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Note: This page contains sample records for the topic "walled pressure vessel" 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.


1

Sapphire tube pressure vessel  

DOE Patents [OSTI]

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

Outwater, John O. (Cambridge, MA)

2000-01-01T23:59:59.000Z

2

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

SciTech Connect (OSTI)

The objective of this work was to provide experimental heat transfer boundary condition and reactor pressure vessel (RPV) section thermal response data that can be used to benchmark computer codes that simulate thermal annealing of RPVS. This specific protect was designed to provide the Electric Power Research Institute (EPRI) with experimental data that could be used to support the development of a thermal annealing model. A secondary benefit is to provide additional experimental data (e.g., thermal response of concrete reactor cavity wall) that could be of use in an annealing demonstration project. The setup comprised a heater assembly, a 1.2 in {times} 1.2 m {times} 17.1 cm thick [4 ft {times} 4 ft {times} 6.75 in] section of an RPV (A533B ferritic steel with stainless steel cladding), a mockup of the {open_quotes}mirror{close_quotes} insulation between the RPV and the concrete reactor cavity wall, and a 25.4 cm [10 in] thick concrete wall, 2.1 in {times} 2.1 in [10 ft {times} 10 ft] square. Experiments were performed at temperature heat-up/cooldown rates of 7, 14, and 28{degrees}C/hr [12.5, 25, and 50{degrees}F/hr] as measured on the heated face. A peak temperature of 454{degrees}C [850{degrees}F] was maintained on the heated face until the concrete wall temperature reached equilibrium. Results are most representative of those RPV locations where the heat transfer would be 1-dimensional. Temperature was measured at multiple locations on the heated and unheated faces of the RPV section and the concrete wall. Incident heat flux was measured on the heated face, and absorbed heat flux estimates were generated from temperature measurements and an inverse heat conduction code. Through-wall temperature differences, concrete wall temperature response, heat flux absorbed into the RPV surface and incident on the surface are presented. All of these data are useful to modelers developing codes to simulate RPV annealing.

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

1994-11-01T23:59:59.000Z

3

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

4

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

5

Pressure Vessel Burst Program: Automated hazard analysis for pressure vessels  

SciTech Connect (OSTI)

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

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

1996-12-31T23:59:59.000Z

6

Tailoring Topology Optimization to Composite Pressure Vessel Design with Simultaneous  

E-Print Network [OSTI]

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

Paulino, Glaucio H.

7

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

8

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

9

Reactor pressure vessel. Status report  

SciTech Connect (OSTI)

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

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

1996-10-01T23:59:59.000Z

10

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

11

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

12

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

E-Print Network [OSTI]

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

Michaleris, Panagiotis

13

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

14

asme pressure vessels: Topics by E-print Network  

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

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

15

asme pressure vessel: Topics by E-print Network  

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

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

16

alloy pressure vessels: Topics by E-print Network  

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

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

17

alloy pressure vessel: Topics by E-print Network  

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

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

18

aged pressure vessel: Topics by E-print Network  

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

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

19

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

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

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

20

Radiation effects on reactor pressure vessel supports  

SciTech Connect (OSTI)

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

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

1996-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "walled pressure vessel" 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

Coalesced Martensite in Pressure Vessel Steels Hector Pous-Romero  

E-Print Network [OSTI]

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

Cambridge, University of

22

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

SciTech Connect (OSTI)

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

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

2006-07-01T23:59:59.000Z

23

Simulation of Diffusive Lithium Evaporation Onto the NSTX Vessel Walls  

SciTech Connect (OSTI)

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

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

2010-12-09T23:59:59.000Z

24

Reactor Pressure Vessel Head Packaging & Disposal  

SciTech Connect (OSTI)

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

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

2003-02-26T23:59:59.000Z

25

Tokamak reactor first wall  

DOE Patents [OSTI]

This invention relates to an improved first wall construction for a tokamak fusion reactor vessel, or other vessels subjected to similar pressure and thermal stresses.

Creedon, R.L.; Levine, H.E.; Wong, C.; Battaglia, J.

1984-11-20T23:59:59.000Z

26

Lightweight cryogenic-compatible pressure vessels for vehicular fuel storage  

DOE Patents [OSTI]

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

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

2004-03-23T23:59:59.000Z

27

Neutron shielding panels for reactor pressure vessels  

DOE Patents [OSTI]

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

Singleton, Norman R. (Murrysville, PA)

2011-11-22T23:59:59.000Z

28

Austenite Grain Growth in a Nuclear Pressure Vessel Steel  

E-Print Network [OSTI]

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

Cambridge, University of

29

Reversible high-pressure carbon nanotube vessel  

E-Print Network [OSTI]

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

Wang, Lifeng

30

International Hydrogen Fuel and Pressure Vessel Forum 2010 Beijing, China  

E-Print Network [OSTI]

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

31

Report of the terawatt laser pressure vessel committee  

SciTech Connect (OSTI)

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

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

2000-09-25T23:59:59.000Z

32

Beryllium pressure vessels for creep tests in magnetic fusion energy  

SciTech Connect (OSTI)

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

Neef, W.S.

1990-07-20T23:59:59.000Z

33

Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database  

SciTech Connect (OSTI)

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

Wang, Jy-An John [ORNL

2010-08-01T23:59:59.000Z

34

Fast neutron fluxes in pressure vessels using Monte Carlo methods  

SciTech Connect (OSTI)

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

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

1986-01-01T23:59:59.000Z

35

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

E-Print Network [OSTI]

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

36

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

SciTech Connect (OSTI)

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

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

2011-01-01T23:59:59.000Z

37

PRESSURIZATION OF CONTAINMENT VESSELS FROM PLUTONIUM OXIDE CONTENTS  

SciTech Connect (OSTI)

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

Hensel, S.

2012-03-27T23:59:59.000Z

38

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

SciTech Connect (OSTI)

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

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

2013-08-26T23:59:59.000Z

39

Fabrication Flaws in Reactor Pressure Vessel Repair Welds  

SciTech Connect (OSTI)

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

Schuster, George J.; Doctor, Steven R.

2007-12-01T23:59:59.000Z

40

A Survey of Pressure Vessel Code Compliance for Superconducting RF Cryomodules  

SciTech Connect (OSTI)

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

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

2011-06-07T23:59:59.000Z

Note: This page contains sample records for the topic "walled pressure vessel" 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

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

E-Print Network [OSTI]

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

Motta, Arthur T.

42

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

SciTech Connect (OSTI)

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

Cadwallader, Lee Charles; Nygren, R. E.

2001-10-01T23:59:59.000Z

43

Gamma ray-induced embrittlement of pressure vessel alloys  

SciTech Connect (OSTI)

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

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

1994-11-01T23:59:59.000Z

44

ASTM Standards for Reactor Dosimetry and Pressure Vessel Surveillance  

SciTech Connect (OSTI)

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

GRIFFIN, PATRICK J.

1999-09-14T23:59:59.000Z

45

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

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

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

46

Field measurement of lateral earth pressures on retaining walls  

E-Print Network [OSTI]

FIELD MEASUREMENT OF LATERAL EARTH PRESSURES ON RETAINING WALLS A Thesis by Michael Riggins Submitted to the Graduate College of Texas ARM University in partial fulfillment of the requirement for the degree of MASTER OF SCIENCE August 1974... Major Subject: Civil Engineering FIELD MEASUREMENT OF LATERAL EARTH PRESSURES ON RETAINING WALLS A Thesis by Michael Riggins Approved as to style and content by: Cha rman of Committee Memb r Head of Departm t P Etc Member August 1974 ABSTRACT...

Riggins, Michael

1974-01-01T23:59:59.000Z

47

Field measurements of earth pressure on a cantilever retaining wall  

E-Print Network [OSTI]

FIELD MEASUREMENTS OF EARTH PRESSURE ON A CANTILEVER RETAINING WALL A Thesis by LARRY WAYNE SCHULZE Submitted to the Graduate College Texas A&M University in Partial fulfillment of the requirements for the degree of MASTER OF SCIENCE... December 1980 Major Subject: Civil Engineering FIELD MEASUREMENTS OF EARTH PRESSURES ON A CANTILEVER RETAINING WALL A Thesis by LARRY WAYNE SCHULZE Approved as to style and content by: Harry M. Coyle ? C airman of Committee Wayne . Dunlap - Member...

Schulze, Larry Wayne

1980-01-01T23:59:59.000Z

48

Creep of A508/533 Pressure Vessel Steel  

SciTech Connect (OSTI)

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

Richard Wright

2014-08-01T23:59:59.000Z

49

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

E-Print Network [OSTI]

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

50

Modelling the Electron Beam Welding of Nuclear Reactor Pressure Vessel Steel  

E-Print Network [OSTI]

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

Cambridge, University of

51

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

E-Print Network [OSTI]

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

Giurgiutiu, Victor

52

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

SciTech Connect (OSTI)

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

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

2012-01-01T23:59:59.000Z

53

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

SciTech Connect (OSTI)

The objective of this project was to provide an assessment of several methods by which the temperature of a commercial nuclear power plant reactor pressure vessel (RPV) could be measured during an annealing process. This project was a coordinated effort between DOE`s Office of Nuclear Energy, Science and Technology; DOE`s Light Water Reactor Technology Center at Sandia National Laboratories; and the Electric Power Research Institute`s Non- Destructive Evaluation Center. Ball- thermocouple probes similar to those described in NUREG/CR-5760, spring-loaded, metal- sheathed thermocouple probes, and 1778 air- suspended thermocouples were investigated in experiments that heated a section of an RPV wall to simulate a thermal annealing treatment. A parametric study of ball material, emissivity, thermal conductivity, and thermocouple function locations was conducted. Also investigated was a sheathed thermocouple failure mode known as shunting (electrical breakdown of insulation separating the thermocouple wires). Large errors were found between the temperature as measured by the probes and the true RPV wall temperature during heat-up and cool-down. At the annealing soak temperature, in this case 454{degrees}C [850`F], all sensors measured the same temperature within about {plus_minus}5% (23.6{degrees}C [42.5{degrees}F]). Because of these errors, actual RPV wall heating and cooling rates differed from those prescribed (by up to 29%). Shunting does not appear to be a problem under these conditions. The large temperature measurement errors led to the development of a thermal model that predicts the RPV wall temperature from the temperature of a ball- probe. Comparisons between the model and the experimental data for ball-probes indicate that the model could be a useful tool in predicting the actual RPV temperature based on the indicated ball- probe temperature. The model does not predict the temperature as well for the spring-loaded and air suspended probes.

Acton, R.U.; Gill, W.; Sais, D.J.; Schulze, D.H.; Nakos, J.T. [Sandia National Labs., Albuquerque, NM (United States)

1996-05-01T23:59:59.000Z

54

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

SciTech Connect (OSTI)

The Soviet-designed VVER (Water-Cooled, Water-Moderated Energy Reactors) pressurized water reactors use horizontal steam generators to transfer energy from the primary to secondary coolant systems (DOE/NE-0084 Revision 2, 1989). Primary coolant flowing from the reactor vessel enters the steam generator through a vertical, circular, manifold header that also serves as the tubesheet distributing coolant to the horizontal tube bundle. Primary coolant exits the tube bundle and steam generator through a second similar vertical manifold header. The header design includes the provision for access by a person to inspect the mainfolds through bolted down closure heads atop each manifold. The internal diameter of each header exceeds that of the connected primary coolant system piping. The postulated failure of a manifold closure head or the manifold itself provides a pathway for primary coolant to enter the secondary system. Steam formation due to flashing of primary coolant inside the steam generator secondary side region can result in pressurization of the steam generator shell to values above the nominal secondary side operating pressure. The present work involves the investigation of the consequences of manifold failure for the case of the VVER-440 reactor system. An analysis has been performed of the loadings upon and the mechanical response of the steam generator shell for the case of a postulated large break in the manifold wall. The objectives were to calculate the maximum pressure attained inside the shell and to predict the shell failure pressure as well as the failure mechanism. 6 refs., 8 figs., 1 tab.

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

1989-01-01T23:59:59.000Z

55

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

SciTech Connect (OSTI)

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

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

2010-04-13T23:59:59.000Z

56

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

SciTech Connect (OSTI)

During cavern leaching in the Strategic Petroleum Reserve (SPR), injected raw water mixes with resident brine and eventually interacts with the cavern salt walls. This report provides a record of data acquired during a series of experiments designed to measure the leaching rate of salt walls in a labscale simulated cavern, as well as discussion of the data. These results should be of value to validate computational fluid dynamics (CFD) models used to simulate leaching applications. Three experiments were run in the transparent 89-cm (35-inch) ID diameter vessel previously used for several related projects. Diagnostics included tracking the salt wall dissolution rate using ultrasonics, an underwater camera to view pre-installed markers, and pre- and post-test weighing and measuring salt blocks that comprise the walls. In addition, profiles of the local brine/water conductivity and temperature were acquired at three locations by traversing conductivity probes to map out the mixing of injected raw water with the surrounding brine. The data are generally as expected, with stronger dissolution when the salt walls were exposed to water with lower salt saturation, and overall reasonable wall shape profiles. However, there are significant block-to-block variations, even between neighboring salt blocks, so the averaged data are considered more useful for model validation. The remedial leach tests clearly showed that less mixing and longer exposure time to unsaturated water led to higher levels of salt wall dissolution. The data for all three tests showed a dividing line between upper and lower regions, roughly above and below the fresh water injection point, with higher salt wall dissolution in all cases, and stronger (for remedial leach cases) or weaker (for standard leach configuration) concentration gradients above the dividing line.

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

2010-10-01T23:59:59.000Z

57

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

E-Print Network [OSTI]

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

Paris-Sud XI, Université de

58

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

59

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

SciTech Connect (OSTI)

Neutron embrittlement of reactor pressure vessels (RPVs) is an unresolved issue for light water reactor life extension, especially since transition temperature shifts (TTS) must be predicted for high 80-year fluence levels up to approximately 1,020 n/cm{sup 2}, far beyond the current surveillance database. Unfortunately, TTS may accelerate at high fluence, and may be further amplified by the formation of late blooming phases that result in severe embrittlement even in low-copper (Cu) steels. Embrittlement by this mechanism is a potentially significant degradation phenomenon that is not predicted by current regulatory models. This project will focus on accurately predicting transition temperature shifts at high fluence using advanced physically based, empirically validated and calibrated models. A major challenge is to develop models that can adjust test reactor data to account for flux effects. Since transition temperature shifts depend on synergistic combinations of many variables, flux-effects cannot be treated in isolation. The best current models systematically and significantly under-predict transition temperature at high fluence, although predominantly for irradiations at much higher flux than actual RPV service. This project will integrate surveillance, test reactor and mechanism data with advanced models to address a number of outstanding RPV embrittlement issues. The effort will include developing new databases and preliminary models of flux effects for irradiation conditions ranging from very low (e.g., boiling water reactor) to high (e.g., accelerated test reactor). The team will also develop a database and physical models to help predict the conditions for the formation of Mn-Ni-Si late blooming phases and to guide future efforts to fully resolve this issue. Researchers will carry out other tasks on a best-effort basis, including prediction of transition temperature shift attenuation through the vessel wall, remediation of embrittlement by annealing, and fracture toughness master curve issues.

Odette, G. Robert; Yamamoto, Takuya

2013-06-17T23:59:59.000Z

60

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

SciTech Connect (OSTI)

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

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

1999-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "walled pressure vessel" 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

Experiment Hazard Class 5.3 High Pressure Vessels  

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

stresses calculated using ASME Code Case 2286 July 17 1998. Verify that pressure relief devices have ASME "UV" certification or documentation of operability tests demonstrating...

62

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

SciTech Connect (OSTI)

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

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

1988-01-01T23:59:59.000Z

63

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

SciTech Connect (OSTI)

The feasibility of nuclear power plant lifetime extension was examined by reducing the fast neutron fluence at the reactor pressure vessel (RPV) and relieving irradiation embrittlement of materials, and thus ensuring enough structural integrity beyond the design lifetime. Two fluence reduction options, peripheral assembly replacement and additional shield installation in the outer core structures, were applied to the Kori Unit-1 reactor, and the fluence reduction effect was carefully analyzed. For an accurate estimate of the neutron fluence at the RPV and a reasonable description of the modified peripheral assemblies, a full-scope explicit modeling of a Monte Carlo simulation was employed in all calculations throughout this study. The Kori Unit-1 cycle-16 core was modeled on a three-dimensional representation by using the MCNP4B code, and the fluence distribution was estimated at the inner wall beltline around the circumferential weld of the RPV. On the basis of fracture toughness requirements of the RPV, the two modified cases were predicted to have an additional life of 7 to 10 effective full-power years. Throughout the core nuclear characteristics analyses, it was confirmed that the critical peaking factors for safe reactor operation were satisfied with the design limits.

Seo, Bo Kyun [Hanyang University (Korea, Republic of); Kim, Jong Kyung [Hanyang University (Korea, Republic of); Shin, Chang Ho [Hanyang University (Korea, Republic of); Kwon, Tae Je [Nuclear Fuel Company (Korea, Republic of)

2001-03-15T23:59:59.000Z

64

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

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

workshop in Washington, D.C. including testing and certification of Type 3 and Type 4 tanks, pressure relief device testing and validation, tank inspection, and end-of-life...

65

Reactor pressure vessel head vents and methods of using the same  

DOE Patents [OSTI]

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

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

2014-10-28T23:59:59.000Z

66

Flowfield and wall pressure characteristics downstream of a boundary layer suction device.  

E-Print Network [OSTI]

Flowfield and wall pressure characteristics downstream of a boundary layer suction device. Meagan A-dimensional slit can significantly reduce the fluctuating wall pressure immediately downstream of the suction slit momentum regions of the flow with the wall at the downstream edge of the suction slit. The third region

Tinney, Charles E.

67

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

E-Print Network [OSTI]

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

Chen, Sheng

68

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

E-Print Network [OSTI]

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

Tijsseling, A.S.

69

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

E-Print Network [OSTI]

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

70

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

E-Print Network [OSTI]

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

Barr, Al

71

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

SciTech Connect (OSTI)

Purpose: Atherosclerosis is the primary cause of heart disease and stroke. The detailed assessment of atherosclerosis of the carotid artery requires high resolution imaging of the vessel wall using multiple MR sequences with different contrast weightings. These images allow manual or automated classification of plaque components inside the vessel wall. Automated classification requires all sequences to be in alignment, which is hampered by patient motion. In clinical practice, correction of this motion is performed manually. Previous studies applied automated image registration to correct for motion using only nondeformable transformation models and did not perform a detailed quantitative validation. The purpose of this study is to develop an automated accurate 3D registration method, and to extensively validate this method on a large set of patient data. In addition, the authors quantified patient motion during scanning to investigate the need for correction. Methods: MR imaging studies (1.5T, dedicated carotid surface coil, Philips) from 55 TIA/stroke patients with ipsilateral <70% carotid artery stenosis were randomly selected from a larger cohort. Five MR pulse sequences were acquired around the carotid bifurcation, each containing nine transverse slices: T1-weighted turbo field echo, time of flight, T2-weighted turbo spin-echo, and pre- and postcontrast T1-weighted turbo spin-echo images (T1W TSE). The images were manually segmented by delineating the lumen contour in each vessel wall sequence and were manually aligned by applying throughplane and inplane translations to the images. To find the optimal automatic image registration method, different masks, choice of the fixed image, different types of the mutual information image similarity metric, and transformation models including 3D deformable transformation models, were evaluated. Evaluation of the automatic registration results was performed by comparing the lumen segmentations of the fixed image and moving image after registration. Results: The average required manual translation per image slice was 1.33 mm. Translations were larger as the patient was longer inside the scanner. Manual alignment took 187.5 s per patient resulting in a mean surface distance of 0.271 ± 0.127 mm. After minimal user interaction to generate the mask in the fixed image, the remaining sequences are automatically registered with a computation time of 52.0 s per patient. The optimal registration strategy used a circular mask with a diameter of 10 mm, a 3D B-spline transformation model with a control point spacing of 15 mm, mutual information as image similarity metric, and the precontrast T1W TSE as fixed image. A mean surface distance of 0.288 ± 0.128 mm was obtained with these settings, which is very close to the accuracy of the manual alignment procedure. The exact registration parameters and software were made publicly available. Conclusions: An automated registration method was developed and optimized, only needing two mouse clicks to mark the start and end point of the artery. Validation on a large group of patients showed that automated image registration has similar accuracy as the manual alignment procedure, substantially reducing the amount of user interactions needed, and is multiple times faster. In conclusion, the authors believe that the proposed automated method can replace the current manual procedure, thereby reducing the time to analyze the images.

Klooster, R. van 't; Staring, M.; Reiber, J. H. C.; Lelieveldt, B. P. F.; Geest, R. J. van der, E-mail: rvdgeest@lumc.nl [Department of Radiology, Division of Image Processing, Leiden University Medical Center, 2300 RC Leiden (Netherlands); Klein, S. [Department of Radiology and Department of Medical Informatics, Biomedical Imaging Group Rotterdam, Erasmus MC, Rotterdam 3015 GE (Netherlands)] [Department of Radiology and Department of Medical Informatics, Biomedical Imaging Group Rotterdam, Erasmus MC, Rotterdam 3015 GE (Netherlands); Kwee, R. M.; Kooi, M. E. [Department of Radiology, Cardiovascular Research Institute Maastricht, Maastricht University Medical Center, Maastricht 6202 AZ (Netherlands)] [Department of Radiology, Cardiovascular Research Institute Maastricht, Maastricht University Medical Center, Maastricht 6202 AZ (Netherlands)

2013-12-15T23:59:59.000Z

72

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

SciTech Connect (OSTI)

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

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

1995-12-01T23:59:59.000Z

73

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

SciTech Connect (OSTI)

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

Cheverton, R.D.

1996-08-01T23:59:59.000Z

74

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

SciTech Connect (OSTI)

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

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

2011-07-01T23:59:59.000Z

75

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

76

Interior Duct Wall Pressure Downstream of a Low-Speed Scott C. Morris  

E-Print Network [OSTI]

Interior Duct Wall Pressure Downstream of a Low-Speed Rotor Scott C. Morris , David B. Stephens The region downstream of a ducted rotor has been experimentally investigated in terms of its wake the description of the flow field and wall pressure in the region downstream of the rotor. Measurements involving

Alonso, Juan J.

77

Elastic properties and pressure-induced phase transitions of single-walled carbon nanotubes  

E-Print Network [OSTI]

Elastic properties and pressure-induced phase transitions of single-walled carbon nanotubes S-walled carbon nanotubes under hydrostatic pressure by first-princi- ples calculations. The circular tubes of carbon nanotubes has been studied with a variety of experimental techniques. Most of these studies seemed

Nabben, Reinhard

78

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

SciTech Connect (OSTI)

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

GJ Schuster, FA Simonen, SR Doctor

2008-04-01T23:59:59.000Z

79

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

DOE Patents [OSTI]

The wall basically consists of an array of small rectangular plates attached to the existing walls with threaded fasteners. The protective wall effectively conceals and protects all mounting hardware beneath the plate array, while providing a substantial surface area that will absorb plasma energy.

Phelps, R.D.; Upham, G.A.; Anderson, P.M.

1985-03-01T23:59:59.000Z

80

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

SciTech Connect (OSTI)

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

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

1994-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "walled pressure vessel" from the National Library of EnergyBeta (NLEBeta).
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to obtain the most current and comprehensive results.


81

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

SciTech Connect (OSTI)

The information fusion technique is used to develop radiation embrittlement prediction models for reactor pressure vessel (RPV) steels from U.S. power reactors, including boiling water reactors and pressurized water reactors. The Charpy transition temperature-shift data is used as the primary index of RPV radiation embrittlement in this study. Six parameters {Cu, Ni, P, neutron fluence, irradiation time, and irradiation temperature {are used in the embrittlement prediction models. The results indicate that this new embrittlement predictor achieved reductions of about 49.5% and 52% in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The implications of dose-rate effect and irradiation temperature effects for the development of radiation embrittlement models are also discussed.

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

2006-01-01T23:59:59.000Z

82

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

SciTech Connect (OSTI)

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

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

1996-12-01T23:59:59.000Z

83

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

SciTech Connect (OSTI)

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

Habil; Ziliukas, A.

1997-04-01T23:59:59.000Z

84

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

E-Print Network [OSTI]

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

American Society for Testing and Materials. Philadelphia

2009-01-01T23:59:59.000Z

85

Seismic Earth Pressure Development in Sheet Pile Retaining Walls: A Numerical Study  

E-Print Network [OSTI]

The design of retaining walls requires the complete knowledge of the earth pressure distribution behind the wall. Due to the complex soil-structure effect, the estimation of earth pressure is not an easy task; even in the static case. The problem becomes even more complex for the dynamic (i.e., seismic) analysis and design of retaining walls. Several earth pressure models have been developed over the years to integrate the dynamic earth pressure with the static earth pressure and to improve the design of retaining wall in seismic regions. Among all the models, MononobeOkabe (M-O) method is commonly used to estimate the magnitude of seismic earth pressures in retaining walls and is adopted in design practices around the world (e.g., EuroCode and Australian Standards). However, the M-O method has several drawbacks and does not provide reliable estimate of the earth pressure in many instances. This study investigates the accuracy of the M-O method to predict the dynamic earth pressure in sheet pile wall. A 2D pl...

Rajeev, P; Sivakugan, N

2015-01-01T23:59:59.000Z

86

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

SciTech Connect (OSTI)

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

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

2009-01-01T23:59:59.000Z

87

Radiation damage characterization in reactor pressure vessel steels with nonlinear ultrasound  

SciTech Connect (OSTI)

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

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

2014-02-18T23:59:59.000Z

88

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

SciTech Connect (OSTI)

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

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

1994-12-31T23:59:59.000Z

89

Wall-pressure and PIV analysis for microbubble drag reduction investigation  

E-Print Network [OSTI]

friction reductions were observed when the microbubbles were injected. Several measurements of wall-pressure were taken at various Reynolds numbers that ranged from 300 up to 6154. No significant drag reduction was observed for flows in the laminar range...

Dominguez Ontiveros, Elvis Efren

2005-11-01T23:59:59.000Z

90

Experimental measurement and analysis of wall pressure distribution for a 50% eccentric whirling annular seal  

E-Print Network [OSTI]

, whirl ratios ranging between ? 0.5 were tested. From the collected data a detailed analysis of wall pressures along the seal surface is performed following the technique described by Winslow (1994) and Robic (1999)....

Suryanarayanan, Arun

2004-11-15T23:59:59.000Z

91

Raman spectroscopy on single and multi-walled nanotubes under high pressure  

E-Print Network [OSTI]

dependence of the high-energy Raman modes in single and multi-walled carbon nanotubes was measuredRaman spectroscopy on single and multi-walled nanotubes under high pressure C. Thomsen, S. Reich, H properties of carbon nanotubes have become of scienti c interest since it was recognized that the low atomic

Nabben, Reinhard

92

Measurements of earth pressures for the design modification of cantilever retaining walls  

E-Print Network [OSTI]

MEASUREMENTS OF EARTH PRESSURES FOR THE DESIGN MODIFICATION OF CANTILEVER RETAINING WALLS A Thesis by William Pri kryl Submitted to the Graduate College Texas ASM University in partial fulfillment of the requirement for the degree of MASTER... OF SCIENCE May 1979 Major Subject: Civil Engineering MEASUREMENTS OF EARTH PRESSURES FOR THE DESIGN MODIFICATION OF CANTILEVER RETAINING WALLS A Thesis by William Prikryl Approved as to style and content by: jl ~, E Harry M. Coyle ? Chairman...

Prikryl, William

1979-01-01T23:59:59.000Z

93

New retaining wall design criteria based on lateral earth pressure measurements  

E-Print Network [OSTI]

NEW RETAINING WALL DESIGN CRITERIA BASED ON LATERAL EARTH PRESSURE MEASUREMENTS A Thesis by William V. ' Wright Submitted to the Graduate College of Texas A & M University in partial fulfillment of the requirement for the degree of MASTER... OF SCIENCE August 1975 Major Subject: Civil Engineering NEW RETAINING WALL DESIGN CRITERIA BASED ON LATERAL EARTH PRESSURE MEASUREMENTS A thesi s by William V. Wright (Chairman of Commi ttee) (M ber) (Head of Department) 0'( (Member) August 1975...

Wright, William Vincent

1975-01-01T23:59:59.000Z

94

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

E-Print Network [OSTI]

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

Motta, Arthur T.

95

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

SciTech Connect (OSTI)

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

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

2012-09-01T23:59:59.000Z

96

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

SciTech Connect (OSTI)

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

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

2011-07-01T23:59:59.000Z

97

Consequence evaluation of radiation embrittlement of Trojan reactor pressure vessel supports  

SciTech Connect (OSTI)

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

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

1990-10-01T23:59:59.000Z

98

Assemblies and methods for mitigating effects of reactor pressure vessel expansion  

DOE Patents [OSTI]

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

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

1999-01-01T23:59:59.000Z

99

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

SciTech Connect (OSTI)

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

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

1996-12-31T23:59:59.000Z

100

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

SciTech Connect (OSTI)

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

Pritam Chakraborty; S. Bulent Biner

2014-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "walled pressure vessel" 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

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

SciTech Connect (OSTI)

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

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

2013-01-01T23:59:59.000Z

102

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

SciTech Connect (OSTI)

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

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

1993-09-01T23:59:59.000Z

103

Reactor pressure vessel integrity research at the Oak Ridge National Laboratory  

SciTech Connect (OSTI)

Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. The RPV is the only key safety-related component of the plant for which a duplicate or redundant backup system does not exist. It is therefore imperative to understand and be able to predict the integrity inherent in the RPV. For this reason, the U.S. Nuclear Regulatory Commission has established the related research programs at ORNL described herein to provide for the development and confirmation of the methods used for: (1) establishing the irradiation exposure conditions within the RPV in the Embrittlement Data Base and Dosimetry Evaluation Program, (2) assessing the effects of irradiation on the RPV materials in the Heavy-Section Steel Irradiation Program, and (3) developing overall structural and fracture analyses of RPVs in the Heavy-Section Steel Technology Program.

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

1995-12-31T23:59:59.000Z

104

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

SciTech Connect (OSTI)

The Light Water Reactor Sustainability Program Plan is a collaborative program between the U.S. Department of Energy and the private sector directed at extending the life of the present generation of nuclear power plants to enable operation to at least 80 years. The reactor pressure vessel (RPV) is one of the primary components requiring significant research to enable such long-term operation. There are significant issues that need to be addressed to reduce the uncertainties in regulatory application, such as, 1) high neutron fluence/long irradiation times, and flux effects, 2) material variability, 3) high-nickel materials, 4)specimen size effects and the fracture toughness master curve, etc. The first issue is the highest priority to obtain the data and mechanistic understanding to enable accurate, reliable embrittlement predictions at high fluences. This paper discusses the major issues associated with long-time operation of existing RPVs and the LWRSP plans to address those issues.

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

2010-01-01T23:59:59.000Z

105

Assemblies and methods for mitigating effects of reactor pressure vessel expansion  

DOE Patents [OSTI]

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

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

1999-07-27T23:59:59.000Z

106

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

SciTech Connect (OSTI)

A severe accident in a nuclear power plant could result in the relocation of large quantities of molten core material onto the lower head of he reactor pressure vessel (RPV). In the absence of inherent cooling mechanisms, failure of the RPV ultimately becomes possible under the combined effects of system pressure and the thermal heat-up of the lower head. Sandia National Laboratories has performed seven experiments at 1:5th scale simulating creep failure of a RPV lower head. This paper describes a modeling program that complements the experimental program. Analyses have been performed using the general-purpose finite-element code ABAQUS-5.6. In order to make ABAQUS solve the specific problem at hand, a material constitutive model that utilizes temperature dependent properties has been developed and attached to ABAQUS-executable through its UMAT utility. Analyses of the LHF-1 experiment predict instability-type failure. Predicted strains are delayed relative to the observed strain histories. Parametric variations on either the yield stress, creep rate, or both (within the range of material property data) can bring predictions into agreement with experiment. The analysis indicates that it is necessary to conduct material property tests on the actual material used in the experimental program. The constitutive model employed in the present analyses is the subject of a separate publication.

Pilch, M.M.; Ludwigsen, J.S.; Chu, T.Y. [Sandia National Labs., Albuquerque, NM (United States); Rashid, Y.R. [Anatech, San Diego, CA (United States)

1998-08-01T23:59:59.000Z

107

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

E-Print Network [OSTI]

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

Michaleris, Panagiotis

108

High-R Walls for New Construction Structural Performance: Wind Pressure Testing  

SciTech Connect (OSTI)

This technical report is focused primarily on laboratory testing that evaluates wind pressure performance characteristics for wall systems constructed with exterior insulating sheathing. This research and test activity will help to facilitate the ongoing use of non-structural sheathing options and provide a more in-depth understanding of how wall system layers perform in response to high wind perturbations normal to the surface.

DeRenzis, A.; Kochkin, V.

2013-01-01T23:59:59.000Z

109

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

SciTech Connect (OSTI)

The complex nonlinear dependencies observed in typical reactor pressure vessel (RPV) material embrittlement data, as well as the inherent large uncertainties and scatter in the radiation embrittlement data, make prediction of radiation embrittlement a difficult task. Conventional statistical and deterministic approaches have only resulted in rather large uncertainties, in part because they do not fully exploit domain-specific mechanisms. The domain models built by researchers in the field, on the other hand, do not fully exploit the statistical and information content of the data. As evidenced in previous studies, it is unlikely that a single method, whether statistical, nonlinear, or domain model, will outperform all others. More generally, considering the complexity of the embrittlement prediction problem, it is highly unlikely that a single best method exists and is tractable, even in theory. In this paper, we propose to combine a number of complementary methods including domain models, neural networks, and nearest neighbor regressions (NNRs). Such a combination of methods has become possible because of recent developments in measurement-based optimal fusers in the area of information fusion. The information fusion technique is used to develop radiation embrittlement prediction models for reactor RPV steels from U.S. power reactors, including boiling water reactors and pressurized water reactors. The Charpy transition temperature-shift data is used as the primary index of RPV radiation embrittlement in this study. Six Cu, Ni, P, neutron fluence, irradiation time, and irradiation-parameters are used in the embrittlement prediction models. The results-temperature indicate that this new embrittlement predictor achieved reductions of about 49.5% and 52% in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The implications of dose-rate effect and irradiation temperature effects for the development of radiation embrittlement models are also discussed.

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

2007-01-01T23:59:59.000Z

110

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

SciTech Connect (OSTI)

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

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

1988-01-01T23:59:59.000Z

111

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

SciTech Connect (OSTI)

Formation of a high density of Ni-Mn-Si nm-scale precipitates in irradiated reactor pressure vessel steels, both with and without Cu, could lead to severe embrittlement. Models long ago predicted that these precipitates, which are not treated in current embrittlement regulations, would emerge only at high fluence. However, the mechanisms and variables that control Ni-Mn- Si precipitate formation, and their detailed characteristics, have not been well understood. High flux irradiations of six steels with systematic variations in Cu and Ni were carried out at ˜ 295±5°C to high and very high neutron fluences of ˜ 1.3x1020 and 1.1x1021 n/cm2. Atom probe tomography (APT) shows that significant mole fractions of these precipitates form in the Cu bearing steels at ˜ 1.3x1020 n/cm2, while they are only beginning to develop in Cu-free steels. However, large mole fractions, far in excess of those found in previous studies, are observed at 1.1x1021 n/cm2 at all Cu levels. The precipitates diffract, and in one case are compositionally and structurally consistent with the Mn6Ni16Si7 G-phase. At the highest fluence, the large precipitate mole fractions primarily depend on the steel Ni content, rather than Cu, and lead to enormous strength increases up to about 700 MPa. The implications of these results to light water reactor life extension are discussed briefly.

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

2014-11-01T23:59:59.000Z

112

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

SciTech Connect (OSTI)

The aim of the current study is the application of local micromechanical models to predict crack initiation in ductile materials. Two reactor pressure vessel materials have been selected for this study: JRQ IAEA monitor base metal (A533B Cl.1) and Doel-IV weld material. Charpy impact tests have been performed in both un-irradiated and irradiated conditions. In addition to standard tensile tests, notched tensile specimens have been tested. The upper shelf energy of the weld material remains almost un-affected by irradiation, whereas a decrease of 20% is detected for the base metal. Accordingly, the tensile properties of the weld material do not reveal a clear irradiation effect on the yield and ultimate stresses, this in contrast to the base material flow properties. The tensile tests have been analyzed in terms of micromechanical models. A good correlation is found between the standard tests and the micromechanical models, that are able to predict the ductile damage evolution in these materials. Additional information on the ductility behavior of these materials is revealed by this micromechanical analysis.

Chaouadi, R.; Walle, E. van; Fabry, A.; Velde, J. van de [SCK-CEN, Mol (Belgium); Meester, P. de [KUL, Heverlee (Belgium). Metals and Materials Science Dept.

1996-12-31T23:59:59.000Z

113

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

SciTech Connect (OSTI)

The results of reactor pressure vessel (RPV) annealing studies are interpreted in light of the current understanding of radiation embrittlement phenomena in RPV steels. An extensive RPV irradiation embrittlement and annealing database has been compiled and the data reveal that the majority of annealing studies completed to date have employed test reactor irradiated weldments. Although test reactor and power reactor irradiations result in similar embrittlement trends, subtle differences between these two damage states can become important in the interpretation of annealing results. Microstructural studies of irradiated steels suggest that there are several different irradiation-induced microstructural features that contribute to embrittlement. The amount of annealing recovery and the post-anneal re-embrittlement behavior of a steel are determined by the annealing response of these microstructural defects. The active embrittlement mechanisms are determined largely by the irradiation temperature and the material composition. Interpretation and thorough understanding of annealing results require a model that considers the underlying physical mechanisms of embrittlement. This paper presents a framework for the construction of a physically based mechanistic model of irradiation embrittlement and annealing behavior.

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

1996-12-31T23:59:59.000Z

114

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

SciTech Connect (OSTI)

One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPV) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes recent experimental results from work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response, or {open_quotes}recovery,{close_quotes} of several irradiated RPV steels; it also includes recent results from both ORNL and the Russian Research Center-Kurchatov Institute (RRC-KI) on a cooperative program of irradiation, annealing and reirradiation of both U.S. and Russian RPV steels. The cooperative program was conducted under the auspices of Working Group 3, U.S./Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS). The materials investigated are an RPV plate and various submerged-arc welds, with tensile, Charpy impact toughness, and fracture toughness results variously determined. Experimental results are compared with applicable prediction guidelines, while observed differences in annealing responses and reirradiation rates are discussed.

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

1996-12-31T23:59:59.000Z

115

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

SciTech Connect (OSTI)

A theoretical model is used to investigate the relative importance of point defect clusters (PDC) and copper-rich precipitates in reactor pressure vessel (RPV) embrittlement and to examine the influence of a broad range of irradiation and material parameters on predicted yield strength changes. The results indicate that there are temperature and displacement rate regimes wherein either CRP or PDC can dominate the material`s response to irradiation, with both interstitial and vacancy type defects contributing to the PDC component. The different dependencies of the CRP and PDC on temperature and displacement rate indicate that simple data extrapolations could lead to poor predictions of RPV embrittlement. It is significant that the yield strength changes predicted by the composite PDC/CRP model exhibit very little dependence on displacement rate below about 10{sup {minus}9} dpa/s. If this result is confirmed, concerns about accelerated displacement rates in power reactor surveillance programs should be minimized. The sensitivity of the model to microstructural parameters highlights the need for more detailed microstructural characterization of RPV steels.

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

1996-12-31T23:59:59.000Z

116

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

SciTech Connect (OSTI)

The Nuclear Regulatory Commission recently initiated a study by NIST to assess the feasibility of using physical-property measurements for evaluating radiation embrittlement in reactor pressure vessel (RPV) steels. Ultrasonic and magnetic measurements provide the most promising approaches for nondestructive characterization of RPV steels because elastic waves and magnetic fields can sense the microstructural changes that embrittle materials. The microstructural changes of particular interest are copper precipitation hardening, which is the likely cause of radiation embrittlement in RPV steels, and the loss of dislocation mobility that is an attribute of the ductile-to-brittle transition. Measurements were made on a 1% copper steel, ASTM grade A710, in the annealed, peak-aged and overaged conditions, and on an RPV steel, ASTM grade A533B. Nonlinear ultrasonic and micromagnetic techniques were the most promising measures of precipitation hardening. Ultrasonic velocity measurements and the magnetic properties associated with hysteresis-loop measurements were not particularly sensitive to either precipitation hardening or the ductile-to-brittle transition. Measurements of internal friction using trapped ultrasonic resonance modes detected energy losses due to the motion of pinned dislocations; however, the ultrasonic attenuation associated with these measurements was small compared to the attenuation caused by beam spreading that would occur in conventional ultrasonic testing of RPVs.

McHenry, H.I.; Alers, G.A. [National Inst. of Standards and Technology, Boulder, CO (United States). Materials Reliability Div.

1998-03-01T23:59:59.000Z

117

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

SciTech Connect (OSTI)

One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPV) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes recent experimental results from work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response, or {open_quotes}recovery,{close_quotes} of several irradiated RPV steels; it also includes recent results from both ORNL and the Russian Research Center-Kurchatov Institute (RRC-KI) on a cooperative program of irradiation, annealing and reirradiation of both U.S. and Russian RPV steels. The cooperative program was conducted under the auspices of Working Group 3, U.S./Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS). The materials investigated are an RPV plate and various submerged-arc welds, with tensile, Charpy impact toughness, and fracture toughness results variously determined. Experimental results are compared with applicable prediction guidelines, while observed differences in annealing responses and reirradiation rates are discussed.

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

1997-02-01T23:59:59.000Z

118

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

SciTech Connect (OSTI)

The International Atomic Energy Agency (IAEA) has conducted a series of Coordinated Research Projects (CRPs) that have focused on irradiated reactor pressure vessel (RPV) steel fracture toughness properties and approaches for assuring structural integrity of RPVs throughout operating life. A series of nine CRPs have been sponsored by the IAEA, starting in the early 1970s, focused on neutron radiation effects on RPV steels. The purpose of the CRPs was to develop comparisons and correlations to test the uniformity of irradiated results through coordinated international research studies and data sharing. Consideration of dose rate effects, effects of alloying (nickel, manganese, silicon, etc.) and residual elements (eg., copper and phosphorus), and drop in upper shelf toughness are also important for assessing neutron embrittlement effects. The ultimate use of embrittlement understanding is assuring structural integrity of the RPV under current and future operation and accident conditions. Material fracture toughness is the key ingredient needed for this assessment, and many of the CRPs have focused on measurement and application of irradiated fracture toughness. This paper presents an overview of the progress made since the inception of the CRPs in the early 1970s. The chronology and importance of each CRP have been reviewed and put into context for continued and long-term safe operation of RPVs.

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

2009-01-01T23:59:59.000Z

119

Dynamic pressure and shear stress measurements on the stator wall of whirling annular seals  

E-Print Network [OSTI]

Dynamic pressure and shear stress measurements on the stator wall of whirling annular seals are presented. Two flow conditions (Re=12,000 & 24,000), two seal speeds (Ta=3,300 & 6,600) and three eccentricity ratios (0, 10, & 50% of the clearance...

Winslow, Robert Bradley

1994-01-01T23:59:59.000Z

120

Experimental measurement of phase averaged wall-pressure distributions for a 25% eccentric whirling annular seal  

E-Print Network [OSTI]

Instantaneous wall-pressure data were recorded for a 25% eccentric whirling annular seal for rotor speeds of 1800RPM and 3600RPM, axial Reynolds numbers of 24000 and 12000, and whirl ratios of 0.1-1.0 following the procedure set forth by Winslow...

Cusano, Domenic

2006-08-16T23:59:59.000Z

Note: This page contains sample records for the topic "walled pressure vessel" 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

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

SciTech Connect (OSTI)

The U.S. Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°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

122

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

SciTech Connect (OSTI)

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

Rucinski, Russ; /Fermilab

1996-11-11T23:59:59.000Z

123

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

SciTech Connect (OSTI)

An industry-government collaborative program was carried out with an aim to promoting the acceptance of fracture mechanics based fitness-for-service assessment methodology for a service-damaged pressure vessel. A collaborative round robin exercise was carried out to predict the fracture behavior of a vessel containing hydrogen damage, fabrication related lack-of-fusion defects, an artificially induced fatigue crack and a localized thinned area. The fracture assessment procedures used include the US ASME Material Property Council`s PREFIS Program based on the British Standard (BS) Published Document (PD) 6493, ASME Section XI and The Central Electricity Generating Board (CEGB) R6 approach; The welding Institute (TWI) CRACKWISE program (based on BS PD6493 Level 2 approach), a variant of the R6 approach, J-tearing instability approaches, various J-estimation schemes, LEFM approach and simplified stress analysis. Assessments were compared with the results obtained from a hydrogen charged burst test of the vessel. Predictions, based on the J-tearing approach, compared well with the actual burst test results. Actual burst pressure was about five times the operating pressure.

Bhuyan, G.S. [Powertech Labs. Inc., Surrey, British Columbia (Canada); Sperling, E.J. [Amoco Corp., Naperville, IL (United States); Shen, G. [CANMET, Ottawa, Ontario (Canada). Metals Technology Labs.; Yin, H. [Mobil Research and Development Corp., Farmers Branch, TX (United States); Rana, M.D. [Praxair, Inc., Tonawanda, NY (United States)

1996-12-01T23:59:59.000Z

124

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

SciTech Connect (OSTI)

An industry-government collaborative program was carried out with an aim to promoting the acceptance of fracture mechanics-based fitness-for-service assessment methodology for a service-damaged pressure vessel. A collaborative round robin exercise was carried out to predict the fracture behavior of a vessel containing hydrogen damage, fabrication-related lack-of-fusion defects, an artificially induced fatigue crack, and a localized thinned area. The fracture assessment procedures used include the US ASME Material Property Council`s PREFIS Program based on the British Standard (BS) Published Document (PD) 6493, ASME Section XI and The Central Electricity Generating Board (CEGB) R6 approach, The Welding Institute (TWI) CRACKWISE program (based on BS PD6493 Level 2 approach), a variant of the R6 approach, J-tearing instability approaches, various J-estimation schemes, LEFM approach, and simplified stress analysis. Assessments were compared with the results obtained from a hydrogen-charged burst test of the vessel. Predictions, based on the J-tearing approach, compared well with the actual burst test results. Actual burst pressure was about five times the operating pressure.

Bhuyan, G.S. [Powertech Labs Inc., Surrey, British Columbia (Canada); Sperling, E.J. [BP-Amoco, Calgary, Alberta (Canada); Shen, G. [CANMET, Ottawa, Ontario (Canada). Metals Technology Labs.; Yin, H. [Mobil Technology Co., Dallas, TX (United States); Rana, M.D. [Praxair, Inc., Tonawanda, NY (United States)

1999-08-01T23:59:59.000Z

125

Field measurements of lateral earth pressures on a pre-cast panel retaining wall  

E-Print Network [OSTI]

OF SCIENCE August 1973 Major Subject: Civil Engineering FIELD MEASUREMENTS OF LATERAL EARTH PRESSURES ON A PRE-CAST PANEL RETAINING WALL A Thesis by DAVID MONROE PRESCOTT Approved as to sty1e and content by: Chai man o Co ttee~ ember ead of Departm... in Appendix I. (The style and format of this thesis follows that used by the Journal of the Soil Mechanics and Foundations Division, Procee ings, of t e American Society of Civil Engineers result of some large scale earth pressure tests at Massachusetts...

Prescott, David Monroe

1973-01-01T23:59:59.000Z

126

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

SciTech Connect (OSTI)

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

Sullivan, Edmund J.; Anderson, Michael T.

2014-06-10T23:59:59.000Z

127

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

SciTech Connect (OSTI)

In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for both the PBMR and prismatic design. The main focus of this report is the RPV for both design concepts with emphasis on material selection.

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

2007-03-21T23:59:59.000Z

128

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

SciTech Connect (OSTI)

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

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

1993-11-01T23:59:59.000Z

129

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

SciTech Connect (OSTI)

To establish the integrity of pressure water reactor (PWR) vessels it is necessary to determine the toughness of A508Cl.3 steel at the end of its life, that is after thermal aging and irradiation embrittlement. In safety analyses the toughness can be deduced from a reference curve set forth in the code (ASME or RCC-M). The validity of the reference curve has been verified for several years for unirradiated French reactor vessels. Tests were performed on specimens taken from materials having heterogeneities in chemical composition. For most of the test results the reference curve is a lower bound. To solve te problem of determining the toughness of SA 508 Cl.3 steel after irradiation and in the presence of possible heterogeneities, the toughness results were gathered. The synthesis shows that the RCC-M code curve is conservative.

Bethmont, M.; Frund, J.M. [Electricite de France, Moret-sur-Loing (France); Housin, B. [Framatome, Paris La Defense (France). Materials and Technology Dept.; Soulat, P. [Commissariat a l`Energie Atomique, Gif-sur-Yvette (France)

1996-12-31T23:59:59.000Z

130

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

SciTech Connect (OSTI)

A recent workshop on next-generation nuclear plant (NGNP) topics underscored the need for research studies on the creep fracture behavior of two materials under consideration for reactor pressure vessel (RPV) applications: 9Cr-1Mo and SA-5XX steels. This research project will provide a fundamental understanding of creep fracture behavior of modified 9Cr-1Mo steel welds for through modeling and experimentation and will recommend a design for an RPV structural health monitoring system. Following are the specific objectives of this research project: • Characterize metallurgical degradation in welded modified 9Cr-1Mo steel resulting from aging processes and creep service conditions. • Perform creep tests and characterize the mechanisms of creep fracture process. • Quantify how the microstructure degradation controls the creep strength of welded steel specimens. • Perform finite element (FE) simulations using polycrystal plasticity to understand how grain texture affects the creep fracture properties of welds. • Develop a microstructure-based creep fracture model to estimate RPVs service life . • Manufacture small, prototypic, cylindrical pressure vessels, subject them to degradation by aging, and measure their leak rates. • Simulate damage evolution in creep specimens by FE analyses. • Develop a model that correlates gas leak rates from welded pressure vessels with the amount of microstructural damage. • Perform large-scale FE simulations with a realistic microstructure to evaluate RPV performance at elevated temperatures and creep strength. • Develop a fracture model for the structural integrity of RPVs subjected to creep loads. • Develop a plan for a non-destructive structural health monitoring technique and damage detection device for RPVs.

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

2013-11-26T23:59:59.000Z

131

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

SciTech Connect (OSTI)

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

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

1989-01-01T23:59:59.000Z

132

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

SciTech Connect (OSTI)

Irradiation embrittlement of the reactor pressure vessel beltline materials of WWER-440/Type 213 reactors is monitored by a material irradiation surveillance program. Due to the high lead factor, the duration of the standard surveillance program is only five years, after which no further surveillance samples remain in the reactor. The large variation and uncertainty in neutron flux over the irradiated materials produce significant scatter in mechanical properties and necessitate a re-evaluation of results using gamma scanning, specimen reconstitution and recalculation. In order to provide information on the impact of changes in plant operation during later years a supplementary surveillance program has been devised.

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

1996-12-31T23:59:59.000Z

133

Application of Computational Physics: Blood Vessel Constrictions and Medical Infuses  

E-Print Network [OSTI]

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

Suprijadi,; Subekti, Petrus; Viridi, Sparisoma

2013-01-01T23:59:59.000Z

134

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

135

Pressure and impulse scaling methods for wall impact in ICF (inertial confinement fusion)  

SciTech Connect (OSTI)

The design of the first structural wall (FSW) in an inertial confinement fusion (ICF) reactor requires some knowledge of the expected wall loading produced by x-ray and neutron deposition; specifically in the High Yield Lithium Injection Fusion Energy (HYLIFE) reactor, wall loading results from two sources -- gas shock and liquid impact. Gas shock is derived from x-ray deposition in the thin layers of exposed blanket material, producing ionized vapor, which will generate gas shock on the FSW. Liquid impact, on the other hand, results from the acceleration of liquid blanket material by two possible forces -- the drag from vapor expansion through the blanket material and the neutron-induced isochoric disassembly process. Both impacts, however, are coupled by the interaction of hot gas expanding through the liquid blanket. This paper discusses scaling methods for estimating pressure and impulse on the HYLIFE FSW from these impacts. In particular, this paper reviews simple analytical and numerical techniques, and the use of experimental results in the estimation of wall impacts for the HYLIFE blanket geometry. Considered important in the analyses are supersonic flow through jet arrays and isochoric disassembly. Given the same initial parameters as those used in previous HYLIFE studies, the techniques described here yield results comparable to the previous studies utilizing heavy numerical simulation.

Liu, J.C.; Chen, X.M.; Schrock, V.E. (California Univ., Berkeley, CA (USA). Dept. of Nuclear Engineering); Orth, C.D. (Lawrence Livermore National Lab., CA (USA))

1990-01-01T23:59:59.000Z

136

Effect of confining wall potential on charged collimated dust beam in low-pressure plasma  

SciTech Connect (OSTI)

The effect of confining wall potential on charged collimated dust beam in low-pressure plasma has been studied in a dusty plasma experimental setup by applying electrostatic field to each channel of a multicusp magnetic cage. Argon plasma is produced by hot cathode discharge method at a pressure of 5×10{sup ?4} millibars and is confined by a full line cusped magnetic field confinement system. Silver dust grains are produced by gas-evaporation technique and move upward in the form of a collimated dust beam due to differential pressure maintained between the dust and plasma chambers. The charged grains in the beam after coming out from the plasma column enter into the diagnostic chamber and are deflected by a dc field applied across a pair of deflector plates at different confining potentials. Both from the amount of deflection and the floating potential, the number of charges collected by the dust grains is calculated. Furthermore, the collimated dust beam strikes the Faraday cup, which is placed above the deflector plates, and the current (?pA) so produced is measured by an electrometer at different confining potentials. The experimental results demonstrate the significant effect of confining wall potential on charging of dust grains.

Kausik, S. S.; Kakati, B.; Saikia, B. K. [Centre of Plasma Physics, Institute for Plasma Research, Sonapur 782 402 (India)] [Centre of Plasma Physics, Institute for Plasma Research, Sonapur 782 402 (India)

2013-05-15T23:59:59.000Z

137

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

SciTech Connect (OSTI)

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

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

1993-12-31T23:59:59.000Z

138

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

SciTech Connect (OSTI)

Fracture properties of Reactor Pressure Vessel (RPV) steels show large variations with changes in temperature and irradiation levels. Brittle behavior is observed at lower temperatures and/or higher irradiation levels whereas ductile mode of failure is predominant at higher temperatures and/or lower irradiation levels. In addition to such temperature and radiation dependent fracture behavior, significant scatter in fracture toughness has also been observed. As a consequence of such variability in fracture behavior, accurate estimates of fracture properties of RPV steels are of utmost importance for safe and reliable operation of reactor pressure vessels. A cohesive zone based approach is being pursued in the present study where an attempt is made to obtain a unified law capturing both stable crack growth (ductile fracture) and unstable failure (cleavage fracture). The parameters of the constitutive model are dependent on both temperature and failure probability. The effect of irradiation has not been considered in the present study. The use of such a cohesive zone based approach would allow the modeling of explicit crack growth at both stable and unstable regimes of fracture. Also it would provide the possibility to incorporate more physical lower length scale models to predict DBT. Such a multi-scale approach would significantly improve the predictive capabilities of the model, which is still largely empirical.

Pritam Chakraborty; S. Bulent Biner

2013-10-01T23:59:59.000Z

139

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

140

Crack opening area estimates in pressurized through-wall cracked elbows under bending  

SciTech Connect (OSTI)

One of the most important aspects in the leak-before-break approach is the estimation of the crack opening area corresponding to potential through-wall cracks at critical locations during plant operation. In order to provide a reasonable lower bound to the leak area under such loading conditions, numerous experimental and numerical programs have been developed in USA, U.K. and FRG and widely discussed in literature. This paper aims to extend these investigations on a class of pipe elbows characteristic of PWR main coolant piping. The paper is divided in three main parts. First, a new simplified estimation scheme for leakage area is described, based on the reference stress method. This approach mainly developed in U.K. and more recently in France provides a convenient way to account for the non-linear behavior of the material. Second, the method is carried out for circumferential through-wall cracks located in PWR elbows subjected to internal pressure. Finite element crack area results are presented and comparisons are made with our predictions. Finally, in the third part, the discussion is extended to elbows under combined pressure and in plane bending moment.

Franco, C.; Gilles, P.; Pignol, M.

1997-04-01T23:59:59.000Z

Note: This page contains sample records for the topic "walled pressure vessel" from the National Library of EnergyBeta (NLEBeta).
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141

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

SciTech Connect (OSTI)

A comprehensive neutron dosimetry experiment was made at one of the prime surveillance sites at the High Flux Isotope Reactor (HFIR) pressure vessel to aid radiation embrittlement studies of the vessel and to benchmark neutron transport calculations. The thermal neutron flux at the key 7, position 5 site was found, from measurements of radioactivation of four cobalt wires and four silver wires, to be 2.4 {times} 10{sup 12} n{center_dot}m{sup {minus}2}{center_dot}s{sup {minus}1}. The thermal flux derived from two helium accumulation monitors was 2.3 {times} 10{sup 12} n{center_dot}m{sup {minus}2}{center_dot}{sup {minus}1}. The thermal flux estimated by neutron transport calculations was 3.7 {times} 10{sup 12} n{center_dot}m{sup {minus}2}s{sup {minus}1}. The fast flux, >1 MeV, determined from two nickel activation wires, was 1.5 {times} 10{sup 12} n{center_dot}m{sup {minus}2}{center_dot}s{sup {minus}1}, in keeping with values obtained earlier from stainless steel surveillance monitors and with a computed value of 1.2 {times} 10{sup 13} n{center_dot}m{sup {minus}2}{center_dot}{sup {minus}1}. The fast fluxes given by two reaction-product-type monitors, neptunium-237 and beryllium, were 2.6 {times} 10{sup 13} n{center_dot}m{sup {minus}2}{center_dot}s {sup {minus}1} and 2.2 {times} 10{sup 13} n{center_dot}m{sup {minus}2}s{sup {minus}1}, respectively. Follow-up experiments indicate that these latter high values of fast flux are reproducible but are false; they are due to the creation of greater levels of reaction products by photonuclear events induced by an exceptionally high ratio of gamma flux to fast neutron flux at the vessel.

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

1994-01-01T23:59:59.000Z

142

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

SciTech Connect (OSTI)

The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750°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

143

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

SciTech Connect (OSTI)

Lincoln Composites operates a facility for designing, testing, and manufacturing composite pressure vessels. Lincoln Composites also has a U.S. Department of Energy (DOE)-funded project to develop composite tanks for high-pressure hydrogen storage. The initial stage of this project involves testing the permeation of high-pressure hydrogen through polymer liners. The company recently moved and is constructing a dedicated research/testing laboratory at their new location. In the meantime, permeation tests are being performed in a corner of a large manufacturing facility. The safety review team visited the Lincoln Composites site on May 25, 2010. The project team presented an overview of the company and project and took the safety review team on a tour of the facility. The safety review team saw the entire process of winding a carbon fiber/resin tank on a liner, installing the boss and valves, and curing and painting the tank. The review team also saw the new laboratory that is being built for the DOE project and the temporary arrangement for the hydrogen permeation tests.

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

2010-12-22T23:59:59.000Z

144

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

SciTech Connect (OSTI)

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

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

2012-09-17T23:59:59.000Z

145

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

SciTech Connect (OSTI)

The safety of commercial light-water nuclear plants is highly dependent on the structural integrity of the reactor pressure vessel (RPV). In the absence of radiation damage to the RPV, fracture of the vessel is difficult to postulate. Exposure to high energy neutrons can result in embrittlement of radiation-sensitive RPV materials. The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory, sponsored by the US Nuclear Regulatory Commission (USNRC), is assessing the effects of neutron irradiation on RPV material behavior, especially fracture toughness. The results of these and other studies are used by the USNRC in the evaluation of RPV integrity and regulation of overall nuclear plant safety. In assessing the effects of irradiation, prototypic RPV materials are characterized in the unirradiated condition and exposed to radiation under varying conditions. Mechanical property tests are conducted to provide data which can be used in the development of guidelines for structural integrity evaluations, while metallurgical examinations and mechanistic modeling are performed to improve understanding of the mechanisms responsible for embrittlement. The results of these investigations, in conjunction with results from commercial reactor surveillance programs, are used to develop a methodology for the prediction of radiation effects on RPV materials. This irradiation-induced degradation of the materials can be mitigated by thermal annealing, i.e., heating the RPV to a temperature above that of normal operation. Thus, thermal annealing and evaluation of reirradiation behavior are major tasks of the HSSI Program. This paper describes the HSSI Program activities by summarizing some past and recent results, as well as current and planned studies. 30 refs., 8 figs., 1 tab.

Nanstad, R.K.; Corwin, W.R.; Alexander, D.J.; Haggag, F.M.; Iskander, S.K.; McCabe, D.E.; Sokolov, M.A.; Stoller, R.E. [Oak Ridge National Lab., TN (United States). Metals and Ceramics Div.

1995-07-01T23:59:59.000Z

146

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

SciTech Connect (OSTI)

In a Pressurized Fluidized Bed Combustion (PFBC)/cyclone/filter system ground coal and sorbent are injected as pastes into the PFBC bed; the hot gases and entrained fine particles of ash and calcined or reacted sorbent are passed through a cyclone (which removes the larger entrained particles); and the very-fine particles that remain are then filtered out, so that the cleaned hot gas can be sent through a non-ruggedized hot-gas turbine. The 70 MWe Tidd PFBC Demonstration Plant in Brilliant, Ohio was completed in late 1990. The initial design utilized seven strings of primary and secondary cyclones to remove 98% of the particulate matter. However, the Plant also included a pressurized filter vessel, placed between the primary and secondary cyclones of one of the seven strings. Coal and dolomitic limestone (i.e, SO{sub 2} sorbent) of various nominal sizes ranging from 12 to 18 mesh were injected into the combustor operating at about 10 atm pressure and 925{degree}C. The cyclone removed elutriated particles larger than about 0.025 mm, and particles larger than ca. 0.0005 mm were filtered at about 750{degree}C by ceramic candle filters. Thus, the chemical reaction times and temperatures, masses of material, particle-size distributions, and chemical compositions were substantially different for particulates removed from the bed drain, the cyclone drain, and the filter unit. Accordingly, we have measured the particle-size distributions and concentrations of calcium, magnesium, sulfur, silicon, and aluminum for material taken from the three units, and also determined the chemical formulas and predominant crystalline forms of the calcium and magnesium sulfate compounds formed. The latter information is particularly novel for the filter-cake material, from which we isolated the ``new`` compound Mg{sub 2}Ca(SO{sub 4}){sub 3}.

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

1995-12-31T23:59:59.000Z

147

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

SciTech Connect (OSTI)

The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

2012-09-01T23:59:59.000Z

148

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

SciTech Connect (OSTI)

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

David H. Bothell

2000-04-30T23:59:59.000Z

149

Neutrino Factory Target Vessel  

E-Print Network [OSTI]

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

McDonald, Kirk

150

Vacuum Vessel Remote Handling  

E-Print Network [OSTI]

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

151

Progress in evaluation and improvement in nondestructive examination reliability for inservice inspection of Light Water Reactors (LWRs) and characterize fabrication flaws in reactor pressure vessels  

SciTech Connect (OSTI)

This paper is a review of the work conducted under two programs. One (NDE Reliability Program) is a multi-year program addressing the reliability of nondestructive evaluation (NDE) for the inservice inspection (ISI) of light water reactor components. This program examines the reliability of current NDE, the effectiveness of evolving technologies, and provides assessments and recommendations to ensure that the NDE is applied at the right time, in the right place with sufficient effectiveness that defects of importance to structural integrity will be reliably detected and accurately characterized. The second program (Characterizing Fabrication Flaws in Reactor Pressure Vessels) is assembling a data base to quantify the distribution of fabrication flaws that exist in US nuclear reactor pressure vessels with respect to density, size, type, and location. These programs will be discussed as two separate sections in this report. 4 refs., 7 figs.

Doctor, S.R.; Bowey, R.E.; Good, M.S.; Friley, J.R.; Kurtz, R.J.; Simonen, F.A.; Taylor, T.T.; Heasler, P.G.; Andersen, E.S.; Diaz, A.A.; Greenwood, M.S.; Hockey, R.L.; Schuster, G.J.; Spanner, J.C.; Vo, T.V.

1991-10-01T23:59:59.000Z

152

Seismic Earth Pressures on Retaining Structures and Basement Walls in Cohesionless Soils  

E-Print Network [OSTI]

? H) Normalized Dynamic Earth Pressure ( ? AE / ? H) Time(H=1 Time(sec) Figure A.64. Total earth pressure time series38 ii 3.7.3. Earth Pressure

Geraili Mikola, Roozbeh

2012-01-01T23:59:59.000Z

153

A scattered-light three-dimensional photoelastic stress analysis of a thick-walled pressure vessel  

E-Print Network [OSTI]

maintain temperature by turning off the heating elements when the temperature exceeds the preset value. This indicated that an exothermic reaction had taken place in the casting, which raised the temperature of the oven. The exothermic reaction was prob... mixture being high enough to sustain an exothermic reaction when placed in the oven. This casting was cured along with the others, but was not used for parts of the model that would undergo anal- ysis. An important secondary nb)ective of the research...

Lednicky, Edward Frank

1971-01-01T23:59:59.000Z

154

Development of automated welding process for field fabrication of thick walled pressure vessels. (First quarterly report, FY 1981)  

SciTech Connect (OSTI)

The choice of sets of root welding parameters is discussed. Thick field demonstration/qualification welds will be performed. A welding procedure handbook which will be prepared is mentioned. (DLC)

Schneider, U.A.

1981-01-01T23:59:59.000Z

155

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

SciTech Connect (OSTI)

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

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

1989-01-01T23:59:59.000Z

156

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

SciTech Connect (OSTI)

Document available in abstract form only, full text of document follows: The dosimetry from the H. B. Robinson Unit 2 Pressure Vessel Benchmark is analyzed with a suite of Westinghouse-developed codes and data libraries. The radiation transport from the reactor core to the surveillance capsule and ex-vessel locations is performed by RAPTOR-M3G, a parallel deterministic radiation transport code that calculates high-resolution neutron flux information in three dimensions. The cross-section library used in this analysis is the ALPAN library, an Evaluated Nuclear Data File (ENDF)/B-VII.0-based library designed for reactor dosimetry and fluence analysis applications. Dosimetry is evaluated with the industry-standard SNLRML reactor dosimetry cross-section data library. (authors)

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

2011-07-01T23:59:59.000Z

157

Modeling and simulation of a high pressure hydrogen storage tank with dynamic wall.  

E-Print Network [OSTI]

??Hydrogen storage is one of the divisions of hydrogen powered vehicles technology. To increase performances of high pressure hydrogen storage tanks, a multilayered design is… (more)

Cumalioglu, Ilgaz

2005-01-01T23:59:59.000Z

158

Modeling and simulation of a high pressure hydrogen storage tank with Dynamic Wall.  

E-Print Network [OSTI]

??Hydrogen storage is one of the divisions of hydrogen powered vehicles technology. To increase performances of high pressure hydrogen storage tanks, a multilayered design is… (more)

Cumalioglu, Ilgaz

2005-01-01T23:59:59.000Z

159

Very long single- and few-walled boron nitride nanotubes via the pressurized vapor/condenser method  

SciTech Connect (OSTI)

Boron nitride nanotubes (BNNTs) are desired for their exceptional mechanical, electronic, thermal, structural, textural, optical, and quantum properties. A new method for producing long, small-diameter, single- and few-walled, boron nitride nanotubes (BNNTs) in macroscopic quantities is reported. The pressurized vapor/condenser (PVC) method produces, without catalysts, highly crystalline, very long, small-diameter, BNNTs. Palm-sized, cotton-like masses of BNNT raw material were grown by this technique and spun directly into centimeters-long yarn. Nanotube lengths were observed to be 100 times that of those grown by the most closely related method. Self-assembly and growth models for these long BNNTs are discussed.

Michael W. Smith, Kevin Jordan, Cheol Park, Jae-Woo Kim, Peter Lillehei, Roy Crooks, Joycelyn Harrison

2009-11-01T23:59:59.000Z

160

The determination of the turbulent intensities in a transitional flow from a smooth to a rough wall with zero pressure gradient in a two-dimensional channel  

E-Print Network [OSTI]

THE DETERMINATION OF THE TURBULENT INTENSITIES IN A TRANSITIONAL FLOW FROM A SMOOTH TO A ROUGH WALL WITH ZERO PRESSURE GRADIENT IN A TWO-DIMENSIONAL CHANNEL A Thesis By Ol3AIDU I. ISLAM Submitted to the Graduate School of. tire Agricultural... WALL WITH ZERO PRESSURE GRADIENT IN A TWO DIMENSIONAL. GHANNEL A Thesis By OBAIDUL ISLAM Approved as to style and content by: F / F Ghairma p'f mm tg Head of Department May 1963 ACKNOWLEDGMENTS Grateful acknowledgment is made to the Texas...

Islam, Obaidul

1963-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "walled pressure vessel" from the National Library of EnergyBeta (NLEBeta).
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We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

Development of a methodology for the assessment of shallow-flaw fracture in nuclear reactor pressure vessels: Generation of biaxial shallow-flaw fracture toughness data  

SciTech Connect (OSTI)

A technology to determine shallow-flaw fracture toughness of reactor pressure vessel (RPV) steels is being developed for application to the safety assessment of RPVs containing postulated shallow-surface flaws. Shallow-flaw fracture toughness of RPV material has been shown to be higher than that for deep flaws, because of the relaxation of crack-tip constraint. This report describes the preliminary test results for a series of cruciform specimens with a uniform depth surface flaw. These specimens are all of the same size with the same depth flaw. Temperature and biaxial load ratio are the independent variables. These tests demonstrated that biaxial loading could have a pronounced effect on shallow-flaw fracture toughness in the lower transition temperature region for RPV materials. Through that temperature range, the effect of full biaxial (1:1) loading on uniaxial, shallow-flaw toughness varied from no effect near the lower shelf to a reduction of approximately 58% at higher temperatures.

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

1998-07-01T23:59:59.000Z

162

Review of industry efforts to manage pressurized water reactor feedwater nozzle, piping, and feedring cracking and wall thinning  

SciTech Connect (OSTI)

This report presents a review of nuclear industry efforts to manage thermal fatigue, flow-accelerated corrosion, and water hammer damage to pressurized water reactor (PWR) feedwater nozzles, piping, and feedrings. The review includes an evaluation of design modifications, operating procedure changes, augmented inspection and monitoring programs, and mitigation, repair and replacement activities. Four actions were taken: (a) review of field experience to identify trends of operating events, (b) review of technical literature, (c) visits to PWR plants and a PWR vendor, and (d) solicitation of information from 8 other countries. Assessment of field experience is that licensees have apparently taken sufficient action to minimize feedwater nozzle cracking caused by thermal fatigue and wall thinning of J-tubes and feedwater piping. Specific industry actions to minimize the wall-thinning in feedrings and thermal sleeves were not found, but visual inspection and necessary repairs are being performed. Assessment of field experience indicates that licensees have taken sufficient action to minimize steam generator water hammer in both top-feed and preheat steam generators. Industry efforts to minimize multiple check valve failures that have allowed backflow of steam from a steam generator and have played a major role in several steam generator water hammer events were not evaluated. A major finding of this review is that analysis, inspection, monitoring, mitigation, and replacement techniques have been developed for managing thermal fatigue and flow-accelerated corrosion damage to feedwater nozzles, piping, and feedrings. Adequate training and appropriate applications of these techniques would ensure effective management of this damage.

Shah, V.N.; Ware, A.G.; Porter, A.M.

1997-03-01T23:59:59.000Z

163

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

SciTech Connect (OSTI)

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

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

2002-07-01T23:59:59.000Z

164

Rheology and microstructural evolution in pressure-driven flow of a magnetorheological fluid with strong particle-wall interactions  

E-Print Network [OSTI]

The interaction between magnetorheological (MR) fluid particles and the walls of the device that retain the field-responsive fluid is critical as this interaction provides the means for coupling the physical device to the ...

Ocalan, Murat

165

Vessel structural support system  

DOE Patents [OSTI]

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

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

1992-01-01T23:59:59.000Z

166

Influence of the reactor wall composition on radicals' densities and total pressure in Cl{sub 2} inductively coupled plasmas: II. During silicon etching  

SciTech Connect (OSTI)

In an industrial inductively coupled plasma reactor dedicated to silicon etching in chlorine-based chemistry, the density of Cl{sub 2} molecules and the gas temperature are measured by means of laser absorption techniques, the density of SiCl{sub x} (x{<=}2) radicals by broadband absorption spectroscopy, the density of SiCl{sub 4} and ions by mass spectrometry, and the total gas pressure with a capacitance gauge. These measurements permit us to estimate the mole fractions of Cl, SiCl{sub 4}, and etch product radicals when etching a 200 mm diameter silicon wafer. The pure Cl{sub 2} plasma is operated in well prepared chamber wall coating with a thin film of SiOCl, AlF, CCl, or TiOCl. The impact of the chemical nature of the reactor wall's coatings on these mole fractions is studied systematically. We show that the reactor wall coatings have a huge influence on the radicals densities, but this is not only from the difference on Cl-Cl recombination coefficient on different surfaces. During silicon etching, SiCl{sub x} radicals sticking on the reactor walls are etched by Cl atoms and recycled into the plasma by forming volatile SiCl{sub 4}. Hence, the loss of Cl atoms in etching the wall deposited silicon is at least as important as their wall recombination in controlling the Cl atoms density. Furthermore, because SiCl{sub 4} is produced at high rate by both the wafer and reactor walls, it is the predominant etching product in the gas phase. However, the percentage of redeposited silicon that can be recycled into the plasma depends on the amount of oxygen present in the plasma: O atoms produced by etching the quartz roof window fix Si on the reactor walls by forming a SiOCl deposit. Hence, the higher the O density is, the lower the SiCl{sub 4} density will be, because silicon is pumped by the reactor walls and the SiOCl layer formed is not isotropically etched by chlorine. As a result, in the same pure Cl{sub 2} plasma at 20 mTorr, the SiCl{sub x} mole fraction can vary from 18% in a SiOCl-coated reactor, where the O density is the highest, to 62% in a carbon-coated reactor, where there is no O. In the latter case, most of the Cl mass injected in the reactor is stored in SiCl{sub 4} molecules, which results in a low silicon etch rate. In this condition, the Cl mass balance is verified within 10%, and from the silicon mass balance we concluded that SiCl{sub x} radicals have a high surface loss probability. The impact of the reactor wall coating on the etching process is thus important, but the mechanisms by which the walls control the plasma chemistry is much more complicated than a simple control through recombination reaction of halogen atoms on these surfaces.

Cunge, G.; Sadeghi, N.; Ramos, R. [Laboratoire des Technologies de la Microelectronique, CNRS, 17 rue des Martyrs (c/o CEA-LETI), 38054 Grenoble Cedex 9 (France); Laboratoire de Spectrometrie Physique (UMR 5588), Universite Joseph Fourier-Grenoble, and CNRS, BP 87, 38402 St. Martin d'Heres (France); Laboratoire des Technologies de la Microelectronique, CNRS, 17 rue des Martyrs (c/o CEA-LETI), 38054 Grenoble Cedex 9 (France)

2007-11-01T23:59:59.000Z

167

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

E-Print Network [OSTI]

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

Melcher, Peter

168

High-Pressure Tube Trailers and Tanks  

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

bending stress: continuous fiber vessels and vessels made of replicants Conformable tanks require internal stiffeners (ribs) to efficiently support the pressure and minimize...

169

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

E-Print Network [OSTI]

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

J. C. Misra; S. Maiti

2012-01-30T23:59:59.000Z

170

Influence of the reactor wall composition on radicals' densities and total pressure in Cl{sub 2} inductively coupled plasmas: I. Without silicon etching  

SciTech Connect (OSTI)

Laser absorption at 355 nm is used to monitor the time variations of the Cl{sub 2} density in high-density industrial inductively coupled plasma. This technique is combined with the measurement of the gas temperature from the Doppler width of the 811.5 nm line of argon, added as a trace gas and with the measurement of the total gas pressure with a Baratron gauge. These measurements permit to estimate the mole fractions of Cl{sub 2} and Cl species in Cl{sub 2} inductively coupled plasmas in a waferless reactor. The impact of the chemical nature of the reactor wall coatings on the Cl and Cl{sub 2} mole fractions is studied systematically. We show that under otherwise identical plasma conditions, the Cl mole fraction is completely different when the plasma is operated in SiOCl, AlF, CCl, or TiOCl coated reactors, because the homogeneous recombination probability of Cl atoms is strongly surface dependant. The Cl atom mole fraction reached at 100 W radiofrequency power in SiOCl coated reactor (80%) is much higher than that obtained at 900 W in a ''clean'' AlF reactor (40%). A simple zero-dimensional model permits to provide the recombination coefficient of Cl atoms, {gamma}{sub rec}: 0.005 on SiOCl film and about 0.3 on the other three coatings. It is proposed to get benefit of this very high sensitivity of Cl{sub 2} dissociation rate to the wall coating for the control of the chamber wall status from the Cl{sub 2} density measurements in standard conditions.

Cunge, G.; Sadeghi, N.; Ramos, R. [Laboratoire des Technologies de la Microelectronique, CNRS, 17 rue des Martyrs (c/o CEA-LETI), 38054 Grenoble Cedex 9 (France); Laboratoire de Spectrometrie Physique (UMR 5588), Universite Joseph Fourier-Grenoble, and CNRS, BP 87, 38402 St. Martin d'Heres (France); Laboratoire des Technologies de la Microelectronique, CNRS, 17 rue des Martyrs (c/o CEA-LETI), 38054 Grenoble Cedex 9 (France)

2007-11-01T23:59:59.000Z

171

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

SciTech Connect (OSTI)

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

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

1994-01-26T23:59:59.000Z

172

Ion transport membrane module and vessel system  

SciTech Connect (OSTI)

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

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

2012-02-14T23:59:59.000Z

173

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

174

Measurements of the hydrogenic recombination coefficient for the TFTR vacuum vessel  

SciTech Connect (OSTI)

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

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

1983-12-01T23:59:59.000Z

175

Applications of ENDF/B-VI and JENDL-3.1 iron data to reactor pressure vessel fluence analysis using continuous energy Monte Carlo code MCNP  

SciTech Connect (OSTI)

A comparison is made of results obtained from neutron transmissions analysis of RPV performed by MCNP with ENDF/B-VI and JENDL-3.1 iron data. At first, a one-dimensional discrete ordinates transport calculation using VITAMIN-C fine-group library based on ENDF/B-IV was performed for a cylindrical model of a PWR to generate the source spectrum at the front of the RPV. And then, the transmission of neutrons through RPV was calculated by MCNP with the moderated fission spectrum incident on the vessel face. For these ENDF/B-IV, -VI and JENDL-3.1 iron data were processed into continuous energy point data form by NJOY91.91. The fast neutron fluxes and dosimeter reaction rates through RPV using each iron data were intercompared.

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

1994-12-31T23:59:59.000Z

176

Ultrasonic Digital Communication System for a Steel Wall Multipath Channel: Methods and Results  

SciTech Connect (OSTI)

As of the development of this thesis, no commercially available products have been identified for the digital communication of instrumented data across a thick ({approx} 6 n.) steel wall using ultrasound. The specific goal of the current research is to investigate the application of methods for digital communication of instrumented data (i.e., temperature, voltage, etc.) across the wall of a steel pressure vessel. The acoustic transmission of data using ultrasonic transducers prevents the need to breach the wall of such a pressure vessel which could ultimately affect its safety or lifespan, or void the homogeneity of an experiment under test. Actual digital communication paradigms are introduced and implemented for the successful dissemination of data across such a wall utilizing solely an acoustic ultrasonic link. The first, dubbed the ''single-hop'' configuration, can communicate bursts of digital data one-way across the wall using the Differential Binary Phase-Shift Keying (DBPSK) modulation technique as fast as 500 bps. The second, dubbed the ''double-hop'' configuration, transmits a carrier into the vessel, modulates it, and retransmits it externally. Using a pulsed carrier with Pulse Amplitude Modulation (PAM), this technique can communicate digital data as fast as 500 bps. Using a CW carrier, Least Mean-Squared (LMS) adaptive interference suppression, and DBPSK, this method can communicate data as fast as 5 kbps. A third technique, dubbed the ''reflected-power'' configuration, communicates digital data by modulating a pulsed carrier by varying the acoustic impedance at the internal transducer-wall interface. The paradigms of the latter two configurations are believed to be unique. All modulation methods are based on the premise that the wall cannot be breached in any way and can therefore be viably implemented with power delivered wirelessly through the acoustic channel using ultrasound. Methods, results, and considerations for future research are discussed herein.

TL Murphy

2006-02-16T23:59:59.000Z

177

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

178

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

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

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

179

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

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

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

180

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

SciTech Connect (OSTI)

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

Simonen, Fredric A.

2001-05-31T23:59:59.000Z

Note: This page contains sample records for the topic "walled pressure vessel" 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

EDS V25 containment vessel explosive qualification test report.  

SciTech Connect (OSTI)

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

Rudolphi, John Joseph

2012-04-01T23:59:59.000Z

182

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

183

Simultaneous Irradiation and Imaging of Blood Vessels During Pulsed  

E-Print Network [OSTI]

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

Barton, Jennifer K.

184

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

SciTech Connect (OSTI)

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

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

2013-07-01T23:59:59.000Z

185

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

186

HFIR vessel probabilistic fracture mechanics analysis  

SciTech Connect (OSTI)

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

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

1997-01-01T23:59:59.000Z

187

Pressure suppression containment system  

DOE Patents [OSTI]

A pressure suppression containment system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and a gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The wetwell pool includes a plenum for receiving the non-condensable gas carried with steam from the drywell following a loss-of-coolant-accident (LOCA). The wetwell plenum is vented to a plenum above the GDCS pool following the LOCA for suppressing pressure rise within the containment vessel. A method of operation includes channeling steam released into the drywell following the LOCA into the wetwell pool for cooling along with the non-condensable gas carried therewith. The GDCS pool is then drained by gravity, and the wetwell plenum is vented into the GDCS plenum for channeling the non-condensable gas thereto. 6 figures.

Gluntz, D.M.; Townsend, H.E.

1994-03-15T23:59:59.000Z

188

Pressure suppression containment system  

DOE Patents [OSTI]

A pressure suppression containment system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and a gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The wetwell pool includes a plenum for receiving the non-condensable gas carried with steam from the drywell following a loss-of coolant-accident (LOCA). The wetwell plenum is vented to a plenum above the GDCS pool following the LOCA for suppressing pressure rise within the containment vessel. A method of operation includes channeling steam released into the drywell following the LOCA into the wetwell pool for cooling along with the non-condensable gas carried therewith. The GDCS pool is then drained by gravity, and the wetwell plenum is vented into the GDCS plenum for channeling the non-condensable gas thereto.

Gluntz, Douglas M. (San Jose, CA); Townsend, Harold E. (San Jose, CA)

1994-03-15T23:59:59.000Z

189

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

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

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

190

Experimental damage-gas flow correlations for cyclically loaded reinforced concrete walls  

E-Print Network [OSTI]

vessels. ” Cement and Concrete research, 32, XTRACT (2007).of air permeability in a concrete shear wall subjected tocharacteristics in cracked concrete. ” Nuclear Engineering

Soppe, Travis E.

2009-01-01T23:59:59.000Z

191

Bonfire Tests of High Pressure Hydrogen Storage Tanks  

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

Bonfire Tests of High Pressure Hydrogen Storage Tanks International Hydrogen Fuel and Pressure Vessel Forum 2010Beijing, P.R. China September 27, 2010 Bonfire Tests of High...

192

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

193

Microsoft Word - 911118_a Vessel_Alternatives-Judy1.doc  

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

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

194

Using SA508/533 for the HTGR Vessel Material  

SciTech Connect (OSTI)

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

Larry Demick

2012-06-01T23:59:59.000Z

195

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

196

Pressurized fluidized bed reactor  

DOE Patents [OSTI]

A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

Isaksson, J.

1996-03-19T23:59:59.000Z

197

Pressurized fluidized bed reactor  

DOE Patents [OSTI]

A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

Isaksson, Juhani (Karhula, FI)

1996-01-01T23:59:59.000Z

198

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

199

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

200

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

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


201

Pressure testing of torispherical heads  

SciTech Connect (OSTI)

Two vessels fabricated from SA516-70 steel with 6% knuckle radius torispherical heads were tested under internal pressure to failure. The D/t ratios of Vessel 1 and Vessel 2 were 238 and 185 respectively. The calculated maximum allowable working pressures of Vessel 1 and 2 heads using the ASME Section 8, Div. 1 rules and measured dimensions were 85 and 110 psi, respectively. Vessel 1 failed at a nozzle weld in the cylindrical shell at 700 psi pressure. Neither buckling nor any other objectionable deformation of the head was observed at a theoretical double-elastic-slope collapse pressure of 241 and a calculated buckling pressure of 270 psi. Buckles were observed developing slowly after 600 psi pressure, and a total of 22 buckles were observed after the test, having the maximum amplitude of 0.15 inch. Vessel 2 failed at the edge of the longitudinal weld of the cylindrical shell at 1,080 psi pressure. Neither buckling nor any other objectionable deformation of the head was observed up to the final pressure, which exceeded the theoretical double-elastic-slope collapse and calculated buckling pressures of 274 psi and 342 psi, respectively.

Rana, M.D. [Praxair, Inc., Tonawanda, NY (United States). Research and Development Dept.; Kalnins, A.; Updike, D.P. [Lehigh Univ., Bethlehem, PA (United States)

1995-12-01T23:59:59.000Z

202

Pressure sensor for sealed containers  

DOE Patents [OSTI]

A magnetic pressure sensor for sensing a pressure change inside a sealed container. The sensor includes a sealed deformable vessel having a first end attachable to an interior surface of the sealed container, and a second end. A magnet mounted to the vessel second end defining a distance away from the container surface provides an externally detectable magnetic field. A pressure change inside the sealed container causes deformation of the vessel changing the distance of the magnet away from the container surface, and thus the detectable intensity of the magnetic field.

Hodges, Franklin R. (Loudon, TN)

2001-01-01T23:59:59.000Z

203

Pressure suppression system  

DOE Patents [OSTI]

A pressure suppression system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and an enclosed gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The GDCS pool includes a plenum for receiving through an inlet the non-condensable gas carried with steam from the drywell following a loss-of-coolant accident (LOCA). A condenser is disposed in the GDCS plenum for condensing the steam channeled therein and to trap the non-condensable gas therein. A method of operation includes draining the GDCS pool following the LOCA and channeling steam released into the drywell following the LOCA into the GDCS plenum for cooling along with the non-condensable gas carried therewith for trapping the gas therein.

Gluntz, Douglas M. (San Jose, CA)

1994-01-01T23:59:59.000Z

204

Pressure suppression system  

DOE Patents [OSTI]

A pressure suppression system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and an enclosed gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The GDCS pool includes a plenum for receiving through an inlet the non-condensable gas carried with steam from the drywell following a loss-of-coolant accident (LOCA). A condenser is disposed in the GDCS plenum for condensing the steam channeled therein and to trap the non-condensable gas therein. A method of operation includes draining the GDCS pool following the LOCA and channeling steam released into the drywell following the LOCA into the GDCS plenum for cooling along with the non-condensable gas carried therewith for trapping the gas therein. 3 figs.

Gluntz, D.M.

1994-10-04T23:59:59.000Z

205

Initial conditioning of the TFTR vacuum vessel  

SciTech Connect (OSTI)

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

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

1984-01-01T23:59:59.000Z

206

Standard guide for mutual inductance bridge applications for wall thickness determinations in boiler tubing  

E-Print Network [OSTI]

1.1 This guide describes a procedure for obtaining relative wall thickness indications in ferromagnetic and non-ferromagnetic steels using the mutual inductance bridge method. The procedure is intended for use with instruments capable of inducing two substantially identical magnetic fields and noting the change in inductance resulting from differing amounts of steel. It is used to distinguish acceptable wall thickness conditions from those which could place tubular vessels or piping at risk of bursting under high temperature and pressure conditions. 1.2 This guide is intended to satisfy two general needs for users of industrial Mutual Inductance Bridge (MIB) equipment: (1) the need for a tutorial guide addressing the general principles of Mutual Inductance Bridges as they apply to industrial piping; and (2) the need for a consistent set of MIB performance parameter definitions, including how these performance parameters relate to MIB system specifications. Potential users and buyers, as well as experienced M...

American Society for Testing and Materials. Philadelphia

2010-01-01T23:59:59.000Z

207

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

208

Pressurizer tank upper support  

DOE Patents [OSTI]

A pressurizer tank in a pressurized water nuclear reactor is mounted between structural walls of the reactor on a substructure of the reactor, the tank extending upwardly from the substructure. For bearing lateral loads such as seismic shocks, a girder substantially encircles the pressurizer tank at a space above the substructure and is coupled to the structural walls via opposed sway struts. Each sway strut is attached at one end to the girder and at an opposite end to one of the structural walls, and the sway struts are oriented substantially horizontally in pairs aligned substantially along tangents to the wall of the circular tank. Preferably, eight sway struts attach to the girder at 90.degree. intervals. A compartment encloses the pressurizer tank and forms the structural wall. The sway struts attach to corners of the compartment for maximum stiffness and load bearing capacity. A valve support frame carrying the relief/discharge piping and valves of an automatic depressurization arrangement is fixed to the girder, whereby lateral loads on the relief/discharge piping are coupled directly to the compartment rather than through any portion of the pressurizer tank. Thermal insulation for the valve support frame prevents thermal loading of the piping and valves. The girder is shimmed to define a gap for reducing thermal transfer, and the girder is free to move vertically relative to the compartment walls, for accommodating dimensional variation of the pressurizer tank with changes in temperature and pressure.

Baker, Tod H. (O'Hara Township, Allegheny County, PA); Ott, Howard L. (Kiski Township, Armstrong County, PA)

1994-01-01T23:59:59.000Z

209

Pressurizer tank upper support  

DOE Patents [OSTI]

A pressurizer tank in a pressurized water nuclear reactor is mounted between structural walls of the reactor on a substructure of the reactor, the tank extending upwardly from the substructure. For bearing lateral loads such as seismic shocks, a girder substantially encircles the pressurizer tank at a space above the substructure and is coupled to the structural walls via opposed sway struts. Each sway strut is attached at one end to the girder and at an opposite end to one of the structural walls, and the sway struts are oriented substantially horizontally in pairs aligned substantially along tangents to the wall of the circular tank. Preferably, eight sway struts attach to the girder at 90[degree] intervals. A compartment encloses the pressurizer tank and forms the structural wall. The sway struts attach to corners of the compartment for maximum stiffness and load bearing capacity. A valve support frame carrying the relief/discharge piping and valves of an automatic depressurization arrangement is fixed to the girder, whereby lateral loads on the relief/discharge piping are coupled directly to the compartment rather than through any portion of the pressurizer tank. Thermal insulation for the valve support frame prevents thermal loading of the piping and valves. The girder is shimmed to define a gap for reducing thermal transfer, and the girder is free to move vertically relative to the compartment walls, for accommodating dimensional variation of the pressurizer tank with changes in temperature and pressure. 10 figures.

Baker, T.H.; Ott, H.L.

1994-01-11T23:59:59.000Z

210

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

211

Cryogenic Pressure Vessel workshop, LLNL, February 15, 2011, p. 1 Cryogenic Pressure Vessels  

E-Print Network [OSTI]

, February 15, 2011, p. 8 In both industrial and laboratory environments, low heat transfer requires remain colder than 150 K due to expansion work during hydrogen extraction Source: BMW #12;Cryogenic

212

Acoustic emission monitoring of HFIR vessel during hydrostatic testing  

SciTech Connect (OSTI)

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

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

1992-08-01T23:59:59.000Z

213

DESIGN OF THE ITER IN-VESSEL COILS  

SciTech Connect (OSTI)

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

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

2010-11-27T23:59:59.000Z

214

Neutrino Factory Mercury Vessel  

E-Print Network [OSTI]

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

McDonald, Kirk

215

HEART AND BLOOD VESSELS CARDIOVASCULARCARDIOVASCULAR  

E-Print Network [OSTI]

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

Cochran-Stafira, D. Liane

216

Sandia National Laboratories: prevent damage to toroid vessel wall  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1developmentturbine bladelifetimepower-to-gas applicationsinternal componentsprevent

217

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

SciTech Connect (OSTI)

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

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

1988-01-01T23:59:59.000Z

218

Single module pressurized fuel cell turbine generator system  

DOE Patents [OSTI]

A pressurized fuel cell system (10), operates within a common pressure vessel (12) where the system contains fuel cells (22), a turbine (26) and a generator (98) where preferably, associated oxidant inlet valve (52), fuel inlet valve (56) and fuel cell exhaust valve (42) are outside the pressure vessel.

George, Raymond A. (Pittsburgh, PA); Veyo, Stephen E. (Murrysville, PA); Dederer, Jeffrey T. (Valencia, PA)

2001-01-01T23:59:59.000Z

219

High pressure furnace  

DOE Patents [OSTI]

A high temperature high pressure furnace has a hybrid partially externally heated construction. A metallic vessel fabricated from an alloy having a composition of at least 45% nickel, 15% chrome, and 10% tungsten is utilized (the preferred alloy including 55% nickel, 22% chrome, 14% tungsten, 2% molybdenum, 3% iron (maximum) and 5% cobalt (maximum)). The disclosed alloy is fabricated into 11/4 or 2 inch, 32 mm or 50 mm bar stock and has a length of about 22 inches, 56 cm. This bar stock has an aperture formed therein to define a closed high temperature, high pressure oxygen chamber. The opposite and closed end of the vessel is provided with a small blind aperture into which a thermocouple can be inserted. The closed end of the vessel is inserted into an oven, preferably heated by standard nickel chrome electrical elements and having a heavily insulated exterior. 19 figures.

Morris, D.E.

1993-09-14T23:59:59.000Z

220

High pressure furnace  

DOE Patents [OSTI]

A high temperature high pressure furnace has a hybrid partially externally heated construction. A metallic vessel fabricated from an alloy having a composition of at least 45% nickel, 15% chrome, and 10% tungsten is utilized (the preferred alloy including 55% nickel, 22% chrome, 14% tungsten, 2% molybdenum, 3% iron (maximum) and 5% cobalt (maximum). The disclosed alloy is fabricated into 11/4 or 2 inch, 32 mm or 50 mm bar stock and has a length of about 22 inches, 56 cm. This bar stock has an aperture formed therein to define a closed high temperature, high pressure oxygen chamber. The opposite and closed end of the vessel is provided with a small blind aperture into which a thermocouple can be inserted. The closed end of the vessel is inserted into an oven, preferably heated by standard nickel chrome electrical elements and having a heavily insulated exterior.

Morris, Donald E. (Kensington, CA)

1993-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "walled pressure vessel" 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

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

SciTech Connect (OSTI)

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

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

1995-11-01T23:59:59.000Z

222

Method of non-destructively inspecting a curved wall portion  

DOE Patents [OSTI]

A method of non-destructively inspecting a curved wall portion of a large and thick walled vessel for a defect by computed tomography is provided. A collimated source of radiation is placed adjacent one side of the wall portion and an array of detectors for the radiation is placed on the other side adjacent the source. The radiation from the source passing through the wall portion is then detected with the detectors over a limited angle, dependent upon the curvature of the wall of the vessel, to obtain a dataset. The source and array are then coordinately moved relative to the wall portion in steps and a further dataset is obtained at each step. The plurality of datasets obtained over the limited angle is then processed to produce a tomogram of the wall portion to determine the presence of a defect therein. In a preferred embodiment, the curved wall portion has a center of curvature so that the source and the array are positioned at each step along a respective arc curved about the center. If desired, the detector array and source can be reoriented relative to a new wall portion and an inspection of the new wall portion can be easily obtained. Further, the source and detector array can be indexed in a direction perpendicular to a plane including the limited angle in a plurality of steps so that by repeating the detecting and moving steps at each index step, a three dimensional image can be created of the wall portion.

Fong, James T. (Bethel Park, PA)

1996-01-01T23:59:59.000Z

223

Near-wall serpentine cooled turbine airfoil  

DOE Patents [OSTI]

A serpentine coolant flow path is formed by inner walls in a cavity between pressure and suction side walls of a turbine airfoil, the cavity partitioned by one or more transverse partitions into a plurality of continuous serpentine cooling flow streams each having a respective coolant inlet.

Lee, Ching-Pang

2014-10-28T23:59:59.000Z

224

Bremsstrahlung Radiation At a Vacuum Bubble Wall  

E-Print Network [OSTI]

When charged particles collide with a vacuum bubble, they can radiate strong electromagnetic waves due to rapid deceleration. Owing to the energy loss of the particles by this bremsstrahlung radiation, there is a non-negligible damping pressure acting on the bubble wall even when thermal equilibrium is maintained. In the non-relativistic region, this pressure is proportional to the velocity of the wall and could have influenced the bubble dynamics in the early universe.

Jae-Weon Lee; Kyungsub Kim; Chul H. Lee; Ji-ho Jang

2007-04-06T23:59:59.000Z

225

CFD Validation of Gas Injection in Flowing Mercury over Vertical Smooth and Grooved Wall  

SciTech Connect (OSTI)

The Spallation Neutron Source (SNS) is an accelerator-based neutron source at Oak Ridge National Laboratory (ORNL).The nuclear spallation reaction occurs when a proton beam hits liquid mercury. This interaction causes thermal expansion of the liquid mercury which produces high pressure waves. When these pressure waves hit the target vessel wall, cavitation can occur and erode the wall. Research and development efforts at SNS include creation of a vertical protective gas layer between the flowing liquid mercury and target vessel wall to mitigate the cavitation damage erosion and extend the life time of the target. Since mercury is opaque, computational fluid dynamics (CFD) can be used as a diagnostic tool to see inside the liquid mercury and guide the experimental efforts. In this study, CFD simulations of three dimensional, unsteady, turbulent, two-phase flow of helium gas injection in flowing liquid mercury over smooth, vertically grooved and horizontally grooved walls are carried out with the commercially available CFD code Fluent-12 from ANSYS. The Volume of Fluid (VOF) model is used to track the helium-mercury interface. V-shaped vertical and horizontal grooves with 0.5 mm pitch and about 0.7 mm depth were machined in the transparent wall of acrylic test sections. Flow visualization data of helium gas coverage through transparent test sections is obtained with a high-speed camera at the ORNL target test facility (TTF). The helium gas mass flow rate is 8 mg/min and introduced through a 0.5 mm diameter port. The local mercury velocity is 0.9 m/s. In this paper, the helium gas flow rate and the local mercury velocity are kept constant for the three cases. Time integration of predicted helium gas volume fraction over time is done to evaluate the gas coverage and calculate the average thickness of the helium gas layer. The predicted time-integrated gas coverage over vertically grooved and horizontally grooved test sections is better than over a smooth wall. The simulations show that the helium gas is trapped inside the grooves. The predicted time-averaged gas coverage is in good qualitative agreement with the measured gas coverage.

Abdou, Ashraf A [ORNL; Wendel, Mark W [ORNL; Felde, David K [ORNL; Riemer, Bernie [ORNL

2009-01-01T23:59:59.000Z

226

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

SciTech Connect (OSTI)

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

Vinson, Dennis

2010-06-01T23:59:59.000Z

227

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

228

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

229

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

E-Print Network [OSTI]

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

230

Retrospective dosimetry analyses of reactor vessel cladding samples  

SciTech Connect (OSTI)

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

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

2011-07-01T23:59:59.000Z

231

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

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

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

232

BWR ex-vessel steam explosion analysis with MC3D code  

SciTech Connect (OSTI)

A steam explosion may occur, during a severe reactor accident, when the molten core comes into contact with the coolant water. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. To resolve the open issues in steam explosion understanding and modeling, the OECD program SERENA phase 2 was launched at the end of year 2007, focusing on reactor applications. To verify the progress made in the understanding and modeling of fuel coolant interaction key phenomena for reactor applications a reactor exercise has been performed. In this paper the BWR ex-vessel steam explosion study, which was carried out with the MC3D code in conditions of the SERENA reactor exercise for the BWR case, is presented and discussed. The premixing simulations were performed with two different jet breakup modeling approaches and the explosion was triggered also at the expected most challenging time. For the most challenging case, at the cavity wall the highest calculated pressure was {approx}20 MPa and the highest pressure impulse was {approx}90 kPa.s. (authors)

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

2012-07-01T23:59:59.000Z

233

Wall to Wall Optimal Transport  

E-Print Network [OSTI]

The calculus of variations is employed to find steady divergence-free velocity fields that maximize transport of a tracer between two parallel walls held at fixed concentration for one of two constraints on flow strength: a fixed value of the kinetic energy or a fixed value of the enstrophy. The optimizing flows consist of an array of (convection) cells of a particular aspect ratio Gamma. We solve the nonlinear Euler-Lagrange equations analytically for weak flows and numerically (and via matched asymptotic analysis in the fixed energy case) for strong flows. We report the results in terms of the Nusselt number Nu, a dimensionless measure of the tracer transport, as a function of the Peclet number Pe, a dimensionless measure of the energy or enstrophy of the flow. For both constraints the maximum transport Nu_{MAX}(Pe) is realized in cells of decreasing aspect ratio Gamma_{opt}(Pe) as Pe increases. For the fixed energy problem, Nu_{MAX} \\sim Pe and Gamma_{opt} \\sim Pe^{-1/2}, while for the fixed enstrophy scenario, Nu_{MAX} \\sim Pe^{10/17} and Gamma_{opt} \\sim Pe^{-0.36}. We also interpret our results in the context of certain buoyancy-driven Rayleigh-Benard convection problems that satisfy one of the two intensity constraints, enabling us to investigate how the transport scalings compare with upper bounds on Nu expressed as a function of the Rayleigh number \\Ra. For steady convection in porous media, corresponding to the fixed energy problem, we find Nu_{MAX} \\sim \\Ra and Gamma_{opt} \\sim Ra^{-1/2}$, while for steady convection in a pure fluid layer between free-slip isothermal walls, corresponding to fixed enstrophy transport, Nu_{MAX} \\sim Ra^{5/12} and Gamma_{opt} \\sim Ra^{-1/4}.

Pedram Hassanzadeh; Gregory P. Chini; Charles R. Doering

2014-04-14T23:59:59.000Z

234

CRAD, Pressurized Systems and Cryogens Assessment Plan  

Broader source: Energy.gov [DOE]

Assure personnel health and safety through regularly scheduled inspections and maintenance on pressure vessels and equipment, compressed gases and gas cylinders, vacuum equipment and systems, hydraulics, and cryogenic materials and systems.

235

The construction of the Browns Bay Vessel  

E-Print Network [OSTI]

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

Amer, Christopher Francis

2012-06-07T23:59:59.000Z

236

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

237

Status of R&D on Mitigating the Effects of Pressure Waves for the Spallation Neutron Source Mercury Target  

SciTech Connect (OSTI)

The Spallation Neutron Source (SNS) at the Oak Ridge National Laboratory has been conducting R&D on mitigating the effects of pressure waves in mercury spallation targets since 2001. More precisely, cavitation damage of the target vessel caused by the short beam pulse threatens to limit its lifetime more severely than radiation damage as well as limit its ultimate power capacity and hence its neutron intensity performance. The R&D program has moved from verification of the beam-induced damage phenomena to study of material and surface treatments for damage resistance to the current emphasis on gas injection techniques for damage mitigation. Two techniques are being worked on: injection of small dispersed gas bubbles that mitigate the pressure waves volumetrically; and protective gas walls that isolate the vessel from the damaging effects of collapsing cavitation bubbles. The latter has demonstrated good damage mitigation during in-beam testing with limited pulses, and adequate gas wall coverage at the beam entrance window has been demonstrated with the SNS mercury target flow configuration using a full scale mercury test loop. A question on the required area coverage remains which depends on results from SNS target post irradiation examination. The small gas bubble technique has been less effective during past in-beam tests but those results were with un-optimized and un-verified bubble populations. Another round of in-beam tests with small gas bubbles is planned for 2011. The first SNS target was removed from service in mid 2009 and samples were cut from two locations at the target s beam entrance window. Through-wall damage was observed at the innermost mercury vessel wall (not a containment wall). The damage pattern suggested correlation with the local mercury flow condition which is nearly stagnant at the peak damage location. Detailed post irradiation examination of the samples is under way that will assess the erosion and measure irradiation-induced changes in mechanical properties. Similar samples were cut from the second SNS target after it was removed from service in mid 2010. More extensive damage was observed on the target inner wall but damage to the containment wall was minimal.

Riemer, Bernie [ORNL] [ORNL; Wendel, Mark W [ORNL] [ORNL; Felde, David K [ORNL] [ORNL; Abdou, Ashraf A [ORNL] [ORNL; McClintock, David A [ORNL] [ORNL

2012-01-01T23:59:59.000Z

238

CHF Enhancement by Vessel Coating for External Reactor Vessel Cooling  

SciTech Connect (OSTI)

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

Fan-Bill Cheung; Joy L. Rempe

2004-06-01T23:59:59.000Z

239

E-Print Network 3.0 - axi-symmetric exit pressures Sample Search...  

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

driven by the imposed high pressure at the top and the lower exit pressure. The inner tube provides... 1 Paper submitted to ASME 2008 ASME Pressure Vessels and Piping ... Source:...

240

Predicting Stenosis in Blood Vessels  

E-Print Network [OSTI]

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

Petta, Jason

Note: This page contains sample records for the topic "walled pressure vessel" 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

Achieve Continuous Injection of Solid Fuels into Advanced Combustion System Pressures  

SciTech Connect (OSTI)

The overall objective of this project is the development of a mechanical rotary-disk feeder, known as the Stamet Posimetric High Pressure Solids Feeder System, to demonstrate feeding of dry granular coal continuously and controllably into pressurized environments of up to 70 kg/cm2 (1,000 psi). This is the Phase III of the ongoing program. Earlier Phases 1 and II successfully demonstrated feeding into pressures up to 35 kg/cm{sup 2} (500 psi). The final report for those phases was submitted in April 2005. Based on the previous work done in Phases I & II using Powder River Basin coal provided by the PSDF facility in Wilsonville, AL, a Phase III feeder system was designed and built to accomplish the target of feeding the coal into a pressure of 70 kg/cm2 (1,000 psi) and to be capable of feed rates of up to 550 kilograms (1,200lbs) per hour. The drive motor system from Phase II was retained for use on Phase III since projected performance calculations indicated it should be capable of driving the Phase III pump to the target levels. The pump & motor system was installed in a custom built test rig comprising an inlet vessel containing an active live-wall hopper mounted on weigh cells in a support frame, transition into the pump inlet, transition from pump outlet and a receiver vessel containing a receiver drum supported on weigh cells. All pressure containment on the rig was rated to105 kg/cm{sup 2} (1,500psi) to accommodate the final pressure requirement of a proposed Phase IV of the program. A screw conveyor and batch hopper were added to transfer coal at atmospheric pressure from the shop floor up into the test rig to enable continuous feeding up to the capacity of the receiving vessel. Control & monitoring systems were up-rated from the Phase II system to cover the additional features incorporated in the Phase III rig, and provide closer control and expanded monitoring of the entire system. A program of testing and modification was carried out in Stamet's facility in CA, culminating in the first successful feeding of coal into the Phase III target of 70 kg/cm{sup 2} (1,000 psi) gas pressure in March 2007. Subsequently, repeated runs at pressure were achieved, and comparison of the data with Phase II results when adjusted for scale differences showed further power reductions of 40% had been achieved from the final Phase II pressure runs. The general design layout of a commercial-scale unit was conducted, and preliminary cost estimates made.

Derek L. Aldred; Timothy Saunders

2007-03-31T23:59:59.000Z

242

Investigation of vessel exterior air cooling for a HLMC reactor  

SciTech Connect (OSTI)

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

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

2000-01-13T23:59:59.000Z

243

Investigation of vessel exterior air cooling for an HLMC reactor  

SciTech Connect (OSTI)

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

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

2000-07-01T23:59:59.000Z

244

Effect of elasticity of wall on diffusion in nano channel  

SciTech Connect (OSTI)

Confining walls of nano channel are taken to be elastic to study their effect on the diffusion coefficient of fluid flowing through the channel. The wall is elastic to the extent that it responses to molecular pressure exerted by fluid. The model to study diffusion is based on microscopic considerations. Results obtained for fluid confining to 20 atomic diameter width contrasted with results obtained by considering rigid and smooth wall. The effect of roughness of wall on diffusion can be compensated by the elastic property of wall.

Tankeshwar, K., E-mail: tankesh@pu.ac.in [Computer Centre, Panjab University Chandigarh,- 160014 (India); Srivastava, Sunita [Department of Physics, Panjab University, Chandigarh 160014 (India)

2014-04-24T23:59:59.000Z

245

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

SciTech Connect (OSTI)

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

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

1992-08-01T23:59:59.000Z

246

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

SciTech Connect (OSTI)

Radioactive sludge was generated in the K-East Basin and K-West Basin fuel storage pools at the Hanford Site while irradiated uranium metal fuel elements from the N Reactor were being stored and packaged. The fuel has been removed from the K Basins, and currently, the sludge resides in the KW Basin in large underwater Engineered Containers. The first phase to the Sludge Treatment Project being led by CH2MHILL Plateau Remediation Company (CHPRC) is to retrieve and load the sludge into sludge transport and storage containers (STSCs) and transport the sludge to T Plant for interim storage. The STSCs will be stored inside T Plant cells that are equipped with secondary containment and leak-detection systems. The sludge is composed of a variety of particulate materials and water, including a fraction of reactive uranium metal particles that are a source of hydrogen gas. If a situation occurs where the reactive uranium metal particles settle out at the bottom of a container, previous studies have shown that a vessel-spanning gas layer above the uranium metal particles can develop and can push the overlying layer of sludge upward. The major concern, in addition to the general concern associated with the retention and release of a flammable gas such as hydrogen, is that if a vessel-spanning bubble (VSB) forms in an STSC, it may drive the overlying sludge material to the vents at the top of the container. Then it may be released from the container into the cell’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

247

aluminum pressure vessels: Topics by E-print Network  

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

angular region on the surface Stokes, Yvonne 204 iCons, 2011 Alzheimers and Aluminum: Lesson Plan Chemistry Websites Summary: iCons, 2011 Alzheimers and Aluminum: Lesson Plan...

248

International Hydrogen Fuel and Pressure Vessel Forum 2010 Proceedings...  

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

ihfpvproceedings.pdf More Documents & Publications Workshop Notes from ""Compressed Natural Gas and Hydrogen Fuels: Lessons Learned for the Safe Deployment of Vehicles""...

249

Digital material skins : for reversible reusable pressure vessels  

E-Print Network [OSTI]

Spacecraft missions have traditionally sacrificed fully functional hardware and entire vehicles to achieve mission objectives. Propellant tanks are typically jettisoned at different stages in a spacecraft mission and left ...

Hovsepian, Sarah

2012-01-01T23:59:59.000Z

250

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel...  

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

of one or more NDE techniques that can assist in the determination of current RPV fracture toughness as well as in prediction of fracture toughness with further aging of the...

251

Forum Agenda: International Hydrogen Fuel and Pressure Vessel Forum |  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:YearRound-UpHeat PumpRecordFederal7.pdfFlash_2010_-24.pdfOverviewPlans |Updated AugustActDepartment

252

Lightweight cryogenic-compatible pressure vessels for vehicular fuel  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: VegetationEquipment Surfaces and Interfaces Sample6, 2011 LawrenceEfeedstocksHomesLighting

253

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergy 0611__Joint_DOE_GoJ_AMS_Data_v3.pptx More DocumentsCommunicationsProvides an overview of

254

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergy 0611__Joint_DOE_GoJ_AMS_Data_v3.pptx More DocumentsCommunicationsProvides an overview

255

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergy 0611__Joint_DOE_GoJ_AMS_Data_v3.pptx More DocumentsCommunicationsProvides an overviewMilestone

256

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research  

Energy Savers [EERE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of Inspector GeneralDepartment of Energy fromCommentsRevolving Loan Funds Revolving LoanA l i c e Land

257

International Hydrogen Fuel and Pressure Vessel Forum 2010 Proceedings  

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

experts presented information and data on testing and certification of storage tanks for compressed hydrogen, CNG, and HCNG fuels. 1 Specific objectives of the Forum were...

258

International Hydrogen Fuel and Pressure Vessel Forum | Department of  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:Year in3.pdfEnergy Health andofIan KalinResearch, Development,CoP)Builders'

259

Cryogenic Pressure Vessels: Progress and Plans | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:YearRound-Up fromDepartmentTieCelebratePartners with Siemens onSite |DepartmentHydrogen

260

High-pressure Storage Vessels for Hydrogen, Natural Gas and  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:YearRound-UpHeatMulti-Dimensional Subject:Ground HawaiiWaste Heat Recovery:| Department of|a d e

Note: This page contains sample records for the topic "walled pressure vessel" 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

International Hydrogen Fuel and Pressure Vessel Forum - Presentations |  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:YearRound-UpHeatMulti-Dimensionalthe10 DOE Vehicle TechnologiesDepartment of Energy

262

International Hydrogen Fuel and Pressure Vessel Forum 2010 Proceedings |  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:YearRound-UpHeatMulti-Dimensionalthe10 DOE Vehicle TechnologiesDepartment of EnergyDepartment

263

E-Print Network 3.0 - atmospheric pressure direct Sample Search...  

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

Collection: Engineering 87 The Centers for Disease Control and Prevention's Vessel Sanitation Program is proud to bring to you the following Summary: pressure 12;Types of...

264

Achieve Continuous Injection of Solid Fuels into Advanced Combustion System Pressures  

SciTech Connect (OSTI)

The overall objective of this project is the development of a mechanical rotary-disk feeder, known as the Stamet Posimetric High Pressure Solids Feeder System, to feed dry granular coal continuously and controllably into pressurized environments of up to 35 kg/cm{sup 2} (500 psi). This was to be accomplished in two phases. The first task was to review materials handling experience in pressurized operations as it related to the target pressures for this project, and review existing coal preparation processes and specifications currently used in advanced combustion systems. Samples of existing fuel materials were obtained and tested to evaluate flow, sealing and friction properties. This provided input data for use in the design of the Stamet Feeders for the project, and ensured that the material specification used met the requirements of advanced combustion & gasification systems. Ultimately, Powder River Basin coal provided by the PSDF facility in Wilsonville, AL was used as the basis for the feeder design and test program. Based on the material property information, a Phase 1 feeder system was designed and built to accomplish feeding the coal to an intermediate pressure up to 21 kg/cm{sup 2} (300 psi) at feed rates of approximately 100 kilograms (220lbs) per hour. The pump & motor system was installed in a custom built test rig comprising an inlet vessel containing an active live-wall hopper mounted in a support frame, transition into the pump inlet, transition from pump outlet and a receiver vessel containing a receiver drum supported on weigh cells. All pressure containment on the rig was rated for the final pressure requirement of 35 kg/cm{sup 2} (500psi). A program of testing and modification was carried out in Stamet's facility in CA, culminating in successful feeding of coal into the Phase 1 target of 21 kg/cm{sup 2} (300psi) gas pressure in December 2003. Further testing was carried out at CQ Inc's facility in PA, providing longer run times and experience of handling and feeding the coal in winter conditions. Based on the data developed through the testing of the Phase I unit, a Phase II system was designed for feeding coal into pressures of up to 35 kg/cm{sup 2} (500 psi). A further program of testing and modification was then carried out in Stamet's facility, with the target pressure being achieved in January 2005. Repeated runs at pressure were achieved, and optimization of the machine resulted in power reductions of 60% from the first successful pressure runs. General design layout of a commercial-scale unit was conducted, and preliminary cost estimates for a commercial unit obtained.

Derek L. Aldred; Timothy Saunders

2005-07-01T23:59:59.000Z

265

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

266

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

267

Near wall cooling for a highly tapered turbine blade  

SciTech Connect (OSTI)

A turbine blade having a pressure sidewall and a suction sidewall connected at chordally spaced leading and trailing edges to define a cooling cavity. Pressure and suction side inner walls extend radially within the cooling cavity and define pressure and suction side near wall chambers. A plurality of mid-chord channels extend radially from a radially intermediate location on the blade to a tip passage at the blade tip for connecting the pressure side and suction side near wall chambers in fluid communication with the tip passage. In addition, radially extending leading edge and trailing edge flow channels are located adjacent to the leading and trailing edges, respectively, and cooling fluid flows in a triple-pass serpentine path as it flows through the leading edge flow channel, the near wall chambers and the trailing edge flow channel.

Liang, George (Palm City, FL)

2011-03-08T23:59:59.000Z

268

First Wall and Operational Diagnostics  

SciTech Connect (OSTI)

In this chapter we review numerous diagnostics capable of measurements at or near the first wall, many of which contribute information useful for safe operation of a tokamak. There are sections discussing infrared cameras, visible and VUV cameras, pressure gauges and RGAs, Langmuir probes, thermocouples, and erosion and deposition measurements by insertable probes and quartz microbalance. Also discussed are dust measurements by electrostatic detectors, laser scattering, visible and IR cameras, and manual collection of samples after machine opening. In each case the diagnostic is discussed with a view toward application to a burning plasma machine such as ITER.

Lasnier, C; Allen, S; Boedo, J; Groth, M; Brooks, N; McLean, A; LaBombard, B; Sharpe, J; Skinner, C; Whyte, D; Rudakov, D; West, W; Wong, C

2006-06-19T23:59:59.000Z

269

Pressurized reactor system and a method of operating the same  

DOE Patents [OSTI]

A method and apparatus are provided for operating a pressurized reactor system in order to precisely control the temperature within a pressure vessel in order to minimize condensation of corrosive materials from gases on the surfaces of the pressure vessel or contained circulating fluidized bed reactor, and to prevent the temperature of the components from reaching a detrimentally high level, while at the same time allowing quick heating of the pressure vessel interior volume during start-up. Superatmospheric pressure gas is introduced from the first conduit into the fluidized bed reactor and heat derived reactions such as combustion and gassification are maintained in the reactor. Gas is exhausted from the reactor and pressure vessel through a second conduit. Gas is circulated from one part of the inside volume to another to control the temperature of the inside volume, such as by passing the gas through an exterior conduit which has a heat exchanger, control valve, blower and compressor associated therewith, or by causing natural convection flow of circulating gas within one or more generally vertically extending gas passages entirely within the pressure vessel (and containing heat exchangers, flow rate control valves, or the like therein). Preferably, inert gas is provided as a circulating gas, and the inert gas may also be used in emergency shut-down situations. In emergency shut-down reaction gas being supplied to the reactor is cut off, while inert gas from the interior gas volume of the pressure vessel is introduced into the reactor.

Isaksson, Juhani M. (Karhula, FI)

1996-01-01T23:59:59.000Z

270

Pressurized reactor system and a method of operating the same  

DOE Patents [OSTI]

A method and apparatus are provided for operating a pressurized reactor system in order to precisely control the temperature within a pressure vessel in order to minimize condensation of corrosive materials from gases on the surfaces of the pressure vessel or contained circulating fluidized bed reactor, and to prevent the temperature of the components from reaching a detrimentally high level, while at the same time allowing quick heating of the pressure vessel interior volume during start-up. Super-atmospheric pressure gas is introduced from the first conduit into the fluidized bed reactor and heat derived reactions such as combustion and gasification are maintained in the reactor. Gas is exhausted from the reactor and pressure vessel through a second conduit. Gas is circulated from one part of the inside volume to another to control the temperature of the inside volume, such as by passing the gas through an exterior conduit which has a heat exchanger, control valve, blower and compressor associated therewith, or by causing natural convection flow of circulating gas within one or more generally vertically extending gas passages entirely within the pressure vessel (and containing heat exchangers, flow rate control valves, or the like therein). Preferably, inert gas is provided as a circulating gas, and the inert gas may also be used in emergency shut-down situations. In emergency shut-down reaction gas being supplied to the reactor is cut off, while inert gas from the interior gas volume of the pressure vessel is introduced into the reactor. 2 figs.

Isaksson, J.M.

1996-06-18T23:59:59.000Z

271

Light Sources on the Nylon Vessels' Surfaces  

E-Print Network [OSTI]

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

272

Liquid Wall Chambers  

SciTech Connect (OSTI)

The key feature of liquid wall chambers is the use of a renewable liquid layer to protect chamber structures from target emissions. Two primary options have been proposed and studied: wetted wall chambers and thick liquid wall (TLW) chambers. With wetted wall designs, a thin layer of liquid shields the structural first wall from short ranged target emissions (x-rays, ions and debris) but not neutrons. Various schemes have been proposed to establish and renew the liquid layer between shots including flow-guiding porous fabrics (e.g., Osiris, HIBALL), porous rigid structures (Prometheus) and thin film flows (KOYO). The thin liquid layer can be the tritium breeding material (e.g., flibe, PbLi, or Li) or another liquid metal such as Pb. TLWs use liquid jets injected by stationary or oscillating nozzles to form a neutronically thick layer (typically with an effective thickness of {approx}50 cm) of liquid between the target and first structural wall. In addition to absorbing short ranged emissions, the thick liquid layer degrades the neutron flux and energy reaching the first wall, typically by {approx}10 x x, so that steel walls can survive for the life of the plant ({approx}30-60 yrs). The thick liquid serves as the primary coolant and tritium breeding material (most recent designs use flibe, but the earliest concepts used Li). In essence, the TLW places the fusion blanket inside the first wall instead of behind the first wall.

Meier, W R

2011-02-24T23:59:59.000Z

273

Method for pressure modulation of turbine sidewall cavities  

DOE Patents [OSTI]

A method is provided for controlling cooling air flow for pressure modulation of turbine components, such as the turbine outer sidewall cavities. The pressure at which cooling and purge air is supplied to the turbine outer side wall cavities is modulated, based on compressor discharge pressure (Pcd), thereby to generally maintain the back flow margin (BFM) so as to minimize excessive leakage and the consequent performance deterioration. In an exemplary embodiment, the air pressure within the third stage outer side wall cavity and the air pressure within the fourth stage outer side wall cavity are each controlled to a respective value that is a respective prescribed percentage of the concurrent compressor discharge pressure. The prescribed percentage may be determined from a ratio of the respective outer side wall pressure to compressor discharge pressure at Cold Day Turn Down (CDTD) required to provide a prescribed back flow margin.

Leone, Sal Albert (Scotia, NY); Book, Matthew David (Altamont, NY); Banares, Christopher R. (Schenectady, NY)

2002-01-01T23:59:59.000Z

274

System for pressure modulation of turbine sidewall cavities  

DOE Patents [OSTI]

A system and method are provided for controlling cooling air flow for pressure modulation of turbine components, such as the turbine outer sidewall cavities. The pressure at which cooling and purge air is supplied to the turbine outer side wall cavities is modulated, based on compressor discharge pressure (Pcd), thereby to generally maintain the back flow margin (BFM) so as to minimize excessive leakage and the consequent performance deterioration. In an exemplary embodiment, the air pressure within the third stage outer side wall cavity and the air pressure within the fourth stage outer side wall cavity are each controlled to a respective value that is a respective prescribed percentage of the concurrent compressor discharge pressure. The prescribed percentage may be determined from a ratio of the respective outer side wall pressure to compressor discharge pressure at Cold Day Turn Down (CDTD) required to provide a prescribed back flow margin.

Leone, Sal Albert (Scotia, NY); Book, Matthew David (Altamont, NY); Banares, Christopher R. (Schenectady, NY)

2002-01-01T23:59:59.000Z

275

Wood Pulp Digetster Wall Corrosion Investigation  

SciTech Connect (OSTI)

The modeling of the flow in a wood pulp digester is but one component of the investigation of the corrosion of digesters. This report describes the development of a Near-Wall-Model (NWM) that is intended to couple with a CFD model that determines the flow, heat, and chemical species transport and reaction within the bulk flow of a digester. Lubrication theory approximations were chosen from which to develop a model that could determine the flow conditions within a thin layer near the vessel wall using information from the interior conditions provided by a CFD calculation of the complete digester. The other conditions will be determined by coupled solutions of the wood chip, heat, and chemical species transport and chemical reactions. The NWM was to couple with a digester performance code in an iterative fashion to provide more detailed information about the conditions within the NW region. Process Simulations, Ltd (PSL) is developing the digester performance code. This more detailed (and perhaps more accurate) information from the NWM was to provide an estimate of the conditions that could aggravate the corrosion at the wall. It is intended that this combined tool (NWM-PSL) could be used to understand conditions at/near the wall in order to develop methods to reduce the corrosion. However, development and testing of the NWM flow model took longer than anticipated and the other developments (energy and species transport, chemical reactions and linking with the PSL code) were not completed. The development and testing of the NWM are described in this report. In addition, the investigation of the potential effects of a clear layer (layer reduced in concentration of wood chips) near the wall is reported in Appendix D. The existence of a clear layer was found to enhance the flow near the wall.

Giles, GE

2003-09-18T23:59:59.000Z

276

Master external pressure charts  

SciTech Connect (OSTI)

This paper presents a method to develop master external pressure charts from which individual external pressure charts for each material specification may be derived. The master external charts can represent a grouping of materials with similar chemical composition, similar stress-strain curves but produced to different strength levels. External pressure charts are used by various Sections of the ASME Boiler and Pressure Vessel and Piping Codes to design various components such as cylinders, sphered, formed heads, tubes, piping, rings and other components, subjected to external pressure or axial compression loads. These charts are pseudo stress-strain curves for groups of materials with similar stress-strain shapes. The traditional approach was originally developed in the 1940`s and is a graphical approach where slopes to the strain curves are drawn graphically from which pseudo-strain levels are calculated. The new method presented in this paper develops mathematical relationships for the material stress-strain curves and the external pressure charts. The method has the ability to calculate stress-strain curves from existing external pressure charts. The relationships are a function of temperature, the modulus of elasticity, yield strength, and two empirical material constants. In this approach, conservative assumptions used to assign materials to lower bound external pressure charts can be removed. This increases the buckling strength capability of many materials in the Code, providing economic benefits while maintaining the margin of safety specified by the Code criteria. The method can also reduce the number of material charts needed in the Code and provides for the capability to extend the existing pressure charts to higher design temperatures. The new method is shown to contain a number of improvements over the traditional approach and is presently under consideration by appropriate ASME Code committees.

Michalopoulos, E. [Hartford Steam Boiler Inspection and Insurance Co., CT (United States). Codes and Standards Dept.

1996-12-01T23:59:59.000Z

277

Webs of Walls  

E-Print Network [OSTI]

Webs of domain walls are constructed as 1/4 BPS states in d=4, N=2 supersymmetric U(Nc) gauge theories with Nf hypermultiplets in the fundamental representation. Web of walls can contain any numbers of external legs and loops like (p,q) string/5-brane webs. We find the moduli space M of a 1/4 BPS equation for wall webs to be the complex Grassmann manifold. When moduli spaces of 1/2 BPS states (parallel walls) and the vacua are removed from M, the non-compact moduli space of genuine 1/4 BPS wall webs is obtained. All the solutions are obtained explicitly and exactly in the strong gauge coupling limit. In the case of Abelian gauge theory, we work out the correspondence between configurations of wall web and the moduli space CP^{Nf-1}.

Minoru Eto; Youichi Isozumi; Muneto Nitta; Keisuke Ohashi; Norisuke Sakai

2005-06-20T23:59:59.000Z

278

Pressure relief valve/safety relief valve testing  

SciTech Connect (OSTI)

Pressure vessels and piping systems are protected form overpressurization by pressure relief valves. These safety features are required to be tested-inspected on some periodic basis and, in most cases witnessed by a third party inspector. As a result nonconformances found by third parties Westinghouse Hanford Company initiated a task team to develop a pressure safety program. This paper reveals their findings.

Murray, W.A.; Hamm, E.R.; Barber, J.R.

1994-02-01T23:59:59.000Z

279

Superheat effects on localized vessel breach enlargement during corium ejection  

SciTech Connect (OSTI)

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

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

1986-01-01T23:59:59.000Z

280

Four-wall turbine airfoil with thermal strain control for reduced cycle fatigue  

DOE Patents [OSTI]

A turbine airfoil (20B) with a thermal expansion control mechanism that increases the airfoil camber (60, 61) under operational heating. The airfoil has four-wall geometry, including pressure side outer and inner walls (26, 28B), and suction side outer and inner walls (32, 34B). It has near-wall cooling channels (31F, 31A, 33F, 33A) between the outer and inner walls. A cooling fluid flow pattern (50C, 50W, 50H) in the airfoil causes the pressure side inner wall (28B) to increase in curvature under operational heating. The pressure side inner wall (28B) is thicker than walls (26, 34B) that oppose it in camber deformation, so it dominates them in collaboration with the suction side outer wall (32), and the airfoil camber increases. This reduces and relocates a maximum stress area (47) from the suction side outer wall (32) to the suction side inner wall (34B, 72) and the pressure side outer wall (26).

Cambell, Christian X

2013-09-17T23:59:59.000Z

Note: This page contains sample records for the topic "walled pressure vessel" 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

Plastic instabilities in statically and dynamically loaded spherical vessels  

SciTech Connect (OSTI)

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

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

2010-01-01T23:59:59.000Z

282

Mineral matter transformations in a pressurized drop-tube furnace  

SciTech Connect (OSTI)

To meet the objectives of the program, a pressurized combustion vessel was built to allow the operating parameters of a direct-fired gas turbine combustor to be simulated. One goal in building this equipment was to design the gas turbine simulator as small as possible to reduce the quantity of test fuel needed, while not undersizing the combustor such that wall effects had a significant effect on the measured combustion performance. Based on computer modeling, a rich-lean, two-stage, nonslagging combustor was constructed to simulate a direct-fired gas turbine. This design was selected to maximize the information that could be obtained on the impact of low-rank coal`s unique properties on the gas turbine combustor, its turbomachinery, and the required hot-gas cleanup devices (such as high-temperature/high-pressure (HTHP) cyclones). Seventeen successful combustion tests using coal-water fuels were completed. These tests included seven tests with a commercially available Otisca Industries-produced, Taggart seam bituminous fuel and five tests each with physically and chemically cleaned Beulah-Zap lignite and a chemically cleaned Kemmerer subbituminous fuel. LRC-fueled heat engine testing conducted at the Energy and Environmental Research Center (EERC) has indicated that LRC fuels perform very well in short residence time heat engine combustion systems. Analyses of the emission and fly ash samples highlighted the superior burnout experienced by the LRC fuels as compared to the bituminous fuel even under a longer residence time profile for the bituminous fuel.

Swanson, M.L.; Tibbetts, J.E.

1992-12-31T23:59:59.000Z

283

Mineral matter transformations in a pressurized drop-tube furnace  

SciTech Connect (OSTI)

To meet the objectives of the program, a pressurized combustion vessel was built to allow the operating parameters of a direct-fired gas turbine combustor to be simulated. One goal in building this equipment was to design the gas turbine simulator as small as possible to reduce the quantity of test fuel needed, while not undersizing the combustor such that wall effects had a significant effect on the measured combustion performance. Based on computer modeling, a rich-lean, two-stage, nonslagging combustor was constructed to simulate a direct-fired gas turbine. This design was selected to maximize the information that could be obtained on the impact of low-rank coal's unique properties on the gas turbine combustor, its turbomachinery, and the required hot-gas cleanup devices (such as high-temperature/high-pressure (HTHP) cyclones). Seventeen successful combustion tests using coal-water fuels were completed. These tests included seven tests with a commercially available Otisca Industries-produced, Taggart seam bituminous fuel and five tests each with physically and chemically cleaned Beulah-Zap lignite and a chemically cleaned Kemmerer subbituminous fuel. LRC-fueled heat engine testing conducted at the Energy and Environmental Research Center (EERC) has indicated that LRC fuels perform very well in short residence time heat engine combustion systems. Analyses of the emission and fly ash samples highlighted the superior burnout experienced by the LRC fuels as compared to the bituminous fuel even under a longer residence time profile for the bituminous fuel.

Swanson, M.L.; Tibbetts, J.E.

1992-01-01T23:59:59.000Z

284

Development of Larger Diameter High Pressure CNG Cylinder Manufactured by Piercing and Drawing for Natural Gas Vehicle  

Broader source: Energy.gov [DOE]

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

285

Tow Vessel | Open Energy Information  

Open Energy Info (EERE)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data CenterFranconia, Virginia:FAQ < RAPID Jump to:Seadov Pty LtdSteen,Ltd JumpOperations JumpTooele County, Utah:JumpVessel Jump to:

286

Great Wall Starbucks  

E-Print Network [OSTI]

along the Great Wall. When you think about it, it's not a bad marketing strategy: the Wall is high, the stairs relentless; what better than an espresso to energize you for the steep climb up? On second thought, make that a double. #ceas #china #tsutsui...

Hacker, Randi; Gatewood, Tyler; Tsutsui, William

2006-03-29T23:59:59.000Z

287

Purge gas protected transportable pressurized fuel cell modules and their operation in a power plant  

DOE Patents [OSTI]

A fuel cell generator apparatus and method of its operation involves: passing pressurized oxidant gas and pressurized fuel gas into modules containing fuel cells, where the modules are each enclosed by a module housing surrounded by an axially elongated pressure vessel, and where there is a purge gas volume between the module housing and pressure vessel; passing pressurized purge gas through the purge gas volume to dilute any unreacted fuel gas from the modules; and passing exhaust gas and circulated purge gas and any unreacted fuel gas out of the pressure vessel; where the fuel cell generator apparatus is transportable when the pressure vessel is horizontally disposed, providing a low center of gravity. 11 figs.

Zafred, P.R.; Dederer, J.T.; Gillett, J.E.; Basel, R.A.; Antenucci, A.B.

1996-11-12T23:59:59.000Z

288

Purge gas protected transportable pressurized fuel cell modules and their operation in a power plant  

DOE Patents [OSTI]

A fuel cell generator apparatus and method of its operation involves: passing pressurized oxidant gas, (O) and pressurized fuel gas, (F), into fuel cell modules, (10 and 12), containing fuel cells, where the modules are each enclosed by a module housing (18), surrounded by an axially elongated pressure vessel (64), where there is a purge gas volume, (62), between the module housing and pressure vessel; passing pressurized purge gas, (P), through the purge gas volume, (62), to dilute any unreacted fuel gas from the modules; and passing exhaust gas, (82), and circulated purge gas and any unreacted fuel gas out of the pressure vessel; where the fuel cell generator apparatus is transpatable when the pressure vessel (64) is horizontally disposed, providing a low center of gravity.

Zafred, Paolo R. (Pittsburgh, PA); Dederer, Jeffrey T. (Valencia, PA); Gillett, James E. (Greensburg, PA); Basel, Richard A. (Plub Borough, PA); Antenucci, Annette B. (Pittsburgh, PA)

1996-01-01T23:59:59.000Z

289

Pressure Safety Program Implementation at ORNL  

SciTech Connect (OSTI)

The Oak Ridge National Laboratory (ORNL) is a US Department of Energy (DOE) facility that is managed by UT-Battelle, LLC. In February 2006, DOE promulgated worker safety and health regulations to govern contractor activities at DOE sites. These regulations, which are provided in 10 CFR 851, Worker Safety and Health Program, establish requirements for worker safety and health program that reduce or prevent occupational injuries, illnesses, and accidental losses by providing DOE contractors and their workers with safe and healthful workplaces at DOE sites. The regulations state that contractors must achieve compliance no later than May 25, 2007. According to 10 CFR 851, Subpart C, Specific Program Requirements, contractors must have a structured approach to their worker safety and health programs that at a minimum includes provisions for pressure safety. In implementing the structured approach for pressure safety, contractors must establish safety policies and procedures to ensure that pressure systems are designed, fabricated, tested, inspected, maintained, repaired, and operated by trained, qualified personnel in accordance with applicable sound engineering principles. In addition, contractors must ensure that all pressure vessels, boilers, air receivers, and supporting piping systems conform to (1) applicable American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (2004) Sections I through XII, including applicable code cases; (2) applicable ASME B31 piping codes; and (3) the strictest applicable state and local codes. When national consensus codes are not applicable because of pressure range, vessel geometry, use of special materials, etc., contractors must implement measures to provide equivalent protection and ensure a level of safety greater than or equal to the level of protection afforded by the ASME or applicable state or local codes. This report documents the work performed to address legacy pressure vessel deficiencies and comply with pressure safety requirements in 10 CFR 851. It also describes actions taken to develop and implement ORNL’s Pressure Safety Program.

Lower, Mark [ORNL; Etheridge, Tom [ORNL; Oland, C. Barry [XCEL Engineering, Inc.

2013-01-01T23:59:59.000Z

290

Pressure &Pressure & TemperatureTemperature  

E-Print Network [OSTI]

to measure atmospheric pressure, and thermometer toprobe to measure atmospheric pressure, and thermometer toprobe to measure atmospheric pressure, and thermometer toprobe to measure atmospheric pressure, and thermometer to measure air temperature.measure air temperature.measure air temperature.measure air temperature

California at Santa Cruz, University of

291

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

SciTech Connect (OSTI)

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

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

2012-10-01T23:59:59.000Z

292

Foam vessel for cryogenic fluid storage  

DOE Patents [OSTI]

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

Spear, Jonathan D (San Francisco, CA)

2011-07-05T23:59:59.000Z

293

Fabrication of Separator Demonstration Facility process vessel  

SciTech Connect (OSTI)

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

Oberst, E.F.

1985-01-15T23:59:59.000Z

294

Application for Amendment 80 Vessel Replacement Page 1 of 6  

E-Print Network [OSTI]

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

295

Radiation embrittlement of PWR vessel supports  

SciTech Connect (OSTI)

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

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

1989-01-01T23:59:59.000Z

296

Wall-collision line broadening of molecular oxygen within nanoporous materials  

SciTech Connect (OSTI)

Wall-collision broadening of near-infrared absorption lines of molecular oxygen confined in nanoporous zirconia is studied by employing high-resolution diode-laser spectroscopy. The broadening is studied for pores of different sizes under a range of pressures, providing new insights on how wall collisions and intermolecular collisions influence the total spectroscopic line profile. The pressure series show that wall-collision broadening is relatively more prominent under reduced pressures, enabling sensitive means to probe pore sizes of porous materials. In addition, we show that the total wall-collision-broadened profile strongly deviates from a Voigt profile and that wall-collision broadening exhibits an additive-like behavior to the pressure and Doppler broadening.

Xu, Can T.; Lewander, Maerta; Andersson-Engels, Stefan; Svensson, Tomas; Svanberg, Sune [Department of Physics, Lund University, P. O. Box 118, SE-221 00 Lund (Sweden); Adolfsson, Erik [Ceramic Materials, SWEREA IVF, Box 104, SE-431 22 Moelndal (Sweden)

2011-10-15T23:59:59.000Z

297

1 SOFE, Chicago, IL, June 27, 2011 Design of the ITER First Wall and  

E-Print Network [OSTI]

First Wall (FW) panel and a Shield Block (SB). It covers ~600 m2 Cooling water (3 MPa and 70°C) is supplied to the BM by manifolds supported off the vacuum vessel behind or to the side of the SB. #12 Pipes Be ?les Be ?les Normal Heat Flux Finger: · q'' = ~ 1-2 MW/m2 · Steel Cooling Pipes · HIP

Raffray, A. René

298

Thermal treatment wall  

DOE Patents [OSTI]

A thermal treatment wall emplaced to perform in-situ destruction of contaminants in groundwater. Thermal destruction of specific contaminants occurs by hydrous pyrolysis/oxidation at temperatures achievable by existing thermal remediation techniques (electrical heating or steam injection) in the presence of oxygen or soil mineral oxidants, such as MnO.sub.2. The thermal treatment wall can be installed in a variety of configurations depending on the specific objectives, and can be used for groundwater cleanup, wherein in-situ destruction of contaminants is carried out rather than extracting contaminated fluids to the surface, where they are to be cleaned. In addition, the thermal treatment wall can be used for both plume interdiction and near-wellhead in-situ groundwater treatment. Thus, this technique can be utilized for a variety of groundwater contamination problems.

Aines, Roger D. (Livermore, CA); Newmark, Robin L. (Livermore, CA); Knauss, Kevin G. (Livermore, CA)

2000-01-01T23:59:59.000Z

299

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

SciTech Connect (OSTI)

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

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

2010-07-01T23:59:59.000Z

300

Thermal wake/vessel detection technique  

DOE Patents [OSTI]

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

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

2012-01-10T23:59:59.000Z

Note: This page contains sample records for the topic "walled pressure vessel" 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

Modelling of a coke oven heating wall M. Landreau, D. Isler, Centre de Pyrolyse de Marienau (CPM)  

E-Print Network [OSTI]

- 1 - Modelling of a coke oven heating wall M. Landreau, D. Isler, Centre de Pyrolyse de Marienau with thermomechanical modelling of a coke oven heating wall. The objective is to define the safe limits of coke oven of walls, roof and larry car, pre-stresses (anchoring system), lateral pressure due to coal pushing A 3D

Boyer, Edmond

302

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

303

Subcooled Boiling Near a Heated Wall  

SciTech Connect (OSTI)

Experimental measurements of void fraction, bubble frequency, and velocity are obtained in subcooled R-134a flowing over a heated flat plate near an unheated wall and compared to analytical predictions. The measurements were obtained for a fixed system pressure and mass flow rate (P = 2.4 MPa and w = 106 kg/hr) at various inlet liquid temperatures. During the experiments, electrical power was applied at a constant rate to one side of the test section. The local void fraction data, acquired with a hot-film anemometer probe, showed the existence of a significant peak near the heated wall and a smaller secondary peak near the unheated wall for the larger inlet subcoolings. Local vapor velocity data, taken with the hot-film probe and a laser Doppler velocimeter, showed broad maxima near the centerline between the heated and unheated plates. Significant temperature gradients near the heated wall were observed for large inlet subcooling. Bubble size data, inferred from measurements of void fraction, bubble frequency and vapor velocity, when combined with the measured bubble chord length distributions illustrate the transition from pure three dimensional spherical to two-dimensional planar bubble flow, the latter being initiated when the bubbles fill the gap between the plates. These various two-phase flow measurements were used for development of a multidimensional, four-field calculational method; comparisons of the data to the calculations show reasonable agreement.

T.A. Trabold; C.C. Maneri; P.F. Vassallo; D.M. Considine

2000-10-27T23:59:59.000Z

304

ASME post construction pressure technology codes  

SciTech Connect (OSTI)

The need to continue to operate pressurized equipment and other facilities in a safe, reliable and cost effective manner has led to the development of many new approaches to in-service inspection, flaw evaluation, and repair. Interest on the part of users, regulatory authorities and others in standardizing these approaches has led to the formation of a new ASME Main Committee on Post Construction under the Board on Pressure Technology Codes and Standards, and a new Division of the Pressure Vessel Research Council on Continued Operation of Equipment. This paper provides a brief overview of these activities.

Sims, J.R. [Exxon Research and Engineering Co., Florham Park, NJ (United States)

1996-12-01T23:59:59.000Z

305

Single-Walle 4. Single-Walled Carbon Nanotubes  

E-Print Network [OSTI]

applications, carbon nanotube research is ac- tively being pursued in diverse areas including energy storage105 Single-Walle 4. Single-Walled Carbon Nanotubes Sebastien Nanot, Nicholas A. Thompson, Ji Single-walled carbon nanotubes (SWCNTs) are hol- low, long cylinders with extremely large aspect ratios

Kono, Junichiro

306

High-R Walls for Remodeling: Wall Cavity Moisture Monitoring  

SciTech Connect (OSTI)

The focus of the study is on the performance of wall systems, and in particular, the moisture characteristics inside the wall cavity and in the wood sheathing. Furthermore, while this research will initially address new home construction, the goal is to address potential moisture issues in wall cavities of existing homes when insulation and air sealing improvements are made.

Wiehagen, J.; Kochkin, V.

2012-12-01T23:59:59.000Z

307

Integrating Biomechanics, Hemodynamics, and Vascular Adaptation to Relate Mechanisms of Vascular Adaptation to Arterial Pulsatile Pressure in Health and Disease  

E-Print Network [OSTI]

local changes in pulsatile blood pressures and flows lead to changes in local endothelial shear stress and circumferential wall stress. The field of mechanobiology has identified how local changes in wall circumferential stress and endothelial shear...

Nguyen, Phuc Hoang

2014-08-07T23:59:59.000Z

308

Earth pressures and deformations in civil infrastructure in expansive soils  

E-Print Network [OSTI]

This dissertation includes the three major parts of the study: volume change, and lateral earth pressure due to suction change in expansive clay soils, and design of civil infrastructure drilled pier, retaining wall and pavement in expansive soils...

Hong, Gyeong Taek

2008-10-10T23:59:59.000Z

309

Negative pressure characteristics of an evaporating meniscus at nanoscale  

E-Print Network [OSTI]

This study aims at understanding the characteristics of negative liquid pressures at the nanoscale using molecular dynamics simulation. A nano-meniscus is formed by placing liquid argon on a platinum wall between two ...

Maroo, Shalabh C.

2011-01-01T23:59:59.000Z

310

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

SciTech Connect (OSTI)

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

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

2012-09-27T23:59:59.000Z

311

Effects of external pressure on the terminal lymphatic flow rate  

E-Print Network [OSTI]

pressure applied to the skin of the canine cause the terminal lymphat- ic flow rate to increase until the external pressure reaches 60mm Hg. At an external pressure of 60mm Hg reduced lymphatic flow is observed in some of the test animals. At 75mm Hg... resulting from the external pressure begins to col- lapse the lymph vessels. External pressure between 60 and 75mm Hg restricts or completely occludes the terminal lymphatic flow rate. ACKNOWLEDGENENTS I would like to express my appreciation...

Seale, James Lewis

1981-01-01T23:59:59.000Z

312

Pressure Safety of JLAB 12GeV Upgrade Cryomodule  

SciTech Connect (OSTI)

This paper reviews pressure safety considerations, per the US Department of Energy (DOE) 10CFR851 Final Rule [1], which are being implemented during construction of the 100 Megavolt Cryomodule (C100 CM) for Jefferson Lab’s 12 GeV Upgrade Project. The C100 CM contains several essential subsystems that require pressure safety measures: piping in the supply and return end cans, piping in the thermal shield and the helium headers, the helium vessel assembly which includes high RRR niobium cavities, the end cans, and the vacuum vessel. Due to the vessel sizes and pressure ranges, applicable national consensus code rules are applied. When national consensus codes are not applicable, equivalent design and fabrication approaches are identified and implemented. Considerations for design, material qualification, fabrication, inspection and examination are summarized. In addition, JLAB’s methodologies for implementation of the 10 CFR 851 requirements are described.

Cheng, Gary [JLAB; Wiseman, Mark A. [JLAB; Daly, Ed [JLAB

2009-11-01T23:59:59.000Z

313

1.3 GHz Cavity Weld to Helium Vessel A. Schmidt, A. Matheisen  

E-Print Network [OSTI]

's preparation and fabrication technique. In the actual design of the DESY cryounit (module), 8 resonators one by heat conductivity of the niobium walls. The eight separated tanks and the superconducting quadrupole and cold pressure resistance of the welds, is the preservation of the cavity characteristics like

314

Near-infrared spectroscopy for the measurement of glucose in an integrated rotating wall vessel  

E-Print Network [OSTI]

culture media cannot fulfill these requirements. Therefore a near-infrared spectroscopy system is proposed that can potentially perform the required measurements on-line and without any interaction with the cell culture media. Two types of solutions...

Galvan, Mark

2012-06-07T23:59:59.000Z

315

Covering Walls With Fabrics.  

E-Print Network [OSTI]

the glue a dull surface to adhere to. Fill any gouges or nail holes with patching plaster and sand smooth after they have dried thoroughly. Minor ripples can be covered with spackling compound, a plaster-like substance that is spread thinly... during dry weather and in a well-ventilated room. Cut each panel 3 inches longer than the ceiling height. Match and cut sufficient fabric widths to cover completely one wall at a time. Start with Corner I nstall the first fabric panel so...

Anonymous,

1979-01-01T23:59:59.000Z

316

Poorly Draining Soil Reinforced with Geosynthetic with in Plane Drainage: Efficiency and Pore Pressure Behavior  

E-Print Network [OSTI]

drainage system contributes to the dissipation of porous pressure when the water content of the soil test, porous pressure 1 INTRODUCTION Sustainable technologies can be defined as the use of methods of cement is to build mechanically stabilized earth walls rather than concrete walls. Conventionally, freely

Zornberg, Jorge G.

317

METC/3M Cooperative Agreement CRADA 94-024 high temperature high pressure filter materials exposure test program. Volume 2, Final report  

SciTech Connect (OSTI)

This report is a summary of the results of activities of the particulate monitoring group in support of the METC/3M CRADA 94024. Online particulate monitoring began in June 1994 and ended in October, 1994. The particulate monitoring group participated in four MGCR runs (No. 7 through No. 10). The instrument used in measuring the particle loadings (particle counts and size distribution) is the Particle Measuring Systems Classical Scattering Aerosol Spectrometer Probe High Temperature and High Pressure (PMS Model CSASP-100-HTHP). This PMS unit is rated to operate at temperatures up to 540{degree}C and gage pressures up to 2.0 MPa. Gas stream conditions, temperature at 540{degree}C, gage pressure at 2.93 MPa, and gas flowrate at 0.0157 SCM per second, precluded the direct measurement of particulate loadings in the gas stream with the PMS unit. A side stream was extracted from the gas stream after it came over to the MGCR, Modular Gas Cleanup Rig, from the FBG, pressurized Fluidized-Bed Gasifier, but before it entered the filter testing vessel. A sampling probe of 0.635 cm O.D. thin wall stainless steel tubing was used for extracting the sample gas isokinetically based on the expected flowrate. The sample gas stream was further split into two streams; one was directed to the PMS unit and the other to the alkali monitor unit. The alkali monitor unit was not used during runs No. 7 through No. 10.

NONE

1995-06-01T23:59:59.000Z

318

METC/Shell Cooperative Agreement CRADA 93-011 high temperature high pressure filtration and sorbent test program. Volume 2, Final report  

SciTech Connect (OSTI)

This report is a summary of the results of activities of the particulate monitoring group in support of the METC/Shell CRADA 93-011. Online particulate monitoring began in August 1993 and ended in October 1994. The particulate monitoring group participated in six MGCR runs (No. 5 through No. 10). The instrument used in measuring the particle loadings (particle counts and size distribution) is the Particle Measuring Systems Classical Scattering Aerosol Spectrometer Probe High Temperature and High Pressure (PMS Model CSASP-100-HTHP). This PMS unit is rated to operate at temperatures up to 540{degree}C and gage pressures up to 2.07 MPa. Gas stream conditions, temperature at 540{degree}C, gage pressure at 2.93 MPa, and gas flowrate at 0.0157 SCM per second, precluded the direct measurement of particulate loadings in the gas stream with the PMS unit. A side stream was extracted from the gas stream after it came over to the MGCR, (Modular Gas Cleanup Rig), from the FBG, pressurized fluidized-bed gasifier, but before it entered the filter testing vessel. A sampling probe of 0.635 cm O.D. thin wall stainless steel tubing was used for extracting the sample gas isokinetically based on the expected flowrate. The sample gas stream was further split into two streams; one was directed to the PMS unit and the other to the alkali monitor unit.

NONE

1995-06-01T23:59:59.000Z

319

Saltstone Osmotic Pressure  

SciTech Connect (OSTI)

Recent research into the moisture retention properties of saltstone suggest that osmotic pressure may play a potentially significant role in contaminant transport (Dixon et al., 2009 and Dixon, 2011). The Savannah River Remediation Closure and Disposal Assessments Group requested the Savannah River National Laboratory (SRNL) to conduct a literature search on osmotic potential as it relates to contaminant transport and to develop a conceptual model of saltstone that incorporates osmotic potential. This report presents the findings of the literature review and presents a conceptual model for saltstone that incorporates osmotic potential. The task was requested through Task Technical Request HLW-SSF-TTR-2013-0004. Simulated saltstone typically has very low permeability (Dixon et al. 2008) and pore water that contains a large concentration of dissolved salts (Flach and Smith 2013). Pore water in simulated saltstone has a high salt concentration relative to pore water in concrete and groundwater. This contrast in salt concentration can generate high osmotic pressures if simulated saltstone has the properties of a semipermeable membrane. Estimates of osmotic pressure using results from the analysis of pore water collected from simulated saltstone show that an osmotic pressure up to 2790 psig could be generated within the saltstone. Most semi-permeable materials are non-ideal and have an osmotic efficiency <1 and as a result actual osmotic pressures are less than theoretical pressures. Observations from laboratory tests of simulated saltstone indicate that it may exhibit the behavior of a semi-permeable membrane. After several weeks of back pressure saturation in a flexible wall permeameter (FWP) the membrane containing a simulated saltstone sample appeared to have bubbles underneath it. Upon removal from the FWP the specimen was examined and it was determined that the bubbles were due to liquid that had accumulated between the membrane and the sample. One possible explanation for the accumulation of solution between the membrane and sample is the development of osmotic pressure within the sample. Osmotic pressure will affect fluid flow and contaminant transport and may result in the changes to the internal structure of the semi-permeable material. B?nard et al. 2008 reported swelling of wet cured Portland cement mortars containing salts of NaNO{sub 3}, KNO{sub 3}, Na{sub 3}PO{sub 4}x12H {sub 2}O, and K{sub 3}PO{sub 4} when exposed to a dilute solution. Typically hydraulic head is considered the only driving force for groundwater in groundwater models. If a low permeability material containing a concentrated salt solution is present in the hydrogeologic sequence large osmotic pressures may develop and lead to misinterpretation of groundwater flow and solute transport. The osmotic pressure in the semi-permeable material can significantly impact groundwater flow in the vicinity of the semi-permeable material. One possible outcome is that groundwater will flow into the semi-permeable material resulting in hydrologic containment within the membrane. Additionally, hyperfiltration can occur within semi-permeable materials when water moves through a membrane into the more concentrated solution and dissolved constituents are retained in the lower concentration solution. Groundwater flow and transport equations that incorporate chemical gradients (osmosis) have been developed. These equations are referred to as coupled flow equations. Currently groundwater modeling to assess the performance of saltstone waste forms is conducted using the PORFLOW groundwater flow and transport model. PORFLOW does not include coupled flow from chemico-osmotic gradients and therefore numerical simulation of the effect of coupled flow on contaminant transport in and around saltstone cannot be assessed. Most natural semi-permeable membranes are non-ideal membranes and do not restrict all movement of solutes and as a result theoretical osmotic potential is not realized. Osmotic efficiency is a parameter in the coupled flow equation that accounts for the

Nichols, Ralph L.; Dixon, Kenneth L.

2013-09-23T23:59:59.000Z

320

Security Walls, LLC WIPP 2009  

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

Walls, LLC Waste Isolation Pilot Plant Department of Energy Voluntary Protection Program Onsite Review March 3-4, 2009 The Department of Energy (DOE) Voluntary Protection Program...

Note: This page contains sample records for the topic "walled pressure vessel" 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

Liquid Walls Innovative Concepts for First Walls and Blankets  

E-Print Network [OSTI]

with existing technology · Size of plasma devices and power plants can be substantially reduced High PoloidalLiquid Walls Innovative Concepts for First Walls and Blankets Mohamed Abdou Professor, Mechanical as part of the US Restructured Fusion Program Strategy to enhance innovation · Natural Questions

Abdou, Mohamed

322

Oven wall panel construction  

DOE Patents [OSTI]

An oven roof or wall is formed from modular panels, each of which comprises an inner fabric and an outer fabric. Each such fabric is formed with an angle iron framework and somewhat resilient tie-bars or welded at their ends to flanges of the angle irons to maintain the inner and outer frameworks in spaced disposition while minimizing heat transfer by conduction and permitting some degree of relative movement on expansion and contraction of the module components. Suitable thermal insulation is provided within the module. Panels or skins are secured to the fabric frameworks and each such skin is secured to a framework and projects laterally so as slidingly to overlie the adjacent frame member of an adjacent panel in turn to permit relative movement during expansion and contraction.

Ellison, Kenneth (20 Avondale Cres., Markham, CA); Whike, Alan S. (R.R. #1, Caledon East, both of Ontario, CA)

1980-04-22T23:59:59.000Z

323

Study Reveals Challenges and Opportunities Related to Vessels...  

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

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

324

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

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

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

325

Universal fuel basket for use with an improved oxide reduction vessel and electrorefiner vessel  

DOE Patents [OSTI]

A basket, for use in the reduction of UO.sub.2 to uranium metal and in the electrorefining of uranium metal, having a continuous annulus between inner and outer perforated cylindrical walls, with a screen adjacent to each wall. A substantially solid bottom and top plate enclose the continuous annulus defining a fuel bed. A plurality of scrapers are mounted adjacent to the outer wall extending longitudinally thereof, and there is a mechanism enabling the basket to be transported remotely.

Herrmann, Steven D. (Idaho Falls, ID); Mariani, Robert D. (Idaho Falls, ID)

2002-01-01T23:59:59.000Z

326

Treating exhaust gas from a pressurized fluidized bed reaction system  

DOE Patents [OSTI]

Hot gases from a pressurized fluidized bed reactor system are purified. Under super atmospheric pressure conditions hot exhaust gases are passed through a particle separator, forming a filtrate cake on the surface of the separator, and a reducing agent--such as an NO{sub x} reducing agent (like ammonia)--is introduced into the exhaust gases just prior to or just after particle separation. The retention time of the introduced reducing agent is enhanced by providing a low gas velocity (e.g. about 1--20 cm/s) during passage of the gas through the filtrate cake while at super atmospheric pressure. Separation takes place within a distinct pressure vessel, the interior of which is at a pressure of about 2--100 bar, and introduction of reducing agent can take place at multiple locations (one associated with each filter element in the pressure vessel), or at one or more locations just prior to passage of clean gas out of the pressure vessel (typically passed to a turbine). 8 figs.

Isaksson, J.; Koskinen, J.

1995-08-22T23:59:59.000Z

327

Apparatus and method for fatigue testing of a material specimen in a high-pressure fluid environment  

DOE Patents [OSTI]

The invention provides fatigue testing of a material specimen while the specimen is disposed in a high pressure fluid environment. A specimen is placed between receivers in an end cap of a vessel and a piston that is moveable within the vessel. Pressurized fluid is provided to compression and tension chambers defined between the piston and the vessel. When the pressure in the compression chamber is greater than the pressure in the tension chamber, the specimen is subjected to a compression force. When the pressure in the tension chamber is greater than the pressure in the compression chamber, the specimen is subjected to a tension force. While the specimen is subjected to either force, it is also surrounded by the pressurized fluid in the tension chamber. In some examples, the specimen is surrounded by hydrogen.

Wang, Jy-An; Feng, Zhili; Anovitz, Lawrence M; Liu, Kenneth C

2013-06-04T23:59:59.000Z

328

Product Sheet Wall Mount Lift  

E-Print Network [OSTI]

Product Sheet Wall Mount Lift Ergotron® Neo-FlexTM 870-05-061, rev. 12/11/07 www. Less effort. Feel the difference. Add greater range of movement to your LCD display or TV with the Neo-Flex Wall Mount Lift! CF patented lift-and-pivot motion technology adjusts with a light touch. Raise

Saskatchewan, University of

329

Moisture Research - Optimizing Wall Assemblies  

SciTech Connect (OSTI)

The Consortium for Advanced Residential Buildings (CARB) evaluated several different configurations of wall assemblies to determine the accuracy of moisture modeling and make recommendations to ensure durable, efficient assemblies. WUFI and THERM were used to model the hygrothermal and heat transfer characteristics of these walls.

Arena, L.; Mantha, P.

2013-05-01T23:59:59.000Z

330

From Cold War to cold vessels  

SciTech Connect (OSTI)

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

Melrath, C.

1996-09-01T23:59:59.000Z

331

TMI-2 reactor vessel head removal  

SciTech Connect (OSTI)

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

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

1984-12-01T23:59:59.000Z

332

TMI-2 reactor vessel head removal  

SciTech Connect (OSTI)

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

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

1985-09-01T23:59:59.000Z

333

Pressurized fluidized bed reactor and a method of operating the same  

DOE Patents [OSTI]

A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

Isaksson, Juhani (Karhula, FI)

1996-01-01T23:59:59.000Z

334

Pressurized fluidized bed reactor and a method of operating the same  

DOE Patents [OSTI]

A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

Isaksson, J.

1996-02-20T23:59:59.000Z

335

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

SciTech Connect (OSTI)

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

Raske, D.T.

1995-06-01T23:59:59.000Z

336

Method for forming a bladder for fluid storage vessels  

DOE Patents [OSTI]

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

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

2000-01-01T23:59:59.000Z

337

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

SciTech Connect (OSTI)

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

Not Available

1984-11-01T23:59:59.000Z

338

Innovative Composite Wall System for Sheathing Masonry Walls  

SciTech Connect (OSTI)

Existing Housing - Much of the older multifamily housing stock in the United States includes units in structures with uninsulated masonry walls. Included in this stock are two- and three-story walk-up apartments, larger apartment complexes, and public housing (both high- rise and townhouse). This older multifamily housing has seen years of heavy use that may have left the plaster wall marred or damaged. Long- term building settlement or movement may have cracked the plaster, sometimes severely. Moisture from invented kitchens and baths may have caused condensation on uninsulated exterior walls. At best this condensation has left stains on the paint or wallpaper. At worst it has supported mold and mildew growth, fouling the air and creating unhealthy living conditions. Deteriorating plaster and flaking paint also result from wet walls. The presence of flaking, lead-based paint in older (pre-1978) housing is a major public health concern. Children can suffer permanent mental handicaps and psychological disorders if they are subjected to elevated levels of lead, while adults can suffer hypertension and other maladies. Studies have found that, in some urban communities with older housing stocks, over 35% of children tested have elevated blood lead levels (Hastings, et al.: 1997). Nationally, nearly 22% of black, non-hispanic children living in pre-1946 housing were found to have elevated levels of lead in their blood (MWWR Article: February 21,1997). The deterioration of many of these walls is to the point that lead can freely enter the living space.

Wendt, Robert L. [Oak Ridge National Lab., TN (United States); Cavallo, James [Argonne National Lab., IL (United States)

1997-09-25T23:59:59.000Z

339

Autonomous Radiation Monitoring of Small Vessels  

SciTech Connect (OSTI)

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

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

2010-01-01T23:59:59.000Z

340

DefectDomain Wall Interactions in Trigonal  

E-Print Network [OSTI]

Defect­Domain Wall Interactions in Trigonal Ferroelectrics Venkatraman Gopalan,1 Volkmar Dierolf,2 walls in the trigonal ferroelectrics lithium niobate and lithium tantalate. It is shown that extrinsic questions re- garding intrinsic widths, defect­domain wall interactions, and static versus dynamic wall

Gopalan, Venkatraman

Note: This page contains sample records for the topic "walled pressure vessel" 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

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

DOE Patents [OSTI]

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

Davin, James (Gilroy, CA)

1992-01-01T23:59:59.000Z

342

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

DOE Patents [OSTI]

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

Davin, J.

1992-12-01T23:59:59.000Z

343

An apparatus for studying scintillator properties at high isostatic pressures  

SciTech Connect (OSTI)

We describe the design and operation of a unique hydraulic press for the study of scintillator materials under isostatic pressure. This press, capable of developing a pressure of a gigapascal, consists of a large sample chamber pressurized by a two-stage hydraulic amplifier. The optical detection of the scintillation light emitted by the sample is performed, through a large aperture optical port, by a photodetector located outside the pressure vessel. In addition to providing essential pressure-dependent studies on the emission characteristics of radioluminescent materials, this apparatus is being developed to elucidate the mechanisms behind the recently observed dependency of light-yield nonproportionality on electronic band structure. The variation of the light output of a Tl:CsI crystal under 511-keV gamma excitation and hydrostatic pressure is given as an example.

Gaume, R. M. [College of Optics and Photonics (CREOL) and NanoScience Technology Center, University of Central Florida, Orlando, Florida 32816 (United States); Lam, S.; Gascon, M.; Feigelson, R. S. [Department of Materials Science and Engineering, Stanford University, Stanford, California 94305 (United States); Setyawan, W. [Pacific Northwest National Laboratory, Richland, Washington 99352 (United States); Curtarolo, S. [Department of Mechanical Engineering and Materials Science, Duke University, Durham, North Carolina 27708 (United States)

2013-01-15T23:59:59.000Z

344

Continuous growth of single-wall carbon nanotubes using chemical vapor deposition  

DOE Patents [OSTI]

The invention relates to a chemical vapor deposition process for the continuous growth of a carbon single-wall nanotube where a carbon-containing gas composition is contacted with a porous membrane and decomposed in the presence of a catalyst to grow single-wall carbon nanotube material. A pressure differential exists across the porous membrane such that the pressure on one side of the membrane is less than that on the other side of the membrane. The single-wall carbon nanotube growth may occur predominately on the low-pressure side of the membrane or, in a different embodiment of the invention, may occur predominately in between the catalyst and the membrane. The invention also relates to an apparatus used with the carbon vapor deposition process.

Grigorian, Leonid; Hornyak, Louis; Dillon, Anne C; Heben, Michael J

2014-09-23T23:59:59.000Z

345

Continuous growth of single-wall carbon nanotubes using chemical vapor deposition  

DOE Patents [OSTI]

The invention relates to a chemical vapor deposition process for the continuous growth of a carbon single-wall nanotube where a carbon-containing gas composition is contacted with a porous membrane and decomposed in the presence of a catalyst to grow single-wall carbon nanotube material. A pressure differential exists across the porous membrane such that the pressure on one side of the membrane is less than that on the other side of the membrane. The single-wall carbon nanotube growth may occur predominately on the low-pressure side of the membrane or, in a different embodiment of the invention, may occur predominately in between the catalyst and the membrane. The invention also relates to an apparatus used with the carbon vapor deposition process.

Grigorian, Leonid (Raymond, OH); Hornyak, Louis (Evergreen, CO); Dillon, Anne C (Boulder, CO); Heben, Michael J (Denver, CO)

2008-10-07T23:59:59.000Z

346

Turbine airfoil with outer wall thickness indicators  

DOE Patents [OSTI]

A turbine airfoil usable in a turbine engine and including a depth indicator for determining outer wall blade thickness. The airfoil may include an outer wall having a plurality of grooves in the outer surface of the outer wall. The grooves may have a depth that represents a desired outer surface and wall thickness of the outer wall. The material forming an outer surface of the outer wall may be removed to be flush with an innermost point in each groove, thereby reducing the wall thickness and increasing efficiency. The plurality of grooves may be positioned in a radially outer region of the airfoil proximate to the tip.

Marra, John J; James, Allister W; Merrill, Gary B

2013-08-06T23:59:59.000Z

347

Webinar: Material Characterization of Storage Vessels for Fuel Cell Forklifts  

Broader source: Energy.gov [DOE]

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

348

A Xenon Condenser with a Remote Liquid Storage Vessel  

E-Print Network [OSTI]

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

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

2009-07-25T23:59:59.000Z

349

Support pedestals for interconnecting a cover and nozzle band wall in a gas turbine nozzle segment  

DOE Patents [OSTI]

A gas turbine nozzle segment has outer and inner band portions. Each band portion includes a nozzle wall, a cover and an impingement plate between the cover and nozzle wall defining two cavities on opposite sides of the impingement plate. Cooling steam is supplied to one cavity for flow through the apertures of the impingement plate to cool the nozzle wall. Structural pedestals interconnect the cover and nozzle wall and pass through holes in the impingement plate to reduce localized stress otherwise resulting from a difference in pressure within the chamber of the nozzle segment and the hot gas path and the fixed turbine casing surrounding the nozzle stage. The pedestals may be cast or welded to the cover and nozzle wall.

Yu, Yufeng Phillip (Simpsonville, SC); Itzel, Gary Michael (Simpsonville, SC); Webbon, Waylon Willard (Greenville, SC); Bagepalli, Radhakrishna (Schenectady, NY); Burdgick, Steven Sebastian (Schenectady, NY); Kellock, Iain Robertson (Simpsonville, SC)

2002-01-01T23:59:59.000Z

350

MHD Electrode and wall constructions  

DOE Patents [OSTI]

Electrode and wall constructions for the walls of a channel transmitting the hot plasma in a magnetohydrodynamic generator. The electrodes and walls are made of a plurality of similar modules which are spaced from one another along the channel. The electrodes can be metallic or ceramic, and each module includes one or more electrodes which are exposed to the plasma and a metallic cooling bar which is spaced from the plasma and which has passages through which a cooling fluid flows to remove heat transmitted from the electrode to the cooling bar. Each electrode module is spaced from and electrically insulated from each adjacent module while interconnected by the cooling fluid which serially flows among selected modules. A wall module includes an electrically insulating ceramic body exposed to the plasma and affixed, preferably by mechanical clips or by brazing, to a metallic cooling bar spaced from the plasma and having cooling fluid passages. Each wall module is, similar to the electrode modules, electrically insulated from the adjacent modules and serially interconnected to other modules by the cooling fluid.

Way, Stewart (Columbia, MD); Lempert, Joseph (Penn Hills, PA)

1984-01-01T23:59:59.000Z

351

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

SciTech Connect (OSTI)

The objective of this project was to determine if the local brittle zone (LBZ) problem, encountered in the testing of the heat-affected zone (HAZ) part of welds in offshore platform construction, can also be found in reactor pressure vessel (RPV) welds. Both structures have multipass welds and grain coarsening along the fusion line. Literature was obtained that described the metallurgical evidence and the type of research work performed on offshore structure welds.

McCabe, D.E.

1999-09-01T23:59:59.000Z

352

Experimental Measurements and Numerical Prediction of the Effect of Waves on Mooring Line Forces for a Container Ship Moored to Pile Supported and Solid Wall Docks  

E-Print Network [OSTI]

710 ft ship and model dock. The dock construction, solid sheet wall or pile supported, was varied to add another aspect of a moored vessel. Mooring lines were modeled using 14 springs in typical mooring line arrangement to simulate the elastic...

Luai, Andres B

2013-05-03T23:59:59.000Z

353

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

SciTech Connect (OSTI)

Purpose. Mechanical thrombectomy (mTE) proved to be effective treating acute vessel occlusions with an acceptable rate of procedural complications. Potential long-term side effects of the vessel wall trauma caused by mechanical irritation of the endothelium are unknown up to now. Methods. From a retrospectively established database of 640 acute stroke treatments, we selected 261 patients with 265 embolic vessel occlusions treated successfully by mTE without permanent implantation of a stent. Analysis comprised the type of devices used and the number of passes performed. Digital subtraction angiography immediately after treatment was evaluated for vasospasm, dissection, and extravasation. Control angiographic images were evaluated for any morphological change compared to the immediate posttreatment angiographic run. Results. Recanalization was achieved with a median of one (range 1-10) mTE maneuvers. Vasospasm occurred in 69 territories (26.0 %) and was treated with glyceroltrinitrate in three. Dissection was observed in one vessel (0.4 %). Intraprocedural hemorrhage in two patients (0.8 %) was either wire or device induced. Follow-up digital subtraction angiography was available for 117 territories after a median of 107 days, revealing target vessel occlusion in one segment (0.9 %) and a de novo stenosis of four segments (3.4 %). All findings were clinically asymptomatic. Posttreatment vasospasm was more frequent in patients with de novo stenosis and occlusion (p = 0.038). Conclusion. De novo stenoses and occlusions occur in a small proportion of patients after mTE. Because all lesions were clinically asymptomatic, this finding does not affect the overall benefit of the treatment. Vasospasm may predict late vessel wall changes.

Kurre, Wiebke, E-mail: w.kurre@klinikum-stuttgart.de; Perez, Marta Aguilar; Horvath, Diana [Klinikum Stuttgart, Klinik fuer Diagnostische und Interventionelle Neuroradiologie (Germany); Schmid, Elisabeth; Baezner, Hansjoerg [Klinikum Stuttgart, Neurologische Klinik (Germany); Henkes, Hans, E-mail: HHHenkes@aol.com [Klinikum Stuttgart, Klinik fuer Diagnostische und Interventionelle Neuroradiologie (Germany)

2013-06-15T23:59:59.000Z

354

Domain walls riding the wave.  

SciTech Connect (OSTI)

Recent years have witnessed a rapid proliferation of electronic gadgets around the world. These devices are used for both communication and entertainment, and it is a fact that they account for a growing portion of household energy consumption and overall world consumption of electricity. Increasing the energy efficiency of these devices could have a far greater and immediate impact than a gradual switch to renewable energy sources. The advances in the area of spintronics are therefore very important, as gadgets are mostly comprised of memory and logic elements. Recent developments in controlled manipulation of magnetic domains in ferromagnet nanostructures have opened opportunities for novel device architectures. This new class of memories and logic gates could soon power millions of consumer electronic devices. The attractiveness of using domain-wall motion in electronics is due to its inherent reliability (no mechanical moving parts), scalability (3D scalable architectures such as in racetrack memory), and nonvolatility (retains information in the absence of power). The remaining obstacles in widespread use of 'racetrack-type' elements are the speed and the energy dissipation during the manipulation of domain walls. In their recent contribution to Physical Review Letters, Oleg Tretiakov, Yang Liu, and Artem Abanov from Texas A&M University in College Station, provide a theoretical description of domain-wall motion in nanoscale ferromagnets due to the spin-polarized currents. They find exact conditions for time-dependent resonant domain-wall movement, which could speed up the motion of domain walls while minimizing Ohmic losses. Movement of domain walls in ferromagnetic nanowires can be achieved by application of external magnetic fields or by passing a spin-polarized current through the nanowire itself. On the other hand, the readout of the domain state is done by measuring the resistance of the wire. Therefore, passing current through the ferromagnetic wire is the preferred method, as it combines manipulation and readout of the domain-wall state. The electrons that take part in the process of readout and manipulation of the domain-wall structure in the nanowire do so through the so-called spin transfer torque: When spin-polarized electrons in the ferromagnet nanowire pass through the domain wall they experience a nonuniform magnetization, and they try to align their spins with the local magnetic moments. The force that the electrons experience has a reaction force counterpart that 'pushes' the local magnetic moments, resulting in movement of the domain wall in the direction of the electron flow through the spin-transfer torque. The forces between the electrons and the local magnetic moments in the ferromagnet also create additional electrical resistance for the electrons passing through the domain wall. By measuring resistance across a segment of the nanowire, one determines if a domain wall is present; i.e., one can read the stored information. The interaction of the spin-polarized electrons with the domain wall in the ferromagnetic nanowire is not very efficient. Even for materials achieving high polarization of the free electrons, it is very difficult to move the magnetic domain wall. Several factors contribute to this problem, with imperfections of the ferromagnetic nanowire that cause domain-wall pinning being the dominant one. Permalloy nanowires, one of the best candidates for domain-wall-based memory and logic devices, require current densities of the order of 10{sup 8} A/cm{sup 2} in order to move a domain wall from a pinning well. Considering that this current has to pass through a relatively long wire, it is not very difficult to imagine that most of the energy will go to Joule heating. The efficiency of the process - the ratio of the energy converted to domain-wall motion to the total energy consumed - is comparable to that of an incandescent light bulb converting electricity to light. A step towards more efficient domain-wall-based memory devices is the advance of using alternating currents or curren

Karapetrov, G.; Novosad, V.; Materials Science Division

2010-11-01T23:59:59.000Z

355

Standard Practice for Determining NeutronExposures for Nuclear Reactor Vessel Support Structures  

E-Print Network [OSTI]

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

American Society for Testing and Materials. Philadelphia

2008-01-01T23:59:59.000Z

356

Coronary artery wall imaging in mice using osmium tetroxide and micro-computed tomography (micro-CT)  

SciTech Connect (OSTI)

The high spatial resolution of micro-computed tomography (micro-CT) is ideal for 3D imaging of coronary arteries in intact mouse heart specimens. Previously, micro-CT of mouse heart specimens utilized intravascular contrast agents that hardened within the vessel lumen and allowed a vascular cast to be made. However, for mouse coronary artery disease models, it is highly desirable to image coronary artery walls and highlight plaques. For this purpose, we describe an ex vivo contrast-enhanced micro-CT imaging technique based on tissue staining with osmium tetroxide (OsO{sub 4}) solution. As a tissue-staining contrast agent, OsO{sub 4} is retained in the vessel wall and surrounding tissue during the fixation process and cleared from the vessel lumens. Its high X-ray attenuation makes the artery wall visible in CT. Additionally, since OsO{sub 4} preferentially binds to lipids, it highlights lipid deposition in the artery wall. We performed micro-CT of heart specimens of 5- to 25-week-old C57BL/6 wild-type mice and 5- to 13-week-old apolipoprotein E knockout (apoE{sup -/-}) mice at 10 {mu}m resolution. The results show that walls of coronary arteries as small as 45 {mu}m in diameter are visible using a table-top micro-CT scanner. Similar image clarity was achieved with 1/2000th the scan time using a synchrotron CT scanner. In 13-week-old apoE mice, lipid-rich plaques are visible in the aorta. Our study shows that the combination of OsO{sub 4} and micro-CT permits the visualization of the coronary artery wall in intact mouse hearts.

Pai, Vinay M.; Kozlowski, Megan; Donahue, Danielle; Miller, Elishiah; Xiao, Xianghui; Chen, Marcus Y.; Yu, Zu-Xi; Connelly, Patricia; Jeffries, Kenneth; Wen, Han (NIH)

2012-05-10T23:59:59.000Z

357

Domain Walls, Triples and Acceleration  

E-Print Network [OSTI]

We present a construction of domain walls in string theory. The domain walls can bridge both Minkowski and AdS string vacua. A key ingredient in the construction are novel classical Yang-Mills configurations, including instantons, which interpolate between toroidal Yang-Mills vacua. Our construction provides a concrete framework for the study of inflating metrics in string theory. In some cases, the accelerating space-time comes with a holographic description. The general form of the holographic dual is a field theory with parameters that vary over space-time.

Travis Maxfield; Savdeep Sethi

2014-04-09T23:59:59.000Z

358

Heat-transfer coefficients in agitated vessels. Sensible heat models  

SciTech Connect (OSTI)

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

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

1995-12-01T23:59:59.000Z

359

Commercial marine vessel contributions to emission inventories. Final report  

SciTech Connect (OSTI)

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

Not Available

1991-10-07T23:59:59.000Z

360

Radial elasticity of multi-walled boron nitride nanotubes  

SciTech Connect (OSTI)

We investigated the radial mechanical properties of multi-walled boron nitride nanotubes (MW-BNNTs) using atomic force microscopy. The employed MW-BNNTs were synthesized using pressurized vapor/condenser (PVC) methods and were dispersed in aqueous solution using ultrasonication methods with the aid of ionic surfactants. Our nanomechanical measurements reveal the elastic deformational behaviors of individual BNNTs with two to four tube walls in their transverse directions. Their effective radial elastic moduli were obtained through interpreting their measured radial deformation profiles using Hertzian contact mechanics models. Our results capture the dependences of the effective radial moduli of MW-BNNTs on both the tube outer diameter and the number of tube layers. The effective radial moduli of double-walled BNNTs are found to be several-fold higher than those of single-walled BNNTs within the same diameter range. Our work contributes directly to a complete understanding of the fundamental structural and mechanical properties of BNNTs and the pursuits of their novel structural and electronics applications.

Michael W. Smith, Cheol Park, Meng Zheng, Changhong Ke ,In-Tae Bae, Kevin Jordan

2012-02-01T23:59:59.000Z

Note: This page contains sample records for the topic "walled pressure vessel" 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

Transpiring wall supercritical water oxidation reactor salt deposition studies  

SciTech Connect (OSTI)

Sandia National Laboratories has teamed with Foster Wheeler Development Corp. and GenCorp, Aerojet to develop and evaluate a new supercritical water oxidation reactor design using a transpiring wall liner. In the design, pure water is injected through small pores in the liner wall to form a protective boundary layer that inhibits salt deposition and corrosion, effects that interfere with system performance. The concept was tested at Sandia on a laboratory-scale transpiring wall reactor that is a 1/4 scale model of a prototype plant being designed for the Army to destroy colored smoke and dye at Pine Bluff Arsenal in Arkansas. During the tests, a single-phase pressurized solution of sodium sulfate (Na{sub 2}SO{sub 4}) was heated to supercritical conditions, causing the salt to precipitate out as a fine solid. On-line diagnostics and post-test observation allowed us to characterize reactor performance at different flow and temperature conditions. Tests with and without the protective boundary layer demonstrated that wall transpiration provides significant protection against salt deposition. Confirmation tests were run with one of the dyes that will be processed in the Pine Bluff facility. The experimental techniques, results, and conclusions are discussed.

Haroldsen, B.L.; Mills, B.E.; Ariizumi, D.Y.; Brown, B.G. [and others

1996-09-01T23:59:59.000Z

362

Steel-framed buildings: Impacts of wall detail configurations on the whole wall thermal performance  

SciTech Connect (OSTI)

The main objective of this paper is the influence of architectural wall details on the whole wall thermal performance. Whole wall thermal performance analysis was performed for six light gage steel-framed wall systems (some with wood components). For each wall system, all wall details were simulated using calibrated 3-D finite difference computer modeling. The thermal performance of the six steel-framed wall systems included various system details and the whole wall system thermal performance for a typical single-story ranch house. Currently, predicted heat losses through building walls are typically based on measurements of the wall system clear wall area using test methods such as ASTM C 236 or are calculated by one of the procedures recommended in the ASHRAE Handbook of Fundamentals that often is carried out for the clear wall area exclusively. In this paper, clear wall area is defined as the part of the wall system that is free of thermal anomalies due to building envelope details or thermally unaffected by intersections with other surfaces of the building envelope. Clear wall experiments or calculations normally do not include the effects of building envelope details such as corners, window and door openings, and structural intersections with roofs, floors, ceilings, and other walls. In steel-framed wall systems, these details typically consist of much more structural components than the clear wall. For this situation, the thermal properties measured or calculated for the clear wall area do not adequately represent the total wall system thermal performance. Factors that would impact the ability of today`s standard practice to accurately predict the total wall system thermal performance are the accuracy of the calculation methods, the area of the total wall that is clear wall, and the quantity and thermal performance of the various wall system details.

Kosny, J.; Desjarlais, A.O.; Christian, J.E.

1998-06-01T23:59:59.000Z

363

Method and apparatus for coupling seismic sensors to a borehole wall  

DOE Patents [OSTI]

A method and apparatus suitable for coupling seismic or other downhole sensors to a borehole wall in high temperature and pressure environments. In one embodiment, one or more metal bellows mounted to a sensor module are inflated to clamp the sensor module within the borehole and couple an associated seismic sensor to a borehole wall. Once the sensing operation is complete, the bellows are deflated and the sensor module is unclamped by deflation of the metal bellows. In a further embodiment, a magnetic drive pump in a pump module is used to supply fluid pressure for inflating the metal bellows using borehole fluid or fluid from a reservoir. The pump includes a magnetic drive motor configured with a rotor assembly to be exposed to borehole fluid pressure including a rotatable armature for driving an impeller and an associated coil under control of electronics isolated from borehole pressure.

West, Phillip B.

2005-03-15T23:59:59.000Z

364

Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor  

SciTech Connect (OSTI)

Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300{degrees}C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered.

Krishnan, S.; Bhasin, V.; Mahajan, S.C. [Bhabha Atomic Research Centre, Bombay (India)] [and others

1997-04-01T23:59:59.000Z

365

Comparison of high pressure transient PVT measurements and model predictions. Part I.  

SciTech Connect (OSTI)

A series of experiments consisting of vessel-to-vessel transfers of pressurized gas using Transient PVT methodology have been conducted to provide a data set for optimizing heat transfer correlations in high pressure flow systems. In rapid expansions such as these, the heat transfer conditions are neither adiabatic nor isothermal. Compressible flow tools exist, such as NETFLOW that can accurately calculate the pressure and other dynamical mechanical properties of such a system as a function of time. However to properly evaluate the mass that has transferred as a function of time these computational tools rely on heat transfer correlations that must be confirmed experimentally. In this work new data sets using helium gas are used to evaluate the accuracy of these correlations for receiver vessel sizes ranging from 0.090 L to 13 L and initial supply pressures ranging from 2 MPa to 40 MPa. The comparisons show that the correlations developed in the 1980s from sparse data sets perform well for the supply vessels but are not accurate for the receivers, particularly at early time during the transfers. This report focuses on the experiments used to obtain high quality data sets that can be used to validate computational models. Part II of this report discusses how these data were used to gain insight into the physics of gas transfer and to improve vessel heat transfer correlations. Network flow modeling and CFD modeling is also discussed.

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

2010-07-01T23:59:59.000Z

366

High Temperature Electrolysis Pressurized Experiment Design, Operation, and Results  

SciTech Connect (OSTI)

A new facility has been developed at the Idaho National Laboratory for pressurized testing of solid oxide electrolysis stacks. Pressurized operation is envisioned for large-scale hydrogen production plants, yielding higher overall efficiencies when the hydrogen product is to be delivered at elevated pressure for tank storage or pipelines. Pressurized operation also supports higher mass flow rates of the process gases with smaller components. The test stand can accommodate planar cells with dimensions up to 8.5 cm x 8.5 cm and stacks of up to 25 cells. It is also suitable for testing other cell and stack geometries including tubular cells. The pressure boundary for these tests is a water-cooled spool-piece pressure vessel designed for operation up to 5 MPa. Pressurized operation of a ten-cell internally manifolded solid oxide electrolysis stack has been successfully demonstrated up 1.5 MPa. The stack is internally manifolded and operates in cross-flow with an inverted-U flow pattern. Feed-throughs for gas inlets/outlets, power, and instrumentation are all located in the bottom flange. The entire spool piece, with the exception of the bottom flange, can be lifted to allow access to the internal furnace and test fixture. Lifting is accomplished with a motorized threaded drive mechanism attached to a rigid structural frame. Stack mechanical compression is accomplished using springs that are located inside of the pressure boundary, but outside of the hot zone. Initial stack heatup and performance characterization occurs at ambient pressure followed by lowering and sealing of the pressure vessel and subsequent pressurization. Pressure equalization between the anode and cathode sides of the cells and the stack surroundings is ensured by combining all of the process gases downstream of the stack. Steady pressure is maintained by means of a backpressure regulator and a digital pressure controller. A full description of the pressurized test apparatus is provided in this report. Results of initial testing showed the expected increase in open-cell voltage associated with elevated pressure. However, stack performance in terms of area-specific resistance was enhanced at elevated pressure due to better gas diffusion through the porous electrodes of the cells. Some issues such as cracked cells and seals were encountered during testing. Full resolution of these issues will require additional testing to identify the optimum test configurations and protocols.

J.E. O'Brien; X. Zhang; G.K. Housley; K. DeWall; L. Moore-McAteer

2012-09-01T23:59:59.000Z

367

Magnetic domain walls driven by interfacial phenomena  

E-Print Network [OSTI]

A domain wall in a ferromagnetic material is a boundary between differently magnetized regions, and its motion provides a convenient scheme to control the magnetization state of the material. Domain walls can be confined ...

Emori, Satoru

2014-01-01T23:59:59.000Z

368

Pressurizer with a mechanically attached surge nozzle thermal sleeve  

DOE Patents [OSTI]

A thermal sleeve is mechanically attached to the bore of a surge nozzle of a pressurizer for the primary circuit of a pressurized water reactor steam generating system. The thermal sleeve is attached with a series of keys and slots which maintain the thermal sleeve centered in the nozzle while permitting thermal growth and restricting flow between the sleeve and the interior wall of the nozzle.

Wepfer, Robert M

2014-03-25T23:59:59.000Z

369

E-Print Network 3.0 - a533b pressure vessel Sample Search Results  

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

SUPPORT STRUCTURE* , G. Ciovati, D. Machie, J. Pitts, J. Preble, W. J. Schneider, K. Smith, L. Turlington Summary: radio frequency (SRF) niobium cavities, which are housed in...

370

Design of pressure vessels using shape optimization: An integrated approach R.C. Carbonari a  

E-Print Network [OSTI]

, and focuses on CNG (Compressed Natural Gas) tank design by means of shape optimization techniques. This paper

Paulino, Glaucio H.

371

Modeling of Late Blooming Phases and Precipitation Kinetics in Aging Reactor Pressure Vessel (RPV) Steels  

SciTech Connect (OSTI)

The principle work at the atomic scale is to develop a predictive quantitative model for the microstructure evolution of RPV steels under thermal aging and neutron radiation. We have developed an AKMC method for the precipitation kinetics in bcc-Fe, with Cu, Ni, Mn and Si being the alloying elements. In addition, we used MD simulations to provide input parameters (if not available in literature). MMC simulations were also carried out to explore the possible segregation/precipitation morphologies at the lattice defects. First we briefly describe each of the simulation algorithms, then will present our results.

Yongfeng Zhang; Pritam Chakraborty; S. Bulent Biner

2013-09-01T23:59:59.000Z

372

Proceedings of PVP2007 2007 ASME Pressure Vessels and Piping Division Conference  

E-Print Network [OSTI]

.K. S. Kundu Materials Science and Metallurgy University of Cambridge, Pembroke Street, Cambridge CB2 3QZ, U.K. H. K. D. H. Bhadeshia Materials Science and Metallurgy University of Cambridge, Pembroke Street, Cambridge CB2 3QZ, U.K. H. J. Stone Materials Science and Metallurgy University of Cambridge

Cambridge, University of

373

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

SciTech Connect (OSTI)

The purpose of the Materials Aging and Degradation Pathway is to develop the scientific basis for understanding and predicting long-term environmental degradation behavior of materials in nuclear power plants and to provide data and methods to assess the performance of systems, structures, and components essential to safe and sustained nuclear power plant operations. The understanding of aging-related phenomena and their impacts on systems, structures, and components is expected to be a significant issue for any nuclear power plant planning for long-term operations (i.e., service beyond the initial license renewal period). Management of those phenomena and their impacts during long-term operations can be better enabled by improved methods and techniques for detection, monitoring, and prediction of systems, structures, and components degradation.

Clayton, Dwight A.; Bakhtiari, Sasan; Smith, Cyrus M.; Simmons, Kevin L.; Ramuhalli, Pradeep; Coble, Jamie B.; Brenchley, David L.; Meyer, Ryan M.

2013-04-16T23:59:59.000Z

374

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

E-Print Network [OSTI]

of electric vehicles to improve environmental quality and energy security, while providing the range of electricity required for liquefying the hydrogen [7]; the evaporation losses that may occur during fueling low, performance, and utility of today's gasoline vehicles. Probably the most significant hurdle for hydrogen

375

Rigorous Simulation of Accidental Leaks from High-Pressure Storage Vessels  

E-Print Network [OSTI]

of nature. The released chemical can form and disperse as vapor cloud leading to fire, explosion, or toxic exposure. The resulting leak could be single phase or multiphase release, choked or non-choked. These releases could result in liquid spills, vapor...

Alisha, -

2014-07-07T23:59:59.000Z

376

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

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankCombustion | Department ofT ib l L d F SSalesOE0000652GrowE-mail onThe2 DOE Hydrogen andProgram In

377

File:06HIGBoilerPressureVesselPermit.pdf | Open Energy Information  

Open Energy Info (EERE)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home5b9fcbce19 NoPublic UtilitiesCABConstructionStormWaterProgram.pdf Jump to: navigation, search FileHIDOtherOverview.pdf

378

Wall Drying in Hot and Humid Climates  

E-Print Network [OSTI]

drying potential while at the same time providing a high potential for mold growth. To reduce moisture accumulation in wall systems, it is important to design wall systems that not only reduce moisture intrusion, but also allow drying. Yet often a wall...

Boone, K.; Weston, T.; Pascual, X.

2004-01-01T23:59:59.000Z

379

Method for reducing pressure drop through filters, and filter exhibiting reduced pressure drop  

DOE Patents [OSTI]

Methods for generating and applying coatings to filters with porous material in order to reduce large pressure drop increases as material accumulates in a filter, as well as the filter exhibiting reduced and/or more uniform pressure drop. The filter can be a diesel particulate trap for removing particulate matter such as soot from the exhaust of a diesel engine. Porous material such as ash is loaded on the surface of the substrate or filter walls, such as by coating, depositing, distributing or layering the porous material along the channel walls of the filter in an amount effective for minimizing or preventing depth filtration during use of the filter. Efficient filtration at acceptable flow rates is achieved.

Sappok, Alexander; Wong, Victor

2014-11-18T23:59:59.000Z

380

Nondestructive Technique Survey for Assessing Integrity of Composite Firing Vessel  

SciTech Connect (OSTI)

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

Tran, A.

2000-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "walled pressure vessel" 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

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

382

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

383

Processing and analysis techniques involving in-vessel material generation  

DOE Patents [OSTI]

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

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

2012-09-25T23:59:59.000Z

384

Processing and analysis techniques involving in-vessel material generation  

DOE Patents [OSTI]

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

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

2011-01-25T23:59:59.000Z

385

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

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

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

386

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

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

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

387

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

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

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

388

Gas-lubricated seal for sealing between a piston and a cylinder wall  

DOE Patents [OSTI]

A piston-cylinder seal uses gas for a lubricant and has a runner supported on a gapless structure and placed in the space between the piston and the cylinder wall. The runner is deformed elastically under the influence of the operating pressures to follow and compensate for variations in the piston-cylinder fit and maintain a seal. 4 figs.

Hoult, D.P.

1985-09-10T23:59:59.000Z

389

Gas-lubricated seal for sealing between a piston and a cylinder wall  

DOE Patents [OSTI]

A piston-cylinder seal uses gas for a lubricant and has a runner supported on a gapless structure and placed in the space between the piston and the cylinder wall. The runner is deformed elastically under the influence of the operating pressures to follow and compensate for variations in the piston-cylinder fit and maintain a seal.

Hoult, David P. (Box 89, Wellesley, MA 02181)

1985-01-01T23:59:59.000Z

390

Pressure Relief Devices for High-Pressure Gaseous Storage Systems: Applicability to Hydrogen Technology  

SciTech Connect (OSTI)

Pressure relief devices (PRDs) are viewed as essential safety measures for high-pressure gas storage and distribution systems. These devices are used to prevent the over-pressurization of gas storage vessels and distribution equipment, except in the application of certain toxic gases. PRDs play a critical role in the implementation of most high-pressure gas storage systems and anyone working with these devices should understand their function so they can be designed, installed, and maintained properly to prevent any potentially dangerous or fatal incidents. As such, the intention of this report is to introduce the reader to the function of the common types of PRDs currently used in industry. Since high-pressure hydrogen gas storage systems are being developed to support the growing hydrogen energy infrastructure, several recent failure incidents, specifically involving hydrogen, will be examined to demonstrate the results and possible mechanisms of a device failure. The applicable codes and standards, developed to minimize the risk of failure for PRDs, will also be reviewed. Finally, because PRDs are a critical component for the development of a successful hydrogen energy infrastructure, important considerations for pressure relief devices applied in a hydrogen gas environment will be explored.

Kostival, A.; Rivkin, C.; Buttner, W.; Burgess, R.

2013-11-01T23:59:59.000Z

391

Effect of temperature and pressure on the dynamics of nanoconfined propane  

SciTech Connect (OSTI)

We report the effect of temperature and pressure on the dynamical properties of propane confined in nanoporous silica aerogel studied using quasielastic neutron scattering (QENS). Our results demonstrate that the effect of a change in the pressure dominates over the effect of temperature variation on the dynamics of propane nano-confined in silica aerogel. At low pressures, most of the propane molecules are strongly bound to the pore walls, only a small fraction is mobile. As the pressure is increased, the fraction of mobile molecules increases. A change in the mechanism of motion, from continuous diffusion at low pressures to jump diffusion at higher pressures has also been observed.

Gautam, Siddharth, E-mail: gautam.25@osu.edu; Liu, Tingting, E-mail: gautam.25@osu.edu; Welch, Susan; Cole, David [School of Earth Sciences, The Ohio State University, 275 Mendenhall Laboratory, 125 S Oval Mall, Columbus, OH 43210 (United States); Rother, Gernot [Geochemistry and Interfacial Science Group, Chemical Science Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Jalarvo, Niina [Jülich Center for Neutron Sciences (JCNS-1), Forschungszentrum Jülich Outstation at Spallation Neutron Source(SNS), Chemical and Engineering Materials Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Mamontov, Eugene [Spallation Neutron Source (SNS), Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

2014-04-24T23:59:59.000Z

392

Factors affecting coking pressures in tall coke ovens  

SciTech Connect (OSTI)

The detrimental effects of excessive coking pressures, resulting in the permanent deformation of coke oven walls, have been recognized for many years. Considerable research has been undertaken worldwide in attempts to define the limits within which a plant may safely operate and to quantify the factors which influence these pressures. Few full scale techniques are available for assessing the potential of a coal blend for causing wall damage. Inference of dangerous swelling pressures may be made however by the measurement of the peak gas pressure which is generated as the plastic layers meet and coalesce at the center of the oven. This pressure is referred to in this report as the carbonizing pressure. At the Dawes Lane cokemaking plant of British Steel`s Scunthorpe Works, a large database has been compiled over several years from the regulator measurement of this pressure. This data has been statistically analyzed to provide a mathematical model for predicting the carbonizing pressure from the properties of the component coals, the results of this analysis are presented in this report.

Grimley, J.J.; Radley, C.E. [British Steel plc, Scunthorpe (United Kingdom). Scunthorpe Works

1995-12-01T23:59:59.000Z

393

Hydrodynamic evaluation of high-speed semi-SWATH vessels  

E-Print Network [OSTI]

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

Guttenplan, Adam (Adam David)

2007-01-01T23:59:59.000Z

394

aging blood vessels: Topics by E-print Network  

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

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

395

Vessel Segmentation with Automatic Centerline Extraction Using Tubular Tree Segmentation  

E-Print Network [OSTI]

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

Paris-Sud XI, Université de

396

Toroid cavities as NMR detectors in high pressure probes  

SciTech Connect (OSTI)

A cylindrical toroid cavity has been developed for application as an NMR detector for high sensitivity and high resolution spectroscopy in metal vessel probes. Those probes are used for in situ investigations at high temperature and pressure. Since the transmitted r.f. field is completely confined within the torus, the cavity can be placed inside the pressurized system without magnetic coupling to the metal vessel. Resonance frequencies up to 400 MHz make the toroid cavity detector especially suited for use in {sup 1}H and {sup 19}F spectroscopy. Typically achieved static {sup 1}H linewidths, measured on CHCl{sub 3} using cavities in Be-Cu pressure vessels, are 2.0 Hz. On the basis of theoretical considerations that include the radial dependence of the r.f. field within cylindrical or circular toroid detectors, equations were evolved to predict the signal intensity as a function of the pulse width. The equations precisely describe the deviations from the sinusoidal approximation, which is generally used for signal intensities derived from Helmholtz or solenoid coils.

Woelk, K.; Rathke, J.W.; Klingler, R.J.

1993-03-01T23:59:59.000Z

397

Resistive wall tearing mode generated finite net electromagnetic torque in a static plasma  

SciTech Connect (OSTI)

The MARS-F code [Y. Q. Liu et al., Phys. Plasmas 7, 3681 (2000)] is applied to numerically investigate the effect of the plasma pressure on the tearing mode stability as well as the tearing mode-induced electromagnetic torque, in the presence of a resistive wall. The tearing mode with a complex eigenvalue, resulted from the favorable averaged curvature effect [A. H. Glasser et al., Phys. Fluids 18, 875 (1975)], leads to a re-distribution of the electromagnetic torque with multiple peaking in the immediate vicinity of the resistive layer. The multiple peaking is often caused by the sound wave resonances. In the presence of a resistive wall surrounding the plasma, a rotating tearing mode can generate a finite net electromagnetic torque acting on the static plasma column. Meanwhile, an equal but opposite torque is generated in the resistive wall, thus conserving the total momentum of the whole plasma-wall system. The direction of the net torque on the plasma is always opposite to the real frequency of the mode, agreeing with the analytic result by Pustovitov [Nucl. Fusion 47, 1583 (2007)]. When the wall time is close to the oscillating time of the tearing mode, the finite net torque reaches its maximum. Without wall or with an ideal wall, no net torque on the static plasma is generated by the tearing mode. However, re-distribution of the torque density in the resistive layer still occurs.

Hao, G. Z., E-mail: haogz@swip.ac.cn; Wang, A. K.; Xu, M.; Qu, H. P.; Peng, X. D.; Wang, Z. H.; Xu, J. Q.; Qiu, X. M. [Southwestern Institute of Physics, Post Office Box 432, Chengdu 610041 (China)] [Southwestern Institute of Physics, Post Office Box 432, Chengdu 610041 (China); Liu, Y. Q. [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)] [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)

2014-01-15T23:59:59.000Z

398

A cog-like vessel from the Netherlands  

E-Print Network [OSTI]

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

Van de Moortel, Aleydis Maria P. A.

1987-01-01T23:59:59.000Z

399

Pressurized solid oxide fuel cell integral air accumular containment  

DOE Patents [OSTI]

A fuel cell generator apparatus contains at least one fuel cell subassembly module in a module housing, where the housing is surrounded by a pressure vessel such that there is an air accumulator space, where the apparatus is associated with an air compressor of a turbine/generator/air compressor system, where pressurized air from the compressor passes into the space and occupies the space and then flows to the fuel cells in the subassembly module, where the air accumulation space provides an accumulator to control any unreacted fuel gas that might flow from the module.

Gillett, James E.; Zafred, Paolo R.; Basel, Richard A.

2004-02-10T23:59:59.000Z

400

High-pressure microhydraulic actuator  

DOE Patents [OSTI]

Electrokinetic ("EK") pumps convert electric to mechanical work when an electric field exerts a body force on ions in the Debye layer of a fluid in a packed bed, which then viscously drags the fluid. Porous silica and polymer monoliths (2.5-mm O.D., and 6-mm to 10-mm length) having a narrow pore size distribution have been developed that are capable of large pressure gradients (250-500 psi/mm) when large electric fields (1000-1500 V/cm) are applied. Flowrates up to 200 .mu.L/min and delivery pressures up to 1200 psi have been demonstrated. Forces up to 5 lb-force at 0.5 mm/s (12 mW) have been demonstrated with a battery-powered DC-DC converter. Hydraulic power of 17 mW (900 psi@ 180 uL/min) has been demonstrated with wall-powered high voltage supplies. The force and stroke delivered by an actuator utilizing an EK pump are shown to exceed the output of solenoids, stepper motors, and DC motors of similar size, despite the low thermodynamic efficiency.

Mosier, Bruce P. (San Francisco, CA) [San Francisco, CA; Crocker, Robert W. (Fremont, CA) [Fremont, CA; Patel, Kamlesh D. (Dublin, CA) [Dublin, CA

2008-06-10T23:59:59.000Z

Note: This page contains sample records for the topic "walled pressure vessel" 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

PressurePressure Indiana Coal Characteristics  

E-Print Network [OSTI]

TimeTime PressurePressure · Indiana Coal Characteristics · Indiana Coals for Coke · Coal Indiana Total Consumption Electricity 59,664 Coke 4,716 Industrial 3,493 Major Coal- red power plantsTransportation in Indiana · Coal Slurry Ponds Evaluation · Site Selection for Coal Gasification · Coal-To-Liquids Study, CTL

Fernández-Juricic, Esteban

402

POROUS WALL, HOLLOW GLASS MICROSPHERES  

SciTech Connect (OSTI)

Hollow Glass Microspheres (HGM) is not a new technology. All one has to do is go to the internet and Google{trademark} HGM. Anyone can buy HGM and they have a wide variety of uses. HGM are usually between 1 to 100 microns in diameter, although their size can range from 100 nanometers to 5 millimeters in diameter. HGM are used as lightweight filler in composite materials such as syntactic foam and lightweight concrete. In 1968 a patent was issued to W. Beck of the 3M{trademark} Company for 'Glass Bubbles Prepared by Reheating Solid Glass Particles'. In 1983 P. Howell was issued a patent for 'Glass Bubbles of Increased Collapse Strength' and in 1988 H. Marshall was issued a patent for 'Glass Microbubbles'. Now Google{trademark}, Porous Wall, Hollow Glass Microspheres (PW-HGMs), the key words here are Porous Wall. Almost every article has its beginning with the research done at the Savannah River National Laboratory (SRNL). The Savannah River Site (SRS) where SRNL is located has a long and successful history of working with hydrogen and its isotopes for national security, energy, waste management and environmental remediation applications. This includes more than 30 years of experience developing, processing, and implementing special ceramics, including glasses for a variety of Department of Energy (DOE) missions. In the case of glasses, SRS and SRNL have been involved in both the science and engineering of vitreous or glass based systems. As a part of this glass experience and expertise, SRNL has developed a number of niches in the glass arena, one of which is the development of porous glass systems for a variety of applications. These porous glass systems include sol gel glasses, which include both xerogels and aerogels, as well as phase separated glass compositions, that can be subsequently treated to produce another unique type of porosity within the glass forms. The porous glasses can increase the surface area compared to 'normal glasses of a 1 to 2 order of magnitude, which can result in unique properties in areas such as hydrogen storage, gas transport, gas separations and purifications, sensors, global warming applications, new drug delivery systems and so on. One of the most interesting porous glass products that SRNL has developed and patented is Porous Wall, Hollow Glass Microspheres (PW-HGMs) that are being studied for many different applications. The European Patent Office (EPO) just recently notified SRS that the continuation-in-part patent application for the PW-HGMs has been accepted. The original patent, which was granted by the EPO on June 2, 2010, was validated in France, Germany and the United Kingdom. The microspheres produced are generally in the range of 2 to 100 microns, with a 1 to 2 micron wall. What makes the SRNL microspheres unique from all others is that the team in Figure 1 has found a way to induce and control porosity through the thin walls on a scale of 100 to 3000 {angstrom}. This is what makes the SRNL HW-HGMs one-of-a-kind, and is responsible for many of their unique properties and potential for various applications, including those in tritium storage, gas separations, H-storage for vehicles, and even a variety of new medical applications in the areas of drug delivery and MRI contrast agents. SRNL Hollow Glass Microspheres, and subsequent, Porous Wall, Hollow Glass Microspheres are fabricated using a flame former apparatus. Figure 2 is a schematic of the apparatus.

Sexton, W.

2012-06-30T23:59:59.000Z

403

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

SciTech Connect (OSTI)

Purpose: To investigate vascular responses during fractionated radiation therapy (F-RT) and the effects of targeting pericytes or bone marrow-derived cells (BMDCs) on the efficacy of F-RT. Methods and Materials: Murine prostate TRAMP-C1 tumors were grown in control mice or mice transplanted with green fluorescent protein-tagged bone marrow (GFP-BM), and irradiated with 60 Gy in 15 fractions. Mice were also treated with gefitinib (an epidermal growth factor receptor inhibitor) or AMD3100 (a CXCR4 antagonist) to examine the effects of combination treatment. The responses of tumor vasculatures to these treatments and changes of tumor microenvironment were assessed. Results: After F-RT, the tumor microvascular density (MVD) was reduced; however, the surviving vessels were dilated, incorporated with GFP-positive cells, tightly adhered to pericytes, and well perfused with Hoechst 33342, suggesting a more mature structure formed primarily via vasculogenesis. Although the gefitinib+F-RT combination affected the vascular structure by dissociating pericytes from the vascular wall, it did not further delay tumor growth. These tumors had higher MVD and better vascular perfusion function, leading to less hypoxia and tumor necrosis. By contrast, the AMD3100+F-RT combination significantly enhanced tumor growth delay more than F-RT alone, and these tumors had lower MVD and poorer vascular perfusion function, resulting in increased hypoxia. These tumor vessels were rarely covered by pericytes and free of GFP-positive cells. Conclusions: Vasculogenesis is a major mechanism for tumor vessel survival during F-RT. Complex interactions occur between vessel-targeting agents and F-RT, and a synergistic effect may not always exist. To enhance F-RT, using CXCR4 inhibitor to block BM cell influx and the vasculogenesis process is a better strategy than targeting pericytes by epidermal growth factor receptor inhibitor.

Chen, Fang-Hsin; Fu, Sheng-Yung [Department of Biomedical Engineering and Environmental Sciences, National Tsing Hua University, Taiwan (China)] [Department of Biomedical Engineering and Environmental Sciences, National Tsing Hua University, Taiwan (China); Yang, Ying-Chieh [Department of Radiation Oncology, National Taiwan University Hospital Hsin-Chu Branch, Taiwan (China)] [Department of Radiation Oncology, National Taiwan University Hospital Hsin-Chu Branch, Taiwan (China); Wang, Chun-Chieh [Department of Radiation Oncology, Chang Gung Memorial Hospital-LinKou, Taiwan (China) [Department of Radiation Oncology, Chang Gung Memorial Hospital-LinKou, Taiwan (China); Department of Medical Imaging and Radiological Science, Chang Gung University, Taiwan (China); Chiang, Chi-Shiun, E-mail: cschiang@mx.nthu.edu.tw [Department of Biomedical Engineering and Environmental Sciences, National Tsing Hua University, Taiwan (China)] [Department of Biomedical Engineering and Environmental Sciences, National Tsing Hua University, Taiwan (China); Hong, Ji-Hong, E-mail: jihong@adm.cgmh.org.tw [Department of Radiation Oncology, Chang Gung Memorial Hospital-LinKou, Taiwan (China) [Department of Radiation Oncology, Chang Gung Memorial Hospital-LinKou, Taiwan (China); Department of Medical Imaging and Radiological Science, Chang Gung University, Taiwan (China)

2013-07-15T23:59:59.000Z

404

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

SciTech Connect (OSTI)

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

Joy Rempe; D. Knudson

2004-12-01T23:59:59.000Z

405

Analysis of Rotating Collectors from the Private Region of JET with Carbon Wall and Metallic ITER-Like Wall  

E-Print Network [OSTI]

Analysis of Rotating Collectors from the Private Region of JET with Carbon Wall and Metallic ITER-Like Wall

406

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

E-Print Network [OSTI]

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

American Society for Testing and Materials. Philadelphia

2010-01-01T23:59:59.000Z

407

SRNL POROUS WALL GLASS MICROSPHERES  

SciTech Connect (OSTI)

The Savannah River National Laboratory (SRNL) has developed a new medium for storage of hydrogen and other gases. This involves fabrication of thin, Porous Walled, Hollow Glass Microspheres (PW-HGMs), with diameters generally in the range of 1 to several hundred microns. What is unique about the glass microballons is that porosity has been induced and controlled within the thin, one micron thick walls, on the scale of 10 to several thousand Angstroms. This porosity results in interesting properties including the ability to use these channels to fill the microballons with special absorbents and other materials, thus providing a contained environment even for reactive species. Gases can now enter the microspheres and be retained on the absorbents, resulting in solid-state and contained storage of even reactive species. Also, the porosity can be altered and controlled in various ways, and even used to filter mixed gas streams within a system. SRNL is involved in about a half dozen different programs involving these PW-HGMs and an overview of some of these activities and results emerging are presented.

Wicks, G; Leung Heung, L; Ray Schumacher, R

2008-04-15T23:59:59.000Z

408

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

SciTech Connect (OSTI)

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

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

1984-01-01T23:59:59.000Z

409

Wall current probe: A non-invasive in situ plasma diagnostic for space and time resolved current density distribution measurement  

SciTech Connect (OSTI)

In the context of low temperature plasma research, we propose a wall current probe to determine the local charged particle fluxes flowing to the chamber walls. This non-intrusive planar probe consists of an array of electrode elements which can be individually biased and for which the current can be measured separately. We detail the probe properties and present the ability of the diagnostic to be used as a space and time resolved measurement of the ion and electron current density at the chamber walls. This diagnostic will be relevant to study the electron transport in magnetized low-pressure plasmas.

Baude, R.; Gaboriau, F.; Hagelaar, G. J. M. [Université de Toulouse, UPS, INPT, LAPLACE (Laboratoire Plasma et Conversion d’énergie), 118 route de Narbonne, F-31062 Toulouse Cedex 9, France and CNRS, LAPLACE, F-31062, Toulouse (France)] [Université de Toulouse, UPS, INPT, LAPLACE (Laboratoire Plasma et Conversion d’énergie), 118 route de Narbonne, F-31062 Toulouse Cedex 9, France and CNRS, LAPLACE, F-31062, Toulouse (France)

2013-08-15T23:59:59.000Z

410

Design and fabrication of a MEMS-array pressure sensor system for passive underwater navigation inspired by the lateral line  

E-Print Network [OSTI]

An object within a fluid flow generates local pressure variations that are unique and characteristic to the object's shape and size. For example, a three-dimensional object or a wall-like obstacle obstructs flow and creates ...

Hou, Stephen Ming-Chang, 1981-

2012-01-01T23:59:59.000Z

411

Performance limits of fusion first-wall structural materials.  

SciTech Connect (OSTI)

Key features of fusion energy relate primarily to potential advantages associated with safety and environmental considerations and the near endless supply of fuel. However, it is generally concluded that high performance fusion power systems will be required in order to be economically competitive with other energy options. As in most energy systems, structural materials operating limits pose a primary constraint to the performance of fusion power systems. It is also recognized that for the case of fusion power, the first-wall/blanket system will have a dominant impact on both the economic and safety/environmental attractiveness of fusion energy. The first-wall blanket structure is particularly critical since it must maintain high integrity at relatively high temperatures during exposure to high radiation levels, high surface heat fluxes, and significant primary stresses. The performance limits of the first-wall/blanket structure will be dependent on the structural material properties, the coolant/breeder system, and the specific design configuration. Key factors associated with high performance structural materials include (1) high temperature operation, (2) a large operating temperature window, and (3) a long operating lifetime. High temperature operation is necessary to provide for high power conversion efficiency. As discussed later, low-pressure coolant systems provide significant advantages. A large operating temperature window is necessary to accommodate high surface heating and high power density. The operating temperature range for the structure must include the temperature gradient through the first wall and the coolant system AT required for efficient energy conversion. This later requirement is dependent on the coolant/breeder operating temperature limits. A long operating lifetime of the structure is important to improve system availability and to minimize waste disposition.

Smith, D. L.; Majumdar, S.; Billone, M.; Mattas, R. F.

1999-11-12T23:59:59.000Z

412

A Helical Coolant Channel Design for the Solid Wall Blanket  

SciTech Connect (OSTI)

A helical coolant channel scheme is proposed for the APEX solid wall blanket module. The self-coolant breeder in this system is FLIBE (LiF)2(BeF2). The structural material is the nanocomposited alloy 12YWT. The neutron multiplier used in the current design is either stationary or slow moving liquid lead. The purpose of this study is to design a blanket that can handle a high wall loading (5 MW/m{sup 2}). In the mean time the design provides means to attain the maximum possible blanket outlet temperature and meet all engineering limits on temperature of structural material and liquids. An important issue for such a design is to optimize the system for minimum pressure loss. For advanced ferritic steel (12YWT) an upper temperature limit of 800 deg. C is expected, and a limit of 700 deg. C at the steel/FLIBE interface is recommended.The blanket module is composed of two main continuous routes. The first route is three helical rectangular channels side-by-side that surround a central box. The helical channels are fed from the bottom and exit at the top to feed the central channels in the central box. The coolant helical channels have a cross sectional area with a length of about 10 cm and a width that changes according to the position around the central box. For instance: the width of the coolant channels facing the plasma is the narrowest while it is the widest in the back (farthest from the plasma).In this design the coolant runs around the central box for only 5 turns to cover the total height of the first wall (6.8 m). The design is optimized with the FW channel width as a parameter with the heat transfer requirements at the first wall as the constraints.

Mogahed, E.A. [University of Wisconsin-Madison (United States)

2003-07-15T23:59:59.000Z

413

Textural break foundation wall construction modules  

DOE Patents [OSTI]

Below-grade, textural-break foundation wall structures are provided for inhibiting diffusion and advection of liquids and gases into and out from a surrounding hydrogeologic environment. The foundation wall structure includes a foundation wall having an interior and exterior surface and a porous medium disposed around a portion of the exterior surface. The structure further includes a modular barrier disposed around a portion of the porous medium. The modular barrier is substantially removable from the hydrogeologic environment.

Phillips, Steven J. (Kennewick, WA)

1990-01-01T23:59:59.000Z

414

Panelized wall system with foam core insulation  

DOE Patents [OSTI]

A wall system includes a plurality of wall members, the wall members having a first metal panel, a second metal panel, and an insulating core between the first panel and the second panel. At least one of the first panel and the second panel include ridge portions. The insulating core can be a foam, such as a polyurethane foam. The foam can include at least one opacifier to improve the k-factor of the foam.

Kosny, Jan (Oak Ridge, TN); Gaskin, Sally (Houston, TX)

2009-10-20T23:59:59.000Z

415

First wall for polarized fusion reactors  

DOE Patents [OSTI]

Depolarization mechanisms arising from the recycling of the polarized fuel at the limiter and the first-wall of a fusion reactor are greater than those mechanisms in the plasma. Rapid depolarization of the plasma is prevented by providing a first-wall or first-wall coating formed of a low-Z, non-metallic material having a depolarization rate greater than 1 sec.sup.-1.

Greenside, Henry S. (Cranbury, NJ); Budny, Robert V. (Princeton, NJ); Post, Jr., Douglass E. (Buttonwood, CT)

1988-01-01T23:59:59.000Z

416

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

SciTech Connect (OSTI)

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

Smith, A.

2010-02-16T23:59:59.000Z

417

Liquid Walls Innovative High Power Density Concepts  

E-Print Network [OSTI]

-CLIFF 3.High-Temperature Refractory Solid Wall -EVOLVE (Two-Phase Lithium Flow) -Helium Cooling erosion as limiting factors -Results in smaller and lower cost components (chamb

California at Los Angeles, University of

418

First wall for polarized fusion reactors  

DOE Patents [OSTI]

A first-wall or first-wall coating for use in a fusion reactor having polarized fuel may be formed of a low-Z non-metallic material having slow spin relaxation, i.e., a depolarization rate greater than 1 sec/sup -1/. Materials having these properties include hydrogenated and deuterated amorphous semiconductors. A method for preventing the rapid depolarization of a polarized plasma in a fusion device may comprise the step of providing a first-wall or first-wall coating formed of a low-Z, non-metallic material having a depolarization rate greater than 1 sec/sup -1/.

Greenside, H.S.; Budny, R.V.; Post, D.E. Jr.

1985-01-29T23:59:59.000Z

419

Multiple moving wall dry coal extrusion pump  

DOE Patents [OSTI]

A pump for transporting particulate material includes a passageway defined on each side between an inlet and an outlet by a moving wall.

Fitzsimmons, Mark Andrew

2013-05-14T23:59:59.000Z

420

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

SciTech Connect (OSTI)

The potential for stainless steel cladding to improve the fracture behavior of an operating nuclear reactor pressure vessel, particularly during certain overcooling transients, may depend greatly on the properties of the irradiated cladding. Therefore, weld overlay cladding irradiated at temperatures and to fluences relevant to power reactor operation was examined. The cladding was applied to a pressure vessel steel plate by the three-wire series-arc commercial method. Cladding was applied in three layers to provide adequate thickness for the fabrication of test specimens. The three-wire series-arc procedure, developed by Combustion Engineering, Inc., Chattanooga, Tennessee, produced a highly controlled weld chemistry, microstructure, and fracture properties in all three layers of the weld. Charpy V-notch and tensile specimens were irradiated at 288/sup 0/C to fluence levels of 2 and 5 x 10/sup 19/ neutrons/cm/sup 2/ (>1 MeV). Postirradiation testing results show that, in the test temperature range from -125 to 288/sup 0/C, the yield strength increased by 8 to 30%, ductility insignificantly increased, while there was almost no change in ultimate tensile strength. All cladding exhibited ductile-to-brittle transition behavior during Charpy impact testing, due to the dominance of delta-ferrite failures at low temperatures. On the upper shelf, energy was reduced, due to irradiation exposure, 15 and 20%, while the lateral expansion was reduced 43 and 41%, at 2 and 5 x 10/sup 19/ neutrons/cm/sup 2/ (>1 MeV), respectively. In addition, radiation damage resulted in 13 and 28/sup 0/C shifts of the Charpy impact transition temperature at the 41-J level for the low and high fluences, respectively.

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

1987-01-01T23:59:59.000Z

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


421

Virtual gap dielectric wall accelerator  

DOE Patents [OSTI]

A virtual, moving accelerating gap is formed along an insulating tube in a dielectric wall accelerator (DWA) by locally controlling the conductivity of the tube. Localized voltage concentration is thus achieved by sequential activation of a variable resistive tube or stalk down the axis of an inductive voltage adder, producing a "virtual" traveling wave along the tube. The tube conductivity can be controlled at a desired location, which can be moved at a desired rate, by light illumination, or by photoconductive switches, or by other means. As a result, an impressed voltage along the tube appears predominantly over a local region, the virtual gap. By making the length of the tube large in comparison to the virtual gap length, the effective gain of the accelerator can be made very large.

Caporaso, George James; Chen, Yu-Jiuan; Nelson, Scott; Sullivan, Jim; Hawkins, Steven A

2013-11-05T23:59:59.000Z

422

Beetle Kill Wall at NREL  

ScienceCinema (OSTI)

When it comes to designing an interior decorative feature for one of the most energy efficient office buildings in the world, very few would consider bringing in a beetle to do the job. But thats what happened at the U.S. Department of Energy's (DOE) Research Support Facility (RSF) located on the National Renewable Energy Laboratory (NREL) campus.In June, the RSF will become home to more than 800 workers from DOE and NREL and building visitors will be greeted with a soaring, two-story high wall entirely covered with wood harvested from the bark beetle infestation that has killed millions of pine trees in the Western U.S. But, the use of beetle kill wood is just one example of the resources being leveraged to make the RSF a model for sustainability and one more step toward NRELs goal to be a net zero energy campus.

None

2013-05-29T23:59:59.000Z

423

Pressure reducing regulator  

DOE Patents [OSTI]

A pressure reducing regulator that controls its downstream or outlet pressure to a fixed fraction of its upstream or inlet pressure. The regulator includes a housing which may be of a titanium alloy, within which is located a seal or gasket at the outlet end which may be made of annealed copper, a rod, and piston, each of which may be made of high density graphite. The regulator is insensitive to temperature by virtue of being without a spring or gas sealed behind a diaphragm, and provides a reference for a system in which it is being used. The rod and piston of the regulator are constructed, for example, to have a 1/20 ratio such that when the downstream pressure is less than 1/20 of the upstream pressure the regulator opens and when the downstream pressure exceeds 1/20 of the upstream pressure the regulator closes.

Whitehead, John C. (Davis, CA); Dilgard, Lemoyne W. (Willits, CA)

1995-01-01T23:59:59.000Z

424

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

425

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

426

Pressurized Testing of Solid Oxide Electrolysis Stacks with Advanced Electrode-Supported Cells  

SciTech Connect (OSTI)

A new facility has been developed at the Idaho National Laboratory for pressurized testing of solid oxide electrolysis stacks. Pressurized operation is envisioned for large-scale hydrogen production plants, yielding higher overall efficiencies when the hydrogen product is to be delivered at elevated pressure for tank storage or pipelines. Pressurized operation also supports higher mass flow rates of the process gases with smaller components. The test stand can accommodate cell dimensions up to 8.5 cm x 8.5 cm and stacks of up to 25 cells. The pressure boundary for these tests is a water-cooled spool-piece pressure vessel designed for operation up to 5 MPa. The stack is internally manifolded and operates in cross-flow with an inverted-U flow pattern. Feed-throughs for gas inlets/outlets, power, and instrumentation are all located in the bottom flange. The entire spool piece, with the exception of the bottom flange, can be lifted to allow access to the internal furnace and test fixture. Lifting is accomplished with a motorized threaded drive mechanism attached to a rigid structural frame. Stack mechanical compression is accomplished using springs that are located inside of the pressure boundary, but outside of the hot zone. Initial stack heatup and performance characterization occurs at ambient pressure followed by lowering and sealing of the pressure vessel and subsequent pressurization. Pressure equalization between the anode and cathode sides of the cells and the stack surroundings is ensured by combining all of the process gases downstream of the stack. Steady pressure is maintained by means of a backpressure regulator and a digital pressure controller. A full description of the pressurized test apparatus is provided in this paper.

J. E. O'Brien; X. Zhang; G. K. Housley; K. DeWall; L. Moore-McAteer; G. Tao

2012-06-01T23:59:59.000Z

427

A new design criterion based on pressure testing of torispherical heads  

SciTech Connect (OSTI)

Two vessels with torispherical heads were pressurized to destruction at the Praxair Tonawanda facility on September 12--13, 1994. The objective was to determine pressures at which observable or measurable indications of failure could be detected. Plastic limit pressures for the two heads were calculated at 190 and 240 psi, respectively. For Vessel 1, the only observable action was a slow formation of some waviness of the knuckle profile at approximately 600 psi. It lost pressure at 700 psi when a crack developed at a nozzle weld at the bottom of the shell. For Vessel 2, no indication of any sign of failure was observed until it burst at a pressure of 1,080 psi by a ductile fracture along the longitudinal weld of the shell. The main conclusion is that there is a problem in the application of the double elastic slope collapse criterion to torispherical heads. It was determined that when using this criterion a collapse pressure signaling excessive deformation cannot be determined with any certainty. Furthermore, the test data do not show anything at any of the calculated collapse pressures that suggests excessive deformation. Thus, the collapse pressures for torispherical heads cannot be confirmed by test. This leads to the inconsistency that if the collapse load is divided by a safety factor, say 1.5, to obtain an allowable pressure, the actual safety margin of the design is not known and may not be 1.5. For a material with sufficient ductility, the use of an estimated burst pressure appears preferable. A design criterion based on the membrane stress at the crown of a torispherical head reaching the ultimate tensile strength is proposed, which is simple, can be supported by theoretical arguments, and is shown to be conservative by current test results as well as by those of two previous test programs.

Kalnins, A. [Lehigh Univ., Bethlehem, PA (United States). Dept. of Mechanical Engineering and Mechanics; Rana, M.D. [Praxair, Inc., Tonawanda, NY (United States). Research and Development Dept.

1996-08-01T23:59:59.000Z

428

Ex-vessel demand by size for the Gulf shrimp  

E-Print Network [OSTI]

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

Chui, Margaret Kam-Too

1980-01-01T23:59:59.000Z

429

Remarks on Liquid Wall Research Mohamed Abdou  

E-Print Network [OSTI]

Remarks on Liquid Wall Research Mohamed Abdou Professor Mechanical and Aerospace Engineering UCLA physicists and engineering scientists · Enhances synergism between IFE and MFE · Provides excellent disciplines. #12;Several "Ideas" Have Been Proposed for Liquid Walls Fluids 1) High-conductivity, low Pr

California at Los Angeles, University of

430

External Insulation of Masonry Walls and Wood Framed Walls  

SciTech Connect (OSTI)

The use of exterior insulation on a building is an accepted and effective means to increase the overall thermal resistance of the assembly that also has other advantages of improved water management and often increased air tightness of building assemblies. For thin layers of insulation (1" to 1 1/2"), the cladding can typically be attached directly through the insulation back to the structure. For thicker insulation layers, furring strips have been added as a cladding attachment location. This approach has been used in the past on numerous Building America test homes and communities (both new and retrofit applications), and has been proven to be an effective and durable means to provide cladding attachment. However, the lack of engineering data has been a problem for many designers, contractors, and code officials. This research project developed baseline engineering analysis to support the installation of thick layers of exterior insulation on existing masonry and frame walls. Furthermore, water management details necessary to integrate windows, doors, decks, balconies and roofs were created to provide guidance on the integration of exterior insulation strategies with other enclosure elements.

Baker, P.

2013-01-01T23:59:59.000Z

431

2003 Plant Cell Walls Gordon Conference  

SciTech Connect (OSTI)

This conference will address recent progress in many aspects of cell wall biology. Molecular, genetic, and genomic approaches are yielding major advances in our understanding of the composition, synthesis, and architecture of plant cell walls and their dynamics during growth, and are identifying the genes that encode the machinery needed to make their biogenesis possible. This meeting will bring together international scientists from academia, industry and government labs to share the latest breakthroughs and perspectives on polysaccharide biosynthesis, wood formation, wall modification, expansion and interaction with other organisms, and genomic & evolutionary analyses of wall-related genes, as well as to discuss recent ''nanotechnological'' advances that take wall analysis to the level of a single cell.

Daniel J. Cosgrove

2004-09-21T23:59:59.000Z

432

Final Report for "Stabilization of resistive wall modes using moving metal walls"  

SciTech Connect (OSTI)

The UW experiment used a linear pinch experiment to study the stabilization of MHD by moving metal walls. The methodology of the experiment had three steps. (1) Identify and understand the no-wall MHD instability limits and character, (2) identify and understand the thin-wall MHD instabilities (re- sistive wall mode), and then (3) add the spinning wall and understand its impact on stability properties. During the duration of the grant we accomplished all 3 of these goals, discovered new physics, and completed the experiment as proposed.

Forest, Cary B.

2014-02-05T23:59:59.000Z

433

Assessment of Wall Shear Stress Changes in Arteries and Veins of Arteriovenous Polytetrafluoroethylene Grafts Using Magnetic Resonance Imaging  

SciTech Connect (OSTI)

The purpose of the study was to determine simultaneously the temporal changes in luminal vessel area, blood flow, and wall shear stress (WSS) in both the anastomosed artery (AA) and vein (AV) of arteriovenous polytetrafluoroethylene (PTFE) grafts. PTFE grafts were placed from the iliac artery to the ipsilateral iliac vein in 12 castrated juvenile male pigs. Contrast-enhanced magnetic resonance angiograpgy with cine phase-contrast magnetic resonance imaging was performed. Luminal vessel area, blood flow, and WSS in the aorta, AA, AV, and inferior vena cava were determined at 3 days (D3), 7 days (D7), and 14 days (D14) after graft placement. Elastin von Gieson staining of the AV was performed. The average WSS of the AA was highest at D3 and then decreased by D7 and D14. In contrast, the average WSS and intima-to-media ratio of the AV increased from D3 to D7 and peaked by D14. Similarly, the average area of the AA was highest by D7 and began to approximate the control artery by D14. The average area of the AV had decreased to its lowest by D7. High blood flows through the AA causes a decrease in average WSS and increase in the average luminal vessel area, whereas at the AV, the average WSS and intima-to-media ratio both increase while the average luminal vessel area decreases.

Misra, Sanjay, E-mail: Misra.sanjay@mayo.edu; Woodrum, David A. [Mayo Clinic, Department of Radiology (United States); Homburger, Jay [Medical College of Georgia, Department of Vascular Surgery (United States); Elkouri, Stephane [Centre Hospitalier de I'Universite de Montreal, Department of Vascular Surgery (Canada); Mandrekar, Jayawant N. [Mayo Clinic, Division of Biostatistics (United States); Barocas, Victor [University of Minnesota, Department of Biomedical Engineering (United States); Glockner, James F. [Mayo Clinic, Department of Radiology (United States); Rajan, Dheeraj K. [Toronto General Hospital, University Health Network, Department of Medical Imaging, Division of Vascular and Interventional Radiology (Canada); Mukhopadhyay, Debabrata [Mayo Clinic, Department of Biochemistry and Molecular Biology (United States)

2006-08-15T23:59:59.000Z

434

Control of linear modes in cylindrical resistive magnetohydrodynamics with a resistive wall, plasma rotation, and complex gain  

SciTech Connect (OSTI)

Feedback stabilization of magnetohydrodynamic (MHD) modes in a tokamak is studied in a cylindrical model with a resistive wall, plasma resistivity, viscosity, and toroidal rotation. The control is based on a linear combination of the normal and tangential components of the magnetic field just inside the resistive wall. The feedback includes complex gain, for both the normal and for the tangential components, and it is known that the imaginary part of the feedback for the former is equivalent to plasma rotation [J. M. Finn and L. Chacon, Phys. Plasmas 11, 1866 (2004)]. The work includes (1) analysis with a reduced resistive MHD model for a tokamak with finite ? and with stepfunction current density and pressure profiles, and (2) computations with a full compressible visco-resistive MHD model with smooth decreasing profiles of current density and pressure. The equilibria are stable for ??=?0 and the marginal stability values ?{sub rp,rw}?wall; resistive plasma, ideal wall; ideal plasma, resistive wall; and ideal plasma, ideal wall) are computed for both models. The main results are: (a) imaginary gain with normal sensors or plasma rotation stabilizes below ?{sub rp,iw} because rotation suppresses the diffusion of flux from the plasma out through the wall and, more surprisingly, (b) rotation or imaginary gain with normal sensors destabilizes above ?{sub rp,iw} because it prevents the feedback flux from entering the plasma through the resistive wall to form a virtual wall. A method of using complex gain G{sub i} to optimize in the presence of rotation in this regime with ??>??{sub rp,iw} is presented. The effect of imaginary gain with tangential sensors is more complicated but essentially destabilizes above and below ?{sub rp,iw}.

Brennan, D. P. [Department of Astrophysical Sciences, Princeton University, Princeton, New Jersey 08544 (United States); Finn, J. M. [Applied Mathematics and Plasma Physics, Theoretical Division, Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States)

2014-10-15T23:59:59.000Z

435

High temperature pressure gauge  

DOE Patents [OSTI]

A high temperature pressure gauge comprising a pressure gauge positioned in fluid communication with one end of a conduit which has a diaphragm mounted in its other end. The conduit is filled with a low melting metal alloy above the diaphragm for a portion of its length with a high temperature fluid being positioned in the remaining length of the conduit and in the pressure gauge.

Echtler, J. Paul (Pittsburgh, PA); Scandrol, Roy O. (Library, PA)

1981-01-01T23:59:59.000Z

436

Thrust allocation with power management functionality on dynamically positioned vessels  

E-Print Network [OSTI]

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

Johansen, Tor Arne

437

Modelling the Induced Magnetic Signature of Naval Vessels  

E-Print Network [OSTI]

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

Low, Robert

438

Response of a vessel to waves at zero ship speed  

E-Print Network [OSTI]

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

439

PublicationsmailagreementNo.40014024 the VeSSeL WILL  

E-Print Network [OSTI]

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

Pedersen, Tom

440

Sampling and Analysis Plan for PUREX canyon vessel flushing  

SciTech Connect (OSTI)

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

Villalobos, C.N.

1995-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "walled pressure vessel" 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

NETL SOFC: Pressurized Systems  

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

(HHV) with greater than 97 percent carbon capture, near-zero emissions, and low water usage. The Pressurized Systems key technology is developing a deeper understanding on the...

442

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

SciTech Connect (OSTI)

The Savannah River Site, in conjunction with AREVA Federal services, has designed a process to treat dissolved radioactive waste solids with ozone. It is known that in this radioactive waste process, radionuclides radiolytically break down water into gaseous hydrogen and oxygen, which presents a well defined flammability hazard. Flammability limits have been established for both ozone and hydrogen separately; however, there is little information on mixtures of hydrogen and ozone. Therefore, testing was designed to provide critical flammability information necessary to support safety related considerations for the development of ozone treatment and potential scale-up to the commercial level. Since information was lacking on flammability issues at low levels of hydrogen and ozone, a testing program was developed to focus on filling this portion of the information gap. A 2-L vessel was used to conduct flammability tests at atmospheric pressure and temperature using a fuse wire ignition source at 1 percent ozone intervals spanning from no ozone to the Lower Flammable Limit (LFL) of ozone in the vessel, determined as 8.4%(v/v) ozone. An ozone generator and ozone detector were used to generate and measure the ozone concentration within the vessel in situ, since ozone decomposes rapidly on standing. The lower flammability limit of hydrogen in an ozone-oxygen mixture was found to decrease from the LFL of hydrogen in air, determined as 4.2 % (v/v) in this vessel. From the results of this testing, Savannah River was able to develop safety procedures and operating parameters to effectively minimize the formation of a flammable atmosphere.

Sherburne, Carol [Savannah River Site (SRS), Aiken, SC (United States). Savannah River Remediation, LLC; Osterberg, Paul [Fauske and Associates, LLC, Burr Ridge, IL (United States); Johnson, Tom [Fauske and Associates, LLC, Burr Ridge, IL (United States); Frawely, Thomas [Fauske and Associates, LLC, Burr Ridge, IL (United States)

2013-01-23T23:59:59.000Z

443

FREE CONVECTIVE LAMINAR FLOW WITHIN THE TROMBE WALL CHANNEL  

E-Print Network [OSTI]

LAMINAR FLOW WITHIN THE TROMBE WALL CHANNEL H. Akbarf andLAMINAR FLOW WITHIN THE TROMBE WALL CHANNEL H. Akbari andchannel surfaces of the Trombe wall has been investigated.

Akbari, H.

2011-01-01T23:59:59.000Z

444

Mathematical modeling of clearance between wall of coke oven and coke cake  

SciTech Connect (OSTI)

A mathematical model was developed for estimating the clearance between the wall of the coke oven and the coke cake. The prediction model is based on the balance between the contractile force and the coking pressure. A clearance forms wh