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Note: This page contains sample records for the topic "ocean-going vessels including" from the National Library of EnergyBeta (NLEBeta).
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1

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

2

Floating vessel  

SciTech Connect

The invention relates to a floating vessel which may be used in oil recovery. The assembly consists of a vertical column having a relatively small diameter. The column has a buoyancy capacity and is supplied with a ballast section having a larger diameter at its end. An upper structure is movably connected to the column. The column and the ballast chamber determine the limits of a shaft. The shaft is open at its lower end and is supplied with means to let fluid into the shaft over a relatively large area. (8 claims)

1974-05-14T23:59:59.000Z

3

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

4

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

5

Vessel structural support system  

SciTech Connect

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

6

Test of 6-inch-thick pressure vessels. Series 1: intermediate test vessels V-1 and V-2  

SciTech Connect

The intermediate vessel tests have been subdivided into four seriesi flaws in cylindrical vessels, A508, class 2 forging steel-two vessels; flaws in cylindrical vessels with longitudinal weld seams, A508, class 2 forging steel, submerged-arc welds-three vessels; flaws in cylindrical vessels wlth longitudinal weld seams, A533, grade B, class l plate steel, submerged-arc weld-two vessels; and cylindrical vessels with radially attached nozzles, vessels of A508, chass 2 forging steel and A533, grade B, class 1 plate steel; nozzle of A508 class 2 forging steel-three vessels. A comprehensive description of the pertinent factors considered in the design of the vessels is presented. Construction of the test facility and documentation of test results and fracture predictions are included. Emphasis is placed on providing the test results in such a manner that they form a resource for amy investigators interested in the problem of fracture. (auth)

Derby, R.W.; Merkle, J.G.; Robinson, G.C.; Whitman, G.D.; Witt, F.J.

1974-02-01T23:59:59.000Z

7

Prediction of Vessel Icing  

Science Conference Proceedings (OSTI)

Vessel icing from wave-generated spray is a severe hazard to expanded marine operations in high latitudes. Hardships in making observations during operations, combined with differences in vessel type and heading, have resulted in great ...

J. E. Overland; C. H. Pease; R. W. Preisendorfer; A. L. Comiskey

1986-12-01T23:59:59.000Z

8

Vacuum Vessel Remote Handling  

E-Print Network (OSTI)

FIRE Vacuum Vessel and Remote Handling Overview B. Nelson, T. Burgess, T. Brown, H-M Fan, G. Jones #12;13 July 2002 Snowmass Review: FIRE Vacuum Vessel and Remote Handling 2 Presentation Outline · Remote Handling - Maintenance Approach & Component Classification - In-Vessel Transporter - Component

9

Reactor vessel using metal oxide ceramic membranes  

DOE Patents (OSTI)

A reaction vessel for use in photoelectrochemical reactions includes as its reactive surface a metal oxide porous ceramic membrane of a catalytic metal such as titanium. The reaction vessel includes a light source and a counter electrode. A provision for applying an electrical bias between the membrane and the counter electrode permits the Fermi levels of potential reaction to be favored so that certain reactions may be favored in the vessel. The electrical biasing is also useful for the cleaning of the catalytic membrane.

Anderson, Marc A. (Madison, WI); Zeltner, Walter A. (Oregon, WI)

1992-08-11T23:59:59.000Z

10

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

11

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

12

Cover Heated, Open Vessels  

SciTech Connect

This revised ITP steam tip sheet on covering heated, open vessels provides how-to advice for improving industrial steam systems using low-cost, proven practices and technologies.

2006-01-01T23:59:59.000Z

13

Reconnecting broken blood vessels  

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

Reconnecting broken blood vessels Reconnecting broken blood vessels Name: Catherine A Kraft Status: N/A Age: N/A Location: N/A Country: N/A Date: N/A Question: While watching the television program "Chicago Hope" the other day, I watched a doctor sew someone's ear back on using an elaborate microscope. I was wondering if a surgeon is required to reconnect all the broken blood vessels, and how you would accomplish this? Thanks for your time! Replies: I'm not a surgeon, but I think the answer to your question is "no." The blood will flow across the wound (out the end of one blood vessel and into the end of another), although not efficiently. I believe they sometimes use leeches sucking on the end of the reconnected part to help induce flow of blood in the right direction through the area. You probably do need to put the ends of the major vessels near each other, so the distribution of blood flow is reasonably like it was before the injury, and so the vessels can eventually reconnect. But probably the microscope is used mostly to be sure the various layers of muscle, connective tissue, and fat are connected together correctly.

14

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

15

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

16

Decisions decisions plant vessels  

Science Conference Proceedings (OSTI)

This paper describes concepts for a family of plant vessels that help users make decisions or reach goals. The concepts use plants to mark time or answer questions for the user, creating a connection between the user and the individual plant. These concepts ...

Jenny Liang

2007-08-01T23:59:59.000Z

17

Radiant vessel auxiliary cooling system  

DOE Patents (OSTI)

In a modular liquid-metal pool breeder reactor, a radiant vessel auxiliary cooling system is disclosed for removing the residual heat resulting from the shutdown of a reactor by a completely passive heat transfer system. A shell surrounds the reactor and containment vessel, separated from the containment vessel by an air passage. Natural circulation of air is provided by air vents at the lower and upper ends of the shell. Longitudinal, radial and inwardly extending fins extend from the shell into the air passage. The fins are heated by radiation from the containment vessel and convect the heat to the circulating air. Residual heat from the primary reactor vessel is transmitted from the reactor vessel through an inert gas plenum to a guard or containment vessel designed to contain any leaking coolant. The containment vessel is conventional and is surrounded by the shell.

Germer, John H. (San Jose, CA)

1987-01-01T23:59:59.000Z

18

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

19

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

20

Reactor Vessel Embrittlement Management Handbook: A Handbook for Managing Reactor Vessel Embrittlement and Vessel Integrity  

Science Conference Proceedings (OSTI)

For many reactor pressure vessels, embrittlement is the primary concern for continued safe operation. The shutdown of the Yankee Rowe plant because of uncertainties related to embrittlement of the vessel demonstrates the importance of adequately addressing embrittlement issues. Managing embrittlement requires integration, management, and implementation of diverse technical, regulatory, planning, and economic activities. An effective embrittlement management program will ensure vessel safety and reliabili...

1994-01-22T23:59:59.000Z

Note: This page contains sample records for the topic "ocean-going vessels including" 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

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

22

BWRVIP-244: BWR Vessel and Internals Project, Nondestructive Evaluation Development 2010  

Science Conference Proceedings (OSTI)

This report provides 2010 results of the nondestructive evaluation NDE development task of the Boiling Water Reactor Vessel and Internals Project BWRVIP Inspection Focus Group. The scope of activity includes applications of various NDE techniques to boiling water reactor vessels and vessel internals components.

2010-12-23T23:59:59.000Z

23

Core Vessel Insert Handling Robot for the Spallation Neutron Source  

Science Conference Proceedings (OSTI)

The Spallation Neutron Source provides the world's most intense pulsed neutron beams for scientific research and industrial development. Its eighteen neutron beam lines will eventually support up to twenty-four simultaneous experiments. Each beam line consists of various optical components which guide the neutrons to a particular instrument. The optical components nearest the neutron moderators are the core vessel inserts. Located approximately 9 m below the high bay floor, these inserts are bolted to the core vessel chamber and are part of the vacuum boundary. They are in a highly radioactive environment and must periodically be replaced. During initial SNS construction, four of the beam lines received Core Vessel Insert plugs rather than functional inserts. Remote replacement of the first Core Vessel Insert plug was recently completed using several pieces of custom-designed tooling, including a highly complicated Core Vessel Insert Robot. The design of this tool are discussed.

Graves, Van B [ORNL; Dayton, Michael J [ORNL

2011-01-01T23:59:59.000Z

24

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

25

LPG storage vessel cracking experience  

SciTech Connect

In order to evaluate liquefied petroleum gas (LPG) handling and storage hazards, Caltex Petroleum Corp. (Dallas) surveyed several installations for storage vessel cracking problems. Cracking was found in approximately one-third of the storage vessels. In most cases, the cracking appeared to be due to original fabrication problems and could be removed without compromising the pressure containment. Several in-service cracking problems found were due to exposure to wet hydrogen sulfide. Various procedures were tried in order to minimize the in-service cracking potential. One sphere was condemned because of extensive subsurface cracking. This article's recommendations concern minimizing cracking on new and existing LPG storage vessels.

Cantwell, J.E. (Caltex Petroleum Corp., P.O. Box 619500, Dallas, TX (US))

1988-10-01T23:59:59.000Z

26

LPG storage vessel cracking experience  

SciTech Connect

As part of an overall company program to evaluate LPG handling and storage hazards the authors surveyed several installations for storage vessel cracking problems. Cracking was found in approximately one third of the storage vessels. In most cases the cracking appeared due to original fabrication problems and could be removed without compromising the pressure containment. Several in-service cracking problems due to exposure to wet hydrogen sulfide were found. Various procedures were tried in order to minimize the in-service cracking potential. One sphere was condemned because of extensive subsurface cracking. Recommendations are made to minimize cracking on new and existing LPG storage vessels.

Cantwell, J.E.

1988-01-01T23:59:59.000Z

27

CHF Enhancement by Vessel Coating for External Reactor Vessel Cooling  

SciTech Connect

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

28

Cryogenic Pressure Vessels: Progress and Plans  

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

Pressure Vessel workshop, LLNL, February 15, 2011, p. 1 Cryogenic Pressure Vessels: Progress and Plans Salvador Aceves, Gene Berry, Francisco Espinosa, Ibo Matthews, Guillaume...

29

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

30

Reactor vessel seal service fixture  

DOE Patents (OSTI)

An apparatus for the preparation of exposed sealing surfaces along the open rim of a nuclear reactor vessel comprised of a motorized mechanism for traveling along the rim and simultaneously brushing the exposed surfaces is described.

Ritz, W.C.

1975-12-01T23:59:59.000Z

31

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

32

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

33

Nuclear reactor pressure vessel support system  

DOE Patents (OSTI)

A support system for nuclear reactor pressure vessels which can withstand all possible combinations of stresses caused by a postulated core disrupting accident during reactor operation. The nuclear reactor pressure vessel is provided with a flange around the upper periphery thereof, and the flange includes an annular vertical extension formed integral therewith. A support ring is positioned atop of the support ledge and the flange vertical extension, and is bolted to both members. The plug riser is secured to the flange vertical extension and to the top of a radially outwardly extension of the rotatable plug. This system eliminates one joint through which fluids contained in the vessel could escape by making the fluid flow path through the joint between the flange and the support ring follow the same path through which fluid could escape through the plug risers. In this manner, the sealing means to prohibit the escape of contained fluids through the plug risers can also prohibit the escape of contained fluid through the securing joint.

Sepelak, George R. (McMurray, PA)

1978-01-01T23:59:59.000Z

34

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations, which govern the operation of commercial nuclear power plants, require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including

35

Annabella: a North American coasting vessel  

E-Print Network (OSTI)

The coasting schooner Annabella was built at Port Elizabeth, New Jersey, in 1834. Originally constructed as a sloop, the vessel was built specifically for transporting raw materials such as cordwood, brick, coal, and perishables to markets and industries along the northeast United States coast. During its lengthy 50-year career, ownership of Annabella was transferred among numerous merchants in Philadelphia, Plymouth, Boston, and, finally, Cape Neddick, Maine. The vessel was finally abandoned on October 17, 1885, in the Cape Neddick River, in Cape Neddick, Maine, beyond repair and no longer fit for service. This study covers the following topics: the 1994 and 1995 archaeological field seasons, including hull and artifact descriptions and analyses; the history of the coasting trade and the cordwood industry during the 19th century in the vicinity of southern Maine; and an analysis of historical documents that detail the history ofannabella. Toward these ends, this thesis will present a description 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 technologies of the 19th century

Claesson, Stefan Hans

1998-01-01T23:59:59.000Z

36

PRESSURE VESSEL FABRICATION USING T-1 STEEL  

SciTech Connect

The fabrication of pressure vessels using C-l steel is described. The welding, welding electrodes, explosionbulge test, and impact and fatigue properties for the pressure vessel are given. (W.L.H.)

Franco-Ferreira, E.A.

1957-11-14T23:59:59.000Z

37

Cryostat including heater to heat a target  

DOE Patents (OSTI)

A cryostat is provided which comprises a vacuum vessel; a target disposed within the vacuum vessel; a heat sink disposed within the vacuum vessel for absorbing heat from the detector; a cooling mechanism for cooling the heat sink; a cryoabsorption mechanism for cryoabsorbing residual gas within the vacuum vessel; and a heater for maintaining the target above a temperature at which the residual gas is cryoabsorbed in the course of cryoabsorption of the residual gas by the cryoabsorption mechanism. 2 figs.

Pehl, R.H.; Madden, N.W.; Malone, D.F.

1990-09-11T23:59:59.000Z

38

Certification Testing and Demonstration of Insulated Pressure Vessels for Vehicular Hydrogen and Natural Gas Storage  

Science Conference Proceedings (OSTI)

We are working on developing an alternative technology for storage of hydrogen or natural gas on light-duty vehicles. This technology has been titled insulated pressure vessels. Insulated pressure vessels are cryogenic-capable pressure vessels that can accept either liquid fuel or ambient-temperature compressed fuel. Insulated pressure vessels offer the advantages of cryogenic liquid fuel tanks (low weight and volume), with reduced disadvantages (fuel flexibility, lower energy requirement for fuel liquefaction and reduced evaporative losses). The work described in this paper is directed at verifying that commercially available pressure vessels can be safely used to store liquid hydrogen or LNG. The use of commercially available pressure vessels significantly reduces the cost and complexity of the insulated pressure vessel development effort. This paper describes a series of tests that have been done with aluminum-lined, fiber-wrapped vessels to evaluate the damage caused by low temperature operation. All analysis and experiments to date indicate that no significant damage has resulted. Future activities include a demonstration project in which the insulated pressure vessels will be installed and tested on two vehicles. A draft standard will also be generated for obtaining insulated pressure vessel certification.

Aceves, S M; Martinez-Frias, J; Espinosa-Loza, F; Schaffer, R; Clapper, W

2002-05-22T23:59:59.000Z

39

Mechanical behavior analysis of CDIO production-blood vessel robot in curved blood vessel  

Science Conference Proceedings (OSTI)

In order to analyze mechanical behavior of blood vessel robot (student's CDIO production) in curved blood, and provide the data for outline design of robot, the flow field out side of robot is numerical simulated. The results show that the vessel shape ... Keywords: blood vessel robot, curved blood vessel, mechanical behavior analysis, numerical simulation

Fan Jiang; Chunliang Zhang; Yijun Wang

2010-10-01T23:59:59.000Z

40

BWRVIP-269: BWR Vessel and Internals Project, Nondestructive Evaluation Development 2012  

Science Conference Proceedings (OSTI)

The purpose of this report is to describe the results of nondestructive evaluation (NDE) activities conducted in the previous year within the NDE Development task of the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Inspection Focus Group. The scope of the ongoing NDE Development task includes the reactor vessel and its internal components. This task attempts to develop solutions for the more difficult ...

2012-12-12T23:59:59.000Z

Note: This page contains sample records for the topic "ocean-going vessels including" 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

Coal gasification vessel. [Patent application  

DOE Patents (OSTI)

A vessel system comprises an outer shell of carbon fibers held in a binder, a coolant circulation mechanism and control mechanism and an inner shell 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 can be computer controlled and can be used to monitor and modulate the coolant which is provided through the circulation mechanism for cooling and protecting the carbon fiber and outer shell. The control mechanism is also used to locate any isolated hot spots which may occur through the local disintegration of the inner refractory shell.

Loo, B.W.

1981-03-17T23:59:59.000Z

42

Marine Casualty and Pollution Database - Vessel Events for 2002 to 2010 |  

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

Vessel Events for 2002 to 2010 Vessel Events for 2002 to 2010 Ocean Data Tools Technical Guide Map Gallery Regional Planning Feedback Ocean You are here Data.gov » Communities » Ocean » Data Marine Casualty and Pollution Database - Vessel Events for 2002 to 2010 Dataset Summary Description The Marine Casualty and Pollution Data files provide details about marine casualty and pollution incidents investigated by Coast Guard Offices throughout the United States. The database can be used to analyze marine accidents and pollution incidents by a variety of factors including vessel or facility type, injuries, fatalities, pollutant details, location, and date. The data collection period began in 1982 for marine casualties and 1973 for polluting incidents, and is ongoing. Documentation includes entity and attribute descriptions along with suggested solutions to general marine pollution, vessel casualty, and personnel injury and death questions.

43

IWTU Construction Workers Set Largest Process Vessel  

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

IWTU Construction Workers Set Largest Process Vessel IWTU Construction Workers Set Largest Process Vessel Click on image to enlarge Construction of the Integrated Waste Treatment Unit (IWTU) took a major step forward on Sept. 2, 2009 as crews lifted into place the largest of the six process vessels that will be used to treat radioactive liquid waste stored at the site. The IWTU will use a steam reforming process to solidify the waste for eventual shipment out of Idaho. The vessel and its skid, or framework, were constructed at Premier Technologies in Blackfoot. (Premier is the main small business partner for CH2M-WG Idaho (CWI), the contractor for DOE's Idaho Cleanup Project.) The Carbon Reduction Reformer vessel and skid weigh approximately 60 tons (120,000 lbs.). Because of the weight of the vessel and the location of the

44

MMA Tugboat/ Barge/ Vessel | Open Energy Information  

Open Energy Info (EERE)

MMA Tugboat/ Barge/ Vessel MMA Tugboat/ Barge/ Vessel Jump to: navigation, search Basic Specifications Facility Name MMA Tugboat/ Barge/ Vessel Overseeing Organization Maine Maritime Academy Hydrodynamic Testing Facility Type Tow Vessel Depth(m) 15.2 Water Type Saltwater Cost(per day) Contact POC Special Physical Features Tug: 73 ft (2)16V-92 Detroits Barge: 43 ft by 230ft Research Vessel Friendship: 40 foot vessel w/ 6 cylinder Cummins diesel engine and A-Frame crane Towing Capabilities Towing Capabilities Yes Maximum Velocity(m/s) 5.1 Wavemaking Capabilities Wavemaking Capabilities None Channel/Tunnel/Flume Channel/Tunnel/Flume None Wind Capabilities Wind Capabilities None Control and Data Acquisition Description Full onbard Navigation, GPS, marine radar and depth plotter; standard PC onboard can be configured as needed for data acquisition needs

45

Neutron Assay System for Confinement Vessel Disposition  

Science Conference Proceedings (OSTI)

Los Alamos National Laboratory has a number of spherical confinement vessels (CVs) remaining from tests involving nuclear materials. These vessels have an inner diameter of 6 feet with 1-inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the CVs. The Confinement Vessel Assay System (CVAS) was developed to measure the amount of special nuclear material (SNM) in CVs before and after cleanout. Prior to cleanout, the system will be used to perform a verification measurement of each vessel. After cleanout, the system will be used to perform safeguards-quality assays of {le}100-g {sup 239}Pu equivalent in a vessel for safeguards termination. The CVAS has been tested and calibrated in preparation for verification and safeguards measurements.

Frame, Katherine C. [Los Alamos National Laboratory; Bourne, Mark M. [Los Alamos National Laboratory; Crooks, William J. [Los Alamos National Laboratory; Evans, Louise [Los Alamos National Laboratory; Mayo, Douglas R. [Los Alamos National Laboratory; Miko, David K. [Los Alamos National Laboratory; Salazar, William R. [Los Alamos National Laboratory; Stange, Sy [Los Alamos National Laboratory; Valdez, Jose I. [Los Alamos National Laboratory; Vigil, Georgiana M. [Los Alamos National Laboratory

2012-07-13T23:59:59.000Z

46

Rancho Seco Reactor Vessel Segmentation Experience Report  

Science Conference Proceedings (OSTI)

This report documents the approach taken by Sacramento Municipal Utility District (SMUD) in the segmentation and disposal of the Reactor Vessel from the Rancho Seco Nuclear Generating Station (RSNGS). The location of the Rancho Seco plant placed major constraints on the shipping options available for large plant components (Steam Generators and Reactor Vessel). This report details the engineering evaluations leading to the segmentation and disposal of the Reactor Vessel (RV). It describes the key element...

2008-03-18T23:59:59.000Z

47

Embrittlement of Nuclear Reactor Pressure Vessels  

Science Conference Proceedings (OSTI)

Boiler and Pressure Vessel Code Section III App. G Protection Against Nonductile Fracture (New York: American Society of Mechanical Engineers, 1986 ). 3.

48

Lightweight bladder lined pressure vessels - Energy Innovation ...  

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 ...

49

Cryostat including heater to heat a target  

DOE Patents (OSTI)

A cryostat is provided which comprises a vacuum vessel; a target disposed within the vacuum vessel; a heat sink disposed within the vacuum vesssel for absorbing heat from the detector; a cooling mechanism for cooling the heat sink; a cryoabsorption mechanism for cryoabsorbing residual gas within the vacuum vessel; and a heater for maintaining the target above a temperature at which the residual gas is cryoabsorbed in the course of cryoabsorption of the residual gas by the cryoabsorption mechanism.

Pehl, Richard H. (Berkeley, CA); Madden, Norman W. (Livermore, CA); Malone, Donald F. (Oakland, CA)

1990-01-01T23:59:59.000Z

50

Pressure vessel and piping codes  

SciTech Connect

Section III of the ASME Boiler and Pressure Vessel Code contains simplified design formulas for placing bounds on the plastic deformations in nuclear power plant piping systems. For Class 1 piping a simple equation is given in terms of primary load stress indices (B/sub 1/ and B/sub 2/) and nominal pressure and bending stresses. The B/sub 1/ and B/sub 2/ stress indices reflect the capacities of various piping products to carry load without gross plastic deformation. In this paper, the significance of the indices, nominal stresses, and limits given in the Code for Class 1 piping and corresponding requirements for Class 2 and Class 3 piping are discussed. Motivation behind recent (1978-1981) changes in the indices and in the associated stress limits is presented.

Moore, S.E.; Rodabaugh, E.C.

1982-11-01T23:59:59.000Z

51

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

52

A review of vessel extraction techniques and algorithms  

Science Conference Proceedings (OSTI)

Vessel segmentation algorithms are the critical components of circulatory blood vessel analysis systems. We present a survey of vessel extraction techniques and algorithms. We put the various vessel extraction approaches and techniques in perspective ... Keywords: Magnetic resonance angiography, X-ray angiography, medical imaging, neurovascular, vessel extraction

Cemil Kirbas; Francis Quek

2004-06-01T23:59:59.000Z

53

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

54

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

55

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel...  

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

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

56

Analysis of Crack Development Involving a Pressure Vessel in a ...  

Science Conference Proceedings (OSTI)

The vessel is part of a by-product refining system comprising a synthetic natural gas production plant. The vessel processes a mixture of chemical species, ...

57

HFIR In-Vessel Irradiation Facilities | ORNL Neutron Sciences  

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

Home Facilities HFIR In-Vessel Irradiation In-Vessel Irradiation Experiment Facilities The HFIR provides a variety of in-core irradiation facilities, allowing for a...

58

Resistance upset welding for vessel fabrication  

SciTech Connect

Solid-state resistance upset welding has been successfully applied to fabrication of small vessels. The process has advantages compared with the fusion welding processes currently used to join the two halves of such vessels. These advantages result from the improved metallurgical properties of the weld zone and the simplicity of the welding process. Spherical and cylindrical shapes have been fabricated using the upset welding process. Nondestructive and destructive tests have shown excellent weld strength. Storage tests have demonstrated long term compatibility of the welds for cylindrical parts made from 304L stainless steel that have been in storage for eight years. Spherical vessels and reinforced desip vessels made from forged 21-6-9 stainless steel have been prepared for storage.

Kanne, W.R. Jr.

1992-01-01T23:59:59.000Z

59

Resistance upset welding for vessel fabrication  

SciTech Connect

Solid-state resistance upset welding has been successfully applied to fabrication of small vessels. The process has advantages compared with the fusion welding processes currently used to join the two halves of such vessels. These advantages result from the improved metallurgical properties of the weld zone and the simplicity of the welding process. Spherical and cylindrical shapes have been fabricated using the upset welding process. Nondestructive and destructive tests have shown excellent weld strength. Storage tests have demonstrated long term compatibility of the welds for cylindrical parts made from 304L stainless steel that have been in storage for eight years. Spherical vessels and reinforced desip vessels made from forged 21-6-9 stainless steel have been prepared for storage.

Kanne, W.R. Jr.

1992-10-01T23:59:59.000Z

60

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

Note: This page contains sample records for the topic "ocean-going vessels including" 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

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)

2012-02-14T23:59:59.000Z

62

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

63

Vessel Sanitation Program 2011 Operations Manual  

E-Print Network (OSTI)

and Prevention (CDC) established the Vessel Sanitation Program (VSP) in the 1970s as a cooperative activity............................................................................................ 1 1.1.1 Cooperative Activity.......................................................................................................... 1 1.2 Activities

64

Final Vitrification Melter And Vessels Evaluation Documentation  

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

DOE has prepared final evaluations and made waste incidental to reprocessing determinations for the vitrification melter and feed vessels (the concentrator feed makeup tank and the melter feed hold...

65

TMI-2 Vessel Investigation Project integration report  

Science Conference Proceedings (OSTI)

The Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project (VIP) was an international effort that was sponsored by the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. The primary objectives of the VIP were to extract and examine samples from the lower head and to evaluate the potential modes of failure and the margin of structural integrity that remained in the TMI-2 reactor vessel during the accident. This report presents a summary of the major findings and conclusions that were developed from research during the VIP. Results from the various elements of the project are integrated to form a cohesive understanding of the vessel`s condition after the accident.

Wolf, J. R.; Rempe, J. L.; Stickler, L. A.; Korth, G. E.; Diercks, D. R.; Neimark, L. A.; Akers, D W; Schuetz, B. K.; Shearer, T L; Chavez, S. A.; Thinnes, G. L.; Witt, R. J.; Corradini, M L; Kos, J. A. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

1994-03-01T23:59:59.000Z

66

PRESSURIZATION OF CONTAINMENT VESSELS FROM PLUTONIUM OXIDE CONTENTS  

Science Conference Proceedings (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

67

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

Assessment of High Value Surveillance Materials Assessment of High Value Surveillance Materials Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Assessment of High Value Surveillance Materials The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely

68

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

DOE Green Energy (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’s secondary containment system at T Plant. A previous study demonstrated that sloped walls on vessels, both cylindrical coned-shaped vessels and rectangular vessels with rounded ends, provided an effective approach for disrupting a VSB by creating a release path for gas as a VSB began to rise. Based on the success of sloped-wall vessels, a similar concept is investigated here where a sloped fin is placed inside the vessel to create a release path for gas. A key potential advantage of using a sloped fin compared to a vessel with a sloped wall is that a small fin decreases the volume of a vessel available for sludge storage by a very small fraction compared to a cone-shaped vessel. The purpose of this study is to quantify the capability of sloped fins to disrupt VSBs and to conduct sufficient tests to estimate the performance of fins in full-scale STSCs. Experiments were conducted with a range of fin shapes to determine what slope and width were sufficient to disrupt VSBs. Additional tests were conducted to demonstrate how the fin performance scales with the sludge layer thickness and the sludge strength, density, and vessel diameter based on the gravity yield parameter, which is a dimensionless ratio of the force necessary to yield the sludge to its weight.( ) Further experiments evaluated the difference between vessels with flat and 2:1 elliptical bottoms and a number of different simulants, including the KW container sludge simulant (complete), which was developed to match actual K-Basin sludge. Testing was conducted in 5-in., 10-in., and 23-in.-diameter vessels to quantify how fin performance is impacted by the size of the test vessel. The most significant results for these scale-up tests are the trend in how behavior changes with vessel size and the results from the 23-in. vessel. The key objective in evaluating fin performance is to determine the conditions that minimize the volume of a VSB when disruption occurs because this reduces the potential for material inside the STSC from being released through vents.

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

69

Decommissioning: Reactor Pressure Vessel Internals Segmentation  

Science Conference Proceedings (OSTI)

Decommissioning a nuclear plant covers a wide variety of challenging projects. One of the most challenging areas is the removal and disposal of the reactor pressure vessel (RPV) and the RPV internals. This report describes commercial reactor pressure vessel segmentation projects that have been completed and discusses several projects that are still in the planning stages. The report also covers lessons learned from each project.

2001-10-11T23:59:59.000Z

70

BWRVIP-277: BWR Vessel and Internals Project, Nondestructive Evaluation Development 2013  

Science Conference Proceedings (OSTI)

This report describes the results of nondestructive evaluation (NDE) activities conducted in the previous year within the NDE development task of the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Inspection Focus Group. The scope of the ongoing task includes the reactor vessel and its internal components. This task attempts to develop solutions for the more difficult inspections recommended by the BWRVIP, to determine limitations to inspection capability, and sometimes to demonstrate ...

2013-11-27T23:59:59.000Z

71

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

DOE Green Energy (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

72

Dr. John L. Gross  

Science Conference Proceedings (OSTI)

... Des Moines Steel Company, where he worked on various projects including the analysis and design of an ocean-going LNG containment system ...

2013-05-06T23:59:59.000Z

73

Welding stainless and 9% nickel steel cryogenic vessels  

SciTech Connect

Gases are often more efficiently stored and shipped as liquids at cryogenic temperatures. Pure gases commonly stored below liquefaction temperatures include oxygen {minus}297 F ({minus}183 C), argon {minus}302 f ({minus}186 C), nitrogen {minus}320 F ({minus}196 C), hydrogen {minus}423 F ({minus}253 C) and helium {minus}452 F ({minus}269 C). Natural gas is also transported and frequently stored as liquefied natural gas (LNG) at temperatures below {minus}261 F ({minus}163 C). Storage tanks for the pure gases are generally shop fabricated in sizes that can be shipped by conventional carriers. Smaller LNG vessels for over-the-road and railroad fuel applications are also shop-fabricated. Shown in a figure is a rail-mounted tank designed to supply liquefied natural gas to locomotives. Another example of a tank installation is also shown. LNG terminal storage tanks are generally field-erected vessels fabricated from 9% nickel steel in sizes of 50,000 to 100,000 m{sup 3} (315,000 to 630,000 bbls). This article focuses on welding practices for shop-fabricated vessels and equipment.

Avery, R.E. [Nickel Development Inst., Londonderry, NH (United States); Parsons, D. [Parsons (David), Hampstead, NH (United States)

1995-11-01T23:59:59.000Z

74

Fast neutron fluence of yonggwang nuclear unit 1 reactor pressure vessel  

Science Conference Proceedings (OSTI)

The Code of Federal Regulations, Title 10, Part 50, Appendix H, requires that the neutron dosimetry be present to monitor the reactor vessel throughout plant life. The Ex-Vessel Neutron Dosimetry System has been installed for Yonggwang Nuclear Unit 1 after complete withdrawal of all six in-vessel surveillance capsules. This system has been installed in the reactor cavity annulus in order to measure the fast neutron spectrum coming out through the reactor pressure vessel. Cycle specific neutron transport calculations were performed to obtain the energy dependent neutron flux throughout the reactor geometry including dosimetry positions. Comparisons between calculations and measurements were performed for the reaction rates of each dosimetry sensors and results show good agreements. (authors)

Yoo, C.; Km, B.; Chang, K.; Leeand, S. [Korea Atomic Energy Research Inst., 150 Dukjin-dong, Yuseung-gu, Daejeon 305-353 (Korea, Republic of); Park, J. [Chungnam National Univ., 220 Gung-dong, Yuseung-gu, Daejeon 305-764 (Korea, Republic of)

2006-07-01T23:59:59.000Z

75

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

76

Nonlinear response of vessel walls due to short-time thermomechanical loading  

Science Conference Proceedings (OSTI)

Maintaining structural integrity of the reactor pressure vessel (RPV) during a postulated core melt accident is an important safety consideration in the design of the vessel. This study addresses the failure predictions of the vessel due to thermal and pressure loadings fro the molten core debris depositing on the lower head of the vessel. Different loading combinations were considered based on the dead load, yield stress assumptions, material response and internal pressurization. The analyses considered only short term failure (quasi static) modes, long term failure modes were not considered. Short term failure modes include plastic instabilities of the structure and failure due to exceeding the failure strain. Long term failure odes would be caused by creep rupture that leads to plastic instability of the structure. Due to the sort time durations analyzed, creep was not considered in the analyses presented.

Pfeiffer, P.A.; Kulak, R.F.

1994-06-01T23:59:59.000Z

77

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

78

EDS V25 containment vessel explosive qualification test report.  

SciTech Connect

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

79

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

80

Upgrade of the DIII-D vacuum vessel protection system  

SciTech Connect

An upgrade of the General Atomics DIII-D tokamak armor protection system has been completed. The upgrade consisted of armoring the outer wall and the divertor gas baffle with monolithic graphite tiles and cleaning the existing floor, ceiling, and inner wall tiles to remove any deposited impurity layer from the tile surfaces. The new tiles replace the graphite tiles used as local armor for neutral beam shine through, three graphite poloidal back-up limiter bands, and miscellaneous Inconel protection tiles. The total number of tiles increased from 1636 to 3200 and corresponding vessel coverage from 40% to 90%. A new, graphite armored, toroidally continuous, gas baffle between the outer wall and the biased divertor ring was installed in order to accommodate the cryocondensation pump that was installed in parallel with the outer wall tiles. To eliminate a source of copper in the plasma, GRAFOIL gaskets replaced the copper felt metal gaskets previously used as a compliant heat transfer interface between the inertially cooled tiles and the vessel wall. GRAFOIL, an exfoliated, flexible graphite material from Union Carbide, Inc., was used between each tile and the vessel wall and also between each tile and its hold-down hardware. Testing was performed to determine the mechanical compliance, thermal conductance, and vacuum characteristics of the GRAFOIL material. To further decrease the quantity of high Z materials exposed to the plasma, the 1636 existing graphite tiles were identified, removed, and grit blasted to eliminate a thin layer of deposited metals which included nickel, chromium, and molybdenum. Prior to any processing, a selected set of tiles was tested for radioactivity, including tritium contamination. The tiles were grit blasted in a negative-pressure blasting cabinet using 37 {mu}m boron carbide powder as the blast media and dry nitrogen as the propellant.

Hollerbach, M.A.; Lee, R.L.; Smith, J.P.; Taylor, P.L.

1993-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "ocean-going vessels including" from the National Library of EnergyBeta (NLEBeta).
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81

Appendix F Cultural Resources, Including  

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

Appendix F Appendix F Cultural Resources, Including Section 106 Consultation STATE OF CALIFORNIA - THE RESOURCES AGENCY EDMUND G. BROWN, JR., Governor OFFICE OF HISTORIC PRESERVATION DEPARTMENT OF PARKS AND RECREATION 1725 23 rd Street, Suite 100 SACRAMENTO, CA 95816-7100 (916) 445-7000 Fax: (916) 445-7053 calshpo@parks.ca.gov www.ohp.parks.ca.gov June 14, 2011 Reply in Reference To: DOE110407A Angela Colamaria Loan Programs Office Environmental Compliance Division Department of Energy 1000 Independence Ave SW, LP-10 Washington, DC 20585 Re: Topaz Solar Farm, San Luis Obispo County, California Dear Ms. Colamaria: Thank you for seeking my consultation regarding the above noted undertaking. Pursuant to 36 CFR Part 800 (as amended 8-05-04) regulations implementing Section

82

Countries Gasoline Prices Including Taxes  

Gasoline and Diesel Fuel Update (EIA)

Countries (U.S. dollars per gallon, including taxes) Countries (U.S. dollars per gallon, including taxes) Date Belgium France Germany Italy Netherlands UK US 01/13/14 7.83 7.76 7.90 8.91 8.76 8.11 3.68 01/06/14 8.00 7.78 7.94 8.92 8.74 8.09 3.69 12/30/13 NA NA NA NA NA NA 3.68 12/23/13 NA NA NA NA NA NA 3.63 12/16/13 7.86 7.79 8.05 9.00 8.78 8.08 3.61 12/9/13 7.95 7.81 8.14 8.99 8.80 8.12 3.63 12/2/13 7.91 7.68 8.07 8.85 8.68 8.08 3.64 11/25/13 7.69 7.61 8.07 8.77 8.63 7.97 3.65 11/18/13 7.99 7.54 8.00 8.70 8.57 7.92 3.57 11/11/13 7.63 7.44 7.79 8.63 8.46 7.85 3.55 11/4/13 7.70 7.51 7.98 8.70 8.59 7.86 3.61 10/28/13 8.02 7.74 8.08 8.96 8.79 8.04 3.64 10/21/13 7.91 7.71 8.11 8.94 8.80 8.05 3.70 10/14/13 7.88 7.62 8.05 8.87 8.74 7.97 3.69

83

MAAP5 BWR Vessel Penetration and Ex-Vessel Equipment Model Enhancement Description  

Science Conference Proceedings (OSTI)

This report describes proposed enhancements to the Modular Accident Analysis Program (MAAP) vessel penetration and ex-vessel equipment model for BWR designs. MAAP is an EPRI-owned and -licensed computer program that simulates the operation of light water and heavy water moderated nuclear power plants for both current and advanced light water reactor (ALWR) designs.The report explores the manner in which the in-core instrument tubes would respond during severe core damage events that ...

2013-02-25T23:59:59.000Z

84

On the interactive 3D reconstruction of Iberian vessels  

Science Conference Proceedings (OSTI)

Reconstructing vessels from sherds is a complex task, specially for hand made pottery. That is the case of the Iberian vessels. The reconstruction process can be done in three steps: orientation of the sherd, computing the symmetry axis and detecting ...

F. J. Melero; J. C. Torres; A. León

2003-11-01T23:59:59.000Z

85

Operating an Acoustic Doppler Current Profiler aboard a Container Vessel  

Science Conference Proceedings (OSTI)

Since October 1992 an acoustic Doppler current profiler (ADCP) has been in near-continuous operation on board a 118-m-long container vessel, the container motor vessel Oleander, which operates on a weekly schedule between Port Elizabeth, New ...

C. N. Flagg; G. Schwartze; E. Gottlieb; T. Rossby

1998-02-01T23:59:59.000Z

86

Materials Reliability Program: Reactor Pressure Vessel Integrity Primer (MRP-278)  

Science Conference Proceedings (OSTI)

This primer is based on two earlier Electric Power Research Institute (EPRI) reports: Reactor Vessel Embrittlement Management Handbook: A Handbook for Managing Reactor Vessel Embrittlement and Vessel Integrity (TR-101975-T2) and Primer: Fracture Mechanics in the Nuclear Power Industry (NP-5792-SR, Rev. 1). The information in those earlier reports has been updated extensively and focuses on todays reactor pressure vessel (RPV) embrittlement, integrity, and plant license renewal issues. This RPV integrity ...

2010-06-09T23:59:59.000Z

87

Responses of wintering humpback whales to vessel traffic  

Science Conference Proceedings (OSTI)

Responses of humpback whales to vessel traffic were monitored over two winter seasons during 1983–1984 in Maui

Gordon B. Bauer; Joseph R. Mobley; Louis M. Herman

1993-01-01T23:59:59.000Z

88

Blood vessel segmentation methodologies in retinal images - A survey  

Science Conference Proceedings (OSTI)

Retinal vessel segmentation algorithms are a fundamental component of automatic retinal disease screening systems. This work examines the blood vessel segmentation methodologies in two dimensional retinal images acquired from a fundus camera and a survey ... Keywords: Blood vessel segmentation, Image segmentation, Medical imaging, Retinal images, Retinopathy, Survey

M. M. Fraz; P. Remagnino; A. Hoppe; B. Uyyanonvara; A. R. Rudnicka; C. G. Owen; S. A. Barman

2012-10-01T23:59:59.000Z

89

Investigation of impulsively loaded pressure vessels  

SciTech Connect

Explosion containment vessels for containing from 2,000 to 3,000 five ton nuclear explosions are considered. Analysis methods appear adequate and lowest weights using the most advanced materials available in the next five years are projected.None of these materials can be fabricated today and all require extensive development. Present material technology limits the choice of materials and defines the weight. The addition of safety factors and fixtures (nozzles, etc.) will add to this weight considerably, and may well radically alter the vessel response. Improvements in the strength weight ratios of metals and glasses over those considered in this report do not appear reasonable at this time. Winding schemes to utilize the high strength of steel wires and somehow maintain a reasonable thickness appear to offer the most promise. A `ductile` beryllium would of course offer vast improvement, but no indications that this is being developed have appeared and all presently known beryllium is much too brittle.

Brown, N.; Cornwell, R.; Hanner, D.; Leichter, H.; Mohr, P.

1963-10-15T23:59:59.000Z

90

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

91

Fracture Toughness Characterization of Japanese Reactor Pressure Vessel Steels: Joint EPRI-CRIEPI RPV Embrittlement Studies  

Science Conference Proceedings (OSTI)

EPRI has examined five Japanese reactor pressure vessel steels to characterize the material properties over a complete temperature range, including the brittle/ductile transition region and the upper shelf typical of normal operation. The test results provide the unirradiated baseline needed for evaluating the effects of radiation embrittlement.

1993-07-01T23:59:59.000Z

92

Test Results Using a Bell Jar to Measure Containment Vessel Pressurization  

SciTech Connect

A bell jar is used to determine containment vessel pressurization due to outgassing of plutonium materials. Fifteen food cans containing plutonium bearing materials, including plutonium packaged in direct contact with plastic and plutonium contaminated enriched oxide have been tested to date.

Hensel, S.J.

2002-05-10T23:59:59.000Z

93

Vehicular Storage of Hydrogen in Insulated Pressure Vessels  

DOE Green Energy (OSTI)

This paper describes the development of an alternative technology for storing hydrogen fuel onboard automobiles. Insulated pressure vessels are cryogenic-capable pressure vessels that can accept cryogenic liquid fuel, cryogenic compressed gas or compressed gas at ambient temperature. Insulated pressure vessels offer advantages over conventional H{sub 2} storage approaches. Insulated pressure vessels are more compact and require less carbon fiber than GH{sub 2} vessels. They have lower evaporative losses than LH{sub 2} tanks, and are much lighter than metal hydrides. After outlining the advantages of hydrogen fuel and insulated pressure vessels, the paper describes the experimental and analytical work conducted to verify that insulated pressure vessels can be used safely for vehicular H{sub 2} storage. The paper describes tests that have been conducted to evaluate the safety of insulated pressure vessels. Insulated pressure vessels have successfully completed a series of DOT, ISO and SAE certification tests. A draft procedure for insulated pressure vessel certification has been generated to assist in a future commercialization of this technology. An insulated pressure vessel has been installed in a hydrogen fueled truck and it is currently being subjected to extensive testing.

Aceves, S M; Berry, G D; Martinez-Frias, J; Espinosa-Loza, F

2005-01-03T23:59:59.000Z

94

Confinement Vessel Assay System: Calibration and Certification Report  

SciTech Connect

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

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

2012-07-17T23:59:59.000Z

95

Thermal energy storage using Prestressed Cast Iron Vessels (PCIV). Final report  

DOE Green Energy (OSTI)

The wide-spread application of thermal energy and high-pressure air storage to electric power generation has so far been hampered by the lack of large high-pressure storage vessels of reasonable cost. Welded steel vessels are too expensive for this purpose. However, the Prestressed Cast Iron Vessel (PCIV), developed as a nuclear reactor pressure vessel by Siempelkamp Giesserei KG of Krefeld, FRG, has the potential of complying with these requirements. Applications of the PCIV include: high-pressure air storage for the quick start-up of open cycle gas turbines; pressurized high-temperature sensible heat storage by means of solids with a gaseous heat transfer medium for closed cycle gas turbines of future solar power stations; and pressurized hot water storage for nuclear, solar, or coal-fired steam power plants, employing either separate peaking turbines or overloadable main turbine sets. A reference PCIV of 8000 m/sup 3/, 275/sup 0/C, with hot going walls and cold going tendons was developed, designed, and stress-analysed. A parametric study showed that pressures between 4 and 8 MPa and L/D ratios larger than 4 should be optimal. Cost of the reference vessel is about $10,000,000 or 33 to 50 $/kWh electric energy stored. Cost of peak power will be from 30 to 100 mills/kWh, depending on many parameters.

Gilli, P.V.; Beckmann, G.; Schilling, F.E.

1977-06-01T23:59:59.000Z

96

Progress in understanding the mechanical behavior of pressure-vessel materials at elevated temperatures  

SciTech Connect

Progress during the 1970's on the production of high-temperature mechanical properties data for pressure vessel materials was reviewed. The direction of the research was toward satisfying new data requirements to implement advances in high-temperature inelastic design methods. To meet these needs, servo-controlled testing machines and high-resolution extensometry were developed to gain more information on the essential behavioral features of high-temperature alloys. The similarities and differences in the mechanical response of various pressure vessel materials were identified. High-temperature pressure vessel materials that have received the most attention included Type 304 stainless steel, Type 316 stainless steel, 2 1/4 Cr-1 Mo steel, alloy 800H, and Hastelloy X.

Swindeman, R.W.; Brinkman, C.R.

1981-01-01T23:59:59.000Z

97

Reactor Vessel Internals Inspection and Reactor Pressure Vessel Surveillance Program Summaries for R.E. Ginna and Nine Mile Point Unit 1  

Science Conference Proceedings (OSTI)

This report provides a summary of project activities involving the reactor pressure vessel and internals for the nuclear power plants included in the joint Electric Power Research Institute (EPRI), Department of Energy (DOE), and Constellation Energy Nuclear Group (CENG) Nuclear Plant Life Extension demonstration project.BackgroundThe project focused on continuing operations at two CENG nuclear units that are currently operating in their extended license ...

2013-12-18T23:59:59.000Z

98

Reactor Vessel Head Disposal Campaign for Nuclear Management Company  

SciTech Connect

After establishing a goal to replace as many reactor vessel heads as possible - in the shortest time and at the lowest cost as possible - Nuclear Management Company (NMC) initiated an ambitious program to replace the heads on all six of its pressurized water reactors. Currently, four heads have been replaced; and four old heads have been disposed of. In 2002, NMC began fabricating the first of its replacement reactor vessel heads for the Kewaunee Nuclear Plant. During its fall 2004 refueling outage, Kewaunee's head was replaced and the old head was prepared for disposal. Kewaunee's disposal project included: - Down-ending, - Draining, - Decontamination, - Packaging, - Removal from containment, - On-Site handling, - Temporary storage, - Transportation, - Disposal. The next two replacements took place in the spring of 2005. Point Beach Nuclear Plant (PBNP) Unit 2 and Prairie Island Nuclear Generating Plant (PINGP) Unit 2 completed their head replacements during their scheduled refueling outages. Since these two outages were scheduled so close to each other, their removal and disposal posed some unique challenges. In addition, changes to the handling and disposal programs were made as a result of lessons learned from Kewaunee. A fourth head replacement took place during PBNP Unit 1's refueling outage during the fall of 2005. A number of additional changes took place. All of these changes and challenges are discussed in the paper. NMC's future schedule includes PINGP Unit 1's installation in Spring 2006 and Palisades' installation during 2007. NMC plans to dispose of these two remaining heads in a similar manner. This paper presents a summary of these activities, plus a discussion of lessons learned. (authors)

Hoelscher, H.L.; Closs, J.W. [Nuclear Management Company, LLC, 700 First Street, Hudson, WI 54016 (United States); Johnson, S.A. [Duratek, Inc., 140 Stoneridge Drive, Columbia, SC 29210 (United States)

2006-07-01T23:59:59.000Z

99

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

100

Engineering Test Reactor (ETR) Vessel Relocated after 50 years.  

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

Printer Friendly Printer Friendly Engineering Test Reactor (ETR) Vessel Relocated Engineering Test Reactor Vessel Pre-startup 1957 Click on image to enlarge. Image 1 of 5 Gantry jacks attached to ETR vessel. Initial lift starts. - Click on image to enlarge. Image 2 of 5 ETR vessel removed from substructure. Vessel lifted approximately 40 ft. - Click on image to enlarge. On Monday, September 24, 2007 the Engineering Test Reactor (ETR) vessel was removed from its location and delivered to the Idaho CERCLA Disposal Facility (ICDF). The long history of the ETR began for this water-cooled reactor with its start up in 1957, after taking only 2 years to build. According to "Proving the Principles," by Susan M. Stacy: When the Engineering Test Reactor started up at the Test Reactor Area in

Note: This page contains sample records for the topic "ocean-going vessels including" 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

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

Science Conference Proceedings (OSTI)

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

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

102

International Hydrogen Fuel and Pressure Vessel Forum 2010 Proceedings  

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

challenges in harmonizing test protocols and requirements for compressed natural gas (CNG), hydrogen, and CNG-hydrogen (HCNG) blend pressure vessels and to define next steps for...

103

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.

Slutsky, S; Breuer, H; Dobi, A; Hall, C; Langford, T; Leonard, D; Kaufman, L J; Strickland, V; Voskanian, N

2009-01-01T23:59:59.000Z

104

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

2009-07-13T23:59:59.000Z

105

Reactor Pressure Vessel Task of Light Water Reactor Sustainability...  

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

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment Reactor Pressure...

106

Damage analysis of composite pressure vessels using acoustic emission monitoring.  

E-Print Network (OSTI)

??Composite pressure vessels (CPVs) fabricated using a metal or plastic liner under a composite structural skin are commonly used for natural gas storage on road… (more)

Chou, H

2012-01-01T23:59:59.000Z

107

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

SciTech Connect

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

108

Retinal vessel segmentation using multiwavelet kernels and multiscale hierarchical decomposition  

Science Conference Proceedings (OSTI)

We propose a comprehensive method for segmenting the retinal vasculature in fundus camera images. Our method does not require preprocessing and training and can therefore be used directly on different images sets. We enhance the vessels using matched ... Keywords: Matched filter, Multiscale hierarchical decomposition, Multiwavelet, Retinal images, Segmentation, Vessel detection

Yangfan Wang; Guangrong Ji; Ping Lin; Emanuele Trucco

2013-08-01T23:59:59.000Z

109

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

SciTech Connect

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

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

110

Casting Apparatus Including A Gas Driven Molten Metal Injector And Method  

DOE Patents (OSTI)

The casting apparatus (50) includes a holding vessel (10) for containing a supply of molten metal (12) and a casting mold (52) located above the holding vessel (10) and having a casting cavity (54). A molten metal injector (14) extends into the holding vessel (10) and is at least partially immersed in the molten metal (12) in the holding vessel (10). The molten metal injector (14) is in fluid communication with the casting cavity (54). The molten metal injector (14) has an injector body (16) defining an inlet opening (24) for receiving molten metal into the injector body (16). A gas pressurization source (38) is in fluid communication with the injector body (16) for cyclically pressurizing the injector body (16) and inducing molten metal to flow from the injector body (16) to the casting cavity (54). An inlet valve (42) is located in the inlet opening (24) in the injector body (16) for filling molten metal into the injector body (16). The inlet valve (42) is configured to prevent outflow of molten metal from the injector body (16) during pressurization and permit inflow of molten metal into the injector body (16) after pressurization. The inlet valve (42) has an inlet valve actuator (44) located above the surface of the supply of molten metal (12) and is operatively connected to the inlet valve (42) for operating the inlet valve (42) between open and closed positions.

Meyer, Thomas N. (Murrysville, PA)

2004-06-01T23:59:59.000Z

111

Science Accelerator content now includes multimedia  

Office of Scientific and Technical Information (OSTI)

Science Accelerator content now includes multimedia Science Accelerator has expanded its suite of collections to include ScienceCinema, which contains videos produced by the U.S....

112

AUTHORIZATION OF THE TROJAN REACTOR VESSEL PACKAGE FOR ONE-TIME SHIPMENT FOR DISPOSAL (1)  

E-Print Network (OSTI)

Request Commission approval, by negative consent, for the staff to grant two specific exemptions from package test requirements specified in 10 CFR Part 71 for the Trojan Reactor Vessel Package (TRVP), and to authorize the TRVP for one-time transport for disposal. BACKGROUND: Portland General Electric Company (PGE) has requested approval of the TRVP (including internals) for transport to the disposal facility operated by US

L. Joseph; Callan Executive; Director Operations; John R. Cook

1998-01-01T23:59:59.000Z

113

Materials Reliability Program: Utility Preparation for Nondestructive Evaluation of Reactor Vessel Upper Head Penetrations (MRP-360)  

Science Conference Proceedings (OSTI)

The purpose of this report is to provide nuclear power plant owners with recommendations for planning and executing reactor vessel upper head (RVUH) penetration examinations in a manner that will minimize the occurrence of human errors while maximizing the probability of success. RVUH penetrations include control rod drive mechanism, control element drive mechanism, in-core instrumentation, and vent line penetrations. These encompass the standard nomenclatures for domestic utilities but might not ...

2013-12-19T23:59:59.000Z

114

Welding the AT-400A Containment Vessel  

SciTech Connect

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

Brandon, E.

1998-11-01T23:59:59.000Z

115

Welding the AT-400A Containment Vessel  

SciTech Connect

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

Brandon, E.

1998-11-01T23:59:59.000Z

116

Petroleum Gasoline & Distillate Needs Including the Energy ...  

U.S. Energy Information Administration (EIA)

Home > Petroleum > Analysis > Petroleum Gasoline & Distillate Needs Including the Energy Independence and Security Act (EISA) ...

117

Petroleum Gasoline & Distillate Needs Including the Energy ...  

U.S. Energy Information Administration (EIA)

Petroleum Gasoline & Distillate Needs Including the Energy Independence and Security Act (EISA) Impacts

118

Confinement Vessel Assay System: Design and Implementation Report  

SciTech Connect

Los Alamos National Laboratory has a number of spherical confinement vessels remaining from tests involving nuclear materials. These vessels have an inner diameter of 6 feet with 1- to 2-inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the vessels. We have developed a neutron assay system for the purposes of Materials Control and Accountability (MC&A) measurements of the vessel prior to and after cleanout. We present our approach to confronting the challenges in designing, building, and testing such a system. The system was designed to meet a set of functional and operational requirements. A Monte Carlo model was developed to aid in optimizing the detector design as well as to predict the systematic uncertainty associated with confinement vessel measurements. Initial testing was performed to optimize and determine various measurement parameters, and then the system was characterized using {sup 252}Cf placed a various locations throughout the measurement system. Measurements were also performed with a {sup 252}Cf source placed inside of small steel and HDPE shells to study the effect of moderation. These measurements compare favorably with their MCNPX model equivalent, making us confident that we can rely on the Monte Carlo simulation to predict the systematic uncertainty due to variations in response to material that may be localized at different points within a vessel.

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

2012-07-18T23:59:59.000Z

119

Thermal Expansion Coefficient of Steels Used in LWR Vessels  

Science Conference Proceedings (OSTI)

Because of the impact that melt relocation and vessel failure have on subsequent progression and associated consequences of a Light Water Reactor (LWR) accident, it is important to accurately predict the heat-up and relocation of materials within the reactor vessel and heat transfer to and from the reactor vessel. Accurate predictions of such heat transfer phenomena require high temperature thermal properties. However, a review of vessel and structural steel material properties in severe accident analysis codes reveals that the required high temperature material properties are extrapolated with little, if any, data above 700ºC. To reduce uncertainties in predictions relying upon this extrapolated high temperature data, new thermal expansion data were obtained using pushrod dilatometry techniques for two metals used in LWR vessels: SA 533 Grade B, Class 1 (SA533B1) low alloy steel, which is used to fabricate most US LWR reactor vessels; and Type 304 Stainless Steel (SS304), which is used in LWR vessel piping, penetration tubes, and internal structures. This paper summarizes the new data and compares it to existing, lower temperature data in the literature.

Joshua E. Daw; Joy L. Rempe; Darrell L. Knudson; John C. Crepeau

2008-05-01T23:59:59.000Z

120

High temperature thermal properties for metals used in LWR vessels  

Science Conference Proceedings (OSTI)

Because of the impact that melt relocation and vessel failure has on subsequent progression and associated consequences of an Light Water Reactor (LWR) accident, it is important to accurately predict the heatup and relocation of materials within the reactor vessel and heat transfer to and from the reactor vessel. Accurate predictions of such heat transfer phenomena require high temperature thermal properties. However, a review of vessel and structural steel material properties in severe accident analysis codes reveals that the required high temperature material properties are extrapolated, with little if any, data above 700 ºC. To reduce uncertainties in predictions relying upon this extrapolated high temperature data, INL obtained data using laser-flash thermal diffusivity techniques for two metals used in LWR vessels: SA533B1 carbon steel, which is used to fabricate most US LWR reactor vessels; and SS304, which is used in LWR vessel piping, penetration tubes, and internal structures. This paper summarizes the new data, compares it to existing data in the literature, and provides recommended correlations for thermal properties based on this data.

Joy L. Rempe

2008-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "ocean-going vessels including" 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

An Enhanced In-Vessel Core Catcher for Improving In-Vessel Retention Margins  

SciTech Connect

In-vessel retention (IVR) of core melt that may relocate to the lower head of a reactor vessel is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for several advanced light water reactors. A U.S.-Korean International Nuclear Energy Research Initiative project has been initiated to explore design enhancements that could increase the margin for IVR for advanced reactors with higher power levels [up to 1500 MW(electric)]. As part of this effort, an enhanced in-vessel core catcher is being designed and evaluated. To reduce cost and simplify manufacture and installation, this new core catcher design consists of several interlocking sections that are machined to fit together when inserted into the lower head. If needed, the core catcher can be manufactured with holes to accommodate lower head penetrations. Each section of the core catcher consists of two material layers with an option to add a third layer (if deemed necessary). The first is a base material that has the capability to support and contain the mass of core materials that may relocate during a severe accident; the second is an oxide coating on top of the base material, which resists interactions with high-temperature core materials; and the third is an optional coating on the bottom side of the base material to protect it from oxidation during the lifetime of the reactor. This paper summarizes results from the invessel core catcher design and evaluation efforts, focusing on recently obtained results from materials interaction tests and prototypic testing activities.

Joy L. Rempe

2005-11-01T23:59:59.000Z

122

Pressure vessel reliability as a function of allowable stress  

SciTech Connect

From Winter meeting of American Society of Mechanical Engineers; Detroit, Michigan, USA (11 Nov 1973). The probability of failure corresponding to specified levels of allowable design stress was calculated for pressure vessels designed in accordance with the ASME Boiler and Pressure Vessel Code. The analysis was performed for maximum shear stress failure and for cyclic stress failure. The significance of such failure prediction is ddscussed and a rationale for selecting an allowable stress is presented. Examples are presented that demonstrate the estimation of vessel failure probability as a function of load variation, strength variation, and design safety factor. (auth)

Arnold, H.G.

1972-01-01T23:59:59.000Z

123

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

124

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

125

In-vessel activation monitors in JET: Progress in modeling  

SciTech Connect

Activation studies were performed in JET with new in-vessel activation monitors. Though primarily dedicated to R and D in the challenging issue of lost {alpha} diagnostics for ITER, which is being addressed at JET with several techniques, these monitors provide for both neutron and charged particle fluences. A set of samples with different orientation with respect to the magnetic field is transported inside the torus by means of a manipulator arm (in contrast with the conventional JET activation system with pneumatic transport system). In this case, radionuclides with longer half-life were selected and ultralow background gamma-ray measurements were needed. The irradiation was closer to the plasma and this potentially reduces the neutron scattering problem. This approach could also be of interest for ITER, where the calibration methods have yet to be developed. The MCNP neutron transport model for JET was modified to include the activation probe and so provide calculations to help assess the new data. The neutron induced activity on the samples are well reproduced by the calculations.

Bonheure, Georges [Laboratory for Plasma Physics, Association 'Euratom-Belgian State', Avenue de la Renaissance 30, B-1000 Brussels (Belgium); Lengar, I. [Slovenian Fusion Association, Jozef Stefan Institute, Jamova 39, SI-1000 Ljubljana (Slovenia); Syme, B.; Popovichev, S. [Euratom/UKAEA Association, Culham Science Centre, Abingdon, OX14 3DB Oxon (United Kingdom); Wieslander, Elisabeth; Hult, Mikael; Gasparro, Joeel; Marissens, Gerd [EC-JRC-IRMM, Institute for Reference Materials and Measurements, Retieseweg 111, B-2440 Geel (Belgium); Arnold, Dirk [Physikalisch-Technische Bundesanstalt, 6.1 Radioactivity, Bundesallee 100, D-38116 Braunschweig (Germany); Laubenstein, Matthias [Laboratori Nazionali del Gran Sasso, S.S, 17/bis, km 18-910, I-67010 Assergi (Italy)

2008-10-15T23:59:59.000Z

126

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR licenses are now being extended from 40y to 60y by the U.S. Nuclear Regulatory Commission (NRC) with intentions to extend licenses to 80y and beyond. The RPV materials exhibit varying degrees of sensitivity to irradiation-induced embrittlement (decreased toughness) , as shown in Fig. 1.1, and extending operation from

127

Experiment Hazard Class 5.3 High Pressure Vessels  

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

3 High Pressure Vessels 3 High Pressure Vessels Applicability This hazard classification applies to working with pressure vessels and systems. Other hazard classifications and associated controls may apply to experiments in this hazard class. Experiment Category Experiments involving previously reviewed hazard controls are catergorized as medium risk experiments. Experiments involving new equipment, processes or materials, or modified hazard control schemes are categorized as high risk experiments. Hazard Control Plan Verification Statements Engineered Controls - The establishment of applicable controls in accordance with the (American Society of Mechanical Engineers) ASME Boiler and Pressure Code, ASME B.31 Piping Code and applicable federal, state, and local codes. Verify vessel is stampled with ASME Code Symbol or allowable

128

NETL: News Release - Ocean Research Vessel Returns with Undersea 'Treasure'  

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

23, 2002 23, 2002 Ocean Research Vessel Returns with Undersea 'Treasure' of Methane Hydrates Largest Amount of Marine Hydrate Core Ever Recovered - The R/V JOIDES Resolution - The R/V JOIDES Resolution VICTORIA, BRITISH COLUMBIA - An internationally funded ocean research vessel has returned to port after a two-month expedition off the Oregon coast, bringing with it the largest amount of marine methane hydrate core samples ever recovered for scientific study. The R/V JOIDES Resolution, the world's largest scientific drillship, docked at Victoria, British Columbia earlier this month and began offloading pressure vessels containing methane hydrates recovered 50 miles offshore of Oregon from an area known as Hydrate Ridge. The pressure vessels, each six feet long and four inches in diameter, will

129

Prediction of Vessel Icing for Near-Freezing Sea Temperatures  

Science Conference Proceedings (OSTI)

The operational NOAA categorical vessel icing algorithm is evaluated with regard to advances in understanding of the icing process and forecasting experience. When sea temperatures are <2–3°C above the saltwater freezing point there is the ...

James E. Overland

1990-03-01T23:59:59.000Z

130

Analysis of the Catastrophic Rupture of a Pressure Vessel  

E-Print Network (OSTI)

occurred at a petroleum refinery in Chicago, killing 17 people and causing extensive property damage [1]. NBS was requested by the Occupational Safety and Health Administration (OSHA) to conduct an investigation into the failure of the pressure vessel that eyewitnesses identified as the initial source of the explosion and fire. This vessel was an amine absorber tower used to strip hydrogen sulfide from a process stream of propane and butane. The vessel was 18.8 m tall, 2.6 m in diameter, and constructed from 25 mm thick plates of type ASTM A516 Grade 70 steel. The investigation was complicated by the damage caused by the explosion and fire. The explosive force had been sufficient to propel the upper 14 m of the vessel a distance of 1 km from its original location,

unknown authors

1984-01-01T23:59:59.000Z

131

Experimental investigation of creep behavior of reactor vessel lower head  

SciTech Connect

The objective of the USNRC supported Lower Head Failure (LHF) Experiment Program at Sandia National Laboratories is to experimentally investigate and characterize the failure of the reactor pressure vessel (RPV) lower head due to the thermal and pressure loads of a severe accident. The experimental program is complemented by a modeling program focused on the development of a constitutive formulation for use in standard finite element structure mechanics codes. The problem is of importance because: lower head failure defines the initial conditions of all ex-vessel events; the inability of state-of-the-art models to simulate the result of the TMI-II accident (Stickler, et al. 1993); and TMI-II results suggest the possibility of in-vessel cooling, and creep deformation may be a precursor to water ingression leading to in-vessel cooling.

Chu, T.Y.; Pilch, M.; Bentz, J.H. [Sandia National Labs., Albuquerque, NM (United States); Behbahani, A. [NRC, Washington, DC (United States)

1998-03-01T23:59:59.000Z

132

Vessel-Spanning Bubble Formation in K-Basin Sludge Stored in Large-Diameter Containers  

DOE Green Energy (OSTI)

The K Basin sludge to be retrieved and stored in the large diameter containers (LDCs) contains some fraction of uranium metal that generates hydrogen gas, which introduces potential upset conditions. One postulated upset condition is a rising plug of sludge supported by a hydrogen bubble that is driven into the vent filters at the top of the container. In laboratory testing with actual K Basin sludge, vessel-spanning bubbles that lifted plugs of sludge were observed in 3-inch-diameter graduated cylinders. This report presents a series of analytical assessments performed by the Pacific Northwest National Laboratory to address the potential for the generation of a vessel spanning bubble in the LDCs. The assessments included the development and evaluation of static and dynamic bubble formation models over the projected range of K Basin sludge physical properties. Additionally, the theory of circular plates was extrapolated to examine conditions under which a plug of sludge would collapse and release a spanning bubble.

Terrones, Guillermo; Gauglitz, Phillip A.

2002-03-01T23:59:59.000Z

133

Using SA508/533 for the HTGR Vessel Material  

SciTech Connect

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

134

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

SciTech Connect

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

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

2012-08-29T23:59:59.000Z

135

Initial observations of cavitation-induced erosion of liquid metal spallation target vessels at the Spallation Neutron Source  

Science Conference Proceedings (OSTI)

During operation of the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory the mechanical properties of the AISI 316L target module are altered by high-energy neutron and proton radiation. The interior surfaces of the target vessel are also damaged by cavitation-induced erosion, which results from repetitive rapid heating of the liquid mercury by high-energy proton beam pulses. Until recently no observations of cavitation-induced erosion were possible for conditions prototypical to the SNS. Post irradiation examination (PIE) of the first and second operational SNS targets was performed to gain insight into the radiation-induced changes in mechanical properties of the 316L target material and the extent of cavitation-induced erosion to the target vessel inner surfaces. Observations of cavitation-induced erosion of the first and second operational SNS target modules are presented here, including images of the target vessel interiors and specimens removed from the target beam-entrance regions.

McClintock, David A [ORNL; Riemer, Bernie [ORNL; Ferguson, Phillip D [ORNL; Carroll, Adam J [ORNL; Dayton, Michael J [ORNL

2012-01-01T23:59:59.000Z

136

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

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

Vessel into the U.S. Form LNG Imports by Vessel into the U.S. Form Excel Version of LNG Imports by Vessel into the U.S. Form.xlsx PDF Version of LNG Imports by Vessel into the U.S....

137

Gas storage materials, including hydrogen storage materials  

DOE Patents (OSTI)

A material for the storage and release of gases comprises a plurality of hollow elements, each hollow element comprising a porous wall enclosing an interior cavity, the interior cavity including structures of a solid-state storage material. In particular examples, the storage material is a hydrogen storage material such as a solid state hydride. An improved method for forming such materials includes the solution diffusion of a storage material solution through a porous wall of a hollow element into an interior cavity.

Mohtadi, Rana F; Wicks, George G; Heung, Leung K; Nakamura, Kenji

2013-02-19T23:59:59.000Z

138

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

139

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.

Hubble, David H. (Export, PA); Ulrich, Klaus H. (Duisburg, DE)

1998-01-01T23:59:59.000Z

140

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

Science Conference Proceedings (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

Note: This page contains sample records for the topic "ocean-going vessels including" 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

Intentionally Including - Engaging Minorities in Physics Careers |  

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

Intentionally Including - Engaging Minorities in Physics Careers Intentionally Including - Engaging Minorities in Physics Careers Intentionally Including - Engaging Minorities in Physics Careers April 24, 2013 - 4:37pm Addthis Joining Director Dot Harris (second from left) were Marlene Kaplan, the Deputy Director of Education and director of EPP, National Oceanic and Atmospheric Administration, Claudia Rankins, a Program Officer with the National Science Foundation and Jim Stith, the past Vice-President of the American Institute of Physics Resources. Joining Director Dot Harris (second from left) were Marlene Kaplan, the Deputy Director of Education and director of EPP, National Oceanic and Atmospheric Administration, Claudia Rankins, a Program Officer with the National Science Foundation and Jim Stith, the past Vice-President of the

142

Transmission line including support means with barriers  

DOE Patents (OSTI)

A gas insulated transmission line includes an elongated outer sheath, a plurality of inner conductors disposed within and extending along the outer sheath, and an insulating gas which electrically insulates the inner conductors from the outer sheath. A support insulator insulatably supports the inner conductors within the outer sheath, with the support insulator comprising a main body portion including a plurality of legs extending to the outer sheath, and barrier portions which extend between the legs. The barrier portions have openings therein adjacent the main body portion through which the inner conductors extend.

Cookson, Alan H. (Pittsburgh, PA)

1982-01-01T23:59:59.000Z

143

DISASTER POLICY Including Extreme Emergent Situations (EES)  

E-Print Network (OSTI)

on the ACGME website with information relating to the ACGME response to the disaster. 3. The University-specific Program Requirements. Defined Responsibilities Following the Declaration of a Disaster or Extreme EmergentPage 123 DISASTER POLICY Including Extreme Emergent Situations (EES) The University of Connecticut

Oliver, Douglas L.

144

Forum Agenda: International Hydrogen Fuel and Pressure Vessel Forum  

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

FORUM AGENDA FORUM AGENDA U.S. Department of Energy and Tsinghua University International Hydrogen Fuel and Pressure Vessel Forum Tsinghua University Beijing, PRC September 27 - 29, 2010 The U.S. Department of Energy (DOE) and Tsinghua University in Beijing co-hosted the International Hydrogen Fuel and Pressure Vessel Forum on September 27 - 29, 2010 in Beijing, China. High pressure vessel experts gathered to share lessons learned from CNG and hydrogen vehicle deployments, and to identify R&D needs to aid the global harmonization of regulations, codes and standards to enable the successful deployment of hydrogen and fuel cell technologies. Forum Objectives: * Address and share data and information on specific technical topics discussed at the workshop in

145

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research  

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

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program The Department of Energy's (DOE's) Light Water Reactor Sustainability (LWRS) Program is a five year effort that 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 operation of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging

146

Tritium permeation and wall loading in the TFTR vacuum vessel  

SciTech Connect

The problems of tritium permeation through and loading of the TFTR vacuum vessel wall structural components are considered. A general analytical solution to the time dependent diffusion equation which takes into account the boundary conditions arising from the tritium filling gas as well as the source function associated with implanted energetic charge exchange tritium is presented. Expressions are derived for two quantities of interest: (1) the total amount of tritium leaving the outer surface of a particular vessel component as a function of time, and (2) the amount retained as a function of time. These quantities are evaluated for specific TFTR operating scenarios and outgassing modes. The results are that permeation through the vessel is important only for the bellows during discharge cleaning if the wall temperature rises above approximately 150/sup 0/C. At 250/sup 0/C, after 72 hours of discharge cleaning 195 Ci would be lost.

Cecchi, J.L.

1978-05-01T23:59:59.000Z

147

The Westinghouse Electric Corporation Reactor Vessel Radiation Surveillance Program  

Science Conference Proceedings (OSTI)

Westinghouse recognized that the disruption of the atomic lattice of metals by collision from energetic neutrons could alter the properties of the metals to such an extent that the changes could be of engineering significance. Furthermore, it was recognized that a physical-metallurgical phenomenon such as aging, both thermal and mechanical, also could alter the properties of a metal over its service life. Because of the potential changes in properties, reactor vessel radiation surveillance programs to monitor the effect of neutron radiation and other environmental factors on the reactor vessel materials during operational conditions over the life of the plant were initiated for Westinghouse plants with the insertion of reactor vessel material radiation surveillance capsules into the Yankee Atomic Company's Yankee Rowe plant in 1961.

Mayer, T.R.; Anderson, S.L.; Yanichko, S.E.

1983-01-01T23:59:59.000Z

148

Sterilization of fermentation vessels by ethanol/water mixtures  

DOE Patents (OSTI)

This invention is comprised of 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, C.E.

1991-03-20T23:59:59.000Z

149

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

150

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

151

Fabrication of toroidal composite pressure vessels. Final report  

DOE Green Energy (OSTI)

A method for fabricating composite pressure vessels having toroidal geometry was evaluated. Eight units were fabricated using fibrous graphite material wrapped over a thin-walled aluminum liner. The material was wrapped using a machine designed for wrapping, the graphite material was impregnated with an epoxy resin that was subsequently thermally cured. The units were fabricated using various winding patterns. They were hydrostatically tested to determine their performance. The method of fabrication was demonstrated. However, the improvement in performance to weight ratio over that obtainable by an all metal vessel probably does not justify the extra cost of fabrication.

Dodge, W.G.; Escalona, A.

1996-11-24T23:59:59.000Z

152

DHCVIM: A direct heating containment vessel interactions module  

SciTech Connect

Models for prediction of direct containment heating phenomena as implemented in the DHCVIM computer module are described. The models were designed to treat thermal, chemical and hydrodynamic processes in the three regions of the Sandia National Laboratory Surtsey DCH test facility: the melt generator, cavity and vessel. The fundamental balance equations, along with constitutive relations are described. A combination of Eulerian treatment for the gas phase and Lagrangian treatment for the droplet phase is used in the modeling. Comparisons of calculations and DCH-1 test results are presented. Reasonable agreement is demonstrated for the vessel pressure rise, melt generator pressure decay and particle size distribution.

Ginsberg, T.; Tutu, N.K.

1987-01-01T23:59:59.000Z

153

Simulation of Diffusive Lithium Evaporation Onto the NSTX Vessel Walls  

Science Conference Proceedings (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.

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

2010-12-09T23:59:59.000Z

154

Buildings Included on EMS Reports"  

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

Office of Legacy Management Office of Legacy Management Buildings Included on EMS Reports" "Site","Property Name","Property ID","GSF","Incl. in Water Baseline (CY2007)","Water Baseline (sq. ft.)","Water CY2008 (sq. ft.)","Water CY2009 (sq. ft.)","Water Notes","Incl. in Energy Baseline (CY2003)","Energy Baseline (sq. ft.)","CY2008 Energy (sq. ft.)","CY2009 Energy (sq. ft.)","Energy Notes","Included as Existing Building","CY2008 Existing Building (sq. ft.)","Reason for Building Exclusion" "Column Totals",,"Totals",115139,,10579,10579,22512,,,3183365,26374,115374,,,99476 "Durango, CO, Disposal/Processing Site","STORAGE SHED","DUD-BLDG-STORSHED",100,"no",,,,,"no",,,,"OSF","no",,"Less than 5,000 GSF"

155

Power generation method including membrane separation  

SciTech Connect

A method for generating electric power, such as at, or close to, natural gas fields. The method includes conditioning natural gas containing C.sub.3+ hydrocarbons and/or acid gas by means of a membrane separation step. This step creates a leaner, sweeter, drier gas, which is then used as combustion fuel to run a turbine, which is in turn used for power generation.

Lokhandwala, Kaaeid A. (Union City, CA)

2000-01-01T23:59:59.000Z

156

Electric power monthly, September 1990. [Glossary included  

SciTech Connect

The purpose of this report is to provide energy decision makers with accurate and timely information that may be used in forming various perspectives on electric issues. The power plants considered include coal, petroleum, natural gas, hydroelectric, and nuclear power plants. Data are presented for power generation, fuel consumption, fuel receipts and cost, sales of electricity, and unusual occurrences at power plants. Data are compared at the national, Census division, and state levels. 4 figs., 52 tabs. (CK)

1990-12-17T23:59:59.000Z

157

Nuclear reactor shield including magnesium oxide  

DOE Patents (OSTI)

An improvement in nuclear reactor shielding of a type used in reactor applications involving significant amounts of fast neutron flux, the reactor shielding including means providing structural support, neutron moderator material, neutron absorber material and other components as described below, wherein at least a portion of the neutron moderator material is magnesium in the form of magnesium oxide either alone or in combination with other moderator materials such as graphite and iron.

Rouse, Carl A. (Del Mar, CA); Simnad, Massoud T. (La Jolla, CA)

1981-01-01T23:59:59.000Z

158

BWRVIP-44-A: BWR Vessel and Internals Project: Underwater Weld Repair of Nickel Alloy Reactor Vessel Internals  

Science Conference Proceedings (OSTI)

The Boiling Water Reactor Vessel and Internals Project (BWRVIP), formed in June 1994, is an association of utilities focused exclusively on BWR vessel and internals issues. This report describes work performed to qualify a flux-core welding process for use in repairing reactor internals at a water depth of up to 50 feet. A previous version of this report was published as BWRVIP-44 (EPRI report TR-108708). The current report, BWRVIP-44-A, incorporates changes proposed by the BWRVIP in response to U.S. Nuc...

2006-08-16T23:59:59.000Z

159

FIRE Vacuum Vessel Cost estimate and R&D needs  

E-Print Network (OSTI)

for remote handling mockup to be used for demonstration of : - Transfer cask docking - Divertor handling - FW and will serve as mockup for remote handling facility #12; tile handling / alignment - Recovery operations - Etc. #12;6 June 2001 FIRE Review: Vacuum Vessel

160

Angiotensin II receptors in rabbit renal preglomerular vessels  

SciTech Connect

Controversy exists regarding the specific sites within the renal microcirculation affected by angiotensin II (ANG II). Under some conditions, ANG II can elicit direct vasoconstrictor responses in the preglomerular vessels and efferent arterioles. These experiments were designed to evaluate the binding of {sup 125}I-ANG II in preglomerular vessels. Arcuate and interlobular arteries, with attached proximal segments of afferent arterioles, were microdissected from rabbit renal cortexes. A membrane preparation was obtained from the pooled freshly dissected vessels and utilized in an ANG II radioreceptor assay on the same day. The dissociation of bound ANG II was enhanced in the presence of a nonhydrolyzable analogue of GTP. Linear Scatchard plots were obtained, indicating the presence of a single class of high-affinity binding sites. In conclusion, there is a single class of specific ANG II receptors in preglomerular vessels. The K{sub D} and N are similar, but the binding inhibition potencies of selected ANG analogues differ in renal and extrarenal vascular tissues. Intrarenal vascular receptors also appear to differ from glomerular receptors. Furthermore, these data support the concept that ANG II may affect renal vascular resistance at sites proximal to the distal afferent arterioles.

Brown, G.P.; Venuto, R.C. (State Univ. of New York, Buffalo (USA))

1988-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "ocean-going vessels including" 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

Fillables: everyday vessels as tangible controllers with adjustable haptics  

Science Conference Proceedings (OSTI)

We introduce Fillables: low-cost and ubiquitous everyday vessels that are appropriated as tangible controllers whose haptics are tuned ad-hoc by filling, e.g., with water. We show how Fillables can assist users in video navigation and drawing tasks with ... Keywords: appropriation, everyday objects, tangible user interfaces, ubiquitous computing, up-and-down transformed response (udtr), weber's law

Christian Corsten; Chat Wacharamanotham; Jan Borchers

2013-04-01T23:59:59.000Z

162

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

163

AN IBM 7090 FORTRAN PROGRAM FOR ASME UNFIRED PRESSURE VESSEL DESIGN AND PRELIMINARY COST ESTIMATION  

SciTech Connect

An IBM 7090 FORTRAN program was written for the preliminary design and cost estimation of unfired pressure vessels with or without a jacket. Both vessel and jacket designs conform to the 1959 ASME Boiler and Pressure Vessel Code, Section VIII, Unfired Pressure Vessels. Vessels and jackets from 5 in. pipe through 84 in. o.d. and 1/4 in. through 1 1/2 in. in metal thickness may be designed by this program as written. Total vessel cost is the sum of metal and fabrication costs, each on a weight basis. (auth)

Prince, C.E.; Milford, R.P.

1962-10-17T23:59:59.000Z

164

Thermovoltaic semiconductor device including a plasma filter  

DOE Patents (OSTI)

A thermovoltaic energy conversion device and related method for converting thermal energy into an electrical potential. An interference filter is provided on a semiconductor thermovoltaic cell to pre-filter black body radiation. The semiconductor thermovoltaic cell includes a P/N junction supported on a substrate which converts incident thermal energy below the semiconductor junction band gap into electrical potential. The semiconductor substrate is doped to provide a plasma filter which reflects back energy having a wavelength which is above the band gap and which is ineffectively filtered by the interference filter, through the P/N junction to the source of radiation thereby avoiding parasitic absorption of the unusable portion of the thermal radiation energy.

Baldasaro, Paul F. (Clifton Park, NY)

1999-01-01T23:59:59.000Z

165

Preliminary investigation on the suitablity of using fiber reinforced concrete in the construction of a hazardous waste disposal vessel  

Science Conference Proceedings (OSTI)

There are certain hazardous wastes that must be contained in an extremely secure vessel for transportation and disposal. The vessel, among other things, must be able to withstand relatively large impacts without rupturing. Such containment vessels therefore must be able to absorb substantial amounts of energy during an impact and still perform their function. One of the impacts that the vessel must withstand is a 30-foot fall onto an unyielding surface. For some disposal scenarios it is proposed to encase the waste in a steel enclosure which is to be surrounded by a thick layer of concrete which, in turn, is encased by a relatively thin steel shell. Tests on concrete in compression and flexure, including static, dynamic and impact tests, have shown that low modulus concretes tend to behave in a less brittle manner than higher modulus concretes. Tests also show that fiber reinforced concretes have significantly greater ductility, crack propagation resistance and toughness than conventional concretes. Since it is known that concrete is a reasonably brittle material, it is necessary to do impact tests on sample containment structures consisting of thin-walled metal containers having closed ends which are filled with concrete, grout, or fiber reinforced concrete. This report presents the results of simple tests aimed at observing the behavior of sample containment structures subjected to impacts due to a fall from 30 feet. 8 figs., 4 tabs.

Ramey, M.R.; Daie-e, G.

1988-07-01T23:59:59.000Z

166

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

SciTech Connect

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

167

TR-105696-R16 (BWRVIP-03) Revision 16: BWR Vessel and Internals Project, Reactor Pressure Vessel and Internals Examination Guidelines  

Science Conference Proceedings (OSTI)

This report provides the boiling water reactor (BWR) fleet with inspection options for all of the safety-related vessel internal components, and provides a stable mechanism for documenting the capability of the evolving inspection technology. It is the sole resource for internals inspection information for BWR ...

2013-12-14T23:59:59.000Z

168

Models of Procyon A including seismic constraints  

E-Print Network (OSTI)

Detailed models of Procyon A based on new asteroseismic measurements by Eggenberger et al (2004) have been computed using the Geneva evolution code including shellular rotation and atomic diffusion. By combining all non-asteroseismic observables now available for Procyon A with these seismological data, we find that the observed mean large spacing of 55.5 +- 0.5 uHz favours a mass of 1.497 M_sol for Procyon A. We also determine the following global parameters of Procyon A: an age of t=1.72 +- 0.30 Gyr, an initial helium mass fraction Y_i=0.290 +- 0.010, a nearly solar initial metallicity (Z/X)_i=0.0234 +- 0.0015 and a mixing-length parameter alpha=1.75 +- 0.40. Moreover, we show that the effects of rotation on the inner structure of the star may be revealed by asteroseismic observations if frequencies can be determined with a high precision. Existing seismological data of Procyon A are unfortunately not accurate enough to really test these differences in the input physics of our models.

P. Eggenberger; F. Carrier; F. Bouchy

2005-01-14T23:59:59.000Z

169

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

Science Conference Proceedings (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

170

LNG Exports by Vessel in ISO Containers out of the U.S. Form...  

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

in ISO Containers out of the U.S. Form LNG Exports by Vessel in ISO Containers out of the U.S. Form Excel Version of LNG Exports by Vessel in ISO Container out of the U.S....

171

Mechanical safety subcommittee guideline for design of thin windows for vacuum vessels  

SciTech Connect

This guideline specifies the usage of thin windows for vacuum vessels in terms of their design and application a Fermilab.

Western, J.L.

1993-03-01T23:59:59.000Z

172

Mechanical Safety Subcommittee guideline for design of thin windows for vacuum vessels  

SciTech Connect

This guideline specifies the usage of thin windows for vacuum vessels in terms of their design and application at Fermilab.

Western, J.L.

1991-03-06T23:59:59.000Z

173

Mechanical safety subcommittee guideline for design of thin windows for vacuum vessels. Revised  

SciTech Connect

This guideline specifies the usage of thin windows for vacuum vessels in terms of their design and application a Fermilab.

Western, J.L.

1993-03-01T23:59:59.000Z

174

LNG Exports by Vessel out of the U.S. Form | Department of Energy  

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

out of the U.S. Form LNG Exports by Vessel out of the U.S. Form Excel Version of LNG Exports by Vessel out of the U.S. Form.xlsx PDF Version of LNG Exports by Vessel out of the...

175

Materials Reliability Program: Reactor Pressure Vessel Integrity Training Module (MRP-286)  

Science Conference Proceedings (OSTI)

For many reactor pressure vessels, embrittlement is the primary concern in ensuring continued safe operation. The shutdown of the Yankee Rowe plant, which occurred because of uncertainties related to embrittlement of the vessel, demonstrated the importance of adequately addressing embrittlement issues. Managing embrittlement involves the integration, management, and implementation of diverse technical, regulatory, planning, and economic activities. Reactor vessel embrittlement management is an essential ...

2010-11-22T23:59:59.000Z

176

An approach to localize the retinal blood vessels using bit planes and centerline detection  

Science Conference Proceedings (OSTI)

The change in morphology, diameter, branching pattern or tortuosity of retinal blood vessels is an important indicator of various clinical disorders of the eye and the body. This paper reports an automated method for segmentation of blood vessels in ... Keywords: Bit plane slicing, Blood vessel segmentation, First order derivative of Gaussian, Image segmentation, Mathematical morphology, Medical imaging, Retinal image, Ocular fundus

M. M. Fraz; S. A. Barman; P. Remagnino; A. Hoppe; A. Basit; B. Uyyanonvara; A. R. Rudnicka; C. G. Owen

2012-11-01T23:59:59.000Z

177

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

Initial Assessment of Thermal Annealing Needs and Challenges Initial Assessment of Thermal Annealing Needs and Challenges Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR licenses are now being extended from 40y to 60y by the U.S. Nuclear Regulatory Commission (NRC) with intentions to extend licenses to 80y and beyond. The RPV materials exhibit varying degrees of sensitivity to irradiation-induced embrittlement (decreased toughness) , as shown in Fig. 1.1, and extending operation from 40y to 80y implies a doubling of the neutron exposure for the RPV. Thus,

178

Inexpensive Delivery of Compressed Hydrogen with Advanced Vessel Technology  

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

delivery of compressed hydrogen delivery of compressed hydrogen with advanced vessel technology Gene Berry Andrew Weisberg Salvador M. Aceves Lawrence Livermore National Laboratory (925) 422-0864 saceves@LLNL.GOV DOE and FreedomCar & Fuel Partnership Hydrogen Delivery and On-Board Storage Analysis Workshop Washington, DC January 25, 2006 LLNL is developing innovative concepts for efficient containment of hydrogen in light duty vehicles concepts may offer advantages for hydrogen delivery Conformable containers efficiently use available space in the vehicle. We are pursuing multiple approaches to conformability High Strength insulated pressure vessels extend LH 2 dormancy 10x, eliminate boiloff, and enable efficiencies of flexible refueling (compressed/cryogenic H 2 /(L)H 2 ) The PVT properties of H

179

Lightweight pressure vessels and unitized regenerative fuel cells  

DOE Green Energy (OSTI)

Energy storage systems have been designed using lightweight pressure vessels with unitized regenerative fuel cells (URFCs). The vessels provide a means of storing reactant gases required for URFCs; they use lightweight bladder liners that act as inflatable mandrels for composite overwrap and provide a permeation barrier. URFC systems have been designed for zero emission vehicles (ZEVs); they are cost competitive with primary FC powered vehicles that operate on H/air with capacitors or batteries for power peaking and regenerative braking. URFCs are capable of regenerative braking via electrolysis and power peaking using low volume/low pressure accumulated oxygen for supercharging the power stack. URFC ZEVs can be safely and rapidly (<5 min.) refueled using home electrolysis units. Reversible operation of cell membrane catalyst is feasible without significant degradation. Such systems would have a rechargeable specific energy > 400 Wh/kg.

Mitlitsky, F.; Myers, B.; Weisberg, A.H.

1996-09-06T23:59:59.000Z

180

DEMO Hot Cell and Ex-Vessel Remote Handling  

E-Print Network (OSTI)

In Europe the work on the specification and design of a Demonstration Power Plant (DEMO) is being carried out by EFDA in the Power Plant Physics and Technology (PPP&T) programme. DEMO will take fusion from experimental research into showing the potential for commercial power generation. This paper describes the first steps being taken towards the design of the DEMO Hot Cell. It will show a comparison of the current DEMO in-vessel maintenance concepts from a Hot Cell perspective, describe a proposed ex-vessel transport system, and summarize the facilities that have been identified as required within the Hot Cell, examine current RH technology and discuss the identified critical development issues.

Thomas, Justin; Bachmann, Christian; Harman, Jon

2013-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "ocean-going vessels including" 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

Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database  

SciTech Connect

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

182

IMPACT OF NUCLEAR MATERIAL DISSOLUTION ON VESSEL CORROSION  

Science Conference Proceedings (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

183

BWRVIP-60-A: BWR Vessel and Internals Project, Evaluation of Stress Corrosion Crack Growth in Low Alloy Steel Vessel Materials in th e BWR Environment  

Science Conference Proceedings (OSTI)

The Boiling Water Reactor Vessel and Internals Project (BWRVIP), formed in June 1994, is an association of utilities focused exclusively on boiling water reactor (BWR) vessel and internals materials issues. This report provides a methodology for assessing crack growth in BWR low alloy steel pressure vessels and nozzles. A previous version of this report was published as BWRVIP-60 (TR-108709). This report (BWRVIP-60-A) incorporates the U.S. Nuclear Regulatory Commission (NRC) Safety Evaluation (SE) and ot...

2003-06-09T23:59:59.000Z

184

Design and Evaluation of an Enhanced In-Vessel Core Catcher  

SciTech Connect

An enhanced in-vessel core catcher is being designed and evaluated as part of a joint United States (U.S.) - Korean International Nuclear Engineering Research Initiative (INERI) investigating methods to insure In-Vessel Retention (IVR) of core materials that may relocate under severe accident conditions in advanced reactors. To reduce cost and simplify manufacture and installation, this new core catcher design consists of several interlocking sections that are machined to fit together when inserted into the lower head. If needed, the core catcher can be manufactured with holes to accommodate lower head penetrations. Each section of the core catcher consists of two material layers with an option to add a third layer (if deemed necessary): a base material, which has the capability to support and contain the mass of core materials that may relocate during a severe accident; an oxide coating material on top of the base material, which resists interactions with high-temperature core materials; and an optional coating on the bottom side of the base material to prevent any potential oxidation of the base material during the lifetime of the reactor. This paper summarizes the status of core catcher design and evaluation efforts, including analyses, materials interaction tests, and prototypic testing efforts.

Joy L. Rempe

2004-06-01T23:59:59.000Z

185

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 individual regions of southwestern New Mexico. Ceramic and clay samples (n=288) from 15 sites in the Gila, Mimbres, and Rio Grande valleys composed the data set. Vessel movement was identified at the regional and site level to determine the degree of regional and site-specific interactions. Fifteen sites composed the regional level data set whereas only five sites contributed large enough sample sizes to determine vessel movement between sites at the site level of analysis. The operating hypotheses of the project were: (1) bowls were distributed throughout the Mimbres cultural system more often than jars; (2) vessel movement between sites within a region exceeded vessel movement between regions; (3) the Mimbres manufactured vessels at the village level; and, (4) elites did not control ceramic vessel distribution. Discriminant function analysis was used to identify vessel movement based on INAA data. The statistical results indicated that bowls were more frequently exchanged than jars, a higher number of vessels were moved between sites within the same region than between regions, vessels were manufactured at the village level, and an elite that controlled vessel distribution most likely did not exist in the Mimbres culture. The absence of a controlling elite was inferred from the overall low levels of exchange and the identification of site-specific production locations.

Dahlin, Eleanor Sherlock

2003-01-01T23:59:59.000Z

186

GRR/Section 6-HI-e - Boiler Pressure Vessel Permit | Open Energy  

Open Energy Info (EERE)

GRR/Section 6-HI-e - Boiler Pressure Vessel Permit GRR/Section 6-HI-e - Boiler Pressure Vessel Permit < GRR Jump to: navigation, search GRR-logo.png GEOTHERMAL REGULATORY ROADMAP Roadmap Home Roadmap Help List of Sections Section 6-HI-e - Boiler Pressure Vessel Permit 06HIGBoilerPressureVesselPermit.pdf Click to View Fullscreen Contact Agencies Hawaii Department of Labor and Industrial Relations Occupational Safety and Health Division Regulations & Policies Boiler and Pressure Vessel Regulations Triggers None specified Click "Edit With Form" above to add content 06HIGBoilerPressureVesselPermit.pdf Error creating thumbnail: Page number not in range. Error creating thumbnail: Page number not in range. Error creating thumbnail: Page number not in range. Flowchart Narrative Boiler/Pressure Vessel Permit

187

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

DOE Patents (OSTI)

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

Bowling, Michael Keith (Blackborough Cullompton, GB)

2000-09-12T23:59:59.000Z

188

Water is used for many purposes, includ-ing growing crops, producing copper,  

E-Print Network (OSTI)

WATER USES Water is used for many purposes, includ- ing growing crops, producing copper, generating electricity, watering lawns, keeping clean, drinking and recreation. Bal- ancing the water budget comes down of the water budget. Reducing demand involves re- ducing how much water each person uses, lim- iting the number

189

An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory  

SciTech Connect

Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs.

Boing, L.E.; Henley, D.R. (Argonne National Lab., IL (USA)); Manion, W.J.; Gordon, J.W. (Nuclear Energy Services, Inc., Danbury, CT (USA))

1989-12-01T23:59:59.000Z

190

Remotely Operated Equipment for Post Irradiation Examination of the SNS Target Vessel  

Science Conference Proceedings (OSTI)

The Spallation Neutron Source produces neutrons by accelerating protons into flowing mercury contained inside a stainless steel target vessel. During facility operation the target vessel is degraded by a combination of high-energy neutrons, the proton beam, and cavitation-induced corrosion. The degradation is primarily concentrated at the nose of the target vessel, where the proton beam passes through. Currently, the Spallation Neutron Source has replaced three target vessels and is operating the fourth. To minimize the operational costs of manufacturing and disposing of target vessels, efforts are underway to increase the operational lifetimes of the target vessels by conducting post irradiation examinations of spent vessels. This examination involves remotely removing multiple coupons from the nose of the target vessel using a single piece of equipment, called the Nose Sampling Cutter, installed inside the Spallation Neutron Source s hot cell. The Cutter produces circular coupons approximately 2 inches in diameter using a carbide-tipped hole saw. The nose of the target vessel consists of four layers of material, and the Nose Sampling Cutter is capable of cutting through the layers in a single stroke. This remote operation has been successfully completed twice. In addition to the Nose Sampling Cutter, a large reciprocation saw capable of removing a sizable section of the nose of the target vessel has been constructed and tested, but never implemented. To support this large reciprocation saw other equipment has also been designed. The details of the Nose Sampling Cutter, reciprocation saw, and associated equipment are discussed.

Carroll, Adam J [ORNL; Graves, Van B [ORNL; Dayton, Michael J [ORNL; Riemer, Bernie [ORNL

2011-01-01T23:59:59.000Z

191

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

DOE Green Energy (OSTI)

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

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

1990-09-01T23:59:59.000Z

192

Method for verification of constituents of a process stream just as they go through an inlet of a reaction vessel  

DOE Patents (OSTI)

A method for validating a process stream for the presence or absence of a substance of interest such as a chemical warfare agent; that is, for verifying that a chemical warfare agent is present in an input line for feeding the agent into a reaction vessel for destruction, or, in a facility for producing commercial chemical products, that a constituent of the chemical warfare agent has not been substituted for the proper chemical compound. The method includes the steps of transmitting light through a sensor positioned in the feed line just before the chemical constituent in the input line enters the reaction vessel, measuring an optical spectrum of the chemical constituent from the light beam transmitted through it, and comparing the measured spectrum to a reference spectrum of the chemical agent and preferably also reference spectra of surrogates. A signal is given if the chemical agent is not entering a reaction vessel for destruction, or if a constituent of a chemical agent is added to a feed line in substitution of the proper chemical compound.

Baylor, Lewis C. (North Augusta, SC); Buchanan, Bruce R. (Aiken, SC); O' Rourke, Patrick E. (Martinez, GA)

1995-01-01T23:59:59.000Z

193

Dye laser amplifier including a low turbulence, stagnation-free dye flow configuration  

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 for example 30 gallons/minute, a specifically designed support vessel for containing the dye cell and a screen device for insuring that the dye stream passes into the dye cell in a substantially turbulent free, stagnation-free manner.

Davin, James (Gilroy, CA)

1992-01-01T23:59:59.000Z

194

Conceptual design of an in-vessel core catcher  

Science Conference Proceedings (OSTI)

An enhanced in-vessel core catcher is being designed and evaluated as part of a joint United States (US)–Korean International Nuclear Energy Research Initiative (INERI) investigating methods to insure retention of materials that may relocate to the lower head of a reactor vessel under severe accident conditions in advanced reactors. This enhanced core catcher design consists of several interlocking sections that are machined to fit together when inserted into the lower head. Each section of the core catcher consists of two material layers with an option to add a third layer (if deemed necessary): a base material, which has the capability to support and contain the mass of core materials that may relocate during a severe accident; an insulator coating material on top of the base material, which resists interactions with high-temperature core materials; and an optional coating on the bottom side of the base material to prevent any potential oxidation of the base material during the lifetime of the reactor. This paper summarizes results from thermal, flow, and structural analyses as well as initial scoping materials interaction tests that were completed to support the conceptual design of the core catcher.

Joy L. Rempe; D. L. Knudson; K. G. Condie; K. Y. Suh; F. B.Cheung; S. B. Kim

2004-05-01T23:59:59.000Z

195

Blowdown of hydrocarbons pressure vessel with partial phase separation  

E-Print Network (OSTI)

We propose a model for the simulation of the blowdown of vessels containing two-phase (gas-liquid) hydrocarbon fluids, considering non equilibrium between phases. Two phases may be present either already at the beginning of the blowdown process (for instance in gas-liquid separators) or as the liquid is formed from flashing of the vapor due to the cooling induced by pressure decrease. There is experimental evidence that the assumption of thermodynamic equilibrium is not appropriate, since the two phases show an independent temperature evolution. Thus, due to the greater heat transfer between the liquid phase with the wall, the wall in contact with the liquid experiences a stronger cooling than the wall in contact with the gas, during the blowdown. As a consequence, the vessel should be designed for a lower temperature than if it was supposed to contain vapor only. Our model is based on a compositional approach, and it takes into account internal heat and mass transfer processes, as well as heat transfer with ...

Speranza, Alessandro; 10.1142/9789812701817_0046

2011-01-01T23:59:59.000Z

196

Calculations to estimate the margin to failure in the TMI-2 vessel  

SciTech Connect

As part of the OECD-sponsored Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project (VIP), margin-to-failure calculations for mechanisms having the potential to threaten the integrity of the vessel were performed to improve understanding of events that occurred during the TMI-2 accident. Analyses considered four failure mechanisms: tube rupture, tube ejection, global vessel failure, and localized vessel failure. Calculational input was based on data from the TMI-2 VIP examinations of the vessel steel samples, the instrument tube nozzles, and samples of the hard layer of debris found on the TMI-2 vessel lower head. Sensitivity studies were performed to investigate the uncertainties in key parameters for these analyses.

Stickler, L.A.; Rempe, J.L.; Chavez, S.A.; Thinnes, G.L.; Snow, S.D. [EG& G Idaho, Inc., Idaho Falls, ID (United States); Witt, R.J.; Corradini, M.L.; Kos, J.A. [Univ. of Wisconsin, Madison, WI (United States)

1994-03-01T23:59:59.000Z

197

PRESTRESSING A TWO-LAYER PRESSURE VESSEL BY CONTROLLED YIELDING OF THE INNER LAYER  

SciTech Connect

A method of designing a two-layer pressure vessel is presented wherein contact between the layers is produced by controlled yielding of the inner vessel by internal pressure. The amount of prestress depends upon the dimensions of the vessel, the properties of the material of construction, and the prestressing pressure. The method takes into account the actual stress-strain curve of the material and satisfies the rales of plastic flow with work hardening. (auth)

Schneider, R.W.

1964-04-01T23:59:59.000Z

198

PRESTRESSING A TWO-LAYER PRESSURE VESSEL BY CONTROLLED YIELDING OF THE INNER LAYER  

SciTech Connect

A method is presented for designing a two-layer pressure vessel wherein contact between the layers is produced by controlled yielding of the inner vessel by internal pressure. The amount of prestress depends upon the dimensions of the vessel, the properties of the material of construction, and the prestressing pressure. The method takes into account the actual stress-strain curve of the material and satisfies the rules of plastic flow with work hardening. (auth)

Schneider, R.W.

1964-01-29T23:59:59.000Z

199

Recent United States and International Experiences in Reactor Vessel and Internals Segmentation  

Science Conference Proceedings (OSTI)

The segmentation of reactor vessels and internals is one of the most challenging tasks in nuclear power plant decommissioning. Many experiences, lessons learned, and best practices have been gained through the execution of the first few reactor vessel and internals segmentation projects. The Electric Power Research Institute (EPRI) previously documented the experiences, lessons learned, best practices, and technologies used in decommissioning reactor vessel and internals segmentation projects in the Unit...

2011-11-29T23:59:59.000Z

200

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

DOE Green Energy (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

Note: This page contains sample records for the topic "ocean-going vessels including" from the National Library of EnergyBeta (NLEBeta).
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201

BWRVIP-189: BWR Vessel and Internals Project, Evaluation of RAMA Fluence Methodology Calculational Uncertainty  

Science Conference Proceedings (OSTI)

This report documents the overall calculational uncertainty associated with the application of the Radiation Application Modeling Application (RAMA) Fluence Methodology to BWR reactor pressure vessel fluence evaluations.

2008-07-07T23:59:59.000Z

202

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

203

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

204

Office of Legacy Management Buildings Included on EMS Reports...  

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

Office of Legacy Management Buildings Included on EMS Reports Office of Legacy Management Buildings Included on EMS Reports Office of Legacy Management Buildings Included on EMS...

205

Experiment DTA report for semiscale transparent vessel countercurrent flow tests  

SciTech Connect

Steady state air-water tests were performed as part of the Semiscale Blowdown and Emergency Core Cooling (ECC) Project to investigate downcomer countercurrent flow and downcomer bypass flow phenomena. These tests were performed in a plexiglass representation of the Semiscale pressure vessel which allowed changes to be madein the geometry of the upper annulus and downcomer for the purpose of investigating the sensitivity of downcomer and bypass flow to changes in system geometry. Tests were also performed to investigate the effects of two-phase inlet flows and different initial system pressures on countercurrent and bypass flow. Results for each test are presented in the form of computer printout of the measurements and of a summary of the pertinent calculated flow rates, pressures, and dimensionless volumetric fluxes. Descriptions of the test facility, instrumentation, operating procedures, and test conditions are also presented. An error analysis is presented for selected volumetric flux calculations. 10 references. (auth)

Hanson, D.J.

1975-10-01T23:59:59.000Z

206

Instrumentation of a prestressed concrete containment vessel model  

Science Conference Proceedings (OSTI)

A series of static overpressurization tests of scale models of nuclear containment structures is being conducted by Sandia National Laboratories for the Nuclear Power Engineering Corporation of Japan and the U.S. Nuclear Regulatory Commission. At present, two tests are being planned: a test of a model of a steel containment vessel (SCV) that is representative of an improved, boiling water reactor (BWR) Mark II design; and a test of a model of a prestressed concrete containment vessel (PCCV). This paper discusses plans and the results of a preliminary investigation of the instrumentation of the PCCV model. The instrumentation suite for this model will consist of approximately 2000 channels of data to record displacements, strains in the reinforcing steel, prestressing tendons, concrete, steel liner and liner anchors, as well as pressure and temperature. The instrumentation is being designed to monitor the response of the model during prestressing operations, during Structural Integrity and Integrated Leak Rate testing, and during test to failure of the model. Particular emphasis has been placed on instrumentation of the prestressing system in order to understand the behavior of the prestressing strands at design and beyond design pressure levels. Current plans are to place load cells at both ends of one third of the tendons in addition to placing strain measurement devices along the length of selected tendons. Strain measurements will be made using conventional bonded foil resistance gages and a wire resistance gage, known as a {open_quotes}Tensmeg{close_quotes}{reg_sign} gage, specifically designed for use with seven-wire strand. The results of preliminary tests of both types of gages, in the laboratory and in a simulated model configuration, are reported and plans for instrumentation of the model are discussed.

Hessheimer, M.F.; Rightley, M.J. [Sandia National Labs., Albuquerque, NM (United States); Matsumoto, T. [Nuclear Power Engineering Corp., Tokyo (Japan)] [and others

1995-09-01T23:59:59.000Z

207

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

Science Conference Proceedings (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°C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

J. K. Wright; R. N. Wright

2010-07-01T23:59:59.000Z

208

Comparison of Alternatives to the 2004 Vacuum Vessel Heat Transfer System  

SciTech Connect

A study comparing different alternatives for the Vacuum Vessel Primary Heat Transfer System has been completed. Three alternatives were proposed in a Project Change Request (PCR-190) by relocating the heat exchangers (HXs) from the roof of the Tokamak building to inside the Vacuum Vessel Pressure Suppression System (VVPSS) tank. The study evaluated the three alternatives and recommended modifications to one of them to arrive at a preferred configuration that included relocating the HXs inside the Tokamak building but outside the VVPSS tank as well as including a small safety-rated pump and HX in parallel to the main circulation pump and HX. The Vacuum Vessel (VV) Primary Heat Transfer System (PHTS) removes heat generated in the VV during normal operation (10 MW, pulsed power) as well as the decay heat from the VV itself and from the structures/components attached to the VV (first wall, blanket, and divertor {approx}0.48 MW peak). Therefore, the VV PHTS has two safety functions: (1) contain contaminated cooling water (similar to the other PHTSs) and (2) provide passive cooling during an accident event. The 2004 design of the VV PHTS consists of two independent loops, each loop cooling half of the 18 VV segments with a nominal flow of 475 kg/s of water at about 1.1 MPa and 100 C. The total flow for both loops is 950 kg/s. Both loops are required to remove the heat load during normal plasma operation. During accident conditions, only one loop is needed to remove by natural convection (no pump needed) the decay heat of the complete VV and attached components. The heat is transferred to heat exchanger (HXs) located on top of the roof, outside the Tokamak building. These HXs are air-to-water (A/W) HXs. Three alternatives have been proposed for this cooling system. For a detailed discussion of these alternatives, please refer to Project Change Request, PCR-190 (Ref. 1). A brief introduction is given here. Alternative 1 includes only one main forced circulation loop with a small safety-rated pump in parallel with the main circulation pump. In addition, this alternative has two natural circulation safety loops. Both the safety and main loops supply water to the bottom of the VV with six branch lines and collect the heated water at the top of the vessel through six branches. The distribution headers are located in the lower pipe chase and the collection headers in the upper pipe chase. Each of these loops (one main and two emergency) has a HX mounted in the Vacuum Vessel Pressure Suppression System (VVPSS) tank. The main HX is cooled using either Component Cooling Water System (CCWS) or Chilled Water System (CHWS) water, and the emergency HXs are cooled by natural circulation of the VVPSS water. See Fig. 1 taken from PCR-190. Alternative 2 is exactly the same as Alternative 1 except that there is only one emergency loop and one emergency HX. See Fig. 2 taken from PCR-190. Alternative 3 also has one main forced circulation loop with a small safety-rated pump in parallel with the main circulation pump and one natural circulation safety loop. In this case, both the safety and main loops supply water to the top of the VV with three branch lines and collect the heated water at the top of the vessel through three branches. Here, the distribution header is located in the upper pipe chase as is the collection header. As before, each of these loops has a HX mounted in the VVPSS tank. The main HX is cooled using either CCWS or CHWS water, and the emergency HXs are cooled by natural circulation of the VVPSS water. See Fig. 3 taken from PCR-190. The preferred configuration is developed by selecting specific attributes of the other configurations analyzed and the logic for selecting this configuration is discussed at the end of the document. It is a modification of Alternative 2 that eliminates the separate safety loop, but incorporates a small safety rated HX and pump in parallel with the main HX and pump. It uses 18 inlet and 18 outlet branches (as did the 2004 design) and locates the HXs outside of the VVPSS tank. Tables 1 and 2 examine alt

Yoder Jr, Graydon L [ORNL; Carbajo, Juan J [ORNL; Kim, Seokho H [ORNL

2010-12-01T23:59:59.000Z

209

BWRVIP-34-A: BWR Vessel and Internals Project, Technical Basis for Part Circumference Weld Overlay Repair of Vessel Internal Core Sp ray Piping  

Science Conference Proceedings (OSTI)

The Boiling Water Reactor Vessel and Internals Project (BWRVIP), formed in June 1994, is an association of utilities focused exclusively on boiling water reactor (BWR) vessel and internals issues. This report summarizes the results of the design and analysis activities and the testing programs conducted to provide BWR utilities with a contingency repair option for internal core spray piping for BWR2/6 plants. A previous version of this report was published as BWRVIP-34 (TR-108198). This report (BWRVIP-34...

2008-03-11T23:59:59.000Z

210

Materials Reliability Project: Benchmark Study of Reactor Pressure Vessel Integrity Probabilistic Computational Results Using the Fracture Analysis of Vessels – Oak Ridge (FAVOR) Software Code (MRP-371)  

Science Conference Proceedings (OSTI)

This report reports the results from the Fracture Analysis of Vessels – Oak Ridge (FAVOR) software analysis of three transients that simulated pressurized thermal shock events in pressurized water reactor (PWR) reactor pressure vessels (RPVs). It was determined that software modifications would be required to complete the probabilistic analyses for the wide range of flaw sizes and locations of interest in the study. Consequently, two software revisions were provided by EPRI to enable ...

2013-08-22T23:59:59.000Z

211

Qualification of In-Service Examination of the Yankee Rowe Reactor Pressure Vessel  

Science Conference Proceedings (OSTI)

An effective in-service examination of the reactor pressure vessel was an essential part of the restart program for the Yankee Atomic Power Company plant in Rowe, Massachusetts. This report describes development of an effective examination strategy, demonstration of performance of the examination procedures, and development of data on the distribution of flaws in reactor pressure vessels.

1993-01-01T23:59:59.000Z

212

BWR Vessel and Internals Project: Quantitative Safety Assessment of BWR Reactor Internals (BWRVIP-09)  

Science Conference Proceedings (OSTI)

The Boiling Water Reactor Vessel and Internals Project (BWRVIP), formed in June, 1994, is an association of utilities focused exclusively on BWR vessel and internals issues. This BWRVIP report documents the results of a quantitative safety assessment conducted to evaluate the safety significance of failures of certain BWR internal components.

1997-02-20T23:59:59.000Z

213

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

Science Conference Proceedings (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

214

HIGH TEMPERATURE THERMAL AND STRUCTURAL MATERIAL PROPERTIES FOR METALS USED IN LWR VESSELS  

Science Conference Proceedings (OSTI)

Because of the impact that melt relocation and vessel failure may have on subsequent progression and associated consequences of a Light Water Reactor (LWR) accident, it is important to accurately predict heating and relocation of materials within the reactor vessel, heat transfer to and from the reactor vessel, and the potential for failure of the vessel and structures within it. Accurate predictions of such phenomena require high temperature thermal and structural properties. However, a review of vessel and structural steel material properties used in severe accident analysis codes reveals that the required high temperature material properties are extrapolated with little, if any, data above 1000 K. To reduce uncertainties in predictions relying upon extrapolated high temperature data, Idaho National Laboratory (INL) obtained high data for two metals used in LWR vessels: SA 533 Grade B, Class 1 (SA533B1) low alloy steel, which is used to fabricate most US LWR reactor vessels; and Type 304 Stainless Steel SS304, which is used in LWR vessel piping, penetration tubes, and internal structures. This paper summarizes the new data, and compares it to existing data.

J.L. Rempe; D.L. Knudson; J. E. Daw; J. C. Crepeau

2008-06-01T23:59:59.000Z

215

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

E-Print Network (OSTI)

RISÃ?-M- 2186 INTERPRETATIOM OF STRAIN HBASUREMEMTS ON NUCLEAR PRESSURE VESSELS Svend Ib Andersen Preben Engbzk Abstract. Selected results from strain measurements on 4 nuclear pressure vessels procedure before and after the test as well as a detailed knowledge of the behaviour of the signal from

216

Repair Technology for Degraded Pressure Vessel and Heat Exchanger Shells: RRAC Task 91  

Science Conference Proceedings (OSTI)

The ability to repair pressure vessels and heat exchangers offers utilities significant cost savings compared to replacing these components. This is especially the case if outage time and loss of production are factored into the cost of replacement. This guide provides a review of various current and proposed repair methods that can be used for pressure vessel and heat exchanger applications.

2002-12-02T23:59:59.000Z

217

REACTION OF DOLPHINS TO A SURVEY VESSEL: EFFECTS ON CENSUS DATA  

E-Print Network (OSTI)

REACTION OF DOLPHINS TO A SURVEY VESSEL: EFFECTS ON CENSUS DATA RoGER P. HEWITI'l ABSTRACf A field of a survey vessel prior to their detection by shipboard observers and that the use of a monotonically decreasing detection function is adequate to minimize bias. Aerial and shipboard estimates of school size

218

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

Science Conference Proceedings (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

219

Method of design for vertical oil shale retorting vessels and retorting therewith  

DOE Patents (OSTI)

A method of designing the gas flow parameters of a vertical shaft oil shale retorting vessel involves determining the proportion of gas introduced in the bottom of the vessel and into intermediate levels in the vessel to provide for lateral distribution of gas across the vessel cross section, providing mixing with the uprising gas, and determining the limiting velocity of the gas through each nozzle. The total quantity of gas necessary for oil shale treatment in the vessel may be determined and the proportion to be injected into each level is then determined based on the velocity relation of the orifice velocity and its feeder manifold gas velocity. A limitation is placed on the velocity of gas issuing from an orifice by the nature of the solid being treated, usually physical tests of gas velocity impinging the solid.

Reeves, Adam A. (Rifle, CO)

1978-01-03T23:59:59.000Z

220

Percentage of Total Natural Gas Residential Deliveries included...  

Gasoline and Diesel Fuel Update (EIA)

City Gate Price Residential Price Percentage of Total Residential Deliveries included in Prices Commercial Price Percentage of Total Commercial Deliveries included in Prices...

Note: This page contains sample records for the topic "ocean-going vessels including" 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

Property:Number of Plants included in Capacity Estimate | Open...  

Open Energy Info (EERE)

of Plants included in Capacity Estimate Jump to: navigation, search Property Name Number of Plants included in Capacity Estimate Property Type Number Retrieved from "http:...

222

Segmented instrumentation tube including a locking sleeve for interlocking the segments of the instrumentation tube  

DOE Patents (OSTI)

Segmented instrumentation tube including a locking sleeve for interlocking the segments of the instrumentation tube, so that the threaded ends of the instrumentation tube do not unthread when subjected to vibration, such an instrumentation tube being suitable for use in a nuclear reactor pressure vessel. The instrumentation tube has a first member having a threaded end portion that has a plurality of first holes circumferentially around the outside surface thereof. The instrumentation tube also has a second member having a threaded end portion that has a plurality of second holes circumferentially around the outside surface thereof. The threads of the second member are caused to threadably engage the threads of the first member for defining a threaded joint therebetween. A sleeve having an inside surface surrounds the end portion of the first member and the end portion of the second member and thus surrounds the threaded joint. The sleeve includes a plurality of first projections and second projections that outwardly extend from the inside surface to engage the first holes and the second holes, respectively. The outside surface of the sleeve is crimped or swaged at the locations of the first projections and second projections such that the first projections and the second projections engage their respective holes. In this manner, independent rotation of the first member with respect to the second member is prevented, so that the instrumentation tube will not unthread at its threaded joint.

Obermeyer, Franklin D. (Pensacola, FL)

1993-01-01T23:59:59.000Z

223

Segmented instrumentation tube including a locking sleeve for interlocking the segments of the instrumentation tube  

DOE Patents (OSTI)

Segmented instrumentation tube including a locking sleeve for interlocking the segments of the instrumentation tube, so that the threaded ends of the instrumentation tube do not unthread when subjected to vibration, such an instrumentation tube being suitable for use in a nuclear reactor pressure vessel. The instrumentation tube has a first member having a threaded end portion that has a plurality of first holes circumferentially around the outside surface thereof. The instrumentation tube also has a second member having a threaded end portion that has a plurality of second holes circumferentially around the outside surface thereof. The threads of the second member are caused to threadably engage the threads of the first member for defining a threaded joint there between. A sleeve having an inside surface surrounds the end portion of the first member and the end portion of the second member and thus surrounds the threaded joint. The sleeve includes a plurality of first projections and second projections that outwardly extend from the inside surface to engage the first holes and the second holes, respectively. The outside surface of the sleeve is crimped or swaged at the locations of the first projections and second projections such that the first projections and the second projections engage their respective holes. In this manner, independent rotation of the first member with respect to the second member is prevented, so that the instrumentation tube will not unthread at its threaded joint. 10 figures.

Obermeyer, F.D.

1993-11-16T23:59:59.000Z

224

AN EXPERIMENTAL STUDY OF THE STABILITY OF VESSEL-SPANNING BUBBLES IN CYLINDRICAL & ANNULAR & OBROUND & AND CONICAL CONTAINERS  

SciTech Connect

This report provides a summary of experiments that were performed by Fauske & Associates on the stability of vessel-spanning bubbles. The report by Fauske & Associates, An Experimental Study of the Stability of Vessel-Spanning Bubbles in Cylindrical, Annular, Obround and Conical Containers, is included in Appendix A. Results from the experiments confirm that the gravity yield parameter, Y{sub G}, correctly includes container size and can be applied to full-scale containers to predict the possibility of the formation of a stable vessel spanning bubble. The results also indicate that a vessel spanning bubble will likely form inside the STSC for KE, KW, and Settler sludges if the shear strengths of these sludges exceed 1820, 2080, and 2120 Pa, respectively. A passive mechanism installed in the STSC is effective at disrupting a rising sludge plug and preventing the sludge from plugging the vent filter or being forced out of the container. The Sludge Treatment Project for Engineered Container and Settler Sludge (EC/ST) Disposition Subproject is being conducted in two phases. Phase 1 of the EC/ST Disposition Subproject will retrieve the radioactive sludge currently stored in the K West (KW) Basin into Sludge Transport and Storage Containers (STSCs) and transport the STSCs to T-Plant for interim storage. Phase 2 of the EC/ST Disposition Subproject will retrieve the sludge from interim storage, treat and package sludge for disposal at the Waste Isolation Pilot Plant. The STSC is a cylindrical container; similar to previously used large diameter containers. A STSC (Figure 1) with a diameter of 58 inches will be used to transport KE and KW originating sludge (located in Engineered Containers 210, 220, 240, 250, and 260) to T-Plant. A STSC with an annulus (Figure 2) will be used to transport Settler Tank sludge, located in Engineered Container 230. An obround small canister design was previously considered to retrieve sludge from the basin. The obround design was selected in Small Canister Design Selection, PRC-STP-00052. However, the small canister was not selected for transporting the sludge. The STSC was selected for sludge loading and transport to T-Plant as discussed in Decision Report for Direct Hydraulic Loading of Sludge into Sludge Transport and Storage Containers, PRC-STP-00112. The STSC will be directly loaded with sludge as described in the Preliminary STP Container and Settler Sludge Process System Description and Material Balance, HNF-41051.

DHALIWAL TK

2010-01-28T23:59:59.000Z

225

Method for forming a bladder for fluid storage vessels  

DOE Green Energy (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

226

BWRVIP-239: BWR Vessel and Internals Project, Updated Evaluation of the Integrated Surveillance Program (ISP) Capsule Withdrawal Sch edule  

Science Conference Proceedings (OSTI)

This report evaluates updated reactor pressure vessel and surveillance capsule fluence data for potential impacts on the Boiling Water Reactor Vessel and Internals Project Integrated Surveillance Program (BWRVIP ISP) capsule withdrawal schedule.

2010-07-16T23:59:59.000Z

227

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

228

Natural Gas Deliveries to Commercial Consumers (Including Vehicle ...  

U.S. Energy Information Administration (EIA)

Natural Gas Deliveries to Commercial Consumers (Including Vehicle Fuel through 1996) in Wisconsin (Million Cubic Feet)

229

Comparison of Pressure Vessel Design and Inspection Requirements as Defined by ASME Code and Germany's TRD Code  

Science Conference Proceedings (OSTI)

This report compares the American Society of Mechanical Engineers (ASME) Code with the German TRD Code for pressure vessel engineering, fabrication, inspection, and other pressure vessel processes. The report compares calculations of minimum required wall thickness for pressure vessels such as boiler tubes, pipes, headers, and drums. It also compares material allowable stress values and reviews the major materials permitted by both codes for use in pressure vessel engineering and manufacturing. The repor...

1994-09-22T23:59:59.000Z

230

USING AN ADAPTER TO PERFORM THE CHALFANT-STYLE CONTAINMENT VESSEL PERIODIC MAINTENANCE LEAK RATE TEST  

Science Conference Proceedings (OSTI)

Recently the Packaging Technology and Pressurized Systems (PT&PS) organization at the Savannah River National Laboratory was asked to develop an adapter for performing the leak-rate test of a Chalfant-style containment vessel. The PT&PS organization collaborated with designers at the Department of Energy's Pantex Plant to develop the adapter currently in use for performing the leak-rate testing on the containment vessels. This paper will give the history of leak-rate testing of the Chalfant-style containment vessels, discuss the design concept for the adapter, give an overview of the design, and will present results of the testing done using the adapter.

Loftin, B.; Abramczyk, G.; Trapp, D.

2011-06-03T23:59:59.000Z

231

Energetics of a Symmetric Circulation Including Momentum Constraints  

Science Conference Proceedings (OSTI)

A theory of available potential energy (APE) for symmetric circulations, which includes momentum constraints, is presented. The theory is a generalization of the classical theory of APE, which includes only thermal constraints on the circulation. ...

Sorin Codoban; Theodore G. Shepherd

2003-08-01T23:59:59.000Z

232

Guam Refinery Thermal Cracking/Other (including Gas Oil ...  

U.S. Energy Information Administration (EIA)

Guam Refinery Thermal Cracking/Other (including Gas Oil) Downstream Charge Capacity as of January 1 (Barrels per Stream Day)

233

Scheduling optimization of a real flexible job shop including side ...  

E-Print Network (OSTI)

Aug 19, 2013 ... including side constraints regarding preventive maintenance, fixture availabil- ...... Engineering and Engineering Management, pp. 787–791.

234

Review of reactor pressure vessel evaluation report for Yankee Rowe Nuclear Power Station (YAEC No. 1735)  

SciTech Connect

The Yankee Atomic Electric Company has performed an Integrated Pressurized Thermal Shock (IPTS)-type evaluation of the Yankee Rowe reactor pressure vessel in accordance with the PTS Rule (10 CFR 50. 61) and a US Regulatory Guide 1.154. The Oak Ridge National Laboratory (ORNL) reviewed the YAEC document and performed an independent probabilistic fracture-mechnics analysis. The review included a comparison of the Pacific Northwest Laboratory (PNL) and the ORNL probabilistic fracture-mechanics codes (VISA-II and OCA-P, respectively). The review identified minor errors and one significant difference in philosophy. Also, the two codes have a few dissimilar peripheral features. Aside from these differences, VISA-II and OCA-P are very similar and with errors corrected and when adjusted for the difference in the treatment of fracture toughness distribution through the wall, yield essentially the same value of the conditional probability of failure. The ORNL independent evaluation indicated RT{sub NDT} values considerably greater than those corresponding to the PTS-Rule screening criteria and a frequency of failure substantially greater than that corresponding to the ``primary acceptance criterion`` in US Regulatory Guide 1.154. Time constraints, however, prevented as rigorous a treatment as the situation deserves. Thus, these results are very preliminary.

Cheverton, R.D.; Dickson, T.L.; Merkle, J.G.; Nanstad, R.K. [Oak Ridge National Lab., TN (United States)

1992-03-01T23:59:59.000Z

235

Review of reactor pressure vessel evaluation report for Yankee Rowe Nuclear Power Station (YAEC No. 1735)  

Science Conference Proceedings (OSTI)

The Yankee Atomic Electric Company has performed an Integrated Pressurized Thermal Shock (IPTS)-type evaluation of the Yankee Rowe reactor pressure vessel in accordance with the PTS Rule (10 CFR 50. 61) and a US Regulatory Guide 1.154. The Oak Ridge National Laboratory (ORNL) reviewed the YAEC document and performed an independent probabilistic fracture-mechnics analysis. The review included a comparison of the Pacific Northwest Laboratory (PNL) and the ORNL probabilistic fracture-mechanics codes (VISA-II and OCA-P, respectively). The review identified minor errors and one significant difference in philosophy. Also, the two codes have a few dissimilar peripheral features. Aside from these differences, VISA-II and OCA-P are very similar and with errors corrected and when adjusted for the difference in the treatment of fracture toughness distribution through the wall, yield essentially the same value of the conditional probability of failure. The ORNL independent evaluation indicated RT{sub NDT} values considerably greater than those corresponding to the PTS-Rule screening criteria and a frequency of failure substantially greater than that corresponding to the primary acceptance criterion'' in US Regulatory Guide 1.154. Time constraints, however, prevented as rigorous a treatment as the situation deserves. Thus, these results are very preliminary.

Cheverton, R.D.; Dickson, T.L.; Merkle, J.G.; Nanstad, R.K. (Oak Ridge National Lab., TN (United States))

1992-03-01T23:59:59.000Z

236

Study of the Neutron Flux and Dpa Attenuation in the Reactor Pressure-Vessel Wall  

Science Conference Proceedings (OSTI)

The study of the neutron flux and dpa attenuation in the reactor pressure vessel (PV) wall presented in this work was performed with state-of-the art methods currently used to determine PV fluxes, the BUGLE-96 cross-section library, and the iron displacement cross sections derived from ENDF/B-VI data. The calculations showed that the RG 1.99, Rev. 2, extrapolation formula predicts slower--and therefore conservative--attenuation of the neutron flux (E > 1MeV) in the PV wall. More importantly, the calculations gave slower attenuation of the dpa rate in the PV wall than the attenuation predicted by the formula. The slower dpa rate attenuation was observed for all the cases considered, which included two different PWRs, and several configurations obtained by varying the PV wall thickness and thermal shield thickness. For example, for a PV wall thickness of {approximately}24 cm, the calculated ratio of the dpa rate at 1/4 and 3/4 of the PV wall thickness to the dpa value on the inner PV surface is {approximately}14% and 19% higher, respectively, than predicted by the RG 1.99, Rev. 2, formula.

Remec, I.

1999-06-01T23:59:59.000Z

237

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

238

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

239

A Small, Low Flow, High Sensitivity Reaction Vessel for NO Chemiluminescence Detectors  

Science Conference Proceedings (OSTI)

Details of a reaction vessel suitable for atmospheric measurements of nitric oxide at parts per trillion mixing ratios using the chemiluminescent reaction with ozone are provided. It is designed to operate at an ambient air flow of 1 standard ...

B. A. Ridley; F. E. Grahek

1990-04-01T23:59:59.000Z

240

CO/sub 2/ welding used to attach inspection manway to NASA hydrogen pressure vessel  

SciTech Connect

Welding of inspection manway for internal survey of a gaseous hydrogen storage vessel is described. Pre-welding activities are reviewed, along with welding operations, and in-process welding control. (JRD)

Palmer, G.; Conklin, D.

1976-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "ocean-going vessels including" 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

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

E-Print Network (OSTI)

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

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

2012-01-01T23:59:59.000Z

242

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

E-Print Network (OSTI)

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

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

2012-03-27T23:59:59.000Z

243

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

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

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...

244

In-Situ Repairs of Oil Industry Pipelines, Tanks and Vessels by ...  

Science Conference Proceedings (OSTI)

Abstract Scope, Metal arc welding under oil (MAW-UO) is a new, revolutionary process to repair a pipeline, tank or vessel by welding in case of flaws and ...

245

Experimental and Numerical Study of the Turbulence Characteristics of Airflow around a Research Vessel  

Science Conference Proceedings (OSTI)

Airflow distortion by research vessels has been shown to significantly affect micrometeorological measurements. This study uses an efficient time-dependent large-eddy simulation numerical technique to investigate the effect of the R/V Tangaroa on ...

Stéphane Popinet; Murray Smith; Craig Stevens

2004-10-01T23:59:59.000Z

246

Draft report: application of organic Rankine cycle heat recovery systems to diesel powered marine vessels  

DOE Green Energy (OSTI)

The analysis and results of an investigation of the application of organic Rankine cycle heat recovery systems to diesel-powered marine vessels are described. The program under which this study was conducted was sponsored jointly by the US Energy Research and Development Administration, the US Navy, and the US Maritime Administration. The overall objective of this study was to investigate diesel bottoming energy recovery systems, currently under development by three US concerns, to determine the potential for application to marine diesel propulsion and auxiliary systems. The study primarily focused on identifying the most promising vessel applications (considering vessel type, size, population density, operational duty cycle, etc.) so the relative economic and fuel conservation merits of energy recovery systems could be determined and assessed. Vessels in the current fleet and the projected 1985 fleet rated at 1000 BHP class and above were investigated.

Not Available

1977-07-15T23:59:59.000Z

247

An accurate model for seaworthy container vessel stowage planning with ballast tanks  

Science Conference Proceedings (OSTI)

Seaworthy container vessel stowage plans generated under realistic assumptions are a key factor for stowage decision support systems in the shipping industry. We propose a linear model with ballast tanks for generating master plans, the first phase of ...

Dario Pacino; Alberto Delgado; Rune Møller Jensen; Tom Bebbington

2012-09-01T23:59:59.000Z

248

Pressure vessel and piping codes. Technical basis for revised reference crack growth rate curves for pressure boundary steels in LWR environment  

SciTech Connect

Since the inception of the pressure vessel and piping codes the reference fatigue crack growth rate curves have been contained in Appendix A of Sect. XI. The curves have been designed to be applicable to carbon and low alloy pressure vessel steels exposed to either air or light water reactor coolant environments. Data obtained over the past several years have shown a different behavior of these steels in the light water reactor environment than that predicted by the present reference curve. A revised set of reference curves has been formulated, incorporating a new curve shape as well as a dependency of growth rate on R ratio (minimum load/maximum load). This work provides the background and justification for such a revision, details the methodology used to develop the revised curves, and includes an evalution of the adequacy and impact of the revised curves as compared with the single curve which they replace. 24 references.

Bamford, W.H.

1980-11-01T23:59:59.000Z

249

BWRVIP-160: BWR Vessel and Internals Project, BWRVIP Inspection Trends, 2006 Update  

Science Conference Proceedings (OSTI)

The BWR Vessel and Internals Project (BWRVIP) is an association of utilities focused exclusively on boiling water reactor (BWR) vessel and internals issues. As a group, the utilities have developed a set of guidelines that recommend inspecting specific BWR internal components. Results of these inspections are reported to EPRI where they are compiled and made available to all member utilities. This report documents results of inspections performed between approximately 1996 and early 2006.

2006-10-03T23:59:59.000Z

250

BWRVIP-198: BWR Vessel and Internals Project, BWRVIP Inspection Trends, 2008 Update  

Science Conference Proceedings (OSTI)

The BWR Vessel and Internals Project (BWRVIP) is an association of utilities focused exclusively on boiling water reactor (BWR) vessel and internals issues. As a group, the utilities have developed a set of guidelines that recommend inspecting specific BWR internal components. Results of these inspections are reported to EPRI where they are compiled and made available to all member utilities. This report documents results of inspections performed between approximately 1996 and spring of 2008.

2008-10-16T23:59:59.000Z

251

BWRVIP-242: BWR Vessel and Internals Project, BWRVIP Inspection Trends, 2010 Update  

Science Conference Proceedings (OSTI)

The BWR Vessel and Internals Project (BWRVIP) is an association of utilities focused exclusively on boiling water reactor (BWR) vessel and internals issues. As a group, the utilities have developed a set of guidelines that recommend inspecting specific BWR internal components. Results of these inspections are reported to EPRI where they are compiled and made available to all member utilities. This report documents results of inspections performed between approximately 1996 and spring of 2010.

2010-12-07T23:59:59.000Z

252

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

SciTech Connect

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

253

Guidelines for Managing Reactor Vessel Material Uncertainties: Part 1: General Approach Part 2: Implementation Guide  

Science Conference Proceedings (OSTI)

Uncertainties about reactor vessel material toughness properties can be a concern for utilities when characterizing vessel integrity. In addition, recent emphasis on variability in material chemistry and initial toughness properties has added to regulatory concerns. This two-part guidelines document provides a general approach (Part 1) for dealing with weld metal property variability and material uncertainties and demonstrates examples of different approaches (Part 2) for dealing with these uncertainties...

1997-04-30T23:59:59.000Z

254

BWRVIP-167: BWR Vessel and Internals Project, Boiling Water Reactor Issue Management Tables  

Science Conference Proceedings (OSTI)

Ongoing issues related to degradation of boiling water reactor (BWR) pressure vessels, reactor internals, and American Society of Mechanical Engineers (ASME) Class 1 piping components have resulted in the need for a summary tool to assist in prioritizing and addressing research and development (R&D) issues. This BWR Vessel and Internals Project (BWRVIP) report provides BWR Issue Management Tables that identify, rank, and describe R&D gaps.

2007-03-20T23:59:59.000Z

255

BWRVIP-167NP, Revision 2: BWR Vessel and Internals Project, Boiling Water Reactor Issue Management Tables  

Science Conference Proceedings (OSTI)

Nuclear utilities face numerous ongoing issues related to degradation of boiling water reactor (BWR) pressure vessels, reactor internals, and American Society of Mechanical Engineers (ASME) Class 1 piping components. These issues have resulted in the need for a summary tool to assist in prioritizing and addressing research and development (R&D) issues and BWR Vessel and Internals Project (BWRVIP) requirements. The BWR Issue Management Tables (IMTs) in the report are living documents that summarize the st...

2010-08-24T23:59:59.000Z

256

BWRVIP-267: BWR Vessel and Internals Project, BWRVIP Inspection Trends, 2012 Update  

Science Conference Proceedings (OSTI)

The BWR Vessel and Internals Project (BWRVIP) is an association of utilities focused exclusively on boiling water reactor (BWR) vessel and internals issues. As a group, the utilities have developed a set of guidelines that recommend inspecting specific BWR internal components. Results of these inspections are reported to the Electric Power Research Institute (EPRI) where they are compiled and made available to all member utilities. This report documents results of inspections performed between ...

2012-10-10T23:59:59.000Z

257

Welding and Repair Technology Center: Repair Technology for Degraded Pressure Vessel and Heat Exchanger Shells  

Science Conference Proceedings (OSTI)

BackgroundPressure vessels and heat exchangers are subject to a number of degradation mechanisms that can cause thinning of component walls and deterioration of internal components. With many repair options available, the Electric Power Research Institute (EPRI) Welding and Repair Technology Center (WRTC) has developed this report to assist operations and engineering personnel who are faced with defective or failed vessel components. Many available repair options allow ...

2013-10-30T23:59:59.000Z

258

Percentage of Total Natural Gas Industrial Deliveries included...  

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

Pipeline and Distribution Use Price City Gate Price Residential Price Percentage of Total Residential Deliveries included in Prices Commercial Price Percentage of Total Commercial...

259

Natural Gas Delivered to Consumers in California (Including ...  

U.S. Energy Information Administration (EIA)

Natural Gas Delivered to Consumers in California (Including Vehicle Fuel) (Million Cubic Feet) Year Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec; ...

260

FAQ 23-How much depleted uranium -- including depleted uranium...  

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

is stored in the United States? How much depleted uranium -- including depleted uranium hexafluoride -- is stored in the United States? In addition to the depleted uranium stored...

Note: This page contains sample records for the topic "ocean-going vessels including" 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

Electrical machines and assemblies including a yokeless stator ...  

Wind Energy; Partners (27) Visual Patent Search; Success Stories; News; Events; Electrical machines and assemblies including a yokeless stator with modular lamination ...

262

U.S. Refinery Thermal Cracking, Other (including Gas Oil ...  

U.S. Energy Information Administration (EIA)

U.S. Refinery Thermal Cracking, Other (including Gas Oil) Downstream Charge Capacity as of January 1 (Barrels per Stream Day)

263

[Article 1 of 7: Motivates and Includes the Consumer  

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

will be diverse and widespread, including renewables, distributed generation, and energy storage. And they will increase rapidly all along the value chain, from suppliers to...

264

Stocks of Total Crude Oil and Petroleum Products (Including SPR)  

U.S. Energy Information Administration (EIA)

-No Data Reported; --= Not Applicable; NA = Not Available; W = Withheld to avoid disclosure of individual company data. Notes: Stocks include those ...

265

Including Retro-Commissioning in Federal Energy Savings Performance...  

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

11.2 Retro-Cx in Federal ESPCs Including Retro-Commissioning In Federal Energy Saving Performance Contracts Retro-commissioning generally reduces operating and maintenance costs,...

266

PLOT: A UNIX PROGRAM FOR INCLUDING GRAPHICS IN DOCUMENTS  

E-Print Network (OSTI)

simple, easy-to-read graphics language designed specificallyPROGRAM FOR INCLUDING GRAPHICS IN DOCUMENTS Pavel Curtismeanings as in the GRAFPAC graphics system. Definl. ~ tions

Curtis, Pavel

2013-01-01T23:59:59.000Z

267

Natural Gas Deliveries to Commercial Consumers (Including Vehicle...  

Gasoline and Diesel Fuel Update (EIA)

Natural Gas Deliveries to Commercial Consumers (Including Vehicle Fuel through 1996) in South Dakota (Million Cubic Feet) Natural Gas Deliveries to Commercial Consumers...

268

Natural Gas Delivered to Consumers in South Dakota (Including...  

Gasoline and Diesel Fuel Update (EIA)

History: Monthly Annual Download Data (XLS File) Natural Gas Delivered to Consumers in South Dakota (Including Vehicle Fuel) (Million Cubic Feet) Natural Gas Delivered to...

269

Solar Energy Education. Reader, Part II. Sun story. [Includes glossary  

DOE Green Energy (OSTI)

Magazine articles which focus on the subject of solar energy are presented. The booklet prepared is the second of a four part series of the Solar Energy Reader. Excerpts from the magazines include the history of solar energy, mythology and tales, and selected poetry on the sun. A glossary of energy related terms is included. (BCS)

Not Available

1981-05-01T23:59:59.000Z

270

Energy Transitions: A Systems Approach Including Marcellus Shale Gas Development  

E-Print Network (OSTI)

Energy Transitions: A Systems Approach Including Marcellus Shale Gas Development A Report: A Systems Approach Including Marcellus Shale Gas Development Executive Summary In the 21st century new we focused on the case of un- conventional natural gas recovery from the Marcellus shale In addition

Walter, M.Todd

271

Energy Transitions: A Systems Approach Including Marcellus Shale Gas Development  

E-Print Network (OSTI)

Energy Transitions: A Systems Approach Including Marcellus Shale Gas Development A Report Transitions: A Systems Approach Including Marcellus Shale Gas Development Executive Summary In the 21st the Marcellus shale In addition to the specific questions identified for the case of Marcellus shale gas in New

Angenent, Lars T.

272

What To Include In The Whistleblower Complaint? | National Nuclear Security  

National Nuclear Security Administration (NNSA)

To Include In The Whistleblower Complaint? | National Nuclear Security To Include In The Whistleblower Complaint? | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog What To Include In The Whistleblower Complaint? Home > About Us > Our Operations > Management and Budget > Whistleblower Program > What To Include In The Whistleblower Complaint? What To Include In The Whistleblower Complaint?

273

In-Vessel Retention of Molten Core Debris in the Westinghouse AP1000 Advanced Passive PWR  

SciTech Connect

In-vessel retention (IVR) of molten core debris via external reactor vessel cooling is the hallmark of the severe accident management strategies in the AP600 passive PWR. The vessel is submerged in water to cool its external surface via nucleate boiling heat transfer. An engineered flow path through the reactor vessel insulation provides cooling water to the vessel surface and vents steam to promote IVR. For the 600 MWe passive plant, the predicted heat load from molten debris to the lower head wall has a large margin to the critical heat flux on the external surface of the vessel, which is the upper limit of the cooling capability. Up-rating the power of the passive plant from 600 to 1000 MWe (AP1000) significantly increases the heat loading from the molten debris to the reactor vessel lower head in the postulated bounding severe accident sequence. To maintain a large margin to the coolability limit for the AP1000, design features and severe accident management (SAM) strategies to increase the critical heat flux on the external surface of the vessel wall need to be implemented. A test program at the ULPU facility at University of California Santa Barbara (UCSB) has been initiated to investigate design features and SAM strategies that can enhance the critical heat flux. Results from ULPU Configuration IV demonstrate that with small changes to the ex-vessel design and SAM strategies, the peak critical heat flux in the AP1000 can be increased at least 30% over the peak critical heat flux predicted for the AP600 configuration. The design and SAM strategy changes investigated in ULPU Configuration IV can be implemented in the AP1000 design and will allow the passive plant to maintain the margin to critical heat flux for IVR, even at the higher power level. Continued testing for IVR phenomena is being performed at UCSB to optimize the AP1000 design and to ensure that vessel failure in a severe accident is physically unreasonable. (authors)

Scobel, James H.; Conway, L.E. [Westinghouse Electric Company LLC, P.O. Box 355, Pittsburgh, PA 15230-0355 (United States); Theofanous, T.G. [Center for Risk Studies and Safety, University of California Santa Barbara (United States)

2002-07-01T23:59:59.000Z

274

U-182: Microsoft Windows Includes Some Invalid Certificates | Department of  

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

82: Microsoft Windows Includes Some Invalid Certificates 82: Microsoft Windows Includes Some Invalid Certificates U-182: Microsoft Windows Includes Some Invalid Certificates June 4, 2012 - 7:00am Addthis PROBLEM: A vulnerability was reported in Microsoft Windows. A remote user may be able to spoof code signing signatures. PLATFORM: Version(s): XP SP3, 2003 SP2, Vista SP2, 2008 SP2, 7 SP1, 2008 R2 SP1; and prior service packs ABSTRACT: The operating system includes some invalid intermediate certificates. The vulnerability is due to the certificate authorities and not the operating system itself. Reference Links: Security tracker ID 1027114 GENERIC-MAP-NOMATCH Vendor Advisory IMPACT ASSESSMENT: High Discussion: The invalid certificates and their thumbprints are: Microsoft Enforced Licensing Intermediate PCA: 2a 83 e9 02 05 91 a5 5f c6

275

Removal of mineral matter including pyrite from coal  

SciTech Connect

Mineral matter, including pyrite, is removed from coal by treatment of the coal with aqueous alkali at a temperature of about 175.degree. to 350.degree. C, followed by acidification with strong acid.

Reggel, Leslie (Pittsburgh, PA); Raymond, Raphael (Bethel Park, PA); Blaustein, Bernard D. (Pittsburgh, PA)

1976-11-23T23:59:59.000Z

276

Free Energy Efficiency Kit includes CFL light bulbs,  

E-Print Network (OSTI)

Free Energy Efficiency Kit Kit includes CFL light bulbs, spray foam, low-flow shower head, and more i ci e n cy On Thursday, March 31st New River Light & Power will sponsor a seminar that is designed

Rose, Annkatrin

277

Characterizations of Aircraft Icing Environments that Include Supercooled Large Drops  

Science Conference Proceedings (OSTI)

Measurements of aircraft icing environments that include supercooled large drops (SLD) greater than 50 ?m in diameter have been made during 38 research flights. These flights were conducted during the First and Third Canadian Freezing Drizzle ...

Stewart G. Cober; George A. Isaac; J. Walter Strapp

2001-11-01T23:59:59.000Z

278

Including costs of supply chain risk in strategic sourcing decisions  

E-Print Network (OSTI)

Cost evaluations do not always include the costs associated with risks when organizations make strategic sourcing decisions. This research was conducted to establish and quantify the impact of risks and risk-related costs ...

Jain, Avani

2009-01-01T23:59:59.000Z

279

Including Atmospheric Layers in Vegetation and Urban Offline Surface Schemes  

Science Conference Proceedings (OSTI)

A formulation to include prognostic atmospheric layers in offline surface schemes is derived from atmospheric equations. Whereas multilayer schemes developed previously need a complex coupling between atmospheric-model levels and surface-scheme ...

Valéry Masson; Yann Seity

2009-07-01T23:59:59.000Z

280

Development of Improved Composite Pressure Vessels for Hydrogen Storage - DOE Hydrogen and Fuel Cells Program FY 2012 Annual Progress Report  

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

0 0 DOE Hydrogen and Fuel Cells Program FY 2012 Annual Progress Report Norman Newhouse (Primary Contact), Jon Knudsen, John Makinson Lincoln Composites, Inc. 5117 NW 40 th Street Lincoln, NE 68524 Phone: (402) 470-5035 Email: nnewhouse@lincolncomposites.com DOE Managers HQ: Ned Stetson Phone: (202) 586-9995 Email: Ned.Stetson@ee.doe.gov GO: Jesse Adams Phone: (720) 356-1421 Email: Jesse.Adams@go.doe.gov Contract Number: DE-FC36-09GO19004 Project Start Date: February 1, 2009 Project End Date: June 30, 2014 Fiscal Year (FY) 2012 Objectives Improve the performance characteristics, including * weight, volumetric efficiency, and cost, of composite pressure vessels used to contain hydrogen in adsorbants. Evaluate design, materials, or manufacturing process *

Note: This page contains sample records for the topic "ocean-going vessels including" 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

Percentage of Total Natural Gas Commercial Deliveries included in Prices  

Gasoline and Diesel Fuel Update (EIA)

City Gate Price Residential Price Percentage of Total Residential Deliveries included in Prices Commercial Price Percentage of Total Commercial Deliveries included in Prices Industrial Price Percentage of Total Industrial Deliveries included in Prices Electric Power Price Period: Monthly Annual City Gate Price Residential Price Percentage of Total Residential Deliveries included in Prices Commercial Price Percentage of Total Commercial Deliveries included in Prices Industrial Price Percentage of Total Industrial Deliveries included in Prices Electric Power Price Period: Monthly Annual Download Series History Download Series History Definitions, Sources & Notes Definitions, Sources & Notes Show Data By: Data Series Area May-13 Jun-13 Jul-13 Aug-13 Sep-13 Oct-13 View History U.S. 63.3 59.3 57.9 57.0 57.4 61.3 1983-2013 Alabama 71.7 71.0 68.5 68.2 68.4 66.7 1989-2013 Alaska 94.1 91.6 91.1 91.0 92.3 92.6 1989-2013 Arizona 84.0 83.0 81.6 80.3 82.8 82.7 1989-2013 Arkansas 37.8 28.3 28.1 28.6 26.7 28.0 1989-2013

282

Percentage of Total Natural Gas Industrial Deliveries included in Prices  

Gasoline and Diesel Fuel Update (EIA)

City Gate Price Residential Price Percentage of Total Residential Deliveries included in Prices Commercial Price Percentage of Total Commercial Deliveries included in Prices Industrial Price Percentage of Total Industrial Deliveries included in Prices Electric Power Price Period: Monthly Annual City Gate Price Residential Price Percentage of Total Residential Deliveries included in Prices Commercial Price Percentage of Total Commercial Deliveries included in Prices Industrial Price Percentage of Total Industrial Deliveries included in Prices Electric Power Price Period: Monthly Annual Download Series History Download Series History Definitions, Sources & Notes Definitions, Sources & Notes Show Data By: Data Series Area May-13 Jun-13 Jul-13 Aug-13 Sep-13 Oct-13 View History U.S. 16.5 16.3 16.0 16.2 16.6 16.9 2001-2013 Alabama 22.1 21.7 21.6 22.8 22.0 22.7 2001-2013 Alaska 100.0 100.0 100.0 100.0 100.0 100.0 2001-2013 Arizona 13.4 15.7 15.3 13.8 13.7 13.9 2001-2013 Arkansas 1.7 1.4 1.2 1.4 1.3 1.5 2001-2013

283

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

DOE Green Energy (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

284

US EPR Tests Performed to confirm the Mechanical and Hydraulic Design of the Vessel Internals  

SciTech Connect

The EPR is an Evolutionary high-Power Reactor which is based on the best French and German experience of the past twenty years in plant design construction and operation. In the present detailed engineering phase of the plant under construction in Finland (Okiluoto 3) or scheduled in France (Flamanville 3), a few actions are still ongoing mainly to complement equipment validation files. Design and validation of the main EPR components were performed within Framatome ANP's engineering teams and its two Technical Centers located in France and Germany, which develop state of the art methods in the field of thermo hydraulic testing. The Reactor Pressure Vessel internals are mainly derived from components already implemented on presently operating plants, but they differ in some features from the design used in French N4 or German Konvoi. The aim of this paper is to present the tests performed to confirm the hydraulic and mechanical design of the EPR vessel internals. - Four different mock-ups are presented to illustrate these tests: - JULIETTE for the reactor pressure vessel lower internals; - ROMEO for the reactor pressure vessel upper internals; - MAGALY for the design of the skeleton-type control rod guide assembly; - HYDRAVIB for the vibratory response of the reactor pressure vessel lowers internals. (authors)

Dolleans, Philippe; Chambrin, Jean-Luc; Muller, Thierry [FRAMATOME ANP, Tour AREVA 1 place de la Coupole, 92084 PARIS La D ense (France)

2006-07-01T23:59:59.000Z

285

In-Vessel Retention of Molten Corium: Lessons Learned and Outstanding Issues  

Science Conference Proceedings (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 Advanced 600 MWe Pressurized Water Reactor (PWR) designed by Westinghouse (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 Light Water Reactors (LWRs). However, it is not clear that the ERVC proposed for the AP600 could provide sufficient heat removal for higher-power reactors (up to 1500 MWe) without additional enhancements. This paper reviews efforts made and results reported regarding the enhancement of IVR in LWRs. Where appropriate, the paper identifies what additional data or analyses are needed to demonstrate that there is sufficient margin for successful IVR in high power thermal reactors.

J.L. Rempe; K.Y. Suh; F. B. Cheung; S. B. Kim

2008-03-01T23:59:59.000Z

286

Introduction to Small-Scale Photovoltaic Systems (Including RETScreen Case  

Open Energy Info (EERE)

Introduction to Small-Scale Photovoltaic Systems (Including RETScreen Case Introduction to Small-Scale Photovoltaic Systems (Including RETScreen Case Study) (Webinar) Jump to: navigation, search Tool Summary LAUNCH TOOL Name: Introduction to Small-Scale Photovoltaic Systems (Including RETScreen Case Study) (Webinar) Focus Area: Solar Topics: Market Analysis Website: www.leonardo-energy.org/webinar-introduction-small-scale-photovoltaic- Equivalent URI: cleanenergysolutions.org/content/introduction-small-scale-photovoltaic Language: English Policies: Deployment Programs DeploymentPrograms: Project Development This video teaches the viewer about photovoltaic arrays and RETscreen's photovoltaic module, which can be used to project the cost and production of an array. An example case study was

287

projects are valued at approximately $67 million (including $15 million  

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

projects are valued at approximately $67 million (including $15 million projects are valued at approximately $67 million (including $15 million in non-Federal cost sharing) over four years. The overall goal of the research is to develop carbon dioxide (CO 2 ) capture and separation technologies that can achieve at least 90 percent CO 2 removal at no more than a 35 percent increase in the cost of electricity. The projects, managed by FE's National Energy Technology Laboratory (NETL), include: (1) Linde, LLC, which will use a post-combustion capture technology incorporating BASF's novel amine-based process at a 1-megawatt electric (MWe) equivalent slipstream pilot plant at the National Carbon Capture Center (NCCC) (DOE contribution: $15 million); (2) Neumann Systems Group, Inc., which will design, construct, and test a patented NeuStreamTM absorber at the Colorado

288

Honda Smart Home to Include Berkeley Lab Ventilation Controller  

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

Honda Smart Home to Include Berkeley Lab Ventilation Controller Honda Smart Home to Include Berkeley Lab Ventilation Controller Honda smart home October 2013 October-November Special Focus: Energy Efficiency, Buildings, and the Electric Grid Honda Motor Company Inc is proceeding with plans to build a Smart Home in Davis, California, to demonstrate the latest in renewable energy technologies and energy efficiency. The home is expected to produce more energy than is consumed, demonstrating how the goal of "zero net energy" can be met in the near term future. A ventilation controller developed by researchers at Berkeley Lab's Environmental Energy Technologies Division (EETD) will be included in the smart home. EETD is currently working with the developers of the home control system to integrate its control algorithms.

289

DOE Revises its NEPA Regulations, Including Categorical Exclusions |  

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

Revises its NEPA Regulations, Including Categorical Exclusions Revises its NEPA Regulations, Including Categorical Exclusions DOE Revises its NEPA Regulations, Including Categorical Exclusions September 30, 2011 - 2:30pm Addthis On September 27, 2011, the Department of Energy (DOE) approved revisions to its National Environmental Policy Act (NEPA) regulations, and on September 28th, submitted the revisions to the Federal Register. The final regulations, which become effective 30 days after publication in the Federal Register, are the culmination of a 2-year process to review and update DOE's NEPA implementing procedures. This process involved internal evaluation, public participation, and Council on Environmental Quality (CEQ) review. The revisions are designed to focus Departmental resources on projects with the potential for significant environmental impact, to better

290

Solar Energy Education. Renewable energy: a background text. [Includes glossary  

SciTech Connect

Some of the most common forms of renewable energy are presented in this textbook for students. The topics include solar energy, wind power hydroelectric power, biomass ocean thermal energy, and tidal and geothermal energy. The main emphasis of the text is on the sun and the solar energy that it yields. Discussions on the sun's composition and the relationship between the earth, sun and atmosphere are provided. Insolation, active and passive solar systems, and solar collectors are the subtopics included under solar energy. (BCS)

1985-01-01T23:59:59.000Z

291

Thin film solar cell including a spatially modulated intrinsic layer  

SciTech Connect

One or more thin film solar cells in which the intrinsic layer of substantially amorphous semiconductor alloy material thereof includes at least a first band gap portion and a narrower band gap portion. The band gap of the intrinsic layer is spatially graded through a portion of the bulk thickness, said graded portion including a region removed from the intrinsic layer-dopant layer interfaces. The band gap of the intrinsic layer is always less than the band gap of the doped layers. The gradation of the intrinsic layer is effected such that the open circuit voltage and/or the fill factor of the one or plural solar cell structure is enhanced.

Guha, Subhendu (Troy, MI); Yang, Chi-Chung (Troy, MI); Ovshinsky, Stanford R. (Bloomfield Hills, MI)

1989-03-28T23:59:59.000Z

292

Solar Energy Education. Renewable energy: a background text. [Includes glossary  

DOE Green Energy (OSTI)

Some of the most common forms of renewable energy are presented in this textbook for students. The topics include solar energy, wind power hydroelectric power, biomass ocean thermal energy, and tidal and geothermal energy. The main emphasis of the text is on the sun and the solar energy that it yields. Discussions on the sun's composition and the relationship between the earth, sun and atmosphere are provided. Insolation, active and passive solar systems, and solar collectors are the subtopics included under solar energy. (BCS)

Not Available

1985-01-01T23:59:59.000Z

293

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

SciTech Connect

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

294

BWRVIP-265: BWR Vessel and Internals Project, Crack Growth in High Fluence BWR Materials-Phase 2  

Science Conference Proceedings (OSTI)

The Boiling Water Reactor Vessel and Internals Project (BWRVIP) has developed a methodology to evaluate crack growth rates (CGR) in irradiated stainless steel components in the BWR vessel. This methodology is documented in BWRVIP-99-A: BWR Vessel and Internals Project, Crack Growth Rates in Irradiated Stainless Steels in BWR Internal Components (EPRI report 1016566), and is applicable to neutron doses of 0.7 to 4.2 displacements per atom (dpa) (5x1020 to 3x1021 ...

2012-11-14T23:59:59.000Z

295

BWRVIP-270, Revision 1: BWR Vessel and Internals Project, Compilation of Fluence Estimates for Boiling Water Reactor Materials  

Science Conference Proceedings (OSTI)

The Boiling Water Reactor Vessel and Internals Project (BWRVIP) is an association of utilities focused on BWR vessel and internals issues. Many of the BWR internal components receive high exposure to neutron flux due to their proximity to the fuel in the Reactor Pressure Vessel (RPV). Identifying how predicted fluence values will impact the materials at these locations is a focus of the BWRVIP proactive materials strategy. As part of this approach, this report provides visual and tabular summaries ...

2013-12-09T23:59:59.000Z

296

Addressing questions about including environmental effects in the DMSO HLA  

SciTech Connect

The Defense Modeling and Simulation Office (DMSO) is developing a High Level Architecture (HLA) to support the DOD Modeling and Simulation (M and S) community. Many, if not all, of the simulations involve the environment in some fashion. In some applications, the simulation takes place in an acknowledged environment without any environmental functionality being taken into account. The Joint Training Federation Prototype (JTFp) is one of several prototype efforts that have been created to provide a test of the DMSO HLA. In addition to addressing the applicability of the HLA to a training community, the JTFp is also one of two prototype efforts that is explicitly including environmental effects in their simulation effort. These two prototyping efforts are examining the issues associated with the inclusion of the environment in an HLA federation. In deciding whether or not to include an environmental federation in the JTFp effort, a number of questions have been raised about the environment and the HLA. These questions have raised the issue of incompatibility between the environment and the HLA and also shown that there is something unique about including the environment in simulations. The purpose of this White Paper, which was developed with inputs from the National Air and Space [Warfare] Model Program among others, is to address the various questions that have been posed about including environmental effects in an HLA simulation.

Hummel, J.R.

1996-10-01T23:59:59.000Z

297

cDNA encoding a polypeptide including a hevein sequence  

DOE Patents (OSTI)

A cDNA clone (HEV1) encoding hevein was isolated via polymerase chain reaction (PCR) using mixed oligonucleotides corresponding to two regions of hevein as primers and a Hevea brasiliensis latex cDNA library as a template. HEV1 is 1,018 nucleotides long and includes an open reading frame of 204 amino acids.

Raikhel, N.V.; Broekaert, W.F.; Namhai Chua; Kush, A.

1993-02-16T23:59:59.000Z

298

Thermal Unit Commitment Including Optimal AC Power Flow Constraints  

E-Print Network (OSTI)

Thermal Unit Commitment Including Optimal AC Power Flow Constraints Carlos Murillo{Sanchez Robert J algorithm for unit commitment that employs a Lagrange relaxation technique with a new augmentation. This framework allows the possibility of committing units that are required for the VArs that they can produce

299

Major initiatives in materials research at Western include  

E-Print Network (OSTI)

in nuclear reactors; and a third in Engineering- J. Jiang, supported by UNENE, working on control in the theory of condensed matter, including its applications to polymers, optical, electronic, and magnetic NSERC Industrial Research Chairs who together make Western a leading university in nuclear power

Christensen, Dan

300

Comparison of ALICE-II code predictions with SRI complex vessel experiments  

Science Conference Proceedings (OSTI)

Several complex vessel experiments on 1/20-scale models of the Clinch River Breeder Reactor Project (CRBR) were performed by SRI International to help evaluate the containment structural integrity subjected to HCDAs. Among these experiments SM-3 is a simple model which consists of a radial shield, core barrel, upper internal structure (UIS), and a primary vessel. Tests SM-4 and SM-5 are more complex models than SM-3. This paper presents comparisons of the ALICE-II code (Arbitrary Lagrangian Implicit-explicit Continuous Fluid Eulerian containment code - second version) with experiments SM-3 through SM-5. Two calculations are performed with ALICE-II on each of these three experiments, using both the pressure-time histories (p-t) and the pressure-volume relationships (p-v) as input to describe the energy source. Pressure profiles, dynamic strains, and vessel deformations are used as the basis of the comparison.

Ku, J.L.; Wang, C.Y.; Zeuch, W.R.

1983-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "ocean-going vessels including" 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

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

SciTech Connect

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

302

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

E-Print Network (OSTI)

In 1953 the tangled, skeletal remains of a ship were pulled from the small harbor of Penetanguishene, Ontario. Local historians had hoped to raise the hull of a War of 1812 veteran, but the vessel pulled from the depths did not meet the criteria. Identified as H.M. Schooner Tecumseth, the vessel was built just after the War of 1812 had ended. Historical research of Tecumseth and her sister ship Newash, which remained in Penetanguishene harbor, illuminated the ships? shadowy past. Conceived and built after the war, the vessels sailed for only two years before being rendered obsolete by the terms of the Rush-Bagot disarmament agreement. Nevertheless, the two vessels offer a unique perspective from which to view the post-war period on the Great Lakes. The schooners? hulls were interpreted and analyzed using archaeological evidence. A theoretical rigging reconstruction was created, using contemporary texts and documentary evidence of the ships themselves. Architectural hull analysis was carried out to explore the nature of these vessels. From these varied approaches, a conception of Newash and Tecumseth has emerged, revealing ways in which the hulls were designed to fulfill their specific duties. The hulls were sharp, yet had capacious cargo areas. The rigs combined square-rigged and fore-and-aft sails for maximum flexibility. The designs of the hulls and rigging also reflect predominant attitudes of the period, in which naval vessels on the lakes gave way to merchant craft. Taken as a whole, Tecumseth and Newash illustrate how ships, while fluid in the nature of their work, are also singular entities that truly encapsulate a specific point in time and place.

Gordon, Leeanne E.

2009-05-01T23:59:59.000Z

303

Spectrometer capillary vessel and method of making same  

DOE Patents (OSTI)

The present invention is an arrangement of a glass capillary tube for use in spectroscopy. In particular, the invention is a capillary arranged in a manner permitting a plurality or multiplicity of passes of a sample material through a spectroscopic measurement zone. In a preferred embodiment, the multi-pass capillary is insertable within a standard NMR sample tube. The present invention further includes a method of making the multi-pass capillary tube and an apparatus for spinning the tube.

Linehan, John C. (Richland, WA); Yonker, Clement R. (Kennewick, WA); Zemanian, Thomas S. (Richland, WA); Franz, James A. (Kennewick, WA)

1995-01-01T23:59:59.000Z

304

Spectrometer capillary vessel and method of making same  

DOE Patents (OSTI)

The present invention is an arrangement of a glass capillary tube for use in spectroscopy. In particular, the invention is a capillary arranged in a manner permitting a plurality or multiplicity of passes of a sample material through a spectroscopic measurement zone. In a preferred embodiment, the multi-pass capillary is insertable within a standard NMR sample tube. The present invention further includes a method of making the multi-pass capillary tube and an apparatus for spinning the tube. 13 figs.

Linehan, J.C.; Yonker, C.R.; Zemanian, T.S.; Franz, J.A.

1995-11-21T23:59:59.000Z

305

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

Science Conference Proceedings (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 ({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

306

Structural integrity of vessels for coal conversion systems. [ASME and ANSI codes  

DOE Green Energy (OSTI)

The integrity of a coal conversion system need not be compromised by material considerations in design or fabrication. The ASME and ANSI Codes assure the structural integrity of the large pressure vessels and piping when they are placed into service. Imposing additional requirements, such as increased impact toughness, will further assure the reliability and safety of the Code-fabricated vessel. Incorporating in-service surveillance as part of the operational plan will ensure the integrity of the pressure-containing components for the anticipated service life.

Canonico, D.A.

1979-09-01T23:59:59.000Z

307

The dissolution vessel for plutonium pits at the U.S. DOE Pantex Plant  

Science Conference Proceedings (OSTI)

The US DOE Pantex Plant has been given the mission to recertify and requalify plutonium pits for reuse in existing War Reserve nuclear weapons. The first process common to both recertification and requalification is cleaning the plutonium pit. The pit will be cleaned in a dissolution vessel using N-methyl pyrrolidone (NMP) solvent. The recertification and requalification programs are both in the design concept phase at Pantex Plant. The US DOE Pantex Plant secures the national security of the United States by using safe vessels for cleaning plutonium pits in a manner that protects the health and safety of employees, the public and the environment.

Eifert, E.J.; Vickers, L.D.

2000-02-01T23:59:59.000Z

308

BWRVIP-53-A: BWR Vessel and lnternals Project, Standby Liquid Control Line Repair Design Criteria  

Science Conference Proceedings (OSTI)

The Boiling Water Reactor Vessel and Internals Project (BWRVIP), formed in June, 1994, is an association of utilities focused exclusively on BWR vessel and internals issues. This BWRVIP report documents criteria which can be used to design a repair for the standby liquid control (SLC) line in a BWR. A previous version of this report was published as BWRVIP-53 (TR-108716). This report (BWRVIP-53- A) incorporates changes proposed by the BWRVIP in response to U.S. Nuclear Regulatory Commission (NRC) Request...

2005-09-07T23:59:59.000Z

309

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

DOE Patents (OSTI)

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

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

1988-01-01T23:59:59.000Z

310

Tank vessels transferring Outer Continental Shelf (OCS) oil proposed design and equipment standards  

SciTech Connect

The US Coast Guard proposes to require US and foreign flag tank vessels engaged in the transfer of OCS oil in bulk as cargo from an offshore oil exploitation or production facility to shore to have segregated ballast tanks, dedicated clean ballast tanks, or special ballast arrangements by 6/1/80. This proposal would implement the Port and Tanker Safety Act of 1978 and would eliminate the mixing of ballast water and oil, thus reducing operational pollution that could occur if there was a substantial increase in vessel traffic. Comments must be received by 6/16/80.

1980-05-01T23:59:59.000Z

311

DOE Considers Natural Gas Utility Service Options: Proposal Includes  

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

Considers Natural Gas Utility Service Options: Proposal Considers Natural Gas Utility Service Options: Proposal Includes 30-mile Natural Gas Pipeline from Pasco to Hanford DOE Considers Natural Gas Utility Service Options: Proposal Includes 30-mile Natural Gas Pipeline from Pasco to Hanford January 23, 2012 - 12:00pm Addthis Media Contacts Cameron Hardy, DOE , (509) 376-5365, Cameron.Hardy@rl.doe.gov RICHLAND, WASH. - The U.S. Department of Energy (DOE) is considering natural gas transportation and distribution requirements to support the Waste Treatment Plant (WTP) and evaporator operations at the Hanford Site in southeastern Washington State. DOE awarded a task order worth up to $5 million to the local, licensed supplier of natural gas in the Hanford area, Cascade Natural Gas Corporation (Cascade). Cascade will support DOE and its Environmental

312

Italy (including San Marino) Fossil-Fuel CO2 Emissions  

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

Western Europe » Italy Western Europe » Italy (including San Marino) Italy (including San Marino) Fossil-Fuel CO2 Emissions Graph graphic Graphics Data graphic Data Trends As occurred in many industrialized nations, CO2 emissions from Italy rose steeply since the late 1940's until the growth was abruptly terminated in 1974. Since 1974, emissions from liquid fuels have vacillated, dropping from 76% to 46% of a static but varying total. Significant increases in natural gas consumption have compensated for the drop in oil consumption. In 2008, 35.8% of Italy's fossil-fuel CO2 emissions were due to natural gas consumption. Coal usage grew steadily until 1985 when CO2 emissions from coal consumption reached 16 million metric tons of carbon. Not until 2004 did coal usage exceed 1985 levels and now accounts for 13.9% of Italy's

313

Coordination). Participants include representatives from Balancing Authorities (BAs), Reliability  

E-Print Network (OSTI)

The MRO Subject Matter Expert Team is an industry stakeholder group which includes subject matter experts from MRO member organizations in various technical areas. Any materials, guidance, and views from stakeholder groups are meant to be helpful to industry participants; but should not be considered approved or endorsed by MRO staff or its board of directors unless specified. Page | 2 Disclaimer The Midwest Reliability Organization (MRO) Standards Committee (SC) is committed to providing training and non-binding guidance to industry stakeholders regarding existing and emerging Reliability Standards. Any materials, including presentations, were developed through the MRO SC by Subject Matter Experts (SMEs) from member organizations within the MRO region. In 2012, SMEs in the field of System Operator Communications were brought together to prepare a guide for complying with NERC Reliability Standard COM-002-2 (Communications and

Will Behnke; Alliant Energy; Jacalynn Bentz; Great River Energy; Marie Knox Miso; Jacalynn Bentz; Marie Knox; Terry Harbour

2013-01-01T23:59:59.000Z

314

Flicker Performance of Modern Lighting Technologies including Impacts of Dimmers  

Science Conference Proceedings (OSTI)

The existing industry standards on flicker measurement and assessment are based on the response of general purpose incandescent lamps. However, worldwide these lamps are being replaced with more energy efficient lamps including Compact Fluorescent Lamps (CFLs) and Light emitting Diode (LED) lamps. In order to keep the flicker standards relevant, the industry standard bodies on the subject are in need of the evidence that compares the flicker performance of new lighting ...

2012-12-12T23:59:59.000Z

315

Conversion of geothermal waste to commercial products including silica  

DOE Patents (OSTI)

A process for the treatment of geothermal residue includes contacting the pigmented amorphous silica-containing component with a depigmenting reagent one or more times to depigment the silica and produce a mixture containing depigmented amorphous silica and depigmenting reagent containing pigment material; separating the depigmented amorphous silica and from the depigmenting reagent to yield depigmented amorphous silica. Before or after the depigmenting contacting, the geothermal residue or depigmented silica can be treated with a metal solubilizing agent to produce another mixture containing pigmented or unpigmented amorphous silica-containing component and a solubilized metal-containing component; separating these components from each other to produce an amorphous silica product substantially devoid of metals and at least partially devoid of pigment. The amorphous silica product can be neutralized and thereafter dried at a temperature from about 25.degree. C. to 300.degree. C. The morphology of the silica product can be varied through the process conditions including sequence contacting steps, pH of depigmenting reagent, neutralization and drying conditions to tailor the amorphous silica for commercial use in products including filler for paint, paper, rubber and polymers, and chromatographic material.

Premuzic, Eugene T. (East Moriches, NY); Lin, Mow S. (Rocky Point, NY)

2003-01-01T23:59:59.000Z

316

Multi-processor including data flow accelerator module  

DOE Patents (OSTI)

An accelerator module for a data flow computer includes an intelligent memory. The module is added to a multiprocessor arrangement and uses a shared tagged memory architecture in the data flow computer. The intelligent memory module assigns locations for holding data values in correspondence with arcs leading to a node in a data dependency graph. Each primitive computation is associated with a corresponding memory cell, including a number of slots for operands needed to execute a primitive computation, a primitive identifying pointer, and linking slots for distributing the result of the cell computation to other cells requiring that result as an operand. Circuitry is provided for utilizing tag bits to determine automatically when all operands required by a processor are available and for scheduling the primitive for execution in a queue. Each memory cell of the module may be associated with any of the primitives, and the particular primitive to be executed by the processor associated with the cell is identified by providing an index, such as the cell number for the primitive, to the primitive lookup table of starting addresses. The module thus serves to perform functions previously performed by a number of sections of data flow architectures and coexists with conventional shared memory therein. A multiprocessing system including the module operates in a hybrid mode, wherein the same processing modules are used to perform some processing in a sequential mode, under immediate control of an operating system, while performing other processing in a data flow mode.

Davidson, George S. (Albuquerque, NM); Pierce, Paul E. (Albuquerque, NM)

1990-01-01T23:59:59.000Z

317

Global Analysis of Solar Neutrino Oscillations Including SNO CC Measurement  

E-Print Network (OSTI)

For active and sterile neutrinos, we present the globally allowed solutions for two neutrino oscillations. We include the SNO CC measurement and all other relevant solar neutrino and reactor data. Five active neutrino oscillation solutions (LMA, LOW, SMA, VAC, and Just So2) are currently allowed at 3 sigma; three sterile neutrino solutions (Just So2, SMA, and VAC) are allowed at 3 sigma. The goodness of fit is satisfactory for all eight solutions. We also investigate the robustness of the allowed solutions by carrying out global analyses with and without: 1) imposing solar model constraints on the 8B neutrino flux, 2) including the Super-Kamiokande spectral energy distribution and day-night data, 3) including a continuous mixture of active and sterile neutrinos, 4) using an enhanced CC cross section for deuterium (due to radiative corrections), and 5) a optimistic, hypothetical reduction by a factor of three of the error of the SNO CC rate. For every analysis strategy used in this paper, the most favored solutions all involve large mixing angles: LMA, LOW, or VAC. The favored solutions are robust, but the presence at 3 sigma of individual sterile solutions and the active Just So2 solution is sensitive to the analysis assumptions.

John N. Bahcall; M. C. Gonzalez-Garcia; Carlos Pena-Garay

2001-06-25T23:59:59.000Z

318

Survey of welding processes for field fabrication of 2 1/4 Cr-1 Mo steel pressure vessels. [128 references  

SciTech Connect

Any evaluation of fabrication methods for massive pressure vessels must consider several welding processes with potential for heavy-section applications. These include submerged-arc and shielded metal-arc, narrow-joint modifications of inert-gas metal-arc and inert-gas tungsten-arc processes, electroslag, and electron beam. The advantage and disadvantages of each are discussed. Electroslag welding can be dropped from consideration for joining of 2 1/4 Cr-1 Mo steel because welds made with this method do not provide the required mechanical properties in the welded and stress relieved condition. The extension of electron-beam welding to sections as thick as 4 or 8 inches (100 or 200 mm) is too recent a development to permit full evaluation. The manual shielded metal-arc and submerged-arc welding processes have both been employed, often together, for field fabrication of large vessels. They have the historical advantage of successful application but present other disadvantages that make them otherwise less attractive. The manual shielded metal-arc process can be used for all-position welding. It is however, a slow and expensive technique for joining heavy sections, requires large amounts of skilled labor that is in critically short supply, and introduces a high incidence of weld repairs. Automatic submerged-arc welding has been employed in many critical applications and for welding in the flat position is free of most of the criticism that can be leveled at the shielded metal-arc process. Specialized techniques have been developed for horizontal and vertical position welding but, used in this manner, the applications are limited and the cost advantage of the process is lost.

Grotke, G.E.

1980-04-01T23:59:59.000Z

319

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

Science Conference Proceedings (OSTI)

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

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

2006-07-01T23:59:59.000Z

320

Information regarding previous INCITE awards including selected highlights  

Office of Science (SC) Website

Information regarding previous INCITE awards including selected highlights Advanced Scientific Computing Research (ASCR) ASCR Home About Research Facilities Accessing ASCR Supercomputers Oak Ridge Leadership Computing Facility (OLCF) Argonne Leadership Computing Facility (ALCF) National Energy Research Scientific Computing Center (NERSC) Energy Sciences Network (ESnet) Research & Evaluation Prototypes (REP) Innovative & Novel Computational Impact on Theory and Experiment (INCITE) ASCR Leadership Computing Challenge (ALCC) Science Highlights Benefits of ASCR Funding Opportunities Advanced Scientific Computing Advisory Committee (ASCAC) News & Resources Contact Information Advanced Scientific Computing Research U.S. Department of Energy SC-21/Germantown Building

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


321

Composite armor, armor system and vehicle including armor system  

DOE Patents (OSTI)

Composite armor panels are disclosed. Each panel comprises a plurality of functional layers comprising at least an outermost layer, an intermediate layer and a base layer. An armor system incorporating armor panels is also disclosed. Armor panels are mounted on carriages movably secured to adjacent rails of a rail system. Each panel may be moved on its associated rail and into partially overlapping relationship with another panel on an adjacent rail for protection against incoming ordnance from various directions. The rail system may be configured as at least a part of a ring, and be disposed about a hatch on a vehicle. Vehicles including an armor system are also disclosed.

Chu, Henry S.; Jones, Warren F.; Lacy, Jeffrey M.; Thinnes, Gary L.

2013-01-01T23:59:59.000Z

322

cDNA encoding a polypeptide including a hevein sequence  

DOE Patents (OSTI)

A cDNA clone (HEV1) encoding hevein was isolated via polymerase chain reaction (PCR) using mixed oligonucleotides corresponding to two regions of hevein as primers and a Hevea brasiliensis latex cDNA library as a template. HEV1 is 1018 nucleotides long and includes an open reading frame of 204 amino acids. The deduced amino acid sequence contains a pu GOVERNMENT RIGHTS This application was funded under Department of Energy Contract DE-AC02-76ER01338. The U.S. Government has certain rights under this application and any patent issuing thereon.

Raikhel, Natasha V. (Okemos, MI); Broekaert, Willem F. (Dilbeek, BE); Chua, Nam-Hai (Scarsdale, NY); Kush, Anil (New York, NY)

1993-02-16T23:59:59.000Z

323

Composite material including nanocrystals and methods of making  

DOE Patents (OSTI)

Temperature-sensing compositions can include an inorganic material, such as a semiconductor nanocrystal. The nanocrystal can be a dependable and accurate indicator of temperature. The intensity of emission of the nanocrystal varies with temperature and can be highly sensitive to surface temperature. The nanocrystals can be processed with a binder to form a matrix, which can be varied by altering the chemical nature of the surface of the nanocrystal. A nanocrystal with a compatibilizing outer layer can be incorporated into a coating formulation and retain its temperature sensitive emissive properties

Bawendi, Moungi G. (Boston, MA); Sundar, Vikram C. (New York, NY)

2008-02-05T23:59:59.000Z

324

Community Assessment Tool for Public Health Emergencies Including Pandemic Influenza  

SciTech Connect

The Community Assessment Tool (CAT) for Public Health Emergencies Including Pandemic Influenza (hereafter referred to as the CAT) was developed as a result of feedback received from several communities. These communities participated in workshops focused on influenza pandemic planning and response. The 2008 through 2011 workshops were sponsored by the Centers for Disease Control and Prevention (CDC). Feedback during those workshops indicated the need for a tool that a community can use to assess its readiness for a disaster - readiness from a total healthcare perspective, not just hospitals, but the whole healthcare system. The CAT intends to do just that - help strengthen existing preparedness plans by allowing the healthcare system and other agencies to work together during an influenza pandemic. It helps reveal each core agency partners (sectors) capabilities and resources, and highlights cases of the same vendors being used for resource supplies (e.g., personal protective equipment [PPE] and oxygen) by the partners (e.g., public health departments, clinics, or hospitals). The CAT also addresses gaps in the community's capabilities or potential shortages in resources. This tool has been reviewed by a variety of key subject matter experts from federal, state, and local agencies and organizations. It also has been piloted with various communities that consist of different population sizes, to include large urban to small rural communities.

ORAU's Oak Ridge Institute for Science Education (HCTT-CHE)

2011-04-14T23:59:59.000Z

325

Community Assessment Tool for Public Health Emergencies Including Pandemic Influenza  

SciTech Connect

The Community Assessment Tool (CAT) for Public Health Emergencies Including Pandemic Influenza (hereafter referred to as the CAT) was developed as a result of feedback received from several communities. These communities participated in workshops focused on influenza pandemic planning and response. The 2008 through 2011 workshops were sponsored by the Centers for Disease Control and Prevention (CDC). Feedback during those workshops indicated the need for a tool that a community can use to assess its readiness for a disaster - readiness from a total healthcare perspective, not just hospitals, but the whole healthcare system. The CAT intends to do just that - help strengthen existing preparedness plans by allowing the healthcare system and other agencies to work together during an influenza pandemic. It helps reveal each core agency partners (sectors) capabilities and resources, and highlights cases of the same vendors being used for resource supplies (e.g., personal protective equipment [PPE] and oxygen) by the partners (e.g., public health departments, clinics, or hospitals). The CAT also addresses gaps in the community's capabilities or potential shortages in resources. This tool has been reviewed by a variety of key subject matter experts from federal, state, and local agencies and organizations. It also has been piloted with various communities that consist of different population sizes, to include large urban to small rural communities.

ORAU' s Oak Ridge Institute for Science Education (HCTT-CHE)

2011-04-14T23:59:59.000Z

326

Detection and characterization of indications in segments of reactor pressure vessels  

Science Conference Proceedings (OSTI)

Studies have been conducted to estimate flaw density in segments cut from light water reactor (LWR) pressure vessels as part of the Oak Ridge National Laboratory's Heavy-Section Steel Technology (HSST) Program. Segments from the Hope Creek Unit 2 vessel and the Pilgrim Unit 2 vessel were purchased from salvage dealers. Hope Creek was a boiling water reactor (BWR) design and Pilgrim was a pressurized water reactor (PWR) design. Neither were ever placed in service. Objectives were to evaluate these LWR segments for flaws with ultrasonic and liquid penetrant techniques and to compare the results with current assumptions related to probabilistic risk assessment. Both objectives were successfully completed. Ultrasonic techniques beyond those required by the 1986 edition of the ASME Boiler and Pressure Vessel Code were necessary for the detection and reporting of the detected discontinuities. Extra care and analysis must be exercised when conducting ultrasonic examination through cladding. The detection of the discontinuities in the arbitrarily selected sections implies that the Marshall report estimates (and others) are nonconservative for such small flaws. 8 refs., 9 figs.

Cook, K.V.; Cunningham, R.A. Jr.; McClung, R.W.

1989-08-01T23:59:59.000Z

327

Vessel routing and scheduling under uncertainty in the liquefied natural gas business  

Science Conference Proceedings (OSTI)

Liquefied natural gas (LNG) is natural gas transformed into liquid state for the purpose of transportation mainly by specially built LNG vessels. This paper considers a real-life LNG ship routing and scheduling problem where a producer is responsible ... Keywords: Liquefied natural gas, Maritime transportation, Ship routing and scheduling, Simulation, Uncertainty

Elin E. Halvorsen-Weare; Kjetil Fagerholt; Mikael RöNnqvist

2013-01-01T23:59:59.000Z

328

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

Science Conference Proceedings (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

329

Pressurized thermal shock probabilistic fracture mechanics sensitivity analysis for Yankee Rowe reactor pressure vessel  

SciTech Connect

The Nuclear Regulatory Commission (NRC) requested Oak Ridge National Laboratory (ORNL) to perform a pressurized-thermal-shock (PTS) probabilistic fracture mechanics (PFM) sensitivity analysis for the Yankee Rowe reactor pressure vessel, for the fluences corresponding to the end of operating cycle 22, using a specific small-break-loss- of-coolant transient as the loading condition. Regions of the vessel with distinguishing features were to be treated individually -- upper axial weld, lower axial weld, circumferential weld, upper plate spot welds, upper plate regions between the spot welds, lower plate spot welds, and the lower plate regions between the spot welds. The fracture analysis methods used in the analysis of through-clad surface flaws were those contained in the established OCA-P computer code, which was developed during the Integrated Pressurized Thermal Shock (IPTS) Program. The NRC request specified that the OCA-P code be enhanced for this study to also calculate the conditional probabilities of failure for subclad flaws and embedded flaws. The results of this sensitivity analysis provide the NRC with (1) data that could be used to assess the relative influence of a number of key input parameters in the Yankee Rowe PTS analysis and (2) data that can be used for readily determining the probability of vessel failure once a more accurate indication of vessel embrittlement becomes available. This report is designated as HSST report No. 117.

Dickson, T.L.; Cheverton, R.D.; Bryson, J.W.; Bass, B.R.; Shum, D.K.M.; Keeney, J.A. [Oak Ridge National Lab., TN (United States)

1993-08-01T23:59:59.000Z

330

Why Web GIS May Not be Enough: A Case Study with the Virtual Research Vessel  

E-Print Network (OSTI)

Why Web GIS May Not be Enough: A Case Study with the Virtual Research Vessel DAWN J. WRIGHT1 infrastructure is desired and needed for ready access to data and the resulting maps via web GIS, in order, and the quantitative evaluation of scientific hypotheses. For widespread data access, web GIS is therefore only

Wright, Dawn Jeannine

331

White Paper on Reactor Vessel Integrity Requirements for Level A and B Conditions  

Science Conference Proceedings (OSTI)

The ASME Section XI Task Group recommends that the Reactor Vessel Integrity Requirements of the ASME code be updated in several areas to reflect current technology. This report reviews current regulations to identify areas of conservatism and potential nonconservatism and to provide recommendations for future code improvements.

1993-03-09T23:59:59.000Z

332

ITER Engineering Design Activities -R & DITER-In-Vessel Remote Handling  

E-Print Network (OSTI)

ITER Engineering Design Activities - R & DITER- In-Vessel Remote Handling Blanket Module Remote Handling Project (L-6) Divertor Remote Handling Project (L-7) Objective To develop and demonstrate handling equipment, port handling equipment, auxiliary remote handling tools and a blanket mockup structure

333

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

SciTech Connect

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

334

Calvert Cliffs 1 Reactor Vessel: Pressurized Thermal Shock Analysis for a Small Steam Line Break  

Science Conference Proceedings (OSTI)

Analysis of this Maryland reactor revealed a wide safety margin in its two-loop Combustion Engineering PWR pressure vessel for transients caused by small steam line breaks. The study employed a new method for analyzing pressurized thermal shock effects that combines several EPRI computer codes.

1984-11-01T23:59:59.000Z

335

Measurements of the hydrogenic recombination coefficient for the TFTR vacuum vessel  

DOE Green Energy (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

336

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, SCPerComp Engineering Inc. (HEI) ­ American Society Of Mechanical Engineers (ASME) ­ Pipeline Working Group (PWG) #12

337

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

DOE Green Energy (OSTI)

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

PIEPHO, M.G.

2000-01-10T23:59:59.000Z

338

BWRVIP-199: BWR Vessel and Internals Project, Testing and Evaluation of the Monticello 300 Degree Capsule  

Science Conference Proceedings (OSTI)

In the late 1990s, a BWR Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) was developed to improve the surveillance of the U. S. BWR fleet. This report describes testing and evaluation of the Monticello 300 Capsule. These results will be used to monitor embrittlement as part of the BWRVIP ISP.

2008-12-03T23:59:59.000Z

339

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

E-Print Network (OSTI)

in this paper. Keywords: Remote inspection, Service robot, Non-destructive test, Nuclear, Climbing robotWalking 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

Chen, Sheng

340

Vessel centerline tracking in CTA and MRA images using hough transform  

Science Conference Proceedings (OSTI)

Vascular disease is characterized by any condition that affects the circulatory system. Recently, a demand for sophisticated software tools that can characterize the integrity and functional state of vascular networks from different vascular imaging ... Keywords: CTA, MRA, angiographic images, hough transform, lumen segmentation, vessel tracking

Maysa M. G. Macedo; Choukri Mekkaoui; Marcel P. Jackowski

2010-11-01T23:59:59.000Z

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


341

Notices ROUTINE USES OF RECORDS MAINTAINED IN THE SYSTEM, INCLUDING  

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

83 Federal Register 83 Federal Register / Vol. 78, No. 51 / Friday, March 15, 2013 / Notices ROUTINE USES OF RECORDS MAINTAINED IN THE SYSTEM, INCLUDING CATEGORIES OF USERS AND THE PURPOSES OF SUCH USES: The Department may disclose information contained in a record in this system of records under the routine uses listed in this system of records without the consent of the individual if the disclosure is compatible with the purposes for which the record was collected. These disclosures may be made on a case-by-case basis or, if the Department has complied with the computer matching requirements of the Privacy Act of 1974, as amended (Privacy Act), under a computer matching agreement. Any disclosure of individually identifiable information from a record in this system must also comply with the requirements of section

342

Copper laser modulator driving assembly including a magnetic compression laser  

DOE Patents (OSTI)

A laser modulator (10) having a low voltage assembly (12) with a plurality of low voltage modules (14) with first stage magnetic compression circuits (20) and magnetic assist inductors (28) with a common core (91), such that timing of the first stage magnetic switches (30b) is thereby synchronized. A bipolar second stage of magnetic compression (42) is coupled to the low voltage modules (14) through a bipolar pulse transformer (36) and a third stage of magnetic compression (44) is directly coupled to the second stage of magnetic compression (42). The low voltage assembly (12) includes pressurized boxes (117) for improving voltage standoff between the primary winding assemblies (34) and secondary winding (40) contained therein.

Cook, Edward G. (Livermore, CA); Birx, Daniel L. (Oakley, CA); Ball, Don G. (Livermore, CA)

1994-01-01T23:59:59.000Z

343

[Article 1 of 7: Motivates and Includes the Consumer]  

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

2 of 7: Research on the Characteristics of a Modern Grid by the NETL 2 of 7: Research on the Characteristics of a Modern Grid by the NETL Modern Grid Strategy Team Accommodates All Generation and Storage Options Last month we presented the first Principal Characteristic of a Modern Grid, "Motivates and Includes the Consumer". This month we present a second characteristic, "Accommodates All Generation and Storage Options". This characteristic will fundamentally transition today's grid from a centralized model for generation to one that also has a more balanced contribution from decentralized generation and storage. This characteristic, along with the other six, define a Modern Grid that will power the 21 st Century economy. For a more detailed discussion on "Accommodates All Generation and Storage Options", please see:

344

Search for Earth-like planets includes LANL star analysis  

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

Search for earth-like planets Search for earth-like planets Search for Earth-like planets includes LANL star analysis The mission will not only be able to search for planets around other stars, but also yield new insights into the parent stars themselves. March 6, 2009 Los Alamos National Laboratory sits on top of a once-remote mesa in northern New Mexico with the Jemez mountains as a backdrop to research and innovation covering multi-disciplines from bioscience, sustainable energy sources, to plasma physics and new materials. Los Alamos National Laboratory sits on top of a once-remote mesa in northern New Mexico with the Jemez mountains as a backdrop to research and innovation covering multi-disciplines from bioscience, sustainable energy sources, to plasma physics and new materials.

345

Dye laser amplifier including a specifically designed diffuser assembly  

SciTech Connect

A large (high flow rate) dye laser amplifier in which a continuous replened 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 relatively high flow rate and a specifically designed diffuser assembly for slowing down the flow of dye while, at the same time, assuring that as the dye stream flows through the diffuser assembly it does so in a stable manner.

Davin, James (Gilroy, CA); Johnston, James P. (Stanford, CA)

1992-01-01T23:59:59.000Z

346

Electra-optical device including a nitrogen containing electrolyte  

DOE Patents (OSTI)

Described is a thin-film battery, especially a thin-film microbattery, and a method for making same having application as a backup or primary integrated power source for electronic devices. The battery includes a novel electrolyte which is electrochemically stable and does not react with the lithium anode and a novel vanadium oxide cathode. Configured as a microbattery, the battery can be fabricated directly onto a semiconductor chip, onto the semiconductor die or onto any portion of the chip carrier. The battery can be fabricated to any specified size or shape to meet the requirements of a particular application. The battery is fabricated of solid state materials and is capable of operation between {minus}15 C and 150 C.

Bates, J.B.; Dudney, N.J.; Gruzalski, G.R.; Luck, C.F.

1995-10-03T23:59:59.000Z

347

Hydraulic engine valve actuation system including independent feedback control  

DOE Patents (OSTI)

A hydraulic valve actuation assembly may include a housing, a piston, a supply control valve, a closing control valve, and an opening control valve. The housing may define a first fluid chamber, a second fluid chamber, and a third fluid chamber. The piston may be axially secured to an engine valve and located within the first, second and third fluid chambers. The supply control valve may control a hydraulic fluid supply to the piston. The closing control valve may be located between the supply control valve and the second fluid chamber and may control fluid flow from the second fluid chamber to the supply control valve. The opening control valve may be located between the supply control valve and the second fluid chamber and may control fluid flow from the supply control valve to the second fluid chamber.

Marriott, Craig D

2013-06-04T23:59:59.000Z

348

A thermovoltaic semiconductor device including a plasma filter  

DOE Patents (OSTI)

A thermovoltaic energy conversion device and related method for converting thermal energy into an electrical potential are disclosed. An interference filter is provided on a semiconductor thermovoltaic cell to pre-filter black body radiation. The semiconductor thermovoltaic cell includes a P/N junction supported on a substrate which converts incident thermal energy below the semiconductor junction band gap into electrical potential. The semiconductor substrate is doped to provide a plasma filter which reflects back energy having a wavelength which is above the band gap and which is ineffectively filtered by the interference filter, through the P/N junction to the source of radiation thereby avoiding parasitic absorption of the unusable portion of the thermal radiation energy.

Baldasaro, Paul F.

1997-12-01T23:59:59.000Z

349

Electra-optical device including a nitrogen containing electrolyte  

SciTech Connect

Described is a thin-film battery, especially a thin-film microbattery, and a method for making same having application as a backup or primary integrated power source for electronic devices. The battery includes a novel electrolyte which is electrochemically stable and does not react with the lithium anode and a novel vanadium oxide cathode Configured as a microbattery, the battery can be fabricated directly onto a semiconductor chip, onto the semiconductor die or onto any portion of the chip carrier. The battery can be fabricated to any specified size or shape to meet the requirements of a particular application. The battery is fabricated of solid state materials and is capable of operation between -15.degree. C. and 150.degree. C.

Bates, John B. (Oak Ridge, TN); Dudney, Nancy J. (Knoxville, TN); Gruzalski, Greg R. (Oak Ridge, TN); Luck, Christopher F. (Knoxville, TN)

1995-01-01T23:59:59.000Z

350

BWRVIP-200: BWR Vessel and Internals Project, Implementation Plan for Two-sided Inspection of BWR Shroud Welds  

Science Conference Proceedings (OSTI)

The Boiling Water Reactor Vessel and Internals Project (BWRVIP) formed in June 1994 and is an association of utilities focused exclusively on BWR vessel and internals issues. This BWRVIP report provides the technical basis for implementing two-sided inspections of BWR shrouds to satisfy the ultrasonic testing (UT) inspection requirements contained in BWRVIP-03 and BWRVIP-76.

2008-11-10T23:59:59.000Z

351

Design of control system for hydraulic lifting platform with jack-up wind-power installation vessel  

Science Conference Proceedings (OSTI)

Jack-up wind-power installation vessel is the most important tool in construction of wind farm. And the control system for hydraulic lifting platform is the key point of jack-up wind-power installation vessel. Therefore the design of the control system ... Keywords: hydraulic control, hydraulic lifting platform, programmable logic controller, wind-power

Xuejin Yang; Dingfang Chen; Mingwang Dong; Taotao Li

2012-11-01T23:59:59.000Z

352

Extractant composition including crown ether and calixarene extractants  

SciTech Connect

An extractant composition comprising a mixed extractant solvent consisting of calix[4] arene-bis-(tert-octylbenzo)-crown-6 ("BOBCalixC6"), 4',4',(5')-di-(t-butyldicyclo-hexano)-18-crown-6 ("DtBu18C6"), and at least one modifier dissolved in a diluent. The DtBu18C6 may be present at from approximately 0.01M to approximately 0.4M, such as at from approximately 0.086 M to approximately 0.108 M. The modifier may be 1-(2,2,3,3-tetrafluoropropoxy)-3-(4-sec-butylphenoxy)-2-propanol ("Cs-7SB") and may be present at from approximately 0.01M to approximately 0.8M. In one embodiment, the mixed extractant solvent includes approximately 0.15M DtBu18C6, approximately 0.007M BOBCalixC6, and approximately 0.75M Cs-7SB modifier dissolved in an isoparaffinic hydrocarbon diluent. The extractant composition further comprises an aqueous phase. The mixed extractant solvent may be used to remove cesium and strontium from the aqueous phase.

Meikrantz, David H. (Idaho Falls, ID); Todd, Terry A. (Aberdeen, ID); Riddle, Catherine L. (Idaho Falls, ID); Law, Jack D. (Pocalello, ID); Peterman, Dean R. (Idaho Falls, ID); Mincher, Bruce J. (Idaho Falls, ID); McGrath, Christopher A. (Blackfoot, ID); Baker, John D. (Blackfoot, ID)

2009-04-28T23:59:59.000Z

353

cDNA encoding a polypeptide including a hevein sequence  

DOE Patents (OSTI)

A cDNA clone (HEV1) encoding hevein was isolated via polymerase chain reaction (PCR) using mixed oligonucleotides corresponding to two regions of hevein as primers and a Hevea brasiliensis latex cDNA library as a template. HEV1 is 1018 nucleotides long and includes an open reading frame of 204 amino acids. The deduced amino acid sequence contains a putative signal sequence of 17 amino acid residues followed by a 187 amino acid polypeptide. The amino-terminal region (43 amino acids) is identical to hevein and shows homology to several chitin-binding proteins and to the amino-termini of wound-induced genes in potato and poplar. The carboxyl-terminal portion of the polypeptide (144 amino acids) is 74--79% homologous to the carboxyl-terminal region of wound-inducible genes of potato. Wounding, as well as application of the plant hormones abscisic acid and ethylene, resulted in accumulation of hevein transcripts in leaves, stems and latex, but not in roots, as shown by using the cDNA as a probe. A fusion protein was produced in E. coli from the protein of the present invention and maltose binding protein produced by the E. coli. 12 figs.

Raikhel, N.V.; Broekaert, W.F.; Chua, N.H.; Kush, A.

1999-05-04T23:59:59.000Z

354

CDNA encoding a polypeptide including a hevein sequence  

DOE Patents (OSTI)

A cDNA clone (HEV1) encoding hevein was isolated via polymerase chain reaction (PCR) using mixed oligonucleotides corresponding to two regions of hevein as primers and a Hevea brasiliensis latex cDNA library as a template. HEV1 is 1018 nucleotides long and includes an open reading frame of 204 amino acids. The deduced amino acid sequence contains a putative signal sequence of 17 amino acid residues followed by a 187 amino acid polypeptide. The amino-terminal region (43 amino acids) is identical to hevein and shows homology to several chitin-binding proteins and to the amino-termini of wound-induced genes in potato and poplar. The carboxyl-terminal portion of the polypeptide (144 amino acids) is 74-79% homologous to the carboxyl-terminal region of wound-inducible genes of potato. Wounding, as well as application of the plant hormones abscisic acid and ethylene, resulted in accumulation of hevein transcripts in leaves, stems and latex, but not in roots, as shown by using the cDNA as a probe. A fusion protein was produced in E. coli from the protein of the present invention and maltose binding protein produced by the E. coli.

Raikhel, Natasha V. (Okemos, MI); Broekaert, Willem F. (Dilbeek, BE); Chua, Nam-Hai (Scarsdale, NY); Kush, Anil (New York, NY)

1995-03-21T23:59:59.000Z

355

cDNA encoding a polypeptide including a hevein sequence  

DOE Patents (OSTI)

A cDNA clone (HEV1) encoding hevein was isolated via polymerase chain reaction (PCR) using mixed oligonucleotides corresponding to two regions of hevein as primers and a Hevea brasiliensis latex cDNA library as a template. HEV1 is 1018 nucleotides long and includes an open reading frame of 204 amino acids. The deduced amino acid sequence contains a putative signal sequence of 17 amino acid residues followed by a 187 amino acid polypeptide. The amino-terminal region (43 amino acids) is identical to hevein and shows homology to several chitin-binding proteins and to the amino-termini of wound-induced genes in potato and poplar. The carboxyl-terminal portion of the polypeptide (144 amino acids) is 74-79% homologous to the carboxyl-terminal region of wound-inducible genes of potato. Wounding, as well as application of the plant hormones abscisic acid and ethylene, resulted in accumulation of hevein transcripts in leaves, stems and latex, but not in roots, as shown by using the cDNA as a probe. A fusion protein was produced in E. coli from the protein of the present invention and maltose binding protein produced by the E. coli.

Raikhel, Natasha V. (Okemos, MI); Broekaert, Willem F. (Dilbeek, BE); Chua, Nam-Hai (Scarsdale, NY); Kush, Anil (New York, NY)

1999-05-04T23:59:59.000Z

356

cDNA encoding a polypeptide including a hevein sequence  

DOE Patents (OSTI)

A cDNA clone (HEV1) encoding hevein was isolated via polymerase chain reaction (PCR) using mixed oligonucleotides corresponding to two regions of hevein as primers and a Hevea brasiliensis latex cDNA library as a template. HEV1 is 1,018 nucleotides long and includes an open reading frame of 204 amino acids. The deduced amino acid sequence contains a putative signal sequence of 17 amino acid residues followed by a 187 amino acid polypeptide. The amino-terminal region (43 amino acids) is identical to hevein and shows homology to several chitin-binding proteins and to the amino-termini of wound-induced genes in potato and poplar. The carboxyl-terminal portion of the polypeptide (144 amino acids) is 74--79% homologous to the carboxyl-terminal region of wound-inducible genes of potato. Wounding, as well as application of the plant hormones abscisic acid and ethylene, resulted in accumulation of hevein transcripts in leaves, stems and latex, but not in roots, as shown by using the cDNA as a probe. A fusion protein was produced in E. coli from the protein of the present invention and maltose binding protein produced by the E. coli. 11 figures.

Raikhel, N.V.; Broekaert, W.F.; Chua, N.H.; Kush, A.

1995-03-21T23:59:59.000Z

357

Analysis of 70 Ophiuchi AB including seismic constraints  

E-Print Network (OSTI)

The analysis of solar-like oscillations for stars belonging to a binary system provides a unique opportunity to probe the internal stellar structure and to test our knowledge of stellar physics. Such oscillations have been recently observed and characterized for the A component of the 70 Ophiuchi system. A model of 70 Ophiuchi AB that correctly reproduces all observational constraints available for both stars is determined. An age of 6.2 +- 1.0 Gyr is found with an initial helium mass fraction Y_i=0.266 +- 0.015 and an initial metallicity (Z/X)_i=0.0300 +- 0.0025 when atomic diffusion is included and a solar value of the mixing-length parameter assumed. A precise and independent determination of the value of the mixing-length parameter needed to model 70 Oph A requires accurate measurement of the mean small separation, which is not available yet. Current asteroseismic observations, however, suggest that the value of the mixing-length parameter of 70 Oph A is lower or equal to the solar calibrated value. The e...

Eggenberger, P; Carrier, F; Fernandes, J; Santos, N C

2008-01-01T23:59:59.000Z

358

Analysis of alpha Centauri AB including seismic constraints  

E-Print Network (OSTI)

Detailed models of alpha Cen A and B based on new seismological data for alpha Cen B by Carrier & Bourban (2003) have been computed using the Geneva evolution code including atomic diffusion. Taking into account the numerous observational constraints now available for the alpha Cen system, we find a stellar model which is in good agreement with the astrometric, photometric, spectroscopic and asteroseismic data. The global parameters of the alpha Cen system are now firmly constrained to an age of t=6.52+-0.30 Gyr, an initial helium mass fraction Y_i=0.275+-0.010 and an initial metallicity (Z/X)_i=0.0434+-0.0020. Thanks to these numerous observational constraints, we confirm that the mixing-length parameter alpha of the B component is larger than the one of the A component, as already suggested by many authors (Noels et al. 1991, Fernandes & Neuforge 1995 and Guenther & Demarque 2000): alpha_B is about 8% larger than alpha_A (alpha_A=1.83+-0.10 and alpha_B=1.97+-0.10). Moreover, we show that asteroseismic measurements enable to determine the radii of both stars with a very high precision (errors smaller than 0.3%). The radii deduced from seismological data are compatible with the new interferometric results of Kervella et al. (2003) even if they are slightly larger than the interferometric radii (differences smaller than 1%).

P. Eggenberger; C. Charbonnel; S. Talon; G. Meynet; A. Maeder; F. Carrier; G. Bourban

2004-01-29T23:59:59.000Z

359

Analysis of 70 Ophiuchi AB including seismic constraints  

E-Print Network (OSTI)

The analysis of solar-like oscillations for stars belonging to a binary system provides a unique opportunity to probe the internal stellar structure and to test our knowledge of stellar physics. Such oscillations have been recently observed and characterized for the A component of the 70 Ophiuchi system. A model of 70 Ophiuchi AB that correctly reproduces all observational constraints available for both stars is determined. An age of 6.2 +- 1.0 Gyr is found with an initial helium mass fraction Y_i=0.266 +- 0.015 and an initial metallicity (Z/X)_i=0.0300 +- 0.0025 when atomic diffusion is included and a solar value of the mixing-length parameter assumed. A precise and independent determination of the value of the mixing-length parameter needed to model 70 Oph A requires accurate measurement of the mean small separation, which is not available yet. Current asteroseismic observations, however, suggest that the value of the mixing-length parameter of 70 Oph A is lower or equal to the solar calibrated value. The effects of atomic diffusion and of the choice of the adopted solar mixture were also studied. We also tested and compared the theoretical tools used for the modeling of stars for which p-modes frequencies are detected by performing this analysis with three different stellar evolution codes and two different calibration methods. We found that the different evolution codes and calibration methods we used led to perfectly coherent results.

P. Eggenberger; A. Miglio; F. Carrier; J. Fernandes; N. C. Santos

2008-02-25T23:59:59.000Z

360

Interim performance criteria for photovoltaic energy systems. [Glossary included  

DOE Green Energy (OSTI)

This document is a response to the Photovoltaic Research, Development, and Demonstration Act of 1978 (P.L. 95-590) which required the generation of performance criteria for photovoltaic energy systems. Since the document is evolutionary and will be updated, the term interim is used. More than 50 experts in the photovoltaic field have contributed in the writing and review of the 179 performance criteria listed in this document. The performance criteria address characteristics of present-day photovoltaic systems that are of interest to manufacturers, government agencies, purchasers, and all others interested in various aspects of photovoltaic system performance and safety. The performance criteria apply to the system as a whole and to its possible subsystems: array, power conditioning, monitor and control, storage, cabling, and power distribution. They are further categorized according to the following performance attributes: electrical, thermal, mechanical/structural, safety, durability/reliability, installation/operation/maintenance, and building/site. Each criterion contains a statement of expected performance (nonprescriptive), a method of evaluation, and a commentary with further information or justification. Over 50 references for background information are also given. A glossary with definitions relevant to photovoltaic systems and a section on test methods are presented in the appendices. Twenty test methods are included to measure performance characteristics of the subsystem elements. These test methods and other parts of the document will be expanded or revised as future experience and needs dictate.

DeBlasio, R.; Forman, S.; Hogan, S.; Nuss, G.; Post, H.; Ross, R.; Schafft, H.

1980-12-01T23:59:59.000Z

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


361

Community Assessment Tool for Public Health Emergencies Including Pandemic Influenza  

SciTech Connect

The Community Assessment Tool (CAT) for Public Health Emergencies Including Pandemic Influenza (hereafter referred to as the CAT) was developed as a result of feedback received from several communities. These communities participated in workshops focused on influenza pandemic planning and response. The 2008 through 2011 workshops were sponsored by the Centers for Disease Control and Prevention (CDC). Feedback during those workshops indicated the need for a tool that a community can use to assess its readiness for a disaster—readiness from a total healthcare perspective, not just hospitals, but the whole healthcare system. The CAT intends to do just that—help strengthen existing preparedness plans by allowing the healthcare system and other agencies to work together during an influenza pandemic. It helps reveal each core agency partners' (sectors) capabilities and resources, and highlights cases of the same vendors being used for resource supplies (e.g., personal protective equipment [PPE] and oxygen) by the partners (e.g., public health departments, clinics, or hospitals). The CAT also addresses gaps in the community's capabilities or potential shortages in resources. While the purpose of the CAT is to further prepare the community for an influenza pandemic, its framework is an extension of the traditional all-hazards approach to planning and preparedness. As such, the information gathered by the tool is useful in preparation for most widespread public health emergencies. This tool is primarily intended for use by those involved in healthcare emergency preparedness (e.g., community planners, community disaster preparedness coordinators, 9-1-1 directors, hospital emergency preparedness coordinators). It is divided into sections based on the core agency partners, which may be involved in the community's influenza pandemic influenza response.

HCTT-CHE

2011-04-14T23:59:59.000Z

362

Community Assessment Tool for Public Health Emergencies Including Pandemic Influenza  

SciTech Connect

The Community Assessment Tool (CAT) for Public Health Emergencies Including Pandemic Influenza (hereafter referred to as the CAT) was developed as a result of feedback received from several communities. These communities participated in workshops focused on influenza pandemic planning and response. The 2008 through 2011 workshops were sponsored by the Centers for Disease Control and Prevention (CDC). Feedback during those workshops indicated the need for a tool that a community can use to assess its readiness for a disaster—readiness from a total healthcare perspective, not just hospitals, but the whole healthcare system. The CAT intends to do just that—help strengthen existing preparedness plans by allowing the healthcare system and other agencies to work together during an influenza pandemic. It helps reveal each core agency partners' (sectors) capabilities and resources, and highlights cases of the same vendors being used for resource supplies (e.g., personal protective equipment [PPE] and oxygen) by the partners (e.g., public health departments, clinics, or hospitals). The CAT also addresses gaps in the community's capabilities or potential shortages in resources. While the purpose of the CAT is to further prepare the community for an influenza pandemic, its framework is an extension of the traditional all-hazards approach to planning and preparedness. As such, the information gathered by the tool is useful in preparation for most widespread public health emergencies. This tool is primarily intended for use by those involved in healthcare emergency preparedness (e.g., community planners, community disaster preparedness coordinators, 9-1-1 directors, hospital emergency preparedness coordinators). It is divided into sections based on the core agency partners, which may be involved in the community's influenza pandemic influenza response.

HCTT-CHE

2011-04-14T23:59:59.000Z

363

Evolution of Design Methodologies for Next Generation of Reactor Pressure Vessels and Extensive Role of Thermal-Hydraulic Numerical Tools  

SciTech Connect

The thermal-hydraulic design of the first pressurized water reactors was mainly based on an experimental approach, with a large series of tests on the main equipment [control rod guide tubes, reactor pressure vessel (RPV) plenums, etc.] to check performance.Development of computational fluid dynamics codes and computers now allows for complex simulations of hydraulics phenomena. Provided adequate qualification, these numerical tools are an efficient means to determine hydraulics in the given design and to perform sensitivities for optimization of new designs. Experiments always play their role, first for qualification and then for validation at the last stage of the design. The design of the European Pressurized Water Reactor (EPR), jointly developed by Framatome ANP, Electricite de France (EDF), and the German utilities, is based on both hydraulics calculations and experiments handled in a complementary approach.This paper describes the collective effort launched by Framatome ANP and EDF on hydraulics calculations for the RPV of the EPR. It concerns three-dimensional calculations of RPV inlets, including the cold legs, the RPV downcomer and lower plenum, and the RPV upper plenum up to and including the hot legs. It covers normal operating conditions but also accidental conditions such as pressurized thermal shock in a small-break loss-of-coolant accident. Those hydraulics studies have provided much useful information for the mechanical design of RPV internals.

Bellet, Serge [Electricite de France - Septen (EDF) (France); Goreaud, Nicolas [Framatome ANP(France); Nicaise, Norbert [Framatome ANP (France)

2005-11-15T23:59:59.000Z

364

DETERMINATION OF LIQUID FILM THICKNESS FOLLOWING DRAINING OF CONTACTORS, VESSELS, AND PIPES IN THE MCU PROCESS  

Science Conference Proceedings (OSTI)

The Department of Energy (DOE) identified the caustic side solvent extraction (CSSX) process as the preferred technology to remove cesium from radioactive waste solutions at the Savannah River Site (SRS). As a result, Washington Savannah River Company (WSRC) began designing and building a Modular CSSX Unit (MCU) in the SRS tank farm to process liquid waste for an interim period until the Salt Waste Processing Facility (SWPF) begins operations. Both the solvent and the strip effluent streams could contain high concentrations of cesium which must be removed from the contactors, process tanks, and piping prior to performing contactor maintenance. When these vessels are drained, thin films or drops will remain on the equipment walls. Following draining, the vessels will be flushed with water and drained to remove the flush water. The draining reduces the cesium concentration in the vessels by reducing the volume of cesium-containing material. The flushing, and subsequent draining, reduces the cesium in the vessels by diluting the cesium that remains in the film or drops on the vessel walls. MCU personnel requested that Savannah River National Laboratory (SRNL) researchers conduct a literature search to identify models to calculate the thickness of the liquid films remaining in the contactors, process tanks, and piping following draining of salt solution, solvent, and strip solution. The conclusions from this work are: (1) The predicted film thickness of the strip effluent is 0.010 mm on vertical walls, 0.57 mm on horizontal walls and 0.081 mm in horizontal pipes. (2) The predicted film thickness of the salt solution is 0.015 mm on vertical walls, 0.74 mm on horizontal walls, and 0.106 mm in horizontal pipes. (3) The predicted film thickness of the solvent is 0.022 mm on vertical walls, 0.91 mm on horizontal walls, and 0.13 mm in horizontal pipes. (4) The calculated film volume following draining is: (a) Salt solution receipt tank--1.6 gallons; (b) Salt solution feed tank--1.6 gallons; (c) Decontaminated salt solution hold tank--1.6 gallons; (d) Contactor drain tank--0.40 gallons; (e) Strip effluent hold tank--0.33 gallons; (f) Decontaminated salt solution decanter--0.37 gallons; (g) Strip effluent decanter--0.14 gallons; (h) Solvent hold tank--0.30 gallon; and (i) Corrugated piping between contactors--16-21 mL. (5) After the initial vessel draining, flushing the vessels with 100 gallons of water using a spray nozzle that produces complete vessel coverage and draining the flush water reduces the source term by the following amounts: (i) Salt solution receipt tank--63X; (ii) Salt solution feed tank--63X; (iii) Decontaminated salt solution hold tank--63X; (iv) Contactor drain tank--250X; (v) Strip effluent hold tank--300X; (vi) Decontaminated salt solution decanter--270X; (vii) Strip effluent decanter--710X; (viii) Solvent hold tank--330X. Understand that these estimates of film thickness are based on laboratory testing and fluid mechanics theory. The calculations assume drainage occurs by film flow. Much of the data used to develop the models came from tests with very ''clean'' fluids. Impurities in the fluids and contaminants on the vessels walls could increase liquid holdup. The application of film thickness models and source term reduction calculations should be considered along with operational conditions and H-Tank Farm/Liquid Waste operating experience. These calculations exclude the PVV/HVAC duct work and piping, as well as other areas that area outside the scope of this report.

Poirier, M; Fernando Fondeur, F; Samuel Fink, S

2006-06-06T23:59:59.000Z

365

Design, fabrication and testing of a model heating and cooling system for a vacuum vessel  

SciTech Connect

A full-size model of a typical cooling and heating system for a vacuum vessel was manufactured and examined in order to clarify and enhance the efficiency and reliability of the designed system. The model consisted of two parts; one of which had the same structure as the other and was located facing each other to simulate the adiabatic condition of the vacuum-side of a vacuum vessel. Its components were rectangular plates, eletric heater units, cooling pipes inside of which water and air flew as cooling fluid. A lot of kinds of tests and measurements were performed to evaluate efficiency and reliability on the model. The numerical and theoretical analyses on the system were also carried out using the dimensional finite difference technique. The analytical results agreed pretty well with the experimental.

Shimizu, M.; Miyauchi, Y.; Nakamura, H.; Kajiura, S.; Koizumi, M.; Hata, M.

1981-01-01T23:59:59.000Z

366

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

367

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

SciTech Connect

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

368

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

SciTech Connect

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

369

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

Science Conference Proceedings (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

370

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

Science Conference Proceedings (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

371

Interim Report: Coiled Tubing Drilling and Intervention System Using Cost Effective Vessel  

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

DOCUMENT TITLE: DOCUMENT TITLE: Self Supporting Riser Technology to Enable Coiled Tubing Intervention for Deepwater Wells Document No.: 08121-1502-12 RPSEA PROJECT TITLE: Coil Tubing Drilling and Intervention System Using a Cost Effective Vessel RPSEA Project No.: 08121-1502 01 April 2011 Charles R. Yemington, PE Project Manager Nautilus International 400 North Sam Houston Parkway East, Suite 105 Houston, Texas 77060 RPSEA Project No.: 08121-1502 Coiled Tubing Drilling and Intervention System Using a Cost Effective Vessel RPSEA Project 08121-1502 01 April 2011 Page 2 of 91 LEGAL NOTICE This report was prepared by Nautilus International, LLC. as an account of work sponsored by the Research Partnership to Secure Energy for America (RPSEA). RPSEA members, the

372

Materials Reliability Program, Reactor Vessel Head Boric Acid Corrosion Testing (MRP-165)  

Science Conference Proceedings (OSTI)

Pressurized water reactor (PWR) coolant leakage from stress corrosion cracking of an Alloy 600 control rod drive mechanism (CRDM) penetration has led to one case of severe corrosion and cavity formation in a low-alloy steel reactor vessel head (RVH). The detailed progression of RVH wastage following initial leakage is complicated and probably involves several corrosion mechanisms. The Materials Reliability Program (MRP) has completed three tasks of a comprehensive program to examine postulated sequential...

2005-12-14T23:59:59.000Z

373

Materials Reliability Program: San Onofre Nuclear Generating Station Reactor Vessel Internals Management Engineering Program (MRP-303)  

Science Conference Proceedings (OSTI)

All operating pressurized water reactors must have a reactor vessel internals aging management document in place by December 2011 according to the mandatory requirement under Nuclear Energy Institute (NEI) 03-08. This program should be developed to meet the guidance provided by Materials Reliability Program (MRP) -227, Rev. 0, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines. For non-license renewal plants, the requirements are valid within the current license period, and the Elec...

2011-02-28T23:59:59.000Z

374

Qualification of in-service examination of the Yankee Rowe reactor pressure vessel  

SciTech Connect

Technical support was provided to assist the Yankee Atomic Electric Company with their restart effort for the Yankee plant in Rowe, Massachusetts. Demonstration of adequate margin during a postulated pressurized thermal shock accident was an important part of the justification for restarting the plant, and effective inservice examination of the critical inner surface of the vessel in the beltline region was a key objective and a significant component of the safety analysis. This report discussed this inservice inspection.

Ammirato, F.; Kietzman, K.; Becker, L.; Ashwin, P.; Selby, G.; Krzywosz, K.; Findlan, S. (Electric Power Research Inst., Charlotte, NC (United States). Nondestructive Evaluation Center); Lance, J. (Yankee Atomic Electric Co., Bolton, MA (United States))

1992-12-01T23:59:59.000Z

375

Analysis of the ANL Test Method for 6CVS Containment Vessels  

Science Conference Proceedings (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

376

Materials Reliability Program: Reactor Vessel Head Boric Acid Corrosion Testing (MRP-199)  

Science Conference Proceedings (OSTI)

PWR coolant leakage from stress corrosion cracking of an Alloy 600 control rod drive mechanism (CRDM) penetration has led to one case of severe corrosion and cavity formation in a low-alloy steel reactor vessel head (RVH). The detailed progression of RVH wastage following initial leakage is complicated and probably involves several corrosion mechanisms. The Materials Reliability Program (MRP) has completed three tasks of a comprehensive program to examine postulated sequential stages of boric acid corros...

2007-06-27T23:59:59.000Z

377

Materials Reliability Program: Destructive Examination of the North Anna 2 Reactor Pressure Vessel Head (MRP-198)  

Science Conference Proceedings (OSTI)

This document is the final of three reports concerning the nondestructive and destructive examinations of selected control rod drive mechanism (CRDM) penetrations from the decommissioned North Anna Unit 2 reactor vessel head (RVH). The phase-1 report of the EPRI-MRP (Materials Reliability Program) managed program described the selection and removal of penetrations from the decommissioned RVH and the penetration decontamination and laboratory nondestructive evaluation (NDE). The phase-2 report detailed th...

2006-11-13T23:59:59.000Z

378

Simulant Melt Experiments on Performance of the In-Vessel Core Catcher  

SciTech Connect

In order to enhance the feasibility of in-vessel retention (IVR) of molten core material during a severe accident for high-power reactors, an in-vessel core catcher (IVCC) was designed and evaluated as part of a joint United States-Korean International Nuclear Energy Research Initiative (INERI). The proposed IVCC is expected to increase the thermal margin for success of IVR by providing an “engineered gap” for heat transfer from materials that relocate during a severe accident and potentially serving as a sacrificial material under a severe accident. In this study, LAVA-GAP experiments were performed to investigate the thermal and mechanical performance of the IVCC using the alumina melt as simulant. The LAVAGAP experiments aim to examine the feasibility and sustainability of the IVCC under the various test conditions using 1/8th scale hemispherical test sections. As a feasibility test of the proposed IVCC in this INERI project, the effects of IVCC base steel materials, internal coating materials, and gap size between the IVCC and the vessel lower head were examined. The test results indicated that the internally coated IVCC has high thermal performance compared with the uncoated IVCC. In terms of integrity of the base steel, carbon steel is superior to stainless steel and the effect of bond coat is found to be trivial for the tests performed in this study. The thermal load is mitigated via boiling heat removal in the gap between the IVCC and the vessel lower head. The current test results imply that gaps less than 10mm are not enough to guarantee effective cooling induced by water ingression and steam venting there through. Selection of endurable material and pertinent gap size is needed to implement the proposed IVCC concept into advanced reactor designs.

Kyoung-Ho Kang; Rae-Joon Park; Sang-Baik Kim; K. Y. Suh; F. B. Cheung; J. L. Rempe

2007-09-01T23:59:59.000Z

379

Development of an Enhanced Core Catcher for Improving In-Vessel Retention Margins  

Science Conference Proceedings (OSTI)

In-vessel retention (IVR) of core melt that may relocate to the lower head of a reactor vessel is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for several advanced light water reactors. A U.S.-Korean International Nuclear Energy Research Initiative project has been initiated to explore design enhancements that could increase the margin for IVR for advanced reactors with higher power levels [up to 1500 MW(electric)]. As part of this effort, an enhanced in-vessel core catcher is being designed and evaluated. To reduce cost and simplify manufacture and installation, this new core catcher design consists of several interlocking sections that are machined to fit together when inserted into the lower head. If needed, the core catcher can be manufactured with holes to accommodate lower head penetrations. Each section of the core catcher consists of two material layers with an option to add a third layer (if deemed necessary). The first is a base material that has the capability to support and contain the mass of core materials that may relocate during a severe accident; the second is an oxide coating on top of the base material, which resists interactions with high-temperature core materials; and the third is an optional coating on the bottom side of the base material to protect it from oxidation during the lifetime of the reactor. This paper summarizes results from the in-vessel core catcher design and evaluation efforts, focusing on recently obtained results from materials interaction tests and prototypic testing activities.

Rempe, J.L. [Idaho National Engineering and Environmental Laboratory (United States); Condie, K.G. [Idaho National Engineering and Environmental Laboratory (United States); Knudson, D.L. [Idaho National Engineering and Environmental Laboratory (United States); Suh, K.Y. [Seoul National University (Korea, Republic of); Cheung, F.B. [The Pennsylvania State University (United States); Kim, S.B. [Korea Atomic Energy Research Institute (Korea, Republic of)

2005-11-15T23:59:59.000Z

380

Trojan Nuclear Power Plant Reactor Vessel and Internals Removal: Trojan Nuclear Plant Decommissioning Experience  

Science Conference Proceedings (OSTI)

One goal of the EPRI Decommissioning Technology Program is to capture the growing utility experience in nuclear plant decommissioning activities for the benefit of other utilities facing similar challenges in the future. This report provides historical information on the background, scope, organization, schedule, cost, contracts, and support activities associated with the Trojan Nuclear Plant Reactor Vessel and Internals Removal (RVAIR) Project. Also discussed are problems, successes, and lessons learned...

2000-10-16T23:59:59.000Z

Note: This page contains sample records for the topic "ocean-going vessels including" 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

BWRVIP-118: BWR Vessel and Internals Project: NMCA Experience Report and Applications Guidelines, 2003 Revision  

Science Conference Proceedings (OSTI)

The boiling water reactor (BWR) fleet has widely embraced noble metal chemical addition (NMCA) to provide protection against intergranular stress corrosion cracking (IGSCC). This report, prepared by a Boiling Water Reactor Vessel and Internals Project (BWRVIP) focus group, updates a report issued in 2001 that compiled data from plants operating on NMCA. It provides guidance for BWRs planning to implement NMCA, and information about expected plant response to operation with NMCA. It also identifies steps ...

2003-11-24T23:59:59.000Z

382

File:06HIGBoilerPressureVesselPermit.pdf | Open Energy Information  

Open Energy Info (EERE)

HIGBoilerPressureVesselPermit.pdf HIGBoilerPressureVesselPermit.pdf Jump to: navigation, search File File history File usage File:06HIGBoilerPressureVesselPermit.pdf Size of this preview: 463 × 599 pixels. Other resolution: 464 × 600 pixels. Full resolution ‎(1,275 × 1,650 pixels, file size: 47 KB, MIME type: application/pdf) File history Click on a date/time to view the file as it appeared at that time. Date/Time Thumbnail Dimensions User Comment current 09:08, 24 October 2012 Thumbnail for version as of 09:08, 24 October 2012 1,275 × 1,650 (47 KB) Dklein2012 (Talk | contribs) 12:32, 23 October 2012 Thumbnail for version as of 12:32, 23 October 2012 1,275 × 1,650 (47 KB) Dklein2012 (Talk | contribs) 16:30, 24 July 2012 Thumbnail for version as of 16:30, 24 July 2012 1,275 × 1,650 (44 KB) Alevine (Talk | contribs)

383

A Flaw Tolerance Approach to Address Reactor Vessel Head Penetration Cracking Issue  

SciTech Connect

Nickel-based alloys and the associated welds are susceptible to Primary Water Stress Corrosion Cracking. In Pressurized Water Reactor nuclear power plants, the reactor vessel closure head upper penetration nozzles used for the Control Rod Drive Mechanisms and other instrumentation systems are made of such nickel-based alloys. Cracking and leakage have been observed in the upper head penetration nozzles in nuclear power plants worldwide. Such cracking and the resulting leakage is a degradation of the reactor vessel pressure boundary. Regulatory requirements have been issued by the Nuclear Regulatory Commission regarding periodic inspection of the susceptible areas to enable detection of indications and provide reasonable assurance of continued structural integrity for reactor vessel closure head. A flaw tolerance approach has been used in the disposition of detected indications to minimize outage delays, by performing up-front fracture mechanics evaluations for the common types of indications detected in the susceptible areas. Details of the flaw tolerance approach are presented in this paper. (authors)

Ng, C. K.; Jirawongkraisorn, S.; Swamy, S. [Westinghouse Electric Company, LLC, Nuclear Services Division, P. O. Box 158, Madison, PA 15663 (United States)

2006-07-01T23:59:59.000Z

384

DESIGN OF A CONTAINMENT VESSEL CLOSURE FOR SHIPMENT OF TRITIUM GAS  

SciTech Connect

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

385

HYDROGEN EFFECTS ON THE BURST PROPERTIES OF TYPE 304L STAINLESS STEEL FLAWED VESSELS  

DOE Green Energy (OSTI)

The effect of hydrogen on the burst properties Type 304L stainless steel vessels was investigated. The purpose of the study was to compare the burst properties of hydrogen-exposed stainless steel vessels burst with different media: water, helium gas, or deuterium gas. A second purpose of the tests was to provide data for the development of a predictive finite-element model. The burst tests were conducted on hydrogen-exposed and unexposed axially-flawed cylindrical vessels. The results indicate that samples burst pneumatically had lower volume ductility than those tested hydraulically. Deuterium gas tests had slightly lower ductility than helium gas tests. Burst pressures were not affected by burst media. Hydrogen-charged samples had lower volume ductility and slightly higher burst pressures than uncharged samples. Samples burst with deuterium gas fractured by quasi-cleavage near the inside wall. The results of the tests were used to improve a previously developed predictive finite-element model. The results show that predicting burst behavior requires as a material input the effect of hydrogen on the plastic strain to fracture from tensile tests. The burst test model shows that a reduction in the plastic strain to fracture of the material will result in lower volume ductility without a reduction in burst pressure which is in agreement with the burst results.

Morgan, M; Monica Hall, M; Ps Lam, P; Dean Thompson, D

2008-03-27T23:59:59.000Z

386

Nondestructive Evaluation: Remote Field Technology Assessment for Piping Inspection Including Buried and Limited Access Components  

Science Conference Proceedings (OSTI)

This document provides results for the following projects: 1. Remote Field Technology Assessment for Piping Inspection 2. Inspection Techniques and NDE for Buried and Limited-Access Components 3. Guideline Development for Above-Ground, Below-Ground, and Limited-Access Storage Vessel Inspection These projects provided the Electric Power Research Institute (EPRI) the opportunity to engage its membership and several vendors in identifying remote field technology for piping inspection and advanced NDE inspec...

2010-11-19T23:59:59.000Z

387

Some mechanistic observations on the crack growth characteristics of pressure vessel and piping steels in PWR environment  

SciTech Connect

The fatigue crack growth behavior of A533B and A508 pressure vessel steel and AISI Types 304 and 316 steels used in reactor coolant piping have been studied in a pressurized water reactor environment at 288/sup 0/C (550/sup 0/F). The influence of stress ratio (P/sub min//P/sub max/), frequency, ramp times, specimen orientation and material microstructures were included in the study. While none of the materials showed evidence of static crack growth in the environment, the ferritic steels did show an enhanced fatigue crack growth rate at test frequencies of five cycles per minute and lower. Based on fractographic examinations the enhanced growth rate is not the result of environmentally induced intergranular or cleavage modes of crack propagation. Instead, striation spacing measurements were found to agree with the macroscopic crack growth rate, demonstrating a time dependent environmental interaction which introduces a frequency dependent enhancement of the mechanically developed striations. Crack growth experiments using hold times have confirmed the absence of any superimposed contribution of static crack growth components. Fatigue crack growth tests were conducted in an environment of Hydrogen Sulfide gas to establish the contribution of hydrogen embrittlement and will also be described.

Bamford, W.H.; Moon, D.M.

1979-01-01T23:59:59.000Z

388

A Multisensor Comparison of Ocean Wave Frequency Spectra from a Research Vessel during the Southern Ocean Gas Exchange Experiment  

Science Conference Proceedings (OSTI)

Obtaining accurate measurements of wave statistics from research vessels remains a challenge due to platform motion. One principal correction is the removal of ship heave and Doppler effects from point measurements. Here, open ocean wave ...

Alejandro Cifuentes-Lorenzen; James. B. Edson; Christopher J. Zappa; Ludovic Bariteau

389

Measurement of Wind Waves and Wave-Coherent Air Pressures on the Open Sea from a Moving SWATH Vessel  

Science Conference Proceedings (OSTI)

The design and implementation on a Small Waterline Area Twin Hull (SWATH) vessel of a complete system for measuring the directional distribution of wind waves and the concomitant fluctuations of air pressure and wind speed immediately above them ...

Mark A. Donelan; Fred W. Dobson; Hans C. Graber; Niels Madsen; Cyril McCormick

2005-07-01T23:59:59.000Z

390

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… (more)

DeWitt, Gregory L

2011-01-01T23:59:59.000Z

391

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

392

Program on Technology Innovation: Weld Metals and Welding Processes for Fabrication of Advanced Light Water Reactor Pressure Vessels  

Science Conference Proceedings (OSTI)

Light water reactors have traditionally been constructed using roll-formed plates for the reactor pressure vessel (RPV) shells, which were assembled via horizontal and vertical seam welds. Weld filler metals often contained significant quantities of copper, other residual elements such as vanadium, and nonmetallic elements such as phosphorous and sulfur. Low-alloy steel weld filler metals of this chemical composition contributed to the degree of neutron radiation-induced embrittlement of vessel ...

2013-06-26T23:59:59.000Z

393

BWRVIP-99-A: BWR Vessel and Internals Project, Crack Growth Rates in Irradiated Stainless Steels in BWR Internal Components  

Science Conference Proceedings (OSTI)

The BWR Vessel and Internals Project (BWRVIP) has developed methodologies to evaluate crack growth in internal components of stainless steel and nickel-base alloys in the BWR vessel. One BWRVIP report—BWRVIP-14—developed an approach to evaluate crack growth by intergranular stress corrosion cracking in austenitic stainless steel core shrouds exposed to a limited amount of neutron irradiation. Subsequently another report—BWRVIP-99—was prepared to provide a crack growth methodology applicable to irradiated...

2008-11-24T23:59:59.000Z

394

BWRVIP-86, Revision 1-A: BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan  

Science Conference Proceedings (OSTI)

This report describes the boiling water reactor (BWR) Integrated Surveillance Program (ISP). Based on recommendations from BWR Vessel and Internals Project (BWRVIP) utilities, it was concluded that combining all separate BWR surveillance programs into a single integrated program would be beneficial. In the integrated program, representative materials chosen for a specific reactor pressure vessel (RPV) can be materials from another plant surveillance program or other source that better represents the ...

2012-10-01T23:59:59.000Z

395

BWRVIP-41, Revision 3: BWR Vessel and Internals Project, BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines  

Science Conference Proceedings (OSTI)

The Boiling Water Reactor Vessel and Internals Project (BWRVIP), formed in June 1994, is an association of utilities focused exclusively on BWR vessel and internals issues. This BWRVIP report provides information on potential failure locations in BWR/3-6 jet pump components and recommends an inspection program designed to ensure that the integrity of all jet pump safety functions is maintained. EPRI published a previous version of this report as BWRVIP-41, Revision 2 (EPRI 1019570). This report (EPRI Rep...

2010-09-10T23:59:59.000Z

396

BWRVIP-18, Revision 1-A: BWR Vessel and Internals Project, BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines  

Science Conference Proceedings (OSTI)

The Boiling Water Reactor Vessel and Internals Project (BWRVIP), formed in June 1994, is an association of utilities focused exclusively on boiling water reactor (BWR) vessel and internals issues. This BWRVIP report contains generic guidelines that describe locations on the core spray piping and spargers for which inspection is needed, categories of plants for which inspection needs would differ, extent of inspection and reinspection for each location, and flaw evaluation procedures to determine ...

2012-04-09T23:59:59.000Z

397

Detection and characterization of flaws in segments of light water reactor pressure vessels  

Science Conference Proceedings (OSTI)

Studies have been conducted to determine flaw density in segments cut from light water reactor (LWR) pressure vessels as part of the Oak Ridge National Laboratory's Heavy-Section Steel Technology (HSST) Program. Segments from the Hope Creek Unit 2 vessil and the Pilgrim Unit 2 Vessel were purchased from salvage dealers. Hope Creek was a boiling water reactor (BWR) design and Pilgrim was a pressurized water reactor (PWR) design. Neither were ever placed in service. Objectives were to evaluate these LWR segments for flaws with ultrasonic and liquid penetrant techniques. Both objectives were successfully completed. One significant indication was detected in a Hope Creek seam weld by ultrasonic techniques and characterized by further analyses terminating with destructive correlation. This indication (with a through-wall dimension of approx.6 mm (approx.0.24 in.)) was detected in only 3 m (10 ft) of weldment and offers extremely limited data when compared to the extent of welding even in a single pressure vessel. However, the detection and confirmation of the flaw in the arbitrarily selected sections implies the Marshall report estimates (and others) are nonconservative for such small flaws. No significant indications were detected in the Pilgrim material by ultrasonic techniques. Unfortunately, the Pilgrim segments contained relatively little weldment; thus, we limited our ultrasonic examinations to the cladding and subcladding regions. Fluorescent liquid penetrant inspection of the cladding surfaces for both LWR segments detected no significant indications (i.e., for a total of approximately 6.8 m/sup 2/ (72 ft/sup 2/) of cladding surface).

Cook, K.V.; Cunningham, R.A. Jr.; McClung, R.W.

1987-01-01T23:59:59.000Z

398

Application of small specimens to fracture mechanics characterization of irradiated pressure vessel steels  

SciTech Connect

In this study, precracked Charpy V-notch (PCVN) specimens were used to characterize the fracture toughness of unirradiated and irradiated reactor pressure vessel steels in the transition region by means of three-point static bending. Fracture toughness at cleavage instability was calculated in terms of elastic-plastic K{sub Jc} values. A statistical size correction based upon weakest-link theory was performed. The concept of a master curve was applied to analyze fracture toughness properties. Initially, size-corrected PCVN data from A 533 grade B steel, designated HSST Plate O2, were used to position the master curve and a 5% tolerance bound for K{sub Jc} data. By converting PCVN data to IT compact specimen equivalent K{sub Jc} data, the same master curve and 5% tolerance bound curve were plotted against the Electric Power Research Institute valid linear-elastic K{sub Jc} database and the ASME lower bound K{sub Ic} curve. Comparison shows that the master curve positioned by testing several PCVN specimens describes very well the massive fracture toughness database of large specimens. These results give strong support to the validity of K{sub Jc} with respect to K{sub Ic} in general and to the applicability of PCVN specimens to measure fracture toughness of reactor vessel steels in particular. Finally, irradiated PCVN specimens of other materials were tested, and the results are compared to compact specimen data. The current results show that PCVNs demonstrate very good capacity for fracture toughness characterization of reactor pressure vessel steels. It provides an opportunity for direct measurement of fracture toughness of irradiated materials by means of precracking and testing Charpy specimens from surveillance capsules. However, size limits based on constraint theory restrict the operational test temperature range for K{sub Jc} data from PCVN specimens. 13 refs., 8 figs., 1 tab.

Sokolov, M.A.; Wallin, K.; McCabe, D.E.

1996-12-31T23:59:59.000Z

399

Dual shell reactor vessel: A pressure-balanced system for high pressure and temperature reactions  

Science Conference Proceedings (OSTI)

The main purpose of this work was to demonstrate the Dual Shell Pressure Balanced Vessel (DSPBV) as a safe and economical reactor for the hydrothermal water oxidation of hazardous wastes. Experimental tests proved that the pressure balancing piston and the leak detection concept designed for this project will work. The DSPBV was sized to process 10 gal/hr of hazardous waste at up to 399{degree}C (750{degree}F) and 5000 psia (34.5 MPa) with a residence time of 10 min. The first prototype reactor is a certified ASME pressure vessel. It was purchased by Innotek Corporation (licensee) and shipped to Pacific Northwest Laboratory for testing. Supporting equipment and instrumentation were, to a large extent, transported here from Battelle Columbus Division. A special air feed system and liquid pump were purchased to complete the package. The entire integrated demonstration system was assembled at PNL. During the activities conducted for this report, the leak detector design was tested on bench top equipment. Response to low levels of water in oil was considered adequate to ensure safety of the pressure vessel. Shakedown tests with water only were completed to prove the system could operate at 350{degree}C at pressures up to 3300 psia. Two demonstration tests with industrial waste streams were conducted, which showed that the DSPBV could be used for hydrothermal oxidation. In the first test with a metal plating waste, chemical oxygen demand, total organic carbon, and cyanide concentrations were reduced over 90%. In the second test with a munitions waste, the organics were reduced over 90% using H{sub 2}O{sub 2} as the oxidant.

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

1995-03-01T23:59:59.000Z

400

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

Science Conference Proceedings (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

Note: This page contains sample records for the topic "ocean-going vessels including" 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

Elliptical Local Vessel Density: a Fast and Robust Quality Metric for Fundus Images  

SciTech Connect

A great effort of the research community is geared towards the creation of an automatic screening system able to promptly detect diabetic retinopathy with the use of fundus cameras. In addition, there are some documented approaches to the problem of automatically judging the image quality. We propose a new set of features independent of Field of View or resolution to describe the morphology of the patient's vessels. Our initial results suggest that they can be used to estimate the image quality in a time one order of magnitude shorter respect to previous techniques.

Giancardo, Luca [ORNL; Chaum, Edward [ORNL; Karnowski, Thomas Paul [ORNL; Meriaudeau, Fabrice [ORNL; Tobin Jr, Kenneth William [ORNL; Abramoff, M.D. [University of Iowa

2008-01-01T23:59:59.000Z

402

Ex-vessel core catcher materials interactions. Annual progress report. [LMFBR  

SciTech Connect

A twelve-month program to investigate ex-vessel core catcher materials interactions has been completed. The investigations, involving depleted uranium dioxide, magnesia brick, stainless steel, and low-carbon steel, were conducted in furnaces and associated facilities existing at Aerospace, which were modified to process molten and solidified radioactive samples. In addition to developing efficient methods for the melting, pouring, and sustained heating of UO/sub 2/, extensive sample characterizations and microanalyses were performed. Theoretical analyses were also made in data interpretation for the purpose of understanding the interaction kinetics.

Swanson, D.G.; Zehms, E.H.; Ang, C.Y.; McClelland, J.D.; Meyer, R.A.; vanPaassen, H.L.L.

1976-10-30T23:59:59.000Z

403

EXPERIMENTAL RESULTS FOR THE ISOTOPIC EXCHANGE OF A 1600 LITER TITANIUM HYDRIDE STORAGE VESSEL  

Science Conference Proceedings (OSTI)

Titanium is used as a low pressure tritium storage material. The absorption/desorption rates and temperature rise during air passivation have been reported previously for a 4400 gram prototype titanium hydride storage vessel (HSV). A desorption limit of roughly 0.25 Q/M was obtained when heating to 700 C which represents a significant residual tritium process vessel inventory. To prepare an HSV for disposal, batchwise isotopic exchange has been proposed to reduce the tritium content to acceptable levels. A prototype HSV was loaded with deuterium and exchanged with protium to determine the effectiveness of a batch-wise isotopic exchange process. A total of seven exchange cycles were performed. Gas samples were taken nominally at the beginning, middle, and end of each desorption cycle. Sample analyses showed the isotopic exchange process does not follow the standard dilution model commonly reported. Samples taken at the start of the desorption process were lower in deuterium (the gas to be removed) than those taken later in the desorption cycle. The results are explained in terms of incomplete mixing of the exchange gas in the low pressure hydride.

Klein, J.

2010-12-14T23:59:59.000Z

404

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

SciTech Connect

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

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

2012-09-25T23:59:59.000Z

405

Consequence evaluation of radiation embrittlement of Trojan reactor pressure vessel supports  

SciTech Connect

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

406

Computation of initial stage of RBMK reactor fuel channel vessel rupture  

SciTech Connect

Objective of this work is estimation of temperature and time characteristics for rupture of the zirconium pipe which is the RBMK reactor fuel channel (FC) vessel under emergencies. As an emergency the zirconium pipe temperature rise process is considered which results in loss of pipe material strength properties and pipe rupture under the action of internal pressure P=80MPa. The work was carried out under Task Order 007 of University of California - VNIIEF Subcontract No. 0002P0004-95. The problem formulation is stated in Protocol (Task 3, Appendix 3) of the Russian-American Workshop which was held in December, 1994 in Los Alamos. Physical-mechanical and geometry characteristics of structure elements (FC vessel with graphite ring and graphite slug) are presented by NIKIET. The temperature mode of the structure is taken in conformity with the NIKIET data obtained with the RELAP5/MOD3 code. Numerical simulation of structure element behavior in an emergency is performed using the DRAKON program comlex oriented to solving strength problems for complex spatial structures at intense dynamic loading. The {open_quotes}DRAKON{close_quotes} program complex is described and compared with similar western codes in its capabilities.

Pevnitsky, A.V.; Solovyev, V.P.; Abakumov, A.I. [and others

1995-12-31T23:59:59.000Z

407

Helium leak testing of a radioactive contaminated vessel under high pressure in a contaminated environment  

Science Conference Proceedings (OSTI)

At ANL-W, with the shutdown of EBR-II, R&D has evolved from advanced reactor design to the safe handling, processing, packaging, and transporting spent nuclear fuel and nuclear waste. New methods of processing spent fuel rods and transforming contaminated material into acceptable waste forms are now in development. Storage of nuclear waste is a high interest item. ANL-W is participating in research of safe storage of nuclear waste, with the WIPP (Waste Isolation Pilot Plant) site in New Mexico the repository. The vessel under test simulates gas generated by contaminated materials stored underground at the WIPP site. The test vessel is 90% filled with a mixture of contaminated material and salt brine (from WIPP site) and pressurized with N2-1% He at 2500 psia. Test acceptance criteria is leakage jar method is used to determine leakage rate using a mass spectrometer leak detector (MSLD). The efficient MSLD and an Al bell jar replaced a costly, time consuming pressure decay test setup. Misinterpretation of test criterion data caused lengthy delays, resulting in the development of a unique procedure. Reevaluation of the initial intent of the test criteria resulted in leak tolerances being corrected and test efficiency improved.

Winter, M.E.

1996-10-01T23:59:59.000Z

408

Helium leak testing of a radioactive contaminated vessel under high pressure in a contaminated environment  

SciTech Connect

At ANL-W, with the shutdown of EBR-II, R&D has evolved from advanced reactor design to the safe handling, processing, packaging, and transporting spent nuclear fuel and nuclear waste. New methods of processing spent fuel rods and transforming contaminated material into acceptable waste forms are now in development. Storage of nuclear waste is a high interest item. ANL-W is participating in research of safe storage of nuclear waste, with the WIPP (Waste Isolation Pilot Plant) site in New Mexico the repository. The vessel under test simulates gas generated by contaminated materials stored underground at the WIPP site. The test vessel is 90% filled with a mixture of contaminated material and salt brine (from WIPP site) and pressurized with N2-1% He at 2500 psia. Test acceptance criteria is leakage < 10{sup -7} cc/seconds at 2500 psia. The bell jar method is used to determine leakage rate using a mass spectrometer leak detector (MSLD). The efficient MSLD and an Al bell jar replaced a costly, time consuming pressure decay test setup. Misinterpretation of test criterion data caused lengthy delays, resulting in the development of a unique procedure. Reevaluation of the initial intent of the test criteria resulted in leak tolerances being corrected and test efficiency improved.

Winter, M.E.

1996-10-01T23:59:59.000Z

409

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

SciTech Connect

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

410

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

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

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

2009-05-22T23:59:59.000Z

411

Documentation of Probabilistic Fracture Mechanics Codes Used for Reactor Pressure Vessels Subjected to Pressurized Thermal Shock Loa ding: Parts 1 and 2  

Science Conference Proceedings (OSTI)

Pressurized thermal shock (PTS) can impact the safety and operability of PWR vessels with significant radiation embrittlement in the vessel walls. This report documents the results of probabilistic fracture mechanics analysis benchmark studies performed to validate the use of several codes for evaluating vessel PTS. Such benchmark studies provide the industry with a standard reference method for verifying probabilistic fracture mechanics codes used in PTS analyses.

1995-08-08T23:59:59.000Z

412

BWRVIP-27-A: BWR Vessel and Internals Project, BWR Standby Liquid Control System / Core Plate Delta-P Inspection and Flaw Evaluation Guidelines  

Science Conference Proceedings (OSTI)

The Boiling Water Reactor Vessel and Internals Project (BWRVIP), formed in June, 1994, is an association of utilities focused exclusively on boiling water reactor (BWR) vessel and internals issues. This BWRVIP report defines inspection requirements for BWR standby liquid control (SLC) system piping from the vessel nozzle safe-end inward. A previous version of this report was published as BWRVIP-27 (TR-107286). This report (BWRVIP-27-A) incorporates changes proposed by the BWRVIP in response to "U.S. Nucl...

2003-08-04T23:59:59.000Z

413

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

SciTech Connect

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

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

1983-01-01T23:59:59.000Z

414

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

Science Conference Proceedings (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

415

Tank vessels of 20,000 dwt or more carrying oil in bulk. proposed design, equipment, and personnel standards  

SciTech Connect

The U.S. Coast Guard (USCG) proposes to add to the rules for tank vessels carrying oil in bulk, standards for segregated ballast tanks, dedicated clean ballast tanks, and crude washing systems, and to withdraw a previous proposal for double bottoms and segregated ballast tanks for certain tank vessels. Adoption of the proposal would reduce the probability of spillage of oil into the navigable waters of the U.S. from vessel accidents, reduce the amount of operational discharges into the oceans by deballasting and tank cleaning, and contribute to the conservation of oil. Written comments must be received by the USCG on or before 4/16/79. Public hearings will be held on 3/21/79 and 3/28/79.

1979-02-12T23:59:59.000Z

416

Materials Reliability Program: Inspection and Evaluation Guidelines for Reactor Vessel Bottom-Mounted Nozzles in U.S. PWR Plants (MR P-206)  

Science Conference Proceedings (OSTI)

This report presents inspection and evaluation guidelines for reactor vessel bottom-mounted nozzles in U.S. pressurized water reactor (PWR) plants.

2009-03-23T23:59:59.000Z

417

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

Science Conference Proceedings (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

418

Tumor blood vessel "normalization" improves the therapeutic efficacy of boron neutron capture therapy (BNCT) in experimental oral cancer  

Science Conference Proceedings (OSTI)

We previously demonstrated the efficacy of BNCT mediated by boronophenylalanine (BPA) to treat tumors in a hamster cheek pouch model of oral cancer with no normal tissue radiotoxicity and moderate, albeit reversible, mucositis in precancerous tissue around treated tumors. It is known that boron targeting of the largest possible proportion of tumor cells contributes to the success of BNCT and that tumor blood vessel normalization improves drug delivery to the tumor. Within this context, the aim of the present study was to evaluate the effect of blood vessel normalization on the therapeutic efficacy and potential radiotoxicity of BNCT in the hamster cheek pouch model of oral cancer.

D. W. Nigg

2012-01-01T23:59:59.000Z

419

ARM: Surface Radiation Measurement Quality Control testing, including climatologically configurable limits  

DOE Data Explorer (OSTI)

Surface Radiation Measurement Quality Control testing, including climatologically configurable limits

Gary Hodges; Tom Stoffel; Mark Kutchenreiter; Bev Kay; Aron Habte; Michael Ritsche; Victor Morris; Mary Anderberg

420

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

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

3-0501 3-0501 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 Sandia National Laboratories Albuquerque, New Mexico 87185 and Livermore, California 94550 Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000. Approved for public release; further dissemination unlimited. 2 Issued by Sandia National Laboratories, operated for the United States Department of Energy by Sandia Corporation. NOTICE: This report was prepared as an account of work sponsored by an agency of the

Note: This page contains sample records for the topic "ocean-going vessels including" 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
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421

Cells Forming Blood Vessels Send Their Copper to the Edge | Advanced Photon  

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

A Molecular Cause for One Form of Deafness A Molecular Cause for One Form of Deafness Water Theory is Watertight Nanowire Micronetworks from Carbon-Black Nanoparticles A Key Step in Repairing DNA Double-Strand Breaks An X-ray Rainbow Science Highlights Archives: 2013 | 2012 | 2011 | 2010 2009 | 2008 | 2007 | 2006 2005 | 2004 | 2003 | 2002 2001 | 2000 | 1998 | Subscribe to APS Science Highlights rss feed Cells Forming Blood Vessels Send Their Copper to the Edge FEBRUARY 20, 2007 Bookmark and Share Areas at the tips of HMVEC filopodia extensions were scanned by XFM at high resolution. The optical image is shown to the right and metal maps are shown to the left. False color images of P, Cu, and Zn are shown in the red, green and blue images respectively, and their overlay is shown to the lower right, demonstrating a transfer of cellular copper across the cell

422

Reactor pressure vessel integrity research at the Oak Ridge National Laboratory  

Science Conference Proceedings (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

423

ORNL/TM-2012/380 Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research  

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

2/380 2/380 Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program September 2012 Prepared by Cyrus Smith Randy Nanstad Robert Odette Dwight Clayton Katie Matlack Pradeep Ramuhalli Glenn Light DOCUMENT AVAILABILITY Reports produced after January 1, 1996, are generally available free via the U.S. Department of Energy (DOE) Information Bridge. Web site http://www.osti.gov/bridge Reports produced before January 1, 1996, may be purchased by members of the public from the following source. National Technical Information Service 5285 Port Royal Road Springfield, VA 22161 Telephone 703-605-6000 (1-800-553-6847) TDD 703-487-4639 Fax 703-605-6900

424

Fluid-structure-interaction analyses of reactor vessel using improved hybrid Lagrangian Eulerian code ALICE-II  

Science Conference Proceedings (OSTI)

This paper describes fluid-structure-interaction and structure response analyses of a reactor vessel subjected to loadings associated with postulated accidents, using the hybrid Lagrangian-Eulerian code ALICE-II. This code has been improved recently to accommodate many features associated with innovative designs of reactor vessels. Calculational capabilities have been developed to treat water in the reactor cavity outside the vessel, internal shield structures and internal thin shells. The objective of the present analyses is to study the cover response and potential for missile generation in response to a fuel-coolant interaction in the core region. Three calculations were performed using the cover weight as a parameter. To study the effect of the cavity water, vessel response calculations for both wet- and dry-cavity designs are compared. Results indicate that for all cases studied and for the design parameters assumed, the calculated cover displacements are all smaller than the bolts` ultimate displacement and no missile generation of the closure head is predicted. Also, solutions reveal that the cavity water of the wet-cavity design plays an important role of restraining the downward displacement of the bottom head. Based on these studies, the analyses predict that the structure integrity is maintained throughout the postulated accident for the wet-cavity design.

Wang, C.Y.

1993-06-01T23:59:59.000Z

425

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

SciTech Connect

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

426

Investigation of leaks in fiberglass-reinforced pressure vessels by direct observation of hollow fibers in glass cloth  

SciTech Connect

A simple method of visual observation of hollow fibers within fiberglass cloth has been developed. This visualization can aid in determining the contribution these fibers make toward leaks observed in fiberglass-reinforced epoxy resin pressure or vacuum vessels. Photographs and frequency data of these hollow fibers are provided. 3 figs.

McAdams, J.

1988-01-01T23:59:59.000Z

427

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

E-Print Network (OSTI)

During the spring of 1992, and again in the winter of 1993, seven graduate students from Texas A&M University's Nautical Archaeology Program participated in a project to document the Louisiana State Museum Vessel, an American Civil War-era submersible boat presently residing in the collections of the Louisiana State Museum in New Orleans. The project initially focused on providing archaeological documentation of the boat's design and construction characteristics, and on compiling some basic historical documentation regarding its known past. Since the turn of the century, this vessel has been presumed by many to be the New Orleans-built Confederate privateer Pioneer, which was scuttled at the time of the evacuation of New Orleans by Federal forces, and last reported in very close proximity to where the Louisiana State Museum Vessel was found in 1878. This was the assumption made at the time the documentation project was conducted, and based upon the information available at that time, an argument in support of this identification was published in a subsequent article summarizing the project's findings. Additional research has clearly determined that this vessel is not the Pioneer. Recent research also indicates that efforts to design, fabricate, and employ submersible vessels within the Confederate States were more widespread than conventionally believed. Furthermore, it now appears that the level of Confederate government support enjoyed by these efforts was more substantial than has traditionally been presumed. The Louisiana State Museum Vessel constitutes the oldest extant example of an important American watercraft tradition. The goals of this thesis are to historically and archaeologically document the Louisiana State Museum Vessel, and, as it cannot presently be identified, to establish likely historical candidates for potential association with it. In this process, the boat is used as a lens through which to view the larger picture of submersible watercraft development efforts undertaken within the Confederacy, and to discuss the collective body of antebellum American experiences and technical knowledge available to the Confederate submersible boat builders. It is also used as a vantage point from which to explore the relationship between Confederate and contemporaneous Federal submersible development efforts, and to acknowledge the common postwar legacy that emerged as a result of these parallel programs.

Wills, Richard Keith

2000-01-01T23:59:59.000Z

428

Results and Analyses of Irradiation/Anneal Experiments Conducted on Yankee Rowe Reactor Pressure Vessel Surrogate Materials: Yankee Atomic Electric Company Test Reactor Program  

Science Conference Proceedings (OSTI)

Many variables influence the response of reactor vessel steels to neutron irradiation. This study looks at the influence of irradiation temperature, steel heat treatment and microstructure, and nickel and phosphorus content on the irradiation response of high-copper reactor vessel steel. Also addressed are several studies evaluating the potential of thermal annealing to restore the mechanical properties of the steels tested.

1996-03-22T23:59:59.000Z

429

BWRVIP-241: BWR Vessel and Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Ve ssel Shell Welds and Nozzle Blend Radii  

Science Conference Proceedings (OSTI)

This report documents supplemental analyses for boiling water reactor (BWR) reactor pressure vessel (RPV) recirculation inlet and outlet nozzle-to-shell welds and nozzle inner radii to address limitations imposed by the U.S. Nuclear Regulatory Commission (NRC) regarding the reduction of inspections specified in Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code.

2010-10-26T23:59:59.000Z

430

Underwater radiated noise measurements of a noise?reduced research vessel: Comparison between a U.S. Navy noise range and a simple hydrophone mooring.  

Science Conference Proceedings (OSTI)

[A feasibility study was undertaken to characterize underwater radiated noise for a new class of noise?reduced fisheries research vessels using a field?deployable hydrophone system. Recent studies have demonstrated that vessel?radiated noise can impact the behavior of fish

Alex De Robertis; Christopher D. Wilson; Peter H. Dahl

2011-01-01T23:59:59.000Z

431

Underwater radiated noise measurements of a noise-reduced research vessel: comparison between a US Navy noise range and a simple hydrophone mooring.  

Science Conference Proceedings (OSTI)

A feasibility study was undertaken to characterize underwater radiated noise for a new class of noise-reduced fisheries research vessels using a field-deployable hydrophone system. Recent studies have demonstrated that vessel-radiated noise can impact the behavior of fish

Alex De Robertis; Christopher Wilson

2014-01-01T23:59:59.000Z

432

Model of the Feed Water System Including a Generic Model of the Deaerator for a Full Scope Combined Cycle Power Plant Simulator  

Science Conference Proceedings (OSTI)

This paper presents the modelling of the Feed water System and an original generic model for closed vessels containing a fluid in two phases at equilibrium conditions with an incondensable gas. The model was used for the deaerator of a Combined Cycle ... Keywords: deaerator, pressurised vessels model, feedwater simulation

Edgardo J. Roldan-Villasana; Ana K. Vazquez

2010-11-01T23:59:59.000Z

433

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 systems. Three small pyrex glass vessels were used to hold the refrigerant and allow for visualization studies of the flashing process. Baseline experiments were run varying initial pressure, initial refrigerant amount, orifice diameter, and vessel geometry. Three sets of experiments were run using two passive enhancement methods (the addition of steel balls and the addition of small amounts of oil) and one active enhancement method (the addition of an immersion heater). Furthermore, a lumped-parameter analytical model was developed from basic thermodynamic principles that predicted the rate of depressurization for the flashing refrigerant. The study showed that the initial refrigerant amount and the orifice size had the greatest influence on the mass flow and pressure characteristics during each sixty second test. The initial pressure and vessel volume had less of an impact under the conditions tested. Two of the enhancement methods consistently increased the amount of refrigerant flashed during the tests as compared to the baseline data for the same initial conditions. The addition a 1 cm layer of 3.6 mm steel balls to the base of the vessel increased the amount flashed from 21% to 81% and the addition of the 215-watt flat-spiral immersion heater the increased the amount flashed from 47% to 111 %. Foaming at the vapor-liquid interface was observed with the refrigerant-oil mixture experiments as two of the eight test conditions averaged an increase while six averaged a decrease, ranging from a 21% increase to a 27% decrease. The analytical depressurization model predicted general pressure and mass flux trends, and revisions to the model improved pressure predictions to within ±11%.

Nutter, Darin Wayne

1994-12-01T23:59:59.000Z

434

Bobbin-Tool Friction-Stir Welding of Thick-Walled Aluminum Alloy Pressure Vessels  

SciTech Connect

It was desired to assemble thick-walled Al alloy 2219 pressure vessels by bobbin-tool friction-stir welding. To develop the welding-process, mechanical-property, and fitness-for-service information to support this effort, extensive friction-stir welding-parameter studies were conducted on 2.5 cm. and 3.8 cm. thick 2219 Al alloy plate. Starting conditions of the plate were the fully-heat-treated (-T62) and in the annealed (-O) conditions. The former condition was chosen with the intent of using the welds in either the 'as welded' condition or after a simple low-temperature aging treatment. Since preliminary stress-analyses showed that stresses in and near the welds would probably exceed the yield-strength of both 'as welded' and welded and aged weld-joints, a post-weld solution-treatment, quenching, and aging treatment was also examined. Once a suitable set of welding and post-weld heat-treatment parameters was established, the project divided into two parts. The first part concentrated on developing the necessary process information to be able to make defect-free friction-stir welds in 3.8 cm. thick Al alloy 2219 in the form of circumferential welds that would join two hemispherical forgings with a 102 cm. inside diameter. This necessitated going to a bobbin-tool welding-technique to simplify the tooling needed to react the large forces generated in friction-stir welding. The bobbin-tool technique was demonstrated on both flat-plates and plates that were bent to the curvature of the actual vessel. An additional issue was termination of the weld, i.e. closing out the hole left at the end of the weld by withdrawal of the friction-stir welding tool. This was accomplished by friction-plug welding a slightly-oversized Al alloy 2219 plug into the termination-hole, followed by machining the plug flush with both the inside and outside surfaces of the vessel. The second part of the project involved demonstrating that the welds were fit for the intended service. This involved determining the room-temperature tensile and elastic-plastic fracture-toughness properties of the bobbin-tool friction-stir welds after a post-weld solution-treatment, quenching, and aging heat-treatment. These mechanical properties were used to conduct fracture-mechanics analyses to determine critical flaw sizes. Phased-array and conventional ultrasonic non-destructive examination was used to demonstrate that no flaws that match or exceed the calculated critical flaw-sizes exist in or near the friction-stir welds.

Dalder, E C; Pastrnak, J W; Engel, J; Forrest, R S; Kokko, E; Ternan, K M; Waldron, D

2007-06-06T23:59:59.000Z

435

DOE Order 440. 1 B: Worker Protection Program for DOE (Including...  

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

0. 1 B: Worker Protection Program for DOE (Including NNSA) Federal Employees DOE Order 440. 1 B: Worker Protection Program for DOE (Including NNSA) Federal Employees Stakeholders:...

436

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

Science Conference Proceedings (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°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic, or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development Program is responsible for performing research and development on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2¼Cr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have concluded, however, that with adequate engineered cooling of the vessel, the A508/533 steels are also acceptable.

J. K. Wright; R. N. Wright

2008-04-01T23:59:59.000Z

437

Impacts of Vessel Noise Perturbations on the Resident Sperm Whale Population in the Gulf of Mexico  

E-Print Network (OSTI)

The Gulf of Mexico is home to two of the world?s ten busiest ports by cargo volume, the Port of New Orleans and the Port of Houston; and in 2008, these ports hosted a combined 14,000 ships, a number which is likely only to increase. Past research shows that this increase in shipping worldwide has historically lead to an increase in ambient noise level of 3-5dB per decade. Sperm whales in the Gulf of Mexico are considered a genetically distinct, resident population. They have a preference for the Louisiana-Mississippi Shelf region which directly overlaps with the entrance to the Mississippi and the Port of New Orleans. Disruptions from vessel noise could influence feeding and breeding patterns essential to the health of the stock. Data used in this analysis were collected continuously over 36 days in the summer of 2001 from bottom moored Navy Environmental Acoustic Recording System (EARS) buoys. Results showed a significant difference (P<0.05) in noise level between hours with ships passing and hours without. Metrics for 56 ship passages were analyzed to compare duration of ship passage with duration of maximum received level (MRL) during ship passage. Results of that analysis showed an average ship passage of 29 minutes with average MRL lasting 23% of the ship passage and an average increase of 40dB. Lastly, click counts were made with the Pamguard. Click counts for ship passages were completed for 35 min and 17.5 min before and after the estimated closest point of approach (CPA) for each ship. Results showed a 36% decrease in the number of detectable clicks as a ship approaches when comparing clicks detected at intervals of both 35 minutes before and 17 minutes before the CPA; additionally, 22% fewer clicks were counted 30 min after the ship than 30 min before (results significant at the P=0.01 level). These results indicate a potential change in sperm whale behavior when exposed to large class size vessel traffic (e.g. tankers and container ships) from major shipping lanes. Recommendations for addressing this issue are discussed.

Azzara, Alyson

2012-05-01T23:59:59.000Z

438

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

SciTech Connect

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

439

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

E-Print Network (OSTI)

Composting is one of the techniques that has evolved as a safe disposal and predisposal alternative to the stringent regulations on hazardous waste disposal. The implementation of this technique needs careful evaluation of the processes a hazardous compound undergoes when subjected to composting. The purpose of this thesis is to define these processes and develop a model for determining the fate of organic compounds in waste during in-vessel composting Volatilization and biodegradation are found to be the major fate determining processes. Following mass balance approach the compound's loss through these processes is evaluated by developing a fate model. Fate of six aromatic compounds which fall into three categories-volatile, semi-volatile, and non volatile, is determined and the results compared to the experimental values for validating the model. A sensitivity analysis has been performed to determine which parameters most influence the model behavior and quantitatively describe their effects on model performance. The results obtained from the model show close agreement with the experimental results. More data is required to quantify the slight differences observed. The volatilization loss is found to exist only for first few hours. Biodegradation rates are found to have very little impact on volatilization of the compound. Air flow rate and volume of the waste are found to have a noticeable effect on the volatilization of a compound. Bulk density is found to effect volatilization to a small extent. Air quality control measures are recommended for the first few days to deal with the volatilized gases.

Bollineni, Prasanthi

1994-01-01T23:59:59.000Z

440

The development of an in-vessel cryopump system for the DIII-D tokamak  

Science Conference Proceedings (OSTI)

The design, testing and initial operation of the DIII-D advanced divertor cryocondensation pumping system is presented. The pump resides inside the tokamak plasma containment vessel where it provides particle exhaust pumping, and it is subjected to Joule heating and hot particle heat loads during each 10 second discharge. In addition, the pump must withstand plasma disruption induced electromagnetic forces and 400{degrees}C bake-out temperatures. Cooling is accomplished by forced flow liquid helium with the two-phase helium exhaust passing through a reliquefier for thermal efficiency. A prototype pump was constructed to study surface temperature rise as a function of flow geometry, applied heat load, helium mass flow rate, and pump outlet conditions. Prototype testing led to the development of a special geometry which was demonstrated to enhance two-phase flow stability and overall heat transfer. During initial operation, deuterium pumping speeds of 32,000 L/s at 2 mTorr pressure were achieved with a helium flow rate of 5 g/s. This speed was maintained during 300 W, 8 s long test beat pulses which meets operational goals.

Schaubel, K.M.; Baxi, C.B.; Campbell, G.L.; Laughon, G.J.; Mahdavi, M.A.; Makariou, C.C.; Smith, J.P.; Schaffer, M.J. [General Atomics, San Diego, CA (United States); Menon, M.M. [Oak Ridge National Lab., TN (United States)

1993-07-01T23:59:59.000Z

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


441

Metallographic and hardness examinations of TMI-2 lower pressure vessel head samples  

SciTech Connect

Fifteen steel samples were removed from the lower pressure vessel head of the damaged TMI-2 nuclear reactor to assess the thermal threat to the head posed by 15 to 20 metric tons of molten core debris relocating there during the accident. Full sections of thirteen of the samples and partial sections of the other two samples underwent hardness and metallographic examinations at the Idaho National Engineering Laboratory. These examinations have shown that eleven of the fifteen samples did not exceed the ferrite-austenite transformation temperature of 727 C during the accident. The remaining four samples did show evidence of having a much more severe thermal history. The samples from core grid positions F-10 and G-8 are believed to have experienced temperatures of 1,040 to 1,060 C for about 30 minutes. Samples from positions E-8 and E-6 appear to have been subjected to 1,075 to 1,100 C for approximately 30 minutes.

Korth, G. E. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

1994-03-01T23:59:59.000Z

442

Composite materials and bodies including silicon carbide and titanium diboride and methods of forming same  

DOE Patents (OSTI)

Methods of forming composite materials include coating particles of titanium dioxide with a substance including boron (e.g., boron carbide) and a substance including carbon, and reacting the titanium dioxide with the substance including boron and the substance including carbon to form titanium diboride. The methods may be used to form ceramic composite bodies and materials, such as, for example, a ceramic composite body or material including silicon carbide and titanium diboride. Such bodies and materials may be used as armor bodies and armor materials. Such methods may include forming a green body and sintering the green body to a desirable final density. Green bodies formed in accordance with such methods may include particles comprising titanium dioxide and a coating at least partially covering exterior surfaces thereof, the coating comprising a substance including boron (e.g., boron carbide) and a substance including carbon.

Lillo, Thomas M.; Chu, Henry S.; Harrison, William M.; Bailey, Derek

2013-01-22T23:59:59.000Z

443

The western river steamboat: structure and machinery, 1811 to 1860  

E-Print Network (OSTI)

The western river steamboat contained the technology that transformed the trans-Appalachian West from a wilderness to an economically significant region of the country. The following study explores the origin and development of this important steamboat type by analyzing archaeological data and historic sources. This information is used to create a thorough study of steamboat construction and machinery. The first steamboat on the western rivers was built by Robert Fulton in 1811. In the next decade many steamboats followed, but these vessels were not well-adapted to the shallow and swift rivers. Typically these steamboats had deep-drafted, stoutly constructed hulls, heavy low-pressure condensing engines, and many other features akin to ocean-going watercraft. In the 1820s, shipwrights began to adapt steamboat hull form and machinery to the river conditions. By the close of this decade the high-pressure engine was universally adopted for use on western steamboats because of its power, light weight, low cost, and ease of repair. Advancements in propulsion machinery were paralleled by the construction of shallow, flat-bottomed hulls and multiple decks rising high above the waterline. In the late 1830s or early 1840s, the construction of steamboats was materially advanced with the invention of hogging chains. These long iron rods prevented steamboat hulls from hogging or sagging, thereby allowing shipwrights to build vessels with lighter timbers, further reducing vessel draft. The first section of this thesis introduces the reader to the subject and outlines the sources consulted for this study, while Sections II and III present the historic context necessary for understanding the western river steamboat's historic importance. Sections IV through VI contain a detailed analysis of steamboat structure and machinery divided into chronological periods. Conclusions are presented in Section VII. Appendices include a table quantifying steamboat construction on western rivers and a table of measurements from steamboats that plied the Ohio River in 1850.

Kane, Adam Isaac

2001-01-01T23:59:59.000Z

444

A re-assembly and reconstruction of the 9th-century AD vessel wrecked off the coast of Bozburun, Turkey  

E-Print Network (OSTI)

In 1973, researchers from the Institute of Nautical Archaeology (INA) were led to the site of a wrecked ship by sponge diver Mehmet A??k??n, near his hometown of Bozburun, Turkey. During further monitoring over the following 21 years by INA, the site was identified as a merchant vessel dating from the 9th century AD. The excavation of the site by INA researchers and students from Texas A&M University occurred over four summer seasons, from 1995 to 1998, and yielded approximately 900 whole or nearly-whole amphorae, personal items, palynological material, and approximately 35 percent of the vessel??s wooden hull. This dissertation is a record of the curation, cataloging, analysis and re-assembly of the preserved elements of the Bozburun vessel??s hull, as well as a theoretical reconstruction of the entire vessel. The Bozburun vessel is unique as it is the only fully-excavated shipwreck from the 9th century AD, and is, indeed, a valuable source of examples of ship construction in the Mediterranean between the 7th and the 11th centuries AD. This dissertation, after discussing the methods of excavation and cataloging methods, posits the hypothesis that the techniques used to build this vessel represent a transitional stage in shipbuilding technology, combining distinctly old and new techniques. While the builders used embedded edge joinery in the ship??s planking, a very old method, they also appear to have used a conceptual framework and standards to design the vessel as well; methods evident in modified forms in Italian shipbuilding treatises from the Renaissance.

Harpster, Matthew Benjamin

2005-08-01T23:59:59.000Z

445

Occult Mediastinal Great Vessel Trauma: The Value of Aortography Performed During Angiographic Screening for Blunt Cervical Vascular Trauma  

SciTech Connect

Purpose. To determine the value of aortography in the assessment of occult aortic and great vessel injuries when routinely performed during screening angiography for blunt cerebrovascular injury (BCVI). Methods. One hundred and one consecutive patients who received both aortography and screening four-vessel angiography over 4 years were identified retrospectively. Angiograms for these patients were evaluated, and the incidence of occult mediastinal vascular injury was determined. Results. Of the 101 patients, 6 (6%) had angiographically documented traumatic aortic injuries. Of these 6 patients, one injury (17%) was unsuspected prior to angiography. Four of the 6 (67%) also had BCVI. One additional patient also had an injury to a branch of the subclavian artery. Conclusion. Routine aortography during screening angiography for BCVI is not warranted due to the low incidence (1%) of occult mediastinal arterial injury. However, in the setting of a BCVI screening study and no CT scan of the chest, aortography may be advantageous.

Ray, Charles E. [Denver Health Medical Center, Denver, Department of Interventional Radiology (United States)], E-mail: cray@dhha.org; Bauer, Jason R. [University of Colorado Health Sciences Center, Department of Diagnostic Radiology (United States); Cothren, C. Clay [Denver Health Medical Center, Department of Surgery (United States); Turner, James H. [Denver Health Medical Center, Denver, Department of Interventional Radiology (United States); Moore, Ernest E. [Denver Health Medical Center, Department of Surgery (United States)

2005-05-15T23:59:59.000Z

446

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

Science Conference Proceedings (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

447

TECHNICAL BASIS AND APPLICATION OF NEW RULES ON FRACTURE CONTROL OF HIGH PRESSURE HYDROGEN VESSEL IN ASME SECTION VIII, DIVISION 3 CODE  

DOE Green Energy (OSTI)

As a part of an ongoing activity to develop ASME Code rules for the hydrogen infrastructure, the ASME Boiler and Pressure Vessel Code Committee approved new fracture control rules for Section VIII, Division 3 vessels in 2006. These rules have been incorporated into new Article KD-10 in Division 3. The new rules require determining fatigue crack growth rate and fracture resistance properties of materials in high pressure hydrogen gas. Test methods have been specified to measure these fracture properties, which are required to be used in establishing the vessel fatigue life. An example has been given to demonstrate the application of these new rules.

Rawls, G

2007-04-30T23:59:59.000Z

448

BWRVIP-120: BWR Vessel and Internals Project - Radiolysis and ECP Improvements to the BWRVIA 2.0 Model  

Science Conference Proceedings (OSTI)

Two measures—moderate hydrogen injection (known as hydrogen water chemistry, or HWC) and noble metal chemical addition (NMCA)—have been applied in boiling water reactors (BWRs) to mitigate intergranular stress corrosion cracking (IGSCC) by lowering primary water electrochemical corrosion potential (ECP). This report describes improvements in radiolysis and ECP analysis made to the EPRI BWR Vessel and Internals Application (BWRVIA) model for use in evaluating HWC and NMCA application in BWRs.

2003-11-11T23:59:59.000Z

449

Toward free-surface modeling of planing vessels: simulation of the Fridsma hull using ALE-VMS  

Science Conference Proceedings (OSTI)

In this paper we focus on a class of applications involving surface vessels moving at high speeds, or "planing". We introduce a Fridsma planing hull benchmark problem, and simulate it using the finite-element-based ALE-VMS (Bazilevs et al. in Math Models ... Keywords: ALE-VMS, Finite elements, Fluid/rigid---body interaction, Free-surface flow, Fridsma planing hull, Level set, Ship hydrodynamics

I. Akkerman; J. Dunaway; J. Kvandal; J. Spinks; Y. Bazilevs

2012-12-01T23:59:59.000Z

450

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

Science Conference Proceedings (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

451

Feasibility of underwater welding of highly irradiated in-vessel components of boiling-water reactors: A literature review  

SciTech Connect

In February 1997, the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research (RES), initiated a literature review to assess the state of underwater welding technology. In particular, the objective of this literature review was to evaluate the viability of underwater welding in-vessel components of boiling water reactor (BWR) in-vessel components, especially those components fabricated from stainless steels that are subjected to high neutron fluences. This assessment was requested because of the recent increased level of activity in the commercial nuclear industry to address generic issues concerning the reactor vessel and internals, especially those issues related to repair options. This literature review revealed a preponderance of general information about underwater welding technology, as a result of the active research in this field sponsored by the U.S. Navy and offshore oil and gas industry concerns. However, the literature search yielded only a limited amount of information about underwater welding of components in low-fluence areas of BWR in-vessel environments, and no information at all concerning underwater welding experiences in high-fluence environments. Research reported by the staff of the U.S. Department of Energy (DOE) Savannah River Site and researchers from the DOE fusion reactor program proved more fruitful. This research documented relevant experience concerning welding of stainless steel materials in air environments exposed to high neutron fluences. It also addressed problems with welding highly irradiated materials, and primarily attributed those problems to helium-induced cracking in the material. (Helium is produced from the neutron irradiation of boron, an impurity, and nickel.) The researchers found that the amount of helium-induced cracking could be controlled, or even eliminated, by reducing the heat input into the weld and applying a compressive stress perpendicular to the weld path.

Lund, A.L.

1997-11-01T23:59:59.000Z

452

NP-3319, January 1984: Physically Based Regression Correlations of Embrittlement Data From Reactor Pressure Vessel Surveillance Programs  

Science Conference Proceedings (OSTI)

The first physically based model for forecasting the embrittlement behavior of irradiation-damaged steels proved much more accurate than earlier models. Thisadvance offers utilities greater precision in establishing operating pressure-temperature limits for PWRs and in assessing the ability of reactor vessels to withstand pressurized thermal shock transients.BackgroundBombardment by high-energy neutrons in the belt line of nuclear reactors can ...

1984-01-31T23:59:59.000Z

453

BWRVIP-275NP: BWR Vessel and Internals Project: Testing and Evaluation of the Susquehanna Unit 1 120° Capsule  

Science Conference Proceedings (OSTI)

In the late 1990s, a Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) was developed to improve the surveillance of the U.S. BWR fleet. This report describes testing and evaluation of the Susquehanna Unit 1 120° capsule. These results will be used to monitor embrittlement as part of the BWRVIP ISP.BackgroundThe BWRVIP ISP represents a major enhancement to the process of monitoring embrittlement for the U.S. ...

2013-10-16T23:59:59.000Z