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Note: This page contains sample records for the topic "railroad vessel bunkering" from the National Library of EnergyBeta (NLEBeta).
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1

The bunkering industry and its effect on shipping tanker operations  

E-Print Network [OSTI]

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 ...

Boutsikas, Angelos

2004-01-01T23:59:59.000Z

2

Bunker Hill Sediment Characterization Study  

SciTech Connect (OSTI)

The long history of mineral extraction in the Coeur d扐lene Basin has left a legacy of heavy metal laden mine tailings that have accumulated along the Coeur d扐lene River and its tributaries (U.S. Environmental Protection Agency, 2001; Barton, 2002). Silver, lead and zinc were the primary metals of economic interest in the area, but the ores contained other elements that have become environmental hazards including zinc, cadmium, lead, arsenic, nickel, and copper. The metals have contaminated the water and sediments of Lake Coeur d扐lene, and continue to be transported downstream to Spokane Washington via the Spokane River. In 1983, the EPA listed the Bunker Hill Mining and Metallurgical Complex on the National Priorities List. Since that time, many of the most contaminated areas have been stabilized or isolated, however metal contaminants continue to migrate through the basin. Designation as a Superfund site causes significant problems for the economically depressed communities in the area. Identification of primary sources of contamination can help set priorities for cleanup and cleanup options, which can include source removal, water treatment or no action depending on knowledge about the mobility of contaminants relative to water flow. The mobility of contaminant mobility under natural or engineered conditions depends on multiple factors including the physical and chemical state (or speciation) of metals and the range of processes, some of which can be seasonal, that cause mobilization of metals. As a result, it is particularly important to understand metal speciation (National Research Council, 2005) and the link between speciation and the rates of metal migration and the impact of natural or engineered variations in flow, biological activity or water chemistry.

Neal A. Yancey; Debby F. Bruhn

2009-12-01T23:59:59.000Z

3

Federal Railroad Administration  

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

Manager Federal Railroad Administration Federal Railroad Administration Dedicated Train Study- Report to Congress FRA' s Research & Development Office (as lead on the study)...

4

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 on Google Bookmark EERE: Alternative Fuels Data CenterFranconia,(Million Barrels) Crude Oil Reserves in NonproducingAdditions to Capacity on CokersA2. ForJanuary 20139,2505,257,810

5

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines AboutDecemberSteam Coal Import96 4.87CBECS Public Use Data0 0 0270,9755,568,0661,983,422

6

Federal Railroad Administration  

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

Transportation Federal Railroad Administration Overview of Proposed Rail Safety & Security Rulemakings Kevin R. Blackwell FRA Hazmat Division Washington, DC Federal Authority DOT...

7

Federal Railroad Administration  

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

- HQ Hazmat Division, Washington, DC. Federal Railroad Administration Dedicated Train Study - Report to Congress November 2003 - FRA' s Ofc. Of Research & Development...

8

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...

9

Intermodal Railroad Daniel Brock  

E-Print Network [OSTI]

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

Nagurney, Anna

10

Plenary II -- Evaluation of Shortline Railroads  

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

Evaluation of Shortline Railroads Tasked for the Transportation of Spent Nuclear Fuel Tasked for the Transportation of Spent Nuclear Fuel Evaluation of Shortline Railroads...

11

Evaluating flexibility in railroad construction projects  

E-Print Network [OSTI]

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 ...

Oh, Choong Ryun, 1972-

2005-01-01T23:59:59.000Z

12

Railroad and locomotive technology roadmap.  

SciTech Connect (OSTI)

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.

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

2003-02-24T23:59:59.000Z

13

Locomotive Fuelling Problem (LFP) in Railroad Operations Bodhibrata Nag1  

E-Print Network [OSTI]

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

Murty, Katta G.

14

Karen Hedlund, Chief Counsel Federal Railroad Administration  

E-Print Network [OSTI]

Karen Hedlund, Chief Counsel Federal Railroad Administration Presentation to the Hagestad Sandhouse-educate, and out-build the rest of the world... That's how we'll win the future." - President Barack Obama, January China are building . . . thousands of miles of high-speed rail." - President Barack Obama, April 19

Bustamante, Fabi谩n E.

15

Monitoring of Olympic National Park Beaches to determine fate and effects of spilled bunker C fuel oil  

SciTech Connect (OSTI)

On December 23, 1988, the barge Nestucca was accidentally struck by its tow, a Souse Brothers Towing Company tug, releasing approximately 230,000 gallons of Bunker C fuel oil and fouling beaches from Grays Harbor north to Vancouver Island. Affected beaches in Washington included a 40-mile-long strip that has been recently added to Olympic National Park. The purpose of the monitoring program documented in this report was to determine the fate of spilled Bunker C fuel oil on selected Washington coastal beaches. We sought to determine (1) how much oil remained in intertidal and shallow subtidal habitats following clean-up and weathering, (2) to what extent intertidal and/or shallow subtidal biotic assemblages have been contaminated, and (3) how rapidly the oil has left the ecosystem. 45 refs., 18 figs., 8 tabs.

Strand, J.A.; Cullinan, V.I.; Crecelius, E.A.; Fortman, T.J.; Citterman, R.J.; Fleischmann, M.L.

1990-10-01T23:59:59.000Z

16

ILLINOIS RAILROAD ENGINEERING Using the RTC Simulation Model to  

E-Print Network [OSTI]

Slide 1 ILLINOIS RAILROAD ENGINEERING Using the RTC Simulation Model to Evaluate Effects) 颅 Simulation models include detailed infrastructure configuration and mimics train dispatcher logic 路 Closest

Barkan, Christopher P.L.

17

Diesel Locomotive Fueling Problem (LFP) in Railroad Operations  

E-Print Network [OSTI]

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

Murty, Katta G.

18

Train derailment and railroad barrier interaction simulation  

E-Print Network [OSTI]

; 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...

Zhang, Heping

1990-01-01T23:59:59.000Z

19

Driver comprehension of railroad-highway grade crossings  

E-Print Network [OSTI]

Over the past twenty years, over $2 billion has been allocated for the improvement of railroad-highway grade crossings. Many passive crossings have been upgraded with active protection which has decreased the number of collisions at grade crossings...

Messick, Jennifer

1995-01-01T23:59:59.000Z

20

Predicting violations at gated active highway-railroad grade crossings  

E-Print Network [OSTI]

-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....

Bean, Jonathan Alan

1997-01-01T23:59:59.000Z

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


21

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

E-Print Network [OSTI]

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 ...

Alpert, Steven M

2007-01-01T23:59:59.000Z

22

GIS for Nevada railroads: 1993 report  

SciTech Connect (OSTI)

This is an interim report on a task within a large, ongoing study by the University of Nevada, Reno to examine the safety of Nevada railroads. The overall goal, of which this year`s research is a middle stage, is to develop models based on the use of geographic information systems (GIS). These models are to enable the selection of the best and safest railway routes for the transport of high-level nuclear waste across Nevada to the proposed repository at Yucca Mountain. Last year`s research concluded that the databases are adequate and that GIS are feasible and desirable for displaying the multi-layered data required to reach decisions about safety. It developed several database layers. This report deals with work during 1993 on the use of geographic information systems (GIS) for rail-route selection. The goal was to identify and assemble many of the databases necessary for the models. In particular, the research aimed to identify (a) any problems with developing database layers; and (b) the level of effort required. This year`s effort developed database layers for two Nevada counties: Clark and Lincoln. The layers dealt with: topographic information, geologic information, and land ownership. These are among the most important database layers. The database layers were successfully created. No significant problems arose in developing them. The level of effort did not exceed the expected level. The most effective approach is by means of digital, shaded relief maps. (Sample maps appear in plates.) Therefore, future database development will be straightforward. Research may proceed on the full development of shaded relief elevation maps for Elko, White Pine, Nye and Eureka counties and with actual modeling for the selection of a route or routes between the UP/SP line in northern Nevada and Yucca Mountain.

Carr, J.R.

1993-12-01T23:59:59.000Z

23

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

SciTech Connect (OSTI)

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

NONE

1991-09-01T23:59:59.000Z

24

Energy Conservation Projects to Benefit the Railroad Industry  

SciTech Connect (OSTI)

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.

Clifford Mirman; Promod Vohra

2009-12-31T23:59:59.000Z

25

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

SciTech Connect (OSTI)

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.

Not Available

1991-08-30T23:59:59.000Z

26

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

27

Railroad Environmental Conference University of Illinois, Urbana-Champaign  

E-Print Network [OSTI]

Use of Ground Penetrating Radar (GPR) to Locate Contaminant Beneath Railroad Track Jim Hyslip 颅 Hy 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

Illinois at Urbana-Champaign, University of

28

Fatigue tests of under-strenghth timber railroad bridge stringers  

E-Print Network [OSTI]

, 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...

Borchers, Christopher Ryan

2002-01-01T23:59:59.000Z

29

Optimal Fueling Strategies for Locomotive Fleets in Railroad Networks  

E-Print Network [OSTI]

Price 3 路 Railroad fuel consumption remains steady 路 Crude oil price sharply increases in recent years 路 Fuel (diesel) price influenced by: 颅 Crude oil price 颅 Refining 颅 Distribution and marketing 颅 Others 4 locomotive j bj=Tank capacity rj=Fuel consumption rate nj=Number of stops fj=Travel frequency gj=Initial fuel

Barkan, Christopher P.L.

30

Neutrino Factory Target Vessel  

E-Print Network [OSTI]

by UT-Battelle for the U.S. Department of Energy Target Vessel Update 26 June 2012 Cooling Channel in both walls for draining 路 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

McDonald, Kirk

31

Vacuum Vessel Remote Handling  

E-Print Network [OSTI]

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

32

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

Open Energy Info (EERE)

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

33

Reactor vessel support system  

DOE Patents [OSTI]

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.

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

1982-01-01T23:59:59.000Z

34

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

E-Print Network [OSTI]

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

Barthelat, Francois

35

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

E-Print Network [OSTI]

[15] 路 Pipelines projects take several years (e.g. ~5 years for the original Keystone) to completeRAILROAD STRATEGY FOR CRUDE OIL TRANSPORT: Considering Public Policy and Pipeline Competition S transported by pipelines in North America, railroads have seen significant growth in this commodity, from just

Entekhabi, Dara

36

RELATIONSHIP BUILDING WITH FREIGHT RAILROADS CRITICAL TO SUPPORT INTERCITY PASSENGER RAIL DEVELOPMENT  

E-Print Network [OSTI]

.4 million (2011) 93% of Amtrak's approximately 21,000 route system utilizes freight railroad owned Relationship between freight railroads and governments remains highly combative HOW CAN GOVERNMENT ENHANCE animosity; address issues through established relationships and dialog over regulation Generate Utility

Illinois at Chicago, University of

37

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 Data Center Home Page on Delicious Rank EERE:YearRound-UpHeatMulti-Dimensional Subject:Ground HawaiiWaste HeatDepartment of EnergyRailroad

38

Reactor vessel annealing system  

DOE Patents [OSTI]

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).

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

39

Sapphire tube pressure vessel  

DOE Patents [OSTI]

A pressure vessel is provided for observing corrosive fluids at high temperatures and pressures. A transparent Teflon bag contains the corrosive fluid and provides an inert barrier. The Teflon bag is placed within a sapphire tube, which forms a pressure boundary. The tube is received within a pipe including a viewing window. The combination of the Teflon bag, sapphire tube and pipe provides a strong and inert pressure vessel. In an alternative embodiment, tie rods connect together compression fittings at opposite ends of the sapphire tube.

Outwater, John O. (Cambridge, MA)

2000-01-01T23:59:59.000Z

40

Recommendations to the NRC for review criteria for alternative methods of low-level radioactive waste disposal: Task 2b: Earth-mounded concrete bunkers  

SciTech Connect (OSTI)

The US Army Engineers Waterways Experiment Station (WES) and US Army Engineer Division, Huntsville (HNDED) have developed general design criteria and specific design review criteria for the earth-mounded concrete bunker (EMCB) alternative method of low-level radioactive waste (LLW) disposal. An EMCB is generally described as a reinforced concrete vault placed below grade, underneath a tumulus, surrounded by filter-blanket and drainage zones. The tumulus is covered over with a low permeability cover layer and top soil with vegetation. Eight major review criteria categories have been developed ranging from the loads imposed on the EMCB structure through material quality and durability considerations. Specific design review criteria have been developed in detail for each of the eight major categories. 63 refs., 13 figs., 2 tabs.

Denson, R.H.; Bennett, R.D.; Wamsley, R.M.; Bean, D.L.; Ainsworth, D.L.

1988-01-01T23:59:59.000Z

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


41

Neutrino Factory Mercury Vessel  

E-Print Network [OSTI]

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 路 Separates functionality, provides double mercury containment, simplifies design and remote handling 路 Each

McDonald, Kirk

42

HEART AND BLOOD VESSELS CARDIOVASCULARCARDIOVASCULAR  

E-Print Network [OSTI]

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

Cochran-Stafira, D. Liane

43

Fatigue behavior of full-size soild-sawn timber railroad stringers  

E-Print Network [OSTI]

As part of the Timber Bridge Life Extension Program sponsored by the Association of American Railroads (AAR), the objective of this thesis is to investigate the fatigue behavior of large solid-sawn timber beams. Specifically, the main flexural...

Maingot, Martin Rex

1999-01-01T23:59:59.000Z

44

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

E-Print Network [OSTI]

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...

Tydlacka, Jonathan Michael

2013-02-22T23:59:59.000Z

45

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

SciTech Connect (OSTI)

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.

Oar, D.L.

1994-09-29T23:59:59.000Z

46

High pressure storage vessel  

DOE Patents [OSTI]

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.

Liu, Qiang

2013-08-27T23:59:59.000Z

47

Reactor pressure vessel nozzle  

DOE Patents [OSTI]

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.

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

1994-10-04T23:59:59.000Z

48

Vessel structural support system  

DOE Patents [OSTI]

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.

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

49

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

50

Coal gasification vessel  

DOE Patents [OSTI]

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).

Loo, Billy W. (Oakland, CA)

1982-01-01T23:59:59.000Z

51

The construction of the Browns Bay Vessel  

E-Print Network [OSTI]

INVESTIGATIVE TECHNIQUES. 10 19 The Site. National Historic Sites Service Excavation and Raising of the Vessel Vessel on Display. The Vessel in 1985. 19 20 27 28 Method of Recording III THE CONSTRUCTION OF THE VESSEL 31 36 The Keel 36 The Stem... A flat-bottomed boat being built. 17 9 Forelocked eye-bolts from the midship beam of the Browne Bay Vessel 21 10 Broad arrow stamped in an eye-bolt from the Browns Bay Vessel. . . . . . . . . . . . . . . . . . . . . . . . . . . 22 11 Pulley...

Amer, Christopher Francis

2012-06-07T23:59:59.000Z

52

Pressure Vessel Burst Program: Automated hazard analysis for pressure vessels  

SciTech Connect (OSTI)

The design, development, and use of a Windows based software tool, PVHAZARD, for pressure vessel hazard analysis is presented. The program draws on previous efforts in pressure vessel research and results of a Pressure Vessel Burst Test Study. Prior papers on the Pressure Vessel Burst Test Study have been presented to the ASME, AIAA, JANNAF, NASA Pressure Systems Seminar, and to a DOD Explosives Safety Board subcommittee meeting. Development and validation is described for simplified blast (overpressure/impulse) and fragment (velocity and travel distance) hazard models. The use of PVHAZARD in making structural damage and personnel injury estimates is discussed. Efforts in-progress are reviewed including the addition of two-dimensional and three-dimensional (2D and 3D) hydrodynamic code analyses to supplement the simplified models, and the ability to assess barrier designs for protection from fragmentation.

Langley, D.R. [Aerospace Corp., Kennedy Space Center, FL (United States); Chrostowski, J.D. [ACTA Inc., Torrance, CA (United States); Goldstein, S. [Aerospace Corp., El Segundo, CA (United States); Cain, M. [General Physics Corp., Titusville, FL (United States)

1996-12-31T23:59:59.000Z

53

Reactor vessel support system. [LMFBR  

DOE Patents [OSTI]

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.

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

1980-05-09T23:59:59.000Z

54

CHF Enhancement by Vessel Coating for External Reactor Vessel Cooling  

SciTech Connect (OSTI)

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.

Fan-Bill Cheung; Joy L. Rempe

2004-06-01T23:59:59.000Z

55

Corrective Action Investigation Plan for Corrective Action Unit 214: Bunkers and Storage Areas Nevada Test Site, Nevada: Revision 0, Including Record of Technical Change No. 1 and No. 2  

SciTech Connect (OSTI)

This Corrective Action Investigation Plan contains the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office's approach to collect the data necessary to evaluate corrective action alternatives appropriate for the closure of Corrective Action Unit (CAU) 214 under the Federal Facility Agreement and Consent Order. Located in Areas 5, 11, and 25 of the Nevada Test Site, CAU 214 consists of nine Corrective Action Sites (CASs): 05-99-01, Fallout Shelters; 11-22-03, Drum; 25-99-12, Fly Ash Storage; 25-23-01, Contaminated Materials; 25-23-19, Radioactive Material Storage; 25-99-18, Storage Area; 25-34-03, Motor Dr/Gr Assembly (Bunker); 25-34-04, Motor Dr/Gr Assembly (Bunker); and 25-34-05, Motor Dr/Gr Assembly (Bunker). These sites are being investigated because existing information on the nature and extent of potential contamination is insufficient to evaluate and recommend corrective action alternatives (CAAs). The suspected contaminants and critical analyte s for CAU 214 include oil (total petroleum hydrocarbons-diesel-range organics [TPH-DRO], polychlorinated biphenyls [PCBs]), pesticides (chlordane, heptachlor, 4,4-DDT), barium, cadmium, chronium, lubricants (TPH-DRO, TPH-gasoline-range organics [GRO]), and fly ash (arsenic). The land-use zones where CAU 214 CASs are located dictate that future land uses will be limited to nonresidential (i.e., industrial) activities. The results of this field investigation will support a defensible evaluation of viable corrective action alternatives that will be presented in the corrective action decision document.

U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office

2003-05-16T23:59:59.000Z

56

Predicting Stenosis in Blood Vessels  

E-Print Network [OSTI]

of plaque Plaque is made up of cholesterol, calcium, and other blood components that stick to the vessel-flow loop is function of degree of stenosis, even at low degrees of stenosis So, stenosis may be detected

Petta, Jason

57

Applications of Engineering and Financial Analysis to the Valuation of Investments in Railroad Infrastructure  

E-Print Network [OSTI]

International Airport............................. 8 1.7 HNTB Companies Corporate Structure ..................................................... 10 1.8 HNTB Corporation Central Division Structure ......................................... 11 1.9 HNTB... Rail Group-Client Relationships .................................................... 12 1.10 Railroad Tank Car used for Transport of Hydrochloric Acid .................... 13 2.1 Traffic Arrival and Departure Patterns at Blocked Roadway...

Roco, Craig E.

2010-01-16T23:59:59.000Z

58

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

E-Print Network [OSTI]

Triaxial TestsTests Direct Shear TestsDirect Shear Tests Clean and Coal Dust Fouled Ballast BehaviorClean1 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 Dust

Barkan, Christopher P.L.

59

A technique for classifying and evaluating railroad-highway grade crossings  

E-Print Network [OSTI]

. 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...

Newton, Thomas Monroe

1974-01-01T23:59:59.000Z

60

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

E-Print Network [OSTI]

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

Barkan, Christopher P.L.

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


61

Tailoring Topology Optimization to Composite Pressure Vessel Design with Simultaneous  

E-Print Network [OSTI]

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

Paulino, Glaucio H.

62

Enhancements to passive warning devices for use at railroad-highway grade crossings  

E-Print Network [OSTI]

The Oregon PUC has currently installed illumination at over 35 rural passively controlled crossings with regular nighttime train movements. The following specifications have been used for installation. 1. At least one luminaire shall be mounted on each... side of the tracks at the crossing. Luminaires should be located so that warning devices at the crossing will be directly illuminated. 28 2. Luminaires shall be oriented toward the railroad track to provide at least one foot-candle of illumination...

Mynier, Charles Leroy

1992-01-01T23:59:59.000Z

63

Light Sources on the Nylon Vessels' Surfaces  

E-Print Network [OSTI]

the buffer thickness between the vessels could enhance -ray background in the corresponding region inside;Chapter 7: Light Sources on the Nylon Vessels' Surfaces 185 or laser). The illuminated spots can be seen the fiber's end to penetrate through the vessel membrane into the scintillator volume. A laser of a specific

64

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)

The Russian Research Reactor Fuel Return (RRRFR) Program, under the U.S. Department of Energy抯 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抯 vessels were licensed for shipping nuclear materials; however, they were not licensed to transport SNF materials. After a thorough review of ASPOL Baltic抯 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.

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

2001-10-01T23:59:59.000Z

65

Reactor pressure vessel vented head  

DOE Patents [OSTI]

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.

Sawabe, James K. (San Jose, CA)

1994-01-11T23:59:59.000Z

66

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 on Google Bookmark EERE: Alternative Fuels Data CenterFranconia, Virginia:FAQ < RAPID Jump to:Seadov Pty LtdSteen,Ltd JumpOperations JumpTooele County, Utah:JumpVessel Jump to:

67

Evaluation of enhanced traffic control devices for use at highway-railroad grade crossings  

E-Print Network [OSTI]

the standard railroad advance warning sign. The LOOK FOR TRAIN AT CROSSING system was installed at a crossing near Temple, Texas. The vehicle-activated strobe light was triggered by a loop detector and powered by a solar charged 12-volt battery. Three study... approximately five hundred speed profiles at each location. Speed profile data is more readily obtained in the field than driver behavior. Also, the low volumes experienced on these roadways and the location of the grade crossing restrict on-site observation...

Carroll, Randy W.

2000-01-01T23:59:59.000Z

68

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

SciTech Connect (OSTI)

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.

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

2007-01-01T23:59:59.000Z

69

Reactor pressure vessel vented head  

DOE Patents [OSTI]

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.

Sawabe, J.K.

1994-01-11T23:59:59.000Z

70

Device for inspecting vessel surfaces  

DOE Patents [OSTI]

A portable, remotely-controlled inspection crawler for use along the walls of tanks, vessels, piping and the like. The crawler can be configured to use a vacuum chamber for supporting itself on the inspected surface by suction or a plurality of magnetic wheels for moving the crawler along the inspected surface. The crawler is adapted to be equipped with an ultrasonic probe for mapping the structural integrity or other characteristics of the surface being inspected. Navigation of the crawler is achieved by triangulation techniques between a signal transmitter on the crawler and a pair of microphones attached to a fixed, remote location, such as the crawler's deployment unit. The necessary communications are established between the crawler and computers external to the inspection environment for position control and storage and/or monitoring of data acquisition.

Appel, D. Keith (Aiken, SC)

1995-01-01T23:59:59.000Z

71

Final report for confinement vessel analysis. Task 2, Safety vessel impact analyses  

SciTech Connect (OSTI)

This report describes two sets of finite element analyses performed under Task 2 of the Confinement Vessel Analysis Program. In each set of analyses, a charge is assumed to have detonated inside the confinement vessel, causing the confinement vessel to fail in either of two ways; locally around the weld line of a nozzle, or catastrophically into two hemispheres. High pressure gases from the internal detonation pressurize the inside of the safety vessel and accelerate the fractured nozzle or hemisphere into the safety vessel. The first set of analyses examines the structural integrity of the safety vessel when impacted by the fractured nozzle. The objective of these calculations is to determine if the high strength bolt heads attached to the nozzle penetrate or fracture the lower strength safety vessel, thus allowing gaseous detonation products to escape to the atmosphere. The two dimensional analyses predict partial penetration of the safety vessel beneath the tip of the penetrator. The analyses also predict maximum principal strains in the safety vessel which exceed the measured ultimate strain of steel. The second set of analyses examines the containment capability of the safety vessel closure when impacted by half a confinement vessel (hemisphere). The predicted response is the formation of a 0.6-inch gap, caused by relative sliding and separation between the two halves of the safety vessel. Additional analyses with closure designs that prevent the gap formation are recommended.

Murray, Y.D. [APTEK, Inc., Colorado Springs, CO (United States)

1994-01-26T23:59:59.000Z

72

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

E-Print Network [OSTI]

Vulnerability of Xylem Vessels to Cavitation in Sugar Maple. Scaling from Individual Vessels 02318 (M.A.Z., N.M.H.) The relation between xylem vessel age and vulnerability to cavitation of sugar-related changes in vulnerability to the overall resistance to cavitation, we combined data on the pressure

Melcher, Peter

73

Foam vessel for cryogenic fluid storage  

DOE Patents [OSTI]

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.

Spear, Jonathan D (San Francisco, CA)

2011-07-05T23:59:59.000Z

74

Fabrication of Separator Demonstration Facility process vessel  

SciTech Connect (OSTI)

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.

Oberst, E.F.

1985-01-15T23:59:59.000Z

75

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

76

Nuclear reactor vessel fuel thermal insulating barrier  

DOE Patents [OSTI]

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.

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

2013-03-19T23:59:59.000Z

77

Radiation embrittlement of PWR vessel supports  

SciTech Connect (OSTI)

Several studies pertaining to radiation damage of PWR vessel supports were conducted between 1978 and 1987. During this period, apparently there was no reason to believe that low-temperature (<100/degree/C) MTR embrittlement data were not appropriate for evaluating embrittlement of PWR vessel supports. However, late in 1986, data from the High Flux Isotope Reactor (HFIR) vessel surveillance program indicated that the embrittlement rates of the several HFIR vessel materials (A212-B, A350-LF3, A105-II) were substantially greater than anticipated on the basis of MTR data. Further evaluation of the HFIR data suggested that a fluence-rate effect was responsible for the apparent discrepancy, and shortly thereafter it became apparent that this rate effect was applicable to the evaluation of LWR vessel supports. As a result, the Nuclear Regulatory Commission (NRC) requested that the Oak Ridge National Laboratory (ORNL) evaluate the impact of the apparent embrittlement rate effect on the integrity of light-water-reactor (LWR) vessel supports. The purpose of the study was to provide an indication of whether the integrity of reactor vessel supports is likely to be challenged by radiation-induced embrittlement. The scope of the evaluation included correlation of the HFIR data for application to the evaluation of LWR vessel supports; a survey and cursory evaluation of all US LWR vessel support designs, selection of two plants for specific-plant evaluation, and a specific-plant evaluation of both plants to determine critical flaw sizes for their vessel supports. 19 refs., 8 figs., 2 tabs.

Cheverton, R.D.; Robinson, G.C.; Pennell, W.E.; Nanstad, R.K.

1989-01-01T23:59:59.000Z

78

Reactor pressure vessel. Status report  

SciTech Connect (OSTI)

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.

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

1996-10-01T23:59:59.000Z

79

Lightweight bladder lined pressure vessels  

DOE Patents [OSTI]

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.

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

80

Lightweight bladder lined pressure vessels  

DOE Patents [OSTI]

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.

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

1998-08-25T23:59:59.000Z

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


81

Thermal wake/vessel detection technique  

DOE Patents [OSTI]

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.

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

2012-01-10T23:59:59.000Z

82

HFIR vessel probabilistic fracture mechanics analysis  

SciTech Connect (OSTI)

The life of the High Flux Isotope Reactor (HFIR) pressure vessel is limited by a radiation induced reduction in the material`s fracture toughness. Hydrostatic proof testing and probabilistic fracture mechanics analyses are being used to meet the intent of the ASME Code, while extending the life of the vessel well beyond its original design value. The most recent probabilistic evaluation is more precise and accounts for the effects of gamma as well as neutron radiation embrittlement. This analysis confirms the earlier estimates of a permissible vessel lifetime of at least 50 EFPY (100 MW).

Cheverton, R.D. [Delta-21 Resources, Inc., Oak Ridge, TN (United States); Dickson, T.L. [Oak Ridge National Lab., TN (United States)

1997-01-01T23:59:59.000Z

83

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

84

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

85

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]

'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...

Neyer, David McAlister

1994-01-01T23:59:59.000Z

86

Ion transport membrane module and vessel system  

SciTech Connect (OSTI)

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.

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

87

Ion transport membrane module and vessel system  

DOE Patents [OSTI]

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.

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

88

Future characteristics of Offshore Support Vessels  

E-Print Network [OSTI]

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 ...

Rose, Robin Sebastian Koske

2011-01-01T23:59:59.000Z

89

Neutron shielding panels for reactor pressure vessels  

DOE Patents [OSTI]

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.

Singleton, Norman R. (Murrysville, PA)

2011-11-22T23:59:59.000Z

90

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

E-Print Network [OSTI]

?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...

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

2008-01-01T23:59:59.000Z

91

EDS V25 containment vessel explosive qualification test report.  

SciTech Connect (OSTI)

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.

Rudolphi, John Joseph

2012-04-01T23:59:59.000Z

92

Initial conditioning of the TFTR vacuum vessel  

SciTech Connect (OSTI)

We report on the initial conditioning of the Tokamak Fusion Test Reactor (TFTR) vacuum vessel prior to the initiation of first plasma discharges, and during subsequent operation with high power ohmically-heated plasmas. Following evacuation of the 86 m/sup 3/ vessel with the 10/sup 4/ 1/s high vacuum pumping system, the vessel was conditioned by a 15 A dc glow discharge in H/sub 2/ at a pressure of 5 mTorr. Rapid-pulse discharge cleaning was used subsequently to preferentially condition the graphite plasma limiters. The effectiveness of the discharge cleaning was monitored by measuring the exhaust rates of the primary discharge products (CO/C/sub 2/H/sub 4/, CH/sub 4/, and H/sub 2/O). After 175 hours of glow discharge treatment, the equivalent of 50 monolayers of C and O was removed from the vessel, and the partial pressures of impurity gases were reduced to the range of 10/sup -9/-10/sup -10/ Torr.

Dylla, H.F.; Blanchard, W.R.; Krawchuk, R.B.; Hawryluk, R.J.; Owens, D.K.

1984-01-01T23:59:59.000Z

93

Reactor pressure vessel with forged nozzles  

DOE Patents [OSTI]

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.

Desai, Dilip R. (Fremont, CA)

1993-01-01T23:59:59.000Z

94

Study Reveals Challenges and Opportunities Related to Vessels...  

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

Study 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,...

95

E-Print Network 3.0 - adjacent vessel sign Sample Search Results  

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

of Maximum Intensity Summary: three-dimensional (3-D) data, which are vessel voxel projection probability, vessel detection... probability, false vessel probability, and...

96

Photoacoustic removal of occlusions from blood vessels  

DOE Patents [OSTI]

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.

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

97

From Cold War to cold vessels  

SciTech Connect (OSTI)

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.

Melrath, C.

1996-09-01T23:59:59.000Z

98

TMI-2 reactor vessel head removal  

SciTech Connect (OSTI)

This report describes the safe removal and storage of the Three Mile Island Unit 2 reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training the head was safely removed and stored and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities.

Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

1984-12-01T23:59:59.000Z

99

TMI-2 reactor vessel head removal  

SciTech Connect (OSTI)

This report describes the safe removal and storage of the Three Mile Island Unit 2 (TMI-2) reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training, the head was safely removed and stored; and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities.

Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

1985-09-01T23:59:59.000Z

100

Water Quality Impacts of Bunker Silos  

E-Print Network [OSTI]

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

Balser, Teri C.

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


101

Unity, Justice and Protection: The Colored Trainmen of America's Struggle to End Jim Crow in the American Railroad Industry [and Elsewhere  

E-Print Network [OSTI]

National Advancement Association for Colored People NMB National Mediation Board NLRB National Labor Relations Board ix NRAB National Railroad Adjustment Board NRLEC Negro Railway... committee of five highly organized black labor unions known as the Negro Railway Labor Executives Committee (NRLEC). Collectively, the CTA, the Association of Colored Railway Trainmen and Locomotive Firemen (ACRTLF), International Association of Railway...

James, Ervin

2012-10-19T23:59:59.000Z

102

Autonomous Radiation Monitoring of Small Vessels  

SciTech Connect (OSTI)

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.

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

2010-01-01T23:59:59.000Z

103

Application of Computational Physics: Blood Vessel Constrictions and Medical Infuses  

E-Print Network [OSTI]

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...

Suprijadi,; Subekti, Petrus; Viridi, Sparisoma

2013-01-01T23:59:59.000Z

104

Application of Computational Physics: Blood Vessel Constrictions and Medical Infuses  

E-Print Network [OSTI]

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.

Suprijadi; Mohamad Rendi; Petrus Subekti; Sparisoma Viridi

2013-12-14T23:59:59.000Z

105

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.

106

A Xenon Condenser with a Remote Liquid Storage Vessel  

E-Print Network [OSTI]

We describe the design and operation of a system for xenon liquefaction in which the condenser is separated from the liquid storage vessel. The condenser is cooled by a pulse tube cryocooler, while the vessel is cooled only by the liquid xenon itself. This arrangement facilitates liquid particle detector research by allowing easy access to the upper and lower flanges of the vessel. We find that an external xenon gas pump is useful for increasing the rate at which cooling power is delivered to the vessel, and we present measurements of the power and efficiency of the apparatus.

S. Slutsky; Y. -R. Yen; H. Breuer; A. Dobi; C. Hall; T. Langford; D. S. Leonard; L. J. Kaufman; V. Strickland; N. Voskanian

2009-07-25T23:59:59.000Z

107

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...

108

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...

109

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...

110

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...

111

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...

112

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...

113

Heat-transfer coefficients in agitated vessels. Sensible heat models  

SciTech Connect (OSTI)

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.

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

1995-12-01T23:59:59.000Z

114

Austenite Grain Growth in a Nuclear Pressure Vessel Steel  

E-Print Network [OSTI]

. 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

Cambridge, University of

115

Coalesced Martensite in Pressure Vessel Steels Hector Pous-Romero  

E-Print Network [OSTI]

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

Cambridge, University of

116

Simultaneous Irradiation and Imaging of Blood Vessels During Pulsed  

E-Print Network [OSTI]

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

Barton, Jennifer K.

117

Commercial marine vessel contributions to emission inventories. Final report  

SciTech Connect (OSTI)

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).

Not Available

1991-10-07T23:59:59.000Z

118

Nuclear reactor construction with bottom supported reactor vessel  

DOE Patents [OSTI]

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.

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

1987-01-01T23:59:59.000Z

119

Radiation effects on reactor pressure vessel supports  

SciTech Connect (OSTI)

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.

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

120

Reactor Pressure Vessel Head Packaging & Disposal  

SciTech Connect (OSTI)

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.

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

2003-02-26T23:59:59.000Z

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


121

Start-up control system and vessel for LMFBR  

DOE Patents [OSTI]

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.

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

1987-01-01T23:59:59.000Z

122

Start-up control system and vessel for LMFBR  

DOE Patents [OSTI]

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.

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

1987-01-01T23:59:59.000Z

123

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

SciTech Connect (OSTI)

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)

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

124

Nondestructive Technique Survey for Assessing Integrity of Composite Firing Vessel  

SciTech Connect (OSTI)

The repeated use and limited lifetime of a composite tiring vessel compel a need to survey techniques for monitoring the structural integrity of the vessel in order to determine when it should be retired. Various nondestructive techniques were researched and evaluated based on their applicability to the vessel. The methods were visual inspection, liquid penetrant testing, magnetic particle testing, surface mounted strain gauges, thermal inspection, acoustic emission, ultrasonic testing, radiography, eddy current testing, and embedded fiber optic sensors. It was determined that embedded fiber optic sensor is the most promising technique due to their ability to be embedded within layers of composites and their immunity to electromagnetic interference.

Tran, A.

2000-08-01T23:59:59.000Z

125

Float level switch for a nuclear power plant containment vessel  

DOE Patents [OSTI]

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.

Powell, James G. (Clifton Park, NY)

1993-01-01T23:59:59.000Z

126

Float level switch for a nuclear power plant containment vessel  

DOE Patents [OSTI]

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.

Powell, J.G.

1993-11-16T23:59:59.000Z

127

Processing and analysis techniques involving in-vessel material generation  

DOE Patents [OSTI]

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.

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

2012-09-25T23:59:59.000Z

128

Processing and analysis techniques involving in-vessel material generation  

DOE Patents [OSTI]

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.

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

2011-01-25T23:59:59.000Z

129

Assessment of Vessel Requirements for the U.S. Offshore Wind...  

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

Wind Sector: Executive Summary Assessment of Vessel Requirements for the U.S. Offshore Wind Sector: Executive Summary Executive summary of the Assessment of Vessel Requirements for...

130

E-Print Network 3.0 - automatic vessel identification Sample...  

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

Review Regulatory Revisions to Summary: .S.C. 971 et seq.) authorize the Secretary to station observers aboard commercial fishing vessels in order... Service). Vessels fishing...

131

E-Print Network 3.0 - artificial blood vessel Sample Search Results  

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

blood vessel without obstructing normal blood flow. Their work takes advantage of microfluidics... by the reactions that prevent clotting (inhibition). When a blood vessel is...

132

Hydrodynamic evaluation of high-speed semi-SWATH vessels  

E-Print Network [OSTI]

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 ...

Guttenplan, Adam (Adam David)

2007-01-01T23:59:59.000Z

133

DESIGN OF THE ITER IN-VESSEL COILS  

SciTech Connect (OSTI)

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.

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

134

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...

135

Microsoft Word - 911118_a Vessel_Alternatives-Judy1.doc  

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

the oil and gas industry. As with other RPV and IHX Pressure Vessel Alternatives Study Report 9111180 6 MOC, more data are needed on compatibility with impure He gas. This...

136

Vessel Segmentation with Automatic Centerline Extraction Using Tubular Tree Segmentation  

E-Print Network [OSTI]

deaths in the United States per year. Vessel Segmentation from CTA data is challenging because of non Mohan1 , Ganesh Sundaramoorthi1,2 , Arthur Stillman3 , and Allen Tannenbaum1 1 School of Electrical

Paris-Sud XI, Universit茅 de

137

Using SA508/533 for the HTGR Vessel Material  

SciTech Connect (OSTI)

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.

Larry Demick

2012-06-01T23:59:59.000Z

138

A cog-like vessel from the Netherlands  

E-Print Network [OSTI]

, 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...

Van de Moortel, Aleydis Maria P. A.

1987-01-01T23:59:59.000Z

139

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

E-Print Network [OSTI]

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

American Society for Testing and Materials. Philadelphia

2010-01-01T23:59:59.000Z

140

Investigation of vessel exterior air cooling for a HLMC reactor  

SciTech Connect (OSTI)

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.

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

2000-01-13T23:59:59.000Z

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


141

Investigation of vessel exterior air cooling for an HLMC reactor  

SciTech Connect (OSTI)

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.

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

2000-07-01T23:59:59.000Z

142

Retrospective dosimetry analyses of reactor vessel cladding samples  

SciTech Connect (OSTI)

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)

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

143

Sterilization of fermentation vessels by ethanol/water mixtures  

DOE Patents [OSTI]

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.

Wyman, Charles E. (Lakewood, CO)

1999-02-09T23:59:59.000Z

144

Sterilization of fermentation vessels by ethanol/water mixtures  

DOE Patents [OSTI]

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.

Wyman, C.E.

1999-02-09T23:59:59.000Z

145

Report of the terawatt laser pressure vessel committee  

SciTech Connect (OSTI)

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.

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

2000-09-25T23:59:59.000Z

146

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

SciTech Connect (OSTI)

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.

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

147

Lightweight cryogenic-compatible pressure vessels for vehicular fuel storage  

DOE Patents [OSTI]

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.

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

2004-03-23T23:59:59.000Z

148

Simulation of Diffusive Lithium Evaporation Onto the NSTX Vessel Walls  

SciTech Connect (OSTI)

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.

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

149

Ex-vessel demand by size for the Gulf shrimp  

E-Print Network [OSTI]

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...

Chui, Margaret Kam-Too

1980-01-01T23:59:59.000Z

150

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)

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

Mark Kauss

2011-06-01T23:59:59.000Z

151

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

DOE Patents [OSTI]

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.

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

2010-02-09T23:59:59.000Z

152

Thrust allocation with power management functionality on dynamically positioned vessels  

E-Print Network [OSTI]

world-wide. The main benefits of diesel-electric propulsion and thrusters are reduced power consumptionThrust allocation with power management functionality on dynamically positioned vessels Aleksander to assist the power management system on dynamically positioned ships is proposed in this paper. Its main

Johansen, Tor Arne

153

Modelling the Induced Magnetic Signature of Naval Vessels  

E-Print Network [OSTI]

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

Low, Robert

154

Acoustic emission monitoring of HFIR vessel during hydrostatic testing  

SciTech Connect (OSTI)

This report discusses the results and conclusions reached from applying acoustic emission monitoring to surveillance of the High Flux Isotope Reactor vessel during pressure testing. The objective of the monitoring was to detect crack growth and/or fluid leakage should it occur during the pressure test. The report addresses the approach, acoustic emission instrumentation, installation, calibration, and test results.

Friesel, M.A.; Dawson, J.F.

1992-08-01T23:59:59.000Z

155

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

156

Beryllium pressure vessels for creep tests in magnetic fusion energy  

SciTech Connect (OSTI)

Beryllium has interesting applications in magnetic fusion experimental machines and future power-producing fusion reactors. Chief among the properties of beryllium that make these applications possible is its ability to act as a neutron multiplier, thereby increasing the tritium breeding ability of energy conversion blankets. Another property, the behavior of beryllium in a 14-MeV neutron environment, has not been fully investigated, nor has the creep behavior of beryllium been studied in an energetic neutron flux at thermodynamically interesting temperatures. This small beryllium pressure vessel could be charged with gas to test pressures around 3, 000 psi to produce stress in the metal of 15,000 to 20,000 psi. Such stress levels are typical of those that might be reached in fusion blanket applications of beryllium. After contacting R. Powell at HEDL about including some of the pressure vessels in future test programs, we sent one sample pressure vessel with a pressurizing tube attached (Fig. 1) for burst tests so the quality of the diffusion bond joints could be evaluated. The gas used was helium. Unfortunately, budget restrictions did not permit us to proceed in the creep test program. The purpose of this engineering note is to document the lessons learned to date, including photographs of the test pressure vessel that show the tooling necessary to satisfactorily produce the diffusion bonds. This document can serve as a starting point for those engineers who resume this task when funds become available.

Neef, W.S.

1990-07-20T23:59:59.000Z

157

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

158

PublicationsmailagreementNo.40014024 the VeSSeL WILL  

E-Print Network [OSTI]

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

Pedersen, Tom

159

Sampling and Analysis Plan for PUREX canyon vessel flushing  

SciTech Connect (OSTI)

A sampling and analysis plan is necessary to provide direction for the sampling and analytical activities determined by the data quality objectives. This document defines the sampling and analysis necessary to support the deactivation of the Plutonium-Uranium Extraction (PUREX) facility vessels that are regulated pursuant to Washington Administrative Code 173-303.

Villalobos, C.N.

1995-03-01T23:59:59.000Z

160

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...

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


161

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

Office of Environmental Management (EM)

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...

162

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

163

Effects of ionizing radiation on normal and tumor-associated lymphatic vessels  

E-Print Network [OSTI]

Lymphatic vessels play a crucial role in both the pathophysiology of tumors and in the spread cancer cells to lymph nodes. The effects of radiation on these vessels, however, are largely unknown. Here, we seek to describe ...

Lobo, Jennifer D

2007-01-01T23:59:59.000Z

164

E-Print Network 3.0 - arch vessel transposition Sample Search...  

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

the effect of applying local vacuum pressure on the temperatures of the epidermis and small vessels during... of skin and blood vessels with different diameters (10-60 mm) at...

165

E-Print Network 3.0 - automatic vessel control Sample Search...  

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

of the vessel boundary, we can ob- tain an automatic computation method... by the fixed-scale GVF snake. In turn, the success of automated vessel ... Source: Acton, Scott -...

166

A Survey of Pressure Vessel Code Compliance for Superconducting RF Cryomodules  

SciTech Connect (OSTI)

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.

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

167

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

E-Print Network [OSTI]

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

Motta, Arthur T.

168

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 injected into the vessel. ITER has focused considerable effort into the ability to thermally remove tritium vessel. Another possible technique to mitigate tritium accumulation in these codeposited surfaces

169

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

E-Print Network [OSTI]

RIS?-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

170

Superheat effects on localized vessel breach enlargement during corium ejection  

SciTech Connect (OSTI)

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.

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

1986-01-01T23:59:59.000Z

171

Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database  

SciTech Connect (OSTI)

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.

Wang, Jy-An John [ORNL

2010-08-01T23:59:59.000Z

172

PRESSURIZATION OF CONTAINMENT VESSELS FROM PLUTONIUM OXIDE CONTENTS  

SciTech Connect (OSTI)

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.

Hensel, S.

2012-03-27T23:59:59.000Z

173

Plastic instabilities in statically and dynamically loaded spherical vessels  

SciTech Connect (OSTI)

Significant changes were made in design limits for pressurized vessels in the 2007 version of the ASME Code (Section VIII, Div. 3) and 2008 and 2009 Addenda. There is now a local damage-mechanics based strain-exhaustion limit as well as the well-known global plastic collapse limit. Moreover, Code Case 2564 (Section VIII, Div. 3) has recently been approved to address impulsively loaded vessels. It is the purpose of this paper to investigate the plastic collapse limit as it applies to dynamically loaded spherical vessels. Plastic instabilities that could potentially develop in spherical shells under symmetric loading conditions are examined for a variety of plastic constitutive relations. First, a literature survey of both static and dynamic instabilities associated with spherical shells is presented. Then, a general plastic instability condition for spherical shells subjected to displacement controlled and impulsive loading is given. This instability condition is evaluated for six plastic and visco-plastic constitutive relations. The role of strain-rate sensitivity on the instability point is investigated. Calculations for statically and dynamically loaded spherical shells are presented, illustrating the formation of instabilities as well as the role of imperfections. Conclusions of this work are that there are two fundamental types of instabilities associated with failure of spherical shells. In the case of impulsively loaded vessels, where the pulse duration is short compared to the fundamental period of the structure, one instability type is found not to occur in the absence of static internal pressure. Moreover, it is found that the specific role of strain-rate sensitivity on the instability strain depends on the form of the constitutive relation assumed.

Duffey, Thomas A [Los Alamos National Laboratory; Rodriguez, Edward A [Los Alamos National Laboratory

2010-01-01T23:59:59.000Z

174

IMPACT OF NUCLEAR MATERIAL DISSOLUTION ON VESSEL CORROSION  

SciTech Connect (OSTI)

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.

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

2012-10-01T23:59:59.000Z

175

Heat-transfer coefficients in agitated vessels. Latent heat models  

SciTech Connect (OSTI)

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.

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

1996-03-01T23:59:59.000Z

176

Fast neutron fluxes in pressure vessels using Monte Carlo methods  

SciTech Connect (OSTI)

The objective of this project is to determine the feasibility of calculating the fast neutron flux in the pressure vessel of a pressurized water reactor by Monte Carlo methods. Neutron reactions reduce the ductility of the steel and thus limit the useful life of this important reactor component. This work was performed for Virginia Power (VEPCO). VIM is a continuous-energy Monte Carlo code which provides a versatile geometrical capability and a neutron physics data base closely representing the EDNF/B-IV data from which it was derived.

Edlund, M.C.; Thomas, J.R.

1986-01-01T23:59:59.000Z

177

Fabrication Flaws in Reactor Pressure Vessel Repair Welds  

SciTech Connect (OSTI)

This paper describes the fabrication flaw distribution and characterization in the repair weld metal of reactor pressure vessels. This work indicates that the large flaws occur in these repairs. These results show that repair flaws are complex in composition and sometimes include cracks on the repair ends. Parametric analysis using an exponential fit is performed on the data. A description of repair flaw morphology is provided. Fabrication flaws in repairs are characterized using high sensitivity nondestructive ultrasonic testing, validation by other nondestructive evaluation (NDE) techniques, and complemented by destructive testing.

Schuster, George J.; Doctor, Steven R.

2007-12-01T23:59:59.000Z

178

Index of /research/alcator/facility/Procedures/IN-VESSEL  

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: Vegetation ProposedUsingFun withconfinementEtching.348ASSEMBLY [ICO] Name Last modified SizeIN-VESSEL

179

Gamma ray-induced embrittlement of pressure vessel alloys  

SciTech Connect (OSTI)

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.

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

180

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 on Google Bookmark EERE: Alternative Fuels Data Center Home5b9fcbce19 NoPublic Utilities Address: 160Benin: EnergyBoston Areais a villageBucyrus, NorthBuhler,Bundesverband

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


181

Nuclear reactor having a polyhedral primary shield and removable vessel insulation  

DOE Patents [OSTI]

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.

Ekeroth, D.E.; Orr, R.

1993-12-07T23:59:59.000Z

182

Nuclear reactor having a polyhedral primary shield and removable vessel insulation  

DOE Patents [OSTI]

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.

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

1993-01-01T23:59:59.000Z

183

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

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

Princeton Plasma Physics Laboratory, Theory Department Collection: Plasma Physics and Fusion 34 LET: a cold neutron chopper spectrometer on ISIS TS2 Summary: Jaws Vacuum Vessel...

184

E-Print Network 3.0 - abnormal blood vessels Sample Search Results  

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

of Summary: ) the blood vessels, which also helps to lower blood pressure. Commonly used brand names in the United States... to treat high blood pressure, heart disease and...

185

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

186

E-Print Network 3.0 - assessing blood vessel Sample Search Results  

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

vascular tortuosity Lambda William E. Hart Summary: as the retinal blood vessel net- work. To assess the relative utility of these mea- sures, they were used... . Nelson Data...

187

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

E-Print Network [OSTI]

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 ...

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

2010-01-01T23:59:59.000Z

188

High-pressure Storage Vessels for Hydrogen, Natural Gas andHydrogen...  

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

Gas and Blends - Materials Testing and Design Requirements for Hydrogen Components and Tanks International Hydrogen Fuel and Pressure Vessel Forum 2010 Proceedings Hydrogen...

189

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

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

Hydrogen Fuel and Pressure Vessel Forum Bonfire Tests of High Pressure Hydrogen Storage Tanks Status and Progress in Research, Development and Demonstration of Hydrogen-Compressed...

190

ASTM Standards for Reactor Dosimetry and Pressure Vessel Surveillance  

SciTech Connect (OSTI)

The ASTM standards provide guidance and instruction on how to field and interpret reactor dosimetry. They provide a roadmap towards understanding the current ''state-of-the-art'' in reactor dosimetry, as reflected by the technical community. The consensus basis to the ASTM standards assures the user of an unbiased presentation of technical procedures and interpretations of the measurements. Some insight into the types of standards and the way in which they are organized can assist one in using them in an expeditious manner. Two example are presented to help orient new users to the breadth and interrelationship between the ASTM nuclear metrology standards. One example involves the testing of a new ''widget'' to verify the radiation hardness. The second example involves quantifying the radiation damage at a pressure vessel critical weld location through surveillance dosimetry and calculation.

GRIFFIN, PATRICK J.

1999-09-14T23:59:59.000Z

191

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

SciTech Connect (OSTI)

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抯 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.

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

192

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

SciTech Connect (OSTI)

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.

Rucinski, Russ; /Fermilab

1996-11-11T23:59:59.000Z

193

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

SciTech Connect (OSTI)

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)

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

194

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

E-Print Network [OSTI]

with increasing diameter, and total contraction time (vessel twitch length, 11.08 1.54 s) slightly increases with increasing diameter. At the intersection of these relationships, there is a critical period, below which the vessel does not have time to fully...

Meisner, Joshua Keith

2009-06-02T23:59:59.000Z

195

RESIDUAL STRESS DISTRIBUTIONS FOR MULTI-PASS WELDS IN PRESSURE VESSEL AND PIPING COMPONENTS  

E-Print Network [OSTI]

RESIDUAL STRESS DISTRIBUTIONS FOR MULTI-PASS WELDS IN PRESSURE VESSEL AND PIPING COMPONENTS distributions in common pressure vessel and piping components is generated by using the multi-pass finite-walled pipes with various radius to thickness ratios. Both single- and double-V weld joints are investigated

Michaleris, Panagiotis

196

Pipeline and Pressure Vessel R&D under the Hydrogen Regional Infrastructure  

E-Print Network [OSTI]

Pipeline and Pressure Vessel R&D under the Hydrogen Regional Infrastructure Program In Pennsylvania Kevin L. Klug, Ph.D. 25 September 2007 DOE Hydrogen Pipeline Working Group Meeting, Aiken, SC & Sensors Hydrogen Delivery Composite Overwrapped Pressure Vessels (COPVs) Pipeline for Off-Board Hydrogen

197

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

198

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

199

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

E-Print Network [OSTI]

-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

Barton, Jennifer K.

200

Modelling the Electron Beam Welding of Nuclear Reactor Pressure Vessel Steel  

E-Print Network [OSTI]

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

Cambridge, University of

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


201

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

E-Print Network [OSTI]

to safety structures (ITS) such as pressure vessels and piping (PVP) in a nuclear reactor. Technologies been tested before. However the irradiation effects, pertinent to nuclear facilities for PWAS, have1 Copyright 漏 2012 by ASME Proceedings of the ASME 2012 Pressure Vessels & Piping Division

Giurgiutiu, Victor

202

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

203

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

SciTech Connect (OSTI)

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.

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

204

Creep of A508/533 Pressure Vessel Steel  

SciTech Connect (OSTI)

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, 2糃r-1Mo in the annealed condition, and Grade 91 steel. The low strength of 2糃r-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癈 from 950 to 1000癈, 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癈 (700癋); 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癈 (700癋) 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.

Richard Wright

2014-08-01T23:59:59.000Z

205

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

E-Print Network [OSTI]

'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...

Cook, Gregory D.

1997-01-01T23:59:59.000Z

206

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

E-Print Network [OSTI]

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 ...

DeWitt, Gregory L

2011-01-01T23:59:59.000Z

207

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

208

Method for forming a bladder for fluid storage vessels  

DOE Patents [OSTI]

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.

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

2000-01-01T23:59:59.000Z

209

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

DOE Patents [OSTI]

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.

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

1997-01-01T23:59:59.000Z

210

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

DOE Patents [OSTI]

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.

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

1997-12-16T23:59:59.000Z

211

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

E-Print Network [OSTI]

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 ...

Brett, Bridget C

2008-01-01T23:59:59.000Z

212

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...

213

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

E-Print Network [OSTI]

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...

Zhang, Ming

2009-05-15T23:59:59.000Z

214

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

SciTech Connect (OSTI)

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.

NONE

1996-08-01T23:59:59.000Z

215

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

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

hydrogen accelerates crack propagation rate of the material and leads to brittle fracture. International Hydrogen Fuel and Pressure Vessel Forum 2010Beijing, P.R. China R&D...

216

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

E-Print Network [OSTI]

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

Paris-Sud XI, Universit茅 de

217

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

E-Print Network [OSTI]

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...

Peters, Andrew

2011-08-04T23:59:59.000Z

218

INAA and distribution patterns of Classic Mimbres Black-on-white vessels during the Classic period  

E-Print Network [OSTI]

Distribution patterns of Classic Mimbres Black-on-white (Style III) bowls and jars were determined by instrumental neutron activation analysis (INAA) to identify vessel movement between geographically defined regions and between villages within...

Dahlin, Eleanor Sherlock

2003-01-01T23:59:59.000Z

219

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

E-Print Network [OSTI]

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 ...

Fuquen-Molano, Rosendo

1982-01-01T23:59:59.000Z

220

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

SciTech Connect (OSTI)

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.

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

2010-07-01T23:59:59.000Z

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


221

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

E-Print Network [OSTI]

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...

American Society for Testing and Materials. Philadelphia

2010-01-01T23:59:59.000Z

222

William W. Hay Railroad Engineering Seminar Freight Railroad Energy  

E-Print Network [OSTI]

(importance v difficulties) North American freight RRs (defining characteristics) Energy density of diesel fuel & alternatives (Btu's per gallon) Biodiesel, Fischer-Tropsch syn fuel & DME Liquefied natural of European RRs (lessons learned) Dual-mode locomotives ("electro-diesels") Unconventional alternatives

Barkan, Christopher P.L.

223

Embolization of Collateral Vessels Using Mechanically Detachable Coils in Young Children with Congenital Heart Disease  

SciTech Connect (OSTI)

Our objective was to evaluate the usefulness of embolizing collateral vessels using mechanically detachable coils (MDCs) in children aged 3 years or younger with congenital heart disease. The subjects were 8 children with congenital heart disease featuring collateral vessels (age 18 days-3 years): 3 with a single ventricle, 2 with the tetralogy of Fallot, 2 with pulmonary atresia, and 1 with a ventricular septal defect. The embolized vessels were the major aortopulmonary collateral artery (MAPCA) in 5 patients, the persistent left superior vena cava in 2, and the coronary arteriovenous fistula in 1. A 4 or a 5 F catheter was used as the guiding device, and embolization was performed using MDCs and other conventional coils introduced through the microcatheter. One patient had growth of new MAPCAs after embolization, and these MAPCAs were also embolized with MDCs. Thus, a total of 9 embolization procedures were performed in 8 patients. Complete occlusion of the collateral vessels was achieved in 8 of 9 procedures (89%). Seven of 8 patients (88%) had uneventful courses after embolization, and MDC procedures appeared to play important roles in avoiding coil migration and achievement of safe coil embolization. One patient who underwent MAPCA embolization showed no improvement in heart function and died 2 months and 19 days later. Embolization of collateral vessels using MDCs in young children with congenital heart disease can be an effective procedure and a valuable adjunct to surgical management.

Sato, Y.; Ogino, H.; Hara, M.; Satake, M.; Oshima, H.; Banno, T. [Nagoya City University Medical School, Department of Radiology (Nagoya) (Japan); Mizuno, K. [Nagoya City University Medical School, Department of Pediatrics (Nagoya) (Japan); Mishima, A. [Nagoya City University Medical School, Department of Cardiovascular Surgery (Nagoya) (Japan); Shibamoto, Y. [Nagoya City University Medical School, Department of Radiology (Nagoya) (Japan)

2003-11-15T23:59:59.000Z

224

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

SciTech Connect (OSTI)

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.

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

1995-12-01T23:59:59.000Z

225

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

SciTech Connect (OSTI)

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

David H. Bothell

2000-04-30T23:59:59.000Z

226

Reactor pressure vessel head vents and methods of using the same  

DOE Patents [OSTI]

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.

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

2014-10-28T23:59:59.000Z

227

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

SciTech Connect (OSTI)

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.

L. C. Cadwallader

2013-01-01T23:59:59.000Z

228

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

E-Print Network [OSTI]

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...

American Society for Testing and Materials. Philadelphia

2003-01-01T23:59:59.000Z

229

PowerPoint Presentation  

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

Boilers Vessel Press Bridge Duct work Railroad Engine Tank Relocation 440T Building Solid Masonry Even a building facade "These guys can move anything" Pressurizer RA...

230

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

SciTech Connect (OSTI)

The Waste Treatment and Immobilization Plant (WTP) at Hanford is being designed and built to pre-treat and vitrify the waste in Hanford抯 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 搄ust-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.

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

231

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

SciTech Connect (OSTI)

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.

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

1994-06-01T23:59:59.000Z

232

An evaluation of life extension of the HFIR pressure vessel. Supplement 1  

SciTech Connect (OSTI)

Preliminary analyses were performed in 1994 to determine the remaining useful life of the HFIR pressure vessel. The estimated total permissible life was {approximately} 50 EFPY (100 MW). More recently, the analyses have been updated, including a more precise treatment of uncertainties in the calculation of the hydrostatic-proof-test conditions and also including the contribution of gammas to the radiation-induced reduction in fracture toughness. These and other refinements had essentially no effect on the predicted useful life of the vessel or on the specified hydrostatic proof-test conditions.

Cheverton, R.D.

1996-08-01T23:59:59.000Z

233

Multidimensional shielding analysis of the JASPER in-vessel fuel storage experiments  

SciTech Connect (OSTI)

The In-Vessel Fuel Storage (IVFS) experiments analyzed in this report were conducted at the Oak Ridge National Laboratory`s Tower Shielding Reactor (TSR) as part of the Japanese-American Shielding Program for Experimental Research (JASPER). These IVFS experiments were designed to study source multiplication and three-dimensional effects related to in-vessel storage of spent fuel elements in liquid metal reactor (LMR) systems. The present report describes the 2-D and 3-D models, analyses, and calculated results corresponding to a limited subset of those IVFS experiments in which the US LMR program has a particular interest.

Bucholz, J.A.

1993-03-01T23:59:59.000Z

234

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]

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...

Perini, Silvia

2014-07-03T23:59:59.000Z

235

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

E-Print Network [OSTI]

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

Chen, Sheng

236

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

E-Print Network [OSTI]

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

Lewandowski, Roger

237

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

E-Print Network [OSTI]

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

Bridges, Tom

238

Acoustic emission monitoring of HFIR vessel during hydrostatic testing. Final report  

SciTech Connect (OSTI)

This report discusses the results and conclusions reached from applying acoustic emission monitoring to surveillance of the High Flux Isotope Reactor vessel during pressure testing. The objective of the monitoring was to detect crack growth and/or fluid leakage should it occur during the pressure test. The report addresses the approach, acoustic emission instrumentation, installation, calibration, and test results.

Friesel, M.A.; Dawson, J.F.

1992-08-01T23:59:59.000Z

239

An Integrated View on Plasma Power Exhaust and In-vessel Components  

E-Print Network [OSTI]

divertor plates able to remove 10 MW/m2 have been produced (ITER water cooled) or are under developmentAn Integrated View on Plasma Power Exhaust and In-vessel Components 20% of the energy produced generated during these transition phases. In addition, instabilities of plasma will produce off

240

Measurements of the hydrogenic recombination coefficient for the TFTR vacuum vessel  

SciTech Connect (OSTI)

Characteristic values of the recombination rate coefficient for hydrogen and deuterium in stainless steel have been measured for the inner wall of the TFTR vacuum vessel for vessel temperatures of 25 to 100 C. In situ measurements of k/sub r/ are important for predicting the hydrogen isotope retention in the wall as a function of time, temperature, and discharge exposure, particularly because existing laboratory measurements of k/sub r/ for stainless steel span a range of four orders of magnitude. The measurement technique involved the observation of the decrease in hydrogen pressure during a glow discharge in the TFTR vacuum vessel with an initial static gas fill. The resulting values of k/sub r/ at 25 C are in the range of (0.4 to 4) x 10/sup -27/cm/sup 4/-s/sup -1/ assuming a value of the hydrogenic diffusivity of 2 x 10/sup -12/cm/sup 2/-s/sup -1/ at room temperature. No significant isotopic dependence was observed and the temperature dependence of k/sub r/ is consistent with the literature value (0.5 eV) of the activation energy. The implications of this range of values of k/sub r/, for the estimation of the in-vessel tritium inventory following D-T operation in TFTR are discussed.

Dylla, H.F.; Cecchi, J.L.; Knize, R.J.

1983-12-01T23:59:59.000Z

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


241

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

SciTech Connect (OSTI)

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.

NONE

1995-01-01T23:59:59.000Z

242

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

E-Print Network [OSTI]

1 Copyright 漏 2012 by ASME Proceedings of the ASME 2012 Pressure Vessels & Piping Division - DOUBLE PIPE - ORIFICE SYSTEM Arris S. TIJSSELING Department of Mathematics and Computer Science Eindhoven Acoustic resonance in a two-pipe system is simulated with four different models for the periodic excitation

Tijsseling, A.S.

243

V1.6 Development of Advanced Manufacturing Technologies for Low Cost Hydrogen Storage Vessels  

SciTech Connect (OSTI)

The goal of this project is to develop an innovative manufacturing process for Type IV high-pressure hydrogen storage vessels, with the intent to significantly lower manufacturing costs. Part of the development is to integrate the features of high precision AFP and commercial FW. Evaluation of an alternative fiber to replace a portion of the baseline fiber will help to reduce costs further.

Leavitt, Mark; Lam, Patrick; Nelson, Karl M.; johnson, Brice A.; Johnson, Kenneth I.; Alvine, Kyle J.; Ruiz, Antonio; Adams, Jesse

2012-10-01T23:59:59.000Z

244

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

E-Print Network [OSTI]

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

Erdem, Erkut

245

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

246

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

247

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

E-Print Network [OSTI]

CALCULATIONS OF TRITIUM FLOW BETWEEN THE BUFFER VESSEL UP TO THE FIRST VACUUM SYSTEM Felix Sharipov diff., Eq.(32) 碌 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

Sharipov, Felix

248

MAGNESIUM MONO POTASSIUM PHOSPHATE GROUT FOR P-REACTOR VESSEL IN-SITU DECOMISSIONING  

SciTech Connect (OSTI)

The objective of this report is to document laboratory testing of magnesium mono potassium phosphate grouts for P-Reactor vessel in-situ decommissioning. Magnesium mono potassium phosphate 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 (pH of about 12.4). A less alkaline material ({<=} 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. Information concerning access points into the P-Reactor vessel and amount of aluminum metal in the vessel is provided elsewhere. Fresh and cured properties were measured for: (1) commercially blended magnesium mono potassium phosphate packaged grouts, (2) commercially available binders blended with inert fillers at SRNL, (3) grouts prepared from technical grade MgO and KH{sub 2}PO{sub 4} and inert fillers (quartz sands, Class F fly ash), and (4) Ceramicrete{reg_sign} magnesium mono potassium phosphate-based grouts prepared at Argonne National Laboratory. Boric acid was evaluated as a set retarder in the magnesium mono potassium phosphate mixes.

Langton, C.; Stefanko, D.

2011-01-05T23:59:59.000Z

249

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

SciTech Connect (OSTI)

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抯 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 搒caling down the operating conditions at full scale to define test conditions at reduced scale and 搒caling 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.

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

250

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

E-Print Network [OSTI]

into a number of product fractions in a single batch column whereas, in continuous distillation several columnsIntegrated 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

Skogestad, Sigurd

251

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

SciTech Connect (OSTI)

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.

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

252

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

E-Print Network [OSTI]

conditions, increasing the infrastructure flexibility and saving energy #12;DOE H2 Program Annual Review, 5 the infrastructure flexibility and saving energy Liquid hydrogen compressed hydrogen #12;DOE H2 Program Annual ReviewDOE H2 Program Annual Review, 5-20-2003 Insulated Pressure Vessels for Vehicular Hydrogen Storage

253

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

E-Print Network [OSTI]

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

Barr, Al

254

A Novel Methodology for Robust Dynamic Positioning of Marine Vessels: Theory and Experiments*  

E-Print Network [OSTI]

, in the offshore, oil, and gas industries many applications are only possible with the use of DP systems operated ship, the Cybership III, under different sea conditions in a towing tank equipped with a hydraulic of the Shell Oil Company. Nowadays, DP systems are used with a wide range of vessel types and in different

Instituto de Sistemas e Robotica

255

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

SciTech Connect (OSTI)

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.

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

1999-01-01T23:59:59.000Z

256

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

SciTech Connect (OSTI)

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.

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

2010-04-13T23:59:59.000Z

257

Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR plants  

SciTech Connect (OSTI)

Recent data from the HFIR vessel surveillance program indicate a substantial radiation embrittlement rate effect at low irradiation temperatures (/approximately/120/degree/F) for A212-B, A350-LF3, A105-II, and corresponding welds. PWR vessel supports are fabricated of similar materials and are subjected to the same low temperatures and fast neutron fluxes (10/sup 8/ to 10/sup 9/ neutrons/cm/sup 2//center dot/s, E > 1.0 MeV) as those in the HFIR vessel. Thus, the embrittlement rate of these structures may be greater than previously anticipated. A study sponsored by the NRC is under way at ORNL to determine the impact of the rate effect on PWR vessel-support life expectancy. The scope includes the interpretation and application of the HFIR data, a survey of all light-water-reactor vessel support designs, and a structural and fracture-mechanics analysis of the supports for two specific PWR plants of particular interest with regard to a potential for support failure as a result of propagation of flaws. Calculations performed thus far indicate best-estimate critical flaw sizes, corresponding to 32 EFPY, of /approximately/0.2 in. for one plant and /approximately/0.4 in. for the other. These flaw sizes are small enough to be of concern. However, it appears that low-cycle fatigue is not a viable mechanism for creation of flaws of this size, and thus, presumably, such flaws would have to exist at the time of fabrication. 59 refs., 128 figs., 49 tabs.

Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

1989-01-01T23:59:59.000Z

258

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)

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.

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

2005-01-01T23:59:59.000Z

259

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

DOE Patents [OSTI]

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.

Hunsbedt, A.; Boardman, C.E.

1995-04-11T23:59:59.000Z

260

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

DOE Patents [OSTI]

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.

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

1995-01-01T23:59:59.000Z

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


261

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

SciTech Connect (OSTI)

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.

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

1984-01-01T23:59:59.000Z

262

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

SciTech Connect (OSTI)

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.

Jordan, W.C.

1993-02-01T23:59:59.000Z

263

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

SciTech Connect (OSTI)

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.

Jordan, W.C.

1993-02-01T23:59:59.000Z

264

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

SciTech Connect (OSTI)

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.

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

2013-08-08T23:59:59.000Z

265

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

SciTech Connect (OSTI)

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)

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

266

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

E-Print Network [OSTI]

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.

J. C. Misra; S. Maiti

2012-01-30T23:59:59.000Z

267

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

SciTech Connect (OSTI)

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.

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

268

Design Analysis and Manufacturing Studies for ITER In-Vessel Coils  

SciTech Connect (OSTI)

ITER is incorporating two types of In Vessel Coils (IVCs): ELM Coils to mitigate Edge Localized Modes and VS Coils to provide Vertical Stabilization of the plasma. Strong coupling with the plasma is required so that the ELM and VS Coils can meet their performance requirements. Accordingly, the IVCs are in close proximity to the plasma, mounted just behind the Blanket Shield Modules. This location results in a radiation and temperature environment that is severe necessitating new solutions for material selection as well as challenging analysis and design solutions. Fitting the coil systems in between the blanket shield modules and the vacuum vessel leads to difficult integration with diagnostic cabling and cooling water manifolds.

Kalish, M.; Heitzenroeder, P.; Neumeyer, C.; Titus, P.; Zhai, Y.; Zatz, I.; Messineo, M.; Gomez, M.; Hause, C. [Princeton Plasma Physics Lab., Princeton, NJ (United States); Daly, E.; Martin, A. [ITER Organization, St. Paul-lez-Duranace (France); Wu, Y.; Jin, J.; Long, F.; Song, Y.; Wang, Z.; Yun, Zan [Chinese Academy of Sciences, Anhui (China). Institute of Plasma Physics; Hsiao, J. [Vector Resources, Annandale, VA (United States); Pillsbury, J. R. [Sherbrooke Consulting, Arlington, VA (United States); Bohm, T.; Sawan, M. [Univ. of Wisconsin-Madison, WI (United States). Fustion Technology Institute; Jiang, NFN [Southwestern Institute of Physics, Chengdu (China)

2014-07-01T23:59:59.000Z

269

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

SciTech Connect (OSTI)

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

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

2012-09-27T23:59:59.000Z

270

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

E-Print Network [OSTI]

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.

American Society for Testing and Materials. Philadelphia

2010-01-01T23:59:59.000Z

271

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

SciTech Connect (OSTI)

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.

Pratt, Joseph William; Harris, Aaron P

2013-01-01T23:59:59.000Z

272

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

SciTech Connect (OSTI)

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.

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

2010-11-10T23:59:59.000Z

273

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

SciTech Connect (OSTI)

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.

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

2011-01-01T23:59:59.000Z

274

Analysis of the ANL Test Method for 6CVS Containment Vessels  

SciTech Connect (OSTI)

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.

Trapp, D.; Crow, G.

2011-06-06T23:59:59.000Z

275

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

DOE Patents [OSTI]

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.

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

2013-01-15T23:59:59.000Z

276

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

SciTech Connect (OSTI)

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.

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

1993-12-31T23:59:59.000Z

277

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

SciTech Connect (OSTI)

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.

GJ Schuster, FA Simonen, SR Doctor

2008-04-01T23:59:59.000Z

278

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

SciTech Connect (OSTI)

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.

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

2009-01-01T23:59:59.000Z

279

DESIGN OF A CONTAINMENT VESSEL CLOSURE FOR SHIPMENT OF TRITIUM GAS  

SciTech Connect (OSTI)

This paper presents a design summary of the containment vessel closure for the Bulk Tritium Shipping Package (BTSP). This new 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 new design is based on changes in the regulatory requirements. The BTSP design incorporates many improvements over its predecessor by implementing improved testing, handling, and maintenance capabilities, while improving manufacturability and incorporating new engineered materials that enhance the package's ability to withstand dynamic loading and thermal effects. This paper will specifically summarize the design philosophy and engineered features of the BTSP containment vessel closure. The closure design incorporates a concave closure lid, metallic C-Ring seals for containing tritium gas, a metal bellows valve and an elastomer O-Ring for leak testing. The efficient design minimizes the overall vessel height and protects the valve housing from damage during postulated drop and crush scenarios. Design features will be discussed.

Eberl, K; Paul Blanton, P

2007-07-03T23:59:59.000Z

280

Refractory lining system for high wear area of high temperature reaction vessel  

DOE Patents [OSTI]

A refractory-lined high temperature reaction vessel comprises a refractory ring lining constructed of refractory brick, a cooler, and a heat transfer medium disposed between the refractory ring lining and the cooler. The refractory brick comprises magnesia (MgO) and graphite. The heat transfer medium contacts the refractory brick and a cooling surface of the cooler, and is composed of a material that accommodates relative movement between the refractory brick and the cooler. The brick is manufactured such that the graphite has an orientation providing a high thermal conductivity in the lengthwise direction through the brick that is higher than the thermal conductivity in directions perpendicular to the lengthwise direction. The graphite preferably is flake graphite, in the range of about 10 to 20 wt %, and has a size distribution selected to provide maximum brick density. The reaction vessel may be used for performing a reaction process including the steps of forming a layer of slag on a melt in the vessel, the slag having a softening point temperature range, and forming a protective frozen layer of slag on the interior-facing surface of the refractory lining in at least a portion of a zone where the surface contacts the layer of slag, the protective frozen layer being maintained at or about the softening point of the slag. 10 figs.

Hubble, D.H.; Ulrich, K.H.

1998-04-21T23:59:59.000Z

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


281

Refractory lining system for high wear area of high temperature reaction vessel  

DOE Patents [OSTI]

A refractory-lined high temperature reaction vessel comprises a refractory ring lining constructed of refractory brick, a cooler, and a heat transfer medium disposed between the refractory ring lining and the cooler. The refractory brick comprises magnesia (MgO) and graphite. The heat transfer medium contacts the refractory brick and a cooling surface of the cooler, and is composed of a material that accommodates relative movement between the refractory brick and the cooler. The brick is manufactured such that the graphite has an orientation providing a high thermal conductivity in the lengthwise direction through the brick that is higher than the thermal conductivity in directions perpendicular to the lengthwise direction. The graphite preferably is flake graphite, in the range of about 10 to 20 wt %, and has a size distribution selected to provide maximum brick density. The reaction vessel may be used for performing a reaction process including the steps of forming a layer of slag on a melt in the vessel, the slag having a softening point temperature range, and forming a protective frozen layer of slag on the interior-facing surface of the refractory lining in at least a portion of a zone where the surface contacts the layer of slag, the protective frozen layer being maintained at or about the softening point of the slag. 10 figs.

Hubble, D.H.; Ulrich, K.H.

1998-09-22T23:59:59.000Z

282

Standard Practice for Determining NeutronExposures for Nuclear Reactor Vessel Support Structures  

E-Print Network [OSTI]

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...

American Society for Testing and Materials. Philadelphia

2008-01-01T23:59:59.000Z

283

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

E-Print Network [OSTI]

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, ...

Hannink, Ryan Christopher

2007-01-01T23:59:59.000Z

284

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

E-Print Network [OSTI]

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 ...

Campbell-Malone, Regina P

2007-01-01T23:59:59.000Z

285

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

E-Print Network [OSTI]

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...

Nutter, Darin Wayne

286

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

SciTech Connect (OSTI)

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.

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

1994-08-01T23:59:59.000Z

287

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

E-Print Network [OSTI]

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

Motta, Arthur T.

288

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

E-Print Network [OSTI]

fulfillment of the requirements for the degree of MASTER OF ARTS August 2000 Major Subject: Anthropology THE LOUISIANA STATE MUSEUM VESSEL: A HISTORICAL AND ARCHAEOLOGICAL ANALYSIS OF AN AMERICAN CIVIL WAR-ERA SUBMERSIBLE BOAT A Thesis by RICHARD KEITH... (Member) David l. . Csrlson (Head of Department) August 2000 Major Subject: Anthropology ABSTRACT The Louisiana State Museum Vessel: A Historical and Archaeological Analysis of an American Civil War-Era Submersible Boat. (August 2000) Richard Keith...

Wills, Richard Keith

2000-01-01T23:59:59.000Z

289

A Review of Proposed Upgrades to the High Flux Isotope Reactor and Potential Impacts to Reactor Vessel Integrity  

SciTech Connect (OSTI)

The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) was scheduled in October 2000 to implement design upgrades that include the enlargement of the HB-2 and HB-4 beam tubes. Higher dose rates and higher radiation embrittlement rates were predicted for the two beam-tube nozzles and surrounding vessel areas. ORNL had performed calculations for the upgraded design to show that vessel integrity would be maintained at acceptable levels. Pacific Northwest National Laboratory (PNNL) was requested by the U.S. Department of Energy Headquarters (DOE/HQ) to perform an independent peer review of the ORNL evaluations. PNNL concluded that the calculated probabilities of failure for the HFIR vessel during hydrostatic tests and for operational conditions as estimated by ORNL are an acceptable basis for selecting pressures and test intervals for hydrostatic tests and for justifying continued operation of the vessel. While there were some uncertainties in the embrittlement predictions, the ongoing efforts at ORNL to measure fluence levels at critical locations of the vessel wall and to test materials from surveillance capsules should be effective in dealing with embrittlement uncertainties. It was recommended that ORNL continue to update their fracture mechanics calculations to reflect methods and data from ongoing research for commercial nuclear power plants. Such programs should provide improved data for vessel fracture mechanics calculations.

Simonen, Fredric A.

2001-05-31T23:59:59.000Z

290

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

SciTech Connect (OSTI)

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.

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

2012-09-01T23:59:59.000Z

291

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

SciTech Connect (OSTI)

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.

Vinson, Dennis

2010-06-01T23:59:59.000Z

292

Spreading of molten corium in Mk I geometry following vessel meltthrough  

SciTech Connect (OSTI)

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.

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

1988-01-01T23:59:59.000Z

293

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

DOE Patents [OSTI]

A large (high flow rate) dye laser amplifier in which a continous 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.

Davin, James (Gilroy, CA)

1992-01-01T23:59:59.000Z

294

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

DOE Patents [OSTI]

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.

Davin, J.

1992-12-01T23:59:59.000Z

295

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

Energy Savers [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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of Inspector General Office0-72.pdfGeorgeDoesn't Happen to HighJosephNOx Traps forLM2 LNG Annual ReportVessel

296

Index of /research/alcator/documentation/6.10.09 vessel close up  

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: Vegetation ProposedUsingFun withconfinementEtching.348 270Energyradsafe/files [ICO] Name6.10.09 vessel

297

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

SciTech Connect (OSTI)

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.

Cadwallader, Lee Charles; Nygren, R. E.

2001-10-01T23:59:59.000Z

298

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

SciTech Connect (OSTI)

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.

Not Available

1984-11-01T23:59:59.000Z

299

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

SciTech Connect (OSTI)

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.

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

1988-01-01T23:59:59.000Z

300

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

SciTech Connect (OSTI)

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.

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

1983-01-01T23:59:59.000Z

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


301

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

SciTech Connect (OSTI)

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.

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

1984-01-01T23:59:59.000Z

302

Radiation damage characterization in reactor pressure vessel steels with nonlinear ultrasound  

SciTech Connect (OSTI)

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.

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

303

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

SciTech Connect (OSTI)

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)

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

2011-07-01T23:59:59.000Z

304

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

SciTech Connect (OSTI)

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.

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

1994-12-31T23:59:59.000Z

305

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

SciTech Connect (OSTI)

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.

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

306

Pressure vessels and piping codes and standards: Volume 2. PVP-Volume 339  

SciTech Connect (OSTI)

The role of Codes and Standards for pressure vessels and piping has increased significantly over the past decade. More and more, developments in Codes and Standards are accommodating the increasing sophistication of analysis methods, the need to address post-construction and operating plant issues, and the efficiencies that may be gained by focusing codes and standards on the areas that present the greatest risk. Codes and Standards for new construction also have had to accommodate greater challenges and more extreme environments imposed by more escalating requirements on piping and pressure vessel design and fabrication. This volume has focused on these challenges faced by Codes and Standards development. The topics in this volume include: (1) International Code Developments; (2) Seismic Developments in Codes and Standards; (3) Fabrication, Repairs, and Installation Issues Relating to Codes and Standards; (4) Application of Risk Based Criteria to In-Service Inspections; (5) Risk Based Codes and Standards; (6) The Code--Then and Now; (7) Reactor Water Fatigue: Fitness for Service; and (8) Two ASME Pressure Technology Code Issues: Post-Construction Codes and Metrication. Separate abstracts were prepared for most of the papers in this volume.

Esselman, T.C. [ed.] [Altran Corp., Boston, MA (United States); Balkey, K. [ed.] [Westinghouse Electric Corp., Pittsburgh, PA (United States); Chao, K.K.N. [ed.] [Consumers Power Co., Jackson, MI (United States); Gosselin, S. [ed.] [Electric Power Research Institute, Charlotte, NC (United States); Hollinger, G. [ed.] [Babcock and Wilcox, Barberton, OH (United States); Lubin, B.T. [ed.] [ABB Combustion Engineering, Windsor, CT (United States); Mohktarain, K. [ed.] [CB and I Technical Services, Plainfield, IL (United States); O`Donnell, W. [ed.] [O`Donnell Consulting Engineers, Inc., Pittsburgh, PA (United States); Rao, K.R. [ed.] [Entergy Operations, Inc, Jackson, MI (United States)

1996-12-01T23:59:59.000Z

307

Consequence evaluation of radiation embrittlement of Trojan reactor pressure vessel supports  

SciTech Connect (OSTI)

This report describes a consequence evaluation to address safety concerns raised by the radiation embrittlement of the reactor pressure vessel (RPV) supports for the Trojan nuclear power plant. The study comprises a structural evaluation and an effects evaluation and assumes that all four reactor vessel supports have completely lost the load carrying capability. By demonstrating that the ASME code requirements governing Level D service limits are satisfied, the structural evaluation concludes that the Trojan reactor coolant loop (RCL) piping is capable of transferring loads to the steam generator (SG) supports and the reactor coolant pump (RCP) supports. A subsequent design margins to accommodate additional loads transferred to them through the RCL piping. The effects evaluation, employing a systems analysis approach, investigates initiating events and the reliability of the engineered safeguard systems as the RPV is subject to movements caused by the RPV support failure. The evaluation identifies a number of areas of additional safety concerns, but further investigation of the above safety concerns, however, concludes that a hypothetical failure of the Trojan RPV supports due to radiation embrittlement will not result in consequences of significant safety concerns.

Lu, S.C.; Sommer, S.C.; Johnson, G.L. (Lawrence Livermore National Lab., CA (USA)); Lambert, H.E. (FTA Associates, Oakland, CA (USA))

1990-10-01T23:59:59.000Z

308

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

SciTech Connect (OSTI)

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.

Habil; Ziliukas, A.

1997-04-01T23:59:59.000Z

309

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

SciTech Connect (OSTI)

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.

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

310

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

E-Print Network [OSTI]

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桽hear 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...

American Society for Testing and Materials. Philadelphia

2009-01-01T23:59:59.000Z

311

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

SciTech Connect (OSTI)

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抯 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.

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

2009-05-22T23:59:59.000Z

312

Effects of 50/degree/C surveillance and test reactor irradiations on ferritic pressure vessel steel embrittlement  

SciTech Connect (OSTI)

The results of surveillance tests on the High-Flux Isotope Reactor (HFIR) pressure vessel at the Oak Ridge National Laboratory revealed that a greater than expected embrittlement had taken place after about 17.5 effective full-power years of operation and an operational assessment program was undertaken to fully evaluate the vessel condition and recommend conditions under which operation could be resumed. A research program was undertaken that included irradiating specimens in the Oak Ridge Research Reactor. Specimens of the A212 grade B vessel shell material were included, along with specimens from a nozzle qualification weld and a submerged-arc weld fabricated at ORNL to reproduce the vessel seam weld. The results of the surveillance program and the materials research program performed in support of the evaluation of the HFIR pressure vessel are presented and show the welds to be more radiation resistant than the A212B. Results of irradiated tensile and annealing experiments are described as well as a discussion of mechanisms which may be responsible for enhanced hardening at low damage rates. 20 refs., 22 figs., 5 tabs.

Nanstad, R.K.; Iskander, S.K.; Rowcliffe, A.F.; Corwin, W.R.; Odette, G.R.

1988-01-01T23:59:59.000Z

313

Impact of an apparent radiation embrittlement rate on the life expectancy of PWR (pressurized-water-reactor) vessel supports  

SciTech Connect (OSTI)

Recent data from the HFIR vessel surveillance program indicate a substantial radiation embrittlement rate effect at low irradiation temperatures (/approximately/120/degree/F) for A212-B, A350-LF3, A105-II and corresponding welds. PWR vessel supports are fabricated of similar materials and are subjected to the same low temperatures and fast neutron fluxes (10/sup 8/ /minus/ 10/sup 9/ n/cm/sup 2//center dot/s, E > 1.0 MeV) as those in the HFIR vessel. Thus, the embrittlement rate of these structures may be greater than previously anticipated. A study sponsored by the NRC is under way at ORNL to determine the impact of the rate effect on PWR vessel support life expectancy. The scope includes the interpretation and application of the HFIR data, a survey of all LWR vessel support designs, and a structural and fracture-mechanics analysis of the supports for two specific PWR plants of particular interest with regard to a potential for support failure as a result of propagation of flaws. Calculations performed for one of the plants indicate best-estimate critical flaw size corresponding to 32 EFPY, of /approximately/0.4 in. It appears that low-cycle fatigue is not a viable mechanism for creation of flaws of this size. Thus, presumably such flaws would have to exist at the time of fabrication. 19 refs., 8 figs., 3 tabs.

Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

1989-01-01T23:59:59.000Z

314

Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR (pressurized-water-reactor) plants  

SciTech Connect (OSTI)

Recent pressure-vessel surveillance data from the High Flux Isotope Reactor (HFIR) indicate an embrittlement fluence-rate effect that is applicable to the evaluation of the integrity of light-water reactor (LWR) pressure vessel supports. A preliminary evaluation using the HFIR data indicated increases in the nil ductility transition temperature at 32 effective full-power years (EFPY) of 100 to 130/degree/C for pressurized-water-reactor (PWR) vessel supports located in the cavity at midheight of the core. This result indicated a potential problem with regard to life expectancy. However, an accurate assessment required a detailed, specific-plant, fracture-mechanics analysis. After a survey and cursory evaluation of all LWR plants, two PWR plants that appeared to have a potential problem were selected. Results of the analyses indicate minimum critical flaw sizes small enough to be of concern before 32 EFPY. 24 refs., 16 figs., 7 tabs.

Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

1988-01-01T23:59:59.000Z

315

Electrical insulation system for the shell-vacuum vessel and poloidal field gap in the ZTH machine  

SciTech Connect (OSTI)

The electrical insulation systems for the ZTH machine have many unusual design problems. The poloidal field gap insulation must be capable of conforming to poloidal and toroidal contours, provide a 25 kV hold off, and sufficiently adhere to the epoxy back fill between the overlapping conductors. The shell-vacuum vessel system will use stretchable and flexible insulation along with protective hats, boots and sleeves. The shell-vacuum vessel system must be able to withstand a 12.5 kV pulse with provision for thermal insulation to limit the effects of the 300{degrees}C vacuum vessel during operation and bakeout. Methodology required to provide the electrical protection along with testing data and material characteristics will be presented. 7 figs.

Reass, W.A.; Ballard, E.O.

1989-01-01T23:59:59.000Z

316

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

SciTech Connect (OSTI)

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.

Raske, D.T.

1995-06-01T23:59:59.000Z

317

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

SciTech Connect (OSTI)

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.

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

318

Assemblies and methods for mitigating effects of reactor pressure vessel expansion  

DOE Patents [OSTI]

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.

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

319

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

SciTech Connect (OSTI)

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.

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

320

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

SciTech Connect (OSTI)

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.

Pritam Chakraborty; S. Bulent Biner

2014-08-01T23:59:59.000Z

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321

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

SciTech Connect (OSTI)

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).

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

322

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

SciTech Connect (OSTI)

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.

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

1993-09-01T23:59:59.000Z

323

Reactor pressure vessel integrity research at the Oak Ridge National Laboratory  

SciTech Connect (OSTI)

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.

Corwin, W.R.; Pennell, W.E.; Pace, J.V.

1995-12-31T23:59:59.000Z

324

REACTOR PRESSURE VESSEL ISSUES FOR THE LIGHT-WATER REACTOR SUSTAINABILITY PROGRAM  

SciTech Connect (OSTI)

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.

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

2010-01-01T23:59:59.000Z

325

THE DEVELOPMENT OF RADIATION EMBRITTLEMENT MODELS FOR U.S. POWER REACTOR PRESSURE VESSEL STEELS  

SciTech Connect (OSTI)

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.

Wang, Jy-An John [ORNL; Rao, Nageswara S [ORNL

2006-01-01T23:59:59.000Z

326

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

SciTech Connect (OSTI)

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.

Smith, A.

2010-02-16T23:59:59.000Z

327

Assemblies and methods for mitigating effects of reactor pressure vessel expansion  

DOE Patents [OSTI]

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.

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

1999-07-27T23:59:59.000Z

328

Measurements for the JASPER program In-Vessel Fuel Storage experiment  

SciTech Connect (OSTI)

The In-Vessel-Fuel-Storage (IVFS) experiment was conducted at the Oak Ridge National Laboratory`s (ORNL) Tower Shielding Facility (TSF) during the first nine months of 1991 as part of the continuing series of eight experiments planned for the Japanese-American Shielding Program for Experimental Research (JASPER) that was started in 1986. This is the fourth in a series of eight experiments that were planned, all of which are intended to provide support in the development of current reactor shield designs proposed for liquid metal reactor (LMR) systems both in Japan and the United States. The program is a cooperative effort between the United States Department of Energy (US DOE) and the Japanese Power Reactor and Nuclear Development Corporation (PNC). This document provides a description of the instrumentation and experimental configuration, test data, and data analysis.

Muckenthaler, F.J.; Spencer, R.R.; Hunter, H.T.; Hull, J.L. [Oak Ridge National Lab., TN (United States)] [Oak Ridge National Lab., TN (United States); Shono, A. [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)] [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)

1992-01-01T23:59:59.000Z

329

Advanced Models of LWR Pressure Vessel Embrittlement for Low Flux-HighFluence Conditions  

SciTech Connect (OSTI)

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.

Odette, G. Robert; Yamamoto, Takuya

2013-06-17T23:59:59.000Z

330

Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)  

SciTech Connect (OSTI)

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癈, 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.

J. K. Wright; R. N. Wright

2010-07-01T23:59:59.000Z

331

Effects of irradiation on strength and toughness of commercial LWR vessel cladding  

SciTech Connect (OSTI)

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.

Haggag, F.M.; Corwin, W.R.; Alexander, D.J.; Nanstad, R.K.

1987-01-01T23:59:59.000Z

332

AN IMPROVED TREATMENT OF RESIDUAL STRESSES IN FLAW ASSESSMENT OF PIPES AND PRESSURE VESSELS FABRICATED FROM FERRITIC STEELS  

E-Print Network [OSTI]

FABRICATED FROM FERRITIC STEELS William C. Mohr, Panagiotis Michaleris, and Mark T. Kirk Edison Welding ferritic steels. Information on these residual stresses are drawn from the literature; both measured treatment of residual stresses produced by welding in pipes and pressure vessels fabricated from ferritic

Michaleris, Panagiotis

333

Validation of vessel-based registration for correction of brain shift I. Reinertsen a,*, M. Descoteaux b,1  

E-Print Network [OSTI]

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

Siddiqi, Kaleem

334

In-Vessel Retention Technology Development and Use for Advanced PWR Designs in the USA and Korea  

SciTech Connect (OSTI)

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.

T.G. Theofanous; S.J. Oh; J.H. Scobel

2004-05-18T23:59:59.000Z

335

BLENDED CALCIUM ALUMINATE-CALCIUM SULFATE CEMENT-BASED GROUT FOR P-REACTOR VESSEL IN-SITU DECOMMISSIONING  

SciTech Connect (OSTI)

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].

Langton, C.; Stefanko, D.

2011-03-10T23:59:59.000Z

336

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

337

Vacuum Vessel Analysis and Design For The ARIES-ACT1 Fusion Power Plant H. H. Toudeshki, F. Najmabadi, X. R. Wang and the ARIES Team  

E-Print Network [OSTI]

Vacuum Vessel Analysis and Design For The ARIES-ACT1 Fusion Power Plant H. H. Toudeshki, F will lead us towards the ribbed structure model analysis. Ribbed structure, increases the strength on the port and door (the largest area of vacuum vessel) is to be determined which includes ribs thickness

338

Margin for In-Vessel Retention in the APR1400 - VESTA and SCDAP/RELAP5-3D Analyses  

SciTech Connect (OSTI)

If cooling is inadequate 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 lower head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with such plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe pressurized water reactor (PWR) (AP600), which relied upon external reactor vessel cooling (ERVC) for in-vessel retention (IVR), resulted in the U.S. Nuclear Regulatory Commission (USNRC) approving the design without requiring certain conventional features common to existing light water reactors (LWRs). IVR of core melt is therefore a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced LWRs. However, it is not clear that currently proposed ERVC without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a three-year, United States (U.S.) -Korean International Nuclear Energy Research Initiative (INERI) project was initiated in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korean Atomic Energy Research Institute (KAERI) explored options, such as enhanced ERVC performance and an enhanced in-vessel core catcher (IVCC), that have the potential to ensure that IVR is feasible for higher power reactors.

Joy Rempe; D. Knudson

2004-12-01T23:59:59.000Z

339

CLOSURE REPORT FOR CORRECTIVE ACTION UNIT 204: STORAGE BUNKERS, NEVADA TEST SITE, NEVADA  

SciTech Connect (OSTI)

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

NONE

2006-04-01T23:59:59.000Z

340

Automated registration of multispectral MR vessel wall images of the carotid artery  

SciTech Connect (OSTI)

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.

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

Note: This page contains sample records for the topic "railroad vessel bunkering" from the National Library of EnergyBeta (NLEBeta).
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341

Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)  

SciTech Connect (OSTI)

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癈 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, 2糃r-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.

J. K. Wright; R. N. Wright

2008-04-01T23:59:59.000Z

342

ILLINOIS RAILROAD ENGINEERING A Quantitative Decision Support  

E-Print Network [OSTI]

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

Barkan, Christopher P.L.

343

Field tests of timber railroad bridge piles  

E-Print Network [OSTI]

increasing the load of trains in return for larger profit, maintenance and replacement of timber bridges has risen throughout recent years. Once chosen for its low cost and ease of construction, timber bridges are being replaced by more efficient concrete...

Donovan, Kendra Ann

2005-02-17T23:59:59.000Z

344

Locomotion : a railroad museum for Chattanooga  

E-Print Network [OSTI]

This thesis is about exploring an architecture that serves a dual purpose: one, as witness to the past, and, two, as evidence of a constantly changing built environment. It is about exploring a landscape rich with associative ...

Kim, Julie Ju Youn

1994-01-01T23:59:59.000Z

345

Optimal Railroad Rail Grinding for Fatigue Mitigation  

E-Print Network [OSTI]

............................................................................. 25 2.4.1 FEA of Wheel-Rail Contact with an Elastic Material ...... 25 2.4.2 FEA of Wheel-Rail Contact with Plasticity Model .......... 27 2.5 Results and Discussion ................................................................. 28... 2.5.1 The Steady State of Residual Stresses .............................. 30 2.5.2 Accumulation of Residual Stresses in Rolling Contact .... 31 2.5.3 Comparisons between the Equivalent Rolling Stresses of Elastic and Hardening...

Tangtragulwong, Potchara

2012-02-14T23:59:59.000Z

346

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 on Google Bookmark EERE: Alternative Fuels Data Center Home5b9fcbce19 No revisionEnvReviewNonInvasiveExplorationUT-g Grant of Access Permit5-ID-aREC SolarRadium HotRail Splitter Wind

347

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 on Google Bookmark EERE: Alternative Fuels Data Center Home5b9fcbce19 No revisionEnvReviewNonInvasiveExplorationUT-g Grant of Access Permit5-ID-aREC SolarRadium HotRail Splitter

348

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power SystemsResources DOEElectricalon Clean DevelopmentCorporation -|Enhancedof Energy Safety

349

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power SystemsResourcesFLASH2011-11-OPAMFY 2007 Total System12Federal Long Tern

350

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines AboutDecemberSteam Coal Import96 4.87CBECS Public Use Data0 0 0270,975

351

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergyENERGY TAX POLICIES AND INTER-JURISDICTIONAL CHALLENGESRailroad Hazardous g Materials

352

Use of MCNP for characterization of reactor vessel internals waste from decommissioned nuclear reactors  

SciTech Connect (OSTI)

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.

Love, E.F.; Pauley, K.A.; Reid, B.D.

1995-09-01T23:59:59.000Z

353

Evolution of Nickel-Manganese-Silicon Dominated Phases in Highly Irradiated Reactor Pressure Vessel Steels  

SciTech Connect (OSTI)

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 2955癈 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.

Peter B Wells; Yuan Wu; Tim Milot; G. Robert Odette; Takuya Yamamoto; Brandon Miller; James Cole

2014-11-01T23:59:59.000Z

354

Application of micromechanical models of ductile fracture initiation to reactor pressure vessel materials  

SciTech Connect (OSTI)

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.

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

355

Modeling of irradiation embrittlement and annealing/recovery in pressure vessel steels  

SciTech Connect (OSTI)

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.

Lott, R.G.; Freyer, P.D. [Westinghouse Science and Technology Center, Pittsburgh, PA (United States)

1996-12-31T23:59:59.000Z

356

Effects of thermal annealing and reirradiation on toughness of reactor pressure vessel steels  

SciTech Connect (OSTI)

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.

Nanstad, R.K.; Iskander, S.K.; Sokolov, M.A. [and others

1996-12-31T23:59:59.000Z

357

Pressure vessel embrittlement predictions based on a composite model of copper precipitation and point defect clustering  

SciTech Connect (OSTI)

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.

Stoller, R.E. [Oak Ridge National Lab., TN (United States). Metals and Ceramics Div.

1996-12-31T23:59:59.000Z

358

Nondestructive characterization of embrittlement in reactor pressure vessel steels -- A feasibility study  

SciTech Connect (OSTI)

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.

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

359

Creep failure of a reactor pressure vessel lower head under severe accident conditions  

SciTech Connect (OSTI)

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.

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

360

Effects of thermal annealing and reirradiation on toughness of reactor pressure vessel steels  

SciTech Connect (OSTI)

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.

Nanstad, R.K.; Iskander, S.K.; Sokolov, M.A. [Oak Ridge National Lab., TN (United States)] [and others

1997-02-01T23:59:59.000Z

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361

A Modification of the Inner and Outer Core for Reactor Pressure Vessel Lifetime Extension  

SciTech Connect (OSTI)

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.

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

362

International Atomic Energy Agency (IAEA) Coordinated Research Projects on Structural Integrity of Reactor Pressure Vessels  

SciTech Connect (OSTI)

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.

Server, W. L. [ATI Consulting, Pinehurst, NC; Nanstad, Randy K [ORNL

2009-01-01T23:59:59.000Z

363

Lessons Learned Following the Successful Decommissioning of a Reaction Vessel Containing Lime Sludge and Technetium-99  

SciTech Connect (OSTI)

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.

Dawson, P. M.; Watson, D. D.; Hylko, J. M.

2002-02-25T23:59:59.000Z

364

On-Site Oxy-Lance Size Reduction of South Texas Project Reactor Vessel Heads - 12324  

SciTech Connect (OSTI)

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)

Posivak, Edward [WMG, inc. (United States); Keeney, Gilbert; Wheeler, Dean [Shaw Group (United States)

2012-07-01T23:59:59.000Z

365

BWR ex-vessel steam explosion analysis with MC3D code  

SciTech Connect (OSTI)

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)

Leskovar, M. [Josef Stefan Inst., Jamova cesta 39, 1001 Ljubljana (Slovenia)

2012-07-01T23:59:59.000Z

366

The toughness of irradiated pressure water reactor (PWR) vessel shell rings and the effect of segregation zones  

SciTech Connect (OSTI)

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.

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

367

FINAL REPORT - HYBRID-MIXING TESTS SUPPORTING THE CONCENTRATE RECEIPT VESSEL (CRV-VSL-00002A/2B) CONFIGURATION  

SciTech Connect (OSTI)

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.

GUERRERO, HECTORN.

2004-09-01T23:59:59.000Z

368

ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R- AND P-REACTOR VESSELS  

SciTech Connect (OSTI)

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.

Wiersma, B.

2009-12-29T23:59:59.000Z

369

Initial Evaluation of the Heat-Affected Zone, Local Embrittlement Phenomenon as it Applies to Nuclear Reactor Vessels  

SciTech Connect (OSTI)

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.

McCabe, D.E.

1999-09-01T23:59:59.000Z

370

THE IMPACT OF OZONE ON THE LOWER FLAMMABLE LIMIT OF HYDROGEN IN VESSELS CONTAINING SAVANNAH RIVER SITE HIGH LEVEL WASTE  

SciTech Connect (OSTI)

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.

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

371

Combination of Vessel-Targeting Agents and Fractionated Radiation Therapy: The Role of the SDF-1/CXCR4 Pathway  

SciTech Connect (OSTI)

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.

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

372

Prediction of failure behavior of a welded pressure vessel containing flaws during a hydrogen-charged burst test  

SciTech Connect (OSTI)

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.

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

373

Prediction of failure behavior of a welded pressure vessel containing flaws during a hydrogen-charged burst test  

SciTech Connect (OSTI)

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.

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

374

Investigation of source-detector separation optimization for an implantable perfusion and oxygenation sensor for liver blood vessels  

SciTech Connect (OSTI)

An implanted system is being developed to monitor transplanted liver health during the critical 7-10 day period posttransplantation. The unit will monitor organ perfusion and oxygen consumption using optically-based probes placed on both the inflow and outflow blood vessels, and on the liver parenchymal surface. Sensing probes are based on a 3- wavelength LED source and a photodiode detector. Sample diffuse reflectance is measured at 735, 805, and 940 nm. To ascertain optimal source-to-photodetector spacing for perfusion measurement in blood vessels, an ex vivo study was conducted. In this work, a dye mixture simulating 80% blood oxygen saturation was developed and perfused through excised porcine arteries while collecting data for various preset probe source-to-photodetector spacings. The results from this study demonstrate a decrease in the optical signal with decreasing LED drive current and a reduction in perfusion index signal with increasing probe spacing. They also reveal a 2- to 4-mm optimal range for blood vessel perfusion probe source-to-photodetector spacing that allows for sufficient perfusion signal modulation depth with maximized signal to noise ratio (SNR). These findings are currently being applied to guide electronic configuration and probe placement for in vivo liver perfusion porcine model studies.

Baba, Justin S [ORNL; Akl, Tony [Texas A& M University; Cote, Gerard L. [Texas A& M University; Wilson, Mark A. [University of Pittsburgh School of Medicine, Pittsburgh PA; Ericson, Milton Nance [ORNL

2011-01-01T23:59:59.000Z

375

Response of Soviet-designed VVER-440 steam generator vessel to pressurization  

SciTech Connect (OSTI)

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.

Kennedy, J.M.; Sienicki, J.J.

1989-01-01T23:59:59.000Z

376

The Information Fusion Embrittlement Models for U.S. Power Reactor Pressure Vessel Steels  

SciTech Connect (OSTI)

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.

Wang, Jy-An John [ORNL; Rao, Nageswara S [ORNL; Konduri, Savanthi [AOL

2007-01-01T23:59:59.000Z

377

An investigation of temperature measurement methods in nuclear power plant reactor pressure vessel annealing  

SciTech Connect (OSTI)

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.

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

378

SPR salt wall leaching experiments in lab-scale vessel : data report.  

SciTech Connect (OSTI)

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.

Webb, Stephen Walter; O'Hern, Timothy John; Hartenberger, Joel David

2010-10-01T23:59:59.000Z

379

EXPERIMENTAL STUDY OF CRITICAL HEAT FLUX WITH ALUMINA-WATER NANOFLUIDS IN DOWNWARD-FACING CHANNELS FOR IN-VESSEL RETENTION APPLICATIONS  

E-Print Network [OSTI]

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 ...

Park, R.J.

380

Three-dimensional discrete ordinates radiation transport calculations of neutron fluxes for beginning-of-cycle at several pressure vessel surveillance positions in the high flux isotope reactor  

SciTech Connect (OSTI)

The objective of this research was to determine improved thermal, epithermal, and fast fluxes and several responses at mechanical test surveillance location keys 2, 4, 5, and 7 of the pressure vessel of the Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) for the beginning of the fuel cycle. The purpose of the research was to provide essential flux data in support of radiation embrittlement studies of the pressure vessel shell and beam tubes at some of the important locations.

Pace, J.V. III; Slater, C.O.; Smith, M.S.

1993-11-01T23:59:59.000Z

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


381

W?a?t?e?r?s?c?h?i?p? NZ42i: a late medieval fishing vessel from Flevoland, The Netherlands  

E-Print Network [OSTI]

WATERSCHIP NZ42i: A LATE MEDIEVAL FISHING VESSEL FROM FLEVOLAND. THE NETHERLANDS A Thesis RALPH KENNETH PEDERSEN Submitted to the Off ice of Graduate Stucf les of Texas ASM University in partial fulfillment of the requirements for the degree... of MASTER OF ARTS May 1991 Major Subject: Anthropology WATERSCHIP NZ42(: A LATE MEDIEVAL FISHING VESSEL FROM FLEVOLAND, THE NETHERLANDS A Thesis by RALPH KENNETH PEDERSEN Approved as to style and content by: F. H. van Doorninck, Jr. (Chair...

Pedersen, Ralph Kenneth

1991-01-01T23:59:59.000Z

382

Coxsackie- and adenovirus receptor (CAR) is expressed in lymphatic vessels in human skin and affects lymphatic endothelial cell function in vitro  

SciTech Connect (OSTI)

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.

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

383

Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.  

SciTech Connect (OSTI)

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.

Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

2007-03-21T23:59:59.000Z

384

The DOS 1 neutron dosimetry experiment at the HB-4-A key 7 surveillance site on the HFIR pressure vessel  

SciTech Connect (OSTI)

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.

Farrell, K.; Kam, F.B.; Baldwin, C.A. [and others

1994-01-01T23:59:59.000Z

385

Prediction and Monitoring Systems of Creep-Fracture Behavior of 9Cr-1Mo Steels for Teactor Pressure Vessels  

SciTech Connect (OSTI)

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.

Potirniche, Gabriel; Barlow, Fred D.; Charit, Indrajit; Rink, Karl

2013-11-26T23:59:59.000Z

386

Microstructure and mechanical properties of WWER-440 reactor vessel metal after service life expiration and recovery anneal  

SciTech Connect (OSTI)

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.

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

387

Surveillance program for WWER-440/Type 213 reactor pressure vessels -- Standard program, re-evaluation of results, supplementary program  

SciTech Connect (OSTI)

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.

Brumovsky, M.; Novosad, P.; Zdarek, J. [Nuclear Research Inst. Rez plc (Czech Republic)

1996-12-31T23:59:59.000Z

388

Modeling the Ductile Brittle Fracture Transition in Reactor Pressure Vessel Steels using a Cohesive Zone Model based approach  

SciTech Connect (OSTI)

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.

Pritam Chakraborty; S. Bulent Biner

2013-10-01T23:59:59.000Z

389

Amtrak and its Host Railroads Paul Vilter, Amtrak AVP Host Railroads  

E-Print Network [OSTI]

Brunswick FDOT VTR NECR Pan Am MBTA Metro- North SCRRA Metra Operated by VIA Rail SCRRA / BNSF / SDN Richmond Springfield Rutland Albany Pontiac Port Huron Niagara Falls St. Albans Brunswick MBTA (Boston

Bustamante, Fabi谩n E.

390

ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R AND P REACTOR VESSELS  

SciTech Connect (OSTI)

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.

Wiersma, B.

2009-10-29T23:59:59.000Z

391

Does Mechanical Thrombectomy in Acute Embolic Stroke Have Long-term Side Effects on Intracranial Vessels? An Angiographic Follow-up Study  

SciTech Connect (OSTI)

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.

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

392

BWRSAR (Boiling Water Reactor Severe Accident Response) calculations of reactor vessel debris pours for Peach Bottom short-term station blackout  

SciTech Connect (OSTI)

This paper describes recent analyses performed by the BWR Severe Accident Technology (BWRSAT) Program at Oak Ridge National Laboratory to estimate the release of debris from the reactor vessel for the unmitigated short-term station blackout accident sequence. Calculations were performed with the BWR Severe Accident Response (BWRSAR) code and are based upon consideration of the Peach Bottom Atomic Power Station. The modeling strategies employed within BWRSAR for debris relocation within the reactor vessel are briefly discussed and the calculated events of the accident sequence, including details of the calculated debris pours, are presented. 4 refs., 13 figs., 3 tabs.

Hodge, S.A.; Ott, L.J.

1988-01-01T23:59:59.000Z

393

Ex-vessel melt-coolant interactions in deep water pool: Studies and accident management for Swedish BWRs  

SciTech Connect (OSTI)

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.

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

394

Final Report - Gas Retention and Release Tests Supporting the Concentrate Receipt Vessel (CRV-VSL-00002A/2B) Configuration  

SciTech Connect (OSTI)

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.

GUERRERO, HECTOR

2004-09-01T23:59:59.000Z

395

CONTAINMENT VESSEL TEMPERATURE FOR PU-238 HEAT SOURCE CONTAINER UNDER AMBIENT, FREE CONVECTION AND LOW EMISSIVITY COOLING CONDITIONS  

SciTech Connect (OSTI)

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.

Gupta, N.; Smith, A.

2011-02-14T23:59:59.000Z

396

Heavy-Section Steel Irradiation Program on irradiation effects in light-water reactor pressure vessel materials  

SciTech Connect (OSTI)

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.

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

397

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]

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

Talke, Stefan

398

Predicting target vessel location on robot-assisted coronary artery bypass graft using CT to ultrasound registration  

SciTech Connect (OSTI)

Purpose: Although robot-assisted coronary artery bypass grafting (RA-CABG) has gained more acceptance worldwide, its success still depends on the surgeon's experience and expertise, and the conversion rate to full sternotomy is in the order of 15%-25%. One of the reasons for conversion is poor pre-operative planning, which is based solely on pre-operative computed tomography (CT) images. In this paper, the authors propose a technique to estimate the global peri-operative displacement of the heart and to predict the intra-operative target vessel location, validated via both an in vitro and a clinical study. Methods: As the peri-operative heart migration during RA-CABG has never been reported in the literatures, a simple in vitro validation study was conducted using a heart phantom. To mimic the clinical workflow, a pre-operative CT as well as peri-operative ultrasound images at three different stages in the procedure (Stage{sub 0}--following intubation; Stage{sub 1}--following lung deflation; and Stage{sub 2}--following thoracic insufflation) were acquired during the experiment. Following image acquisition, a rigid-body registration using iterative closest point algorithm with the robust estimator was employed to map the pre-operative stage to each of the peri-operative ones, to estimate the heart migration and predict the peri-operative target vessel location. Moreover, a clinical validation of this technique was conducted using offline patient data, where a Monte Carlo simulation was used to overcome the limitations arising due to the invisibility of the target vessel in the peri-operative ultrasound images. Results: For the in vitro study, the computed target registration error (TRE) at Stage{sub 0}, Stage{sub 1}, and Stage{sub 2} was 2.1, 3.3, and 2.6 mm, respectively. According to the offline clinical validation study, the maximum TRE at the left anterior descending (LAD) coronary artery was 4.1 mm at Stage{sub 0}, 5.1 mm at Stage{sub 1}, and 3.4 mm at Stage{sub 2}. Conclusions: The authors proposed a method to measure and validate peri-operative shifts of the heart during RA-CABG. In vitro and clinical validation studies were conducted and yielded a TRE in the order of 5 mm for all cases. As the desired clinical accuracy imposed by this procedure is on the order of one intercostal space (10-15 mm), our technique suits the clinical requirements. The authors therefore believe this technique has the potential to improve the pre-operative planning by updating peri-operative migration patterns of the heart and, consequently, will lead to reduced conversion to conventional open thoracic procedures.

Cho, Daniel S.; Linte, Cristian; Chen, Elvis C. S.; Bainbridge, Daniel; Wedlake, Chris; Moore, John; Barron, John; Patel, Rajni; Peters, Terry [Imaging Research Laboratories, Robarts Research Institute and Biomedical Engineering Graduate Program, University of Western Ontario, Ontario N6A 5K8 (Canada); Imaging Research Laboratories, Robarts Research Institute and Biomedical Imaging Resource, Mayo Clinic, Rochester, Minnesota 55905 (United States); Imaging Research Laboratories, Robarts Research Institute, University of Western Ontario, Ontario N6A 5K8 (Canada); Canadian Surgical Technologies and Advanced Robotics, London, Ontario N6A 5A5 (Canada); Imaging Research Laboratories, Robarts Research Institute, University of Western Ontario, Ontario N6A 5K8 (Canada); Department of Computer Science, University of Western Ontario, Ontario N6A 5B7 (Canada); Canadian Surgical Technologies and Advanced Robotics, London, Ontario N6A 5A5 (Canada); Imaging Research Laboratories, Robarts Research Institute, Biomedical Engineering Graduate Program, University of Western Ontario N6A 5K8 (Canada); and Canadian Surgical Technologies and Advanced Robotics, London, Ontario N6A 5A5 (Canada)

2012-03-15T23:59:59.000Z

399

Debris dispersal in reactor material experiments on corium-water thermal interactions in ex-vessel geometry  

SciTech Connect (OSTI)

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.

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

1984-01-01T23:59:59.000Z

400

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program  

SciTech Connect (OSTI)

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.

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

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


401

Safety Evaluation Report: Development of Improved Composite Pressure Vessels for Hydrogen Storage, Lincoln Composites, Lincoln, NE, May 25, 2010  

SciTech Connect (OSTI)

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.

Fort, III, William C.; Kallman, Richard A.; Maes, Miguel; Skolnik, Edward G.; Weiner, Steven C.

2010-12-22T23:59:59.000Z

402

Analysis of dosimetry from the H.B. Robinson unit 2 pressure vessel benchmark using RAPTOR-M3G and ALPAN  

SciTech Connect (OSTI)

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)

Fischer, G.A. [Westinghouse Electric Company, LLC, 1000 Westinghouse Dr., Cranberry Township, PA 16066 (United States)

2011-07-01T23:59:59.000Z

403

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)

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.

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

404

Specific corrosion product on interior surface of a bronze wine vessel with loop-handle and its growth mechanism, Shang Dynasty, China  

SciTech Connect (OSTI)

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.

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

405

Optical Measurement Technologies for High Temperature, Radiation Exposure, and Corrosive Environments桽ignificant Activities and Findings: In-vessel Optical Measurements for Advanced SMRs  

SciTech Connect (OSTI)

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

Anheier, Norman C.; Cannon, Bret D.; Qiao, Hong (Amy) [Amy; Suter, Jonathan D.

2012-09-01T23:59:59.000Z

406

Composition and chemistry of particulates from the Tidd Clean Coal Demonstration Plant pressurized fluidized bed combustor, cyclone, and filter vessel  

SciTech Connect (OSTI)

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}.

Smith, D.H.; Grimm, U.; Haddad, G.

1995-12-31T23:59:59.000Z

407

Calculating Vessel Volumes  

E-Print Network [OSTI]

This summarizes a demonstration at the "Metron" conference (April 2002, Yale University), using a computer program to calculate the capacity of pots drawn on the computer screen and then given one dimension (e.g., its ...

Younger, John G.

2003-01-01T23:59:59.000Z

408

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

409

1-Dimensional simulation of thermal annealing in a commercial nuclear power plant reactor pressure vessel wall section  

SciTech Connect (OSTI)

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.

Nakos, J.T.; Rosinski, S.T.; Acton, R.U.

1994-11-01T23:59:59.000Z

410

Application of ex-vessel neutron dosimetry combined with in-core measurements for correction of neutron source used for RPV fluence calculations  

SciTech Connect (OSTI)

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)

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

411

Fracture Analysis of Vessels Oak Ridge FAVOR, v06.1, Computer Code: Theory and Implementation of Algorithms, Methods, and Correlations  

SciTech Connect (OSTI)

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.

Williams, P. T. [ORNL; Dickson, T. L. [ORNL; Yin, S. [ORNL

2007-12-01T23:59:59.000Z

412

Suggestion of typical phases of in-vessel fuel-debris by thermodynamic calculation for decommissioning technology of Fukushima-Daiichi nuclear power station  

SciTech Connect (OSTI)

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)

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

413

ILLINOIS -RAILROAD ENGINEERING Understanding the Transition Behavior of  

E-Print Network [OSTI]

at Level Crossings on US High Speed Rail Shared Corridors Francesco Bedini Jacobini PhD Student Rail

Illinois at Urbana-Champaign, University of

414

Application of CAST and STPA to railroad safety in China  

E-Print Network [OSTI]

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 ...

Dong, Airong, S.M. Massachusetts Institute of Technology

2012-01-01T23:59:59.000Z

415

Remote Sensing of Railroad Locomotive Emissions: A Feasibility Study  

E-Print Network [OSTI]

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

Denver, University of

416

North American Short Line and Regional Railroads Industry Report  

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

45G Short Line Tax Credit 45G Short Line Tax Credit government owned short lines government owned short lines State grants State grants D.O.E. investment D.O.E. investment...

417

Optimal Location of Railroad Wayside Defect Detection Installations  

E-Print Network [OSTI]

Background, motivation and objective #12;5 5 Background New wayside detector technologies are used to monitor-gauge sensors on rails in reverse curves 路 Increase safety, reduce fuel costs, reduce damage to track Components Acoustic (e.g. TADSTM) Acoustic Bearing Detection Leaking Air Sensors Ultrasonic Cracked Wheel

Barkan, Christopher P.L.

418

Paper No. 01-0238 Railroads in Greece  

E-Print Network [OSTI]

and motor carriers, gas price, inflation and GNP. These were used to estimate time-series models. Forecasts operations is not hopeful. Additional emphasis may be placed on lines with historical and cultural and costs. #12;3 1. Introduction and Historical Summary This paper investigates information on the past

Prevedouros, Panos D.

419

An analysis of railroad accident dynamics in College Station, Texas  

E-Print Network [OSTI]

anniversary of the Waco standoff between Federal agents and members of the religious sect headed by David Koresh and one year after the Oklahoma bombing, bomb threats by an anonymous caller claiming to represent an international terrorist group shut down...-PROTECTION For a terrorist rail incident at Texas ARM University, considerations for the protection of responders and bystanders include possible exposure to asphixiative agents, mechanical hazards, thermal hazards, and psychological effects. The asphixiative...

Brown, Hart Stefan

1998-01-01T23:59:59.000Z

420

U.S. Railroad Safety Statistics and Trends  

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

Track Rail steels Welded rail Fasteners Detection of flaws, weak spots Improved Equipment Heat treated curved plate wheels Hot box detectors, roller bearings, acoustic detection...

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


421

Effect of Railroad Deregulation on Export - Grain Transportation Rate Structures.  

E-Print Network [OSTI]

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...

Fuller, Stephen; MacDonald, James; Bessler, David; Wohlgenant, Michael

1988-01-01T23:59:59.000Z

422

Permeability, Resistivity and Strength of Fouled Railroad Ballast  

E-Print Network [OSTI]

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...

Rahman, A. J.

2013-08-31T23:59:59.000Z

423

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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of Inspector GeneralDepartmentAUDIT REPORTOpenWende New EnergyAnatolia Jump to: navigation,

424

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power SystemsResources DOEElectricalon Clean DevelopmentCorporation -|Enhanced

425

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 on Google Bookmark EERE: Alternative Fuels Data Center Home5b9fcbce19 NoPublic Utilities Address: 160 East 300 SouthWaterBrasil JumpAerowattOpen2008District No 3formPermits | Open Energy

426

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the Contributions andData andFleet TestAccounts andThe Role of Mesoscale Eddies9085023

427

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)

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.

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

428

In-vessel thermohydraulics evaluation of an unprotected transient overpower accident and delayed neutron precursor concentration transport analysis using a multidimensional code  

SciTech Connect (OSTI)

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.

Muramatsu, T.; Ninokata, H. (Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan))

1992-02-01T23:59:59.000Z

429

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)

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.

McAfee, W.J.; Bass, B.R.; Bryson, J.W.

1998-07-01T23:59:59.000Z

430

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)

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.

Kim, Jungo-Do; Gil, Choong-Sup [Korea Atomic Energy Research Institute, Taejon (Korea, Democratic People`s Republic of)

1994-12-31T23:59:59.000Z

431

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)

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.

N /A

2004-02-12T23:59:59.000Z

432

For Halibut Processor or Vessel Name Vessels Delivering to Processor  

E-Print Network [OSTI]

Explorer 3011 Ocean Harvester 5130 Pacific Explorer 3010 Pacific Ram 4305 Pacific Viking 422 Pegasus 1265 Moon Bay 249 Miss Berdie 3679 Nordic Fury 1094 Ocean Hope 3 652397 Pacific Fury 421 Poseidon 1164 Royal Caravelle 3402 Dusk 4 Sea Mac 1043 Topaz 405 Ocean Beauty Seafoods, Kodiak, Alaska 30883 Pacific Star 2781

433

Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.  

SciTech Connect (OSTI)

This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition from normal high-flow operation to natural circulation. Low-flow coolant events are the most difficult to design for because they involve the most complex thermal-hydraulic behavior induced by the dominance of thermal-buoyancy forces acting on the coolants. Such behavior can cause multiple-component flow interaction phenomena, which are not adequately understood or appreciated by reactor designers as to their impact on reactor performance and safety. Since the early 1990s, when DOE canceled the U.S. Liquid Metal Fast Breeder Reactor (LMFBR) program, little has been done experimentally to further understand the importance of the complex thermal-buoyancy phenomena and their impact on reactor design or to improve the ability of three-dimensional (3-D) transient computational fluid dynamics (CFD) and structures codes to model the phenomena. An improved experimental data base and the associated improved validated codes would provide needed design tools to the reactor community. The improved codes would also facilitate scale-up from small-scale testing to prototype size and would facilitate comparing performance of one reactor/component design with another. The codes would also have relevance to the design and safety of water-cooled reactors. To accomplish the preceding, it is proposed to establish a national GNEP-LMR research and development center at Argonne having as its foundation state-of-art science-based infrastructure consisting of: (a) thermal-hydraulic experimental capabilities for conducting both water and sodium testing of individual reactor components and complete reactor in-vessel models and (b) a computational modeling development and validation capability that is strongly interfaced with the experimental facilities. The proposed center would greatly advance capabilities for reactor development by establishing the validity of high-fidelity (i.e., close to first principles) models and tools. Such tools could be used directly for reactor design or for qualifying/tuning of lower-fidelity models, which now require costly experimental qualification for each different type of design

Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

2007-06-30T23:59:59.000Z

434

State of Advancement of the International REVE Project: Computational Modelling of Irradiation-Induced Hardening in Reactor Pressure Vessel Steels and Relevant Experimental Validation Programme  

SciTech Connect (OSTI)

The REVE (Reactor for Virtual Experiments) project is an international joint effort aimed at developing multi-scale modelling computational toolboxes capable of simulating the behaviour of materials under irradiation at different time and length scales. Well grounded numerical techniques such as molecular dynamics (MD) and Monte Carlo (MC) algorithms, as well as rate equation (RE) and dislocation-defect interaction theory, form the basis on which the project is built. The goal is to put together a suite of integrated codes capable of deducing the changes in macroscopic properties starting from a detailed simulation of the microstructural changes produced by irradiation in materials. To achieve this objective, several European laboratories are closely collaborating, while exchanging data with American and Japanese laboratories currently pursuing similar approaches. The material chosen for the first phase of this project is reactor pressure vessel (RPV) steel, the target macroscopic magnitude to be predicted being the yield strength increase ({delta}{sigma}y) due, essentially, to irradiation-enhanced formation of intragranular solute atom precipitates or clouds, as well as irradiation induced defects in the matrix, such as point defect clusters and dislocation loops. A description of the methodological approach used in the project and its current state is given in the paper. The development of the simulation tools requires a continuous feedback from ad hoc experimental data. In the framework of the REVE project SCK EN has therefore performed a neutron irradiation campaign of model alloys of growing complexity (from pure Fe to binary and ternary systems and a real RPV steel) in the Belgian test reactor BR2 and is currently carrying on the subsequent materials characterisation using its hot cell facilities. The paper gives the details of this experimental programme - probably the first large-scale one devoted to the validation of numerical simulation tools - and presents and discusses the first available results, with a view to their use as feedback for the improvement of the computational modelling. (authors)

Malerba, Lorenzo; Van Walle, Eric [SCK.CEN, Boeretang 200, 2400 Mol (Belgium); Domain, Christophe; Jumel, Stephanie; Van Duysen, Jean-Claude [EDR R and D (France)

2002-07-01T23:59:59.000Z

435

COMMERCIAL FISHING VESSELS AND GEAR  

E-Print Network [OSTI]

traps for salmon and the extensive Great Lakes trap nets. Sail and steam have given way to gasoline included. General range of length, beam, draft, net tonnage, construction, engine, refrigeration, speed

436

The Borobudur Vessels in Context  

E-Print Network [OSTI]

and Indonesian Archipelago. Created by Douglas Inglis using a portion of the 1:10m Natural Earth II map and wind pattern data presented by Hall as well as Glover and Bellwood (Hall 1985, 22, Map 1; Bellwood and Glover 2004, 10, Fig. 1.4; Natural Earth 2014... Natural Earth II map and wind pattern data presented by Hall as well as Glover and Bellwood (Hall 1985, 22, Map 1; Bellwood and Glover 2004, 10, Fig. 1.4; Natural Earth 2014). 20 interaction between these turns of topography and the prevailing winds...

Inglis, Douglas Andrew

2014-07-28T23:59:59.000Z

437

Record of Technical Change 1 for Closure Report for Corrective Action Unit 204: Storage Bunkers, Nevada Test Site, Nevada, Revision 0  

SciTech Connect (OSTI)

Downgrade the FFACO UR at CAU 204, CAS 05-18-02, Chemical Explosives Storage to an Administrative UR.

Krauss, Mark J; Birney, Cathleen

2006-04-01T23:59:59.000Z

438

In-service Inspection Ultrasonic Testing of Reactor Pressure Vessel Welds for Assessing Flaw Density and Size Distribution per 10 CFR 50.61a, Alternate Fracture Toughness Requirements  

SciTech Connect (OSTI)

Pressurized thermal shock (PTS) events are system transients in a pressurized water reactor (PWR) in which there is a rapid operating temperature cool-down that results in cold vessel temperatures with or without repressurization of the vessel. The rapid cooling of the inside surface of the reactor pressure vessel (RPV) causes thermal stresses that can combine with stresses caused by high pressure. The aggregate effect of these stresses is an increase in the potential for fracture if a pre-existing flaw is present in a material susceptible to brittle failure. The ferritic, low alloy steel of the reactor vessel beltline adjacent to the core, where neutron radiation gradually embrittles the material over the lifetime of the plant, can be susceptible to brittle fracture. The PTS rule, described in the Code of Federal Regulations, Title 10, Section 50.61 (50.61), 揊racture Toughness Requirements for Protection against Pressurized Thermal Shock Events, adopted on July 23, 1985, establishes screening criteria to ensure that the potential for a reactor vessel to fail due to a PTS event is deemed to be acceptably low. The U.S. Nuclear Regulatory Commission (NRC) completed a research program that concluded that the risk of through-wall cracking due to a PTS event is much lower than previously estimated. The NRC subsequently developed a rule, 50.61a, published on January 4, 2010, entitled 揂lternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events (75 FR 13). Use of the new rule by licensees is optional. The 50.61a rule differs from 50.61 in that it requires licensees who choose to follow this alternate method to analyze the results from periodic volumetric examinations required by the ASME Code, Section XI, Rules for Inservice Inspection (ISI) of Nuclear Power Plants. These analyses are intended to determine if the actual flaw density and size distribution in the licensee抯 reactor vessel beltline welds are bounded by the flaw density and size distribution values used in the PTS technical basis. Under a contract with the NRC, Pacific Northwest National Laboratory (PNNL) has been working on a program to assess the ability of current inservice inspection (ISI)-ultrasonic testing (UT) techniques, as qualified through ASME Code, Appendix VIII, Supplements 4 and 6, to detect small fabrication or inservice-induced flaws located in RPV welds and adjacent base materials. As part of this effort, the investigators have pursued an evaluation, based on the available information, of the capability of UT to provide flaw density/distribution inputs for making RPV weld assessments in accordance with 50.61a. This paper presents the results of an evaluation of data from the 1993 Browns Ferry Nuclear Plant, Unit 3, Spirit of Appendix VIII reactor vessel examination, a comparison of the flaw density/distribution from this data with the distribution in 50.61a, possible reasons for differences, and plans and recommendations for further work in this area.

Sullivan, Edmund J.; Anderson, Michael T.; Norris, Wallace

2012-09-17T23:59:59.000Z

439

Evaluation on the Feasibility of Using Ultrasonic Testing of Reactor Pressure Vessel Welds for Assessing Flaw Density/Distribution per 10 CFR 50.61a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock  

SciTech Connect (OSTI)

This technical letter report provides the status of an assessment undertaken by PNNL at the request of the NRC to verify the capability of periodic ASME-required volumetric examinations of reactor vessels to characterize the density and distribution of flaws of interest for applying 50.61a on a plant-by-plant basis. The PTS rule, described in the Code of Federal Regulations, Title 10, Section 50.61 (50.61), "Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events," establishes screening criteria to ensure that the potential for a reactor vessel to fail due to a PTS event is deemed to be acceptably low. Recently, the NRC completed a research program that concluded that the risk of through-wall cracking due to a PTS event is much lower than previously estimated. The NRC subsequently developed and promulgated an alternate PTS rule, 50.61a, that can be implemented by PWR licensees. The 50.61a rule differs from 50.61 in that it requires licensees who choose to follow this alternate method to analyze the results from periodic volumetric examinations required by the ASME Code, Section XI, Rules for Inservice Inspection (ISI) of Nuclear Power Plants.

Sullivan, Edmund J.; Anderson, Michael T.

2014-06-10T23:59:59.000Z

440

Technical Review Report for the Mound 1KW Package Safety Analysis Report for Packaging Waiver for the Use of Modified Primary Containment Vessel (PCV)  

SciTech Connect (OSTI)

This Technical Review Report (TRR) documents the review, performed by the Lawrence Livermore National Laboratory (LLNL) staff, at the request of the U.S. Department of Energy (DOE), on the Waiver for the Use of Modified Primary Containment Vessels (PCV). The waiver is to be used to support a limited number of shipments of fuel for the Multi-Mission Radioisotope Thermoelectric Generator (MMRTG) Project in support of the National Aeronautics and Space Administration's (NASA's) Mars Science Laboratory (MSL) mission. Under the waiver, an inventory of existing national security PCVs will be converted to standard PCVs. Both types of PCVs are currently approved for use by the Office of Nuclear Energy. LLNL has previously reviewed the national security PCVs under Mound 1KW Package Safety Analysis Report for Packaging, Addendum No. 1, Revision c, dated June 2007 (Addendum 1). The safety analysis of the package is documented in the Safety Analysis Report for Packaging (SARP) for the Mound 1KW Package (i.e., the Mound 1KW SARP, or the SARP) where the standard PCVs have been reviewed by LLNL. The Mound 1KW Package is certified by DOE Certificate of Compliance (CoC) number USA/9516/B(U)F-85 for the transportation of Type B quantities of plutonium heat source material. The waiver requests an exemption, claiming safety equivalent to the requirements specified in 10 CFR 71.12, Specific Exemptions, and will lead to a letter amendment to the CoC. Under the waiver, the Office of Radioisotope Power Systems, NE-34, is seeking an exemption from 10 CFR 71.19(d)(1), Previously Approved Package,[5] which states: '(d) NRC will approve modifications to the design and authorized contents of a Type B package, or a fissile material package, previously approved by NRC, provided--(1) The modifications of a Type B package are not significant with respect to the design, operating characteristics, or safe performance of the containment system, when the package is subjected to the tests specified in {section}71.71 and 71.73.' The LLNL staff had previously reviewed a request from Idaho National Laboratory (INL) to reconfigure national security PCVs to standard PCVs. With a nominal 50% reduction in both the height and the volume, the LLNL staff initially deemed the modifications to be significant, which would not be allowed under the provisions of 10 CFR 71.19(d)(1)--see above. As a follow-up, the DOE requested additional clarification from the Nuclear Regulatory Commission (NRC). The NRC concluded that the reconfiguration would be a new fabrication, and that an exemption to the regulations would be required to allow its use, as per the requirements specified in 10 CFR 71.19(c)(1), Previously Approved Package: '(c) A Type B(U) package, a Type B(M) package, or a fissile material package previously approved by the NRC with the designation '-85' in the identification number of the NRC CoC, may be used under the general license of {section}71.17 with the following additional conditions: (1) Fabrication of the package must be satisfactorily completed by December 31, 2006, as demonstrated by application of its model number in accordance with 71.85(c).' Although the preferred approach toward the resolution of this issue would be for the applicant to submit an updated SARP, the applicant has stated that the process of updating the Model Mound 1KW Package SARP is a work that is in progress, but that the updated SARP is not yet ready for submittal. The applicant has to provide a submittal, proving that the package meets the '-96' requirements of International Atomic Energy Agency (IAEA) Safety Standards Series No. TS-R-1, in order to fabricate approved packagings after December 31, 2006. The applicant has further stated that all other packaging features, as described in the currently approved Model Mound 1KW Package SARP, remain unchanged. This report documents the LLNL review of the waiver request. The specific review for each SARP Chapter is documented.

West, M; Hafner, R

2008-05-05T23:59:59.000Z

Note: This page contains sample records for the topic "railroad vessel bunkering" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
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441

A Science Service Feature-Released upon receipt but intended f o r use  

E-Print Network [OSTI]

& Just as air-conditioning and artifictal cooling on railroads was at f i r s t applied only to dining. As deodorizing can easily be made an adjunct of air- conditioning, shipboard atmospheres in the future w i l l s t vessels to be equipped a i t h air-conditioning and air- cooling aoparatus is the S.S. Mariposa, which

442

An implicit centered finite-difference simulation for free surface flows in a rocking tank  

E-Print Network [OSTI]

include the liquid movement in closed containers such as tank trucks on highways and railroads, liquid fuel tanks in space vehicles' and contained liquid cargo in oceangoing vessels. Interest in this particular fluid phenomenon has grown consider...AN IMPLICIT CENTERED FINITE-DIFFERENCE SIMULATION FOR FREE SURFACE FLOWS IN A ROCKING TANK A Thesis by WILLIAM EDWARD JOBST Submitted to the Graduate College of Texas A8M University in partial fulfillment of the requirement for the degree...

Jobst, William Edward

1982-01-01T23:59:59.000Z

443

IN SITU CHEMICAL AND BIOLOGICAL TREATMENT OF TCE IN GROUNDWATER AT A LEGACY RAILROAD SITE  

E-Print Network [OSTI]

.0 1.5 2.0 2.5 3.0 -50 0 50 100 150 200 250 300 350 400 Ethene(mol/L) PCE,TCE,DCE,VC(mol/L) Days (Day 0 = July 18, 2013) MW-19 Molar Concentration PCE TCE Total DCE VC Total ethene #12;Results 0.00 0.20 0.40 0 300 350 400 Ethene(mol/L) PCE,TCE,DCE,VC(mol/L) Days (Day 0 = July 15, 2013) MW-26 Molar Concentration

Illinois at Urbana-Champaign, University of

444

Column generation heuristic for a rich arc routing Application to railroad track inspection routing  

E-Print Network [OSTI]

/LAAS) Optimising maintenance routing ATMOS 2010 7 / 24 #12;Literature review Industrial arc routing problems Hasle

Ingrand, Fran莽ois

445

iit magazine My father worked for the CB&Q Railroad all his life, so  

E-Print Network [OSTI]

: that of a small businessman who owned and operated a steam engine that he traveled around the countryside so he's engineers experience a wide variety of professional activities, but "drive a train" or "operate a steam knew what an engineer did: he drove the train. Many years later, as a college student, I read Thomas

Heller, Barbara

446

Field evaluation of a standard test method for screening fuels in soils at a railroad site  

SciTech Connect (OSTI)

American Society for Testing and Materials (ASTM) Method D-5831-95 is a standard test method for screening fuel contamination in soils. This method uses low-toxicity chemicals and can be used to screen organic-rich soils. It is also fast, easy, and inexpensive to perform. The screening method calls for extracting a sample of soil with isopropyl alcohol (IPA) following treatment with calcium oxide. The resulting extract is filtered, and the ultraviolet (UV) absorbance of the extract is measured at 254 nm. Depending on the information available concerning the contaminant fuel type and availability of the contaminant fuel for calibration, the method can be used to determine the approximate concentration of fuel contamination, an estimated value of fuel contamination, or an indication of the presence or absence of fuel contamination. Fuels containing aromatic compounds, such as diesel fuel and gasoline, as well as other aromatic-containing hydrocarbon materials, such as motor oil, crude oil, and coal oil can be determined. ASTM Method D-5831 was evaluated by using the method to screen soil samples at an actual field site. Soil contaminated with weathered and fresh diesel fuel was sampled and tested for its contaminant concentration. Soil samples were screened in the field using ASTM Method D-5831 and a portable soil test kit. In addition, splits of the soil samples were analyzed in the laboratory using an extractable petroleum hydrocarbon method. Field and laboratory data were compared and show good correlation between field screening and laboratory results.

Schabron, J.F.; Sorini, S.S. [Western Research Institute, Laramie, WY (United States); Butler, E.L. [Gradient Corp., Cambridge, MA (United States); Frisbie, S. [Johnson Co., Inc., Montpelier, VT (United States)

1997-12-31T23:59:59.000Z

447

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

E-Print Network [OSTI]

delay are given less consideration or ignored completely. Consequently, state-of-the-practice strategies may cause serious pedestrian safety and efficiency problems at IHRGCs. Therefore, there is a definite need for research on how to improve traffic...

Cho, Hanseon

2004-09-30T23:59:59.000Z

448

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

E-Print Network [OSTI]

awareness. Americans began to ask, "What if something similar happened here?" Chemicals with hazardous properties have become part of daily life. Industry, government, and the public have become aware of the need to respond to problems involving hazardous...

Reeder, Geoffrey Benton

1995-01-01T23:59:59.000Z

449

An investigation into the use of highway traffic signals at highway-railroad grade crossings  

E-Print Network [OSTI]

2 Highway Speed Limit Classification 3 Light Condition Classification 4 Presence of Drunk Driver Classification 5 Driver Age Group Classification . . 43 50 52 57 LIST OF TABLES TABLE Page 1 Fatal Accidents Based on FARS Traffic Control... Device 2 Traffic Control Comparison between FARS and DOT/AAR Databases. . . 3 Fatal Accident Rates Based on DOT/AAR Traffic Control Device . 4 Driver Related Factor Results 5 FARS Variables Used in Analysis . 6 Driver Related Factor Classification...

Frieslaar, Andre Henry

1997-01-01T23:59:59.000Z

450

The development of a life cycle cost model for railroad tunnels  

E-Print Network [OSTI]

Today, Life Cycle Costing is one of the most popular ways of assessing a project's or an investment's worth to a company. This method of assessment is often applied to all stages of a investment's lifecycle, starting from ...

Angeles, Jon Virgil V

2011-01-01T23:59:59.000Z

451

Understanding government and railroad strategy for crude oil transportation in North America  

E-Print Network [OSTI]

On July 6, 2013, an oil-laden unit train derailed and exploded in Lac-Megantic, Quebec, Canada, killing 47 people, shocking and saddening many, and leading to significantly increased public scrutiny of crude oil transported ...

Carlson, S. Joel (Stephen Joel)

2014-01-01T23:59:59.000Z

452

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

E-Print Network [OSTI]

perform the specified task or job and (2) the path by which work and/or information flows through the team in order to optimally perform the specified task or job (Arthur, Villado, & Bennett, in press). 4 Theories of team mental models suggest... successful team performance requires the individual to work with members of the team in order to optimally perform the specified task or job (p. 3). Although tasks are often referred to as 搃nterdependent if to some degree they require input from...

Kyte, Tobin Bruce

2008-10-10T23:59:59.000Z

453

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

E-Print Network [OSTI]

. 18 18 23 28 40 40 42 42 47 50 CHAPI'ER V CONCLUSIONS CONCLUSIONS RECOMMENDATIONS Page 59 59 60 REFERENCES 62 APPENDIX A FIELD DATA FORMS . APPENDIX B TEST SITE DESCRIPTIONS APPENDIX C SPEED GRAPHS AND SUMMARIES... TO TRAINS Sign Page Figure 4. Experimental Sign System with YIELD TO TRAINS Sign . . . . . 20 Figure 5. Experimental Sign System with LOOK FOR TRAINS Sign . . . . 22 Figure 6. Texas Counties Selected for Study Sites Figure 7. Results of Driver Opinion...

Beitler, Michael Morgan

2012-06-07T23:59:59.000Z

454

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

E-Print Network [OSTI]

. This research focused on three specific areas. First, average vehicle delay was examined, and this delay was compared for seven specific train speed distributions, including existing conditions. Furthermore, each distribution was associated with train...

Tydlacka, Jonathan Michael

2004-11-15T23:59:59.000Z

455

Development of an Ontology to Assist the Modeling of Accident Scenarii "Application on Railroad Transport "  

E-Print Network [OSTI]

In a world where communication and information sharing are at the heart of our business, the terminology needs are most pressing. It has become imperative to identify the terms used and defined in a consensual and coherent way while preserving linguistic diversity. To streamline and strengthen the process of acquisition, representation and exploitation of scenarii of train accidents, it is necessary to harmonize and standardize the terminology used by players in the security field. The research aims to significantly improve analytical activities and operations of the various safety studies, by tracking the error in system, hardware, software and human. This paper presents the contribution of ontology to modeling scenarii for rail accidents through a knowledge model based on a generic ontology and domain ontology. After a detailed presentation of the state of the art material, this article presents the first results of the developed model.

Maalel, Ahmed; Mejri, Lassad; Ghezela, Henda Hajjami Ben

2012-01-01T23:59:59.000Z

456

Guidelines for storage distance between highway-highway and highway-railroad intersections  

E-Print Network [OSTI]

movemenLs are subject to progression, and/or for design volume-to-capacity ratios less than one. The following two equations were selected to evaluate the queuing probability model: Arrival, X = 1. 1 x V/3600 where X = an ival rate in equivalent...). The an ival rate and service rate can be expressed m any unit of time as long as the units are the same for both rate. TABLE I Queuing Performance Equations for Random Arrival-Random Service Single Channel Systems Queuing Characteristics M/M/1 (, FIFO...

Harry, Scott Raymond

2000-01-01T23:59:59.000Z

457

E-Print Network 3.0 - american railroads mechanical Sample Search...  

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

Environmental Engineering Alumni Association--SpringSummer As a boy, Jim Brix loved trains Summary: to the department in 1998 from a position with the Association of American...

458

16 TAC 3.30 - Memorandum of Understanding between the Railroad Commission  

Open Energy Info (EERE)

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459

Evaluation of Shortline Railroads & SNF/HLW Rail Shipment Inspections  

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power SystemsResources DOEElectricalonJustice EnvironmentalDISTRIBUTIO192-01Tasked for the

460

Annual Energy Outlook 2014 foresees growth of LNG as a fuel for railroads  

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 on Google Bookmark EERE: Alternative Fuels Data CenterFranconia, Virginia:FAQ < RAPID Jump to:SeadovCooperativeA2. World liquids consumption by region,Purchases211AlabamaAnnual14, 2014

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


461

Thermal And-Or Near Infrared At Railroad Valley Area (Laney, 2005) | Open  

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 on Google Bookmark EERE: Alternative Fuels Data CenterFranconia, Virginia:FAQ < RAPID Jump to:Seadov Pty LtdSteen,Ltd Jump Jump to:InformationTheInformation 9) JumpEnergyEnergy

462

Polynuclear aromatic hydrocarbons on the vegetation of a railroad right-of-way  

E-Print Network [OSTI]

(a)pyrone (P/BaP) ratios in leaf waxes to ratios reported in urban air pollution indicated that the py1'ene found in the leaf waxes was more stable than pyrene in the atmosphere. This suggested that pyrene, and possibly other PAH, were endogenous leaf wax... studied trace com- ponents in polluted air (HOFFi~WuN and WYDHER, 1968). They have been identified in the atmosphere of all large American cities (SAWACKI, 1967) and are present in a variety of environmental substances with which humans are in contact...

Hancock, James Leonard

1969-01-01T23:59:59.000Z

463

Formal Safety analysis of a radiobased railroad crossing using Deductive CauseConsequence  

E-Print Network [OSTI]

#ects analysis (FMEA) and fault tree analysis (FTA). We apply the method to a real world case study: a radio (DCCA). This technique is a formal generalization of well璳nown safety analysis methods like FMEA [10 by analyzed) than traditional FMEA. We show, that the results of DCCA have the same semantics as those

Reif, Wolfgang

464

Formal Safety analysis of a radio-based railroad crossing using Deductive Cause-Consequence  

E-Print Network [OSTI]

and effects analysis (FMEA) and fault tree analysis (FTA). We apply the method to a real world case study like FMEA [10], FMECA [4] and FTA [3]. The logical framework of DCCA may be used to rigorously verify of what can by analyzed) than traditional FMEA. We show, that the results of DCCA have the same semantics

Reif, Wolfgang

465

Reversible high-pressure carbon nanotube vessel  

E-Print Network [OSTI]

Applying a full pressure loop, i.e., loading and unloading, on a nanocrystal with in situ observation remains a challenge to experimentalists up until now. Using a multiwalled carbon nanotube, we realize the pressure loop ...

Wang, Lifeng

466

Submarine cargo vessels : opportunities for future transport.  

E-Print Network [OSTI]

??The strategic importance of submarines from a defence perspective is generally well understood by the public and notwithstanding the secrecy surrounding their activities, they have (more)

Jones, B. D.

2008-01-01T23:59:59.000Z

467

MMA Tugboat/ Barge/ 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 on Google Bookmark EERE: Alternative Fuels Data Center Home5b9fcbce19 No revision hasInformation Earth's Heat JumpIncMAKGalway Bay IEOWCCatcher.pngWavemill <

468

Final Vitrification Melter And Vessels Evaluation Documentation |  

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power SystemsResourcesFLASH2011-11-OPAMFY 2007 TotalFinal Design Review ModuleDecemberDepartment

469

Annabella: a North American coasting vessel  

E-Print Network [OSTI]

and analysis of a type of craft that once was common to the eastern seaboard, including discussions about how the craft was designed and built for transporting specific cargoes, and how this ship may be representative of maritime activities and shipbuilding...

Claesson, Stefan Hans

1998-01-01T23:59:59.000Z

470

FIRE Vacuum Vessel Design and Analysis  

E-Print Network [OSTI]

electrical breaks required) 路 Provide access ports for heating (no NBI), diagnostics and remote maintenance 路 Aid in plasma stabilization - conducting shell - internal control coils 路 Maximum access for heating lifetime component - remotely welded joints are double contained - all bellows are double contained 路 High

471

Engineering functional blood vessels in vivo  

E-Print Network [OSTI]

At the present time, there are many hurdles to overcome in order to create a long-lasting and engineered tissue for tissue transplant in patients. The challenges include the isolation and expansion of appropriate cells, ...

Au, Pakwai

2008-01-01T23:59:59.000Z

472

IWTU Construction Workers Set Largest Process Vessel  

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: Vegetation ProposedUsingFun withconfinementEtching. | EMSLtheIndustry | Department ofT. Bonoli,