Sample records for rectangular pressure vessel

  1. Dual shell pressure balanced vessel

    DOE Patents [OSTI]

    Fassbender, Alexander G. (West Richland, WA)

    1992-01-01T23:59:59.000Z

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

  2. Reactor pressure vessel nozzle

    DOE Patents [OSTI]

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

    1994-10-04T23:59:59.000Z

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

  3. High pressure storage vessel

    DOE Patents [OSTI]

    Liu, Qiang

    2013-08-27T23:59:59.000Z

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

  4. Level indicator for pressure vessels

    DOE Patents [OSTI]

    Not Available

    1982-04-28T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Paulino, Glaucio H.

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

  6. Reactor pressure vessel vented head

    DOE Patents [OSTI]

    Sawabe, James K. (San Jose, CA)

    1994-01-11T23:59:59.000Z

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

  7. Reactor pressure vessel vented head

    DOE Patents [OSTI]

    Sawabe, J.K.

    1994-01-11T23:59:59.000Z

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

  8. Reactor pressure vessel. Status report

    SciTech Connect (OSTI)

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

    1996-10-01T23:59:59.000Z

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

  9. Lightweight bladder lined pressure vessels

    DOE Patents [OSTI]

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

    1998-08-25T23:59:59.000Z

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

  10. Lightweight bladder lined pressure vessels

    DOE Patents [OSTI]

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

    1998-01-01T23:59:59.000Z

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

  11. PURE NIOBIUM AS A PRESSURE VESSEL MATERIAL

    SciTech Connect (OSTI)

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

    2010-04-09T23:59:59.000Z

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

  12. Reactor pressure vessel with forged nozzles

    DOE Patents [OSTI]

    Desai, Dilip R. (Fremont, CA)

    1993-01-01T23:59:59.000Z

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

  13. 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...

  14. 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...

  15. 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...

  16. 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...

  17. 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...

  18. Pressure drop and heat transfer distributions in three-pass rectangular channels with rib turbulators

    E-Print Network [OSTI]

    Zhang, Peng

    1988-01-01T23:59:59.000Z

    PRESSURE DROP AND HEAT TRANSFER DISTRIBUTIONS IN THREE-PASS RECTANGULAR CHANNELS WITH RIB TURBULATORS A THESIS by PENG ZHANG Submitted to the Graduate College of Texas AkM University in partial fulfillment of the requirements for the degree... of MASTER OF SCIENCE August 1988 Major Subject: Mechanical Engineering PRESSURE DROP AND HEAT TRANSFER DISTRIBUTIONS IN THREE-PASS RECTANGULAR CHANNELS WITH RIB TURBULATORS A THESIS by PENG ZHANG Approved as to style and content by: J. C. Han...

  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)

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

    1996-05-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Cambridge, University of

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

  2. Reactor Pressure Vessel Head Packaging & Disposal

    SciTech Connect (OSTI)

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

    2003-02-26T23:59:59.000Z

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

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

    DOE Patents [OSTI]

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

    2004-03-23T23:59:59.000Z

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

  4. Neutron shielding panels for reactor pressure vessels

    DOE Patents [OSTI]

    Singleton, Norman R. (Murrysville, PA)

    2011-11-22T23:59:59.000Z

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

  5. Austenite Grain Growth in a Nuclear Pressure Vessel Steel

    E-Print Network [OSTI]

    Cambridge, University of

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

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

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

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

  7. 2005 ASME Pressure Vessels and Piping Conference Denver, Colorado, USA

    E-Print Network [OSTI]

    Özer, Mutlu

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

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

    E-Print Network [OSTI]

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

  9. Report of the terawatt laser pressure vessel committee

    SciTech Connect (OSTI)

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

    2000-09-25T23:59:59.000Z

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

  10. Technical Appendix to Cryogenic Pressure Vessels

    SciTech Connect (OSTI)

    Mulholland, G.T.; Rucinski, R.A; /Fermilab

    1990-02-22T23:59:59.000Z

    The 20,000 gls. Liquid Argon dewar stores up to 15,000 gls. of high purity (<1.0 ppm O{sub 2}, 0.999995) LAr for use in the Liquid Argon calorimeters of E740, the D0 collider detector, at elevation 707-feet. The dewar provides for the total detector volume of 11,000 gls and a 4,000 gls. storage inventory. The large gas volume ({ge}5,000 gls.) serves operational needs and guards against overfill concerns. The LAr dewar functions in two modes: (1) low pressure (16 psi relief) storage, and liquid and gas transfer operations to and from the low pressure (13 psi relief) detector cryostats, and (2) high pressure (65 psi relief) liquid transfer operations to and from a delivery trailer at elevation 743-feet. The storage function is intended to be long term and nonventing. The dewar is equipped with a 40 kW LN{sub 2} condenser that operates to maintain the pressure constant in the storage mode. This service exactly parallels the NeH{sub 2} and D{sub 2} storage dewar services provided at the 15-feet bubble chamber for its operation.

  11. Design Considerations For Blast Loads In Pressure Vessels.

    SciTech Connect (OSTI)

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

    2007-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Wang, Jy-An John [ORNL

    2010-08-01T23:59:59.000Z

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

  13. 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

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

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

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

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

    SciTech Connect (OSTI)

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

    1994-10-01T23:59:59.000Z

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

  16. Measurement of heat transfer and pressure drop in rectangular channels with turbulence promoters. Final report

    SciTech Connect (OSTI)

    Han, J. C.; Park, J. S.; Ibrahim, M. Y.

    1986-09-01T23:59:59.000Z

    Periodic rib turbulators were used in advanced turbine cooling designs to enhance the internal heat transfer. The objective of the present project was to investigate the combined effects of the rib angle of attack and the channel aspect ratio on the local heat transfer and pressure drop in rectangular channels with two opposite ribbed walls for Reynolds number varied from 10,000 to 60,000. The channel aspect ratio (W/H) was varied from 1 to 2 to 4. The rib angle of attack (alpha) was varied from 90 to 60 to 45 to 30 degree. The highly detailed heat transfer coefficient distribution on both the smooth side and the ribbed side walls from the channel sharp entrance to the downstream region were measured. The results showed that, in the square channel, the heat transfer for the slant ribs (alpha = 30 -45 deg) was about 30% higher that of the transverse ribs (alpha = 90 deg) for a constant pumping power. However, in the rectangular channels (W/H = 2 and 4, ribs on W side), the heat transfer at alpha = 30 -45 deg was only about 5% higher than 90 deg. The average heat transfer and friction correlations were developed to account for rib spacing, rib angle, and channel aspect ratio over the range of roughness Reynolds number.

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

    SciTech Connect (OSTI)

    Pilch, M.; Tarbell, W.M.

    1985-09-01T23:59:59.000Z

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

  18. PRESSURIZATION OF CONTAINMENT VESSELS FROM PLUTONIUM OXIDE CONTENTS

    SciTech Connect (OSTI)

    Hensel, S.

    2012-03-27T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2011-06-07T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Motta, Arthur T.

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

  1. Gamma ray-induced embrittlement of pressure vessel alloys

    SciTech Connect (OSTI)

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

    1994-11-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Cambridge, University of

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

  3. Creep of A508/533 Pressure Vessel Steel

    SciTech Connect (OSTI)

    Richard Wright

    2014-08-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Horn, R.M.

    2011-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2012-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2010-04-13T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Chapman, Clark R.

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

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

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

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

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

    E-Print Network [OSTI]

    Fuquen-Molano, Rosendo

    1982-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2006-07-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1999-01-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

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

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

    SciTech Connect (OSTI)

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

    1998-12-31T23:59:59.000Z

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

  14. 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...

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

    DOE Patents [OSTI]

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

    2014-10-28T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Barr, Al

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

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

    E-Print Network [OSTI]

    Chen, Sheng

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

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

    E-Print Network [OSTI]

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

  19. 1 Copyright 2012 by ASME Proceedings of the ASME 2012 Pressure Vessels & Piping Division Conference

    E-Print Network [OSTI]

    Tijsseling, A.S.

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

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

    SciTech Connect (OSTI)

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

    1995-12-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Nguyen, Ba Nghiep; Simmons, Kevin L.

    2013-07-22T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2011-07-01T23:59:59.000Z

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

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

    DOE Patents [OSTI]

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

    2013-01-15T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2011-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    GJ Schuster, FA Simonen, SR Doctor

    2008-04-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1994-08-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2013-08-26T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2006-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Habil; Ziliukas, A.

    1997-04-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2009-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2009-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2014-02-18T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1994-12-31T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Motta, Arthur T.

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

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

    SciTech Connect (OSTI)

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

    2012-09-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2011-07-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1989-01-01T23:59:59.000Z

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

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

    DOE Patents [OSTI]

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

    1999-07-27T23:59:59.000Z

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

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

    DOE Patents [OSTI]

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

    1999-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2012-07-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1996-12-31T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Pritam Chakraborty; S. Bulent Biner

    2014-08-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2013-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1993-09-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1995-12-31T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2010-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1998-08-01T23:59:59.000Z

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

  8. 1 Copyright 2010 by ASME Proceedings of the ASME 2010 Pressure Vessels & Piping Division / K-PVP Conference

    E-Print Network [OSTI]

    Tijsseling, A.S.

    1 Copyright © 2010 by ASME Proceedings of the ASME 2010 Pressure Vessels & Piping Division / K-6]. Developers and users of computational codes need full- scale data with which to compare their theoretical;2 Copyright © 2010 by ASME Here, a large-scale pipeline test rig at Deltares, Delft, The Netherlands has been

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

    SciTech Connect (OSTI)

    Odette, G. Robert; Yamamoto, Takuya

    2013-06-17T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2007-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2014-11-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1996-12-31T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1996-12-31T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1996-12-31T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1996-12-31T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1998-03-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1997-02-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2001-03-15T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2009-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    J. K. Wright; R. N. Wright

    2008-04-01T23:59:59.000Z

    The U.S. Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic, or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development Program is responsible for performing research and development on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2¼Cr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have concluded, however, that with adequate engineered cooling of the vessel, the A508/533 steels are also acceptable.

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

    SciTech Connect (OSTI)

    Rucinski, Russ; /Fermilab

    1996-11-11T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1996-05-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1996-12-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1999-08-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2007-03-21T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1996-12-31T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Rawls, G

    2007-04-30T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2013-11-26T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1996-12-31T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1993-12-31T23:59:59.000Z

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

  11. The influence of selected containment structures on debris dispersal and transport following high pressure melt ejection from the reactor vessel

    SciTech Connect (OSTI)

    Pilch, M.; Tarbell, W.W.; Brockmann, J.E.

    1988-09-01T23:59:59.000Z

    High pressure expulsion of molten core debris from the reactor pressure vessel may result in dispersal of the debris from the reactor cavity. In most plants, the cavity exits into the containment such that the debris impinges on structures. Retention of the debris on the structures may affect the further transport of the debris throughout the containment. Two tests were done with scaled structural shapes placed at the exit of 1:10 linear scale models of the Zion cavity. The results show that the debris does not adhere significantly to structures. The lack of retention is attributed to splashing from the surface and reentrainment in the gas flowing over the surface. These processes are shown to be applicable to reactor scale. A third experiment was done to simulate the annular gap between the reactor vessel and cavity wall. Debris collection showed that the fraction of debris exiting through the gap was greater than the gap-to-total flow area ratio. Film records indicate that dispersal was primarily by entrainment of the molten debris in the cavity. 29 refs., 36 figs., 11 tabs.

  12. Effect of silicon on ultra-low temperature toughness of Nb–Ti microalloyed cryogenic pressure vessel steels

    SciTech Connect (OSTI)

    Qiu, J.A. [Hubei Collaborative Innovation Center for Advanced Steels, International Research Institute for Steel Technology, Wuhan University of Science and Technology, Wuhan 430081 (China); Wu, K.M., E-mail: wukaiming2000@yahoo.com [Hubei Collaborative Innovation Center for Advanced Steels, International Research Institute for Steel Technology, Wuhan University of Science and Technology, Wuhan 430081 (China); Li, J.H. [Hubei Collaborative Innovation Center for Advanced Steels, International Research Institute for Steel Technology, Wuhan University of Science and Technology, Wuhan 430081 (China); Research and Development Center of WISCO, Wuhan 430080 (China); Hodgson, P.D. [Hubei Collaborative Innovation Center for Advanced Steels, International Research Institute for Steel Technology, Wuhan University of Science and Technology, Wuhan 430081 (China); Institute for Frontier Materials, Deakin University, Geelong, Victoria 3220 (Australia); Hou, T.P. [Hubei Collaborative Innovation Center for Advanced Steels, International Research Institute for Steel Technology, Wuhan University of Science and Technology, Wuhan 430081 (China); Ding, Q.F. [Research and Development Center of WISCO, Wuhan 430080 (China)

    2013-09-15T23:59:59.000Z

    The effect of Si on the ultra-low temperature toughness of Nb–Ti microalloyed cryogenic pressure vessel steels was investigated by electron back-scattered diffraction and transmission electron microscope with energy dispersive spectroscopy. Equiaxed ferrite and bainite were obtained in the tempered steels with small Si additions. Nanosized Nb–Ti carbides (< 10 nm) were formed in the steel containing 0.05% Si, whereas much coarser carbides (> 30 nm) were found in the steel containing 0.47% Si. The ultra-low temperature toughness of the Nb–Ti microalloyed cryogenic pressure vessel steel was remarkably enhanced by the reduction in the Si content, which was attributed to the pre-existing iron carbide formation before the precipitation of nanosized Nb–Ti carbides during tempering. - Highlights: • Nanosized Nb-Ti carbides formed in the tempered steel with smaller Si addition. • Coarser Nb-Ti carbides formed in the tempered steel with more Si addition. • Pre-existing cememtites provide nucleation sites for Nb-Ti carbide precipitation. • Ultra-low temperature toughness was remarkably enhanced by Si content reduction.

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

    SciTech Connect (OSTI)

    Pritam Chakraborty; S. Bulent Biner

    2013-10-01T23:59:59.000Z

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

  14. Assessment of Negligible Creep, Off-Normal Welding and Heat Treatment of Gr91 Steel for Nuclear Reactor Pressure Vessel Application

    SciTech Connect (OSTI)

    Ren, Weiju [ORNL; Terry, Totemeier [Idaho National Laboratory (INL)

    2006-10-01T23:59:59.000Z

    Two different topics of Grade 91 steel are investigated for Gen IV nuclear reactor pressure vessel application. On the first topic, negligible creep of Grade 91 is investigated with the motivation to design the reactor pressure vessel in negligible creep regime and eliminate costly surveillance programs during the reactor operation. Available negligible creep criteria and creep strain laws are reviewed, and new data needs are evaluated. It is concluded that modifications of the existing criteria and laws, together with their associated parameters, are needed before they can be reliably applied to Grade 91 for negligible creep prediction and reactor pressure vessel design. On the second topic, effects of off-normal welding and heat treatment on creep behavior of Grade 91 are studied with the motivation to better define the control over the parameters in welding and heat treatment procedures. The study is focused on off-normal austenitizing temperatures and improper cooling after welding but prior to post-weld heat treatment.

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

    SciTech Connect (OSTI)

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

    1994-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    J. K. Wright; R. N. Wright

    2010-07-01T23:59:59.000Z

    The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750°C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

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

    SciTech Connect (OSTI)

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

    2010-12-22T23:59:59.000Z

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

  18. Experimental results of direct containment heating by high-pressure melt ejection into the Surtsey vessel: The DCH-3 and DCH-4 tests

    SciTech Connect (OSTI)

    Allen, M.D.; Pilch, M.; Brockmann, J.E.; Tarbell, W.W. (Sandia National Labs., Albuquerque, NM (United States)); Nichols, R.T. (Ktech Corp., Albuquerque, NM (United States)); Sweet, D.W. (AEA Technology, Winfrith (United Kingdom))

    1991-08-01T23:59:59.000Z

    Two experiments, DCH-3 and DCH-4, were performed at the Surtsey test facility to investigate phenomena associated with a high-pressure melt ejection (HPME) reactor accident sequence resulting in direct containment heating (DCH). These experiments were performed using the same experimental apparatus with identical initial conditions, except that the Surtsey test vessel contained air in DCH-3 and argon in DCH-4. Inerting the vessel with argon eliminated chemical reactions between metallic debris and oxygen. Thus, a comparison of the pressure response in DCH-3 and DCH-4 gave an indication of the DCH contribution due to metal/oxygen reactions. 44 refs., 110 figs., 43 tabs.

  19. Behavior of core debris ejected from a pressurized vessel into scaled reactor cavities

    SciTech Connect (OSTI)

    Tarbell, W.W.; Pilch, M.; Brockmann, J.E.

    1984-01-01T23:59:59.000Z

    Results from four recent 1:10 scale experiments are presented along with analyses of the possible consequences for plant geometries. The tests cover a range in initial system pressure from 4 to 12 MPa, with either dry or water-filled cavities. Nearly all of the core debris is dispersed from the cavity with less than five percent (5%) of the original mass found adhered to the exposed cavity surfaces. Those tests involving water in the cavity show the water being expelled as a slug ahead of the dispersed melt. Models for the interaction of the ejected core debris with the containment atmosphere show that both thermal and chemical energy is liberated from the debris. The calculated pressurization from direct heating of the containment atmosphere can threaten even the most robust containments. Models and experiments are currently being devised to study the possible mitigating effects of the above-cavity structures.

  20. TEMP-STRESS analysis of a reinforced concrete vessel under internal pressure

    SciTech Connect (OSTI)

    Marchertas, A.H.; Kennedy, J.M.; Pfeiffer, P.A.

    1987-01-01T23:59:59.000Z

    Prediction of the response of the Sandia National laboratory 1/6-scale reinforced concrete containment model test was obtained by Argonne National Laboratory (ANL) employing a computer program developed by ANL. The test model was internally pressurized to failure. The two-dimensional code TEMP-STRESS (1-5) has been developed at ANL for stress analysis of plane and axisymmetric 2-D reinforced structures under various thermal conditions. The program is applicable to a wide variety of nonlinear problems, and is utilized in the present study. The comparison of these pretest computations with test data on the containment model should be a good indication of the state of the code.

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

    SciTech Connect (OSTI)

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

    1995-07-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Grotke, G.E.

    1980-04-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1995-12-31T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2012-09-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    David H. Bothell

    2000-04-30T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1991-10-01T23:59:59.000Z

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

  7. 1-Dimensional simulation of thermal annealing in a commercial nuclear power plant reactor pressure vessel wall section

    SciTech Connect (OSTI)

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

    1994-11-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2011-07-01T23:59:59.000Z

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

  9. Comparison of MELCOR modeling techniques and effects of vessel water injection on a low-pressure, short-term, station blackout at the Grand Gulf Nuclear Station

    SciTech Connect (OSTI)

    Carbajo, J.J.

    1995-06-01T23:59:59.000Z

    A fully qualified, best-estimate MELCOR deck has been prepared for the Grand Gulf Nuclear Station and has been run using MELCOR 1.8.3 (1.8 PN) for a low-pressure, short-term, station blackout severe accident. The same severe accident sequence has been run with the same MELCOR version for the same plant using the deck prepared during the NUREG-1150 study. A third run was also completed with the best-estimate deck but without the Lower Plenum Debris Bed (BH) Package to model the lower plenum. The results from the three runs have been compared, and substantial differences have been found. The timing of important events is shorter, and the calculated source terms are in most cases larger for the NUREG-1150 deck results. However, some of the source terms calculated by the NUREG-1150 deck are not conservative when compared to the best-estimate deck results. These results identified some deficiencies in the NUREG-1150 model of the Grand Gulf Nuclear Station. Injection recovery sequences have also been simulated by injecting water into the vessel after core relocation started. This marks the first use of the new BH Package of MELCOR to investigate the effects of water addition to a lower plenum debris bed. The calculated results indicate that vessel failure can be prevented by injecting water at a sufficiently early stage. No pressure spikes in the vessel were predicted during the water injection. The MELCOR code has proven to be a useful tool for severe accident management strategies.

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

    SciTech Connect (OSTI)

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

    1998-07-01T23:59:59.000Z

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

  11. Numerical simulation of thermoconvective flows and more uniform depositions in a cold wall rectangular APCVD reactor.

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    rectangular APCVD reactor. Xavier NICOLAS1 , Abderrahmane BENZAOUI1 , Shihe XIN2 1 Université Paris-Est, LETEM in the cold wall horizontal rectangular thermal CVD reactors correspond to steady longitudinal the pressure in the reactor. It consists in adequately exciting the parallel thermoconvective rolls at channel

  12. Random rectangular Graphs

    E-Print Network [OSTI]

    Estrada, Ernesto

    2015-01-01T23:59:59.000Z

    A generalization of the random geometric graph (RGG) model is proposed by considering a set of points uniformly and independently distributed on a rectangle of unit area instead of on a unit square \\left[0,1\\right]^{2}. The topological properties, such as connectivity, average degree, average path length and clustering, of the random rectangular graphs (RRGs) generated by this model are then studied as a function of the rectangle sides lengths a and b=1/a, and the radius r used to connect the nodes. When a=1 we recover the RGG, and when a\\rightarrow\\infty the very elongated rectangle generated resembles a one-dimensional RGG. We provided computational and analytical evidence that the topological properties of the RRG differ significantly from those of the RGG. The connectivity of the RRG depends not only on the number of nodes as in the case of the RGG, but also on the side length of the rectangle. As the rectangle is more elongated the critical radius for connectivity increases following first a power-law an...

  13. 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)

    Sullivan, Edmund J.; Anderson, Michael T.

    2014-06-10T23:59:59.000Z

    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.

  14. 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)

    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

    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)

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

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

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

  16. 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)

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

    2012-09-17T23:59:59.000Z

    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.

  17. Optimization Online - Rectangular sets of probability measures

    E-Print Network [OSTI]

    Alexander Shapiro

    2014-10-24T23:59:59.000Z

    Oct 24, 2014 ... We define rectangularity as a property of dynamic decomposition of a distributionally robust stochastic optimization problem and show how it ...

  18. Ion transport membrane module and vessel system

    DOE Patents [OSTI]

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

    2008-02-26T23:59:59.000Z

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

  19. Ion transport membrane module and vessel system

    DOE Patents [OSTI]

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

    2012-02-14T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1994-12-31T23:59:59.000Z

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

  1. Anosov maps with rectangular holes. Nonergodic cases.

    E-Print Network [OSTI]

    Ingenier'ia. Universidad de la Rep'ublica C.C. 30, Montevideo, Uruguay E­mail: roma@fing.edu.uy; Fax: (598 Partially supported by CONICYT (Uruguay). 1 #12; Running head: Anosov maps with rectangular holes Address

  2. Enhancements of a Combustion Vessel to Determine Laminar Flame Speeds of Hydrocarbon Blends with Helium Dilution at Elevated Temperatures and Pressures

    E-Print Network [OSTI]

    Plichta, Drew

    2013-04-03T23:59:59.000Z

    of importance comes the need for baseline data such as laminar flame speed of said fuels. While flame speeds at standard temperature and pressure have been extensively studied in the literature, experimental data at turbine-like conditions are still lacking...

  3. Ultrasonic liquid-level detector for varying temperature and pressure environments

    DOE Patents [OSTI]

    Anderson, R.L.; Miller, G.N.

    1981-10-26T23:59:59.000Z

    An ultrasonic liquid level detector for use in varying temperature and pressure environments, such as a pressurized water nuclear reactor vessel, is provided. The detector employs ultrasonic extensional and torsional waves launched in a multiplexed alternating sequence into a common sensor. The sensor is a rectangular cross section stainless steel rod which extends into the liquid medium whose level is to be detected. The sensor temperature derived from the extensional wave velocity measurements is used to compensate for the temperature dependence of the torsional wave velocity measurements which are also level dependent. The torsional wave velocity measurements of a multiple reflection sensor then provide a measurement of liquid level over a range of several meters with a small uncertainty over a temperature range of 20 to 250/sup 0/C and pressures up to 15 MPa.

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

    E-Print Network [OSTI]

    Melcher, Peter

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

  5. EDS V25 containment vessel explosive qualification test report.

    SciTech Connect (OSTI)

    Rudolphi, John Joseph

    2012-04-01T23:59:59.000Z

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

  6. Coal gasification vessel

    DOE Patents [OSTI]

    Loo, Billy W. (Oakland, CA)

    1982-01-01T23:59:59.000Z

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

  7. Electrochemical apparatus comprising modified disposable rectangular cuvette

    DOE Patents [OSTI]

    Dattelbaum, Andrew M; Gupta, Gautam; Morris, David E

    2013-09-10T23:59:59.000Z

    Electrochemical apparatus includes a disposable rectangular cuvette modified with at least one hole through a side and/or the bottom. Apparatus may include more than one cuvette, which in practice is a disposable rectangular glass or plastic cuvette modified by drilling the hole(s) through. The apparatus include two plates and some means of fastening one plate to the other. The apparatus may be interfaced with a fiber optic or microscope objective, and a spectrometer for spectroscopic studies. The apparatus are suitable for a variety of electrochemical experiments, including surface electrochemistry, bulk electrolysis, and flow cell experiments.

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

    DOE Patents [OSTI]

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

    2010-02-09T23:59:59.000Z

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

  9. Pressure suppression containment system

    DOE Patents [OSTI]

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

    1994-03-15T23:59:59.000Z

    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.

  10. Pressure suppression containment system

    DOE Patents [OSTI]

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

    1994-03-15T23:59:59.000Z

    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.

  11. 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

  12. Neutrino Factory Target Vessel

    E-Print Network [OSTI]

    McDonald, Kirk

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

  13. ac rectangular wave: Topics by E-print Network

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

    vibration analysis. I. Mansour Nikkhah-bahrami; Masih Loghmani; Mostafa Pooyanfar 6 Propagation of Electromagnetic Waves on a Rectangular Lattice of Polarizable Points...

  14. Using SA508/533 for the HTGR Vessel Material

    SciTech Connect (OSTI)

    Larry Demick

    2012-06-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

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

    2010-01-01T23:59:59.000Z

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

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

    Broader source: Energy.gov [DOE]

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

  17. HEART AND BLOOD VESSELS CARDIOVASCULARCARDIOVASCULAR

    E-Print Network [OSTI]

    Cochran-Stafira, D. Liane

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

  18. Viscous lock-exchange in rectangular channels

    E-Print Network [OSTI]

    Jerome Martin; Nicole Rakotomalala; Laurent Talon; Dominique Salin

    2010-11-29T23:59:59.000Z

    In a viscous lock-exchange gravity current, which describes the reciprocal exchange of two fluids of different densities in a horizontal channel, the front between two Newtonian fluids spreads as the square root of time. The resulting diffusion coefficient reflects the competition between the buoyancy driving effect and the viscous damping, and depends on the geometry of the channel. This lock-exchange diffusion coefficient has already been computed for a porous medium, a 2D Stokes flow between two parallel horizontal boundaries separated by a vertical height, H, and, recently, for a cylindrical tube. In the present paper, we calculate it, analytically, for a rectangular channel (horizontal thickness b, vertical height, H) of any aspect ratio (H/b) and compare our results with experiments in horizontal rectangular channels for a wide range of aspect ratios (1/10-10). We also discuss the 2D Stokes-Darcy model for flows in Hele-Shaw cells and show that it leads to a rather good approximation, when an appropriate Brinkman correction is used.

  19. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, Juhani (Karhula, FI)

    1996-01-01T23:59:59.000Z

    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.

  20. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, J.

    1996-03-19T23:59:59.000Z

    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.

  1. Reactor vessel support system

    DOE Patents [OSTI]

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

    1982-01-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Suprijadi; Mohamad Rendi; Petrus Subekti; Sparisoma Viridi

    2013-12-14T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Suprijadi,; Subekti, Petrus; Viridi, Sparisoma

    2013-01-01T23:59:59.000Z

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

  4. Method and structure for cache aware transposition via rectangular subsections

    DOE Patents [OSTI]

    Gustavson, Fred Gehrung; Gunnels, John A

    2014-02-04T23:59:59.000Z

    A method and structure for transposing a rectangular matrix A in a computer includes subdividing the rectangular matrix A into one or more square submatrices and executing an in-place transposition for each of the square submatrices A.sub.ij.

  5. Fresnel approximations for acoustic fields of rectangularly symmetric sources

    E-Print Network [OSTI]

    Mast, T. Douglas

    Fresnel approximations for acoustic fields of rectangularly symmetric sources T. Douglas Masta for determining the acoustic fields of rectangularly symmetric, baffled, time-harmonic sources under the Fresnel. The expressions presented are generalized to three different Fresnel approximations that correspond, respectively

  6. The Number of Hamiltonian Paths in a Rectangular Grid

    E-Print Network [OSTI]

    Collins, Karen L.

    The Number of Hamiltonian Paths in a Rectangular Grid@wesleyan.edu Abstract It is easy to find out which rectangular m vertex by n vertex grids have answers for grids with fixed m for m = 1, 2, 3, 4, 5. 1 Introduction Given a grid with m vertices

  7. Pressure testing of torispherical heads

    SciTech Connect (OSTI)

    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

    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.

  8. Physics from Angular Projection of Rectangular Grids

    E-Print Network [OSTI]

    Singh, Ashmeet

    2015-01-01T23:59:59.000Z

    In this paper, we present a mathematical model for the angular projection of a rectangular arrangement of points in a grid. This simple, yet interesting problem, has both a scholarly value and applications for data extraction techniques to study the physics of various systems. Our work can interest undergraduate students to understand subtle points in the angular projection of a grid and describes various quantities of interest in the projection with completeness and sufficient rigour. We show that for certain angular ranges, the projection has non-distinctness, and calculate the details of such angles, and correspondingly, the number of distinct points and the total projected length. We focus on interesting trends obtained for the projected length of the grid elements and present a simple application of the model to determine the geometry of an unknown grid whose spatial extensions are known, using measurement of the grid projection at two angles only. Towards the end, our model is shown to have potential ap...

  9. Pressure sensor for sealed containers

    DOE Patents [OSTI]

    Hodges, Franklin R. (Loudon, TN)

    2001-01-01T23:59:59.000Z

    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.

  10. Reactor vessel annealing system

    DOE Patents [OSTI]

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

    1991-01-01T23:59:59.000Z

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

  11. Pressure suppression system

    DOE Patents [OSTI]

    Gluntz, D.M.

    1994-10-04T23:59:59.000Z

    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.

  12. Pressure suppression system

    DOE Patents [OSTI]

    Gluntz, Douglas M. (San Jose, CA)

    1994-01-01T23:59:59.000Z

    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.

  13. 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

  14. Physics from Angular Projection of Rectangular Grids

    E-Print Network [OSTI]

    Ashmeet Singh

    2014-12-08T23:59:59.000Z

    In this paper, we present a mathematical model for the angular projection of a rectangular arrangement of points in a grid. This simple, yet interesting problem, has both a scholarly value and applications for data extraction techniques to study the physics of various systems. Our work can interest undergraduate students to understand subtle points in the angular projection of a grid and describes various quantities of interest in the projection with completeness and sufficient rigour. We show that for certain angular ranges, the projection has non-distinctness, and calculate the details of such angles, and correspondingly, the number of distinct points and the total projected length. We focus on interesting trends obtained for the projected length of the grid elements and present a simple application of the model to determine the geometry of an unknown grid whose spatial extensions are known, using measurement of the grid projection at two angles only. Towards the end, our model is shown to have potential applications in various branches of physical sciences including crystallography, astrophysics and bulk properties of materials.

  15. Neutrino Factory Mercury Vessel

    E-Print Network [OSTI]

    McDonald, Kirk

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

  16. The optimization of solar radiation upon a rectangular building

    E-Print Network [OSTI]

    Ingle, James Allen

    1973-01-01T23:59:59.000Z

    THE OPTIMIZATION OF SOLAR RADIATION UPON A RECTANGULAR BUILDING A Thesis by JAMES ALLEN INGLF. Submitted to the Graduate College of Texas AAM University in Partial fulfillment of the requirement for the degree of MASTER OF SCIENCE August 19... 3 Major Subjeoti Meteorology THE OPTIMIZATION OF SOLAR RADIATION UPON A RECTANGULAR BUILDING A Thesis by JAILS ALLEN INGLE Approved as to style and content bye Cha an of ittee Mr, ' ohn F, Griffiths Member Dro Robert CD Runnels Head...

  17. OSS 19.4 Pressure Safety 3/27/95

    Broader source: Energy.gov [DOE]

    The objective of this surveillance is to evaluate the contractor's implementation of programs to ensure the integrity of pressure vessels and minimize risks from failure of vessels to the public...

  18. High pressure melt ejection

    SciTech Connect (OSTI)

    Tarbell, W.W.; Brockmann, J.E.; Pilch, M.

    1983-01-01T23:59:59.000Z

    Recent probabilistic risk assessments have identified the potential for reactor pressure vessel failure while the reactor coolant system is at elevated pressure. The analyses postulate that the blowdown of steam and hydrogen into the reactor cavity will cause the core material to be swept from the cavity region into the containment building. The High Pressure Melt Streaming (HIPS) program is an experimental study of the high pressure ejection of molten material and subsequent interactions within a concrete cavity. The program focuses on using prototypic system conditions and scaled models of reactor geometries to accurately simulate the ex-vessel processes during high-pressure accident sequences. Scaling analyses of the experiment show that the criteria established for core debris removal from the cavity are met or exceeded. Tests are performed at two scales, representing 1/10th and 1/20th linear reproductions of the Zion reactor plant. Results of the 1/20th scale tests are presented.

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

    SciTech Connect (OSTI)

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

    1995-11-01T23:59:59.000Z

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

  20. High pressure liquid level monitor

    DOE Patents [OSTI]

    Bean, Vern E. (Frederick, MD); Long, Frederick G. (Ijamsville, MD)

    1984-01-01T23:59:59.000Z

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

  1. Single module pressurized fuel cell turbine generator system

    DOE Patents [OSTI]

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

    2001-01-01T23:59:59.000Z

    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.

  2. High pressure furnace

    DOE Patents [OSTI]

    Morris, D.E.

    1993-09-14T23:59:59.000Z

    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.

  3. High pressure furnace

    DOE Patents [OSTI]

    Morris, Donald E. (Kensington, CA)

    1993-01-01T23:59:59.000Z

    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.

  4. Start-up control system and vessel for LMFBR

    DOE Patents [OSTI]

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

    1987-01-01T23:59:59.000Z

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

  5. Start-up control system and vessel for LMFBR

    DOE Patents [OSTI]

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

    1987-01-01T23:59:59.000Z

    A reflux condensing start-up system 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.

  6. Bonfire Tests of High Pressure Hydrogen Storage Tanks

    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.

  7. 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

  8. Vessel structural support system

    DOE Patents [OSTI]

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

    1992-01-01T23:59:59.000Z

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

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

    DOE Patents [OSTI]

    Bowling, Michael Keith (Blackborough Cullompton, GB)

    2000-09-12T23:59:59.000Z

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

  10. Retrospective dosimetry analyses of reactor vessel cladding samples

    SciTech Connect (OSTI)

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

    2011-07-01T23:59:59.000Z

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

  11. Nuclear reactor construction with bottom supported reactor vessel

    DOE Patents [OSTI]

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

    1987-01-01T23:59:59.000Z

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

  12. The study and characterization of the major flow through rectangular slit virtual impactor

    E-Print Network [OSTI]

    Gupta, Amit

    2002-01-01T23:59:59.000Z

    The major flow through a rectangular slit virtual impactor was experimentally studied. The impactor was a two-dimensional rectangular slit dichotomous sampler. A series of experiments were performed using monodisperse liquid oleic acid particles...

  13. Flow adjustment and interior flow associated with a rectangular porous obstruction

    E-Print Network [OSTI]

    Rominger, Jeffrey Tsaros

    The flow at the leading edge and in the interior of a rectangular porous obstruction is described through experiments and scaling. The porous obstruction consists of an emergent, rectangular array of cylinders in shallow ...

  14. Random Young Diagrams in a Rectangular Box Dan Beltoft

    E-Print Network [OSTI]

    - portional to the exponential of their area (grand-canonical ensemble), and confined in a rectangular box for the unconfined case lead to a two-sided stationary Ornstein-Uhlenbeck process. Keywords: Young diagrams, Gauss of the classical Gauss polynomials, well- known in combinatorics as the generating functions for the number

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

    SciTech Connect (OSTI)

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

    2010-12-09T23:59:59.000Z

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

  16. Reactor vessel support system. [LMFBR

    DOE Patents [OSTI]

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

    1980-05-09T23:59:59.000Z

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

  17. 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.

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

    SciTech Connect (OSTI)

    Fan-Bill Cheung; Joy L. Rempe

    2004-06-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2010-09-22T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2000-01-13T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2000-07-01T23:59:59.000Z

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

  2. High pressure xenon ionization detector

    DOE Patents [OSTI]

    Markey, John K. (New Haven, CT)

    1989-01-01T23:59:59.000Z

    A method is provided for detecting ionization comprising allowing particles that cause ionization to contact high pressure xenon maintained at or near its critical point and measuring the amount of ionization. An apparatus is provided for detecting ionization, the apparatus comprising a vessel containing a ionizable medium, the vessel having an inlet to allow high pressure ionizable medium to enter the vessel, a means to permit particles that cause ionization of the medium to enter the vessel, an anode, a cathode, a grid and a plurality of annular field shaping rings, the field shaping rings being electrically isolated from one another, the anode, cathode, grid and field shaping rings being electrically isolated from one another in order to form an electric field between the cathode and the anode, the electric field originating at the anode and terminating at the cathode, the grid being disposed between the cathode and the anode, the field shaping rings being disposed between the cathode and the grid, the improvement comprising the medium being xenon and the vessel being maintained at a pressure of 50 to 70 atmospheres and a temperature of 0.degree. to 30.degree. C.

  3. Pressurized security barrier and alarm system

    DOE Patents [OSTI]

    Carver, D.W.

    1995-04-11T23:59:59.000Z

    A security barrier for placement across a passageway is made up of interconnected pressurized tubing made up in a grid pattern with openings too small to allow passage. The tubing is connected to a pressure switch, located away from the barrier site, which activates an alarm upon occurrence of a pressure drop. A reinforcing bar is located inside and along the length of the tubing so as to cause the tubing to rupture and set off the alarm upon an intruder`s making an attempt to crimp and seal off a portion of the tubing by application of a hydraulic tool. Radial and rectangular grid patterns are disclosed. 7 figures.

  4. Pressurized security barrier and alarm system

    DOE Patents [OSTI]

    Carver, Don W. (Knoxville, TN)

    1995-01-01T23:59:59.000Z

    A security barrier for placement across a passageway is made up of interconnected pressurized tubing made up in a grid pattern with openings too small to allow passage. The tubing is connected to a pressure switch, located away from the barrier site, which activates an alarm upon occurrence of a pressure drop. A reinforcing bar is located inside and along the length of the tubing so as to cause the tubing to rupture and set off the alarm upon an intruder's making an attempt to crimp and seal off a portion of the tubing by application of a hydraulic tool. Radial and rectangular grid patterns are disclosed.

  5. Dynamic response of containment vessels to blast loading

    SciTech Connect (OSTI)

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

    1982-01-01T23:59:59.000Z

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

  6. Corner heating in rectangular solid oxide electrochemical cell generators

    DOE Patents [OSTI]

    Reichner, Philip (Plum Boro, PA)

    1989-01-01T23:59:59.000Z

    Disclosed is an improvement in a solid oxide electrochemical cell generator 1 having a rectangular design with four sides that meet at corners, and containing multiplicity of electrically connected fuel cells 11, where a fuel gas is passed over one side of said cells and an oxygen containing gas is passed into said cells, and said fuel is burned to form heat, electricity, and an exhaust gas. The improvement comprises passing the exhaust gases over the multiplicity of cells 11 in such a way that more of the heat in said exhaust gases flows at the corners of the generator, such as through channels 19.

  7. The Gross-Pitaevskii hierarchy on general rectangular tori

    E-Print Network [OSTI]

    Sebastian Herr; Vedran Sohinger

    2014-10-20T23:59:59.000Z

    In this work, we study the Gross-Pitaevskii hierarchy on general --rational and irrational-- rectangular tori of dimension two and three. This is a system of infinitely many linear partial differential equations which arises in the rigorous derivation of the nonlinear Schr\\"{o}dinger equation. We prove a conditional uniqueness result for the hierarchy. In two dimensions, this result allows us to obtain a rigorous derivation of the defocusing cubic nonlinear Schr\\"{o}dinger equation from the dynamics of many-body quantum systems. On irrational tori, this question was posed as an open problem in previous work of Kirkpatrick, Schlein, and Staffilani.

  8. DEVELOPING FLOW AND HEAT TRANSFER IN STRONGLY CURVED DUCTS OF RECTANGULAR CROSS-SECTION

    E-Print Network [OSTI]

    Yee, G.

    2010-01-01T23:59:59.000Z

    DEVELOpiNG FLOW AND HEAT TRANSFER IN STRONGLY CURVED DUCTS9092 Developing Flow and Heat Transfer in Strongly CurvedForced Convection Heat Transfer in Curved Rectangular

  9. Measurement strategy for rectangular electrical capacitance tomography sensor

    SciTech Connect (OSTI)

    Ye, Jiamin; Ge, Ruihuan; Qiu, Guizhi; Wang, Haigang [Institute of Engineering Thermophysics, Chinese Academy of Sciences, Beijing, 100190 (China)

    2014-04-11T23:59:59.000Z

    To investigate the influence of the measurement strategy for the rectangular electrical capacitance tomography (ECT) sensor, a Finite Element Method (FEM) is utilized to create the model for simulation. The simulation was carried out using COMSOL Multiphysics(trade mark, serif) and Matlab(trade mark, serif). The length-width ratio of the rectangular sensing area is 5. Twelve electrodes are evenly arranged surrounding the pipe. The covering ratio of the electrodes is 90%. The capacitances between different electrode pairs are calculated for a bar distribution. The air of the relative permittivity 1.0 and the material of the permittivity 3.0 are used for the calibration. The relative permittivity of the second phase is 3.0. The noise free and noise data are used for the image reconstruction using the Linear Back Projection (LBP). The measurement strategies with 1-, 2- and 4- electrode excitation are compared using the correlation coefficient. Preliminary results show that the measurement strategy with 2-electrode excitation outperforms other measurement strategies with 1- or 4-electrode excitation.

  10. 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...

  11. International Hydrogen Fuel and Pressure Vessel Forum 2010 Proceedings

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

    Hythane can be odorized, and, unlike hydrogen, its flame is not invisible in daylight. Mr. Lynch noted that worldwide most CNG cylinders are Type 1 (metal only), and,...

  12. 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""...

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

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

    and Progress in Research, Development and Demonstration of Hydrogen - Compressed Natural Gas Vehicles in China Professor Z.Q. Mao Tsinghua University and Chair of the China...

  14. Digital material skins : for reversible reusable pressure vessels

    E-Print Network [OSTI]

    Hovsepian, Sarah

    2012-01-01T23:59:59.000Z

    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 ...

  15. 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 DataDepartment of Energy Your Density Isn't YourTransport inEnergy0.pdf Flash2010-60.pdf2 DOE HydrogenPlans |Former WorkerFortDepartment

  16. 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 DataDepartment of Energy Your Density Isn't YourTransport(FactDepartment of EnergyIndustry15Among Statesfor aInternational

  17. 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 DataDepartment of Energy Your Density Isn't YourTransport(FactDepartment of EnergyIndustry15Among Statesfor aInternationalDepartment of

  18. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn'tOrigin ofEnergy at Waste-to-Energy usingofRetrofittingFundA l i c e L i p p e rDepartmentand

  19. 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: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOEThe Bonneville PowerCherries 82981-1cnHigh SchoolIn12electron 9 5Let us countLighting Sign InMilitarystorage -

  20. 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 DataDepartment of Energy Your Density Isn't Your Destiny: Theof"Wave theJuly 30,Crafty Gifts for theof EnergyRev.Hydrogen

  1. 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: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn April 23, 2014, an OHASeptember 2010In addition to 1National Broadband PlanDr. James

  2. 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: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn April 23, 2014, an OHASeptember 2010In addition to 1National Broadband PlanDr.

  3. 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: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn April 23, 2014, an OHASeptember 2010In addition to 1National Broadband PlanDr.Milestone

  4. 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 onYouTube YouTube Note: Since the.pdfBreaking ofOil & Gas » Methane Hydrate » InternationalEnergy Hydrogen Fuel

  5. 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 DataDepartment of Energy Your Density Isn't YourTransport(Fact Sheet), GeothermalGridHYDROGEND D e e p p a a rDepartment| Departmenta d e

  6. Boiling heat transfer with three fluids in small circular and rectangular channels

    SciTech Connect (OSTI)

    Tran, T.N.; Wambsganss, M.W. [Argonne National Lab., IL (United States); France, D.M. [Illinois Univ., Chicago, IL (United States). Dept. of Mechanical Engineering

    1995-01-01T23:59:59.000Z

    Small circular and noncircular channels are representative of flow passages act evaporators and condensers. This report describes results of an ental study on heat transfer to the flow boiling of refrigerants (R-12) and refrigerant-134a (R-134a) in a small horizontal circular-cross-section tube. The tube diameter of 2.46 mm was chosen to approximate the hydraulic diameter of a 4.06 {times} 1.70 mm rectangular channel previously studied with R-12, and a 2.92-mm-diameter circular tube previously studied with R-113. The objective of this study was to assess the effects of channel geometry and fluid properties on the heat transfer coefficient and to obtain additional insights relative to the heat transfer mechanism(s). The current circular flow channel for the R-12 and R-134a tests was made of brass and had an overall length of 0.9 in. The channel wall was electrically heated, and thermocouples were installed on the channel wall and in the bulk fluid stream. Voltage taps were located at the same axial locations as the stream thermocouples to allow testing over an exit quality range to 0.94 and a large range of mass flux (58 to 832 kg/m{sup 2}s) and heat flux (3.6 to 59 kW/m{sup 2}). Saturation pressure was nearly constant, averaging 0.82 MPa for most of the testing, with some tests performed at a lower pressure of 0.4--0.5 MPa. Local heat transfer coefficients were determined experimentally as a function of quality along the length of the test section. Analysis of all data for three tubes and three fluids supported the conclusion that a nucleation mechanism dominates for flow boiling in small channels. Nevertheless, a convection-dominant region was obtained experimentally in this study at very low values of wall superheat (< {approx} 2.75{degrees}C). The circular and rectangular tube data for three fluids were successfully correlated in the nucleation-dominant region.

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

    SciTech Connect (OSTI)

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

    2013-07-01T23:59:59.000Z

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

  8. Light Sources on the Nylon Vessels' Surfaces

    E-Print Network [OSTI]

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

  9. Study of instabilities and quasi-two-dimensional turbulence in volumetrically heated magnetohydrodynamic flows in a vertical rectangular duct

    E-Print Network [OSTI]

    Abdou, Mohamed

    magnetohydrodynamic flows in a vertical rectangular duct N. Vetcha, S. Smolentsev, M. Abdou, and R. Moreau Citation in a vertical rectangular duct N. Vetcha,1 S. Smolentsev,1,a) M. Abdou,1 and R. Moreau2 1 Mechanical

  10. Pressurized reactor system and a method of operating the same

    DOE Patents [OSTI]

    Isaksson, J.M.

    1996-06-18T23:59:59.000Z

    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.

  11. Pressurized reactor system and a method of operating the same

    DOE Patents [OSTI]

    Isaksson, Juhani M. (Karhula, FI)

    1996-01-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    1986-01-01T23:59:59.000Z

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

  13. Pressure relief valve/safety relief valve testing

    SciTech Connect (OSTI)

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

    1994-02-01T23:59:59.000Z

    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.

  14. Scaling regimes of a semi-flexible polymer in a rectangular channel

    E-Print Network [OSTI]

    E. Werner; B. Mehlig

    2015-03-06T23:59:59.000Z

    We derive scaling relations for the extension statistics and the confinement free energy for a semi-flexible polymer confined to a channel with a rectangular cross-section. Our motivation are recent numerical results [Gupta {\\em et al.}, JCP {\\bf 140} (2014) 214901] indicating that extensional fluctuations are quite different in rectangular channels compared to square channels. Our results are of direct relevance for interpreting current experiments on DNA molecules confined to nano-channels, as many experiments are performed for rectangular channels with large aspect ratios while theoretical and simulation results are usually obtained for square channels.

  15. Tow Vessel | Open Energy Information

    Open Energy Info (EERE)

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

  16. 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.

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

    SciTech Connect (OSTI)

    Porter, V.L.

    1993-12-31T23:59:59.000Z

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

  18. Diffraction of surface wave on conducting rectangular wedge

    E-Print Network [OSTI]

    Igor A. Kotelnikov; Vasily V. Gerasimov; Boris A. Knyazev

    2013-01-16T23:59:59.000Z

    Diffraction of a surface wave on a rectangular wedge with impedance faces is studied using the Sommerfeld-Malyuzhinets technique. An analog of Landau's bypass rule in the theory of plasma waves is introduced for selection of a correct branch of the Sommerfeld integral, and the exact solution is given in terms of imaginary error function. The formula derived is valid both in the near-field and far-wave zones. It is shown that a diffracted surface wave is completely scattered into freely propagating electromagnetic waves and neither reflected nor transmitted surface waves are generated in case of bare metals which have positive real part of surface impedance. The scattered waves propagate predominantly at a grazing angle along the direction of propagation of the incident surface wave and mainly in the upper hemisphere regarding the wedge face. The profile of radiated intensity is nonmonotonic and does not resemble the surface wave profile which exponentially evanesces with the distance from the wedge face. Comparison with experiments carried out in the terahertz spectral range at Novosibirsk free electron laser has shown a good agreement of the theory and the experiments.

  19. Experimental Two-Phase Flow Characterization of Subcooled Boiling in a Rectangular Channel

    E-Print Network [OSTI]

    Estrada Perez, Carlos E.

    2010-01-16T23:59:59.000Z

    On the efforts to provide a reliable source of experimental information on turbulent subcooled boiling ow, time resolved Particle Tracking Velocimetry (PTV) experiments were carried out using HFE-301 refrigerant ow through a vertical rectangular...

  20. Two-Layer Error Control Codes Combining Rectangular and Hamming Product Codes for Cache Error

    E-Print Network [OSTI]

    Zhang, Meilin

    We propose a novel two-layer error control code, combining error detection capability of rectangular codes and error correction capability of Hamming product codes in an efficient way, in order to increase cache error ...

  1. Jet impingement heat transfer in two-pass rotating rectangular channels

    E-Print Network [OSTI]

    Zhang, Yuming

    1996-01-01T23:59:59.000Z

    The combined effects of rotation and jet impingement on local heat transfer in a two-pass rotating rectangular channel is studied. The results of an experimental investigation on the surface heat transfer coefficients under a perforated plate...

  2. RECTANGULAR POLYOMINO SET WEAK (1,2)-ACHIEVEMENT GAMES EDGAR FISHER AND NNDOR SIEBEN

    E-Print Network [OSTI]

    Sieben, Nándor

    RECTANGULAR POLYOMINO SET WEAK (1,2)-ACHIEVEMENT GAMES EDGAR FISHER AND NÁNDOR SIEBEN Abstract Classi#28;cation. 05B50, 91A46. Key words and phrases. achievement games, polyomino. 1 #12; 2 EDGAR

  3. Degree of mixing downstream of rectangular bends and design of an inlet for ambient aerosol

    E-Print Network [OSTI]

    Seo, Youngjin

    2006-04-12T23:59:59.000Z

    Tests were conducted to characterize mixing in a square and a rectangular duct with respect to suitability for single point sampling of contaminants. Several configurations, such as a straight duct with unidirectional flow at the entrance section...

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

    DOE Patents [OSTI]

    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

    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.

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

    DOE Patents [OSTI]

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

    1996-11-12T23:59:59.000Z

    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.

  6. Thermal vibration of a rectangular single-layered graphene sheet with quantum effects

    SciTech Connect (OSTI)

    Wang, Lifeng, E-mail: walfe@nuaa.edu.cn; Hu, Haiyan [State Key Laboratory of Mechanics and Control of Mechanical Structures, Nanjing University of Aeronautics and Astronautics, 210016 Nanjing (China)

    2014-06-21T23:59:59.000Z

    The thermal vibration of a rectangular single-layered graphene sheet is investigated by using a rectangular nonlocal elastic plate model with quantum effects taken into account when the law of energy equipartition is unreliable. The relation between the temperature and the Root of Mean Squared (RMS) amplitude of vibration at any point of the rectangular single-layered graphene sheet in simply supported case is derived first from the rectangular nonlocal elastic plate model with the strain gradient of the second order taken into consideration so as to characterize the effect of microstructure of the graphene sheet. Then, the RMS amplitude of thermal vibration of a rectangular single-layered graphene sheet simply supported on an elastic foundation is derived. The study shows that the RMS amplitude of the rectangular single-layered graphene sheet predicted from the quantum theory is lower than that predicted from the law of energy equipartition. The maximal relative difference of RMS amplitude of thermal vibration appears at the sheet corners. The microstructure of the graphene sheet has a little effect on the thermal vibrations of lower modes, but exhibits an obvious effect on the thermal vibrations of higher modes. The quantum effect is more important for the thermal vibration of higher modes in the case of smaller sides and lower temperature. The relative difference of maximal RMS amplitude of thermal vibration of a rectangular single-layered graphene sheet decreases monotonically with an increase of temperature. The absolute difference of maximal RMS amplitude of thermal vibration of a rectangular single-layered graphene sheet increases slowly with the rising of Winkler foundation modulus.

  7. Stress analysis, by photoelastic methods, of a rectangular bar subjected to a torsion load

    E-Print Network [OSTI]

    Rajagopalan, Tiruchengode Chinnappan

    1961-01-01T23:59:59.000Z

    STRESS ANALYSIS BY PHOTOELASTIC METHODS OF A RECTANGULAR BAR SUBJECTED TO A TORSION LOAD A Thesis by T. C. RAJAGOPALAN Submitted to the Graduate School of the Agricultural and Mechanical College of Texas in partial fulfillment... of the requirements for the degree of MASTER OF SCIENCE January 1961 Major Subject: Mechanical Engineering STRESS ANALYSIS, BY PHOTOELASTIC METHODS, OF A RECTANGULAR BAR SUBJECTED TO A TORSION LOAD A Thesis by T. C. RAJAGOPALAN Approved as to style...

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

    SciTech Connect (OSTI)

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

    1996-12-31T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2012-07-01T23:59:59.000Z

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

  10. Pressure Safety Program Implementation at ORNL

    SciTech Connect (OSTI)

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

    2013-01-01T23:59:59.000Z

    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.

  11. Fabrication of Separator Demonstration Facility process vessel

    SciTech Connect (OSTI)

    Oberst, E.F.

    1985-01-15T23:59:59.000Z

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

  12. Foam vessel for cryogenic fluid storage

    DOE Patents [OSTI]

    Spear, Jonathan D (San Francisco, CA)

    2011-07-05T23:59:59.000Z

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

  13. 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

  14. Pressurized subsampling system for pressured gas-hydrate-bearing sediment: Microscale imaging using X-ray computed tomography

    SciTech Connect (OSTI)

    Jin, Yusuke, E-mail: u-jin@aist.go.jp; Konno, Yoshihiro; Nagao, Jiro [Production Technology Team, Methane Hydrate Research Center, National Institute of Advanced Industrial Science and Technology (AIST), Sapporo 062-8517 (Japan)

    2014-09-01T23:59:59.000Z

    A pressurized subsampling system was developed for pressured gas hydrate (GH)-bearing sediments, which have been stored under pressure. The system subsamples small amounts of GH sediments from cores (approximately 50 mm in diameter and 300 mm in height) without pressure release to atmospheric conditions. The maximum size of the subsamples is 12.5 mm in diameter and 20 mm in height. Moreover, our system transfers the subsample into a pressure vessel, and seals the pressure vessel by screwing in a plug under hydraulic pressure conditions. In this study, we demonstrated pressurized subsampling from artificial xenon-hydrate sediments and nondestructive microscale imaging of the subsample, using a microfocus X-ray computed tomography (CT) system. In addition, we estimated porosity and hydrate saturation from two-dimensional X-ray CT images of the subsamples.

  15. Nuclear reactor vessel fuel thermal insulating barrier

    DOE Patents [OSTI]

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

    2013-03-19T23:59:59.000Z

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

  16. Power Generation by Pressure-Driven Transport of Ions in Nanofluidic

    E-Print Network [OSTI]

    Dekker, Cees

    Power Generation by Pressure-Driven Transport of Ions in Nanofluidic Channels Frank H. J. van der in the context of micro- and nanofluidic devices,2-7 whose geometries and material properties can be engineered in individual rectangular nanofluidic channels whose geometries remain well-defined down to the regime where

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

    SciTech Connect (OSTI)

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

    2010-07-01T23:59:59.000Z

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

  18. Thermal wake/vessel detection technique

    DOE Patents [OSTI]

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

    2012-01-10T23:59:59.000Z

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

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

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

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

  20. Automatic Lung Vessel Segmentation via Stacked Multiscale Feature Learning

    E-Print Network [OSTI]

    Toronto, University of

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

  1. Future characteristics of Offshore Support Vessels

    E-Print Network [OSTI]

    Rose, Robin Sebastian Koske

    2011-01-01T23:59:59.000Z

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

  2. Pressure safety program Lawrence Livermore National Laboratory

    SciTech Connect (OSTI)

    Borzileri, C.; Traini, M.

    1992-10-01T23:59:59.000Z

    The Lawrence Livermore National Laboratory (LLNL) is a Research and Development facility. Programs include research in: nuclear weapons, energy, environmental, biomedical, and other DOE funded programs. LLNL is managed by the University of California for the Department of Energy. Many research and development programs require the use of pressurized fluid systems. In the early 1960`s, courses were developed to train personnel to safely work with pressurized systems. These courses served as a foundation for the Pressure Safety Program. The Pressure Safety Program is administered by the Pressure Safety Manager through the Hazards Control Department, and responsibilities include: (1) Pressure Safety course development and training, (2) Equipment documentation, tracking and inspections/retests, (3) Formal and informal review of pressure systems. The program uses accepted codes and standards and closely follows the DOE Pressure Safety Guidelines Manual. This manual was developed for DOE by Lawrence Livermore National Laboratory. The DOE Pressure Safety Guidelines Manual defines five (5) basic elements which constitute this Pressure Safety Program. These elements are: (1) A Pressure Safety Manual, (2) A Safety Committee, (3) Personnel who are trained and qualified, (4) Documentation and accountability for each pressure vessel or system, (5) Control of the selection and the use of high pressure hardware.

  3. Consolidated solutions to rectangular air duct design by graphical methods

    E-Print Network [OSTI]

    Otts, John Graves

    1953-01-01T23:59:59.000Z

    . It is the purpose of this paper to study the present buo de- sign charts an4 te present sam other ~ of solving tbe seas data in less coaplieated consolidated instrueents. PROCR?gQ LLvv ~ prccoihne in 4esigaiag ~ duct systea is ss follows& l. Study tb? buil41ag... to the outLets. With this infoniistioa it is possible to calculate the duct nines, deter?inc the friction Loss of each section, and. obtain the total pressure loss in the systea. yriction thod. The initial velocity in ths duct at the fan is deterainsd...

  4. Pressurized melt ejection into scaled reactor cavities

    SciTech Connect (OSTI)

    Tarbell, W.W.; Pilch, M.; Brockmann, J.E.; Ross, J.W.; Gilbert, D.W.

    1986-10-01T23:59:59.000Z

    This report describes four tests performed in the High-Pressure Melt Streaming Program (HIPS) using linear-scaled cavities of the Zion Nuclear Power Plant. These experiments were conducted to study the phenomena involved in high-pressure ejection of core debris into the cavity beneath the reactor pressure vessel. One-tenth and one-twentieth linear scale models of reactor cavities were constructed and instrumented. The first test used an apparatus constructed of alumina firebrick to minimize the potential interaction between the ejected melt and cavity material. The remaining three experiments used scaled representations of the Zion nuclear plant geometry, constructed of prototypic concrete composition.

  5. Upflow bioreactor with septum and pressure release mechanism

    DOE Patents [OSTI]

    Hansen, Conly L.; Hansen, Carl S.; Pack, Kevin; Milligan, John; Benefiel, Bradley C.; Tolman, C. Wayne; Tolman, Kenneth W.

    2010-04-20T23:59:59.000Z

    An upflow bioreactor includes a vessel having an inlet and an outlet configured for upflow operation. A septum is positioned within the vessel and defines a lower chamber and an upper chamber. The septum includes an aperture that provides fluid communication between the upper chamber and lower chamber. The bioreactor also includes means for releasing pressure buildup in the lower chamber. In one configuration, the septum includes a releasable portion having an open position and a closed position. The releasable portion is configured to move to the open position in response to pressure buildup in the lower chamber. In the open position fluid communication between the lower chamber and the upper chamber is increased. Alternatively the lower chamber can include a pressure release line that is selectively actuated by pressure buildup. The pressure release mechanism can prevent the bioreactor from plugging and/or prevent catastrophic damage to the bioreactor caused by high pressures.

  6. THERMO-HYDRODYNAMICS OF DEVELOPING FLOW IN A RECTANGULAR MINI-CHANNEL ARRAY Gaurav Agarwal

    E-Print Network [OSTI]

    Khandekar, Sameer

    THERMO-HYDRODYNAMICS OF DEVELOPING FLOW IN A RECTANGULAR MINI-CHANNEL ARRAY Gaurav Agarwal Dept of Technology Kanpur Kanpur (UP) 208016, India samkhan@iitk.ac.in ABSTRACT Thermo-hydrodynamic performance on developing flows. Thus, the study reveals that conventional theory, which predicts thermo

  7. EVALUATION OF THE RECTANGULAR RAPID FLASH BEACON AT A PINELLAS TRAIL CROSSING IN ST. PETERSBURG, FLORIDA

    E-Print Network [OSTI]

    North Carolina at Chapel Hill, University of

    rectangular yellow LED indicators which flash rapidly in a wig-wag sequence. It is solar-powered, radio controlled, and activated by trail users. The experimental design was to collect data of trail users before of more benefit would be periodic police enforcement operations, or the development of a passive system

  8. Helicity invariants of force-free field for a rectangular box

    E-Print Network [OSTI]

    Rudenko, G V

    2010-01-01T23:59:59.000Z

    An algorithm for calculating three gauge-invariant helicities (self-, mutual- and Berger relative helicity) for a magnetic field specified in a rectangular box is described. The algorithm is tested on a well-known force-free model (Low and Lou, 1990) presented in vector-potential form.

  9. RECTANGULAR POLYOMINO SET WEAK (1,2)-ACHIEVEMENT GAMES EDGAR FISHER AND NNDOR SIEBEN

    E-Print Network [OSTI]

    Sieben, Nándor

    RECTANGULAR POLYOMINO SET WEAK (1,2)-ACHIEVEMENT GAMES EDGAR FISHER AND NÁNDOR SIEBEN Abstract. achievement games, polyomino. #12;2 EDGAR FISHER AND NÁNDOR SIEBEN P1,1 P2,1 P3,1 P3,2 P4,1 P4,2 P4,3 P4,4 P4

  10. A Quadrature Finite Element Galerkin Scheme for a Biharmonic Problem on a Rectangular Polygon

    E-Print Network [OSTI]

    Aitbayev, Rakhim

    to the OSC method (see [3]), which uses Gaussian quadrature nodes as collocation points and a finite elementA Quadrature Finite Element Galerkin Scheme for a Biharmonic Problem on a Rectangular Polygon with an underdetermined orthogonal spline collocation scheme. © 2007 Wiley Periodicals, Inc. Numer Methods Partial

  11. Recharge/seepage from an array of rectangular Mahender Choudhary a

    E-Print Network [OSTI]

    Chahar, B. R.

    Recharge/seepage from an array of rectangular channels Mahender Choudhary a , Bhagu R. Chahar b@civil.iitd.ac.in, chahar_br@yahoo.com (B.R. Chahar). Journal of Hydrology (2007) 343, 71­79 available at www

  12. Flow past a rectangular cylinder in a stratified fluid Harish Dixit, A Sameen 1

    E-Print Network [OSTI]

    Dixit, Harish

    ); Zdravkovich (1996); Matsumoto (1999). When the Reynolds number, Re exceeds a critical value, a Karman vortex street associated with a periodic shedding appears. Traditionally, studies on bluff body wakes have been the Strouhal numbers for vortex shedding behind rectangular cylinders. Davis et al. (1984) carried out

  13. Pressure Safety of JLAB 12GeV Upgrade Cryomodule

    SciTech Connect (OSTI)

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

    2009-11-01T23:59:59.000Z

    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.

  14. Effects of external pressure on the terminal lymphatic flow rate

    E-Print Network [OSTI]

    Seale, James Lewis

    1981-01-01T23:59:59.000Z

    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...

  15. A multiresolution finite element method based on a new locking-free rectangular Mindlin plate element

    E-Print Network [OSTI]

    Xia, Yi-Ming

    2015-01-01T23:59:59.000Z

    A locking-free rectangular Mindlin plate element with a new multi-resolution analysis (MRA) is proposed and a multireolution finite element method is hence presented. The MRA framework is formulated out of a mutually nesting displacement subspace sequence. The MRA endows the proposed element with the resolution level (RL) to adjust the element node number, thus modulating structural analysis accuracy accordingly. As a result, the traditional 4-node rectangular Mindlin plate element and method is a mono-resolution one and also a special case of the proposed element and method. The meshing for the monoresolution plate element model is based on the empiricism while the RL adjusting for the multiresolution is laid on the rigorous mathematical basis. The accuracy of a structural analysis is actually determined by the RL, not by the mesh. The rational MRA enables the implementation of the multiresolution Mindlin plate element method to be more rational and efficient than that of the conventional monoresolution or o...

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

    E-Print Network [OSTI]

    Miller, John V. (John Vaala)

    2007-01-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

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

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

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

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

  19. Structural loading of cross deck connections for trimaran vessels

    E-Print Network [OSTI]

    Rhoads, Jason L

    2004-01-01T23:59:59.000Z

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

  20. Two Channel Dielectric-Lined Rectangular High Transformer Ratio Accelerator Structure Experiment

    SciTech Connect (OSTI)

    Shchelkunov, S. V.; LaPointe, M. A. [Beam Physics Laboratory, Yale University, 272 Whitney Avenue, New Haven, CT 06511 (United States); Hirshfield, J. L. [Beam Physics Laboratory, Yale University, 272 Whitney Avenue, New Haven, CT 06511 (United States); Omega-P, Inc., 258 Bradley St., New Haven, CT 06510 (United States); Marshall, T. C. [Columbia University, New York, NY 10027 (United States); Omega-P, Inc., 258 Bradley St., New Haven, CT 06510 (United States); Sotnikov, G. [NSC Kharkov Institute of Physics and Technology, Kharkov (Ukraine); Omega-P, Inc., 258 Bradley St., New Haven, CT 06510 (United States); Gai, Wei; Conde, M.; Power, J.; Mihalcea, D. [Argonne National Laboratory, Argonne, IL 60439 (United States)

    2010-11-04T23:59:59.000Z

    Current status of a two-channel cm-scale rectangular dielectric lined wakefield accelerator structure is described. This structure is installed at the Argonne Wakefield Accelerator facility (AWA), and is presently being evaluated. The device has a transformer ratio of {approx}12.5:1. When driven by a {approx}50 nC single drive bunch it is expected to obtain {approx}6 MV/m acceleration gradient. Related issues are discussed.

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

    DOE Patents [OSTI]

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

    2013-06-04T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Cohen, Laurent

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

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

    E-Print Network [OSTI]

    Leow, Wee Kheng

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

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

    SciTech Connect (OSTI)

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

    1989-11-01T23:59:59.000Z

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

  5. Treating exhaust gas from a pressurized fluidized bed reaction system

    DOE Patents [OSTI]

    Isaksson, J.; Koskinen, J.

    1995-08-22T23:59:59.000Z

    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.

  6. From Cold War to cold vessels

    SciTech Connect (OSTI)

    Melrath, C.

    1996-09-01T23:59:59.000Z

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

  7. Photoacoustic removal of occlusions from blood vessels

    DOE Patents [OSTI]

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

    2002-01-01T23:59:59.000Z

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

  8. Pressurized fluidized bed reactor and a method of operating the same

    DOE Patents [OSTI]

    Isaksson, Juhani (Karhula, FI)

    1996-01-01T23:59:59.000Z

    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.

  9. Pressurized fluidized bed reactor and a method of operating the same

    DOE Patents [OSTI]

    Isaksson, J.

    1996-02-20T23:59:59.000Z

    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.

  10. Confinement Vessel Assay System: Calibration and Certification Report

    SciTech Connect (OSTI)

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

    2012-07-17T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    J. C. Misra; S. Maiti

    2012-01-30T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Raske, D.T.

    1995-06-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1983-01-01T23:59:59.000Z

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

  14. Design of a high-pressure research flow loop for the experimental investigation of liquid loading in gas wells 

    E-Print Network [OSTI]

    Fernandez Alvarez, Juan Jose

    2009-05-15T23:59:59.000Z

    body with top and bottom welded-flat head covers with multiple openings to minimize its weight. The pipelines connecting major equipment and injection manifold located at the pressure vessel were selected based on the superficial velocities for air...

  15. A photoelastic study of rectangular beams in bending with a hole on the vertical center line

    E-Print Network [OSTI]

    Bailey, James Fletcher

    1965-01-01T23:59:59.000Z

    . To plot the stress concentration factors in the tested beams so that they may be used in design work. 4. To compare the results of this study with those obtained by Cox (1)~, Morgan (2), and Lindsay (3) . * Numbers in parentheses refer to references... that the maximum stresses occurred at one of four points on each of the beams tested. The four points of stress concentration were at the top or bottom of the beam and of the hole. In a rectangular beam in bending, without a hole on the vertical center line...

  16. Method for forming a bladder for fluid storage vessels

    DOE Patents [OSTI]

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

    2000-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Not Available

    1984-11-01T23:59:59.000Z

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

  18. Autonomous Radiation Monitoring of Small Vessels

    SciTech Connect (OSTI)

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

    2010-01-01T23:59:59.000Z

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

  19. Finite temperature Casimir energy in closed rectangular cavities: a rigorous derivation based on zeta function technique

    E-Print Network [OSTI]

    S. C. Lim; L. P. Teo

    2008-04-24T23:59:59.000Z

    We derive rigorously explicit formulas of the Casimir free energy at finite temperature for massless scalar field and electromagnetic field confined in a closed rectangular cavity with different boundary conditions by zeta regularization method. We study both the low and high temperature expansions of the free energy. In each case, we write the free energy as a sum of a polynomial in temperature plus exponentially decay terms. We show that the free energy is always a decreasing function of temperature. In the cases of massless scalar field with Dirichlet boundary condition and electromagnetic field, the zero temperature Casimir free energy might be positive. In each of these cases, there is a unique transition temperature (as a function of the side lengths of the cavity) where the Casimir energy change from positive to negative. When the space dimension is equal to two and three, we show graphically the dependence of this transition temperature on the side lengths of the cavity. Finally we also show that we can obtain the results for a non-closed rectangular cavity by letting the size of some directions of a closed cavity going to infinity, and we find that these results agree with the usual integration prescription adopted by other authors.

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

    DOE Patents [OSTI]

    Davin, J.

    1992-12-01T23:59:59.000Z

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

  1. 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.

  2. Method and device for supporting blood vessels during anastomosis

    DOE Patents [OSTI]

    Doss, J.D.

    1985-05-20T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    McCabe, D.E.

    1999-09-01T23:59:59.000Z

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

  4. An apparatus for studying scintillator properties at high isostatic pressures

    SciTech Connect (OSTI)

    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

    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.

  5. Aerosol source term in high pressure melt ejection

    SciTech Connect (OSTI)

    Brockmann, J.E.; Tarbell, W.W.

    1984-11-01T23:59:59.000Z

    Pressurized ejection of melt from a reactor pressure vessel has been identified as an important element of a severe reactor accident. Copious aerosol production is observed when thermitically generated melts pressurized with nitrogen or carbon dioxide to 1.3 to 17 MPa are ejected into an air atmosphere. Aerosol particle size distributions measured in the tests have modes of about 0.5, 5, and > 10 ..mu..m. Mechanisms leading to formation of these multimodal size distributions are suggested. This aerosol is a potentially important fission product source term that has not been considered in previous severe accident analyses.

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

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2008-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1995-12-01T23:59:59.000Z

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

  8. Simultaneous Irradiation and Imaging of Blood Vessels During Pulsed

    E-Print Network [OSTI]

    Barton, Jennifer K.

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

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

    SciTech Connect (OSTI)

    Not Available

    1991-10-07T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Smith, Robert L.

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

  11. Density modification by two superposing TE{sub 10} modes in a plasma filled rectangular waveguide

    SciTech Connect (OSTI)

    Tomar, Sanjay K.; Malik, Hitendra K. [Plasma Waves and Particle Acceleration Laboratory, Department of Physics, Indian Institute of Technology Delhi, New Delhi 110 016 (India)] [Plasma Waves and Particle Acceleration Laboratory, Department of Physics, Indian Institute of Technology Delhi, New Delhi 110 016 (India)

    2013-07-15T23:59:59.000Z

    Microwave and plasma interaction is examined via two fundamental TE{sub 10} modes propagating in a plasma filled rectangular waveguide after superposing at a smaller angle. The propagation of the resultant mode realized from these two modes is governed by a wave equation obtained using the Maxwell's equations. This equation is solved numerically using fourth order Runge-Kutta method for the field amplitude of the microwave in the waveguide considering the waveguide to be made up of a perfect conductor and filled with different types of initial plasma density distributions, viz. homogeneous density, linear density with gradient in the propagation direction, and the density with Gaussian profile along the waveguide width. A phenomenon similar to the duct formation by high power microwaves is found to take place, where the plasma density attains interesting profiles. These profiles can be controlled by the angle of superposition, phase difference between the fields of the two modes, microwave frequency and microwave field amplitude.

  12. Microwave whirlpools in a rectangular-waveguide cavity with a thin ferrite disk

    E-Print Network [OSTI]

    E. O. Kamenetskii; Michael Sigalov; Reuven Shavit

    2007-07-09T23:59:59.000Z

    We study a three dimensional system of a rectangular-waveguide resonator with an inserted thin ferrite disk. The interplay of reflection and transmission at the disk interfaces together with material gyrotropy effect, gives rise to a rich variety of wave phenomena. We analyze the wave propagation based on full Maxwell-equation numerical solutions of the problem. We show that the power-flow lines of the microwave-cavity field interacting with a ferrite disk, in the proximity of its ferromagnetic resonance, form whirlpool-like electromagnetic vortices. Such vortices are characterized by the dynamical symmetry breaking. The role of ohmic losses in waveguide walls and dielectric and magnetic losses in a disk is a subject of our investigations.

  13. LDV measurement and Navier-Stokes computation of parallel jet mixing in a rectangular confinement

    SciTech Connect (OSTI)

    Kunz, R.F.; D`Amico, S.W.; Vassallo, P.F.; Zaccaria, M.A. [Knolls Atomic Power Lab., Schenectady, NY (United States); Aksoy, H.; So, R.M.C. [Arizona State Univ., Tempe, AZ (United States). Dept. of Mechanical and Aerospace Engineering

    1995-06-01T23:59:59.000Z

    Laser Doppler Velocimetry (LDV) measurements were taken in a rectangular confinement into which issues a row of parallel jets. Two-component measurements were taken with two optics orientations yielding three mean velocity components and four Reynolds stress components. As observed in isolated three dimensional wall bounded jets, the transverse diffusion of the jets is quite large. The data indicates that this rapid mixing process is due to strong secondary flows, transport of large inlet intensities and Reynolds stress anisotropy effects. Navier-Stokes analyses of this configuration underpredict the rate of transverse jet diffusion. Detailed numerical accuracy studies show that this is attributed to shortcomings in low-Reynolds number two-equation turbulence modelling. A low-Reynolds number full-Reynolds stress model is shown to provide improvement.

  14. Lyapunov exponents for products of rectangular real, complex and quaternionic Ginibre matrices

    E-Print Network [OSTI]

    J. R. Ipsen

    2015-03-26T23:59:59.000Z

    We study the joint density of eigenvalues for products of independent rectangular real, complex and quaternionic Ginibre matrices. In the limit where the number of matrices tends to infinity, it is shown that the joint probability density function for the eigenvalues forms a permanental point process for all three classes. The moduli of the eigenvalues become uncorrelated and log-normal distributed, while the distribution for the phases of the eigenvalues depends on whether real, complex or quaternionic Ginibre matrices are considered. In the derivation for a product of real matrices, we explicitly use the fact that all eigenvalues become real when the number of matrices tends to infinity. Finally, we compare our results with known results for the Lyapunov exponents as well as numerical simulations.

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

    SciTech Connect (OSTI)

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

    2010-07-01T23:59:59.000Z

    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.

  16. Welding the AT-400A Containment Vessel

    SciTech Connect (OSTI)

    Brandon, E.

    1998-11-01T23:59:59.000Z

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

  17. Numerical simulation of large amplitude liquid sloshing in a rigid rectangular tank

    E-Print Network [OSTI]

    Bridges, Thomas J.

    1981-01-01T23:59:59.000Z

    oscillations, harbor oscillations, tank trucks on highways, liquid fuel in space craft, and sloshing of liquid cargo in oceangoing vessels. Throughout recent history, investigators have used various methods to mathematically represent. liquid sloshing... loads in cargo tanks is not restricted to LNG carriers since similar problems have been experienced in other types of liquid transport ships such as bulk oil carriers. However, several factors make slosh loads more important with regard to LNG ship...

  18. High Temperature Electrolysis Pressurized Experiment Design, Operation, and Results

    SciTech Connect (OSTI)

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

    2012-09-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Blanton, P.; Eberl, K.

    2010-07-09T23:59:59.000Z

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

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

    DOE Patents [OSTI]

    Powell, J.G.

    1993-11-16T23:59:59.000Z

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

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

    DOE Patents [OSTI]

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

    2011-01-25T23:59:59.000Z

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

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

    DOE Patents [OSTI]

    Powell, James G. (Clifton Park, NY)

    1993-01-01T23:59:59.000Z

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

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

    DOE Patents [OSTI]

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

    2012-09-25T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    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

    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.

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

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

    September 27 - 29, 2010, in Beijing, China. ihfpvlynch.pdf More Documents & Publications Properties, Behavior and Material Compatibility of Hydrogen, Natural Gas and Blends -...

  6. 1 Copyright 2012 by ASME Proceedings of the ASME 2012 Pressure Vessels & Piping Division Conference

    E-Print Network [OSTI]

    Giurgiutiu, Victor

    requirements for monitoring and inspection of dry storage systems as part of aging management plans. FIGURE 1.S. Department of Energy (DOE) as a high priority cross-cutting need. Monitoring is necessary to determine: DRY CASK STORAGE SYSTEM FOR SPENT NUCLEAR FUEL Structural health monitoring offers the solution

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

    E-Print Network [OSTI]

    Award Number DE-FC36-04GO14229 · Partners ­ Savannah River National Laboratory (SRNL) ­ Hy volunteered to coordinate the RRT with participating laboratories and organizations; plan endorsed by PWG

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

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

    Workshop on Critical Property Needs American Society of Mechanical EngineersSavannah River National Laboratory (ASMESRNL) Materials and Components for Hydrogen Infrastructure...

  9. Pipeline and Pressure Vessel R&D under the Hydrogen Regional Infrastructure Program In Pennsylvania

    Broader source: Energy.gov [DOE]

    Began as rapid data generator for numerical modeling efforts, Cathodic charging applied to base, Heat Affected Zone (HAZ) and weld metal

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

    E-Print Network [OSTI]

    Paulino, Glaucio H.

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

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

    SciTech Connect (OSTI)

    Yongfeng Zhang; Pritam Chakraborty; S. Bulent Biner

    2013-09-01T23:59:59.000Z

    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.

  12. Proceedings of PVP2007 2007 ASME Pressure Vessels and Piping Division Conference

    E-Print Network [OSTI]

    Cambridge, University of

    .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

  13. Rigorous Simulation of Accidental Leaks from High-Pressure Storage Vessels

    E-Print Network [OSTI]

    Alisha, -

    2014-07-07T23:59:59.000Z

    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...

  14. Systems Engineering of Chemical Hydrogen Storage, Pressure Vessel and Balance of Plant for Onboard Hydrogen Storage

    SciTech Connect (OSTI)

    Brooks, Kriston P.; Simmons, Kevin L.; Weimar, Mark R.

    2014-09-02T23:59:59.000Z

    This is the annual report for the Hydrogen Storage Engineering Center of Excellence project as required by DOE EERE's Fuel Cell Technologies Office. We have been provided with a specific format. It describes the work that was done with cryo-sorbent based and chemical-based hydrogen storage materials. Balance of plant components were developed, proof-of-concept testing performed, system costs estimated, and transient models validated as part of this work.

  15. Rigorous Simulation of Accidental Leaks from High-Pressure Storage Vessels 

    E-Print Network [OSTI]

    Alisha, -

    2014-07-07T23:59:59.000Z

    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...

  16. 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 onYou are now leaving Energy.gov You are now leaving Energy.gov You are beingZealandORCEncroachment.pdf Jump to:-FD-a -NVBSundryNotice6HICDrinkingWaterPermit.pdf Jump

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

    E-Print Network [OSTI]

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

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

    E-Print Network [OSTI]

    Collette, Eric Peter

    2013-07-22T23:59:59.000Z

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

  19. DESIGN OF THE ITER IN-VESSEL COILS

    SciTech Connect (OSTI)

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

    2010-11-27T23:59:59.000Z

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

  20. 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...

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

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

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

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

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

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

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

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

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

  4. Simple program calculates partial liquid volumes in vessels

    SciTech Connect (OSTI)

    Koch, P.

    1992-04-13T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Guttenplan, Adam (Adam David)

    2007-01-01T23:59:59.000Z

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

  6. A cog-like vessel from the Netherlands

    E-Print Network [OSTI]

    Van de Moortel, Aleydis Maria P. A.

    1987-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2013-11-01T23:59:59.000Z

    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.

  8. Pressurized solid oxide fuel cell integral air accumular containment

    DOE Patents [OSTI]

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

    2004-02-10T23:59:59.000Z

    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.

  9. Effect of permeable ribs on heat transfer and friction in a rectangular channel

    SciTech Connect (OSTI)

    Hwang, J.J. [Chung-Hua Polytechnic Inst., Hsinchu (Taiwan, Province of China). Dept. of Mechanical Engineering; Liou, T.M. [National Tsing Hua Univ., Hsinchu (Taiwan, Province of China). Dept. of Power Mechanical Engineering

    1995-04-01T23:59:59.000Z

    To increase specific thrust and to reduce specific fuel consumption (SFC), high turbine entry gas temperature (1,400--1,600 C) has become the trend in advanced aero-engine design. Such a high gas temperature is far above the allowable metal temperature; therefore, turbine blades must be cooled in order to operate in the high gas temperature environment. Heat transfer and friction characteristics in a rectangular channel with perforated ribs arranged in-line on two opposite walls are investigated experimentally. Five perforated rib open-area ratios (0, 10, 22, 38, and 44%) and three rib pitch-to-height ratios (10, 15, and 20) are examined. The Reynolds number ranges from 5,000 to 50,000. The rib height-to-channel hydraulic diameter ratio and the channel aspect ratio are 0.081 and 4, respectively. Laser holographic interferometry is employed not only to measure the heat transfer coefficients of the ribbed wall but also to determine the rib apparent permeability. It is found that ribs with appropriate high open-area ratio and high Reynolds number are permeable, and the critical Reynolds number for evidence of flow permeability decreases with increasing rib open-area ratio. Results of local heat transfer coefficients further show that the permeable ribs have an advantage of obviating hot spots. Moreover, the duct with permeable ribs gives a higher thermal performance than that with solid ribs.

  10. Heat transfer and friction characteristics in rectangular channels with rib turbulators

    SciTech Connect (OSTI)

    Han. J.C. (Texas A and M Univ., College Station (USA))

    1988-05-01T23:59:59.000Z

    The effect of the channel aspect ratio on the distribution of the local heat transfer coefficient in rectangular channels with two opposite ribbed walls (to simulate turbine airfoil cooling passages) was determined for a Reynolds number range of 10,000 to 60,000. The channel width-to-height ratios (W/H, ribs on side W) were 1/4, 1/2, 1, 2, and 4. The test channels were heated by passing current through thin, stainless steel foils instrumented with thermocouples. The local heat transfer coefficients on the ribbed side wall and on the smooth side wall of each test channel from the channel entrance to the fully developed regions were measured for two rib spacings (P/e = 10 and 20). The rib angle-of-attack was kept at 90 deg. The local data in the fully developed region were averaged and correlated, based on the heat transfer and friction similarity laws developed for ribbed channels, to cover the ranges of channel aspect ratio, rib spacing, rib height, and Reynolds number. The results compare well with the published data for flow in a square channel with two opposite ribbed walls. The correlations can be used in the design of turbine airfoil cooling passages.

  11. Heat transfer augmentation in a rectangular channel with slit rib-turbulators on two opposite walls

    SciTech Connect (OSTI)

    Hwang, J.J. [Chung-Hua Polytechnic Inst., Hsinchu (Taiwan, Province of China). Dept. of Mechanical Engineering; Liou, T.M. [National Tsing-Hua Univ., Hsinchu (Taiwan, Province of China). Dept. of Power Mechanical Engineering

    1997-07-01T23:59:59.000Z

    The effect of slit ribs on heat transfer and friction in a rectangular channel is investigated experimentally. The slit ribs are arranged in-line on two opposite walls of the channel. Three rib open-area ratios ({beta} = 24, 37, and 46%), three rib pitch-to-height ratios (Pi/H = 10, 20, and 30), and two rib height-to-channel hydraulic diameter ratios (H/De = 0.081, and 0.162) are examined. The Reynolds number ranges from 10,000 to 50,000. Laser holographic interferometry is employed to measure the local heat transfer coefficients of the ribbed wall quantitatively, and observe the flow over the ribbed wall qualitatively. The results show that the slit rib has an advantage of avoiding hot spots. In addition, the heat transfer performance of the slit-ribbed channel is much better than that of the solid-ribbed channel. Semi-empirical correlations for friction and heat transfer are developed to account for rib spacings and open-area ratios. These correlations may be used in the design of turbine blade cooling passages.

  12. Natural convection heat transfer for a staggered array of heated, horizontal cylinders within a rectangular enclosure

    SciTech Connect (OSTI)

    Triplett, C.E.

    1996-12-01T23:59:59.000Z

    This thesis presents the results of an experimental investigation of natural convection heat transfer in a staggered array of heated cylinders, oriented horizontally within a rectangular enclosure. The main purpose of this research was to extend the knowledge of heat transfer within enclosed bundles of spent nuclear fuel rods sealed within a shipping or storage container. This research extends Canaan`s investigation of an aligned array of heated cylinders that thermally simulated a boiling water reactor (BWR) spent fuel assembly sealed within a shipping or storage cask. The results are presented in terms of piecewise Nusselt-Rayleigh number correlations of the form Nu = C(Ra){sup n}, where C and n are constants. Correlations are presented both for individual rods within the array and for the array as a whole. The correlations are based only on the convective component of the heat transfer. The radiative component was calculated with a finite-element code that used measured surface temperatures, rod array geometry, and measured surface emissivities as inputs. The correlation results are compared to Canaan`s aligned array results and to other studies of natural convection in horizontal tube arrays.

  13. Numerical simulation of the non-isothermal developing flow of a nonlinear viscoelastic fluid in a rectangular channel

    E-Print Network [OSTI]

    Nikoleris, Teo

    1988-01-01T23:59:59.000Z

    Fluid in a Rectangular Channel (December 1988) Teo Nikoleris, B. S. , Reed College Chairman of Advisory Committee: Dr. R. Darby An orthogonal collocation finite element program was used to numerically model the hydrodynamicslly and thermally... in negligible increase of Nw~ ~?. Also, the approach of Chang and Finlayson [6], [7] who applied orthogonal collocation finite elements in conjunction with bicubic Hermitian polynomials to approximate various viscoelastic flow problems, also met with little...

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

    SciTech Connect (OSTI)

    Vinson, Dennis

    2010-06-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2010-01-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Leow, Wee Kheng

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

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

    DOE Patents [OSTI]

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

    2005-03-29T23:59:59.000Z

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

  18. PressurePressure Indiana Coal Characteristics

    E-Print Network [OSTI]

    Fernández-Juricic, Esteban

    TimeTime PressurePressure · Indiana Coal Characteristics · Indiana Coals for Coke · CoalTransportation in Indiana · Coal Slurry Ponds Evaluation · Site Selection for Coal Gasification · Coal-To-Liquids Study, CTL · Indiana Coal Forecasting · Under-Ground Coal Gasification · Benefits of Oxyfuel Combustion · Economic

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

    SciTech Connect (OSTI)

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

    1988-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1984-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Smith, A.

    2010-02-16T23:59:59.000Z

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

  2. Prediction of turbulent flow and heat transfer in a ribbed rectangular duct with and without rotation

    SciTech Connect (OSTI)

    Prakash, C.; Zerkle, R. [General Electric Co., Cincinnati, OH (United States)

    1995-04-01T23:59:59.000Z

    The present study deals with the numerical prediction of turbulent flow and heat transfer in a 2:1 aspect ratio rectangular duct with ribs don the two shorter sides. The ribs are of square cross section, staggered and aligned normal (90 deg) to the main flow direction. The ratio of rib height to duct hydraulic diameter equals 0.063, and the ratio of rib spacing to rib height equals 10. The duct may be stationary or rotating. The axis of rotation is normal to the axis of the duct and parallel to the ribbed walls (i.e., the ribbed walls form the leading and the trailing faces). The problem is three dimensional and fully elliptic; hence, for computational economy, the present analysis deals only with a periodically fully developed situation where the calculation domain is limited to the region between two adjacent ribs. Turbulence is modeled with the {kappa}-{epsilon} model in conjunction with wall functions. However, since the rib height is small, use of wall functions necessitates that the Reynolds number be kept high. (Attempts to use a two-layer model that permits integration to the wall did not yield satisfactory results and such modeling issues are discussed at length.) Computations are made here for Reynolds number in the range 30,000--100,000 and for Rotation number = 0 (stationary), 0.06, and 0.12. For the stationary case, the predicted heat transfer agrees well with the experimental correlations. Due to the Coriolis-induced secondary flow, rotation is found to enhance heat transfer from the trailing and the side walls, while decreasing heat transfer from the leading face. Relative to the corresponding stationary case, the effect of rotation is found to be less for a ribbed channel as compared to a smooth channel.

  3. Three-dimensional gravity modeling and focusing inversion using rectangular meshes.

    SciTech Connect (OSTI)

    Commer, M.

    2011-03-01T23:59:59.000Z

    Rectangular grid cells are commonly used for the geophysical modeling of gravity anomalies, owing to their flexibility in constructing complex models. The straightforward handling of cubic cells in gravity inversion algorithms allows for a flexible imposition of model regularization constraints, which are generally essential in the inversion of static potential field data. The first part of this paper provides a review of commonly used expressions for calculating the gravity of a right polygonal prism, both for gravity and gradiometry, where the formulas of Plouff and Forsberg are adapted. The formulas can be cast into general forms practical for implementation. In the second part, a weighting scheme for resolution enhancement at depth is presented. Modelling the earth using highly digitized meshes, depth weighting schemes are typically applied to the model objective functional, subject to minimizing the data misfit. The scheme proposed here involves a non-linear conjugate gradient inversion scheme with a weighting function applied to the non-linear conjugate gradient scheme's gradient vector of the objective functional. The low depth resolution due to the quick decay of the gravity kernel functions is counteracted by suppressing the search directions in the parameter space that would lead to near-surface concentrations of gravity anomalies. Further, a density parameter transformation function enabling the imposition of lower and upper bounding constraints is employed. Using synthetic data from models of varying complexity and a field data set, it is demonstrated that, given an adequate depth weighting function, the gravity inversion in the transform space can recover geologically meaningful models requiring a minimum of prior information and user interaction.

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

    SciTech Connect (OSTI)

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

    1987-01-01T23:59:59.000Z

    The potential for stainless steel cladding to improve the fracture behavior of an operating nuclear reactor pressure vessel, particularly during certain overcooling transients, may depend greatly on the properties of the irradiated cladding. Therefore, weld overlay cladding irradiated at temperatures and to fluences relevant to power reactor operation was examined. The cladding was applied to a pressure vessel steel plate by the three-wire series-arc commercial method. Cladding was applied in three layers to provide adequate thickness for the fabrication of test specimens. The three-wire series-arc procedure, developed by Combustion Engineering, Inc., Chattanooga, Tennessee, produced a highly controlled weld chemistry, microstructure, and fracture properties in all three layers of the weld. Charpy V-notch and tensile specimens were irradiated at 288/sup 0/C to fluence levels of 2 and 5 x 10/sup 19/ neutrons/cm/sup 2/ (>1 MeV). Postirradiation testing results show that, in the test temperature range from -125 to 288/sup 0/C, the yield strength increased by 8 to 30%, ductility insignificantly increased, while there was almost no change in ultimate tensile strength. All cladding exhibited ductile-to-brittle transition behavior during Charpy impact testing, due to the dominance of delta-ferrite failures at low temperatures. On the upper shelf, energy was reduced, due to irradiation exposure, 15 and 20%, while the lateral expansion was reduced 43 and 41%, at 2 and 5 x 10/sup 19/ neutrons/cm/sup 2/ (>1 MeV), respectively. In addition, radiation damage resulted in 13 and 28/sup 0/C shifts of the Charpy impact transition temperature at the 41-J level for the low and high fluences, respectively.

  5. Understanding Blood Pressure

    E-Print Network [OSTI]

    Understanding Blood Pressure · Monitorathomewithadigitalmonitor. · Useleftarmwithcorrectsizecuff. · Avoidcaffeine,alcohol,andtobacco. Steps to Follow FOR AN ACCURATE MEASUREMENT Blood pressure is the measurement of the force of blood on the walls of the arteries. Bottom number = Diastolic (force between heart beats) Top

  6. Sterilization of fermentation vessels by ethanol/water mixtures

    DOE Patents [OSTI]

    Wyman, C.E.

    1999-02-09T23:59:59.000Z

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

  7. Sterilization of fermentation vessels by ethanol/water mixtures

    DOE Patents [OSTI]

    Wyman, Charles E. (Lakewood, CO)

    1999-02-09T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Chui, Margaret Kam-Too

    1980-01-01T23:59:59.000Z

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

  9. Pressurized Testing of Solid Oxide Electrolysis Stacks with Advanced Electrode-Supported Cells

    SciTech Connect (OSTI)

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

    2012-06-01T23:59:59.000Z

    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.

  10. A new design criterion based on pressure testing of torispherical heads

    SciTech Connect (OSTI)

    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

    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.

  11. Atmospheric Pressure Reactor System | EMSL

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

    Atmospheric Pressure Reactor System Atmospheric Pressure Reactor System The atmospheric pressure reactor system is designed for testing the efficiency of various catalysts for the...

  12. Rarefied gas flow in a rectangular enclosure induced by non-isothermal walls

    SciTech Connect (OSTI)

    Vargas, Manuel; Tatsios, Giorgos; Valougeorgis, Dimitris, E-mail: diva@mie.uth.gr [Department of Mechanical Engineering, University of Thessaly, 38334 Volos (Greece)] [Department of Mechanical Engineering, University of Thessaly, 38334 Volos (Greece); Stefanov, Stefan [Institute of Mechanics, Bulgarian Academy of Sciences, Sofia (Bulgaria)] [Institute of Mechanics, Bulgarian Academy of Sciences, Sofia (Bulgaria)

    2014-05-15T23:59:59.000Z

    The flow of a rarefied gas in a rectangular enclosure due to the non-isothermal walls with no synergetic contributions from external force fields is investigated. The top and bottom walls are maintained at constant but different temperatures and along the lateral walls a linear temperature profile is assumed. Modeling is based on the direct numerical solution of the Shakhov kinetic equation and the Direct Simulation Monte Carlo (DSMC) method. Solving the problem both deterministically and stochastically allows a systematic comparison and verification of the results as well as the exploitation of the numerical advantages of each approach in the investigation of the involved flow and heat transfer phenomena. The thermally induced flow is simulated in terms of three dimensionless parameters characterizing the problem, namely, the reference Knudsen number, the temperature ratio of the bottom over the top plates, and the enclosure aspect ratio. Their effect on the flow configuration and bulk quantities is thoroughly examined. Along the side walls, the gas flows at small Knudsen numbers from cold-to-hot, while as the Knudsen number is increased the gas flows from hot-to-cold and the thermally induced flow configuration becomes more complex. These flow patterns with the hot-to-cold flow to be extended to the whole length of the non-isothermal side walls may exist even at small temperature differences and then, they are enhanced as the temperature difference between the top and bottom plates is increased. The cavity aspect ratio also influences this flow configuration and the hot-to-cold flow is becoming more dominant as the depth compared to the width of the cavity is increased. To further analyze the flow patterns a novel solution decomposition into ballistic and collision parts is introduced. This is achieved by accordingly modifying the indexing process of the typical DSMC algorithm. The contribution of each part of the solution is separately examined and a physical interpretation of the flow configuration, including the hot-to-cold flow close to the side walls, in the whole range of the Knudsen number is provided.

  13. Inexpensive delivery of compressed hydrogen with advanced vessel technology

    E-Print Network [OSTI]

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

  14. 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

  15. Modelling the Induced Magnetic Signature of Naval Vessels

    E-Print Network [OSTI]

    Low, Robert

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

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

    E-Print Network [OSTI]

    Luo, Xiaoyu

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

  17. PublicationsmailagreementNo.40014024 the VeSSeL WILL

    E-Print Network [OSTI]

    Pedersen, Tom

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

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

    SciTech Connect (OSTI)

    Sherburne, Carol [Savannah River Site (SRS), Aiken, SC (United States). Savannah River Remediation, LLC; Osterberg, Paul [Fauske and Associates, LLC, Burr Ridge, IL (United States); Johnson, Tom [Fauske and Associates, LLC, Burr Ridge, IL (United States); Frawely, Thomas [Fauske and Associates, LLC, Burr Ridge, IL (United States)

    2013-01-23T23:59:59.000Z

    The Savannah River Site, in conjunction with AREVA Federal services, has designed a process to treat dissolved radioactive waste solids with ozone. It is known that in this radioactive waste process, radionuclides radiolytically break down water into gaseous hydrogen and oxygen, which presents a well defined flammability hazard. Flammability limits have been established for both ozone and hydrogen separately; however, there is little information on mixtures of hydrogen and ozone. Therefore, testing was designed to provide critical flammability information necessary to support safety related considerations for the development of ozone treatment and potential scale-up to the commercial level. Since information was lacking on flammability issues at low levels of hydrogen and ozone, a testing program was developed to focus on filling this portion of the information gap. A 2-L vessel was used to conduct flammability tests at atmospheric pressure and temperature using a fuse wire ignition source at 1 percent ozone intervals spanning from no ozone to the Lower Flammable Limit (LFL) of ozone in the vessel, determined as 8.4%(v/v) ozone. An ozone generator and ozone detector were used to generate and measure the ozone concentration within the vessel in situ, since ozone decomposes rapidly on standing. The lower flammability limit of hydrogen in an ozone-oxygen mixture was found to decrease from the LFL of hydrogen in air, determined as 4.2 % (v/v) in this vessel. From the results of this testing, Savannah River was able to develop safety procedures and operating parameters to effectively minimize the formation of a flammable atmosphere.

  19. Jet impingement heat transfer in two-pass rotating rectangular channels 

    E-Print Network [OSTI]

    Zhang, Yuming

    1996-01-01T23:59:59.000Z

    compared with previously reported correlations. The pressure distributions show that the major effect on heat transfer is the jet impingement in the beginning of the channel and the cross flow at the end of the channel. The results also show that heat...

  20. BWR ex-vessel steam explosion analysis with MC3D code

    SciTech Connect (OSTI)

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

    2012-07-01T23:59:59.000Z

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

  1. Pressure cryocooling protein crystals

    DOE Patents [OSTI]

    Kim, Chae Un (Ithaca, NY); Gruner, Sol M. (Ithaca, NY)

    2011-10-04T23:59:59.000Z

    Preparation of cryocooled protein crystal is provided by use of helium pressurizing and cryocooling to obtain cryocooled protein crystal allowing collection of high resolution data and by heavier noble gas (krypton or xenon) binding followed by helium pressurizing and cryocooling to obtain cryocooled protein crystal for collection of high resolution data and SAD phasing simultaneously. The helium pressurizing is carried out on crystal coated to prevent dehydration or on crystal grown in aqueous solution in a capillary.

  2. High temperature pressure gauge

    DOE Patents [OSTI]

    Echtler, J. Paul (Pittsburgh, PA); Scandrol, Roy O. (Library, PA)

    1981-01-01T23:59:59.000Z

    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.

  3. Pressure-sensitive optrode

    DOE Patents [OSTI]

    Hirschfeld, T.B.

    1985-04-09T23:59:59.000Z

    An apparatus and method are disclosed for sensing changes in pressure and for generating optical signals related to changes in pressure. Light from a fiber optic is directed to a movable surface which is coated with a light-responsive material, and which moves relative to the end of the fiber optic in response to changes in pressure. The same fiber optic collects a portion of the reflected or emitted light from the movable surface. Changes in pressure are determined by measuring changes in the amount of light collected. 5 figs.

  4. Aspect ratio dependence of heat transport by turbulent Rayleigh-B\\'{e}nard convection in rectangular cells

    E-Print Network [OSTI]

    Zhou, Quan; Li, Chun-Mei; Zhong, Bao-Chang

    2012-01-01T23:59:59.000Z

    We report high-precision measurements of the Nusselt number $Nu$ as a function of the Rayleigh number $Ra$ in water-filled rectangular Rayleigh-B\\'{e}nard convection cells. The horizontal length $L$ and width $W$ of the cells are 50.0 cm and 15.0 cm, respectively, and the heights $H=49.9$, 25.0, 12.5, 6.9, 3.5, and 2.4 cm, corresponding to the aspect ratios $(\\Gamma_x\\equiv L/H,\\Gamma_y\\equiv W/H)=(1,0.3)$, $(2,0.6)$, $(4,1.2)$, $(7.3,2.2)$, $(14.3,4.3)$, and $(20.8,6.3)$. The measurements were carried out over the Rayleigh number range $6\\times10^5\\lesssim Ra\\lesssim10^{11}$ and the Prandtl number range $5.2\\lesssim Pr\\lesssim7$. Our results show that for rectangular geometry turbulent heat transport is independent of the cells' aspect ratios and hence is insensitive to the nature and structures of the large-scale mean flows of the system. This is slightly different from the observations in cylindrical cells where $Nu$ is found to be in general a decreasing function of $\\Gamma$, at least for $\\Gamma=1$ and l...

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

    Energy Savers [EERE]

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

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

    E-Print Network [OSTI]

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

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

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

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

  8. High pressure feeder and method of operating to feed granular or fine materials

    DOE Patents [OSTI]

    Vimalchand, Pannalal; Liu, Guohai; Peng, Wan Wang

    2014-10-07T23:59:59.000Z

    A coal feed system to feed pulverized low rank coals containing up to 25 wt % moisture to gasifiers operating up to 1000 psig pressure is described. The system includes gas distributor and collector gas permeable pipes imbedded in the lock vessel. Different methods of operation of the feed system are disclosed to minimize feed problems associated with bridging and packing of the pulverized coal. The method of maintaining the feed system and feeder device exit pressures using gas addition or extraction with the pressure control device is also described.

  9. Aerosol source term in high-pressure-melt ejection. [PWR; BWR

    SciTech Connect (OSTI)

    Brockmann, J.E.; Tarbell, W.W.

    1983-01-01T23:59:59.000Z

    Pressurized ejection of melt from a reactor pressure vessel has been identified as an important element of a severe reactor accident. Copious aerosol production is observed when thermitically generated melts pressurized with nitrogen or carbon dioxide to 1.3 to 17 MPa are ejected into an air atmosphere. Aerosol particle size distributions measured in the tests have modes of about 0.5, 5, and > 10..mu..m. Mechanisms leading to formation of these multimodal size distributions are suggested. This aerosol is a potentially important fission product source term which has not been considered in previous severe accident analyses.

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

    DOE Patents [OSTI]

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

    2006-02-14T23:59:59.000Z

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

  11. PRESSURE ACTIVATED SEALANT TECHNOLOGY

    SciTech Connect (OSTI)

    Michael A. Romano

    2004-04-01T23:59:59.000Z

    The objective of this project is to develop new, efficient, cost effective methods of internally sealing natural gas pipeline leaks through the application of differential pressure activated sealants. In researching the current state of the art for gas pipeline sealing technologies we concluded that if the project was successful, it appeared that pressure activated sealant technology would provide a cost effective alternative to existing pipeline repair technology. From our analysis of current field data for a 13 year period from 1985 to 1997 we were able to identify 205 leaks that were candidates for pressure activated sealant technology, affirming that pressure activated sealant technology is a viable option to traditional external leak repairs. The data collected included types of defects, areas of defects, pipe sizes and materials, incident and operating pressures, ability of pipeline to be pigged and corrosion states. This data, and subsequent analysis, was utilized as a basis for constructing applicable sealant test modeling.

  12. IMPACT OF NUCLEAR MATERIAL DISSOLUTION ON VESSEL CORROSION

    SciTech Connect (OSTI)

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

    2012-10-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Goodarz Ahmadi

    2002-07-01T23:59:59.000Z

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

  14. Start-up control system and vessel for lmfbr

    SciTech Connect (OSTI)

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

    1987-04-07T23:59:59.000Z

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

  15. Pressurizer tank upper support

    DOE Patents [OSTI]

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

    1994-01-01T23:59:59.000Z

    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.

  16. Pressurizer tank upper support

    DOE Patents [OSTI]

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

    1994-01-11T23:59:59.000Z

    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.

  17. Design of a high-pressure research flow loop for the experimental investigation of liquid loading in gas wells

    E-Print Network [OSTI]

    Fernandez Alvarez, Juan Jose

    2009-05-15T23:59:59.000Z

    compressors working in parallel is the most technical and economic configuration for the TowerLab based on the overall costs provided by the supplier, the footprint but most importantly the flexibility. The design of the pressure vessel required a cylindrical...

  18. Extremal transmission through a microwave photonic crystal and the observation of edge states in a rectangular Dirac billiard

    E-Print Network [OSTI]

    S. Bittner; B. Dietz; M. Miski-Oglu; A. Richter

    2011-12-20T23:59:59.000Z

    This article presents experimental results on properties of waves propagating in an unbounded and a bounded photonic crystal consisting of metallic cylinders which are arranged in a triangular lattice. First, we present transmission measurements of plane waves traversing a photonic crystal. The experiments are performed in the vicinity of a Dirac point, i.e., an isolated conical singularity of the photonic band structure. There, the transmission shows a pseudodiffusive 1/L dependence, with $L$ being the thickness of the crystal, a phenomenon also observed in graphene. Second, eigenmode intensity distributions measured in a microwave analog of a relativistic Dirac billiard, a rectangular microwave billiard that contains a photonic crystal, are discussed. Close to the Dirac point states have been detected which are localized at the straight edge of the photonic crystal corresponding to a zigzag edge in graphene.

  19. Effect of rib spacing on heat transfer and friction in a rotating two-pass rectangular (AR=1:2) channel 

    E-Print Network [OSTI]

    Liu, Yao-Hsien

    2006-10-30T23:59:59.000Z

    The research focuses on testing the heat transfer enhancement in a channel for different spacing of the rib turbulators. Those ribs are put on the surface in the two pass rectangular channel with an aspect ratio of AR=1:2. The cross section...

  20. Capacitance pressure sensor

    DOE Patents [OSTI]

    Eaton, William P. (Tijeras, NM); Staple, Bevan D. (Albuquerque, NM); Smith, James H. (Albuquerque, NM)

    2000-01-01T23:59:59.000Z

    A microelectromechanical (MEM) capacitance pressure sensor integrated with electronic circuitry on a common substrate and a method for forming such a device are disclosed. The MEM capacitance pressure sensor includes a capacitance pressure sensor formed at least partially in a cavity etched below the surface of a silicon substrate and adjacent circuitry (CMOS, BiCMOS, or bipolar circuitry) formed on the substrate. By forming the capacitance pressure sensor in the cavity, the substrate can be planarized (e.g. by chemical-mechanical polishing) so that a standard set of integrated circuit processing steps can be used to form the electronic circuitry (e.g. using an aluminum or aluminum-alloy interconnect metallization).

  1. High pressure counterflow CHF.

    E-Print Network [OSTI]

    Walkush, Joseph Patrick

    1975-01-01T23:59:59.000Z

    This is a report of the experimental results of a program in countercurrent flow critical heat flux. These experiments were performed with Freon 113 at 200 psia in order to model a high pressure water system. An internally ...

  2. 196 MATHEMATICS MAGAZINE Blood Vessel Branching: Beyond the

    E-Print Network [OSTI]

    Adam, John A.

    them to consist of thin layers that slide past one another, developing a resistance to the flow, it is not reasonable to think of layers of fluid sliding past each other, so our models do not apply. Furthermore, the pressure driving the whole system is far from constant; there are short time lags between the high pressure

  3. Pressurized melt ejection into water pools

    SciTech Connect (OSTI)

    Tarbell, W.W.; Pilch, M. (Sandia National Labs., Albuquerque, NM (USA)); Ross, J.W.; Oliver, M.S.; Gilbert, D.W.; Nichols, R.T. (Ktech Corp., Albuquerque, NM (USA))

    1991-03-01T23:59:59.000Z

    Direct Containment Heating is important because it is one of the postulated methods for early containment failure. If the reactor pressure vessel (RPV) should fail at an instrument tube penetration in the lower head, the resulting aperture would allow the molten core material to be discharged at high velocity into the cavity. Scaled experiments have demonstrated that the gas discharged during blowdown of the pressure system can entrain core debris and carry it out of the cavity region. Although these experiments were performed with the cavity initially devoid of water, other tests with the cavity partially filled with water exhibited similar results. The objective of the work described here is twofold: (1) to study the jet ejection and debris dispersal behavior when water is in contact with the lower head of the RPV and completely fills the cavity; and, (2) to compare the results of an experiment where the cavity is partially filled with water. These tests are of interest not only because they consider the dispersal of water and debris from the cavity but they also consider the potential consequences of codispersing water with core debris into the containment. Because the core debris may impart sufficient energy to the containment atmosphere to raise the pressure to potentially threatening levels, it is important to identify possible mitigating mechanisms. Analytical efforts have suggested that the codispersed water may act as a finely distributed heat sink that would have the beneficial effect of absorbing debris energy. This has not been confirmed experimentally, although the work presented here does attempt to identify the potential for water preexisting in the cavity to be dispersed as small droplets. 17 refs., 41 figs., 12 tabs.

  4. Pressurized water nuclear reactor system with hot leg vortex mitigator

    DOE Patents [OSTI]

    Lau, Louis K. S. (Monroeville, PA)

    1990-01-01T23:59:59.000Z

    A pressurized water nuclear reactor system includes a vortex mitigator in the form of a cylindrical conduit between the hot leg conduit and a first section of residual heat removal conduit, which conduit leads to a pump and a second section of residual heat removal conduit leading back to the reactor pressure vessel. The cylindrical conduit is of such a size that where the hot leg has an inner diameter D.sub.1, the first section has an inner diameter D.sub.2, and the cylindrical conduit or step nozzle has a length L and an inner diameter of D.sub.3 ; D.sub.3 /D.sub.1 is at least 0.55, D.sub.2 is at least 1.9, and L/D.sub.3 is at least 1.44, whereby cavitation of the pump by a vortex formed in the hot leg is prevented.

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

    SciTech Connect (OSTI)

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

    2004-12-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1996-03-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Not Available

    1981-08-01T23:59:59.000Z

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

  8. Head Loss Evaluation in a PWR Reactor Vessel Using CFD Analysis

    SciTech Connect (OSTI)

    Ji Hwan Jeong; Jong Pil Park [School of Mechanical Engineering, Pusan National University, Enesys Jangjeon-dong, Geumjeong-gu, Busan (Korea, Republic of); Byoung-Sub Han [Jangdae-dong, Yusong-gu, Daejeon (Korea, Republic of)

    2006-07-01T23:59:59.000Z

    Nuclear vendors and utilities perform lots of simulations and analyses in order to ensure the safe operation of nuclear power plants (NPPs). In general, the simulations are carried out using vendor-specific design codes and best-estimate system analysis codes and most of them were developed based on 1-dimensional lumped parameter models. During the past decade, however, computers, parallel computation methods, and 3-dimensional computational fluid dynamics (CFD) codes have been dramatically enhanced. It is believed to be beneficial to take advantage of advanced commercial CFD codes in safety analysis and design of NPPs. The present work aims to analyze the flow distribution in downcomer and lower plenum of Korean standard nuclear power plants (KSNPs) using STAR-CD. The lower plenum geometry of a PWR is very complicated since there are so many reactor internals, which hinders in CFD analysis for real reactor geometry up to now. The present work takes advantage of 3D CAD model so that real geometry of lower plenum is used. The results give a clear figure about flow fields in the reactor vessel, which is one of major safety concerns. The calculated pressure drop across downcomer and lower plenum appears to be in good agreement with the data in engineering calculation note. A algorithm which can evaluate head loss coefficient which is necessary for thermal-hydraulic system code running was suggested based on this CFD analysis results. (authors)

  9. High-Pressure Melt Streaming (HIPS) program plan

    SciTech Connect (OSTI)

    Tarbell, W.; Brockmann, J.; Pilch, M.

    1984-08-01T23:59:59.000Z

    The Zion Probabilistic Safety Study (ZPSS) envisions accident sequences that could lead to failure of the reactor vessel while the primary system is pressurized. The resulting ejection of molten core material into the reactor cavity followed by the blowdown of steam and hydrogen is shown to cause the debris to enter into the containment region. The High Pressure Melt Streaming (HIPS) program has been developed to provide an experimental and analytical investigation of the scenario described above. One-tenth linear scale models of the Zion cavity region will be used to investigate the debris dispersal phenomena. Smaller-scale experiments (SPIT-tests) are also used to study high-velocity jets, jet-water interactions, and 1/20th scale cavity geometries. Both matrices are developed using a factorial design approach. The document describes certain aspects of the ZPSS ex-vessel phenomena, the experimental matrices, test equipment, and instrumentation, and the program's analytical efforts. Preliminary data from SPIT testing are included.

  10. Oxygen partial pressure sensor

    DOE Patents [OSTI]

    Dees, D.W.

    1994-09-06T23:59:59.000Z

    A method for detecting oxygen partial pressure and an oxygen partial pressure sensor are provided. The method for measuring oxygen partial pressure includes contacting oxygen to a solid oxide electrolyte and measuring the subsequent change in electrical conductivity of the solid oxide electrolyte. A solid oxide electrolyte is utilized that contacts both a porous electrode and a nonporous electrode. The electrical conductivity of the solid oxide electrolyte is affected when oxygen from an exhaust stream permeates through the porous electrode to establish an equilibrium of oxygen anions in the electrolyte, thereby displacing electrons throughout the electrolyte to form an electron gradient. By adapting the two electrodes to sense a voltage potential between them, the change in electrolyte conductivity due to oxygen presence can be measured. 1 fig.

  11. Plating under reduced pressure

    SciTech Connect (OSTI)

    Dini, J.W.; Beat, T.G.; Cowden, W.C. (Lawrence Livermore National Lab., CA (United States)); Ryan, L.E.; Hewitt, W.B. (TRW, Inc., Redondo Beach, CA (United States))

    1992-06-01T23:59:59.000Z

    Plating under reduced pressure was evaluated for both electroless nickel and electrodeposited copper systems. The objective was to reduce pitting of these coatings thereby further enhancing their usage for diamond turning applications. Cursory experiments with electroless nickel showed reduced porosity when deposition was done at around 500 torr. Detailed experiments with electrodeposited copper at around 100 torr provided similar results. Scanning tunneling microscopy was effectively used to show the improvement in the copper deposits plated under reduced pressure. Benefits included reduced surface roughness and finer and denser grain structure.

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

    DOE Patents [OSTI]

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

    1985-03-01T23:59:59.000Z

    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.

  13. Saltstone Osmotic Pressure

    SciTech Connect (OSTI)

    Nichols, Ralph L.; Dixon, Kenneth L.

    2013-09-23T23:59:59.000Z

    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

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

    DOE Patents [OSTI]

    Ekeroth, D.E.; Orr, R.

    1993-12-07T23:59:59.000Z

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

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

    DOE Patents [OSTI]

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

    1993-01-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

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

  17. Modeling the pressure increase in liquid helium cryostats after failure of the insulating vacuum

    SciTech Connect (OSTI)

    Heidt, C.; Grohmann, S. [Karlsruhe Institute of Technology, Institute for Technical Physics, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen, Germany and Karlsruhe Institute of Technology, Institute for Technical Thermodynamics and Refrigeration, Engler-Bunte (Germany); Süßer, M. [Karlsruhe Institute of Technology, Institute for Technical Physics, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany)

    2014-01-29T23:59:59.000Z

    The pressure relief system of liquid helium cryostats requires a careful design, due to helium's low enthalpy of vaporization and due to the low operating temperature. Hazard analyses often involve the failure of the insulating vacuum in the worst-case scenario. The venting of the insulating vacuum and the implications for the pressure increase in the helium vessel, however, have not yet been fully analyzed. Therefore, the dimensioning of safety devices often requires experience and reference to very few experimental data. In order to provide a better foundation for the design of cryogenic pressure relief systems, this paper presents an analytic approach for the strongly dynamic process induced by the loss of insulating vacuum. The model is based on theoretical considerations and on differential equation modeling. It contains only few simplifying assumptions, which will be further investigated in future experiments. The numerical solutions of example calculations are presented with regard to the heat flux into the helium vessel, the helium pressure increase and the helium flow rate through the pressure relief device. Implications concerning two-phase flow and the influence of kinetic energy are discussed.

  18. Optical Measurement Technologies for High Temperature, Radiation Exposure, and Corrosive Environments—Significant Activities and Findings: In-vessel Optical Measurements for Advanced SMRs

    SciTech Connect (OSTI)

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

    2012-09-01T23:59:59.000Z

    Development of advanced Small Modular Reactors (aSMRs) is key to providing the United States with a sustainable, economically viable, and carbon-neutral energy source. The aSMR designs have attractive economic factors that should compensate for the economies of scale that have driven development of large commercial nuclear power plants to date. For example, aSMRs can be manufactured at reduced capital costs in a factory and potentially shorter lead times and then be shipped to a site to provide power away from large grid systems. The integral, self-contained nature of aSMR designs is fundamentally different than conventional reactor designs. Future aSMR deployment will require new instrumentation and control (I&C) architectures to accommodate the integral design and withstand the extreme in-vessel environmental conditions. Operators will depend on sophisticated sensing and machine vision technologies that provide efficient human-machine interface for in-vessel telepresence, telerobotic control, and remote process operations. The future viability of aSMRs is dependent on understanding and overcoming the significant technical challenges involving in-vessel reactor sensing and monitoring under extreme temperatures, pressures, corrosive environments, and radiation fluxes

  19. Aging study of boiling water reactor high pressure injection systems

    SciTech Connect (OSTI)

    Conley, D.A.; Edson, J.L.; Fineman, C.F. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1995-03-01T23:59:59.000Z

    The purpose of high pressure injection systems is to maintain an adequate coolant level in reactor pressure vessels, so that the fuel cladding temperature does not exceed 1,200{degrees}C (2,200{degrees}F), and to permit plant shutdown during a variety of design basis loss-of-coolant accidents. This report presents the results of a study on aging performed for high pressure injection systems of boiling water reactor plants in the United States. The purpose of the study was to identify and evaluate the effects of aging and the effectiveness of testing and maintenance in detecting and mitigating aging degradation. Guidelines from the United States Nuclear Regulatory Commission`s Nuclear Plant Aging Research Program were used in performing the aging study. Review and analysis of the failures reported in databases such as Nuclear Power Experience, Licensee Event Reports, and the Nuclear Plant Reliability Data System, along with plant-specific maintenance records databases, are included in this report to provide the information required to identify aging stressors, failure modes, and failure causes. Several probabilistic risk assessments were reviewed to identify risk-significant components in high pressure injection systems. Testing, maintenance, specific safety issues, and codes and standards are also discussed.

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

    E-Print Network [OSTI]

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

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

    E-Print Network [OSTI]

    Lednicky, Edward Frank

    1971-01-01T23:59:59.000Z

    of the "standard epoxy" of Leven (14) were cast using Bailey's (5) procedures. The blocks measured approximately 16" x 18" x 4". The composition of "standard epoxy" is (by weight) 100 parts of Bakelite ERL 2774 epoxy resin, 42 parts of phathalic anhydride...

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

    SciTech Connect (OSTI)

    Schneider, U.A.

    1981-01-01T23:59:59.000Z

    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)

  3. Single crystal silicon as a macro-world structural material : application to compact, lightweight high pressure vessels

    E-Print Network [OSTI]

    Garza, Tanya Cruz

    2011-01-01T23:59:59.000Z

    Single crystal silicon has promising inherent structural properties which are attractive for weight sensitive applications. Single crystal silicon, however, is a brittle material which makes the usable strength that can ...

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

    E-Print Network [OSTI]

    Horn, R.M.

    2011-01-01T23:59:59.000Z

    gained through use in the petrochemical industry i n d i c ae s s f u l l y f o r petrochemical a p p l i c a t i o n .Code does not However, petrochemical experience has shown

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

    E-Print Network [OSTI]

    Michaleris, Panagiotis

    draft of API RP-579, Recommended Practice for Fitness-for-Service. INTRODUCTION When performing standards. While welding residual stresses can reach or exceed yield in certain situations. These recommendations have been proposed for inclusion in the current draft of API RP-579, Recommended Practice

  6. The modelling of irradiation-enhanced phosphorus segregation in neutron irradiated reactor pressure vessel submerged-arc welds

    SciTech Connect (OSTI)

    Druce, S.G.; English, C.A.; Foreman, A.J.E.; McElroy, R.J.; Vatter, I.A. [AEA Technology, Didcot (United Kingdom). Harwell Lab.; Bolton, C.J.; Buswell, J.T.; Jones, R.B. [Nuclear Electric, Berkeley (United Kingdom). Berkeley Technology Centre

    1996-12-31T23:59:59.000Z

    Recent results on neutron-irradiated RPV submerged-arc welds have revealed grain boundary segregation of phosphorus during irradiation, which may lead to intergranular fracture. However, the experimental database is insufficient to define the dependence of the process on variables such ad dose, dose-rate and temperature. This paper describes work in which two existing models of phosphorus segregation, under thermal or irradiation conditions, have been developed to obtain predictions of these dependencies. The critical parameters in the models have been adjusted to give consistency with the available reference data, and predictions have been made of the dependence of segregation on a number of variables.

  7. The design of a reduced diameter Pebble Bed Modular Reactor for reactor pressure vessel transport by railcar

    E-Print Network [OSTI]

    Everson, Matthew S

    2009-01-01T23:59:59.000Z

    Many desirable locations for Pebble Bed Modular Reactors are currently out of consideration as construction sites for current designs due to limitations on the mode of transportation for large RPVs. In particular, the ...

  8. A flexible pressure monitoring system for pressure ulcer prevention

    E-Print Network [OSTI]

    Yip, Marcus

    Pressure ulcers are painful sores that arise from prolonged exposure to high pressure points, which restricts blood flow and leads to tissue necrosis. This is a common occurrence among patients with impaired mobility, ...

  9. Continuous pressure letdown system

    DOE Patents [OSTI]

    Sprouse, Kenneth M.; Matthews, David R.; Langowski, Terry

    2010-06-08T23:59:59.000Z

    A continuous pressure letdown system connected to a hopper decreases a pressure of a 2-phase (gas and solid) dusty gas stream flowing through the system. The system includes a discharge line for receiving the dusty gas from the hopper, a valve, a cascade nozzle assembly positioned downstream of the discharge line, a purge ring, an inert gas supply connected to the purge ring, an inert gas throttle, and a filter. The valve connects the hopper to the discharge line and controls introduction of the dusty gas stream into the discharge line. The purge ring is connected between the discharge line and the cascade nozzle assembly. The inert gas throttle controls a flow rate of an inert gas into the cascade nozzle assembly. The filter is connected downstream of the cascade nozzle assembly.

  10. High pressure oxygen furnace

    DOE Patents [OSTI]

    Morris, Donald E. (Kensington, CA)

    1992-01-01T23:59:59.000Z

    A high temperature high pressure oxygen furnace having a hybrid partially externally heated construction is disclosed. A metallic bar 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 inch bar stock and has a length of about 17 inches. This bar stock is gun drilled for over 16 inches of its length with 0.400 inch aperture to define a closed high temperature, high pressure oxygen chamber. The opposite and closed end of the bar is provided with a small support aperture into which both a support and a thermocouple can be inserted. The closed end of the gun drilled bar is inserted into an oven, preferably heated by standard nickel chrome electrical elements and having a heavily insulated exterior.

  11. High pressure oxygen furnace

    DOE Patents [OSTI]

    Morris, D.E.

    1992-07-14T23:59:59.000Z

    A high temperature high pressure oxygen furnace having a hybrid partially externally heated construction is disclosed. A metallic bar 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 inch bar stock and has a length of about 17 inches. This bar stock is gun drilled for over 16 inches of its length with 0.400 inch aperture to define a closed high temperature, high pressure oxygen chamber. The opposite and closed end of the bar is provided with a small support aperture into which both a support and a thermocouple can be inserted. The closed end of the gun drilled bar is inserted into an oven, preferably heated by standard nickel chrome electrical elements and having a heavily insulated exterior. 5 figs.

  12. Device for automating in vitro characterization of lymphatic vessel function

    E-Print Network [OSTI]

    Rajagopalan, Shruti

    2005-02-17T23:59:59.000Z

    pump and resistive effects that informs the direction of the present work. . Figure 7: Effect of outflow pressure on lung lymph flow. As the outflow pressure increases, the lymph flow decreases. Digitized and reproduced from Drake et al., J Appl... edemagenic stress. Am J Physiol 257: H2059-2069, 1989. 3. Brady AJ. The three element model of muscle mechanics: its applicability to cardiac muscle. Physiologist 10: 75-86, 1967. 4. Drake R, Giesler M, Laine G, Gabel J, and Hansen T. Effect of outflow...

  13. The analysis of cracks in high-pressure piping and their effects on strength and lifetime of construction components at the Ignalina nuclear plant

    SciTech Connect (OSTI)

    Aleev, A.; Petkevicius, K.; Senkus, V. [and others

    1997-04-01T23:59:59.000Z

    A number of cracks and damages of other sorts have been identified in the high-pressure parts at the Ignalina Nuclear Plant. They are caused by inadequate production- and repair technologies, as well as by thermal, chemical and mechanical processes of their performance. Several techniques are available as predictions of cracks and other defects of pressurized vessels. The choice of an experimental technique should be based on the level of its agreement with the actual processes.

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

    E-Print Network [OSTI]

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

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

    E-Print Network [OSTI]

    Skogestad, Sigurd

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

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

    E-Print Network [OSTI]

    Skogestad, Sigurd

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

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

    E-Print Network [OSTI]

    Barton, Jennifer K.

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

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

    E-Print Network [OSTI]

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

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

    E-Print Network [OSTI]

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

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

    SciTech Connect (OSTI)

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

    2012-07-01T23:59:59.000Z

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

  1. Economics of Steam Pressure Reduction

    E-Print Network [OSTI]

    Sylva, D. M.

    Economics of Steam Pressure Reduction is a technical paper that addresses the operating and economic advantages associated with the program to lower the steam operating pressure. Evaluation of a testing program will be discussed. The paper...

  2. Measurements of wall heat (mass) transfer for flow through blockages with round and square holes in a wide rectangular channel 

    E-Print Network [OSTI]

    Cervantes, Joel

    2002-01-01T23:59:59.000Z

    . . . n1 DEDICATION. ACKNOWLEGDEMENTS . . . V1 TABLE OF CONTENTS . vn LIST OF FIGURES. NOMENCLATURE . . INTRODUCTION. LITERATURE SURVEY. EXPERIMENTAL APPARATUS . Xt EXPERIMENTAL PROCEDURE DATA REDUCTION. PRESENTATION & DISCUSSION OF RESULTS..., kg/(m s) Nun~ local Nusselt number //u pa average Nusselt number Nus reference Nusselt number for fully developed turbulent flow in smooth channel POIIII Pv, w atmospheric pressure, N/m 2 vapor pressure on naphthalene surface, N/m 2 P...

  3. Nitrogen at very high pressure

    SciTech Connect (OSTI)

    Nellis, W.J.

    1987-07-01T23:59:59.000Z

    High-pressure results for nitrogen are reviewed and discussed in terms of phenomena that occur at extreme conditions.

  4. Pressure Data Within BOP- ODS

    Broader source: Energy.gov [DOE]

    This file describes the components within the BOP and the pressure readings taken during diagnostic operations on May 25.

  5. Pressure Data Within BOP- XLS

    Broader source: Energy.gov [DOE]

    This file describes the components within the BOP and the pressure readings taken during diagnostic operations on May 25.

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

    SciTech Connect (OSTI)

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

    2007-12-01T23:59:59.000Z

    The current regulations to insure that nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to transients such as pressurized thermal shock (PTS) events were derived from computational models developed in the early-to-mid 1980s. Since that time, advancements and refinements in relevant technologies that impact RPV integrity assessment have led to an effort by the NRC to re-evaluate its PTS regulations. Updated computational methodologies have been developed through interactions between experts in the relevant disciplines of thermal hydraulics, probabilistic risk assessment, materials embrittlement, fracture mechanics, and inspection (flaw characterization). Contributors to the development of these methodologies include the NRC staff, their contractors, and representatives from the nuclear industry. These updated methodologies have been integrated into the Fracture Analysis of Vessels – Oak Ridge (FAVOR, v06.1) computer code developed for the NRC by the Heavy Section Steel Technology (HSST) program at Oak Ridge National Laboratory (ORNL). The FAVOR, v04.1, code represents the baseline NRC-selected applications tool for re-assessing the current PTS regulations. This report is intended to document the technical bases for the assumptions, algorithms, methods, and correlations employed in the development of the FAVOR, v06.1, code.

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

    SciTech Connect (OSTI)

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

    1984-01-01T23:59:59.000Z

    An analysis has been performed of corium sweepout behavior in the ANL/EPRI CWTI-series reactor material experiments involving the gas pressure-driven injection of molten corium into the reactor cavity region of a 1:30 scale mockup of a PWR containment. A computer model was developed to calculate the sweepout versus retention of corium and water from the cavity. The model consists of hydrodynamics and freezing calculations describing the pressure-driven two-phase flow of corium, water, steam and gas out of the cavity, freezing of corium upon structural surfaces, and levitation of corium within the cavity by the vessel blowdown gas jet. The model has had good success predicting the disposition of corium for the available CWTI tests, indicating retention in the cavity of between 40 and 70% of the injected corium masses. For conditions representative of the TMLB' sequence in the reactor system, the model predicts essentially complete sweepout of corium from the full-scale cavity region before the dispersive forces arising from the blowdown of the primary system have decayed. However, this large sweepout does not imply that the swept out material would deliver its energy directly to the containment atmosphere.

  8. Blood Pressure Medicine: Special Instructions

    E-Print Network [OSTI]

    Bandettini, Peter A.

    Blood Pressure Medicine: Special Instructions: U.S. DEPARTMENT OF HEALTH AND HUMAN SERVICES National Institutes of Health National Heart, Lung, and Blood Institute · What is my blood pressure reading in numbers? · What is my goal blood pressure? · Is there a healthy eating plan that I should follow to help

  9. Incident IR Bandwidth Effects on Efficiency and Shaping for Third Harmonic Generation of Quasi-Rectangular UV Longitudinal Profiles

    SciTech Connect (OSTI)

    Not Available

    2010-12-07T23:59:59.000Z

    The photocathode of the proposed LCLS RF Photoinjector will be irradiated by uv laser light which is generated as the third harmonic of incident fundamental ir laser light. We have investigated quantitatively the effect of input ir spectral bandwidth on the exiting longitudinal intensity profiles, energy conversion efficiencies and spectral bandwidths that characterize the third harmonic generation (THG) process with a pair of crystals. These profiles, efficiencies and bandwidths include the residual fundamental and residual second harmonic light exiting the second crystal. The intrinsic acceptance bandwidth for THG is determined by crystal material and thickness as well as the type of phase matching that is used. For our case of BBO material with type I phase matching these bandwidths are approximately 0.9 nm*cm and 0.1 nm*cm for second and third harmonic generation respectively. Consequently for fixed crystal thicknesses and a fixed input ir longitudinal profile, the specified input ir bandwidth will determine the profiles, efficiencies and bandwidths exiting the second crystal. The results reported here are predictions of the SNLO code that is available as 'freeware' from the Sandia National Laboratories. It has been modified for this work. It is critical to note that this modification has enabled us to generate SNLO predictions of the 'coupled' case in which the output of the first crystal is used as input to the second crystal. Our focus is the dependence of uv longitudinal intensity profile and THG efficiency on the input ir bandwidth and crystal thicknesses. We include here cases that best illustrate input bandwidth effects. The criteria for selection of reported cases are highest efficiency generation of quasi-rectangular uv profiles with proportional intensity ripple less than 5% rms on the plateau of the pulse. Maximizing THG efficiency typically amounts to maximizing the crystal thicknesses with the longitudinal profile constraint. The specified incident ir longitudinal profile is quasi-rectangular (i.e. nonzero risetime and falltime with small intensity variation on the plateau) with a 10 psec pulse duration (FWHM). By assumption, this profile has been established upstream of the crystals at the fundamental ir wavelength. The simplest possible optical configuration is used in this work as shown in figure 1. The first crystal is the site of second harmonic generation (SHG) driven by the incident ir irradiation of central wavelength, 800nm. Downstream of the first crystal, the second crystal is the site of third harmonic generation (THG) which occurs by sum frequency mixing. Inter-crystal optics (such as a half waveplate) are assumed to be lossless at the fundamental and second harmonic wavelengths. As shown in figure 1, a portion of the incident ir irradiation is not sequestered from the first crystal for subsequent THG in the second crystal. Also, quasi-phase matching configurations and other complex compensation schemes have not been investigated at this point. The simplistic geometry better elucidates the intrinsic acceptance bandwidth limitations imposed by the crystals. Our goal in this endeavor has been to conduct a quantitative assessment of incident ir bandwidth effects on the THG process for BBO material of varied thicknesses and not, at this stage, to comply with all uv pulse specifications for the LCLS RF Photoinjector. Nonetheless, our results can be compared with LCLS photoinjector uv pulse requirements which call for a nominal 10 psec FWHM with 1 psec risetime and falltime and a nominally flat plateau (allowing for slope adjustments) with no more than a 5% rms proportional intensity variation. Furthermore, the results of this work can be used to suggest crystal thicknesses that would likely comply with all uv pulse requirements given the appropriate longitudinal profile and bandwidth for an input ir pulse.

  10. Peer review of the Three Mile Island Unit 2 Vessel Investigation Project metallurgical examinations

    SciTech Connect (OSTI)

    Bohl, R.W.; Gaydos, R.G.; Vander Voort, G.F.; Diercks, D.R. [Argonne National Lab., IL (United States)

    1994-07-01T23:59:59.000Z

    Fifteen samples recovered from the lower head of the Three Mile Island (TMI) Unit 2 nuclear reactor pressure vessel were subjected to detailed metallurgical examinations by the Idaho National Engineering Laboratory (INEL), with supporting work carried out by Argonne National Laboratory (ANL) and several of the European participants. These examinations determined that a portion of the lower head, a so-called elliptical ``hot spot`` measuring {approx}0.8 {times} 1 m, reached temperatures as high as 1100{degrees}C during the accident and cooled from these temperatures at {approx}10--100{degrees}C/min. The remainder of the lower head was found to have remained below the ferrite-toaustenite transformation temperature of 727{degrees}C during the accident. Because of the significance of these results and their importance to the overall analysis of the TMI accident, a panel of three outside peer reviewers, Dr. Robert W. Bohl, Mr. Richard G. Gaydos, and Mr. George F. Vander Voort, was formed to conduct an independent review of the metallurgical analyses. After a thorough review of the previous analyses and examination of photo-micrographs and actual lower head specimens, the panel determined that the conclusions resulting from the INEL study were fundamentally correct. In particular, the panel reaffirmed that four lower head samples attained temperatures as high as 1100{degrees}C, and perhaps as high as 1150--1200{degrees}C in one case, during the accident. They concluded that these samples subsequently cooled at a rate of {approx}50--125{degrees}C/min in the temperature range of 600--400{degrees}C, in good agreement with the original analysis. The reviewers also agreed that the remainder of the lower head samples had not exceeded the ferrite-to-austenite transformation temperature during the accident and suggested several refinements and alternative procedures that could have been employed in the original analysis.

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

    E-Print Network [OSTI]

    Cook, Gregory D.

    1997-01-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    DeWitt, Gregory L

    2011-01-01T23:59:59.000Z

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

  13. Direct containment heating and aerosol generation during high-pressure-melt expulsion experiments

    SciTech Connect (OSTI)

    Tarbell, W.W.; Brockmann, J.E.; Washington, K.E.; Pilch, M.; Marx, K.D.

    1988-01-01T23:59:59.000Z

    Severe nuclear plant accidents can involve the degradation of the reactor core while the primary coolant system remains pressurized. Molten fuel reaching the lower head of the reactor pressure vessel (RPV) may attack and fail the instrument guide tube penetrations, allowing the tube to be expelled from the vessel. The resulting aperture allows the molten fuel to be ejected into the cavity, followed by the blowdown of the contents of the primary system (high-pressure-melt ejection). Entrainment of the core debris in the cavity by the blowdown gases may cause high-temperature fuel particles to be carried into the containment building. Energy exchange between the particles and the atmosphere may cause heating and pressurizing of the containment (direct containment heating (DCH)). The complex phenomena associated with direct containment heating accident sequences are not well understood. This work describes a series of four experiments that have been performed to study and quantify the processes involved. The data from the experiments are used to guide the development of computer models to describe the response of containments under accident conditions.

  14. Measurement and interpretation of threshold stress intensity factors for steels in high-pressure hydrogen gas.

    SciTech Connect (OSTI)

    Dadfarnia, Mohsen (University of Illinois at Urbana-Champaign, Urbana, IL); Nibur, Kevin A.; San Marchi, Christopher W.; Sofronis, Petros (University of Illinois at Urbana-Champaign, Urbana, IL); Somerday, Brian P.; Foulk, James W., III; Hayden, Gary A. (CP Industries, McKeesport, PA)

    2010-07-01T23:59:59.000Z

    Threshold stress intensity factors were measured in high-pressure hydrogen gas for a variety of low alloy ferritic steels using both constant crack opening displacement and rising crack opening displacement procedures. The sustained load cracking procedures are generally consistent with those in ASME Article KD-10 of Section VIII Division 3 of the Boiler and Pressure Vessel Code, which was recently published to guide design of high-pressure hydrogen vessels. Three definitions of threshold were established for the two test methods: K{sub THi}* is the maximum applied stress intensity factor for which no crack extension was observed under constant displacement; K{sub THa} is the stress intensity factor at the arrest position for a crack that extended under constant displacement; and K{sub JH} is the stress intensity factor at the onset of crack extension under rising displacement. The apparent crack initiation threshold under constant displacement, K{sub THi}*, and the crack arrest threshold, K{sub THa}, were both found to be non-conservative due to the hydrogen exposure and crack-tip deformation histories associated with typical procedures for sustained-load cracking tests under constant displacement. In contrast, K{sub JH}, which is measured under concurrent rising displacement and hydrogen gas exposure, provides a more conservative hydrogen-assisted fracture threshold that is relevant to structural components in which sub-critical crack extension is driven by internal hydrogen gas pressure.

  15. High-pressure neutron diffraction

    SciTech Connect (OSTI)

    Xu, Hongwu [Los Alamos National Laboratory

    2011-01-10T23:59:59.000Z

    This lecture will cover progress and prospect of applications of high-pressure neutron diffraction techniques to Earth and materials sciences. I will first introduce general high-pressure research topics and available in-situ high-pressure techniques. Then I'll talk about high-pressure neutron diffraction techniques using two types of pressure cells: fluid-driven and anvil-type cells. Lastly, I will give several case studies using these techniques, particularly, those on hydrogen-bearing materials and magnetic transitions.

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

    E-Print Network [OSTI]

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

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

    E-Print Network [OSTI]

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

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

    E-Print Network [OSTI]

    Schulze, Jürgen P.

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

  19. External pressure limitations for 0--15 psi storage tanks

    SciTech Connect (OSTI)

    Dib, M.W. [ICF Kaiser Hanford Co., Richland, WA (United States); Shrivastava, H.P. [Westinghouse Hanford Co., Richland, WA (United States)

    1995-12-01T23:59:59.000Z

    Large cylindrical storage tanks are designed in accordance with design rules of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section 3, Subsection NC, Article NC-3900 or American Petroleum Institute (API) Standard 620. Both of these Codes have identical requirements. These Codes provide a limit on the partial vacuum in the gas or vapor space not to exceed 1 oz/in{sup 2} to ensure stability of cylindrical walls against collapse. This criterion seems to be too conservative for the underground double shell storage tanks to be built at Hanford for the Department of Energy. The analysis presented herein shows that the bottom plate of the Hanford tank is the most critical component when an empty tank is subjected to partial vacuum. However, the allowable external pressures for both cylindrical walls and the bottom plate are significantly higher than 1 oz/in{sup 2}. The allowable external pressure for the bottom plate is largely dependent upon the plate uplift considerations which in turns depends on the plate thickness. The large displacement non-linear elastic analyses and the eigenvalue buckling solutions indicate that considerable wrinkling can occur before a snap-through buckling failure occurs.

  20. Pressure effect on ionic conductivity in yttrium-oxide-doped single-crystal zirconium oxide

    SciTech Connect (OSTI)

    Park, E.T.; Park, J.H.

    1998-06-01T23:59:59.000Z

    In this study, the authors investigated the effect of pressure on the ionic conductivity of a 9.5 mol% yttria-stabilized zirconia (YSZ) single crystal. The experiment was conducted in the elastic region, and the oxygen ion transport number was unity (t{sub ion} > 0.99999). A conventional four-probe DC method was used to measure the ionic conductivity of the rectangular-shaped sample under uniaxial pressures up to 600 atm at 750 C in air. Measured ionic conductivity decreased as applied pressure increased. Based on henry Eyring`s absolute reaction rate theory, which states that the calculated activation volume has a positive value ({Delta}V{sup 2} = 2.08 cm{sup 3}/mol of O{sup {minus}2}) for oxygen ion transport in the fluoride cubic lattice, they concluded that the results they obtained could be explained by an oxygen ion transport mechanism. This mechanism can explain the fact that the interionic distance increases during oxygen ion transport from one unit cell to neighboring unit cells.

  1. Theoretical collapse pressures for two pressurized torispherical heads

    SciTech Connect (OSTI)

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

    1995-12-01T23:59:59.000Z

    In order to determine the pressures at which real torispherical heads fail upon a single application of pressure, two heads were pressurized in recent Praxair tests, and displacements and strains were recorded at various locations. In this paper, theoretical results for the two test heads are presented in the form of curves of pressure versus crown deflections, using the available geometry and material parameters. From these curves, limit and collapse pressures are calculated, using procedures permitted by the ASME B and PV Code Section 8/Div.2. These pressures are shown to vary widely, depending on the method and model used to calculate them. The effect of no stress relief on the behavior of the Praxair test heads is also evaluated and found to be of no significance for neither the objectives of the tests nor the objectives of this paper. The results of this paper are submitted as an enhancement to the experimental results recorded during the Praxair tests.

  2. Cradle and pressure grippers

    DOE Patents [OSTI]

    Muniak, John E. (New York, NY)

    2001-01-01T23:59:59.000Z

    A gripper that is designed to incorporate the functions of gripping, supporting and pressure tongs into one device. The gripper has two opposing finger sections with interlocking fingers that incline and taper to form a wedge. The interlocking fingers are vertically off-set so that the opposing finger sections may close together allowing the inclined, tapered tips of the fingers to extend beyond the plane defined by the opposing finger section's engagement surface. The range of motion defined by the interlocking relationship of the finger sections allows the gripper to grab, lift and support objects of varying size and shape. The gripper has one stationary and one moveable finger section. Power is provided to the moveable finger section by an actuating device enabling the gripper to close around an object to be lifted. A lifting bail is attached to the gripper and is supported by a crane that provides vertical lift.

  3. Steam Oxidation at High Pressure

    SciTech Connect (OSTI)

    Holcomb, Gordon R. [NETL; Carney, Casey [URS

    2013-07-19T23:59:59.000Z

    A first high pressure test was completed: 293 hr at 267 bar and 670{degrees}C; A parallel 1 bar test was done for comparison; Mass gains were higher for all alloys at 267 bar than at 1 bar; Longer term exposures, over a range of temperatures and pressures, are planned to provide information as to the commercial implications of pressure effects; The planned tests are at a higher combination of temperatures and pressures than in the existing literature. A comparison was made with longer-term literature data: The short term exposures are largely consistent with the longer-term corrosion literature; Ferritic steels--no consistent pressure effect; Austenitic steels--fine grain alloys less able to maintain protective chromia scale as pressure increases; Ni-base alloys--more mass gains above 105 bar than below. Not based on many data points.

  4. High Pressure Hydrogen Materials Compatibility of Piezoelectric...

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

    Pressure Hydrogen Materials Compatibility of Piezoelectric Films. High Pressure Hydrogen Materials Compatibility of Piezoelectric Films. Abstract: Abstract: Hydrogen is being...

  5. To estimate vapor pressure easily

    SciTech Connect (OSTI)

    Yaws, C.L.; Yang, H.C. (Lamar Univ., Beaumont, TX (USA))

    1989-10-01T23:59:59.000Z

    Vapor pressures as functions of temperature for approximately 700 major organic chemical compounds are given. The tabulation also gives the temperature range for which the data are applicable. Minimum and maximum temperatures are denoted by TMIN and TMAX. The Antoine equation that correlates vapor pressure as a function of temperature is described. A representative comparison of calculated and actual data values for vapor pressure is shown for ethyl alcohol. The coefficient tabulation is based on both literature (experimental data) and estimated values.

  6. Measurements of wall heat (mass) transfer for flow through blockages with round and square holes in a wide rectangular channel

    E-Print Network [OSTI]

    Cervantes, Joel

    2002-01-01T23:59:59.000Z

    ) transfer on the channel wall by 4.7 to 6.3 times, and increased the pressure drop along the test channel by up to almost 490 times that for fully developed turbulent flow through a smooth channel at the same mass flow rates. The blockages with round holes...

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

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

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

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

    E-Print Network [OSTI]

    Peters, Andrew

    2011-08-04T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Brett, Bridget C

    2008-01-01T23:59:59.000Z

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

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

    DOE Patents [OSTI]

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

    1997-12-16T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Bollineni, Prasanthi

    1994-01-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Zhang, Ming

    2009-05-15T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    NONE

    1996-08-01T23:59:59.000Z

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

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

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

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

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

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

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

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

    E-Print Network [OSTI]

    Normore, Stephen S. (Stephen Selwyn)

    2013-01-01T23:59:59.000Z

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

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

    DOE Patents [OSTI]

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

    1997-01-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Kastrup, Christian J.

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

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

    E-Print Network [OSTI]

    Meisner, Joshua Keith

    2009-06-02T23:59:59.000Z

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

  20. Effect of Blending on High-Pressure Laminar Flame Speed Measurements, Markstein Lengths, and Flame Stability of Hydrocarbons

    E-Print Network [OSTI]

    Lowry, William Baugh

    2012-02-14T23:59:59.000Z

    . Hydrocarbon blends of methane, ethane, and propane make up a large portion of natural gas and it has been shown that dimethyl ether can be used as a supplement or in its pure form for gas turbine combustion. Because of this, a fundamental understanding... include the flame speeds for binary blends of methane, ethane, propane, and dimethyl ether performed at elevated pressures, up to 10-atm initial pressure, using a spherically expanding flame in a constant-volume vessel. Also included in this thesis is a...

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

    E-Print Network [OSTI]

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

    2011-01-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2010-01-01T23:59:59.000Z

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

  3. CFD Analyses of Damaged Fuel Inside a Cleaning Vessel

    SciTech Connect (OSTI)

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

    2006-07-01T23:59:59.000Z

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

  4. Electrokinetically pumped high pressure sprays

    DOE Patents [OSTI]

    Schoeniger, Joseph S. (Oakland, CA); Paul, Phillip H. (Livermore, CA); Schoeniger, Luke (Pittsford, NY)

    2002-01-01T23:59:59.000Z

    An electrokinetic pump capable of producing high pressure is combined with a nozzle having a submicron orifice to provide a high pressure spray device. Because of its small size, the device can be contained within medical devices such as an endoscope for delivering biological materials such as DNA, chemo therapeutic agents, or vaccines to tissues and cells.

  5. Possible Pressure Effect for Superconductors

    E-Print Network [OSTI]

    A. Kwang-Hua Chu

    2005-08-30T23:59:59.000Z

    We make an estimate of the possible range of $\\Delta T_c$ induced by high-pressure effects in post-metallic superconductors by using the theory of {\\it extended irreversible/reversible thermodynamics} and Pippard's length scale. The relationship between the increment of the superconducting temperature and the increase of the pressure is parabolic.

  6. Balanced pressure gerotor fuel pump

    DOE Patents [OSTI]

    Raney, Michael Raymond; Maier, Eugen

    2004-08-03T23:59:59.000Z

    A gerotor pump for pressurizing gasoline fuel is capable of developing pressures up to 2.0 MPa with good mechanical and volumetric efficiency and satisfying the durability requirements for an automotive fuel pump. The pump has been designed with optimized clearances and by including features that promote the formation of lubricating films of pressurized fuel. Features of the improved pump include the use of a shadow port in the side plate opposite the outlet port to promote balancing of high fuel pressures on the opposite sides of the rotors. Inner and outer rotors have predetermined side clearances with the clearances of the outer rotor being greater than those of the inner rotor in order to promote fuel pressure balance on the sides of the outer rotor. Support of the inner rotor and a drive shaft on a single bushing with bearing sleeves maintains concentricity. Additional features are disclosed.

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

    SciTech Connect (OSTI)

    Ikeuchi, Hirotomo; Yano, Kimihiko; Kaji, Naoya; Washiya, Tadahiro [Japan Atomic Energy Agency, 4-33 Muramatsu, Tokai-mura, Ibaraki-ken, 319-1194 (Japan); Kondo, Yoshikazu; Noguchi, Yoshikazu [PESCO Co.Ltd. (Korea, Republic of)

    2013-07-01T23:59:59.000Z

    For the decommissioning of the Fukushima-Daiichi Nuclear Power Station (1F), the characterization of fuel-debris in cores of Units 1-3 is necessary. In this study, typical phases of the in-vessel fuel-debris were estimated using a thermodynamic equilibrium (TDE) calculation. The FactSage program and NUCLEA database were applied to estimate the phase equilibria of debris. It was confirmed that the TDE calculation using the database can reproduce the phase separation behavior of debris observed in the Three Mile Island accident. In the TDE calculation of 1F, the oxygen potential [G(O{sub 2})] was assumed to be a variable. At low G(O{sub 2}) where metallic zirconium remains, (U,Zr)O{sub 2}, UO{sub 2}, and ZrO{sub 2} were found as oxides, and oxygen-dispersed Zr, Fe{sub 2}(Zr,U), and Fe{sub 3}UZr{sub 2} were found as metals. With an increase in zirconium oxidation, the mass of those metals, especially Fe{sub 3}UZr{sub 2}, decreased, but the other phases of metals hardly changed qualitatively. Consequently, (U,Zr)O{sub 2} is suggested as a typical phase of oxide, and Fe{sub 2}(Zr,U) is suggested as that of metal. However, a more detailed estimation is necessary to consider the distribution of Fe in the reactor pressure vessel through core-melt progression. (authors)

  8. Mineral matter transformations in a pressurized drop-tube furnace

    SciTech Connect (OSTI)

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

    1992-12-31T23:59:59.000Z

    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.

  9. Mineral matter transformations in a pressurized drop-tube furnace

    SciTech Connect (OSTI)

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

    1992-01-01T23:59:59.000Z

    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.

  10. An Improved Probabilistic Fracture Mechanics Model for Pressurized Thermal Shock

    SciTech Connect (OSTI)

    Dickson, T.L.

    2001-10-29T23:59:59.000Z

    This paper provides an overview of an improved probabilistic fracture mechanics (PFM) model used for calculating the conditional probabilities of fracture and failure of a reactor pressure vessel (RPV) subjected to pressurized-thermal-shock (PTS) transients. The updated PFM model incorporates several new features: expanded databases for the fracture toughness properties of RPV steels; statistical representations of the fracture toughness databases developed through application of rigorous mathematical procedures; and capability of generating probability distributions for RPV fracture and failure. The updated PFM model was implemented into the FAVOR fracture mechanics program, developed at Oak Ridge National Laboratory as an applications tool for RPV integrity assessment; an example application of that implementation is discussed herein. Applications of the new PFM model are providing essential input to a probabilistic risk assessment (PRA) process that will establish an improved technical basis for re-assessment of current PTS regulations by the US Nuclear Regulatory Commission (NRC). The methodology described herein should be considered preliminary and subject to revision in the PTS re-evaluation process.

  11. TENSILE TESTING OF CARBON STEEL IN HIGH PRESSURE HYDROGEN

    SciTech Connect (OSTI)

    Duncan, A; Thad Adams, T; Ps Lam, P

    2007-05-02T23:59:59.000Z

    An infrastructure of new and existing pipelines and systems will be required to carry and to deliver hydrogen as an alternative energy source under the hydrogen economy. Carbon and low alloy steels of moderate strength are currently used in hydrogen delivery systems as well as in the existing natural gas systems. It is critical to understand the material response of these standard pipeline materials when they are subjected to pressurized hydrogen environments. The methods and results from a testing program to quantify hydrogen effects on mechanical properties of carbon steel pipeline and pipeline weld materials are provided. Tensile properties of one type of steel (A106 Grade B) in base metal, welded and heat affected zone conditions were tested at room temperature in air and high pressure (10.34 MPa or 1500 psig) hydrogen. A general reduction in the materials ability to plastically deform was noted in this material when specimens were tested in hydrogen. Furthermore, the primary mode of fracture was changed from ductile rupture in air to cleavage with secondary tearing in hydrogen. The mechanical test results will be applied in future analyses to evaluate service life of the pipelines. The results are also envisioned to be part of the bases for construction codes and structural integrity demonstrations for hydrogen service pipeline and vessels.

  12. Electrokinetic high pressure hydraulic system

    DOE Patents [OSTI]

    Paul, Phillip H. (Livermore, CA); Rakestraw, David J. (Fremont, CA)

    2000-01-01T23:59:59.000Z

    A compact high pressure hydraulic pump having no moving mechanical parts for converting electric potential to hydraulic force. The electrokinetic pump, which can generate hydraulic pressures greater than 2500 psi, can be employed to compress a fluid, either liquid or gas, and manipulate fluid flow. The pump is particularly useful for capillary-base systems. By combining the electrokinetic pump with a housing having chambers separated by a flexible member, fluid flow, including high pressure fluids, is controlled by the application of an electric potential, that can vary with time.

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

    SciTech Connect (OSTI)

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

    2014-03-01T23:59:59.000Z

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

  14. OVERBURDEN PRESSURE AFFECTS FRACTURE APERTURE

    E-Print Network [OSTI]

    Schechter, David S.

    OVERBURDEN PRESSURE AFFECTS FRACTURE APERTURE AND FRACTURE PERMEABILITY IN A FRACTURED RESERVOIR are in integrated reservoir study, reservoir charac- terization, naturally fractured reservoirs, waterflooding in Hydraulically and Naturally Fractured Reservoirs." His research areas include experimental analysis

  15. Electrokinetic high pressure hydraulic system

    DOE Patents [OSTI]

    Paul, Phillip H.; Rakestraw, David J.; Arnold, Don W.; Hencken, Kenneth R.; Schoeniger, Joseph S.; Neyer, David W.

    2003-06-03T23:59:59.000Z

    An electrokinetic high pressure hydraulic pump for manipulating fluids in capillary-based system. The pump uses electro-osmotic flow to provide a high pressure hydraulic system, having no moving mechanical parts, for pumping and/or compressing fluids, for providing valve means and means for opening and closing valves, for controlling fluid flow rate, and manipulating fluid flow generally and in capillary-based systems (microsystems), in particular. The compact nature of the inventive high pressure hydraulic pump provides the ability to construct a micro-scale or capillary-based HPLC system that fulfills the desire for small sample quantity, low solvent consumption, improved efficiency, the ability to run samples in parallel, and field portability. Control of pressure and solvent flow rate is achieved by controlling the voltage applied to an electrokinetic pump.

  16. Electrokinetic high pressure hydraulic system

    DOE Patents [OSTI]

    Paul, Phillip H. (Livermore, CA); Rakestraw, David J. (Fremont, CA); Arnold, Don W. (Livermore, CA); Hencken, Kenneth R. (Pleasanton, CA); Schoeniger, Joseph S. (Oakland, CA); Neyer, David W. (Castro Valley, CA)

    2001-01-01T23:59:59.000Z

    An electrokinetic high pressure hydraulic pump for manipulating fluids in capillary-based systems. The pump uses electro-osmotic flow to provide a high pressure hydraulic system, having no moving mechanical parts, for pumping and/or compressing fluids, for providing valve means and means for opening and closing valves, for controlling fluid flow rate, and manipulating fluid flow generally and in capillary-based systems (Microsystems), in particular. The compact nature of the inventive high pressure hydraulic pump provides the ability to construct a micro-scale or capillary-based HPLC system that fulfills the desire for small sample quantity, low solvent consumption, improved efficiency, the ability to run samples in parallel, and field portability. Control of pressure and solvent flow rate is achieved by controlling the voltage applied to an electrokinetic pump.

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

    E-Print Network [OSTI]

    Neimark, Alexander V.

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

  18. Spray bottle apparatus with pressure multiplying pistons

    DOE Patents [OSTI]

    Moss, Owen R. (Kennewick, WA); Gordon, Norman R. (Kennewick, WA); DeFord, Henry S. (Kennewick, WA)

    1990-01-01T23:59:59.000Z

    The present invention comprises a spray bottle in which the pressure resulting from the gripping force applied by the user is amplified and this increased pressure used in generating a spray such as an aerosol or fluid stream. In its preferred embodiment, the invention includes a high pressure chamber and a corresponding piston which is operative for driving fluid out of this chamber at high pressure through a spray nozzle and a low pressure chamber and a corresponding piston which is acted upon the hydraulic pressure within the bottle resulting from the gripping force. The low pressure chamber and piston are of larger size than the high pressure chamber and piston. The pistons are rigidly connected so that the force created by the pressure acting on the piston in the low pressure chamber is transmitted to the piston in the high pressure chamber where it is applied over a more limited area thereby generating greater hydraulic pressure for use in forming the spray.

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

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2003-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1993-04-01T23:59:59.000Z

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