel reaffirmed that four lower head samples attained temperatures as high as 1100{degrees}C, and perhaps as high as 1150--1200{degrees}C in one case, during the accident. They concluded that these samples subsequently cooled at a rate of {approx}50--125{degrees}C/min in the temperature range of 600--400{degrees}C, in good agreement with the original analysis. The reviewers also agreed that the remainder of the lower head samples had not exceeded the ferrite-to-austenite transformation temperature during the accident and suggested several refinements and alternative procedures that could have been employed in the original analysis.

  20. Response of Soviet-designed VVER-440 steam generator vessel to pressurization

    SciTech Connect (OSTI)

    Kennedy, J.M.; Sienicki, J.J.

    1989-01-01T23:59:59.000Z

    The Soviet-designed VVER (Water-Cooled, Water-Moderated Energy Reactors) pressurized water reactors use horizontal steam generators to transfer energy from the primary to secondary coolant systems (DOE/NE-0084 Revision 2, 1989). Primary coolant flowing from the reactor vessel enters the steam generator through a vertical, circular, manifold header that also serves as the tubesheet distributing coolant to the horizontal tube bundle. Primary coolant exits the tube bundle and steam generator through a second similar vertical manifold header. The header design includes the provision for access by a person to inspect the mainfolds through bolted down closure heads atop each manifold. The internal diameter of each header exceeds that of the connected primary coolant system piping. The postulated failure of a manifold closure head or the manifold itself provides a pathway for primary coolant to enter the secondary system. Steam formation due to flashing of primary coolant inside the steam generator secondary side region can result in pressurization of the steam generator shell to values above the nominal secondary side operating pressure. The present work involves the investigation of the consequences of manifold failure for the case of the VVER-440 reactor system. An analysis has been performed of the loadings upon and the mechanical response of the steam generator shell for the case of a postulated large break in the manifold wall. The objectives were to calculate the maximum pressure attained inside the shell and to predict the shell failure pressure as well as the failure mechanism. 6 refs., 8 figs., 1 tab.

  1. The Information Fusion Embrittlement Models for U.S. Power Reactor Pressure Vessel Steels

    SciTech Connect (OSTI)

    Wang, Jy-An John [ORNL; Rao, Nageswara S [ORNL; Konduri, Savanthi [AOL

    2007-01-01T23:59:59.000Z

    The complex nonlinear dependencies observed in typical reactor pressure vessel (RPV) material embrittlement data, as well as the inherent large uncertainties and scatter in the radiation embrittlement data, make prediction of radiation embrittlement a difficult task. Conventional statistical and deterministic approaches have only resulted in rather large uncertainties, in part because they do not fully exploit domain-specific mechanisms. The domain models built by researchers in the field, on the other hand, do not fully exploit the statistical and information content of the data. As evidenced in previous studies, it is unlikely that a single method, whether statistical, nonlinear, or domain model, will outperform all others. More generally, considering the complexity of the embrittlement prediction problem, it is highly unlikely that a single best method exists and is tractable, even in theory. In this paper, we propose to combine a number of complementary methods including domain models, neural networks, and nearest neighbor regressions (NNRs). Such a combination of methods has become possible because of recent developments in measurement-based optimal fusers in the area of information fusion. The information fusion technique is used to develop radiation embrittlement prediction models for reactor RPV steels from U.S. power reactors, including boiling water reactors and pressurized water reactors. The Charpy transition temperature-shift data is used as the primary index of RPV radiation embrittlement in this study. Six Cu, Ni, P, neutron fluence, irradiation time, and irradiation-parameters are used in the embrittlement prediction models. The results-temperature indicate that this new embrittlement predictor achieved reductions of about 49.5% and 52% in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The implications of dose-rate effect and irradiation temperature effects for the development of radiation embrittlement models are also discussed.

  2. An investigation of temperature measurement methods in nuclear power plant reactor pressure vessel annealing

    SciTech Connect (OSTI)

    Acton, R.U.; Gill, W.; Sais, D.J.; Schulze, D.H.; Nakos, J.T. [Sandia National Labs., Albuquerque, NM (United States)

    1996-05-01T23:59:59.000Z

    The objective of this project was to provide an assessment of several methods by which the temperature of a commercial nuclear power plant reactor pressure vessel (RPV) could be measured during an annealing process. This project was a coordinated effort between DOE`s Office of Nuclear Energy, Science and Technology; DOE`s Light Water Reactor Technology Center at Sandia National Laboratories; and the Electric Power Research Institute`s Non- Destructive Evaluation Center. Ball- thermocouple probes similar to those described in NUREG/CR-5760, spring-loaded, metal- sheathed thermocouple probes, and 1778 air- suspended thermocouples were investigated in experiments that heated a section of an RPV wall to simulate a thermal annealing treatment. A parametric study of ball material, emissivity, thermal conductivity, and thermocouple function locations was conducted. Also investigated was a sheathed thermocouple failure mode known as shunting (electrical breakdown of insulation separating the thermocouple wires). Large errors were found between the temperature as measured by the probes and the true RPV wall temperature during heat-up and cool-down. At the annealing soak temperature, in this case 454{degrees}C [850`F], all sensors measured the same temperature within about {plus_minus}5% (23.6{degrees}C [42.5{degrees}F]). Because of these errors, actual RPV wall heating and cooling rates differed from those prescribed (by up to 29%). Shunting does not appear to be a problem under these conditions. The large temperature measurement errors led to the development of a thermal model that predicts the RPV wall temperature from the temperature of a ball- probe. Comparisons between the model and the experimental data for ball-probes indicate that the model could be a useful tool in predicting the actual RPV temperature based on the indicated ball- probe temperature. The model does not predict the temperature as well for the spring-loaded and air suspended probes.

  3. SPR salt wall leaching experiments in lab-scale vessel : data report.

    SciTech Connect (OSTI)

    Webb, Stephen Walter; O'Hern, Timothy John; Hartenberger, Joel David

    2010-10-01T23:59:59.000Z

    During cavern leaching in the Strategic Petroleum Reserve (SPR), injected raw water mixes with resident brine and eventually interacts with the cavern salt walls. This report provides a record of data acquired during a series of experiments designed to measure the leaching rate of salt walls in a labscale simulated cavern, as well as discussion of the data. These results should be of value to validate computational fluid dynamics (CFD) models used to simulate leaching applications. Three experiments were run in the transparent 89-cm (35-inch) ID diameter vessel previously used for several related projects. Diagnostics included tracking the salt wall dissolution rate using ultrasonics, an underwater camera to view pre-installed markers, and pre- and post-test weighing and measuring salt blocks that comprise the walls. In addition, profiles of the local brine/water conductivity and temperature were acquired at three locations by traversing conductivity probes to map out the mixing of injected raw water with the surrounding brine. The data are generally as expected, with stronger dissolution when the salt walls were exposed to water with lower salt saturation, and overall reasonable wall shape profiles. However, there are significant block-to-block variations, even between neighboring salt blocks, so the averaged data are considered more useful for model validation. The remedial leach tests clearly showed that less mixing and longer exposure time to unsaturated water led to higher levels of salt wall dissolution. The data for all three tests showed a dividing line between upper and lower regions, roughly above and below the fresh water injection point, with higher salt wall dissolution in all cases, and stronger (for remedial leach cases) or weaker (for standard leach configuration) concentration gradients above the dividing line.

  4. CLOSURE REPORT FOR CORRECTIVE ACTION UNIT 204: STORAGE BUNKERS, NEVADA TEST SITE, NEVADA

    SciTech Connect (OSTI)

    NONE

    2006-04-01T23:59:59.000Z

    Corrective Action Unit (CAU) 330 consists of four Corrective Action Sites (CASs) located in Areas 6, 22, and 23 of the Nevada Test Site (NTS). The unit is listed in the Federal Facility Agreement and Consent Order (FFACO, 1996) as CAU 330: Areas 6, 22, and 23 Tanks and Spill Sites. CAU 330 consists of the following CASs: CAS 06-02-04, Underground Storage Tank (UST) and Piping CAS 22-99-06, Fuel Spill CAS 23-01-02, Large Aboveground Storage Tank (AST) Farm CAS 23-25-05, Asphalt Oil Spill/Tar Release

  5. CLOSURE REPORT FOR CORRECTIVE ACTION UNIT 214: BUNKERS AND STORAGE AREAS NEVADA TEST SITE, NEVADA

    SciTech Connect (OSTI)

    NONE

    2006-09-01T23:59:59.000Z

    The purpose of this Closure Report is to document that the closure of CAU 214 complied with the Nevada Division of Environmental Protection-approved Corrective Action Plan closure requirements. The closure activities specified in the Corrective Action Plan were based on the approved corrective action alternatives presented in the CAU 214 Corrective Action Decision Document.

  6. The Ozark Uplift and related structural elements, a view from the north bunker

    SciTech Connect (OSTI)

    Bunker, B.J.; Witzke, B.J.; Ludvigson, G.A. (Iowa Dept. of Natural Resources, Iowa City, IA (United States). Geological Survey Bureau)

    1993-03-01T23:59:59.000Z

    The structural/stratigraphic framework of the central Midcontinent region has been evaluated utilizing a series of isopach and paleogeographic maps constructed within the framework of Sloss' (1963) sequences. General northward-trending Sauk structural patterns were replaced by more easterly-trending patterns across Iowa and Illinois during the M, and U. Ordovician, coincident with initial upwarping of the Ozark Uplift. Silurian strata in this region are largely restricted to the East-Central Iowa and North Kansas basins and the east-west trending structural sag connecting these two basins. The structural framework influencing early Kaskaskia deposition was largely inherited from that which developed during the late Tippecanoe. The Iowa Basin, which formed during the late Middle to Late Devonian, represents an intrashelf basin which developed on the Midcontinent Carbonate shelf in which shallow-water and mudflat sedimentation kept pace with increased subsidence. The present thickness of Mississippian rocks in the Midcontinent reflects extensive pre-Absaroka uplift and erosion. Structural deformation during the Early-Middle Pennsylvanian (Nemaha Uplift) bisected the region of the North Kansas Basin and cut off the southwestern extension of the Kaskaskia Iowa Basin. Up to 320 m of pre-Missourian rocks accumulated in the structural depression east of the Humboldt Fault Zone (Forest City Basin). The Forest City Basin was a relatively short-lived asymmetric fault-bounded sedimentary basin that subsided in synchrony with the ascension of the Nemaha Uplift. Earlier and subsequent Phanerozoic sedimentation in the area occurred in short-lived depositional basins whose structural geometries were strikingly dissimilar.

  7. Lucrative Opportunities in Asia Pacific to Help Global Bunker Fuel Market

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are beingZealand Jump to: navigation, searchOf KilaueaInformation Other4Q07) WindLow VoltageGroup Jump

  8. Blood flow measurement and slow flow detection in retinal vessels with Joint Spectral and Time domain method in ultrahigh speed OCT

    E-Print Network [OSTI]

    Gorczynska, I.

    We present an application of the Joint Spectral and Time domain OCT (STdOCT) method for detection of wide range of flows in the retinal vessels. We utilized spectral/Fourier domain OCT (SOCT) technique for development of ...

  9. Multimodal Vessel Visualization of Mouse Aorta PET/CT Scans Timo Ropinski, Member, IEEE, Sven Hermann, Rainer Reich, Michael Schafers, and Klaus Hinrichs, Member, IEEE

    E-Print Network [OSTI]

    Hinrichs, Klaus

    Multimodal Vessel Visualization of Mouse Aorta PET/CT Scans Timo Ropinski, Member, IEEE, Sven present a visualization system for the visual analysis of PET/CT scans of aortic arches of mice

  10. EXPERIMENTAL STUDY OF CRITICAL HEAT FLUX WITH ALUMINA-WATER NANOFLUIDS IN DOWNWARD-FACING CHANNELS FOR IN-VESSEL RETENTION APPLICATIONS

    E-Print Network [OSTI]

    Park, R.J.

    The Critical Heat Flux (CHF) of water with dispersed alumina nanoparticles was measured for the geometry and flow conditions relevant to the In-Vessel Retention (IVR) situation which can occur during core melting sequences ...

  11. Coxsackie- and adenovirus receptor (CAR) is expressed in lymphatic vessels in human skin and affects lymphatic endothelial cell function in vitro

    SciTech Connect (OSTI)

    Vigl, Benjamin; Zgraggen, Claudia; Rehman, Nadia; Banziger-Tobler, Nadia E.; Detmar, Michael [Institute of Pharmaceutical Sciences, Swiss Federal Institute of Technology, ETH Zurich, Wolfgang-Pauli Str. 10, CH-8093 Zurich (Switzerland); Halin, Cornelia [Institute of Pharmaceutical Sciences, Swiss Federal Institute of Technology, ETH Zurich, Wolfgang-Pauli Str. 10, CH-8093 Zurich (Switzerland)], E-mail: cornelia.halin@pharma.ethz.ch

    2009-01-15T23:59:59.000Z

    Lymphatic vessels play an important role in tissue fluid homeostasis, intestinal fat absorption and immunosurveillance. Furthermore, they are involved in pathologic conditions, such as tumor cell metastasis and chronic inflammation. In comparison to blood vessels, the molecular phenotype of lymphatic vessels is less well characterized. Performing comparative gene expression analysis we have recently found that coxsackie- and adenovirus receptor (CAR) is significantly more highly expressed in cultured human, skin-derived lymphatic endothelial cells (LECs), as compared to blood vascular endothelial cells. Here, we have confirmed these results at the protein level, using Western blot and FACS analysis. Immunofluorescence performed on human skin confirmed that CAR is expressed at detectable levels in lymphatic vessels, but not in blood vessels. To address the functional significance of CAR expression, we modulated CAR expression levels in cultured LECs in vitro by siRNA- and vector-based transfection approaches. Functional assays performed with the transfected cells revealed that CAR is involved in distinct cellular processes in LECs, such as cell adhesion, migration, tube formation and the control of vascular permeability. In contrast, no effect of CAR on LEC proliferation was observed. Overall, our data suggest that CAR stabilizes LEC-LEC interactions in the skin and may contribute to lymphatic vessel integrity.

  12. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

    SciTech Connect (OSTI)

    Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

    2007-03-21T23:59:59.000Z

    In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for both the PBMR and prismatic design. The main focus of this report is the RPV for both design concepts with emphasis on material selection.

  13. The DOS 1 neutron dosimetry experiment at the HB-4-A key 7 surveillance site on the HFIR pressure vessel

    SciTech Connect (OSTI)

    Farrell, K.; Kam, F.B.; Baldwin, C.A. [and others

    1994-01-01T23:59:59.000Z

    A comprehensive neutron dosimetry experiment was made at one of the prime surveillance sites at the High Flux Isotope Reactor (HFIR) pressure vessel to aid radiation embrittlement studies of the vessel and to benchmark neutron transport calculations. The thermal neutron flux at the key 7, position 5 site was found, from measurements of radioactivation of four cobalt wires and four silver wires, to be 2.4 {times} 10{sup 12} n{center_dot}m{sup {minus}2}{center_dot}s{sup {minus}1}. The thermal flux derived from two helium accumulation monitors was 2.3 {times} 10{sup 12} n{center_dot}m{sup {minus}2}{center_dot}{sup {minus}1}. The thermal flux estimated by neutron transport calculations was 3.7 {times} 10{sup 12} n{center_dot}m{sup {minus}2}s{sup {minus}1}. The fast flux, >1 MeV, determined from two nickel activation wires, was 1.5 {times} 10{sup 12} n{center_dot}m{sup {minus}2}{center_dot}s{sup {minus}1}, in keeping with values obtained earlier from stainless steel surveillance monitors and with a computed value of 1.2 {times} 10{sup 13} n{center_dot}m{sup {minus}2}{center_dot}{sup {minus}1}. The fast fluxes given by two reaction-product-type monitors, neptunium-237 and beryllium, were 2.6 {times} 10{sup 13} n{center_dot}m{sup {minus}2}{center_dot}s {sup {minus}1} and 2.2 {times} 10{sup 13} n{center_dot}m{sup {minus}2}s{sup {minus}1}, respectively. Follow-up experiments indicate that these latter high values of fast flux are reproducible but are false; they are due to the creation of greater levels of reaction products by photonuclear events induced by an exceptionally high ratio of gamma flux to fast neutron flux at the vessel.

  14. Prediction and Monitoring Systems of Creep-Fracture Behavior of 9Cr-1Mo Steels for Teactor Pressure Vessels

    SciTech Connect (OSTI)

    Potirniche, Gabriel; Barlow, Fred D.; Charit, Indrajit; Rink, Karl

    2013-11-26T23:59:59.000Z

    A recent workshop on next-generation nuclear plant (NGNP) topics underscored the need for research studies on the creep fracture behavior of two materials under consideration for reactor pressure vessel (RPV) applications: 9Cr-1Mo and SA-5XX steels. This research project will provide a fundamental understanding of creep fracture behavior of modified 9Cr-1Mo steel welds for through modeling and experimentation and will recommend a design for an RPV structural health monitoring system. Following are the specific objectives of this research project: • Characterize metallurgical degradation in welded modified 9Cr-1Mo steel resulting from aging processes and creep service conditions. • Perform creep tests and characterize the mechanisms of creep fracture process. • Quantify how the microstructure degradation controls the creep strength of welded steel specimens. • Perform finite element (FE) simulations using polycrystal plasticity to understand how grain texture affects the creep fracture properties of welds. • Develop a microstructure-based creep fracture model to estimate RPVs service life . • Manufacture small, prototypic, cylindrical pressure vessels, subject them to degradation by aging, and measure their leak rates. • Simulate damage evolution in creep specimens by FE analyses. • Develop a model that correlates gas leak rates from welded pressure vessels with the amount of microstructural damage. • Perform large-scale FE simulations with a realistic microstructure to evaluate RPV performance at elevated temperatures and creep strength. • Develop a fracture model for the structural integrity of RPVs subjected to creep loads. • Develop a plan for a non-destructive structural health monitoring technique and damage detection device for RPVs.

  15. Microstructure and mechanical properties of WWER-440 reactor vessel metal after service life expiration and recovery anneal

    SciTech Connect (OSTI)

    Gorynin, I.V.; Nesterova, E.V.; Nikolaev, V.A.; Rybin, V.V. [Central Research Inst. of Structural Materials Prometey, St. Petersburg (Russian Federation)

    1996-12-31T23:59:59.000Z

    The microstructure of base and weld metals (st. 15kH2MFA) of Novovoronezh Nuclear Power Plant Units 1 and 4 reactor vessels was studied after service life expiration and recovery anneal by means of light metallography and transmission electron microscopy. The qualitative characteristics of flow structure were determined. The estimates were made for the contributions of different flows to the radiation hardening and its total value. The conclusion was made that the marked difference in the mechanical properties of irradiated weld and base metals can be caused either by different original structure conditions or by the difference of alloying and impurity elements content.

  16. Surveillance program for WWER-440/Type 213 reactor pressure vessels -- Standard program, re-evaluation of results, supplementary program

    SciTech Connect (OSTI)

    Brumovsky, M.; Novosad, P.; Zdarek, J. [Nuclear Research Inst. Rez plc (Czech Republic)

    1996-12-31T23:59:59.000Z

    Irradiation embrittlement of the reactor pressure vessel beltline materials of WWER-440/Type 213 reactors is monitored by a material irradiation surveillance program. Due to the high lead factor, the duration of the standard surveillance program is only five years, after which no further surveillance samples remain in the reactor. The large variation and uncertainty in neutron flux over the irradiated materials produce significant scatter in mechanical properties and necessitate a re-evaluation of results using gamma scanning, specimen reconstitution and recalculation. In order to provide information on the impact of changes in plant operation during later years a supplementary surveillance program has been devised.

  17. Engineering Evaluation/Cost Analysis for Power Burst Facility (PER-620) Final End State and PBF Vessel Disposal

    SciTech Connect (OSTI)

    B. C. Culp

    2007-05-01T23:59:59.000Z

    Preparation of this engineering evaluation/cost analysis is consistent with the joint U.S. Department of Energy and U.S. Environmental Protection Agency Policy on Decommissioning of Department of Energy Facilities Under the Comprehensive Environmental Response, Compensation, and Liability Act, (DOE and EPA 1995) which establishes the Comprehensive Environmental, Response, Compensation, and Liability Act non-time critical removal action process as an approach for decommissioning. The scope of this engineering evaluation/cost analysis is to evaluate alternatives and recommend a preferred alternative for the final end state of the PBF and the final disposal location for the PBF vessel.

  18. The influence of selected containment structures on debris dispersal and transport following high pressure melt ejection from the reactor vessel

    SciTech Connect (OSTI)

    Pilch, M.; Tarbell, W.W.; Brockmann, J.E.

    1988-09-01T23:59:59.000Z

    High pressure expulsion of molten core debris from the reactor pressure vessel may result in dispersal of the debris from the reactor cavity. In most plants, the cavity exits into the containment such that the debris impinges on structures. Retention of the debris on the structures may affect the further transport of the debris throughout the containment. Two tests were done with scaled structural shapes placed at the exit of 1:10 linear scale models of the Zion cavity. The results show that the debris does not adhere significantly to structures. The lack of retention is attributed to splashing from the surface and reentrainment in the gas flowing over the surface. These processes are shown to be applicable to reactor scale. A third experiment was done to simulate the annular gap between the reactor vessel and cavity wall. Debris collection showed that the fraction of debris exiting through the gap was greater than the gap-to-total flow area ratio. Film records indicate that dispersal was primarily by entrainment of the molten debris in the cavity. 29 refs., 36 figs., 11 tabs.

  19. Effect of silicon on ultra-low temperature toughness of Nb–Ti microalloyed cryogenic pressure vessel steels

    SciTech Connect (OSTI)

    Qiu, J.A. [Hubei Collaborative Innovation Center for Advanced Steels, International Research Institute for Steel Technology, Wuhan University of Science and Technology, Wuhan 430081 (China); Wu, K.M., E-mail: wukaiming2000@yahoo.com [Hubei Collaborative Innovation Center for Advanced Steels, International Research Institute for Steel Technology, Wuhan University of Science and Technology, Wuhan 430081 (China); Li, J.H. [Hubei Collaborative Innovation Center for Advanced Steels, International Research Institute for Steel Technology, Wuhan University of Science and Technology, Wuhan 430081 (China); Research and Development Center of WISCO, Wuhan 430080 (China); Hodgson, P.D. [Hubei Collaborative Innovation Center for Advanced Steels, International Research Institute for Steel Technology, Wuhan University of Science and Technology, Wuhan 430081 (China); Institute for Frontier Materials, Deakin University, Geelong, Victoria 3220 (Australia); Hou, T.P. [Hubei Collaborative Innovation Center for Advanced Steels, International Research Institute for Steel Technology, Wuhan University of Science and Technology, Wuhan 430081 (China); Ding, Q.F. [Research and Development Center of WISCO, Wuhan 430080 (China)

    2013-09-15T23:59:59.000Z

    The effect of Si on the ultra-low temperature toughness of Nb–Ti microalloyed cryogenic pressure vessel steels was investigated by electron back-scattered diffraction and transmission electron microscope with energy dispersive spectroscopy. Equiaxed ferrite and bainite were obtained in the tempered steels with small Si additions. Nanosized Nb–Ti carbides (< 10 nm) were formed in the steel containing 0.05% Si, whereas much coarser carbides (> 30 nm) were found in the steel containing 0.47% Si. The ultra-low temperature toughness of the Nb–Ti microalloyed cryogenic pressure vessel steel was remarkably enhanced by the reduction in the Si content, which was attributed to the pre-existing iron carbide formation before the precipitation of nanosized Nb–Ti carbides during tempering. - Highlights: • Nanosized Nb-Ti carbides formed in the tempered steel with smaller Si addition. • Coarser Nb-Ti carbides formed in the tempered steel with more Si addition. • Pre-existing cememtites provide nucleation sites for Nb-Ti carbide precipitation. • Ultra-low temperature toughness was remarkably enhanced by Si content reduction.

  20. The B and W Owners Group program for microstructural characterization and radiation embrittlement modelling of Linde 80 reactor vessel welds

    SciTech Connect (OSTI)

    Pavinich, W.A. [Grove Engineering, Knoxville, TN (United States); Harbison, L.S. [B and W Nuclear Technologies, Lynchburg, VA (United States)

    1996-12-31T23:59:59.000Z

    The Babcock and Wilcox Owners Group (B and WOG) is embrittlement of Linde 80 reactor vessel welds from a micro-mechanical viewpoint. Previous work that focused on characterizing the large microstructural features indicated that a large portion of the bulk copper content is in precipitate/inclusion/carbide form. This result indicates that copper in solid solution is considerably less than the bulk composition. Field-ion microscope atom probe investigations on unirradiated weld metals with bulk copper contents ranging from 0.22 to 0.38 wt%, also indicate significant amount of copper are tied up in precipitate/inclusion/carbide form. This results is significant since the bulk copper content (which includes both copper in solid solution and copper contained in precipitates, inclusions, and carbides) is used in Regulatory Guide 1.99, Revision 2 to determine radiation damage. This paper reviews these results. Existing radiation embrittlement models superpose the changes in yield strength due to defect clusters and copper-rich precipitates induced by neutron irradiation. Low-copper Linde 80 welds display little or no increase in the 41 joule (30 ft-lb) transition temperature as a result of neutron irradiation which indicates that precipitation is the dominant component of radiation embrittlement for Linde 80 welds. Future work will include further microstructural characterizations of Linde 80 reactor vessel welds and applying the existing radiation embrittlement models to Linde 80 welds. This paper describes the detailed plans for future work.

  1. Modeling the Ductile Brittle Fracture Transition in Reactor Pressure Vessel Steels using a Cohesive Zone Model based approach

    SciTech Connect (OSTI)

    Pritam Chakraborty; S. Bulent Biner

    2013-10-01T23:59:59.000Z

    Fracture properties of Reactor Pressure Vessel (RPV) steels show large variations with changes in temperature and irradiation levels. Brittle behavior is observed at lower temperatures and/or higher irradiation levels whereas ductile mode of failure is predominant at higher temperatures and/or lower irradiation levels. In addition to such temperature and radiation dependent fracture behavior, significant scatter in fracture toughness has also been observed. As a consequence of such variability in fracture behavior, accurate estimates of fracture properties of RPV steels are of utmost importance for safe and reliable operation of reactor pressure vessels. A cohesive zone based approach is being pursued in the present study where an attempt is made to obtain a unified law capturing both stable crack growth (ductile fracture) and unstable failure (cleavage fracture). The parameters of the constitutive model are dependent on both temperature and failure probability. The effect of irradiation has not been considered in the present study. The use of such a cohesive zone based approach would allow the modeling of explicit crack growth at both stable and unstable regimes of fracture. Also it would provide the possibility to incorporate more physical lower length scale models to predict DBT. Such a multi-scale approach would significantly improve the predictive capabilities of the model, which is still largely empirical.

  2. ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R AND P REACTOR VESSELS

    SciTech Connect (OSTI)

    Wiersma, B.

    2009-10-29T23:59:59.000Z

    The R- and P-reactor buildings were retired from service and are now being prepared for deactivation and decommissioning (D&D). D&D activities will consist primarily of immobilizing contaminated components and structures in a grout-like formulation. Aluminum corrodes very rapidly when it comes in contact with the alkaline grout materials and as a result produces hydrogen gas. To address this potential deflagration/explosion hazard, the Materials Science and Technology Directorate (MS&T) of the Savannah River National Laboratory (SRNL) has been requested to review and evaluate existing experimental and analytical studies of this issue to determine if any process constraints on the chemistry of the fill material and the fill operation are necessary. Various options exist for the type of grout material that may be used for D&D of the reactor vessels. The grout formulation options include ceramicrete (pH 6-8), low pH portland cement + silica fume grout (pH 10.4), or portland cement grout (pH 12.5). The assessment concluded that either ceramicrete or the silica fume grout may be used to safely grout the R- and P- reactor vessels. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Conservative calculations estimate that either ceramicrete or the silica fume grout may be used to safely grout the R- and P- reactor vessels. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Although these calculations are conservative, there are some measures that may be taken to further minimize the potential for hydrogen evolution. (1) Minimize the temperature of the grout as much as practical. Lower temperatures will mean lower hydrogen generation rates. Grout temperatures less than 100 C should however, still provide an adequate safety margin for the pH 8 and pH 10.4 grout formulations. (2) Minimize the fill rate as much as practical. Lowering the fill rate takes advantage of passivation of the aluminum components and hence lower hydrogen generation rates. Fill rates that are less than 2 inches/min will reduce the chance of significant hydrogen build-up. (3) Ventilate the building as much as practical (e.g., leave doors open) to further disperse hydrogen. The volumetric hydrogen generation rates however, are low for the pH 8 and pH 10.4 grout, i.e., less than 0.32 ft{sup 3}/min. Portland cement grout, on the other hand, for the same range of process parameters does not provide a significant margin of safety against the accumulation of flammable gas in the reactor vessel during grouting operations. It is recommended that this grout not be utilized for this task. If further walk-down inspections of the reactor vessels suggest an increase in the actual areal density of aluminum, the calculations should be re-visited.

  3. Assessment of Negligible Creep, Off-Normal Welding and Heat Treatment of Gr91 Steel for Nuclear Reactor Pressure Vessel Application

    SciTech Connect (OSTI)

    Ren, Weiju [ORNL; Terry, Totemeier [Idaho National Laboratory (INL)

    2006-10-01T23:59:59.000Z

    Two different topics of Grade 91 steel are investigated for Gen IV nuclear reactor pressure vessel application. On the first topic, negligible creep of Grade 91 is investigated with the motivation to design the reactor pressure vessel in negligible creep regime and eliminate costly surveillance programs during the reactor operation. Available negligible creep criteria and creep strain laws are reviewed, and new data needs are evaluated. It is concluded that modifications of the existing criteria and laws, together with their associated parameters, are needed before they can be reliably applied to Grade 91 for negligible creep prediction and reactor pressure vessel design. On the second topic, effects of off-normal welding and heat treatment on creep behavior of Grade 91 are studied with the motivation to better define the control over the parameters in welding and heat treatment procedures. The study is focused on off-normal austenitizing temperatures and improper cooling after welding but prior to post-weld heat treatment.

  4. Does Mechanical Thrombectomy in Acute Embolic Stroke Have Long-term Side Effects on Intracranial Vessels? An Angiographic Follow-up Study

    SciTech Connect (OSTI)

    Kurre, Wiebke, E-mail: w.kurre@klinikum-stuttgart.de; Perez, Marta Aguilar; Horvath, Diana [Klinikum Stuttgart, Klinik fuer Diagnostische und Interventionelle Neuroradiologie (Germany); Schmid, Elisabeth; Baezner, Hansjoerg [Klinikum Stuttgart, Neurologische Klinik (Germany); Henkes, Hans, E-mail: HHHenkes@aol.com [Klinikum Stuttgart, Klinik fuer Diagnostische und Interventionelle Neuroradiologie (Germany)

    2013-06-15T23:59:59.000Z

    Purpose. Mechanical thrombectomy (mTE) proved to be effective treating acute vessel occlusions with an acceptable rate of procedural complications. Potential long-term side effects of the vessel wall trauma caused by mechanical irritation of the endothelium are unknown up to now. Methods. From a retrospectively established database of 640 acute stroke treatments, we selected 261 patients with 265 embolic vessel occlusions treated successfully by mTE without permanent implantation of a stent. Analysis comprised the type of devices used and the number of passes performed. Digital subtraction angiography immediately after treatment was evaluated for vasospasm, dissection, and extravasation. Control angiographic images were evaluated for any morphological change compared to the immediate posttreatment angiographic run. Results. Recanalization was achieved with a median of one (range 1-10) mTE maneuvers. Vasospasm occurred in 69 territories (26.0 %) and was treated with glyceroltrinitrate in three. Dissection was observed in one vessel (0.4 %). Intraprocedural hemorrhage in two patients (0.8 %) was either wire or device induced. Follow-up digital subtraction angiography was available for 117 territories after a median of 107 days, revealing target vessel occlusion in one segment (0.9 %) and a de novo stenosis of four segments (3.4 %). All findings were clinically asymptomatic. Posttreatment vasospasm was more frequent in patients with de novo stenosis and occlusion (p = 0.038). Conclusion. De novo stenoses and occlusions occur in a small proportion of patients after mTE. Because all lesions were clinically asymptomatic, this finding does not affect the overall benefit of the treatment. Vasospasm may predict late vessel wall changes.

  5. Experimental results of direct containment heating by high-pressure melt ejection into the Surtsey vessel: The DCH-3 and DCH-4 tests

    SciTech Connect (OSTI)

    Allen, M.D.; Pilch, M.; Brockmann, J.E.; Tarbell, W.W. (Sandia National Labs., Albuquerque, NM (United States)); Nichols, R.T. (Ktech Corp., Albuquerque, NM (United States)); Sweet, D.W. (AEA Technology, Winfrith (United Kingdom))

    1991-08-01T23:59:59.000Z

    Two experiments, DCH-3 and DCH-4, were performed at the Surtsey test facility to investigate phenomena associated with a high-pressure melt ejection (HPME) reactor accident sequence resulting in direct containment heating (DCH). These experiments were performed using the same experimental apparatus with identical initial conditions, except that the Surtsey test vessel contained air in DCH-3 and argon in DCH-4. Inerting the vessel with argon eliminated chemical reactions between metallic debris and oxygen. Thus, a comparison of the pressure response in DCH-3 and DCH-4 gave an indication of the DCH contribution due to metal/oxygen reactions. 44 refs., 110 figs., 43 tabs.

  6. Control of Plasma-Stored Energy for Burn Control using DIII-D In-Vessel Coils

    SciTech Connect (OSTI)

    Hawryluk, R. J. [PPPL; Eidietis, N. W. [General Atomics; Grierson, B. A. [PPPL; Hyatt, A. W. [General Atomics; Koleman, E. [PPPL; Logan, N. C. [PPPL; Nazikian, R. [PPPL; Paz-Soldan, C. [General Atomics; Wolf, S. [MIT

    2014-09-01T23:59:59.000Z

    A new approach has been experimentally demonstrated to control the stored energy by applying a non-axisymmetric magnetic field using the DIII-D in-vessel coils to modify the energy confinement time. In future burning plasma experiments as well as magnetic fusion energy power plants, various concepts have been proposed to control the fusion power. The fusion power in a power plant operating at high gain can be related to the plasma-stored energy and hence, is a strong function of the energy confinement time. Thus, an actuator, that modifies the confinement time, can be used to adjust the fusion power. In relatively low collisionality DIII-D discharges, the application of non-axisymmetric magnetic fields results in a decrease in confinement time and density pumpout. Gas puffing was used to compensate the density pumpout in the pedestal while control of the stored energy was demonstrated by the application of non-axisymmetric fields.

  7. Ex-vessel melt-coolant interactions in deep water pool: Studies and accident management for Swedish BWRs

    SciTech Connect (OSTI)

    Sienicki, J.J.; Chu, C.C.; Spencer, B.W. (Argonne National Lab., IL (United States)); Frid, W. (Swedish Nuclear Power Inspectorate, Stockholm (Sweden)); Loewenhielm, G. (Vattenfall AB, Vaellingby (Sweden))

    1993-01-01T23:59:59.000Z

    In Swedish BWRs having an annular suppression pool, the lower drywell beneath the reactor vessel is flooded with water to mitigate against the effects of melt release into the drywell during a severe accident. The THIRMAL code has been used to analyze the effectiveness of the water pool to protect lower drywell penetrations by fragmenting and quenching the melt as it relocates downward through the water. Experiments have also been performed to investigate the benefits of adding surfactants to the water to reduce the likelihood of fine-scale debris formation from steam explosions. This paper presents an overview of the accident management approach and surfactant investigations together with results from the THIRMAL analyses.

  8. Fracture toughness testing of Linde 1092 reactor vessel welds in the transition range using Charpy-sized specimens

    SciTech Connect (OSTI)

    Pavinich, W.A. [Framatome Technologies Inc., Knoxville, TN (United States); Yoon, K.K. [Framatome Technologies Inc., Lynchburg, VA (United States); Hour, K.Y. [Babcock and Wilcox Co., Lynchburg, VA (United States). Research and Development Div.; Hoffman, C.L. [ABB-CE, Windsor, CT (United States)

    1999-10-01T23:59:59.000Z

    The present reference toughness method for predicting the change in fracture toughness can provide over estimates of these values because of uncertainties in initial RT{sub NDT} and shift correlations. It would be preferable to directly measure fracture toughness. However, until recently, no standard method was available to characterize fracture toughness in the transition range. ASTM E08 has developed a draft standard that shows promise for providing lower bound transition range fracture toughness using the master curve approach. This method has been successfully implemented using 1T compact fracture specimens. Combustion Engineering reactor vessel surveillance programs do not have compact fracture specimens. Therefore, the CE Owners Group developed a program to validate the master curve method for Charpy-sized and reconstituted Charpy-sized specimens for future application on irradiated specimens. This method was validated for Linde 1092 welds using unirradiated Charpy-sized and reconstituted Charpy-sized specimens by comparison of results with those from compact fracture specimens.

  9. Final Report - Gas Retention and Release Tests Supporting the Concentrate Receipt Vessel (CRV-VSL-00002A/2B) Configuration

    SciTech Connect (OSTI)

    GUERRERO, HECTOR

    2004-09-01T23:59:59.000Z

    Gas Retention and Release (GR and R) tests were performed in the scaled Concentrate Receipt Vessel (CRV) Test Stand at the Savannah River National Laboratory to validate the capability of candidate Hybrid-Mixing systems for the CRV to safely release hydrogen during normal and upset conditions. Hydrogen is generated in the radioactive waste as a result of natural and plant processes and must not be allowed to accumulate above flammability limits. Two types of tests were conducted. Gas holdup tests determined the steady state amount of gas accumulated in the simulant under normal PJM only or PJM plus sparging conditions. Gas release tests determined what operating conditions are necessary to fully release gas after a steady state gas fraction of 4 per cent tank volume or more was reached in the simulant.

  10. CONTAINMENT VESSEL TEMPERATURE FOR PU-238 HEAT SOURCE CONTAINER UNDER AMBIENT, FREE CONVECTION AND LOW EMISSIVITY COOLING CONDITIONS

    SciTech Connect (OSTI)

    Gupta, N.; Smith, A.

    2011-02-14T23:59:59.000Z

    The EP-61 primary containment vessel of the 5320 shipping package has been used for storage and transportation of Pu-238 plutonium oxide heat source material. For storage, the material in its convenience canister called EP-60 is placed in the EP-61 and sealed by two threaded caps with elastomer O-ring seals. When the package is shipped, the outer cap is seal welded to the body. While stored, the EP-61s are placed in a cooling water bath. In preparation for welding, several containers are removed from storage and staged to the welding booth. The significant heat generation of the contents, and resulting rapid rise in component temperature necessitates special handling practices. The test described here was performed to determine the temperature rise with time and peak temperature attained for an EP-61 with 203 watts of internal heat generation, upon its removal from the cooling water bath.

  11. Heavy-Section Steel Irradiation Program on irradiation effects in light-water reactor pressure vessel materials

    SciTech Connect (OSTI)

    Nanstad, R.K.; Corwin, W.R.; Alexander, D.J.; Haggag, F.M.; Iskander, S.K.; McCabe, D.E.; Sokolov, M.A.; Stoller, R.E. [Oak Ridge National Lab., TN (United States). Metals and Ceramics Div.

    1995-07-01T23:59:59.000Z

    The safety of commercial light-water nuclear plants is highly dependent on the structural integrity of the reactor pressure vessel (RPV). In the absence of radiation damage to the RPV, fracture of the vessel is difficult to postulate. Exposure to high energy neutrons can result in embrittlement of radiation-sensitive RPV materials. The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory, sponsored by the US Nuclear Regulatory Commission (USNRC), is assessing the effects of neutron irradiation on RPV material behavior, especially fracture toughness. The results of these and other studies are used by the USNRC in the evaluation of RPV integrity and regulation of overall nuclear plant safety. In assessing the effects of irradiation, prototypic RPV materials are characterized in the unirradiated condition and exposed to radiation under varying conditions. Mechanical property tests are conducted to provide data which can be used in the development of guidelines for structural integrity evaluations, while metallurgical examinations and mechanistic modeling are performed to improve understanding of the mechanisms responsible for embrittlement. The results of these investigations, in conjunction with results from commercial reactor surveillance programs, are used to develop a methodology for the prediction of radiation effects on RPV materials. This irradiation-induced degradation of the materials can be mitigated by thermal annealing, i.e., heating the RPV to a temperature above that of normal operation. Thus, thermal annealing and evaluation of reirradiation behavior are major tasks of the HSSI Program. This paper describes the HSSI Program activities by summarizing some past and recent results, as well as current and planned studies. 30 refs., 8 figs., 1 tab.

  12. Survey of welding processes for field fabrication of 2 1/4 Cr-1 Mo steel pressure vessels. [128 references

    SciTech Connect (OSTI)

    Grotke, G.E.

    1980-04-01T23:59:59.000Z

    Any evaluation of fabrication methods for massive pressure vessels must consider several welding processes with potential for heavy-section applications. These include submerged-arc and shielded metal-arc, narrow-joint modifications of inert-gas metal-arc and inert-gas tungsten-arc processes, electroslag, and electron beam. The advantage and disadvantages of each are discussed. Electroslag welding can be dropped from consideration for joining of 2 1/4 Cr-1 Mo steel because welds made with this method do not provide the required mechanical properties in the welded and stress relieved condition. The extension of electron-beam welding to sections as thick as 4 or 8 inches (100 or 200 mm) is too recent a development to permit full evaluation. The manual shielded metal-arc and submerged-arc welding processes have both been employed, often together, for field fabrication of large vessels. They have the historical advantage of successful application but present other disadvantages that make them otherwise less attractive. The manual shielded metal-arc process can be used for all-position welding. It is however, a slow and expensive technique for joining heavy sections, requires large amounts of skilled labor that is in critically short supply, and introduces a high incidence of weld repairs. Automatic submerged-arc welding has been employed in many critical applications and for welding in the flat position is free of most of the criticism that can be leveled at the shielded metal-arc process. Specialized techniques have been developed for horizontal and vertical position welding but, used in this manner, the applications are limited and the cost advantage of the process is lost.

  13. Seasonal patterns of coarse sediment transport on a mixed sand and gravel beach due to vessel wakes, wind waves, and tidal currents

    E-Print Network [OSTI]

    Talke, Stefan

    Seasonal patterns of coarse sediment transport on a mixed sand and gravel beach due to vessel wakes, wind waves, and tidal currents Gregory M. Curtiss a, , Philip D. Osborne b,1 , Alexander R. Horner December 2008 Accepted 29 December 2008 Keywords: mixed sand and gravel beach ferry wake wash beach

  14. HSE 1 HSE 2 HSE 3 GE 1 GE 2 GE 3 Residual effects of Large Vessels in GE BOLD Differential Mapping of Ocular Dominance Columns

    E-Print Network [OSTI]

    HSE 1 HSE 2 HSE 3 GE 1 GE 2 GE 3 Residual effects of Large Vessels in GE BOLD Differential Mapping these techniques in humans. Previous human studies (4-6) instead used the conventional GE BOLD technique, combined and limitations of GE BOLD differential mapping as compared to HSE BOLD differential mapping of ocular dominance

  15. Safety Evaluation Report: Development of Improved Composite Pressure Vessels for Hydrogen Storage, Lincoln Composites, Lincoln, NE, May 25, 2010

    SciTech Connect (OSTI)

    Fort, III, William C.; Kallman, Richard A.; Maes, Miguel; Skolnik, Edward G.; Weiner, Steven C.

    2010-12-22T23:59:59.000Z

    Lincoln Composites operates a facility for designing, testing, and manufacturing composite pressure vessels. Lincoln Composites also has a U.S. Department of Energy (DOE)-funded project to develop composite tanks for high-pressure hydrogen storage. The initial stage of this project involves testing the permeation of high-pressure hydrogen through polymer liners. The company recently moved and is constructing a dedicated research/testing laboratory at their new location. In the meantime, permeation tests are being performed in a corner of a large manufacturing facility. The safety review team visited the Lincoln Composites site on May 25, 2010. The project team presented an overview of the company and project and took the safety review team on a tour of the facility. The safety review team saw the entire process of winding a carbon fiber/resin tank on a liner, installing the boss and valves, and curing and painting the tank. The review team also saw the new laboratory that is being built for the DOE project and the temporary arrangement for the hydrogen permeation tests.

  16. Debris dispersal in reactor material experiments on corium-water thermal interactions in ex-vessel geometry

    SciTech Connect (OSTI)

    Sienicki, J.J.; Spencer, B.W.; Squarer, D.

    1984-01-01T23:59:59.000Z

    An analysis has been performed of corium sweepout behavior in the ANL/EPRI CWTI-series reactor material experiments involving the gas pressure-driven injection of molten corium into the reactor cavity region of a 1:30 scale mockup of a PWR containment. A computer model was developed to calculate the sweepout versus retention of corium and water from the cavity. The model consists of hydrodynamics and freezing calculations describing the pressure-driven two-phase flow of corium, water, steam and gas out of the cavity, freezing of corium upon structural surfaces, and levitation of corium within the cavity by the vessel blowdown gas jet. The model has had good success predicting the disposition of corium for the available CWTI tests, indicating retention in the cavity of between 40 and 70% of the injected corium masses. For conditions representative of the TMLB' sequence in the reactor system, the model predicts essentially complete sweepout of corium from the full-scale cavity region before the dispersive forces arising from the blowdown of the primary system have decayed. However, this large sweepout does not imply that the swept out material would deliver its energy directly to the containment atmosphere.

  17. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

    SciTech Connect (OSTI)

    Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

    2012-09-01T23:59:59.000Z

    The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

  18. Progress in evaluation and improvement in nondestructive examination reliability for inservice inspection of Light Water Reactors (LWRs) and characterize fabrication flaws in reactor pressure vessels

    SciTech Connect (OSTI)

    Doctor, S.R.; Bowey, R.E.; Good, M.S.; Friley, J.R.; Kurtz, R.J.; Simonen, F.A.; Taylor, T.T.; Heasler, P.G.; Andersen, E.S.; Diaz, A.A.; Greenwood, M.S.; Hockey, R.L.; Schuster, G.J.; Spanner, J.C.; Vo, T.V.

    1991-10-01T23:59:59.000Z

    This paper is a review of the work conducted under two programs. One (NDE Reliability Program) is a multi-year program addressing the reliability of nondestructive evaluation (NDE) for the inservice inspection (ISI) of light water reactor components. This program examines the reliability of current NDE, the effectiveness of evolving technologies, and provides assessments and recommendations to ensure that the NDE is applied at the right time, in the right place with sufficient effectiveness that defects of importance to structural integrity will be reliably detected and accurately characterized. The second program (Characterizing Fabrication Flaws in Reactor Pressure Vessels) is assembling a data base to quantify the distribution of fabrication flaws that exist in US nuclear reactor pressure vessels with respect to density, size, type, and location. These programs will be discussed as two separate sections in this report. 4 refs., 7 figs.

  19. Analysis of dosimetry from the H.B. Robinson unit 2 pressure vessel benchmark using RAPTOR-M3G and ALPAN

    SciTech Connect (OSTI)

    Fischer, G.A. [Westinghouse Electric Company, LLC, 1000 Westinghouse Dr., Cranberry Township, PA 16066 (United States)

    2011-07-01T23:59:59.000Z

    Document available in abstract form only, full text of document follows: The dosimetry from the H. B. Robinson Unit 2 Pressure Vessel Benchmark is analyzed with a suite of Westinghouse-developed codes and data libraries. The radiation transport from the reactor core to the surveillance capsule and ex-vessel locations is performed by RAPTOR-M3G, a parallel deterministic radiation transport code that calculates high-resolution neutron flux information in three dimensions. The cross-section library used in this analysis is the ALPAN library, an Evaluated Nuclear Data File (ENDF)/B-VII.0-based library designed for reactor dosimetry and fluence analysis applications. Dosimetry is evaluated with the industry-standard SNLRML reactor dosimetry cross-section data library. (authors)

  20. Specific corrosion product on interior surface of a bronze wine vessel with loop-handle and its growth mechanism, Shang Dynasty, China

    SciTech Connect (OSTI)

    Li Yang; Bao Zhirong; Wu Taotao [School of Physics and Technology, Center for Electron Microscopy and MOE Key Laboratory of Artificial Micro- and Nano-structures, Wuhan University, Wuhan 430072 (China); Jiang, Junchun [Xiaogan Museum, Xiaogan 432000 (China); Chen Guantao [Center for Archaeometry, Wuhan University, Wuhan 430072 (China); Pan Chunxu, E-mail: cxpan@whu.edu.cn [School of Physics and Technology, Center for Electron Microscopy and MOE Key Laboratory of Artificial Micro- and Nano-structures, Wuhan University, Wuhan 430072 (China); Center for Archaeometry, Wuhan University, Wuhan 430072 (China)

    2012-06-15T23:59:59.000Z

    In this paper, a kind of specific stalactitic product was found on the interior surface of a covered bronze wine vessel with loop-handle (Chinese name is you), which was fabricated in Shang Dynasty (1700 B.C.-1100 B.C.) and now is collected in Xiaogan Museum, Hubei province of China. The microstructures of the product were characterized systematically by using optical microscopy, scanning electron microscope, transmission electron microscope, X-ray diffraction, and Raman microscopy. The experimental results revealed that the product belonged to a kind of malachite with high purity and high crystallinity. The growth of the product was considered to be a possible reason that the vessel was overly airtight within a museum display cabinet besides a lid of the vessel, which made the excess of H{sub 2}O and CO{sub 2} gas concentrations inside the vessel during long-term storage. This corrosion product is very harmful to bronze cultural relics, because of a large amount of copper consumption from the matrix which will reduce its life. The growth mechanism of the specific stalactitic product and the suggestions for preservation of the similar bronze relics in museum were proposed. - Highlights: Black-Right-Pointing-Pointer The stalactitic product was the high purity and good crystallinity malachite. Black-Right-Pointing-Pointer Its growth was related to the excess of H{sub 2}O and CO{sub 2} gas concentrations in museum. Black-Right-Pointing-Pointer It is harmful to the bronzes, because copper will be consumed from the matrix. Black-Right-Pointing-Pointer The suggestions for preservation of the similar bronzes in museum were proposed.

  1. Optical Measurement Technologies for High Temperature, Radiation Exposure, and Corrosive Environments—Significant Activities and Findings: In-vessel Optical Measurements for Advanced SMRs

    SciTech Connect (OSTI)

    Anheier, Norman C.; Cannon, Bret D.; Qiao, Hong (Amy) [Amy; Suter, Jonathan D.

    2012-09-01T23:59:59.000Z

    Development of advanced Small Modular Reactors (aSMRs) is key to providing the United States with a sustainable, economically viable, and carbon-neutral energy source. The aSMR designs have attractive economic factors that should compensate for the economies of scale that have driven development of large commercial nuclear power plants to date. For example, aSMRs can be manufactured at reduced capital costs in a factory and potentially shorter lead times and then be shipped to a site to provide power away from large grid systems. The integral, self-contained nature of aSMR designs is fundamentally different than conventional reactor designs. Future aSMR deployment will require new instrumentation and control (I&C) architectures to accommodate the integral design and withstand the extreme in-vessel environmental conditions. Operators will depend on sophisticated sensing and machine vision technologies that provide efficient human-machine interface for in-vessel telepresence, telerobotic control, and remote process operations. The future viability of aSMRs is dependent on understanding and overcoming the significant technical challenges involving in-vessel reactor sensing and monitoring under extreme temperatures, pressures, corrosive environments, and radiation fluxes

  2. Functional and operational design requirements for decontamination and decommissioning of the EBR-I Mark-II NaK: Final report. [NaK eutectics

    SciTech Connect (OSTI)

    Brown, B.W.; Crandall, D.L.; Dafoe, R.E.; Dolenc, M.R.; LaRue, D.M.

    1987-09-01T23:59:59.000Z

    Approximately 180 gal of sodium/potassium (NaK) eutectic liquid metal were severely radioactively contaminated during a meltdown of the Mark-II core of the Experimental Breeder Reactor-I (EBR-I) in November 1955. This contaminated NaK, which is contained in four vessels, is currently stored in an underground bunker located at the Army Reentry Vehicle Facility Site (ARVFS) located approximately at the center of the Idaho National Engineering Laboratory (INEL). This document presents the Functional and Operational Requirements (F and ORs) for the D and D of the contaminated NaK and the ARVFS bunker site. This project will chemically deactivate the NaK; dispose of the radioactively contaminated product at a designated burial site; chemically deactivate any residual NaK in the containers, and dispose of the containers at a designated burial site; decontaminate and decommission any contaminated process equipment used in these operations, and decontaminate and decommission the ARVFS bunker site. Completion of the above technical objectives will allow for the effective disposition of the NaK, and will return the ARFVS bunker and immediate area to a reusable condition. Upon completion, the ARVFS NaK, which is now considered a significant potential hazard, will be removed from the Surplus Facilities Management Program priority listing of projects. 33 refs., 8 figs.

  3. Composition and chemistry of particulates from the Tidd Clean Coal Demonstration Plant pressurized fluidized bed combustor, cyclone, and filter vessel

    SciTech Connect (OSTI)

    Smith, D.H.; Grimm, U.; Haddad, G.

    1995-12-31T23:59:59.000Z

    In a Pressurized Fluidized Bed Combustion (PFBC)/cyclone/filter system ground coal and sorbent are injected as pastes into the PFBC bed; the hot gases and entrained fine particles of ash and calcined or reacted sorbent are passed through a cyclone (which removes the larger entrained particles); and the very-fine particles that remain are then filtered out, so that the cleaned hot gas can be sent through a non-ruggedized hot-gas turbine. The 70 MWe Tidd PFBC Demonstration Plant in Brilliant, Ohio was completed in late 1990. The initial design utilized seven strings of primary and secondary cyclones to remove 98% of the particulate matter. However, the Plant also included a pressurized filter vessel, placed between the primary and secondary cyclones of one of the seven strings. Coal and dolomitic limestone (i.e, SO{sub 2} sorbent) of various nominal sizes ranging from 12 to 18 mesh were injected into the combustor operating at about 10 atm pressure and 925{degree}C. The cyclone removed elutriated particles larger than about 0.025 mm, and particles larger than ca. 0.0005 mm were filtered at about 750{degree}C by ceramic candle filters. Thus, the chemical reaction times and temperatures, masses of material, particle-size distributions, and chemical compositions were substantially different for particulates removed from the bed drain, the cyclone drain, and the filter unit. Accordingly, we have measured the particle-size distributions and concentrations of calcium, magnesium, sulfur, silicon, and aluminum for material taken from the three units, and also determined the chemical formulas and predominant crystalline forms of the calcium and magnesium sulfate compounds formed. The latter information is particularly novel for the filter-cake material, from which we isolated the ``new`` compound Mg{sub 2}Ca(SO{sub 4}){sub 3}.

  4. BULLETIN OF THE UNITED STATES FISH COMMISSION. 187 90s-AN ACT T O P R O H I B I T PIRHZS\\'61 BY STEAM VESSELS WIT'R

    E-Print Network [OSTI]

    BY STEAM VESSELS WIT'R WEIRRED O R PURSE CJEINES IN ANY O F THE WAWERS WITHIN THE JURISDICTION O F TRE, That it shall not be lawfuI for any person with steam ves- sels to take with purse or shirred nets any menhaden directed by scction four of this act j and the said steam vessel used and employed in the conmission

  5. 1-Dimensional simulation of thermal annealing in a commercial nuclear power plant reactor pressure vessel wall section

    SciTech Connect (OSTI)

    Nakos, J.T.; Rosinski, S.T.; Acton, R.U.

    1994-11-01T23:59:59.000Z

    The objective of this work was to provide experimental heat transfer boundary condition and reactor pressure vessel (RPV) section thermal response data that can be used to benchmark computer codes that simulate thermal annealing of RPVS. This specific protect was designed to provide the Electric Power Research Institute (EPRI) with experimental data that could be used to support the development of a thermal annealing model. A secondary benefit is to provide additional experimental data (e.g., thermal response of concrete reactor cavity wall) that could be of use in an annealing demonstration project. The setup comprised a heater assembly, a 1.2 in {times} 1.2 m {times} 17.1 cm thick [4 ft {times} 4 ft {times} 6.75 in] section of an RPV (A533B ferritic steel with stainless steel cladding), a mockup of the {open_quotes}mirror{close_quotes} insulation between the RPV and the concrete reactor cavity wall, and a 25.4 cm [10 in] thick concrete wall, 2.1 in {times} 2.1 in [10 ft {times} 10 ft] square. Experiments were performed at temperature heat-up/cooldown rates of 7, 14, and 28{degrees}C/hr [12.5, 25, and 50{degrees}F/hr] as measured on the heated face. A peak temperature of 454{degrees}C [850{degrees}F] was maintained on the heated face until the concrete wall temperature reached equilibrium. Results are most representative of those RPV locations where the heat transfer would be 1-dimensional. Temperature was measured at multiple locations on the heated and unheated faces of the RPV section and the concrete wall. Incident heat flux was measured on the heated face, and absorbed heat flux estimates were generated from temperature measurements and an inverse heat conduction code. Through-wall temperature differences, concrete wall temperature response, heat flux absorbed into the RPV surface and incident on the surface are presented. All of these data are useful to modelers developing codes to simulate RPV annealing.

  6. Comparison of MELCOR modeling techniques and effects of vessel water injection on a low-pressure, short-term, station blackout at the Grand Gulf Nuclear Station

    SciTech Connect (OSTI)

    Carbajo, J.J.

    1995-06-01T23:59:59.000Z

    A fully qualified, best-estimate MELCOR deck has been prepared for the Grand Gulf Nuclear Station and has been run using MELCOR 1.8.3 (1.8 PN) for a low-pressure, short-term, station blackout severe accident. The same severe accident sequence has been run with the same MELCOR version for the same plant using the deck prepared during the NUREG-1150 study. A third run was also completed with the best-estimate deck but without the Lower Plenum Debris Bed (BH) Package to model the lower plenum. The results from the three runs have been compared, and substantial differences have been found. The timing of important events is shorter, and the calculated source terms are in most cases larger for the NUREG-1150 deck results. However, some of the source terms calculated by the NUREG-1150 deck are not conservative when compared to the best-estimate deck results. These results identified some deficiencies in the NUREG-1150 model of the Grand Gulf Nuclear Station. Injection recovery sequences have also been simulated by injecting water into the vessel after core relocation started. This marks the first use of the new BH Package of MELCOR to investigate the effects of water addition to a lower plenum debris bed. The calculated results indicate that vessel failure can be prevented by injecting water at a sufficiently early stage. No pressure spikes in the vessel were predicted during the water injection. The MELCOR code has proven to be a useful tool for severe accident management strategies.

  7. Sponsored by William W. Hay Railroad Engineering Seminar

    E-Print Network [OSTI]

    Illinois at Urbana-Champaign, University of

    Expenditures (In Billions) 4 #12;5 Network Planning's role is critical to company success Service Growth Capital Network Planning... -- Identifies where to add capacity to accommodate future growth -- Ensures Planning makes sure that the network is in the right place, with the right capabilities, to serve the right

  8. Railroad Environmental Conference University of Illinois, Urbana-Champaign

    E-Print Network [OSTI]

    Illinois at Urbana-Champaign, University of

    for Ethanol 2% - 5% Natural Gasoline (Low octane) Contains BTEX which are the regulated compounds Geology Elevated track bed with perched water Below track bed is a 35 foot clay later about 10 feet below grade Affected D.O. in Surface Water Impacted Sediment in Drainage Ditches #12;Typical methods=3 weeks

  9. Performance and alertness of railroad engineers on long duration trips

    E-Print Network [OSTI]

    Karnali, Leona D., 1978-

    2004-01-01T23:59:59.000Z

    The present study investigated the performance and alertness of 8 healthy male participants between the age of 31 years and 65 years old in nighttime, long duration simulated train operation. Participants completed the ...

  10. Remote Sensing of Railroad Locomotive Emissions: A Feasibility Study

    E-Print Network [OSTI]

    Denver, University of

    of General Motors (EMD) or General Electric Transportation Systems (GETS). A diesel- electric locomotive and Donald H. Stedman Department of Chemistry and Biochemistry University of Denver Denver, CO 80208 July. Carbon monoxide (CO) levels become elevated primarily due to direct emission of the gas. Ground

  11. ONE WAY ONE WAY KB&S Railroad

    E-Print Network [OSTI]

    Ginzel, Matthew

    VAWT SHLY WOOD DUHM DMNT FWLR HNLY MRRT VLAB DOYL LCC MATH BRNG REC ME KCTR YONG UPSB BCHM NUCL KNOY

  12. William W. Hay Railroad Engineering Seminar Union Pacific's Next

    E-Print Network [OSTI]

    Barkan, Christopher P.L.

    Cluster VTI System #12;1010 Mid Chord Offset (10') Exception · 10' Mid Chord Offset (MCO) ­ Axle and calculate MCO. ­ 10' chord recommended and detection added to VTI's. ­ MCO initially developed to find the short profile exceptions in concrete tie track 5 ft 5 ft +MCO Y+5 Y=0 y-5 Rail Surface #12;1111 Mid

  13. An analysis of railroad accident dynamics in College Station, Texas

    E-Print Network [OSTI]

    Brown, Hart Stefan

    1998-01-01T23:59:59.000Z

    and requires a management decision. A contingency plan for the release of hazardous materials along the track appears warranted in an attempt to reduce the effects of a hazardous release of flammable or toxic materials. Based on the analysis presented herein...

  14. U.S. Railroad Safety Statistics and Trends

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

    injuries with no days away from work. Includes fatal injuries. Data for Sweden & Norway (1993-96) and for U.S. (all years) is per 200,000 hours actually worked, hence higher....

  15. Train Resistance and Railroad Emissions and Efficiency Mark Stehly

    E-Print Network [OSTI]

    Barkan, Christopher P.L.

    with respect to the wind. R is in lbs force or lbs force per ton of train weight. Train Resistance #12 lbs force per bearing at a typical wheel size. This is 0.6 lbs force per ton of train weight. Removal by 30% or more. Wheel/rail friction on tangent track without lubrication typically is 2 lbs force per

  16. Driver comprehension of railroad-highway grade crossings

    E-Print Network [OSTI]

    Messick, Jennifer

    1995-01-01T23:59:59.000Z

    each year; however, collisions at grade crossings are still a major concern. In order to continue improving safety at grade crossings, driver comprehension and driver attitudes must be investigated. This research investigated driver comprehension...

  17. ILLINOIS -RAILROAD ENGINEERING Understanding the Transition Behavior of

    E-Print Network [OSTI]

    Illinois at Urbana-Champaign, University of

    at Level Crossings on US High Speed Rail Shared Corridors Francesco Bedini Jacobini PhD Student Rail

  18. american railroads mechanical: Topics by E-print Network

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

    company's series. It seems more... Levine, Stuart 1976-01-01 4 Federal Deposit Insurance Corporation Center for Financial ResearchAsset Salability and Debt Maturity: Evidence...

  19. APPLICATION OF CAST AND STPA TO RAILROAD SAFETY IN CHINA

    E-Print Network [OSTI]

    Leveson, Nancy

    The accident analysis method called STAMP (System-Theoretic Accident Model), developed by Prof. Nancy Leveson

  20. Application of CAST and STPA to railroad safety in China

    E-Print Network [OSTI]

    Dong, Airong, S.M. Massachusetts Institute of Technology

    2012-01-01T23:59:59.000Z

    The accident analysis method called STAMP (System-Theoretic Accident Model), developed by Prof. Nancy Leveson from MIT, was used here to re-analyze a High Speed Train accident in China. On July 23rd, 2011, 40 people were ...

  1. Permeability, Resistivity and Strength of Fouled Railroad Ballast

    E-Print Network [OSTI]

    Rahman, A. J.

    2013-08-31T23:59:59.000Z

    particles between the ballast particles; therefore, permeability and resistivity were also correlated. The strength properties of clean and fouled ballast were also evaluated using large direct shear box and modified direct shear box (extension in height...

  2. Optimal Location of Railroad Wayside Defect Detection Installations

    E-Print Network [OSTI]

    Barkan, Christopher P.L.

    reduce inspection costs (e.g., labor) Examples of new technologies Wheel impact load detector (WILD systems #12;6 6 Wheel Impact Load Detectors (WILD) 1.5 x 5 in. Shell, 106 k1.5 x 5 in. Shell, 106 k. #12;10 10 Newer Technologies Image-Based (e.g. FactISTM) Wheel Profile Missing, Worn or Damaged

  3. Effect of Railroad Deregulation on Export - Grain Transportation Rate Structures.

    E-Print Network [OSTI]

    Fuller, Stephen; MacDonald, James; Bessler, David; Wohlgenant, Michael

    1988-01-01T23:59:59.000Z

    costs. A stratified ran dom sample of rail waybills drawn from the ICC Waybill Statistics show unit train use spread rapidly on the Plains region export routes in the early 1980s, especially be tween 1982 and 1984. In particular, in the 1981...

  4. Enhancing Railroad Hazardous Materials Transportation Safety Rail Routing |

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic Plan Departmentof EnergyPublic LawEnergyEnhanced Reduce

  5. Railroad Commission of Texas, Oil and Gas Division | Open Energy

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directedAnnual Siteof Energy 2,AUDITCalifornia Sector: Wind energy Product:Anatolia JumpRSE

  6. Alaska Administrative Code - Title 17, Chapter 15 - Utility and Railroad

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are beingZealand Jump to:Ezfeedflag JumpID-fTriWildcat 1AMEEAisin Seiki G60Alameda County,90 -Definitions

  7. Name: Lisa Anderson Organization: Tri-City Railroad Company

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What's Possible for Renewable Energy: GridTruck PlatooningJefferson7593 LECTIJHES ONThe Role9085023 Name:

  8. Texas Railroad Commission - Pollution Discharge Regulations | Open Energy

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov YouKizildere IRaghuraji Agro Industries PvtStratosolar JumpTennessee/WindPetroleum Storage Tanks Webpage Jump

  9. Historic Railroad Building Goes Net Zero | Department of Energy

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn't YourTransport(Fact Sheet), GeothermalGridHYDROGEND D e e p p a aHilton

  10. Suggestion of typical phases of in-vessel fuel-debris by thermodynamic calculation for decommissioning technology of Fukushima-Daiichi nuclear power station

    SciTech Connect (OSTI)

    Ikeuchi, Hirotomo; Yano, Kimihiko; Kaji, Naoya; Washiya, Tadahiro [Japan Atomic Energy Agency, 4-33 Muramatsu, Tokai-mura, Ibaraki-ken, 319-1194 (Japan); Kondo, Yoshikazu; Noguchi, Yoshikazu [PESCO Co.Ltd. (Korea, Republic of)

    2013-07-01T23:59:59.000Z

    For the decommissioning of the Fukushima-Daiichi Nuclear Power Station (1F), the characterization of fuel-debris in cores of Units 1-3 is necessary. In this study, typical phases of the in-vessel fuel-debris were estimated using a thermodynamic equilibrium (TDE) calculation. The FactSage program and NUCLEA database were applied to estimate the phase equilibria of debris. It was confirmed that the TDE calculation using the database can reproduce the phase separation behavior of debris observed in the Three Mile Island accident. In the TDE calculation of 1F, the oxygen potential [G(O{sub 2})] was assumed to be a variable. At low G(O{sub 2}) where metallic zirconium remains, (U,Zr)O{sub 2}, UO{sub 2}, and ZrO{sub 2} were found as oxides, and oxygen-dispersed Zr, Fe{sub 2}(Zr,U), and Fe{sub 3}UZr{sub 2} were found as metals. With an increase in zirconium oxidation, the mass of those metals, especially Fe{sub 3}UZr{sub 2}, decreased, but the other phases of metals hardly changed qualitatively. Consequently, (U,Zr)O{sub 2} is suggested as a typical phase of oxide, and Fe{sub 2}(Zr,U) is suggested as that of metal. However, a more detailed estimation is necessary to consider the distribution of Fe in the reactor pressure vessel through core-melt progression. (authors)

  11. Application of ex-vessel neutron dosimetry combined with in-core measurements for correction of neutron source used for RPV fluence calculations

    SciTech Connect (OSTI)

    Borodkin, P.G.; Borodkin, G.I.; Khrennikov, N.N. [Scientific and Engineering Centre for Nuclear and Radiation Safety SEC NRS, Malaya Krasnoselskaya ul., 2/8, Bld. 5, 107140 Moscow (Russian Federation); Konheiser, J. [Helmholz Zentrum Dresden-Rossendorf HZDR, Postfach 510119, D-01314 Dresden (Germany)

    2011-07-01T23:59:59.000Z

    This paper deals with calculated and semi-analytical evaluations of VVER-1000 reactor core neutron source distributions and their influence on measurements and calculations of the integral through-vessel neutron leakage. Neutron activation measurements analyzed in the paper were carried out in an ex-vessel air cavity at different nuclear power plant units with VVER-1000 during different fuel cycles. The time-integrated neutron source distributions used for DORT calculations were prepared via two different approaches based on (a) calculated fuel burnup (standard routine procedure) and (b) in-core measurements by means of self-powered detectors (SPDs) and thermocouples (TCs) (new approach). Considering that fuel burnup distributions in operating VVER may be evaluated now by the use of analytical methods (calculations) only, it is necessary to develop new approaches for the testing and correction of calculated evaluations of a neutron source. The results presented in this paper allow one to consider the reverse task of the alternative estimation of fuel burnup distributions. The proposed approach is based on the adjustment (fitting) of time-integrated neutron source distributions, and thus fuel burnup patterns, in some part of the reactor core, taking into account neutron leakage measurements, neutron-physical calculations, and in-core SPD and TC measurement data. (authors)

  12. Fracture Analysis of Vessels – Oak Ridge FAVOR, v06.1, Computer Code: Theory and Implementation of Algorithms, Methods, and Correlations

    SciTech Connect (OSTI)

    Williams, P. T. [ORNL; Dickson, T. L. [ORNL; Yin, S. [ORNL

    2007-12-01T23:59:59.000Z

    The current regulations to insure that nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to transients such as pressurized thermal shock (PTS) events were derived from computational models developed in the early-to-mid 1980s. Since that time, advancements and refinements in relevant technologies that impact RPV integrity assessment have led to an effort by the NRC to re-evaluate its PTS regulations. Updated computational methodologies have been developed through interactions between experts in the relevant disciplines of thermal hydraulics, probabilistic risk assessment, materials embrittlement, fracture mechanics, and inspection (flaw characterization). Contributors to the development of these methodologies include the NRC staff, their contractors, and representatives from the nuclear industry. These updated methodologies have been integrated into the Fracture Analysis of Vessels – Oak Ridge (FAVOR, v06.1) computer code developed for the NRC by the Heavy Section Steel Technology (HSST) program at Oak Ridge National Laboratory (ORNL). The FAVOR, v04.1, code represents the baseline NRC-selected applications tool for re-assessing the current PTS regulations. This report is intended to document the technical bases for the assumptions, algorithms, methods, and correlations employed in the development of the FAVOR, v06.1, code.

  13. Development of the front end test stand and vessel for extraction and source plasma analyses negative hydrogen ion sources at the Rutherford Appleton Laboratory

    SciTech Connect (OSTI)

    Lawrie, S. R., E-mail: scott.lawrie@stfc.ac.uk [STFC ISIS Pulsed Spallation Neutron and Muon Facility, Rutherford Appleton Laboratory, Harwell Oxford, Harwell (United Kingdom); John Adams Institute of Accelerator Science, University of Oxford, Oxford (United Kingdom); Faircloth, D. C.; Letchford, A. P.; Perkins, M.; Whitehead, M. O.; Wood, T. [STFC ISIS Pulsed Spallation Neutron and Muon Facility, Rutherford Appleton Laboratory, Harwell Oxford, Harwell (United Kingdom)] [STFC ISIS Pulsed Spallation Neutron and Muon Facility, Rutherford Appleton Laboratory, Harwell Oxford, Harwell (United Kingdom); Gabor, C. [ASTeC Intense Beams Group, Rutherford Appleton Laboratory, Harwell Oxford, Harwell (United Kingdom)] [ASTeC Intense Beams Group, Rutherford Appleton Laboratory, Harwell Oxford, Harwell (United Kingdom); Back, J. [High Energy Physics Department, University of Warwick, Coventry (United Kingdom)] [High Energy Physics Department, University of Warwick, Coventry (United Kingdom)

    2014-02-15T23:59:59.000Z

    The ISIS pulsed spallation neutron and muon facility at the Rutherford Appleton Laboratory (RAL) in the UK uses a Penning surface plasma negative hydrogen ion source. Upgrade options for the ISIS accelerator system demand a higher current, lower emittance beam with longer pulse lengths from the injector. The Front End Test Stand is being constructed at RAL to meet the upgrade requirements using a modified ISIS ion source. A new 10% duty cycle 25 kV pulsed extraction power supply has been commissioned and the first meter of 3 MeV radio frequency quadrupole has been delivered. Simultaneously, a Vessel for Extraction and Source Plasma Analyses is under construction in a new laboratory at RAL. The detailed measurements of the plasma and extracted beam characteristics will allow a radical overhaul of the transport optics, potentially yielding a simpler source configuration with greater output and lifetime.

  14. In-vessel thermohydraulics evaluation of an unprotected transient overpower accident and delayed neutron precursor concentration transport analysis using a multidimensional code

    SciTech Connect (OSTI)

    Muramatsu, T.; Ninokata, H. (Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan))

    1992-02-01T23:59:59.000Z

    This paper reports on a three-dimensional in-vessel thermohydraulics analysis that is carried out for the early phase of an unprotected transient overpower (UTOP) accident and delayed neutron precursor concentration transport in a typical loop-type fast breeder reactor plant. In the UTOP calculations, the time at which the sodium temperature reaches the reactor trip level is evaluated based on calculated upper plenum flow and temperature distributions. For fission product release from the core assemblies, the delayed neutron precursor concentration in the sodium that reaches the detectors depends on the location of the faulted assembly. Three-dimensional flow patterns, and hence, the residence time in the upper plenum. Delayed neutron precursors that bypassed the recirculation flow to appear in the plenum primarily contribute to the peak concentration.

  15. Development of a methodology for the assessment of shallow-flaw fracture in nuclear reactor pressure vessels: Generation of biaxial shallow-flaw fracture toughness data

    SciTech Connect (OSTI)

    McAfee, W.J.; Bass, B.R.; Bryson, J.W.

    1998-07-01T23:59:59.000Z

    A technology to determine shallow-flaw fracture toughness of reactor pressure vessel (RPV) steels is being developed for application to the safety assessment of RPVs containing postulated shallow-surface flaws. Shallow-flaw fracture toughness of RPV material has been shown to be higher than that for deep flaws, because of the relaxation of crack-tip constraint. This report describes the preliminary test results for a series of cruciform specimens with a uniform depth surface flaw. These specimens are all of the same size with the same depth flaw. Temperature and biaxial load ratio are the independent variables. These tests demonstrated that biaxial loading could have a pronounced effect on shallow-flaw fracture toughness in the lower transition temperature region for RPV materials. Through that temperature range, the effect of full biaxial (1:1) loading on uniaxial, shallow-flaw toughness varied from no effect near the lower shelf to a reduction of approximately 58% at higher temperatures.

  16. Applications of ENDF/B-VI and JENDL-3.1 iron data to reactor pressure vessel fluence analysis using continuous energy Monte Carlo code MCNP

    SciTech Connect (OSTI)

    Kim, Jungo-Do; Gil, Choong-Sup [Korea Atomic Energy Research Institute, Taejon (Korea, Democratic People`s Republic of)

    1994-12-31T23:59:59.000Z

    A comparison is made of results obtained from neutron transmissions analysis of RPV performed by MCNP with ENDF/B-VI and JENDL-3.1 iron data. At first, a one-dimensional discrete ordinates transport calculation using VITAMIN-C fine-group library based on ENDF/B-IV was performed for a cylindrical model of a PWR to generate the source spectrum at the front of the RPV. And then, the transmission of neutrons through RPV was calculated by MCNP with the moderated fission spectrum incident on the vessel face. For these ENDF/B-IV, -VI and JENDL-3.1 iron data were processed into continuous energy point data form by NJOY91.91. The fast neutron fluxes and dosimeter reaction rates through RPV using each iron data were intercompared.

  17. Supplement Analysis to the 1999 Site-Wide Environmental Impact Statement for Continued Operation of Los Alamos National Laboratory for the Proposed Disposition of Certain Large Containment Vessels

    SciTech Connect (OSTI)

    N /A

    2004-02-12T23:59:59.000Z

    This Supplement Analysis (SA) has been prepared to determine if the Site-Wide Environmental Impact Statement for Continued Operations of Los Alamos National Laboratory (SWEIS) (DOE/EIS-0238) (DOE 1999a) adequately addresses the environmental effects of introducing a proposed project for the clean-out and decontamination (DECON) of certain large containment vessels into the Chemistry and Metallurgy Research (CMR) Building located at Los Alamos National Laboratory (LANL) Technical Area (TA) 3, or if the SWEIS needs to be supplemented. After undergoing the clean-out and DECON steps, the subject containment vessels would be disposed of at LANL's TA-54 low-level waste (LLW) disposal site or, as appropriate, at a DOE or commercial offsite permitted LLW-regulated landfill; after actinides were recovered from the DECON solution within the CMR Building, they would be moved to LANL's TA-55 Plutonium Facility and undergo subsequent processing at that facility for reuse. Council on Environmental Quality regulations at Title 40, Section 1502.9(c) of the Code of Federal Regulations (40 CFR 1502.9[c]) require federal agencies to prepare a supplement to an environmental impact statement (EIS) when an agency makes substantial changes in the proposed action that are relevant to environmental concerns, or there are changed circumstances or new or changed information relevant to concerns and bearing on the proposed action or its impacts. This SA is prepared in accordance with Section 10 CFR 10211.314(c) of the DOE's regulations for National Environmental Policy Act (NEPA) implementation that states: ''When it is unclear whether or not an EIS supplement is required, DOE shall prepare a Supplement Analysis''. This SA specifically compares key impact assessment parameters of the proposed project action with the LANL operations capabilities evaluated in the 1999 SWEIS in support DOE's long-term hydrodynamic testing program at LANL, as well as the waste disposal capabilities evaluated in the SWEIS in support of LANL operations. It also provides an explanation of any differences between the proposed action and activities described in the SWEIS analysis. The SWEIS analyzed the impacts of performing plutonium (Pu) and actinide activities, including hydrodynamic testing support activity, at the Plutonium Facility and at the CMR Building.

  18. Application of the base catalyzed decomposition process to treatment of PCB-contaminated insulation and other materials associated with US Navy vessels. Final report

    SciTech Connect (OSTI)

    Schmidt, A.J.; Zacher, A.H.; Gano, S.R.

    1996-09-01T23:59:59.000Z

    The BCD process was applied to dechlorination of two types of PCB-contaminated materials generated from Navy vessel decommissioning activities at Puget Sound Naval Shipyard: insulation of wool felt impregnated with PCB, and PCB-containing paint chips/debris from removal of paint from metal surfaces. The BCD process is a two-stage, low-temperature chemical dehalogenation process. In Stage 1, the materials are mixed with sodium bicarbonate and heated to 350 C. The volatilized halogenated contaminants (eg, PCBs, dioxins, furans), which are collected in a small volume of particulates and granular activated carbon, are decomposed by the liquid-phase reaction (Stage 2) in a stirred-tank reactor, using a high-boiling-point hydrocarbon oil as the reaction medium, with addition of a hydrogen donor, a base (NaOH), and a catalyst. The tests showed that treating wool felt insulation and paint chip wastes with Stage 2 on a large scale is feasible, but compared with current disposal costs for PCB-contaminated materials, using Stage 2 would not be economical at this time. For paint chips generated from shot/sand blasting, the solid-phase BCD process (Stage 1) should be considered, if paint removal activities are accelerated in the future.

  19. The reconstruction of the Lake Champlain sidewheel steamer Champlain II

    E-Print Network [OSTI]

    Baldwin, Elizabeth Robinson

    1997-01-01T23:59:59.000Z

    The steamship Champlain II, ex-Oakes Ames, was built as a railroad car transfer ferry in 1868 at Marks Bay, Burlington, Vermont in the private shipyard of Napoleon B. Proctor. The vessel was later converted to a passenger line boat in 1873...

  20. Blood Vessels of the Fetal Pig Dissection Anterior Vessels Protocol

    E-Print Network [OSTI]

    Loughry, Jim

    the same with a second piece of string for the back legs. a. Begin by making two incisions through the skin the skin back away from the rib cage. This will allow you to see more clearly where the sternum is and what the heart out into the body. 5. Identify these veins that drain the anterior portion of the fetal pig: a