National Library of Energy BETA

Sample records for walled pressure vessel

  1. Dual shell pressure balanced vessel

    DOEpatents

    Fassbender, Alexander G.

    1992-01-01

    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. Fracture resistance of welded thick-walled high-pressure vessels in power plants. Report No. 2. Approach to evaluating static strength

    SciTech Connect

    Gorynin, I.V.; Filatov, V.M.; Ignatov, V.A.; Timofeev, B.T.; Zvezdin, Yu. I.

    1986-07-01

    The authors examine data on the effect of defects on the fracture resistance of high-pressure vessels and their models obtained within the framework of the HSST program. Results of internal-pressure tests of two types of vessels with a wall thickness of 152 mm made from forgings of steels SA508 and SA533, as well as small vessels with a wall thickness of 11.5 and 23mm made of steel SA533 are shown. The authors state that testing thick-walled welded high-pressure vessels and thin-walled vessels with surface defects of different sizes has demonstrated that there are substantial static-strength reserves in structures designed by existing domestic and foreign standards on the strength of power-plant equipment. A correction was proposed for the presently used method of calculating the resistance of highpressure vessels to brittle fracture that allows for the dimensions of the defects in relation to the type of vessel, the manufacturing technology, and the method of inspection.

  3. Sapphire tube pressure vessel

    DOEpatents

    Outwater, John O. (Cambridge, MA)

    2000-01-01

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

  4. GOLD PRESSURE VESSEL SEAL

    DOEpatents

    Smith, A.E.

    1963-11-26

    An improved seal between the piston and die member of a piston-cylinder type pressure vessel is presented. A layer of gold, of sufficient thickness to provide an interference fit between the piston and die member, is plated on the contacting surface of at least one of the members. (AEC)

  5. Reactor pressure vessel nozzle

    DOEpatents

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

    1994-10-04

    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.

  6. Reactor pressure vessel nozzle

    DOEpatents

    Challberg, Roy C.; Upton, Hubert A.

    1994-01-01

    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.

  7. High pressure storage vessel

    DOEpatents

    Liu, Qiang

    2013-08-27

    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.

  8. Level indicator for pressure vessels

    DOEpatents

    Not Available

    1982-04-28

    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.

  9. Conformable pressure vessel for high pressure gas storage

    DOEpatents

    Simmons, Kevin L.; Johnson, Kenneth I.; Lavender, Curt A.; Newhouse, Norman L.; Yeggy, Brian C.

    2016-01-12

    A non-cylindrical pressure vessel storage tank is disclosed. The storage tank includes an internal structure. The internal structure is coupled to at least one wall of the storage tank. The internal structure shapes and internally supports the storage tank. The pressure vessel storage tank has a conformability of about 0.8 to about 1.0. The internal structure can be, but is not limited to, a Schwarz-P structure, an egg-crate shaped structure, or carbon fiber ligament structure.

  10. Reactor pressure vessel vented head

    DOEpatents

    Sawabe, James K.

    1994-01-11

    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.

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

    DOEpatents

    Bowling, Michael Keith (Blackborough Cullompton, GB)

    2000-09-12

    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.

  12. Reactor pressure vessel vented head

    DOEpatents

    Sawabe, J.K.

    1994-01-11

    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.

  13. Hydrogen storage in insulated pressure vessels

    SciTech Connect

    Aceves, S.M.; Garcia-Villazana, O.

    1998-08-01

    Insulated pressure vessels are cryogenic-capable pressure vessels that can be fueled with liquid hydrogen (LH{sub 2}) or ambient-temperature compressed hydrogen (CH{sub 2}). Insulated pressure vessels offer the advantages of liquid hydrogen tanks (low weight and volume), with reduced disadvantages (lower energy requirement for hydrogen liquefaction and reduced evaporative losses). This paper shows an evaluation of the applicability of the insulated pressure vessels for light-duty vehicles. The paper shows an evaluation of evaporative losses and insulation requirements and a description of the current analysis and experimental plans for testing insulated pressure vessels. The results show significant advantages to the use of insulated pressure vessels for light-duty vehicles.

  14. Lightweight bladder lined pressure vessels

    DOEpatents

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

    1998-08-25

    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.

  15. Lightweight bladder lined pressure vessels

    DOEpatents

    Mitlitsky, Fred; Myers, Blake; Magnotta, Frank

    1998-01-01

    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.

  16. PURE NIOBIUM AS A PRESSURE VESSEL MATERIAL

    SciTech Connect

    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.

    2010-04-09

    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.

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

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Agenda for the International Hydrogen Fuel and Pressure Vessel Forum held Sept. 27-29, 2010, in Beijing, China Forum Agenda: International Hydrogen Fuel and Pressure Vessel Forum ...

  18. Cavity closure arrangement for high pressure vessels

    DOEpatents

    Amtmann, Hans H.

    1981-01-01

    A closure arrangement for a pressure vessel such as the pressure vessel of a high temperature gas-cooled reactor wherein a liner is disposed within a cavity penetration in the reactor vessel and defines an access opening therein. A closure is adapted for sealing relation with an annular mounting flange formed on the penetration liner and has a plurality of radially movable locking blocks thereon having outer serrations adapted for releasable interlocking engagement with serrations formed internally of the upper end of the penetration liner so as to effect high strength closure hold-down. In one embodiment, ramping surfaces are formed on the locking block serrations to bias the closure into sealed relation with the mounting flange when the locking blocks are actuated to locking positions.

  19. Reactor pressure vessel with forged nozzles

    DOEpatents

    Desai, Dilip R.

    1993-01-01

    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.

  20. (Irradiation embrittlement of reactor pressure vessels)

    SciTech Connect

    Corwin, W.R.

    1990-09-24

    The traveler served as a member of the two-man US Nuclear Regulatory Commission sponsored team who visited the Prometey Complex in Leningrad to assess the potential for expanded cooperative research concerning integrity of the primary pressure boundary in commercial light-water reactors. The emphasis was on irradiation embrittlement, structural analysis, and fracture mechanics research for reactor pressure vessels. At the irradiation seminar in Cologne, presentations were made by German, French, Finnish, Russian, and US delegations concerning many aspects of irradiation of pressure vessel steels. The traveler made presentations on mechanisms of irradiation embrittlement and on important aspects of the Heavy-Section Steel Irradiation Program results of irradiated fracture mechanics tests.

  1. Code System to Calculate Pressure Vessel Failure Probabilities.

    Energy Science and Technology Software Center

    2001-03-27

    Version 00 OCTAVIA (Operationally Caused Transients And Vessel Integrity Analysis) calculates the probability of pressure vessel failure from operationally-caused pressure transients which can occur in a pressurized water reactor (PWR). For specified vessel and operating environment characteristics the program computes the failure pressure at which the vessel will fail for different-sized flaws existing in the beltline and the probability of vessel failure per reactor year due to the flaw. The probabilities are summed over themore » various flaw sizes to obtain the total vessel failure probability. Sensitivity studies can be performed to investigate different vessel or operating characteristics in the same computer run.« less

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

    Office of Energy Efficiency and Renewable Energy (EERE)

    Agenda for the International Hydrogen Fuel and Pressure Vessel Forum held Sept. 27-29, 2010, in Beijing, China

  3. Nuclear reactor pressure vessel support system

    DOEpatents

    Sepelak, George R.

    1978-01-01

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

  4. Jam proof closure assembly for lidded pressure vessels

    DOEpatents

    Cioletti, Olisse C.

    1992-01-01

    An expendable closure assembly is provided for use (in multiple units) with a lockable pressure vessel cover along its rim, such as of an autoclave. This assembly is suited to variable compressive contact and locking with the vessel lid sealing gasket. The closure assembly consists of a thick walled sleeve insert for retention in the under bores fabricated in the cover periphery and the sleeve is provided with internal threading only. A snap serves as a retainer on the underside of the sleeve, locking it into an under bore retention channel. Finally, a standard elongate externally threaded bolt is sized for mating cooperation with the so positioned sleeve, whereby the location of the bolt shaft in the cover bore hole determines its compressive contact on the underlying gasket.

  5. Midland reactor pressure vessel flaw distribution

    SciTech Connect

    Foulds, J.R.; Kennedy, E.L.; Rosinski, S.T.

    1993-12-01

    The results of laboratory nondestructive examination (NDE), and destructive cross-sectioning of selected weldment sections of the Midland reactor pressure vessel were analyzed per a previously developed methodology in order to develop a flaw distribution. The flaw distributions developed from the NDE results obtained by two different ultrasonic test (UT) inspections (Electric Power Research Institute NDE Center and Pacific Northwest Laboratories) were not statistically significantly different. However, the distribution developed from the NDE Center`s (destructive) cross-sectioning-based data was found to be significantly different than those obtained through the UT inspections. A fracture mechanics-based comparison of the flaw distributions showed that the cross-sectioning-based data, conservatively interpreted (all defects considered as flaws), gave a significantly lower vessel failure probability when compared with the failure probability values obtained using the UT-based distributions. Given that the cross-sectioning data were reportedly biased toward larger, more significant-appearing (by UT) indications, it is concluded that the nondestructive examinations produced definitively conservative results. In addition to the Midland vessel inspection-related analyses, a set of twenty-seven numerical simulations, designed to provide a preliminary quantitative assessment of the accuracy of the flaw distribution method used here, were conducted. The calculations showed that, in more than half the cases, the analysis produced reasonably accurate predictions.

  6. Reactor pressure vessel structural integrity research

    SciTech Connect

    Pennell, W.E.; Corwin, W.R.

    1995-04-01

    Development continues on the technology used to assess the safety of irradiation-embrittled nuclear reactor pressure vessels (RPVs) containing flaws. Fracture mechanics tests on RPV steel, coupled with detailed elastic-plastic finite-element analyses of the crack-tip stress fields, have shown that (1) constraint relaxation at the crack tip of shallows surface flaws results in increased data scatter but no increase in the lower-bound fracture toughness, (2) the nil ductility temperature (NDT) performs better than the reference temperature for nil ductility transition (RT{sub NDT}) as a normalizing parameter for shallow-flaw fracture toughness data, (3) biaxial loading can reduce the shallow-flaw fracture toughness, (4) stress-based dual-parameter fracture toughness correlations cannot predict the effect of biaxial loading on a shallow-flaw fracture toughness because in-plane stresses at the crack tip are not influenced by biaxial loading, and (5) an implicit strain-based dual-parameter fracture toughness correlation can predict the effect of biaxial loading on shallow-flaw fracture toughness. Experimental irradiation investigations have shown that (1) the irradiation-induced shift in Charpy V-notch vs temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement, and (2) the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties.

  7. International Hydrogen Fuel and Pressure Vessel Forum - Presentations |

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Department of Energy International Hydrogen Fuel and Pressure Vessel Forum - Presentations International Hydrogen Fuel and Pressure Vessel Forum - Presentations These presentations were given at the International Hydrogen Fuel and Pressure Vessel Forum held September 27-29, 2010 in Beijing, China. September 27, 2010 Keynote: Status and Progress in Research, Development and Demonstration of Hydrogen-Compressed Natural Gas Vehicles in China Professor Z.Q. Mao Tsinghua University and Chair of

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

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Department of Energy Fuel and Pressure Vessel Forum 2010 Proceedings International Hydrogen Fuel and Pressure Vessel Forum 2010 Proceedings Proceedings from the forum, which took place in Beijing, China, on September 27-29, 2010. International Hydrogen Fuel and Pressure Vessel Forum 2010 Proceedings (284.25 KB) More Documents & Publications Workshop Notes from ""Compressed Natural Gas and Hydrogen Fuels: Lessons Learned for the Safe Deployment of Vehicles"" Workshop,

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

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    FORUM AGENDA U.S. Department of Energy and Tsinghua University International Hydrogen Fuel and Pressure Vessel Forum Tsinghua University Beijing, PRC September 27 - 29, 2010 The U.S. Department of Energy (DOE) and Tsinghua University in Beijing co-hosted the International Hydrogen Fuel and Pressure Vessel Forum on September 27 - 29, 2010 in Beijing, China. High pressure vessel experts gathered to share lessons learned from CNG and hydrogen vehicle deployments, and to identify R&D needs to

  10. High-pressure Storage Vessels for Hydrogen, Natural Gas and

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Hydrogen-Natural Gas Blends | Department of Energy High-pressure Storage Vessels for Hydrogen, Natural Gas and Hydrogen-Natural Gas Blends High-pressure Storage Vessels for Hydrogen, Natural Gas and Hydrogen-Natural Gas Blends These slides were presented at the International Hydrogen Fuel and Pressure Vessel Forum on September 27 - 29, 2010, in Beijing, China. ihfpv_lynch.pdf (4.21 MB) More Documents & Publications Properties, Behavior and Material Compatibility of Hydrogen, Natural Gas

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

    SciTech Connect

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

    2010-12-09

    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.

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

    Energy Saver

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

  13. Stress analysis and evaluation of a rectangular pressure vessel...

    Office of Scientific and Technical Information (OSTI)

    States)) 42 ENGINEERING; 12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES; PRESSURE VESSELS; STRESS ANALYSIS; RADIOACTIVE WASTE STORAGE;...

  14. Radiation effects on reactor pressure vessel supports

    SciTech Connect

    Johnson, R.E.; Lipinski, R.E.

    1996-05-01

    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.

  15. Tokamak reactor first wall

    DOEpatents

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

    1984-11-20

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

  16. Neutron shielding panels for reactor pressure vessels

    DOEpatents

    Singleton, Norman R.

    2011-11-22

    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.

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

    DOEpatents

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

    2004-03-23

    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.

  18. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    and Development by the Light Water Reactor Sustainability Program | Department of Energy Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program The Department of Energy's (DOE's) Light Water Reactor Sustainability (LWRS) Program is a five year effort that works to

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

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

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

  20. Stress analysis and evaluation of a rectangular pressure vessel

    SciTech Connect

    Rezvani, M.A.; Ziada, H.H.; Shurrab, M.S.

    1992-10-01

    This study addresses structural analysis and evaluation of an abnormal rectangular pressure vessel, designed to house equipment for drilling and collecting samples from Hanford radioactive waste storage tanks. It had to be qualified according to ASME boiler and pressure vessel code, Section VIII; however, it had the cover plate bolted along the long face, a configuration not addressed by the code. Finite element method was used to calculate stresses resulting from internal pressure; these stresses were then used to evaluate and qualify the vessel. Fatigue is not a concern; thus, it can be built according to Section VIII, Division I instead of Division 2. Stress analysis was checked against the code. A stayed plate was added to stiffen the long side of the vessel.

  1. 700 bar Type IV H2 Pressure Vessel Cost Projections

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    bar Type IV H2 Pressure Vessel Cost Projections Brian D. James and Cassidy Houchins Department of Energy Physical-Based Hydrogen Storage Workshop: Identifying Potential Pathways for Lower Cost 700 Bar Storage Vessels 24 August 2016 USCAR, Southfield, MI Outline * System design * Cost analysis methodology * Cost projections * Key opportunities for cost reduction * Recent focus areas - Composites - BOP - Winding time 2 * Overview assumptions & results of latest cost analyses * Categorize

  2. Report of the terawatt laser pressure vessel committee

    SciTech Connect

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

    2000-09-25

    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

  3. Cryogenic Pressure Vessels: Progress and Plans | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Pressure Vessels: Progress and Plans Cryogenic Pressure Vessels: Progress and Plans Presented at the R&D Strategies for Compressed, Cryo-Compressed and Cryo-Sorbent Hydrogen Storage Technologies Workshops on February 14 and 15, 2011. compressed_hydrogen2011_9_aceves.pdf (1.88 MB) More Documents & Publications OEM Perspective on Cryogenic H2 Storage Cryo-Compressed Hydrogen Storage: Performance and Cost Review Proceedings of the 1998 U.S. DOE Hydrogen Program Review: April 28-30, 1998

  4. Threaded insert for compact cryogenic-capable pressure vessels

    DOEpatents

    Espinosa-Loza, Francisco; Ross, Timothy O.; Switzer, Vernon A.; Aceves, Salvador M.; Killingsworth, Nicholas J.; Ledesma-Orozco, Elias

    2015-06-16

    An insert for a cryogenic capable pressure vessel for storage of hydrogen or other cryogenic gases at high pressure. The insert provides the interface between a tank and internal and external components of the tank system. The insert can be used with tanks with any or all combinations of cryogenic, high pressure, and highly diffusive fluids. The insert can be threaded into the neck of a tank with an inner liner. The threads withstand the majority of the stress when the fluid inside the tank that is under pressure.

  5. Lightweight pressure vessels and unitized regenerative fuel cells

    SciTech Connect

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

    1996-12-31

    High specific energy (>400 Wh/kg) energy storage systems have been designed using lightweight pressure vessels in conjunction with unitized regenerative fuel cells (URFCs). URFCs produce power and electrolytically regenerate their reactants using a single stack of reversible cells. Although a rechargeable energy storage system with such high specific energy has not yet been fabricated, we have made progress towards this goal. A primary fuel cell (FC) test rig with a single cell (0.05 ft{sup 2} active area) has been modified and operated reversibly as a URFC. This URFC uses bifunctional electrodes (oxidation and reduction electrodes reverse roles when switching from charge to discharge, as with a rechargeable battery) and cathode feed electrolysis (water is fed from the oxygen side of the cell). Lightweight pressure vessels with state-of-the-art performance factors (burst pressure * internal volume/tank weight = Pb V/W) have been designed and fabricated. These vessels provide a lightweight means of storing reactant gases required for fuel cells (FCs) or URFCs. The vessels use lightweight bladder liners that act as inflatable mandrels for composite overwrap and provide the permeation barrier for gas storage. The bladders are fabricated using materials that are compatible with humidified gases which may be created by the electrolysis of water and are compatible with elevated temperatures that occur during fast fills.

  6. Corrosion fatigue characterization of reactor pressure vessel steels. [PWR; BWR

    SciTech Connect

    Van Der Sluys, W.A.

    1982-12-01

    During routine operation, light water reactor (LWR) pressure vessels are subjected to a variety of transients that result in time-varying stresses. Consequently, fatigue and environmentally-assisted fatigue are mechanisms of growth relevant to flaws in these pressure vessels. To provide a better understanding of the resistance of nuclear pressure vessel steels to these flaw growth processes, fracture mechanics data were generated on the rates of fatigue crack growth for SA508-2 and SA533B-1 steels in both room temperature air and 288/sup 0/C water. Areas investigated were: the relationship of crack growth rate to prior loading history; the effects of loading frequency and R ratio (K/sub min//K/sub max/) on crack growth rate as a function of the stress intensity factor range (..delta..K); transient aspects of the fatigue crack growth behavior; the effect of material chemistry (sulphur content) on fatigue crack; and growth rate; water chemistry effects (high-purity water versus simulated pressurized water reactotr (PWR) primary coolant).

  7. Design Considerations For Blast Loads In Pressure Vessels.

    SciTech Connect

    Rodriguez, E. A.; Nickell, Robert E.; Pepin, J. E.

    2007-01-01

    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.

  8. The coolability limits of a reactor pressure vessel lower head

    SciTech Connect

    Theofanous, T.G.; Syri, S.

    1995-09-01

    Configuration II of the ULPU experimental facility is described, and from a comprehensive set of experiments are provided. The facility affords full-scale simulations of the boiling crisis phenomenon on the hemispherical lower head of a reactor pressure vessel submerged in water, and heated internally. Whereas Configuration I experiments (published previously) established the lower limits of coolability under low submergence, pool-boiling conditions, with Configuration II we investigate coolability under conditions more appropriate to practical interest in severe accident management; that is, heat flux shapes (as functions of angular position) representative of a core melt contained by the lower head, full submergence of the reactor pressure vessel, and natural circulation. Critical heat fluxes as a function of the angular position on the lower head are reported and related the observed two-phase flow regimes.

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

    SciTech Connect

    Wang, Jy-An John

    2010-08-01

    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.

  10. File:06HIGBoilerPressureVesselPermit.pdf | Open Energy Information

    OpenEI (Open Energy Information) [EERE & EIA]

    6HIGBoilerPressureVesselPermit.pdf Jump to: navigation, search File File history File usage Metadata File:06HIGBoilerPressureVesselPermit.pdf Size of this preview: 463 599...

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

    SciTech Connect

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

    1994-10-01

    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).

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

    SciTech Connect

    Jetter, Robert I; Sham, Sam; Swindeman, Robert W

    2011-01-01

    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.

  13. PRESSURIZATION OF CONTAINMENT VESSELS FROM PLUTONIUM OXIDE CONTENTS

    SciTech Connect

    Hensel, S.

    2012-03-27

    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.

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

    SciTech Connect

    Cannell, Gary L.; Huth, Ralph J.; Hallum, Randall T.

    2013-08-26

    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.

  15. Measuring precrack initiation fatigue state in reactor pressure vessel steels

    SciTech Connect

    Nakagawa, Y.; Fukuoka, C.

    1993-08-01

    Identification of a precursor whose measurement could lead to damage avoidance or mitigation in a power plant structure, system or component that is expensive to repair or replace would substantially improve the economics of electrical power generation, The accent is identification of measurable microstructural precursors of those macroscopic damages that limit the useful life of metallic, power plant, pressure boundary components that degrade by fatigue, creep, stress corrosion and embrittlement. This report presents the results of a five year research program to develop an inspection technique to predict when fatigue initiation will occur in pressure vessel steels operating at LWR nuclear plant conditions. The report serves two primary objectives. One is to describe an inspection method that appears to predict when fatigue crack initiation will occur. The second is to illustrate the nature and scope of a program required to validate a microstructural precursor program for the engineering materials and service conditions of utility practice.

  16. Fatigue of weldments in nuclear pressure vessels and piping

    SciTech Connect

    Booker, M.K.; Booker, B.L.P.; Meieran, H.B.; Heuschkel J.

    1980-03-01

    Current (ASME) Code fatigue design rules for nuclear pressure vessels and piping include no special considerations for weldments other than purely geometric factors. Research programs aimed at nonnuclear applications have found weldments to display fatigue behavior inferior to that of pure base material. Available information on fatigue of weldments relevant to nuclear pressure vessels and piping was reviewed and determined changes in the current design rules appear to be dictated by the available information. Information was obtained and summarized and stored in a computerized data management system to facilitate correlation of facts and development of conclusions. Significant areas where development of additional data would substantially increase the ability to judge the adequacy of the current ASME design rules include: a better understanding of the relative importance of crack initiation and crack propagation to fatigue life; additional fatigue data for prototypic commercial weldments, including cumulative damage; properties of repair welds; significance of reheat cracks; quantitative effect of Code-allowable weld defects; and the effect of variable microstructure across the weld joint. Based on the information that is available, there is no evidence that the ASME Code fatigue design procedures need to be changed at this time. The current ASME design procedures, which form the general basis for fatigue evaluation both in the US and abroad are reviewed. Included is a review of various factors that influence the fatigue of weldments and of service experience with nuclear systems regarding fatigue of weldments. Research programs that may contribute to available information are reviewed.

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

    SciTech Connect

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

    2011-06-07

    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.

  18. Structural design, analysis, and code evaluation of an odd-shaped pressure vessel

    SciTech Connect

    Rezvani, M.A.; Ziada, H.H.

    1992-12-01

    This paper is the result of an effort to design, analyze and evaluate a rectangular pressure vessel. Normally pressure vessels are designed in circular or spherical shapes to prevent stress concentrations. In this case, because of operational limitations, the choice of vessels was limited to a rectangular pressure box with a removable cover plate. The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code is used as a guideline for pressure containments whose width or depth exceeds 15.24 cm (6.0 in.) and where pressures will exceed 103.4 KPa (15.0 lbf/in[sup 2]). This evaluation used Section VIII of this Code, hereafter referred to as the Code. The dimensions and working pressure of the subject vessel fall within the pressure vessel category of the Code. The Code design guidelines and rules do not directly apply to this vessel. Therefore, finite-element methodology was used to analyze the pressure vessel, and the Code then was used in qualifying the vessel to be stamped to the Code. Section VIII, Division 1 of the Code was used for evaluation. This action was justified by selecting a material for which fatigue damage would not be a concern. The stress analysis results were then chocked against the Code, and the thicknesses adjusted to satisfy Code requirements. Although not directly applicable, the Code design formulas for rectangular vessels were also considered and presented in this study.

  19. Structural design, analysis, and code evaluation of an odd-shaped pressure vessel

    SciTech Connect

    Rezvani, M.A.; Ziada, H.H.

    1992-12-01

    This paper is the result of an effort to design, analyze and evaluate a rectangular pressure vessel. Normally pressure vessels are designed in circular or spherical shapes to prevent stress concentrations. In this case, because of operational limitations, the choice of vessels was limited to a rectangular pressure box with a removable cover plate. The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code is used as a guideline for pressure containments whose width or depth exceeds 15.24 cm (6.0 in.) and where pressures will exceed 103.4 KPa (15.0 lbf/in{sup 2}). This evaluation used Section VIII of this Code, hereafter referred to as the Code. The dimensions and working pressure of the subject vessel fall within the pressure vessel category of the Code. The Code design guidelines and rules do not directly apply to this vessel. Therefore, finite-element methodology was used to analyze the pressure vessel, and the Code then was used in qualifying the vessel to be stamped to the Code. Section VIII, Division 1 of the Code was used for evaluation. This action was justified by selecting a material for which fatigue damage would not be a concern. The stress analysis results were then chocked against the Code, and the thicknesses adjusted to satisfy Code requirements. Although not directly applicable, the Code design formulas for rectangular vessels were also considered and presented in this study.

  20. Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

    SciTech Connect

    Wang, Jy-An John

    2015-08-01

    The embrittlement trend curve development project for HFIR reactor pressure vessel (RPV) steels was carried out with three major tasks. Which are (1) data collection to match that used in HFIR steel embrittlement trend published in 1994 Journal Nuclear Material by Remec et. al, (2) new embrittlement data of A212B steel that are not included in earlier HFIR RPV trend curve, and (3) the adjustment of nil-ductility-transition temperature (NDTT) shift data with the consideration of the irradiation temperature effect. An updated HFIR RPV steel embrittlement trend curve was developed, as described below. NDTT( C) = 23.85 log(x) + 203.3 log (x) + 434.7, with 2- uncertainty of 34.6 C, where parameter x is referred to total dpa. The developed update HFIR RPV embrittlement trend curve has higher embrittlement rate compared to that of the trend curve developed in 1994.

  1. Crystal Plasticity Model of Reactor Pressure Vessel Embrittlement in GRIZZLY

    SciTech Connect

    Chakraborty, Pritam; Biner, Suleyman Bulent; Zhang, Yongfeng; Spencer, Benjamin Whiting

    2015-07-01

    The integrity of reactor pressure vessels (RPVs) is of utmost importance to ensure safe operation of nuclear reactors under extended lifetime. Microstructure-scale models at various length and time scales, coupled concurrently or through homogenization methods, can play a crucial role in understanding and quantifying irradiation-induced defect production, growth and their influence on mechanical behavior of RPV steels. A multi-scale approach, involving atomistic, meso- and engineering-scale models, is currently being pursued within the GRIZZLY project to understand and quantify irradiation-induced embrittlement of RPV steels. Within this framework, a dislocation-density based crystal plasticity model has been developed in GRIZZLY that captures the effect of irradiation-induced defects on the flow stress behavior and is presented in this report. The present formulation accounts for the interaction between self-interstitial loops and matrix dislocations. The model predictions have been validated with experiments and dislocation dynamics simulation.

  2. D-Zero Central Calorimeter Pressure Vessel and Vacuum Vessel Safety Notes

    SciTech Connect

    Rucinski, R.; Luther, R.; /Fermilab

    1990-10-25

    The relief valve and relief piping capacity was calculated to be 908 sefm air. This exceeds all relieving conditions. The vessel also has a rupture disc with a 2640 scfm air stamped capacity. In order to significantly decrease the amount of time required to fill the cryostats, it is desired to raise the setpoint of the 'operating' relief valve on the argon storage dewar to 20 psig from its existing 16 psig setting. This additional pressure increases the flow to the cryostats and will overwhelm the relief capacity if the temperature of the modules within these vessels is warm enough. Using some conservative assumptions and simple calculations within this note, the maximum average temperature that the modules within each cryostat can be at prior to filling from the storage dewar with liquid argon is at least 290 K. The average temperature of the module mass for any of the three cryostats can be as high as 290 K prior to filling that particular cryostat. This should not be confused with the average temperature of a single type or location which is useful in protecting the modules-not necessarily the vessel itself. A few modules of each type and at different elevations should be used in an average which would account for the different weights of each module. Note that at 290 K, the actual flow of argon through the relief valve and the rupture disk was under the maximum theoretical flows for each relief device. This means that the bulk temperature could actually have been raised to flow argon through the reliefs at their maximum capacity. Therefore, the temperature of 290 K is a conservative value for the calculated flow rate of 12.3 gpm. Safeguards in addition to and used in conjunction with operating procedures shall be implemented in such a way so that the above temperature limitation is not exceeded and such that it is exclusive of the programmable logic controller (PLC). One suggestion is using a toggle switch for each cryostat mounted in the PLC I/O box which

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

    DOEpatents

    Phelps, Richard D.; Upham, Gerald A.; Anderson, Paul M.

    1988-01-01

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

  4. Cleavage Fracture Modeling of Pressure Vessels under Transient Thermo-Mechanical Loading

    SciTech Connect

    Qian, Xudong; Dodds, Robert; Yin, Shengjun; Bass, Bennett Richard

    2008-02-01

    The next generation of fracture assessment procedures for nuclear reactor pressure vessels (RPVs) will combine nonlinear analyses of crack-front response with stochastic treatments of crack size, shape, orientation, location, material properties and thermal-pressure transients. The projected computational demands needed to support stochastic approaches with detailed 3-D, nonlinear stress analyses of vessels containing defects appear well beyond current and near-term capabilities. In the interim, 2-D models become appealing to approximate certain classes of critical flaws in RPVs, and have computational demands within reach for stochastic frameworks. The present work focuses on the capability of 2-D models to provide values for the Weibull stress fracture parameter with accuracy comparable to those from very detailed 3-D models. Weibull stress approaches provide one route to connect nonlinear vessel response with fracture toughness values measured using small laboratory specimens. The embedded axial flaw located in the RPV wall near the cladding-vessel interface emerges from current linear-elastic, stochastic investigations as a critical contributor to the conditional probability of initiation. Three different types of 2-D models reflecting this configuration are subjected to a thermal-pressure transient characteristic of a critical pressurized thermal shock event. The plane-strain, 2-D models include: the modified boundary layer (MBL) model, the middle tension (M(T)) model, and the 2-D RPV model. The 2-D MBL model provides a high quality estimate for the Weibull stress but only in crack-front regions with a positive T-stress. For crack-front locations with low constraint (T-stress < 0), the M(T) specimen provides very accurate Weibull stress values but only for pressure load acting alone on the RPV. For RPVs under a combined thermal-pressure transient, Weibull stresses computed from the 2-D RPV model demonstrate close agreement with those computed from the

  5. Technical Forum Participants at the International Hydrogen Fuel and Pressure Vessel Forum

    Energy.gov [DOE]

    Photo of the Technical Forum Participants at the International Hydrogen Fuel and Pressure Vessel Forum, which was held on September 27–29, 2010, in Beijing, China.

  6. Improved mechanical properties of A 508 class 3 steel for nuclear pressure vessel through steelmaking

    SciTech Connect

    Kim, J.T.; Kwon, H.K.; Kim, K.C.; Kim, J.M.

    1997-12-31

    The present work is concerned with the steelmaking practices which improve the mechanical properties of the A 508 class 3 steel for reactor pressure vessel. Three kinds of steelmaking practices were applied to manufacture the forged heavy wall shell for reactor pressure vessel, that is, the vacuum carbon deoxidation (VCD), modified VCD containing aluminum and silicon-killing. The segregation of the chemical elements through the thickness was quite small so that the variations of the tensile properties at room temperature were small and the anisotropy of the impact properties was hardly observed regardless of the steelmaking practices. The Charpy V-notch impact properties and the reference nil-ductile transition temperature by drop weight test were significantly improved by the modified VCD and silicon-killing as compared with those of the steel by VCD. Moreover, the plane strain fracture toughness values of the materials by modified VCD and silicon-killing practices was much higher than those of the steel by VCD. These were resulted from the fining of austenite grain size. It was observed that the grain size was below 20 {micro}m (ASTM No. 8.5) when using the modified VCD and silicon-killing, compared to 50 {micro}m (ASTM No. 7.0) when using VCD.

  7. Creep of A508/533 Pressure Vessel Steel

    SciTech Connect

    Richard Wright

    2014-08-01

    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

  8. A DISLOCATION-BASED CLEAVAGE INITIATION MODEL FOR PRESSURE VESSEL

    SciTech Connect

    Cochran, Kristine B; Erickson, Marjorie A; Williams, Paul T; Klasky, Hilda B; Bass, Bennett Richard

    2012-01-01

    Efforts are under way to develop a theoretical, multi-scale model for the prediction of fracture toughness of ferritic steels in the ductile-to-brittle transition temperature (DBTT) region that accounts for temperature, irradiation, strain rate, and material condition (chemistry and heat treatment) effects. This new model is intended to address difficulties associated with existing empirically-derived models of the DBTT region that cannot be extrapolated to conditions for which data are unavailable. Dislocation distribution equations, derived from the theories of Yokobori et al., are incorporated to account for the local stress state prior to and following initiation of a microcrack from a second-phase particle. The new model is the basis for the DISlocation-based FRACture (DISFRAC) computer code being developed at the Oak Ridge National Laboratory (ORNL). The purpose of this code is to permit fracture safety assessments of ferritic structures with only tensile properties required as input. The primary motivation for the code is to assist in the prediction of radiation effects on nuclear reactor pressure vessels, in parallel with the EURATOM PERFORM 60 project.

  9. Fracture-mechanics data deduced from thermal-shock and related experiments with LWR pressure-vessel material

    SciTech Connect

    Cheverton, R.D.; Canonico, D.A.; Iskander, S.K.; Bolt, S.E.; Holz, P.P.; Nanstad, R.K.; Stelzman, W.J.

    1982-01-01

    Pressurized water reactors (PWRs) are susceptible to certain types of hypothetical accidents that can subject the reactor pressure vessel to severe thermal shock, that is, a rapid cooling of the inner surface of the vessel wall. The thermal-shock loading, coupled with the radiation-induced reduction in the material fracture toughness, introduces the possibility of propagation of preexistent flaws and what at one time were regarded as somewhat unique fracture-oriented conditions. Several postulated reactor accidents have been analyzed to discover flaw behavior trends; seven intermediate-scale thermal-shock experiments with steel cylinders have been conducted; and corresponding materials characterization studies have been performed. Flaw behavior trends and related fracture-mechanics data deduced from these studies are discussed.

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

    SciTech Connect

    Walter, Matthew; Yin, Shengjun; Stevens, Gary; Sommerville, Daniel; Palm, Nathan; Heinecke, Carol

    2012-01-01

    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

  11. Reactor Pressure Vessel Fracture Analysis Capabilities in Grizzly

    SciTech Connect

    Spencer, Benjamin; Backman, Marie; Chakraborty, Pritam; Hoffman, William

    2015-03-01

    Efforts have been underway to develop fracture mechanics capabilities in the Grizzly code to enable it to be used to perform deterministic fracture assessments of degraded reactor pressure vessels (RPVs). Development in prior years has resulted a capability to calculate -integrals. For this application, these are used to calculate stress intensity factors for cracks to be used in deterministic linear elastic fracture mechanics (LEFM) assessments of fracture in degraded RPVs. The -integral can only be used to evaluate stress intensity factors for axis-aligned flaws because it can only be used to obtain the stress intensity factor for pure Mode I loading. Off-axis flaws will be subjected to mixed-mode loading. For this reason, work has continued to expand the set of fracture mechanics capabilities to permit it to evaluate off-axis flaws. This report documents the following work to enhance Grizzly’s engineering fracture mechanics capabilities for RPVs: • Interaction Integral and -stress: To obtain mixed-mode stress intensity factors, a capability to evaluate interaction integrals for 2D or 3D flaws has been developed. A -stress evaluation capability has been developed to evaluate the constraint at crack tips in 2D or 3D. Initial verification testing of these capabilities is documented here. • Benchmarking for axis-aligned flaws: Grizzly’s capabilities to evaluate stress intensity factors for axis-aligned flaws have been benchmarked against calculations for the same conditions in FAVOR. • Off-axis flaw demonstration: The newly-developed interaction integral capabilities are demon- strated in an application to calculate the mixed-mode stress intensity factors for off-axis flaws. • Other code enhancements: Other enhancements to the thermomechanics capabilities that relate to the solution of the engineering RPV fracture problem are documented here.

  12. Structural characterization of nanoscale intermetallic precipitates in highly neutron irradiated reactor pressure vessel steels

    SciTech Connect

    Sprouster, D. J.; Sinsheimer, J.; Dooryhee, E.; Ghose, S.; Wells, P.; Stan, T.; Almirall, N.; Odette, G. R.; Ecker, L. E.

    2015-10-21

    Here, massive, thick-walled pressure vessels are permanent nuclear reactor structures that are exposed to a damaging flux of neutrons from the adjacent core. The neutrons cause embrittlement of the vessel steel that increases with dose (fluence or service time), as manifested by an increasing temperature transition from ductile-to-brittle fracture. Moreover, extending reactor life requires demonstrating that large safety margins against brittle fracture are maintained at the higher neutron fluence associated with 60 to 80 years of service. Here synchrotron-based x-ray diffraction and small angle x-ray scattering measurements are used to characterize a new class of highly embrittling nm-scale Mn-Ni-Si precipitates that develop in the irradiated steels at high fluence. Furthermore, these precipitates can lead to severe embrittlement that is not accounted for in current regulatory models. Application of the complementarity techniques has, for the very first time, successfully characterized the crystal structures of the nanoprecipitates, while also yielding self-consistent compositions, volume fractions and size distributions.

  13. Structural characterization of nanoscale intermetallic precipitates in highly neutron irradiated reactor pressure vessel steels

    DOE PAGES [OSTI]

    Sprouster, D. J.; Sinsheimer, J.; Dooryhee, E.; Ghose, S.; Wells, P.; Stan, T.; Almirall, N.; Odette, G. R.; Ecker, L. E.

    2015-10-21

    Here, massive, thick-walled pressure vessels are permanent nuclear reactor structures that are exposed to a damaging flux of neutrons from the adjacent core. The neutrons cause embrittlement of the vessel steel that increases with dose (fluence or service time), as manifested by an increasing temperature transition from ductile-to-brittle fracture. Moreover, extending reactor life requires demonstrating that large safety margins against brittle fracture are maintained at the higher neutron fluence associated with 60 to 80 years of service. Here synchrotron-based x-ray diffraction and small angle x-ray scattering measurements are used to characterize a new class of highly embrittling nm-scale Mn-Ni-Si precipitatesmore » that develop in the irradiated steels at high fluence. Furthermore, these precipitates can lead to severe embrittlement that is not accounted for in current regulatory models. Application of the complementarity techniques has, for the very first time, successfully characterized the crystal structures of the nanoprecipitates, while also yielding self-consistent compositions, volume fractions and size distributions.« less

  14. A wall-crawling robot for reactor vessel inspection in advanced reactors

    SciTech Connect

    Spelt, P.F.; Crane, C.; Feng, L.; Abidi, M.; Tosunoglu, S.

    1994-06-01

    A consortium of four universities and the Center for Engineering Systems Advanced Research of the Oak Ridge National Laboratory has designed a prototype wall-crawling robot to perform weld inspection in advanced nuclear reactors. Design efforts for the reactor vessel inspection robot (RVIR) concentrated on the Advanced Liquid Metal Reactor because it presents the most demanding environment in which such a robot must operate. The RVIR consists of a chassis containing two sets of suction cups that can alternately grasp the side of the vessel being inspected, providing both locomotion and steering functions. Sensors include three CCD cameras and a weld inspection device based on new shear-wave technology. The restrictions of the inspection environment presented major challenges to the team. These challenges were met in the prototype, which has been tested in a non-radiation, room-temperature mockup of the robot work environment and shown to perform as expected.

  15. Stress-intensity-factor influence coefficients for semielliptical inner-surface flaws in clad pressure vessels

    SciTech Connect

    Keeney, J.A.; Bryson, J.W.

    1995-12-31

    A problem of particular interest in pressure vessel technology is the calculation of accurate stress-intensity factors for semielliptical surface cracks in cylinders. Computing costs for direct solution techniques can be prohibitive when applied to three-dimensional (3-D) geometries with time-varying boundary conditions such as those associated with pressurized thermal shock. An alternative superposition technique requires the calculation of a set of influence coefficients for a given 3-D crack model that can be superimposed to obtain mode-I stress-intensity factors. This paper presents stress-intensity-factor influence coefficients (SIFICs) for axially and circumferentially oriented finite-length semielliptical inner-surface flaws with aspect ratios (total crack length (2c) to crack depth (a)) of 2, 6, and 10 for clad cylinders having an internal radius to wall thickness (t) ratio of 10. SIFICs are computed for flaw depths in the range of 0.01 {le} a/t {le} 0.5 and two cladding thicknesses. The incorporate of this SIFIC data base in fracture mechanics codes will facilitate the generation of fracture mechanics solutions for a wide range of flaw geometries as may be required in structural integrity assessments of pressurized-water and boiling-water reactors.

  16. Irradiation effects on magnetic properties in neutron and proton irradiated reactor pressure vessel steel

    SciTech Connect

    Park, D.G.; Hong, J.H.; Kim, I.S.; Kim, H.C.

    1999-09-01

    The effects of neutron and proton dose on the magnetic properties of a reactor pressure vessel (RPV) steel were investigated. The coercivity and maximum induction increased in two stages with respect to neutron dose, being nearly constant up to a dose of 1.5 x 10{sup {minus}7} dpa, followed by a rapid increase up to a dose of 1.5 x 10{sup {minus}5} dpa. The coercivity and maximum induction in the proton irradiated specimens also showed a two stage variation with respect to proton dose, namely a rapid increase up to a dose of 0.2 x 10{sup {minus}2} dpa, then a decrease up to 1.2 x 10{sup {minus}2} dpa. The Barkhausen noise (BN) amplitude in neutron irradiated specimens also varied in two stages in a reverse manner, the transition at the same dose of 1.5 x 10{sup {minus}7} dpa. The BN amplitude in proton irradiated specimens decreased by 60% up to 0.2 x 10{sup {minus}2} dpa followed by an increase up to 1.2 x 10{sup {minus}2} dpa. The results were in good accord with the one dimensional domain wall model considering the density of defects and wall energy.

  17. Photoacoustic sample vessel and method of elevated pressure operation

    DOEpatents

    Autrey, Tom; Yonker, Clement R.

    2004-05-04

    An improved photoacoustic vessel and method of photoacoustic analysis. The photoacoustic sample vessel comprises an acoustic detector, an acoustic couplant, and an acoustic coupler having a chamber for holding the acoustic couplant and a sample. The acoustic couplant is selected from the group consisting of liquid, solid, and combinations thereof. Passing electromagnetic energy through the sample generates an acoustic signal within the sample, whereby the acoustic signal propagates through the sample to and through the acoustic couplant to the acoustic detector.

  18. U.S. DOE Hydrogen and Fuel Cell Activities: 2010 International Hydrogen Fuel and Pressure Vessel Forum

    Energy.gov [DOE]

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

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

    SciTech Connect

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

    2010-04-13

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

  20. Micromechanisms of ductile stable crack growth in nuclear pressure vessel steels

    SciTech Connect

    Belcher, W.P.A.; Druce, S.G.

    1981-10-01

    The objective of this work was to investigate the relationship between the micromechanisms of ductile crack growth, the microstructural constituent phases present in nuclear pressure vessel steel, and the observed fracture behavior as determined by impact and fracture mechanics tests. Results from a microstructural and mechanical property comparison of an A508 Class 3 pressurized water reactor nozzle forging cutout and a 150-mm-thick A533B Class 1 plate are reported. The variation of upper-shelf toughness between the two steels and its orientation sensitivity are discussed on the basis of inclusion and precipitate distributions. Inclusion clusters in A533B, deformed to elongated disks in the rolling plane, have a profound effect on short transverse fracture properties. Data derived using the multi-specimen J-integral method to characterize the initiation of ductile crack extension and resistance to stable crack growth are compared with equivalent Charpy results. Results of the J /SUB R/ -curve analyses indicate (1) that the A533B short transverse crack growth resistance is approximately half that observed from transverse and longitudinal specimen orientations, and (2) that the A508 initiation toughness and resistance to stable crack growth are insensitive to position through the forging wall, and are higher than exhibited by A533B at any orientation in the midthickness position.

  1. Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Assessment of High Value Surveillance Materials

    Office of Energy Efficiency and Renewable Energy (EERE)

    The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the...

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

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    These slides were presented at the International Hydrogen Fuel and Pressure Vessel Forum on September 27 - 29, 2010, in Beijing, China. ihfpvlynch.pdf (4.21 MB) More Documents & ...

  3. LWR Pressure-Vessel Surveillance Dosimetry-Improvement Program. Quarterly progress report, April 1982-June 1982

    SciTech Connect

    Guthrie, G.L.; McElroy, W.N.

    1983-01-01

    The Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program (LWR-PV-SDIP) has been established by NRC to improve, test, verify, and standardize the physics-dosimetry-metallurgy, damage correlation, and the associated reactor analysis methods, procedures and data that are used to predict the integrated effect of neutron exposure to LWR pressure vessels and their support structures. Research activities by the Hanford Engineering Development Laboratory and the Oak Ridge National Laboratory are reviewed.

  4. Stress and Fracture Mechanics Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles

    SciTech Connect

    Yin, Shengjun; Bass, Bennett Richard; Stevens, Gary; Kirk, Mark

    2011-01-01

    This paper describes stress analysis and fracture mechanics work performed to assess boiling water reactor (BWR) and pressurized water reactor (PWR) nozzles located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Various RPV nozzle geometries were investigated: 1. BWR recirculation outlet nozzle; 2. BWR core spray nozzle3 3. PWR inlet nozzle; ; 4. PWR outlet nozzle; and 5. BWR partial penetration instrument nozzle. The above nozzle designs 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-license (EOL) to require evaluation as part of establishing the allowed limits on heatup, cooldown, and hydrotest (leak test) conditions. These nozzles analyzed represent one each of the nozzle types potentially requiring evaluation. The purpose of the analyses performed on these nozzle designs was as follows: To model and understand differences in pressure and thermal stress results using a two-dimensional (2-D) axi-symmetric finite element model (FEM) versus a three-dimensional (3-D) FEM for all nozzle types. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated; To verify the accuracy of a selected linear elastic fracture mechanics (LEFM) hand solution for stress intensity factor for a postulated nozzle corner crack for both thermal and pressure loading for all nozzle types; To assess the significance of attached piping loads on the stresses in the nozzle corner region; and To assess the significance of applying pressure on the crack face with respect to the stress intensity factor for a postulated nozzle corner crack.

  5. Utilization of reactor pressure vessel surveillance data in support of aging management

    SciTech Connect

    Mager, T.R.

    1993-12-01

    The majority of pressurized water reactors (PWR) operating in the US have a design basis life of 30 to 40 years. These design basis life estimations were not based on technical studies of material degradation in general but rather on fatigue usage factors for the most part. Recognizing this fact, the subject of operating an existing nuclear steam supply system (NSSS) for longer periods than originally intended has become an important issue worldwide. Radiation embrittlement of the reactor vessel beltline region (the area surrounding the height of the reactor core) is the main concern in extending the operation of the NSSS. Radiation damage, if any, to the reactor pressure vessel material is monitored by a material radiation surveillance program. If the data from the reactor vessel materials surveillance program indicate that the reactor pressure vessel will not meet the rules of the Code of Federal Register (CFR) and various regulatory guides (RG), there are a number of options a utility may take to ensure reactor pressure vessel design life attainment or extension. This paper describes the results from 12 surveillance capsule programs, which encompass four reactor vessel materials and five reactor vessel manufacturers.

  6. International Hydrogen Fuel and Pressure Vessel Forum | Department...

    Office of Environmental Management (EM)

    and certification of Type 3 and Type 4 tanks, pressure relief device testing and ... codes and standards for on-board hydrogen tanks, including the Society of Automotive ...

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

    SciTech Connect

    Odette, G. Robert; Yamamoto, Takuya

    2013-06-17

    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

  8. OVERVIEW OF PRESSURE VESSEL DESIGN CRITERIA FOR INTERNAL DETONATION (BLAST) LOADING

    SciTech Connect

    T. A. DUFFEY; E. A. RODRIGUEZ

    2001-05-01

    Spherical and cylindrical pressure vessels are often used to completely contain the effects of high explosions. These vessels generally fall into two categories. The first includes vessels designed for multiple use ([1]-[6]). Applications of such multiple-use vessels include testing of explosive components and bomb disposal. Because of the multiple-use requirement, response of the vessel is restricted to the elastic range. The second category consists of vessels designed for one-time use only ([7]-[9]). Vessels in this category are typically used to contain accidental explosions and are designed to efficiently utilize the significant plastic energy absorption capacity of ductile materials. Because these vessels may undergo large permanent plastic deformations, they may not be reusable. Ideally one would design a Containment Vessel according to some National or International Consensus Standard, such as the ASME Boiler and Pressure Vessel Code. Unfortunately, however, a number of issues preclude direct use of the ASME Code in its present form to the design of Containment Vessels. These issues are described in Section 2, along with a request for guidance from the PVRC as to a suitable path forward for developing appropriate ASME B&PV design guidance for Containment Vessels. Next, a discussion of the nature of impulsive loading as a result of an internal detonation of the high explosive within a Containment Vessel is described in Section 3. Ductile failure criteria utilized for LANL Containment Vessels are described in Section 4. Finally, brittle fracture criteria currently utilized by LANL are presented in Section 5. This memo is concluded with a brief summary of results and an appeal to PVRC to recommend and develop an appropriate path forward (Section 6). This path forward could be of a short-term specialized nature (e.g., Code Case) for specific guidance regarding design of the LANL Containment Vessels; a long-term development of a general design approach

  9. Creep rupture failure of reactor pressure vessel lower head during severe accidents

    SciTech Connect

    Shah, V.N.

    1986-09-12

    A creep rupture analysis of a reactor pressure vessel lower head subjected to high temperature and pressure during a severe accident is presented herein. Preliminary results show that creep rupture failure will take place at temperatures of approximately 1000/sup 0/K. These temperatures are significantly lower than the melting temperatures of steel (1700 to 1800/sup 0/K) when the system pressure is high (6.8 to 13.8 MPa).

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

    SciTech Connect

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

    1981-01-01

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

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

    SciTech Connect

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

    2014-10-28

    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.

  12. Fast neutron fluence of yonggwang nuclear unit 1 reactor pressure vessel

    SciTech Connect

    Yoo, C.; Km, B.; Chang, K.; Leeand, S.; Park, J.

    2006-07-01

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

  13. Development of design criteria for a high pressure vessel construction code

    SciTech Connect

    Mraz, G.J.

    1987-05-01

    Out of concern for public safety, most legal jurisdictions now require unfired pressure vessel construction to comply with the ASME Boiler and Pressure Vessel Code. Because the present two divisions of Section VIII of that Code are not well suited for high pressure design, a new division is needed. The currently anticipated main design criteria of the proposed division are full plastic flow or full overstrain pressure, stress intensity in the bore, fatigue, and fracture mechanics. The rules are expected to allow better utilization of high strength steels already included in the present Section VIII. At the same time materials of even higher strength are introduced. The benefits of compressive prestress are recognized. Construction methods allowing it's achievement, such as autofrettage, shrink fitting and wire winding are included. Reasons for selection of the criteria are given.

  14. Reactor Pressure Vessel Temperature Analysis for Prismatic and Pebble-Bed VHTR Designs

    SciTech Connect

    H. D. Gougar; C. B. Davis

    2006-04-01

    Analyses were performed to determine maximum temperatures in the reactor pressure vessel for two potential Very-High Temperature Reactor (VHTR) designs during normal operation and during a depressurized conduction cooldown accident. The purpose of the analyses was to aid in the determination of appropriate reactor vessel materials for the VHTR. The designs evaluated utilized both prismatic and pebble-bed cores that generated 600 MW of thermal power. Calculations were performed for fluid outlet temperatures of 900 and 950 °C, corresponding to the expected range for the VHTR. The analyses were performed using the RELAP5-3D and PEBBED-THERMIX computer codes. Results of the calculations were compared with preliminary temperature limits derived from the ASME pressure vessel code.

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

    SciTech Connect

    Rana, M.D.; Zawierucha, R.

    1995-12-01

    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.

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

    SciTech Connect

    Nguyen, Ba Nghiep; Simmons, Kevin L.

    2013-07-22

    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].

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

    DOEpatents

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

    1985-03-01

    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.

  18. Fabrication Flaw Density and Distribution in the Repairs of Reactor Pressure Vessels

    SciTech Connect

    Schuster, George J.; Doctor, Steven R.; Simonen, Fredric A.

    2006-02-15

    The Pacific Northwest National Laboratory (PNNL) is developing a generalized flaw size and density distribution for the population of U.S. reactor pressure vessels (RPVs). The purpose of the generalized flaw distribution is to predict vessel specific flaw rates for use in probabilistic fracture mechanics calculations that estimate vessel failure probability. Considerable progress has been made on the construction of an engineering data base of fabrication flaws in U.S. nuclear RPVs. The fabrication processes and product forms used to construct U.S. RPVs are represented in the data base. A validation methodology has been developed for characterizing the flaws for size, shape, orientation, and composition. The relevance of construction records has been established for describing fabrication processes and product forms. The fabrication flaws were detected in material removed from cancelled nuclear power plants using high sensitivity nondestructive ultrasonic testing, and validated by other nondestructive evaluation (NDE) techniques, and complemented by destructive testing. This paper describes research that has generated data on welding flaws, which indicated that the largest flaws occur in weld repairs. Recent research results confirm that repair flaws are complex in composition and may include cracks on the repair ends. Section III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code for nuclear power plant components requires radiographic examinations (RT) of welds and requires repairs for RT indications that exceed code acceptable sizes. PNNL has previously obtained the complete construction records for two RPVs. Analysis of these records show a significant change in repair frequency.

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

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    and Pressure Vessel R&D under the Hydrogen Regional Infrastructure Program In Pennsylvania Kevin L. Klug, Ph.D. 25 September 2007 DOE Hydrogen Pipeline Working Group Meeting, Aiken, SC Acknowledgments * This material is based upon work supported by the Department of Energy under Award Number DE-FC36-04GO14229 * Partners - Savannah River National Laboratory (SRNL) - HyPerComp Engineering Inc. (HEI) - American Society Of Mechanical Engineers (ASME) - Pipeline Working Group (PWG) Program

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

    DOEpatents

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

    2013-01-15

    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.

  1. Fatigue crack growth rates in pressure vessel and piping steels in LWR environments: Final report

    SciTech Connect

    Cullen, W.H.

    1987-03-01

    The measurement of fatigue crack growth rates for pressure vessel and piping steels in high-temperature, pressurized water has been carried out using compact fracture specimens. Over the last ten years, the programs sponsored by the NRC and carried out at the Naval Research Laboratory and Materials Engineering Associates have provided data for over three hundred tests of these specimens, which have been published in a series of NUREG topical reports and annual reports. This is the final report in this series and describes briefly the significant findings of the program, reports on the most recent data which have been acquired, and indicates some directions for future research in this area.

  2. REACTOR PRESSURE VESSEL TEMPERATURE ANALYSIS OF CANDIDATE VERY HIGH TEMPERATURE REACTOR DESIGNS

    SciTech Connect

    Hans D. Gougar; Cliff B. Davis; George Hayner; Kevan Weaver

    2006-10-01

    Analyses were performed to determine maximum temperatures in the reactor pressure vessel for two potential Very-High Temperature Reactor (VHTR) designs during normal operation and during a depressurized conduction cooldown accident. The purpose of the analyses was to aid in the determination of appropriate reactor vessel materials for the VHTR. The designs evaluated utilized both prismatic and pebble-bed cores that generated 600 MW of thermal power. Calculations were performed for fluid outlet temperatures of 900 and 950 °C, corresponding to the expected range for the VHTR. The analyses were performed using the RELAP5-3D and PEBBED-THERMIX computer codes. Results of the calculations were compared with preliminary temperature limits derived from the ASME pressure vessel code. Because PEBBED-THERMIX has not been extensively validated, confirmatory calculations were also performed with RELAP5-3D for the pebble-bed design. During normal operation, the predicted axial profiles in reactor vessel temperature were similar with both codes and the predicted maximum values were within 2 °C. The trends of the calculated vessel temperatures were similar during the depressurized conduction cooldown accident. The maximum value predicted with RELAP5-3D during the depressurized conduction cooldown accident was about 40 °C higher than that predicted with PEBBED. This agreement is considered reasonable based on the expected uncertainty in either calculation. The differences between the PEBBED and RELAP5-3D calculations were not large enough to affect conclusions concerning comparisons between calculated and allowed maximum temperatures during normal operation and the depressurized conduction cooldown accident.

  3. Update to Risk-Informed Pressurized Water Reactor Vessel 10 to 20 Year Inspection Interval Extension

    SciTech Connect

    Palm, Nathan A.; Bishop, Bruce A.; Boggess, Cheryl L.

    2006-07-01

    The Pressurized Water Reactor Owners Group (formerly the Westinghouse Owners Group (WOG)) methodology for extending the inservice inspection interval for welds in pressurized water reactor (PWR) reactor pressure vessel (RPV) was introduced as ICONE12-49429. The paper presented a risk informed basis for extending the interval between inspections from the current interval of 10 years to 20 years. In the paper presented at ICONE-12, results of pilot studies on typical Westinghouse and Combustion Engineering Nuclear Steam Supply System (NSSS) designs of PWR vessels showed that the change in risk associated with the proposed inspection interval extension was within the guidelines specified in the United States Nuclear Regulatory Commission (NRC) Regulatory Guide 1.174 for an acceptably small change in risk. Since the methodology was originally presented, the evaluation has been updated to incorporate the latest changes in the NRC Pressurized Thermal Shock (PTS) Risk Reevaluation Program and expanded to include the Babcock and Wilcox NSSS RPV design. The results of these evaluations demonstrate that the proposed RPV inspection interval extension remains a viable option for the industry. The updates to the methodology and input, pilot plant evaluations, results, process for demonstrating applicability of the pilot plant analysis to non-pilot lead plants and lessons learned from the evaluations performed are summarized in this paper. (authors)

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

    SciTech Connect

    GJ Schuster, FA Simonen, SR Doctor

    2008-04-01

    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.

  5. Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges

    Office of Energy Efficiency and Renewable Energy (EERE)

    The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR...

  6. Prestressed-concrete pressure vessels and their applicability to advanced-energy-system concepts

    SciTech Connect

    Naus, D.J

    1983-01-01

    Prestressed concrete pressure vessels (PCPVs) are, in essence, spaced steel structures since their strength is derived from a multitude of steel elements made up of deformed reinforcing bars and prestressing tendons which are present in sufficient quantities to carry tension loads imposed on the vessel. Other major components of a PCPV include the concrete, liner and cooling system, and insulation. PCPVs exhibit a number of advantages which make them ideally suited for application to advanced energy concepts: fabricability in virtually any size and shape using available technology, improved safety, reduced capital costs, and a history of proven performance. PCPVs have many applications to both nuclear- and non-nuclear-based energy systems concepts. Several of these concepts will be discussed as well as the research and development activities conducted at ORNL in support of PCPV development.

  7. Measurement of Fatigue Crack Growth Relationships in Hydrogen Gas for Pressure Swing Adsorber Vessel Steels

    SciTech Connect

    Somerday, Brian P.; Barney, Monica

    2014-12-04

    We measured the hydrogen-assisted fatigue crack growth rates (da/dN) for SA516 Grade 70 steel as a function of stress-intensity factor range (ΔK) and load-cycle frequency to provide life-prediction data relevant to pressure swing adsorber (PSA) vessels. For ΔK values up to 18.5 MPa m1/2, the baseline da/dN versus ΔK relationship measured at 1Hz in 2.8 MPa hydrogen gas represents an upper bound with respect to crack growth rates measured at lower frequency. However, at higher ΔK values, we found that the baseline da/dN data had to be corrected to account for modestly higher crack growth rates at the lower frequencies relevant to PSA vessel operation.

  8. Measurement of Fatigue Crack Growth Relationships in Hydrogen Gas for Pressure Swing Adsorber Vessel Steels

    DOE PAGES [OSTI]

    Somerday, Brian P.; Barney, Monica

    2014-12-04

    We measured the hydrogen-assisted fatigue crack growth rates (da/dN) for SA516 Grade 70 steel as a function of stress-intensity factor range (ΔK) and load-cycle frequency to provide life-prediction data relevant to pressure swing adsorber (PSA) vessels. For ΔK values up to 18.5 MPa m1/2, the baseline da/dN versus ΔK relationship measured at 1Hz in 2.8 MPa hydrogen gas represents an upper bound with respect to crack growth rates measured at lower frequency. However, at higher ΔK values, we found that the baseline da/dN data had to be corrected to account for modestly higher crack growth rates at the lower frequenciesmore » relevant to PSA vessel operation.« less

  9. D-Zero Central Calorimeter Technical Appendix to Cryogenic Pressure Vessels

    SciTech Connect

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

    1990-11-19

    DO (D Zero) is a large Liquid Argon (LAr) HEP Calorimeter designed to function in the laboratories P-Pbar collider at the DO section of the Tevatron accelerator. It contains 5,000 gls. of LAr in the CC cryostat, and 3,000 gls. in each of two, a north and south, EC cryostats. These low pressure vessels are filled with detector modules built of stainless steel, copper and depleted uranium. The LAr functions as the ionization medium, and the spatial and temporal of the collection of the charge of the electrons produced signals the passsage of charged particles. The collection of these charges in 4 pi is related to the energy of the particles, and their measurement is called calorimetry. The contained LAr (T=90K) is isolated from the ambient temperatures in specially designed, vacuum and superinsulated, vessels (cryostats) provided with liquid nitrogen, heat of vaporization, cooling.

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

    SciTech Connect

    Rana, M.D. . Research and Development Dept.)

    1994-08-01

    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.

  11. Containment of explosions in spherical vessels

    SciTech Connect

    Duffey, T.A.; Greene, J.M. ); Baker, W.E. . Dept. of Mechanical Engineering); Lewis, B.B. )

    1992-01-01

    A correlation of the experimentally recorded dynamic response of a spherical containment vessel with theoretical finite element calculations is presented. Three experiments were performed on the 6-ft-diameter steel vessel using centrally located 12-lb. and 40-lb. high explosive charges. Pressure-time loading on the inner wall of the vessel was recorded for each test using pressure transducers. Resulting dynamic response of the vessel was recorded for each test using strain gages mounted at selected locations on the outer surface of the vessel. Response of the vessel was primarily elastic. A finite element model of the vessel was run using DYNA3D, a dynamic structural analysis code. Pressure loading for the finite element model was based on results from a one-dimensional reactive hydrodynamics code. Correlations between experiments and analysis were generally good for the tests for frequency and strain magnitude at most locations. Comparisons of experimental and calculated pressure-time histories were less satisfactory.

  12. Containment of explosions in spherical vessels

    SciTech Connect

    Duffey, T.A.; Greene, J.M.; Baker, W.E.; Lewis, B.B.

    1992-12-31

    A correlation of the experimentally recorded dynamic response of a spherical containment vessel with theoretical finite element calculations is presented. Three experiments were performed on the 6-ft-diameter steel vessel using centrally located 12-lb. and 40-lb. high explosive charges. Pressure-time loading on the inner wall of the vessel was recorded for each test using pressure transducers. Resulting dynamic response of the vessel was recorded for each test using strain gages mounted at selected locations on the outer surface of the vessel. Response of the vessel was primarily elastic. A finite element model of the vessel was run using DYNA3D, a dynamic structural analysis code. Pressure loading for the finite element model was based on results from a one-dimensional reactive hydrodynamics code. Correlations between experiments and analysis were generally good for the tests for frequency and strain magnitude at most locations. Comparisons of experimental and calculated pressure-time histories were less satisfactory.

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

    SciTech Connect

    Habil; Ziliukas, A.

    1997-04-01

    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.

  14. Supplementary neutron-flux calculations for the ORNL Pool Critical Assembly Pressure Vessel Facility

    SciTech Connect

    Maudlin, P.J.; Maerker, R.E.

    1982-01-01

    A three-dimensional Monte Carlo calculation using the MORSE code was performed to validate a procedure previously adopted in the ORNL discrete ordinate analysis of measurements made in the ORNL Pool Critical Assembly Pressure Vessel Facility. The results of these flux calculations agree, within statistical undertainties of about 5%, with those obtained from a discrete ordinate analysis employing the same procedure. This study therefore concludes that the procedure for combining several one- and two-dimensional discrete ordinate calculations into a three-dimensional flux is sufficiently accurate that it does not account for the existing discrepancies observed between calculations and measurements in this facility.

  15. Review of the Palisades pressure vessel accumulated fluence estimate and of the least squares methodology employed

    SciTech Connect

    Griffin, P.J.

    1998-05-01

    This report provides a review of the Palisades submittal to the Nuclear Regulatory Commission requesting endorsement of their accumulated neutron fluence estimates based on a least squares adjustment methodology. This review highlights some minor issues in the applied methodology and provides some recommendations for future work. The overall conclusion is that the Palisades fluence estimation methodology provides a reasonable approach to a {open_quotes}best estimate{close_quotes} of the accumulated pressure vessel neutron fluence and is consistent with the state-of-the-art analysis as detailed in community consensus ASTM standards.

  16. Welding and brazing qualifications (supplement to ASME Boiler and Pressure Vessel Code, Section IX)

    SciTech Connect

    Not Available

    1981-11-01

    This standard supplements the requirements of the 1980 edition of the ASME Boiler and Pressure Vessel Code (the Code), Section IX. When this standard is invoked or referenced, the applicable subsections of Section IX of the Code are also invoked or referenced. The paragraph numbers in this standard apply only to the 1980 edition of Section IX and its addenda. The user of this standard is responsible for obtaining and applying the edition and revision of this standard that supplement the edition and addenda of Section IX that are in legal effect at the time of use.

  17. Welding and brazing qualifications (supplement to ASME boiler and pressure vessel code, Section IX)

    SciTech Connect

    Not Available

    1980-01-01

    This standard supplements the requirements of the 1977 edition of the ASME Boiler and Pressure Vessel Code (the Code), Section IX. When this standard is invoked or referenced, the applicable subsections of Section IX of the Code are also invoked or referenced. The paragraph numbers apply only to the 1977 edition of Section IX and its addenda. The user of this standard is responsible for obtaining and applying the edition and revision of this standard that supplement the edition and Addenda of Section IX that are in legal effect at the time of use.

  18. HFIR (High Flux Isotope Reactor) pressure vessel and structural components materials surveillance program: Supplement 1

    SciTech Connect

    Cheverton, R.D.; McGinty, D.M.; McWherter, J.R.; Nanstad, R.K.

    1987-10-01

    Extending the life of the HFIR vessel by the proposed 10 effective full-power years is contingent upon a continuation of the materials surveillance program and the application of hydrostatic proof testing. As a part of the surveillance program, Charpy V-notch (CVN) specimens of shell, weld and nozzle materials are installed adjacent to the inner surface of the vessel and are removed periodically for testing to determine the radiation-induced increase in the nil-ductility transition temperature. Hydro testing is conducted to prove that a critical combination of flaw size, stress and fracture toughness does not exist. Information from the materials surveillance program is used in a fracture mechanics analysis to confirm that the hydro-test pressure being applied is appropriate for the desired life extension of the vessel. This report specifies (1) the number, type, location and schedule for removal-testing of the CVN specimens for the continuing materials surveillance program, and (2) the procedures and test conditions for the hydro test.

  19. Effect of compression on individual pressure vessel nickel/hydrogen components

    SciTech Connect

    Manzo, M.A.; Perez-Davis, M.E.

    1988-08-01

    Compression tests were performed on representative Individual Pressure Vessel (IPV) Nickel/Hydrogen cell components in an effort to better understand the effects of force on component compression and the interactions of components under compression. It appears that the separator is the most easily compressed of all of the stack components. It will typically partially compress before any of the other components begin to compress. The compression characteristics of the cell components in assembly differed considerably from what would be predicted based on individual compression characteristics. Component interactions played a significant role in the stack response to compression. The results of the compression tests were factored into the design and selection of Belleville washers added to the cell stack to accommodate nickel electrode expansion while keeping the pressure on the stack within a reasonable range of the original preset.

  20. Initial Probabilistic Evaluation of Reactor Pressure Vessel Fracture with Grizzly and Raven

    SciTech Connect

    Spencer, Benjamin; Hoffman, William; Sen, Sonat; Rabiti, Cristian; Dickson, Terry; Bass, Richard

    2015-10-01

    The Grizzly code is being developed with the goal of creating a general tool that can be applied to study a variety of degradation mechanisms in nuclear power plant components. The first application of Grizzly has been to study fracture in embrittled reactor pressure vessels (RPVs). Grizzly can be used to model the thermal/mechanical response of an RPV under transient conditions that would be observed in a pressurized thermal shock (PTS) scenario. The global response of the vessel provides boundary conditions for local models of the material in the vicinity of a flaw. Fracture domain integrals are computed to obtain stress intensity factors, which can in turn be used to assess whether a fracture would initiate at a pre-existing flaw. These capabilities have been demonstrated previously. A typical RPV is likely to contain a large population of pre-existing flaws introduced during the manufacturing process. This flaw population is characterized stastistically through probability density functions of the flaw distributions. The use of probabilistic techniques is necessary to assess the likelihood of crack initiation during a transient event. This report documents initial work to perform probabilistic analysis of RPV fracture during a PTS event using a combination of the RAVEN risk analysis code and Grizzly. This work is limited in scope, considering only a single flaw with deterministic geometry, but with uncertainty introduced in the parameters that influence fracture toughness. These results are benchmarked against equivalent models run in the FAVOR code. When fully developed, the RAVEN/Grizzly methodology for modeling probabilistic fracture in RPVs will provide a general capability that can be used to consider a wider variety of vessel and flaw conditions that are difficult to consider with current tools. In addition, this will provide access to advanced probabilistic techniques provided by RAVEN, including adaptive sampling and parallelism, which can dramatically

  1. Neutron damage in reactor pressure-vessel steel examined with positron annihilation lifetime spectroscopy

    SciTech Connect

    Cumblidge, S.E.; Motta, A.T.; Catchen, G.L.

    1997-11-01

    The authors have used positron annihilation lifetime spectroscopy to study the development of damage and annealing behavior of neutron-irradiated reactor pressure-vessel steels. They irradiated samples of ASTM A508 nuclear reactor pressure-vessel steel to fast neutron fluences of up to 10{sup 17} n/cm{sup 2}, and they examined these samples using positron annihilation lifetime spectroscopy (PALS) to study the effects of neutron damage in the steels on positron lifetimes. Non-irradiated samples show two positron lifetimes: a 110 ps component corresponding to annihilations in the bulk material, and a 165 ps lifetime corresponding to annihilations in dislocation defects. The irradiated samples show an additional lifetime component of 300 ps in the PAl spectra and an increase in the proportion of annihilations with a 165 ps lifetime, suggesting that vacancies and vacancy clusters are present in the material after room temperature irradiation. The samples were then annealed to temperatures ranging from 210 C to 450 C. The positron lifetimes introduced by neutron damage disappear after annealing the samples at 280 C.

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

    SciTech Connect

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

    2012-09-01

    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

  3. Example calculations illustrating methods for analyzing ductile flaw stability in nuclear pressure vessels

    SciTech Connect

    Merkle, J.G.; Johnson, R.E.

    1983-05-01

    This report contains example calculations of ductile flaw instability stresses for hypothetical flaws in nuclear pressure vessels. For comparison, three different methods of estimating upper shelf toughness as a function of Charpy impact energy were used, namely: a power law R-curve correlation, the Rolfe-Novak correlation and the Paris J/sub 50/ correlation. All three methods were used in LEFM calculations including a plastic zone size correction, and gave similar results, with the Paris J/sub 50/ method being the most conservative at low Charpy upper shelf energy levels. Safety factors based on the tearing modulus ratio T/sub mat//T/sub appl/ can exceed those based on load by considerable amounts and use of them at this time is not recommended. The use of resistance curve data obtained from actual vessel material test specimens is recommended over the use of correlations. Furthermore, evaluation of a recently proposed modified crack extension adjustment procedure for R-curve data, which is not overconservative, is recommended.

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

    SciTech Connect

    DeRenzis, A.; Kochkin, V.

    2013-01-01

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

  5. Irradiation effects in low-alloy reactor pressure vessel steels (Heavy-Section Steel Technology Program Series 4 and 5)

    SciTech Connect

    Berggren, R.G.; McGowan, J.J.; Menke, B.H.; Nanstad, R.K.; Thoms, K.R.

    1984-01-01

    Multiple testing is done at two laboratories of typical nuclear pressure vessel materials (both irradiated and unirradiated) and statistical analyses of the test results. Multiple tests are conducted at each of several test temperatures for each material, standard deviations are determined, and results from the two laboratories are compared. The Fourth Heavy-Section Steel Technology (HSST) Irradiation Series, almost completed, was aimed at elastic-plastic and fully plastic fracture toughness of low-copper weldments (current practice welds). A typical nuclear pressure vessel plate steel was included for statistical purposes. The Fifth HSST Irradiation Series, now in progress, is aimed at determining the shape of the K/sub IR/ curve after significant radiation-induced shift of the transition temperatures. This series includes irradiated test specimens of thicknesses up to 100 mm and weldment compositions typical of early nuclear power reactor pressure vessel welds.

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

    DOEpatents

    Challberg, Roy C.; Gou, Perng-Fei; Chu, Cherk Lam; Oliver, Robert P.

    1999-01-01

    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.

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

    DOEpatents

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

    1999-07-27

    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.

  8. Tensile properties of irradiated nuclear grade pressure vessel welds for the third HSST irradiation series

    SciTech Connect

    McGowan, J.J.

    1985-03-01

    The Heavy Section Steel Technology (HSST) Program conducted a series of experiments to investigate the effect of neutron irradiation on the fracture toughness of nuclear pressure vessel materials. Four welds of A 508 class 2 steel were examined in this Third HSST Irradiation Series. The welds were fabricated according to ''early'' (pre-1972) light-water reactor weld practice (i.e., copper-coated electrodes). As part of this study, tensile properties were measured after irradiation to 2 to 10 x 10/sup 22/ neutrons/m/sup 2/ (E > 1 MeV) at temperatures between 250 and 290/sup 0/C. Strength properties of all four welds increased with exposure to irradiation. Yield strength was more sensitive to irradiation than was ultimate strength. Tensile ductility was not affected significantly by exposure to irradiation.

  9. The effect of an anisotropic pressure of thermal particles on resistive wall mode stability

    SciTech Connect

    Berkery, J. W. Sabbagh, S. A.; Betti, R.; Guazzotto, L.; Manickam, J.

    2014-11-15

    The effect of an anisotropic pressure of thermal particles on resistive wall mode stability in tokamak fusion plasmas is derived through kinetic theory and assessed through calculation with the MISK code [B. Hu et al., Phys. Plasmas 12, 0?57301 (2005)]. The fluid anisotropy is treated as a small perturbation on the plasma equilibrium and modeled with a bi-Maxwellian distribution function. A complete stability treatment without an assumption of high frequency mode rotation leads to anisotropic kinetic terms in the dispersion relation in addition to anisotropy corrections to the fluid terms. With the density and the average pressure kept constant, when thermal particles have a higher temperature perpendicular to the magnetic field than parallel, the fluid pressure-driven ballooning destabilization term is reduced. Additionally, the stabilizing kinetic effects of the trapped thermal ions can be enhanced. Together these two effects can lead to a modest increase in resistive wall mode stability.

  10. Overview of new rules and recent changes in ASME code, Section VIII, pressure vessels

    SciTech Connect

    Farr, J.R.

    1995-12-01

    In this presentation, some of the new rules and recent changes to the ASME Boiler and Pressure Vessel Code, Section 8, Divisions 1 and 2, are reviewed. On July 1, 1995, the 1995 Edition of the ASME Code was issued. This 1995 Edition incorporates those items which were added of changed in the 1992, 1993, and 1994 Addenda to the 1992 Edition of the Code. The 1995 Edition contains no new items which were not included in the previous edition and three addenda. With the possibility of an extended time before some of the new rules are able to appear in the addenda, the recent trend is to put the rules in Code Cases which are approved earlier. Consequently, it is necessary to review new Code Cases as well as Code changes. Updates continue for impact requirements for standard components as well as for materials other than UCS, carbon steel and low alloys. Extensive changes have been made for UHA, high-alloy, materials regarding impact requirements. Example problems have been revised to include these effects. Significant changes are reviewed.

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

    SciTech Connect

    McHenry, H.I.; Alers, G.A.

    1998-03-01

    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.

  12. Investigation of mechanisms of environmentally accelerated crack growth in reactor pressure vessel steels

    SciTech Connect

    Kobayashi, T. )

    1990-08-01

    The fracture surface topography analysis (FRASTA) technique was applied to several pressure vessel steels tested under simulated PWR service conditions in attempting to establish the mechanism underlying environmentally accelerated cyclic crack growth. FRASTA, which seeks to reconstruct the crack propagation process in microscopic detail by comparing the topographies of conjugate fracture surfaces, showed differences in the process zone microfeatures in A533B-1 and A508-2 steel and also differences in A508-2 tested in a dry argon environment. Conclusions from these observations in terms of the slip dissolution and hydrogen embrittlement crack propagation mechanisms must await further FRASTA studies aimed at clarifying the effects of the test environment and the post-test cleaning solution on the fracture surface topography. Constant extension rate tests performed in simulated PWR environments containing purposefully high sulfur concentrations resulted in no stress corrosion cracking. These results do not support the hypothesis that sulfur in the environment promotes crack propagation. 14 refs., 26 figs., 3 tabs.

  13. Fracture properties of specially heat treated Type A 508 Class 2 pressure vessel steel

    SciTech Connect

    Alexander, D.J.; Cheverton, R.D.

    1991-01-01

    The mechanical properties of A 508 class two pressure vessel steel quenched and tempered to simulate the effects of irradiation have been measured. The tensile, Charpy impact, fracture toughness, and crack arrest properties were measured as a function of temperature. The fracture toughness was measured using 1T, 2T, and 4T compact specimens. The maximum dimensions of the 4T specimens exceeded the test cylinder thickness, so compound specimens were successfully fabricated by electron-beam welding arms to the material. Compact crack arrest specimens were used to determine the arrest toughness. Two sizes of specimens were tested: 150 {times} 150 {times} 32 mm, and 75 {times} 75 {times} 25 mm. The larger specimens used a brittle weld bead and notch as the crack starter. Two different notches (blunt and fatigue precracked) were used with the smaller specimens. The blunt notch was not successful, but the fatigue-precracked specimens provided valid data without the need for warm prestressing. the results from the small fatigue-precracked specimens were consistent with the data from the larger weld-embrittled specimens.

  14. Fracture properties of specially heat treated Type A 508 Class 2 pressure vessel steel

    SciTech Connect

    Alexander, D.J.; Cheverton, R.D.

    1991-12-31

    The mechanical properties of A 508 class two pressure vessel steel quenched and tempered to simulate the effects of irradiation have been measured. The tensile, Charpy impact, fracture toughness, and crack arrest properties were measured as a function of temperature. The fracture toughness was measured using 1T, 2T, and 4T compact specimens. The maximum dimensions of the 4T specimens exceeded the test cylinder thickness, so compound specimens were successfully fabricated by electron-beam welding arms to the material. Compact crack arrest specimens were used to determine the arrest toughness. Two sizes of specimens were tested: 150 {times} 150 {times} 32 mm, and 75 {times} 75 {times} 25 mm. The larger specimens used a brittle weld bead and notch as the crack starter. Two different notches (blunt and fatigue precracked) were used with the smaller specimens. The blunt notch was not successful, but the fatigue-precracked specimens provided valid data without the need for warm prestressing. the results from the small fatigue-precracked specimens were consistent with the data from the larger weld-embrittled specimens.

  15. Environmental acceleration of fatigue crack growth in reactor pressure vessel materials. Volume 2: Final report supplement

    SciTech Connect

    Van Der Sluys, W.A.; Emanuelson, R.H.; Vaccaro, F.P.

    1993-08-01

    Testing revealed that sulfur content and sulfur morphology are the most important variables determining the susceptibility of low alloy pressure vessel steels (SA508C and SA533B) to environmentally enhanced fatigue crack growth. The next most significant engineering variables are loading frequency and temperature. As each applied stress intensity factor level increased during crack growth tests, the frequency at which environmentally assisted cracking (EAC) occurred decreased. EAC was most pronounced at 200{degrees}C, with the effect over a wide range in frequency. A host of other variables exert some influence on the existence of EAC, including oxygen content and R-ratio of the loading wave form. Sufficient information now exists regarding the magnitude and controlling variables for EAC so that, in many cases, realistic predictions of crack growth rates can be made. Overall, the research described in this report contributes substantially to the scientific understanding of EAC under fatigue loading conditions. Volume 1 of this report summarizes key results from the fatigue crack growth studies. Volume 2 provides details of the test methods and data from the studies.

  16. Cryogenic Pressure Vessels for H2 Vehicles Rapidly Refueled by LH2 pump to 700 bar

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Vessels for H 2 Vehicles Rapidly Refueled by LH 2 pump to 700 bar Salvador Aceves, Gene Berry, Guillaume Petitpas, Vernon Switzer Lawrence Livermore National Laboratory CAMX meeting October 29 th , 2015 LLNL-PRES-678629 * Cryogenic H 2 Onboard Storage * Temperature as a Degree of Freedom in H 2 storage * LLNL Cryocompressed Project History * 350 Bar Test Vehicle Park & Drive Results * Current Project * 700 bar prototype (cryogenic) vessels * Refueling with LH 2 Pump * Test Vessel Cycling

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

    SciTech Connect

    J. K. Wright; R. N. Wright

    2008-04-01

    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

  18. Life assessment of a C-1/2Mo petroleum refinery pressure vessel operating in the creep regime

    SciTech Connect

    Brown, R.G.; Osage, D.A.; Buchheim, G.M.; Dobis, J.D.

    1995-12-31

    A comprehensive fitness-for-service assessment was conducted to evaluate a C-1/2Mo pressure vessel which has operated at temperatures in the creep range for almost 45 years. An initial damage assessment based on elastic stress analysis results indicated that this vessel was approaching its predicted failure life and thus there was little potential for increasing the operating temperature. Creep tests were conducted on samples removed from high stress regions of the vessel according to the MPC Omega Program protocol. The creep test results indicated that the material possesses creep strength superior to average new material and therefore has substantial remaining life. A nonlinear finite element analysis incorporating the MPC Project Omega creep law was performed to assess creep and fatigue damage. The results of this assessment indicated that future operation at increased temperatures was indeed feasible.

  19. Pressure vessel fracture studies pertaining to the PWR thermal-shock issue: experiment TSE-7

    SciTech Connect

    Cheverton, R.D.; Ball, D.G.; Bolt, S.E.; Iskander, S.K.; Nanstad, R.K.

    1985-08-01

    Thermal-shock experiment TSE-7 was conducted for the purpose of investigating the behavior of surface flaws under pressurized-water reactor (PWR) overcooling-accident conditions. This experiment was the eighth in a series of thermal-shock experiments conducted for this purpose with large steel cylinders (A 508, class-2 chemistry; 991-mm OD x 152-mm wall x 1.2-m length) as a part of the Heavy-Section Steel Technology (HSST) Program. The initial flaw for TSE-7 was a shallow, semielliptical, inner-surface, axially oriented, sharp crack located at midlength of the test cylinder. The thermal shock was applied to the inner surface only, and this was accomplished by effectively dunking the test cylinder, initially at approx.93/sup 0/C, into a large volume of liquid nitrogen. The specific purpose of TSE-7 was to determine whether, in agreement with analysis, a short and shallow surface flaw, in the absence of cladding, would extend on the surface to effectively become a very long flaw as a result of severe thermal-shock loading. During the experiment, there were three major initiation-arrest events. The first event consisted of some radial propagation and very extensive surface extension, with many bifurcations taking place. The second and third events consisted primarily of radial propagation. A fourth initiation event was prevented by warm prestressing. These results were in good agreement with predictions. 50 refs., 77 figs., 13 tabs.

  20. Reactor pressure vessel steels ASTM A533B and A508 c1. 2: crack opening displacement (COD) test results

    SciTech Connect

    Pelli, R.; Kemppainen, M.; Toeroenen, K.

    1980-06-01

    This report describes the crack opening displacement (COD) test results for the steels ASTM A533B and A508 C1.2 obtained in connection with a program initiated to gather and create information concerning the manufacturing variables, e.g. heat treatment, needed in the assessment of the structural integrity of reactor pressure vessels. The elastic-plastic fracture toughness was studied after applying various austenitizing and tempering temperatures.

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

    SciTech Connect

    Bhuyan, G.S.; Sperling, E.J.; Shen, G.; Yin, H.; Rana, M.D.

    1996-12-01

    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.

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

    SciTech Connect

    Bhuyan, G.S.; Sperling, E.J.; Shen, G.; Yin, H.; Rana, M.D.

    1999-08-01

    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.

  3. Effects of irradiation temperature on embrittlement of nuclear pressure vessel steels

    SciTech Connect

    Haggag, F.M.

    1992-12-31

    The effects of neutron irradiation on the steel reactor vessel for the modular high-temperature gas-cooled reactor (MHTGR) are being investigated, primarily because the operating temperatures are low [121 to 288{degrees}C (250--550{degrees}F)] compared to those for commercial light-water reactors (LWRs) [{approximately}288{degrees}C (550{degrees}F)]. The need for design data on the reference temperature (RT{sub NDT}) shift necessitated the irradiation at different temperatures of A 533 grade B class 1 plates, A 508 class 3 forging, and welds used for the vessel shell, vessel closure head, and vessel flange. This paper presents regular- and mini-tensile, Automated Ball Indentation (ABI), and Charpy V-notch (CVN) impact test results from five irradiation capsules of this program.

  4. Effects of irradiation temperature on embrittlement of nuclear pressure vessel steels

    SciTech Connect

    Haggag, F.M.

    1992-01-01

    The effects of neutron irradiation on the steel reactor vessel for the modular high-temperature gas-cooled reactor (MHTGR) are being investigated, primarily because the operating temperatures are low [121 to 288[degrees]C (250--550[degrees]F)] compared to those for commercial light-water reactors (LWRs) [[approximately]288[degrees]C (550[degrees]F)]. The need for design data on the reference temperature (RT[sub NDT]) shift necessitated the irradiation at different temperatures of A 533 grade B class 1 plates, A 508 class 3 forging, and welds used for the vessel shell, vessel closure head, and vessel flange. This paper presents regular- and mini-tensile, Automated Ball Indentation (ABI), and Charpy V-notch (CVN) impact test results from five irradiation capsules of this program.

  5. Design practices in Japan for the super high pressure vessels and comparison with the ASME Code Sect. VIII Div. 3 (under preparation)

    SciTech Connect

    Onozawa, Tsutomu; Tahara, Takayasu

    1995-12-01

    Recently, super high pressure facilities have been increasing in the industrial area so that to establish the regulatory standard to regulate the super high pressure vessels is a matter of great urgency world widely to keep the industrial safety. Under such a situation, the author shows respect to the ASME Code Committee for their efforts to publish the super high pressure vessel code. Mr. Leslie P. Antalffy, Fluor Daniel, Incorporated, Houston, Texas presented a paper during the 1993 and 1994 ASME PVP Conferences that ASME Code Committee has been preparing the rules of Division 3 of Section 8 of the Boiler and Pressure Vessel Code and explained its outline. In this paper, the authors shows the current super high pressure vessel design practices in Japan and explain the merit and problem area of these formulas comparing with the ASME formula and necessary conditions for the fatigue analysis.

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

    SciTech Connect

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

    2007-03-21

    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

  7. Effect of dielectric wall temperature on plasma plume in an argon atmospheric pressure discharge

    SciTech Connect

    Song, Jian; Huo, Yuxin; Wang, Youyin; Yu, Daren, E-mail: yudaren@hit.edu.cn [School of Energy Science and Engineering, Harbin Institute of Technology, Harbin 150001 (China); Tang, Jingfeng; Wei, Liqiu [Academy of Fundamental and Interdisciplinary Sciences, Harbin Institute of Technology, Harbin 150001 (China)

    2014-10-15

    In this letter, the effect of the dielectric wall temperature on the length and volume of an atmospheric pressure plasma jet (APPJ) is investigated using a single-electrode configuration driven with an AC power supply. To distinguish the APPJ status from the argon flow rate, the three modes, laminar, transition, and turbulent, are separated. When the dielectric wall is heated, the APPJ length and volume are enhanced. Also, the transition regions remarkably expand over a large range of flow rates. The results indicate that different factors contribute to the expansion of the transition region. The increase in the radial and axial velocities is the main cause of the expansion of the transition region to the low-velocity region. The expansion to the high-velocity region is dominantly induced by a change in the viscosity.

  8. 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

    Sullivan, Edmund J.; Anderson, Michael T.

    2014-06-10

    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.

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

    SciTech Connect

    Kausik, S. S.; Kakati, B.; Saikia, B. K.

    2013-05-15

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

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

    Energy.gov [DOE]

    The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations, which govern the...

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

    SciTech Connect

    Bethmont, M.; Frund, J.M.; Housin, B.; Soulat, P.

    1996-12-31

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

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

    SciTech Connect

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

    2013-11-26

    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; and Develop a plan for a non-destructive structural health monitoring technique and damage detection device for RPVs.

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

    SciTech Connect

    Rawls, G

    2007-04-30

    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.

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

    SciTech Connect

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

    1997-04-01

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

  15. Creep behavior of a nuclear pressure vessel under severe accident conditions

    SciTech Connect

    Beghini, M.; Bertini, L.; Vitale, E.

    1996-12-31

    The results of a study on the creep behavior of the vessel lower head under severe accident conditions are reported. An experimental program aimed at the evaluation of the creep properties of A533grB steel at high temperature (800--1,100 C) and under biaxial loading is summarized and the main results reported. A Finite Element simulation of the lower head under severe accident conditions allows to show the effect of the main parameters affecting the time to rupture.

  16. Use of miniature and standard specimens to evaluate effects of irradiation temperature on pressure vessel steels

    SciTech Connect

    Haggag, F.M.; Nanstad, R.K. ); Byrne, S.T. )

    1991-01-01

    The effects of neutron irradiation on the steel reactor vessel for the modular high-temperature gas-cooled reactor (MHTGR) are being investigated, primarily because the operating temperatures are low (121 to 210{degrees}C (250--410{degrees}F)) compared to those for commercial light-water reactors (LWRs) ({approximately}288{degrees}C (550{degrees}F)). The need for design data on the reference temperature shift necessitated the irradiation at different temperatures of A 533 grade B class 1 plate. A 508 class 3 forging, and welds used for the vessel shell, vessel closure head, the vessel flange. This paper presents results from the first four irradiation capsules of this program. The four capsules were irradiated in the University of Buffalo Reactor to an effective fast fluence of 1 {times}10{sup 18} neutron/cm{sup 2} (0.68 {times} 10{sup 18} neutron/cm{sup 2} (>1 MeV)) at temperatures of 288, 204, 163, and 121{degrees}C (550, 400, 325, and 250{degrees}F), respectively. The yield and ultimate strengths of both steel plate materials of the MHTGR Program increased with decreasing irradiation temperature. Similarly, the 41-J Charpy V-notch (CVN) transition temperature shift increased with decreasing irradiation temperature (in agreement with the increase in yield strength). The miniature tensile and automated ball indentation (ABI) test results (yield strength and flow properties) were in good agreement with those from standard tensile specimens. The miniature tensile and ABI test results were also used in a model that utilizes the changes in yield strength to estimate the CVN ductile-to-brittle transition temperature shift due to irradiation. The model predictions were compared with CVN test results obtained here and in earlier work. 5 refs., 11 figs., 6 tabs.

  17. Radial pressure flange seal

    DOEpatents

    Batzer, Thomas H.; Call, Wayne R.

    1989-01-01

    This invention provides an all metal seal for vacuum or pressure vessels or systems. This invention does not use gaskets. The invention uses a flange which fits into a matching groove. Fluid pressure is applied in a chamber in the flange causing at least one of the flange walls to radially press against a side of the groove creating the seal between the flange wall and the groove side.

  18. Radial pressure flange seal

    DOEpatents

    Batzer, T.H.; Call, W.R.

    1989-01-24

    This invention provides an all metal seal for vacuum or pressure vessels or systems. This invention does not use gaskets. The invention uses a flange which fits into a matching groove. Fluid pressure is applied in a chamber in the flange causing at least one of the flange walls to radially press against a side of the groove creating the seal between the flange wall and the groove side. 5 figs.

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

    SciTech Connect

    Qiu, J.A.; Wu, K.M.; Li, J.H.; Hodgson, P.D.; Hou, T.P.; Ding, Q.F.

    2013-09-15

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

  20. Evaluation and prediction of neutron embrittlement in reactor pressure vessel materials. Final report. [PWR; BWR

    SciTech Connect

    Hawthorne, J.R.; Menke, B.H.; Loss, F.J.; Watson, H.E.; Hiser, A.L.; Gray, R.A.

    1982-12-01

    This study evaluates the effects of fast neutron irradiation on the mechanical properties of eight nuclear reactor vessel materials. The materials include submerged arc weldments, three plates, and one forging. The materials are in the unirradiated and irradiated conditions with regard to tensile, Charpy impact, and static and dynamic fracture toughness properties. Correlations between impact and fracture toughness parameters are developed from the experimental results. The observed shifts in transition temperature and the drop in upper-shelf energy are compared with predictions developed from the Regulatory Guide 1.99.1 trend curves.

  1. Determining the syringyl/guaiacyl lignin ratio in the vessel and fiber cell walls of transgenic Populus plants

    DOE PAGES [OSTI]

    Tolbert, Allison K.; Ma, Tao; Kalluri, Udaya C.; Ragauskas, Arthur J.

    2016-06-20

    Observation of the spatial lignin distribution throughout the plant cell wall provides insight into the physicochemical characteristics of lignocellulosic biomass. The distribution of syringyl (S) and guaiacyl (G) lignin in cell walls of a genetically modified Populus deltoides and its corresponding empty vector control were analyzed with time-of-flight secondary ion mass spectrometry (ToF-SIMS) and then mapped to determine the S/G lignin ratio of the sample surface and specific regions of interest (ROIs). The surface characterizations of transgenic cross-sections within 1 cm vertical distance of each other on the stem possess similar S/G lignin ratios. Furthermore, the analysis of the ROIsmore » determined that there was a 50% decrease in the S/G lignin ratio of the transgenic xylem fiber cell walls.« less

  2. Nuclear reactor construction with bottom supported reactor vessel

    DOEpatents

    Sharbaugh, John E.

    1987-01-01

    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

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

    SciTech Connect

    J. K. Wright; R. N. Wright

    2010-07-01

    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.

  4. High-Pressure Effects on Boron Nitride Multi-Walled Nanotubes: An X-ray Diffraction Study

    SciTech Connect

    Muthu, D.; Midgley, A; Petruska, E; Sood, A; Bando, Y; Golberg, D; Kruger, M

    2008-01-01

    An X-ray diffraction study has been carried out on boron nitride multi-walled nanotubes up to 19.1 GPa. The fit to the unit cell volumes yields a zero pressure bulk modulus K0 = 34.9 {+-} 1 GPa and its pressure derivative K{prime}{sub 0} {+-} 0.3. The compression is highly anisotropic, with the intertube-axis being close to 19 times more compressible than the a-axis. At the highest pressure the diffraction peaks were present even though they were weak and broad, which is in contrast to previous high-pressure Raman results.

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

    SciTech Connect

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

    2010-12-22

    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.

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

    SciTech Connect

    Ren, Weiju; Terry, Totemeier

    2006-10-01

    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.

  7. LWR Pressure Vessel Surveillance Dosimetry Improvement Program for NRC (Nuclear Regulatory Commission) Materials Engineering Branch: 1985 summary annual report

    SciTech Connect

    McElroy, W.N.

    1988-04-01

    The objective of this program is to make measurements in neutron fields (''Benchmark'' and reactor ''Test Surveillance Regions'') for the subsequent validation/calibration of available state-of-the-art physics-dosimetry-metallurgy, damage correlation, and associated reactor analysis procedures and data. These procedures and data are in turn used for predicting the integrated effects of neutron exposure to Light Water Reactor (LWR) Pressure Vessel (PV) and support structure steels from the results of research reactor tests and power reactor surveillance programs. The program work includes: (1) selection of the neutron field, (2) validation/calibration of physics-dosimetry-metallurgy, damage correlations, and the associated reactor analysis procedures and data using these fields, (3) preparation and editing of a series of 22 supporting NUREG reports, and (4) preparation and establishment of a set of 21 ASTM-recommended standard Practices, Guides, and Methods. 53 refs., 12 figs., 14 tabs.

  8. Determination of the threshold values for corrosion fatigue crack growth rate of pressure vessel steels in PWR primary water

    SciTech Connect

    Haenninen, H.E.; Arilahti, E.; Ehrnsten, U.

    1992-12-31

    Corrosion fatigue crack growth rates over a range of frequencies from 10 Hz to 0.00001 Hz in two materials that have exhibited low-rate (A508 Class 3) and high-rate (A533B) crack growth behaviour at 288{degrees}C were studied at 200{degrees}C in PWR primary water. The frequency values above which marked environmental enhancement was observed were determined. Also the threshold values in terms of {Delta}K{sub th}, above which the marked environmental enhancement was observed in the crack growth rate, were determined both for A533B steel and Soviet pressure vessel steels with certain test parameters. Based on the extensive fractography the crack growth rate results are discussed mechanistically.

  9. Preparation of reconstituted Charpy V-notch impact specimens for generating pressure vessel steel fracture toughness data

    SciTech Connect

    Perrin, J.S.; Fromm, E.O.; Server, W.L.; McConnell, P.E.

    1982-01-01

    The arc stud welding process has been adapted for use in producing reconstituted Charpy V-notch impact specimens. In this process, each half of a tested and fractured Charpy specimen is used as the central region of a reconstituted specimen. End tabs are joined to one half of a fractured specimen by a specially designed stud welding apparatus. SA533B-1 and SA508-2 unirradiated and irradiated pressure vessel steel specimens have been produced. Both conventional and precracked reconstituted specimen data have been produced. Both types of data have been shown to be in excellent agreement with original specimen data. The arc stud welding process can therefore be used to increase the amount of data obtainable from a limited number of specimens or to obtain Charpy data when full size specimens cannot otherwise be obtained.

  10. Fracture toughness characterization of Japanese reactor pressure vessel steels: Joint EPRI-CRIEPI RPV embrittlement studies. Final report

    SciTech Connect

    Loss, F.J.; Graham, S.M.; Menke, B.H.; Server, W.L.

    1993-05-01

    The properties of five Japanese reactor pressure vessel steels have been characterized. These steels represent three A533-B C1 1 plates and two A 508 C1 3 forgings produced over a 20-year period, from 1969 to 1989. The fracture toughness was characterized by means of the J-R curve. Also investigated were the Charpy V-Notch and tensile properties as well as chemical composition and metallography. The five specimens were found to be essentially identical in the areas investigated. The test results for the five Japanese steels provide the unirradiated baseline needed for evaluating the effects of radiation embrittlement. Several of these Japanese materials are being irradiated in power and test reactors as part of the EPRI/CRIEPI Joint Research Project.

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

    SciTech Connect

    Allen, M.D.; Pilch, M.; Brockmann, J.E.; Tarbell, W.W. ); Nichols, R.T. ); Sweet, D.W. )

    1991-08-01

    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.

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

    SciTech Connect

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

    1987-01-01

    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.

  13. Pressure and concentration dependences of the autoignition temperature for normal butane + air mixtures in a closed vessel

    SciTech Connect

    Chandraratna, M.R.; Griffiths, J.F. . School of Chemistry)

    1994-12-01

    The condition at which autoignition occurs in lean premixed n-butane + air mixtures over the composition range 0.2%--2.5% n-butane by volume (0.06 < [phi] < 0.66) were investigated experimentally. Total reactant pressure from 0.1 to 0.6 MPa (1--6 atm) were studied in a spherical, stainless-steel, closed vessel (0.5 dm[sup 3]). There is a critical transition from nonignition to ignition, at pressures above 0.1 MPa, as the mixture is enriched in the vicinity of 1% fuel vapor by volume. There is also a region of multiplicity, which exhibits three critical temperatures at a given composition. Chemical analyses show that partially oxygenated components,including many o-heterocyclic compounds, are important products of the lean combustion of butane at temperatures up to 800 K. The critical conditions for autoignition are discussed with regard to industrial ignition hazards, especially in the context of the autoignition temperature of alkanes given by ASTM or BS tests. The differences between the behavior of n-butane and the higher n-alkanes are explained. The experimental results are also used as a basis for testing a reduced kinetic model to represent the oxidation and autoignition of n-butane or other alkanes.

  14. Effects of Pore Distributions on Ductility of Thin-Walled High Pressure Die-Cast Magnesium

    SciTech Connect

    Choi, Kyoo Sil; Li, Dongsheng; Sun, Xin; Li, Mei; Allison, John

    2013-06-01

    In this paper, a microstructure-based three-dimensional (3D) finite element modeling method is adopted to investigate the effects of porosity in thin-walled high pressure die-cast (HPDC) Magnesium alloys on their ductility. For this purpose, the cross-sections of AM60 casting samples are first examined using optical microscope and X-ray tomography to obtain the general information on the pore distribution features. The experimentally observed pore distribution features are then used to generate a series of synthetic microstructure-based 3D finite element models with different pore volume fractions and pore distribution features. Shear and ductile damage models are adopted in the finite element analyses to induce the fracture by element removal, leading to the prediction of ductility. The results in this study show that the ductility monotonically decreases as the pore volume fraction increases and that the effect of skin region on the ductility is noticeable under the condition of same local pore volume fraction in the center region of the sample and its existence can be beneficial for the improvement of ductility. The further synthetic microstructure-based 3D finite element analyses are planned to investigate the effects of pore size and pore size distribution.

  15. HTGR Base Technology Program. Task 2: concrete properties in nuclear environment. A review of concrete material systems for application to prestressed concrete pressure vessels

    SciTech Connect

    Naus, D.J.

    1981-05-01

    Prestressed concrete pressure vessels (PCPVs) are designed to serve as primary pressure containment structures. The safety of these structures depends on a correct assessment of the loadings and proper design of the vessels to accept these loadings. Proper vessel design requires a knowledge of the component (material) properties. Because concrete is one of the primary constituents of PCPVs, knowledge of its behavior is required to produce optimum PCPV designs. Concrete material systems are reviewed with respect to constituents, mix design, placing, curing, and strength evaluations, and typical concrete property data are presented. Effects of extreme loadings (elevated temperature, multiaxial, irradiation) on concrete behavior are described. Finally, specialty concrete material systems (high strength, fibrous, polymer, lightweight, refractory) are reviewed. 235 references.

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

    SciTech Connect

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

    1995-12-31

    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}.

  17. BIOASSAY VESSEL FAILURE ANALYSIS

    SciTech Connect

    Vormelker, P

    2008-09-22

    Two high-pressure bioassay vessels failed at the Savannah River Site during a microwave heating process for biosample testing. Improper installation of the thermal shield in the first failure caused the vessel to burst during microwave heating. The second vessel failure is attributed to overpressurization during a test run. Vessel failure appeared to initiate in the mold parting line, the thinnest cross-section of the octagonal vessel. No material flaws were found in the vessel that would impair its structural performance. Content weight should be minimized to reduce operating temperature and pressure. Outer vessel life is dependent on actual temperature exposure. Since thermal aging of the vessels can be detrimental to their performance, it was recommended that the vessels be used for a limited number of cycles to be determined by additional testing.

  18. 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

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

    2012-09-17

    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

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

    SciTech Connect

    Smith, Cyrus M; Nanstad, Randy K; Clayton, Dwight A; Matlack, Katie; Ramuhalli, Pradeep; Light, Glenn

    2012-09-01

    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.

  20. Analysis of A508 Class 2 and A533B pressure vessel steel fatigue tests in air. Final report

    SciTech Connect

    Eason, E.D.; Andrew, S.P.; Warmbrodt, S.B.; Nelson, E.E.

    1993-08-01

    A large set of well-documented A508 Class 2 and A533B pressure vessel steel specimens, fatigue-tested in air at various temperatures up to 300{degrees}C, were analyzed on a consistent basis starting from the basic experimental data. These data were from a variety of sources worldwide, and were analyzed using an advanced computer code, FATDAC. The reduction of experimental (a,N) data to (da/dN,{Delta}K) data utilized an optimization technique that minimizes scatter arising from numerical differentiation. All points exhibiting threshold behavior or not meeting elastic validity criteria were deleted; thus, the resulting model applies to Region II crack growth only. A model including a load ratio (R) factor was fitted to a randomly selected four-fifths sample of the specimens, using non-linear least squares techniques. The data were weighted to give more weight to points representing greater crack extension, and to compensate for the lack of experimental data in certain ranges in {Delta}K and R-Ratio. The resulting model shows a flatter slope than does the current ASME Section XI Code model for air data and subsurface flaws. The remaining one-fifth sample and other sets of data not used for fitting were compared with the model to verify its accuracy, with excellent agreement. Appropriate statistical bounds were constructed on the data and on predicted mean values using the model. Changes to the Section XI reference line are recommended.

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

    SciTech Connect

    David H. Bothell

    2000-04-30

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

  2. Method of fabricating a prestressed cast iron vessel

    DOEpatents

    Lampe, Robert F.

    1982-01-01

    A method of fabricating a prestressed cast iron vessel wherein double wall cast iron body segments each have an arcuate inner wall and a spaced apart substantially parallel outer wall with a plurality of radially extending webs interconnecting the inner wall and the outer wall, the bottom surface and the two exposed radial side surfaces of each body segment are machined and eight body segments are formed into a ring. The top surfaces and outer surfaces of the outer walls are machined and keyways are provided across the juncture of adjacent end walls of the body segments. A liner segment complementary in shape to a selected inner wall of one of the body segments is mounted to each of the body segments and again formed into a ring. The liner segments of each ring are welded to form unitary liner rings and thereafter the cast iron body segments are prestressed to complete the ring assembly. Ring assemblies are stacked to form the vessel and adjacent unitary liner rings are welded. A top head covers the top ring assembly to close the vessel and axially extending tendons retain the top and bottom heads in place under pressure.

  3. Atom probe tomography characterization of neutron irradiated surveillance samples from the R. E. Ginna reactor pressure vessel

    SciTech Connect

    Edmondson, Philip D.; Miller, Michael K.; Powers, Kathy A.; Nanstad, Randy K.

    2015-12-29

    Surveillance samples of a low copper (nominally 0.05 wt.% Cu) forging and a higher copper (0.23 wt.% Cu) submerged arc weld from the R. E. Ginna reactor pressure vessel have been characterized by atom probe tomography (APT) after exposure to three levels of neutron irradiation, i.e., fluences of 1.7, 3.6 and 5.8 × 1023 n.m–2 (E > 1 MeV), and inlet temperatures of ~289 °C (~552 °F). As no copper-enriched precipitates were observed in the low copper forging, and the measured copper content in the ferrite matrix was 0.04± <0.01 at.% Cu, after neutron irradiation to a fluence of 1.7 × 1023 n.m–3, this copper level was below the solubility limit. A number density of 2 × 1022 m–3 of Ni–, Mn– Si-enriched precipitates with an equivalent radius of gyration of 1.7 ± 0.4 nm were detected in the sample. However, Cu-, Ni-, Mn-enriched precipitates were observed in specimens cut from different surveillance specimens from the same forging material in which the overall measured copper level was 0.08± <0.01 at.% (fluence of 3.6 × 1023 n.m–3) and 0.09± <0.01 at.% Cu (fluence of 5.8 × 1023 n.m–3). Therefore, these slightly higher copper contents were above the solubility limit of Cu under these irradiation conditions. A best fit of all the composition data indicated that the size and number density of the Cu-enriched precipitates increased slightly in both size and number density by additional exposure to neutron irradiation. High number densities of Cu-enriched precipitates were observed in the higher Cu submerged arc weld for all irradiated conditions. The size and number density of the precipitates in the welds were higher than in the same fluence forgings. Some Cu-enriched precipitates were found to have Ni-, Mn- Si-, and P-enriched regions on their surfaces suggesting a preferential nucleation site. Furthermore, atom maps revealed P, Ni, and Mn

  4. Atom probe tomography characterization of neutron irradiated surveillance samples from the R. E. Ginna reactor pressure vessel

    DOE PAGES [OSTI]

    Edmondson, Philip D.; Miller, Michael K.; Powers, Kathy A.; Nanstad, Randy K.

    2015-12-29

    Surveillance samples of a low copper (nominally 0.05 wt.% Cu) forging and a higher copper (0.23 wt.% Cu) submerged arc weld from the R. E. Ginna reactor pressure vessel have been characterized by atom probe tomography (APT) after exposure to three levels of neutron irradiation, i.e., fluences of 1.7, 3.6 and 5.8 × 1023 n.m–2 (E > 1 MeV), and inlet temperatures of ~289 °C (~552 °F). As no copper-enriched precipitates were observed in the low copper forging, and the measured copper content in the ferrite matrix was 0.04± <0.01 at.% Cu, after neutron irradiation to a fluence of 1.7more » × 1023 n.m–3, this copper level was below the solubility limit. A number density of 2 × 1022 m–3 of Ni–, Mn– Si-enriched precipitates with an equivalent radius of gyration of 1.7 ± 0.4 nm were detected in the sample. However, Cu-, Ni-, Mn-enriched precipitates were observed in specimens cut from different surveillance specimens from the same forging material in which the overall measured copper level was 0.08± <0.01 at.% (fluence of 3.6 × 1023 n.m–3) and 0.09± <0.01 at.% Cu (fluence of 5.8 × 1023 n.m–3). Therefore, these slightly higher copper contents were above the solubility limit of Cu under these irradiation conditions. A best fit of all the composition data indicated that the size and number density of the Cu-enriched precipitates increased slightly in both size and number density by additional exposure to neutron irradiation. High number densities of Cu-enriched precipitates were observed in the higher Cu submerged arc weld for all irradiated conditions. The size and number density of the precipitates in the welds were higher than in the same fluence forgings. Some Cu-enriched precipitates were found to have Ni-, Mn- Si-, and P-enriched regions on their surfaces suggesting a preferential nucleation site. Furthermore, atom maps revealed P, Ni, and Mn segregation to, and preferential precipitation of, Cu-enriched precipitates over the surface of a grain

  5. Pressurized water reactor flow skirt apparatus

    DOEpatents

    Kielb, John F.; Schwirian, Richard E.; Lee, Naugab E.; Forsyth, David R.

    2016-04-05

    A pressurized water reactor vessel having a flow skirt formed from a perforated cylinder structure supported in the lower reactor vessel head at the outlet of the downcomer annulus, that channels the coolant flow through flow holes in the wall of the cylinder structure. The flow skirt is supported at a plurality of circumferentially spaced locations on the lower reactor vessel head that are not equally spaced or vertically aligned with the core barrel attachment points, and the flow skirt employs a unique arrangement of hole patterns that assure a substantially balanced pressure and flow of the coolant over the entire underside of the lower core support plate.

  6. Specifications for the development of BUGLE-93: An ENDF/B-VI multigroup cross section library for LWR shielding and pressure vessel dosimetry

    SciTech Connect

    White, J.E.; Wright, R.Q.; Roussin, R.W.; Ingersoll, D.T.

    1992-11-01

    This report discusses specifications which have been developed for a new multigroup cross section library based on ENDF/B-VI data for light water reactor shielding and reactor pressure vessel dosimetry applications. The resulting broad-group library and an intermediate fine-group library are defined by the specifications provided in this report. Processing ENDF/B-VI into multigroup format for use in radiation transport codes will provide radiation shielding analysts with the most currently available nuclear data. it is expected that the general nature of the specifications will be useful in other applications such as reactor physics.

  7. Dissolver vessel bottom assembly

    DOEpatents

    Kilian, Douglas C.

    1976-01-01

    An improved bottom assembly is provided for a nuclear reactor fuel reprocessing dissolver vessel wherein fuel elements are dissolved as the initial step in recovering fissile material from spent fuel rods. A shock-absorbing crash plate with a convex upper surface is disposed at the bottom of the dissolver vessel so as to provide an annular space between the crash plate and the dissolver vessel wall. A sparging ring is disposed within the annular space to enable a fluid discharged from the sparging ring to agitate the solids which deposit on the bottom of the dissolver vessel and accumulate in the annular space. An inlet tangential to the annular space permits a fluid pumped into the annular space through the inlet to flush these solids from the dissolver vessel through tangential outlets oppositely facing the inlet. The sparging ring is protected against damage from the impact of fuel elements being charged to the dissolver vessel by making the crash plate of such a diameter that the width of the annular space between the crash plate and the vessel wall is less than the diameter of the fuel elements.

  8. Conceptual Engineering Method for Attenuating He Ion Interactions on First Wall Components in the Fusion Test Facility (FTF) Employing a Low-Pressure Noble Gas

    SciTech Connect

    C.A.Gentile, W.R.Blanchard, T.Kozub, C.Priniski, I.Zatz, S.Obenschain

    2009-09-21

    It has been shown that post detonation energetic helium ions can drastically reduce the useful life of the (dry) first wall of an IFE reactor due to the accumulation of implanted helium. For the purpose of attenuating energetic helium ions from interacting with first wall components in the Fusion Test Facility (FTF) target chamber, several concepts have been advanced. These include magnetic intervention (MI), deployment of a dynamically moving first wall, use of a sacrificial shroud, designing the target chamber large enough to mitigate the damage caused by He ions on the target chamber wall, and the use of a low pressure noble gas resident in the target chamber during pulse power operations. It is proposed that employing a low-pressure (~ 1 torr equivalent) noble gas in the target chamber will thermalize energetic helium ions prior to interaction with the wall. The principle benefit of this concept is the simplicity of the design and the utilization of (modified) existing technologies for pumping and processing the noble ambient gas. Although the gas load in the system would be increased over other proposed methods, the use of a "gas shield" may provide a cost effective method of greatly extending the first wall of the target chamber. An engineering study has been initiated to investigate conceptual engineering metmethods for implementing a viable gas shield strategy in the FTF.

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

    SciTech Connect

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

    1997-03-01

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

  10. Statistical evaluation of the through-thickness copper variation and the K{sub Ic} and K{sub Ia} curves for reactor pressure vessels

    SciTech Connect

    Simonen, F.A.; Khaleel, M.A.

    1995-11-01

    This paper describes a statistical evaluation of the through-thickness copper variation for welds in reactor pressure vessels, and reviews the historical basis for the static and arrest fracture toughness (K{sub Ic} and K{sub Ia}) equations used in the VISA-II code. Copper variability in welds is due to fabrication procedures with copper contents being randomly distributed, variable from one location to another through the thickness of the vessel. The VISA-II procedure of sampling the copper content from a statistical distribution for every 6.35- to 12.7-mm (1/4- to 1/2-in.) layer through the thickness was found to be consistent with the statistical observations. However, the parameters of the VISA-II distribution and statistical limits required further investigation. Copper contents at few locations through the thickness were found to exceed the 0.4% upper limit of the VISA-II code. The data also suggest that the mean copper content varies systematically through the thickness. While, the assumption of normality is not clearly supported by the available data, a statistical evaluation based on all the available data results in mean and standard deviations within the VISA-II code limits.

  11. Influence of long-term thermal aging on the microstructural evolution of nuclear reactor pressure vessel materials: An atom probe study

    SciTech Connect

    Pareige, P.; Russell, K.F.; Stoller, R.E.; Miller, M.K.

    1998-03-01

    Atom probe field ion microscopy (APFIM) investigations of the microstructure of unaged (as-fabricated) and long-term thermally aged ({approximately} 100,000 h at 280 C) surveillance materials from commercial reactor pressure vessel steels were performed. This combination of materials and conditions permitted the investigation of potential thermal-aging effects. This microstructural study focused on the quantification of the compositions of the matrix and carbides. The APFIM results indicate that there was no significant microstructural evolution after a long-term thermal exposure in weld, plate, or forging materials. The matrix depletion of copper that was observed in weld materials was consistent with the copper concentration in the matrix after the stress-relief heat treatment. The compositions of cementite carbides aged for 100,000 h were compared with the Thermocalc{trademark} prediction. The APFIM comparisons of materials under these conditions are consistent with the measured change in mechanical properties such as the Charpy transition temperature.

  12. Vessel V-7 and V-8 repair and characterization of insert material. Final report

    SciTech Connect

    Domian, H.A.

    1984-05-01

    Pieces of Type SA508-2 steel, specially tempered to produce a high-impact-transition temperature, were welded in the side walls of Intermediate Test Vessels V-7 and V-8. These vessels are to be tested by the Oak Ridge National Laboratory (ORNL) in the Pressurized-Thermal-Shock (PTS) Project of the Heavy-Section Steel Technology (HSST) Program. A comparable piece of forging taken from the same source and heat treated with the vessels was characterized for its mechanical properties to provide data for use in the PTS tests.

  13. Predicting the Influence of Pore Characteristics on Ductility of Thin-Walled High Pressure Die Casting Magnesium

    SciTech Connect

    Sun, Xin; Choi, Kyoo Sil; Li, Dongsheng

    2013-06-10

    In this paper, a two-dimensional microstructure-based finite element modeling method is adopted to investigate the effects of porosity in thin-walled high pressure die casting Mg materials on their ductility. For this purpose, the cross-sections of AM50 and AM60 casting samples are first examined using optical microscope to obtain the overall information on the pore characteristics. The experimentally quantified pore characteristics are then used to generate a series of synthetic microstructures with different pore sizes, pore volume fractions and pore size distributions. Pores are explicitly represented in the synthetic microstructures and meshed out for the subsequent finite element analysis. In the finite element analysis, an intrinsic critical strain value is used for the Mg matrix material, beyond which work-hardening is no longer permissible. With no artificial failure criterion prescribed, ductility levels are predicted for the various microstructures in the form of strain localization. Mesh size effect study is also conducted, from which a mesh size dependent critical strain curve is determined. A concept of scalability of pore size effects is then presented and examined with the use of the mesh size dependent critical strain curve. The results in this study show that, for the regions with lower pore size and lower volume fraction, the ductility generally decreases as the pore size and pore volume fraction increase whereas, for the regions with larger pore size and larger pore volume fraction, other factors such as the mean distance between the pores begin to have some substantial influence on the ductility. The results also indicate that the pore size effects may be scalable for the models with good-representative pore shape and distribution with the use of the mesh size dependent critical strain curve.

  14. Particle Imaging Velocimetry Technique Development for Laboratory Measurement of Fracture Flow Inside a Pressure Vessel Using Neutron Imaging

    SciTech Connect

    Polsky, Yarom; Bingham, Philip R; Bilheux, Hassina Z; Carmichael, Justin R

    2015-01-01

    This paper will describe recent progress made in developing neutron imaging based particle imaging velocimetry techniques for visualizing and quantifying flow structure through a high pressure flow cell with high temperature capability (up to 350 degrees C). This experimental capability has great potential for improving the understanding of flow through fractured systems in applications such as enhanced geothermal systems (EGS). For example, flow structure measurement can be used to develop and validate single phase flow models used for simulation, experimentally identify critical transition regions and their dependence on fracture features such as surface roughness, and study multiphase fluid behavior within fractured systems. The developed method involves the controlled injection of a high contrast fluid into a water flow stream to produce droplets that can be tracked using neutron radiography. A description of the experimental setup will be provided along with an overview of the algorithms used to automatically track droplets and relate them to the velocity gradient in the flow stream. Experimental results will be reported along with volume of fluids based simulation techniques used to model observed flow.

  15. Effects of composition and heat treatment on the toughness of ASTM A508 Grade 3 Class 1 material for pressure vessels

    SciTech Connect

    Lin, M.; Hansen, S.S.; Nelson, T.D.; Focht, R.B.

    1997-12-31

    A laboratory study was conducted to evaluate the effects of composition and heat treatment on the toughness of ASTM A508 Grade 3 Class 1 material for pressure vessels. Five steels were vacuum induction melted and cast as ingots in the laboratory. These heats included a base steel representing the specification mid-range analysis, a steel containing higher levels of Si, Ni, and Cr (high-side composition) as compared to the base steel, and three steels derived from the high-side composition by adding Al, Al/N, and Nb, respectively. The ingots were rolled to plate, heat treated, and evaluated. Among these steels, the high-side composition with additions of Al and N displays the best strength/toughness combination. For example, a 75 mm-thick plate of this steel has acceptable strength and a reference nil ductility transition temperature (RT{sub NDT}) of {le} {minus} 29 C after austenitizing at 875 C, air cooling, and tempering at 660 C for up to 20 hours. Upper-nose temper embrittlement (UNTE) occurs in all these steels. This UNTE is attributed to the precipitation of needle-like Mo-rich carbides during tempering, and is significantly reduced by increasing the cooling rate after austenitizing.

  16. Upgrade of the DIII-D vacuum vessel protection system

    SciTech Connect

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

    1993-10-01

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

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

    SciTech Connect

    Park, R. J.; Lee, J. R.; Kim, S. B.; Jin, Y.; Kim, H. Y.

    2012-07-01

    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)

  18. Vessel structural support system

    DOEpatents

    Jenko, James X.; Ott, Howard L.; Wilson, Robert M.; Wepfer, Robert M.

    1992-01-01

    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.

  19. A preliminary assessment of the effects of heat flux distribution and penetration on the creep rupture of a reactor vessel lower head

    SciTech Connect

    Chu, T.Y.; Bentz, J.; Simpson, R.; Witt, R.

    1997-02-01

    The objective of the Lower Head Failure (LHF) Experiment Program is to experimentally investigate and characterize the failure of the reactor vessel lower head due to thermal and pressure loads under severe accident conditions. The experiment is performed using 1/5-scale models of a typical PWR pressure vessel. Experiments are performed for various internal pressure and imposed heat flux distributions with and without instrumentation guide tube penetrations. The experimental program is complemented by a modest modeling program based on the application of vessel creep rupture codes developed in the TMI Vessel Investigation Project. The first three experiments under the LHF program investigated the creep rupture of simulated reactor pressure vessels without penetrations. The heat flux distributions for the three experiments are uniform (LHF-1), center-peaked (LHF-2), and side-peaked (LHF-3), respectively. For all the experiments, appreciable vessel deformation was observed to initiate at vessel wall temperatures above 900K and the vessel typically failed at approximately 1000K. The size of failure was always observed to be smaller than the heated region. For experiments with non-uniform heat flux distributions, failure typically occurs in the region of peak temperature. A brief discussion of the effect of penetration is also presented.

  20. Development of automated welding process for field fabrication of thick walled pressure vessels. Technical progress report, second quarter, FY 1980, ending March 28, 1980

    SciTech Connect

    Schneider, U.A.

    1980-01-01

    Progress on a metallurgical contract is reported: (1) specifications of 2 1/4 chromium-1 molybdenum low alloy steel plate for a coal gasification project; (2) methods of welding and analyses of helium-argon mixtures for welding; and (3) tensile properties of welded joints. (LTN)

  1. Cover Heated, Open Vessels | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Cover Heated, Open Vessels Cover Heated, Open Vessels This tip sheet on covering heated, open vessels provides how-to advice for improving industrial steam systems using low-cost, proven practices and technologies. STEAM TIP SHEET #19 Cover Heated, Open Vessels (January 2012) (386.38 KB) More Documents & Publications Improving Steam System Performance: A Sourcebook for Industry, Second Edition Use Steam Jet Ejectors or Thermocompressors to Reduce Venting of Low-Pressure Steam Improving

  2. LPG storage vessel cracking experience

    SciTech Connect

    Cantwell, J.E. )

    1988-10-01

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

  3. LPG storage vessel cracking experience

    SciTech Connect

    Cantwell, J.E.

    1988-01-01

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

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

    SciTech Connect

    Tzanos, C.P.; Tessier, J.H.; Pedersen, D.R. )

    1989-11-01

    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. An optimization study for the reactor vessel auxiliary cooling system of a pool liquid-metal reactor

    SciTech Connect

    Tzanos, C.P.; Tessier, H.; Pedersen, D.R. )

    1991-04-01

    This paper reports on the effects of design parameters on the performance of the reactor vessel auxiliary cooling system (RVACS) of a pool liquid-metal reactor (LMR). These parameters include stack height, size of the airflow gap, system pressure loss, fins on the guard vessel or the baffle wall, and repeated ribs on the airflow channel walls. As a measure of performance , the peak sodium pool temperature during transient following a reactor scram from full power was used. Horizontal ribs with a 0.003-m height and a 0.015-m pitch gave the best performance, i.e., the lowest peak sodium pool temperature during the scram transient. For a 3500-MW(thermal) LMR, they gave peak hot pool and peak cladding temperatures that were 52{degrees}C lower than those obtained with a reference RVACS having smooth airflow channel walls.

  6. Wall Insulation

    SciTech Connect

    2000-10-01

    This fact sheet provides information on advanced wall framing, including insulating walls, airtight construction, and moisture control.

  7. 6151 Pressure Systems

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    For design, fabrication, testing, repair, modification and inspection are based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section ...

  8. Statistical Analyses for Probabilistic Assessments of the Reactor Pressure Vessel Structural Integrity: Building a Master Curve on an Extract of the 'Euro' Fracture Toughness Dataset, Controlling Statistical Uncertainty for Both Mono-Temperature and multi-temperature tests

    SciTech Connect

    Josse, Florent; Lefebvre, Yannick; Todeschini, Patrick; Turato, Silvia; Meister, Eric

    2006-07-01

    Assessing the structural integrity of a nuclear Reactor Pressure Vessel (RPV) subjected to pressurized-thermal-shock (PTS) transients is extremely important to safety. In addition to conventional deterministic calculations to confirm RPV integrity, Electricite de France (EDF) carries out probabilistic analyses. Probabilistic analyses are interesting because some key variables, albeit conventionally taken at conservative values, can be modeled more accurately through statistical variability. One variable which significantly affects RPV structural integrity assessment is cleavage fracture initiation toughness. The reference fracture toughness method currently in use at EDF is the RCCM and ASME Code lower-bound K{sub IC} based on the indexing parameter RT{sub NDT}. However, in order to quantify the toughness scatter for probabilistic analyses, the master curve method is being analyzed at present. Furthermore, the master curve method is a direct means of evaluating fracture toughness based on K{sub JC} data. In the framework of the master curve investigation undertaken by EDF, this article deals with the following two statistical items: building a master curve from an extract of a fracture toughness dataset (from the European project 'Unified Reference Fracture Toughness Design curves for RPV Steels') and controlling statistical uncertainty for both mono-temperature and multi-temperature tests. Concerning the first point, master curve temperature dependence is empirical in nature. To determine the 'original' master curve, Wallin postulated that a unified description of fracture toughness temperature dependence for ferritic steels is possible, and used a large number of data corresponding to nuclear-grade pressure vessel steels and welds. Our working hypothesis is that some ferritic steels may behave in slightly different ways. Therefore we focused exclusively on the basic french reactor vessel metal of types A508 Class 3 and A 533 grade B Class 1, taking the sampling

  9. Molten metal containment vessel with rare earth oxysulfide protective coating thereon and method of making same

    DOEpatents

    Krikorian, Oscar H.; Curtis, Paul G.

    1992-01-01

    An improved molten metal containment vessel is disclosed in which wetting of the vessel's inner wall surfaces by molten metal is inhibited by coating at least the inner surfaces of the containment vessel with one or more rare earth oxysulfide or rare earth sulfide compounds to inhibit wetting and or adherence by the molten metal to the surfaces of the containment vessel.

  10. Microsoft Word - Errata for the Pressure and Vacuum Systems Safety Supplement 3-15

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    PRESSURE VESSEL REGISTRATION FORM PS-4 Pressure System Number: Date: Pressure System Name: Pressure Vessel Number: P&ID Number: Pressure Vessel Description: MAWP/Design Pressure: Design Temperature: Operating Pressure: Operating Temperature: Code: Code Year: System Fluid: Fluid Category: Fluid State: VESSEL DATA ASME Stamp Type ___U Stamp ____UM Stamp ___Other (specify) Vessel Type: __Air Tank __Water Tank __Non-Flam Gas Tank __Flam Gas Tank __Other (specify) Vessel Manufacturer National

  11. Ion transport membrane module and vessel system

    DOEpatents

    Stein, VanEric Edward; Carolan, Michael Francis; Chen, Christopher M.; Armstrong, Phillip Andrew; Wahle, Harold W.; Ohrn, Theodore R.; Kneidel, Kurt E.; Rackers, Keith Gerard; Blake, James Erik; Nataraj, Shankar; van Doorn, Rene Hendrik Elias; Wilson, Merrill Anderson

    2007-02-20

    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.

  12. Ion transport membrane module and vessel system

    DOEpatents

    Stein, VanEric Edward; Carolan, Michael Francis; Chen, Christopher M.; Armstrong, Phillip Andrew; Wahle, Harold W.; Ohrn, Theodore R.; Kneidel, Kurt E.; Rackers, Keith Gerard; Blake, James Erik; Nataraj, Shankar; van Doorn, Rene Hendrik Elias; Wilson, Merrill Anderson

    2008-02-26

    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.

  13. Ion transport membrane module and vessel system

    DOEpatents

    Stein, VanEric Edward; Carolan, Michael Francis; Chen, Christopher M.; Armstrong, Phillip Andrew; Wahle, Harold W.; Ohrn, Theodore R.; Kneidel, Kurt E.; Rackers, Keith Gerard; Blake, James Erik; Nataraj, Shankar; Van Doorn, Rene Hendrik Elias; Wilson, Merrill Anderson

    2012-02-14

    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.

  14. Nuclear reactor vessel fuel thermal insulating barrier

    DOEpatents

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

    2013-03-19

    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.

  15. Structural Analysis of the NCSX Vacuum Vessel

    SciTech Connect

    Fred Dahlgren; Art Brooks; Paul Goranson; Mike Cole; Peter Titus

    2004-09-28

    The NCSX (National Compact Stellarator Experiment) vacuum vessel has a rather unique shape being very closely coupled topologically to the three-fold stellarator symmetry of the plasma it contains. This shape does not permit the use of the common forms of pressure vessel analysis and necessitates the reliance on finite element analysis. The current paper describes the NCSX vacuum vessel stress analysis including external pressure, thermal, and electro-magnetic loading from internal plasma disruptions and bakeout temperatures of up to 400 degrees centigrade. Buckling and dynamic loading conditions are also considered.

  16. Pressure surge attenuator

    DOEpatents

    Christie, Alan M.; Snyder, Kurt I.

    1985-01-01

    A pressure surge attenuation system for pipes having a fluted region opposite crushable metal foam. As adapted for nuclear reactor vessels and heads, crushable metal foam is disposed to attenuate pressure surges.

  17. Stress corrosion cracking of reactor pressure vessel steel in 288 C water: The effect of oxygen, electrochemical potential, and steel composition

    SciTech Connect

    Magdowski, R.; Kraus, A.; Speidel, M.O.

    1994-12-31

    A new set of fracture mechanics stress corrosion crack growth rate data is presented for transgranular cracking of reactor pressure steels in high temperature water. The essential observations are as follows. Fast stress corrosion crack growth rates between 10{sup {minus}9} and 10{sup {minus}8} m/s may be observed down to 400 ppb dissolved oxygen at water conductivities of 1.0 {mu}s/cm in refreshed autoclaves. Stress corrosion crack growth could not be observed below minus 270 mV on the hydrogen scale. Low and medium sulfur contents in the steels have no measurable influence on the stress corrosion crack growth rates in 288 C water with a conductivity of 1.0 {mu}S/cm.

  18. Coal gasification vessel

    DOEpatents

    Loo, Billy W.

    1982-01-01

    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).

  19. Dynamic high-pressure studies of an electrothermal capillary

    SciTech Connect

    Benson, D.A.; Cahill, P.A.

    1990-01-01

    This paper describes arc discharge tests conducted in a prepressurized, constant-volume pressure vessel to study arc behavior over a wide range of current densities, discharge durations and initial vessel pressures. This method allows controlled access to a wider range of conditions than those previously studied in capillary tests. We have investigated aspects of the radiative heat transfer by calculating the material opacity and mean free paths of photons for conditions typical of arc diagnostics. We also performed one-dimensional Eulerian hydrodynamic calculations of the boundary layer behavior in the radiative diffusion approximation. These calculations, which describe the radial mass flow and heat transfer in the absence of turbulent flow effects, show the characteristic times for equilibrium of the high-pressure arc. Finally, we describe progress on a promising means for increasing the mass flux from the capillary discharge through the use of chemically reactive media on the capillary walls. 20 refs., 7 figs.

  20. EDS V25 containment vessel explosive qualification test report.

    SciTech Connect

    Rudolphi, John Joseph

    2012-04-01

    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.

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

    SciTech Connect

    Cooke, Conrad; Spann, Holger

    2013-07-01

    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

  2. Static-stress analysis of dual-axis safety vessel

    SciTech Connect

    Bultman, D.H.

    1992-11-01

    An 8-ft-diameter safety vessel, made of HSLA-100 steel, is evaluated to determine its ability to contain the quasi-static residual pressure from a high-explosive (HE) blast. The safety vessel is designed for use with the Dual-Axis Radiographic Hydrotest (DARHT) facility being developed at Los Alamos National Laboratory. A smaller confinement vessel fits inside the safety vessel and contains the actual explosion, and the safety vessel functions as a second layer of containment in the unlikely case of a confinement vessel leak. The safety vessel is analyzed as a pressure vessel based on the ASME Boiler and Pressure Vessel Code, Section VIII, Division 1, and the Welding Research Council Bulletin, WRC107. Combined stresses that result from internal pressure and external loads on nozzles are calculated and compared to the allowable stresses for HSLA-100 steel. Results confirm that the shell and nozzle components are adequately designed for a static pressure of 830 psi, plus the maximum expected external loads. Shell stresses at the shellto-nozzle interface, produced from external loads on the nozzles, were less than 700 psi. The maximum combined stress resulting from the internal pressure plus external loads was 17,384 psi, which is significantly less than the allowable stress of 42,375 psi for HSLA-100 steel.

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

    DOEpatents

    Holmes, Michael Jerome; Ohrn, Theodore R.; Chen, Christopher Ming-Poh

    2010-02-09

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

  4. Cryogenic Pressure Vessels: Progress and Plans

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Salvador Aceves, Gene Berry, Francisco Espinosa, Ibo Matthews, Guillaume Petitpas, Tim Ross, Ray Smith, Vernon Switzer Lawrence Livermore National Laboratory February 15, 2011 This ...

  5. Static-stress analysis of dual-axis confinement vessel

    SciTech Connect

    Bultman, D.H.

    1992-11-01

    This study evaluates the static-pressure containment capability of a 6-ft-diameter, spherical vessel, made of HSLA-100 steel, to be used for high-explosive (HE) containment. The confinement vessel is designed for use with the Dual-Axis Radiographic Hydrotest Facility (DARHT) being developed at Los Alamos National Laboratory. Two sets of openings in the vessel are covered with x-ray transparent covers to allow radiographic imaging of an explosion as it occurs inside the vessel. The confinement vessel is analyzed as a pressure vessel based on the ASME Boiler and Pressure Vessel Code, Section VIII, Division 1, and the Welding Research Council Bulletin, WRC-107. Combined stresses resulting from internal pressure and external loads on nozzles are calculated and compared with the allowable stresses for HSLA-100 steel. Results confirm that the shell and nozzles of the confinement vessel are adequately designed to safely contain the maximum residual pressure of 1675 psi that would result from an HE charge of 24.2 kg detonated in a vacuum. Shell stresses at the shell-to-nozzle interface, produced from external loads on the nozzles, were less than 400 psi. The maximum combined stress resulting from the internal pressure plus external loads was 16,070 psi, which is less than half the allowable stress of 42,375 psi for HSLA-100 steel.

  6. Relief device for a vacuum vessel

    DOEpatents

    Fast, Ronald W.

    1987-04-28

    A pressure relief device 5 for a vessel having redundant pressure relief capabilities. An annular plate 12 overlies a surface 11 which has an aperature to the vessel. A seal is formed between the surface 11 and annular plate 12. A solid plate 13 overlies the annular plate 12. A seal is formed between the solid plate 13 and annular plate 12. The relief device 5 will open at a first predetermined pressure by lifting the solid plate 13. In the event the seal between solid plate 13 and annular plate 12 should stick the relief device 5 will open at a second slightly higher, predetermined pressure by lifting the annular plate 12 and solid plate 13 together. Hinging means 6 are provided to reclose the pressure relief device 5 when conditions return to normal.

  7. Bonfire Tests of High Pressure Hydrogen Storage Tanks | Department of

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Energy Bonfire Tests of High Pressure Hydrogen Storage Tanks Bonfire Tests of High Pressure Hydrogen Storage Tanks These slides were presented at the International Hydrogen Fuel and Pressure Vessel Forum on September 27 - 29, 2010, in Beijing, China. ihfpv_zheng1.pdf (986.67 KB) More Documents & Publications R&D of Large Stationary Hydrogen/CNG/HCNG Storage Vessels Forum Agenda: International Hydrogen Fuel and Pressure Vessel Forum International Hydrogen Fuel and Pressure Vessel

  8. Pressure suppression containment system

    DOEpatents

    Gluntz, Douglas M.; Townsend, Harold E.

    1994-03-15

    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.

  9. Pressure suppression containment system

    DOEpatents

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

    1994-03-15

    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. Confinement Vessel Assay System: Calibration and Certification Report

    SciTech Connect

    Frame, Katherine C.; Bourne, Mark M.; Crooks, William J.; Evans, Louise; Gomez, Cipriano; Mayo, Douglas R.; Miko, David K.; Salazar, William R.; Stange, Sy; Vigil, Georgiana M.

    2012-07-17

    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. Device for inspecting vessel surfaces

    DOEpatents

    Appel, D. Keith

    1995-01-01

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

  12. Inexpensive Delivery of Compressed Hydrogen with Advanced Vessel Technology

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    | Department of Energy Inexpensive Delivery of Compressed Hydrogen with Advanced Vessel Technology Inexpensive Delivery of Compressed Hydrogen with Advanced Vessel Technology Presentation on inexpensive delivery of compressed hydrogen with advanced vessel technology. wkshp_storage_berry.pdf (367.52 KB) More Documents & Publications Overview of FreedomCAR & Fuels Partnership/DOE Delivery Program President's Hydrogen Fuel Initiative High-Pressure Tube Trailers and Tanks

  13. A comparison of hydrogen vs. helium glow discharge effects on fusion device first-wall conditioning

    SciTech Connect

    Dylla, H.F.

    1989-09-01

    Hydrogen- and deuterium-fueled glow discharges are used for the initial conditioning of magnetic fusion device vacuum vessels following evacuation from atmospheric pressure. Hydrogenic glow discharge conditioning (GDC) significantly reduces the near-surface concentration of simple adsorbates, such as H/sub 2/O, CO, and CH/sub 4/, and lowers ion-induced desorption coefficients by typically three orders of magnitude. The time evolution of the residual gas production observed during hydrogen-glow discharge conditioning of the carbon first-wall structure of the TFTR device is similar to the time evolution observed during hydrogen GDC of the initial first-wall configuration in TFTR, which was primarily stainless steel. Recently, helium GDC has been investigated for several wall-conditioning tasks on a number of tokamaks including TFTR. Helium GDC shows negligible impurity removal with stainless steel walls. For impurity conditioning with carbon walls, helium GDC shows significant desorption of H/sub 2/O, CO, and CO/sub 2/; however, the total desorption yield is limited to the monolayer range. In addition, helium GDC can be used to displace hydrogen isotopes from the near-surface region of carbon first-walls in order to lower hydrogenic retention and recycling. 38 refs., 6 figs.

  14. Reactor vessel support system

    DOEpatents

    Golden, Martin P.; Holley, John C.

    1982-01-01

    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.

  15. Using SA508/533 for the HTGR Vessel Material

    SciTech Connect

    Larry Demick

    2012-06-01

    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.

  16. Tow Vessel | Open Energy Information

    OpenEI (Open Energy Information) [EERE & EIA]

    Tow Vessel Jump to: navigation, search Retrieved from "http:en.openei.orgwindex.php?titleTowVessel&oldid596390" Feedback Contact needs updating Image needs updating...

  17. LANL Robotic Vessel Scanning

    SciTech Connect

    Webber, Nels W.

    2015-11-25

    Los Alamos National Laboratory in J-1 DARHT Operations Group uses 6ft spherical vessels to contain hazardous materials produced in a hydrodynamic experiment. These contaminated vessels must be analyzed by means of a worker entering the vessel to locate, measure, and document every penetration mark on the vessel. If the worker can be replaced by a highly automated robotic system with a high precision scanner, it will eliminate the risks to the worker and provide management with an accurate 3D model of the vessel presenting the existing damage with the flexibility to manipulate the model for better and more in-depth assessment.The project was successful in meeting the primary goal of installing an automated system which scanned a 6ft vessel with an elapsed time of 45 minutes. This robotic system reduces the total time for the original scope of work by 75 minutes and results in excellent data accumulation and transmission to the 3D model imaging program.

  18. Pressurized fluidized bed reactor

    DOEpatents

    Isaksson, J.

    1996-03-19

    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.

  19. Pressurized fluidized bed reactor

    DOEpatents

    Isaksson, Juhani

    1996-01-01

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

    DOEpatents

    Ekeroth, D.E.; Orr, R.

    1993-12-07

    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.

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

    DOEpatents

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

    1993-01-01

    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.

  2. Radioactive material release from a containment vessel during a fire accident

    SciTech Connect

    Hensel, S.; Norkus, J.

    2015-02-26

    A methodology is presented to determine the source term for leaks and ruptures of pressurized vessels. The generic methodology is applied to a 9975 Primary Containment Vessel (PCV) which losses containment due to a hypothesized fire accident. The release due to a vessel rupture is approximately two orders of magnitude greater than the release due to a leak.

  3. Reactor vessel annealing system

    DOEpatents

    Miller, Phillip E.; Katz, Leonoard R.; Nath, Raymond J.; Blaushild, Ronald M.; Tatch, Michael D.; Kordalski, Frank J.; Wykstra, Donald T.; Kavalkovich, William M.

    1991-01-01

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

  4. Pressure testing of torispherical heads

    SciTech Connect

    Rana, M.D.; Kalnins, A.; Updike, D.P.

    1995-12-01

    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.

  5. Photoacoustic removal of occlusions from blood vessels

    DOEpatents

    Visuri, Steven R.; Da Silva, Luiz B.; Celliers, Peter M.; London, Richard A.; Maitland, IV, Duncan J.; Esch, Victor C.

    2002-01-01

    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.

  6. Modeling and measurement of the motion of the DIII-D vacuum vessel during vertical instabilities

    SciTech Connect

    Reis, E.; Blevins, R.D.; Jensen, T.H.; Luxon, J.L.; Petersen, P.I.; Strait, E.J.

    1991-11-01

    The motions of the D3-D vacuum vessel during vertical instabilities of elongated plasmas have been measured and studied over the past five years. The currents flowing in the vessel wall and the plasma scrapeoff layer were also measured and correlated to a physics model. These results provide a time history load distribution on the vessel which were input to a dynamic analysis for correlation to the measured motions. The structural model of the vessel using the loads developed from the measured vessel currents showed that the calculated displacement history correlated well with the measured values. The dynamic analysis provides a good estimate of the stresses and the maximum allowable deflection of the vessel. In addition, the vessel motions produce acoustic emissions at 21 Hertz that are sufficiently loud to be felt as well as heard by the D3-D operators. Time history measurements of the sounds were correlated to the vessel displacements. An analytical model of an oscillating sphere provided a reasonable correlation to the amplitude of the measured sounds. The correlation of the theoretical and measured vessel currents, the dynamic measurements and analysis, and the acoustic measurements and analysis show that: (1) The physics model can predict vessel forces for selected values of plasma resistivity. The model also predicts poloidal and toroidal wall currents which agree with measured values; (2) The force-time history from the above model, used in conjunction with an axisymmetric structural model of the vessel, predicts vessel motions which agree well with measured values; (3) The above results, input to a simple acoustic model predicts the magnitude of sounds emitted from the vessel during disruptions which agree with acoustic measurements; (4) Correlation of measured vessel motions with structural analysis shows that a maximum vertical motion of the vessel up to 0.24 in will not overstress the vessel or its supports. 11 refs., 10 figs., 1 tab.

  7. Center Stack Vacuum Vessel

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Stack Vacuum Vessel Current in center stack: 2 million amps, enough to light 2 million 100 watt light bulbs. Overall diameter and vacuum vessel diameter: 16.6 feet, 11.2 feet Height and weight of whole machine: 23.3 feet, 85 tons Components of the NSTX-U ENERGY U.S. DEPARTMENT OF ENERGY U.S. DEPARTMENT OF ENERGY U.S. DEPARTMENT OF ENERGY U.S. DEPARTMENT OF ENERGY ENERGY U.S. DEPARTMENT OF ENERGY U.S. DEPARTMENT OF ENERGY U.S. DEPARTMENT OF Heating power: 10 million amps, enough to produce

  8. Pressurizer tank upper support

    DOEpatents

    Baker, Tod H.; Ott, Howard L.

    1994-01-01

    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.

  9. Pressurizer tank upper support

    DOEpatents

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

    1994-01-11

    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.

  10. Economic advantages of Division 2 design for vessels per ASME Code Section VIII

    SciTech Connect

    Lengsfeld, M.; Holman, R.; Lengsfeld, P.F.

    1995-12-01

    ASME Boiler and Pressure Vessel Code Section 8, Division 2 has been available since 1968 for the design of pressure equipment. Industry has generally accepted this code for the design of high pressure vessels, high pressure being relative. Some consider high pressure above 3,000 PSIG, others look at high pressure above 1,000 or 1,500 PSIG. There are organizations who tie the use of Division 2 to thickness, meaning vessels in a thickness range above 3 to 4 inches as worthwhile to design to Division 2. In this paper the authors discuss the use of Division 2 strictly as an economic issue. Independent of thickness, if say a 3/4 in. thick vessel is lower in cost designed to Division 2 vs Division 1 why would one not build this vessel using Division 2 as the design basis?

  11. Pressure sensor for sealed containers

    DOEpatents

    Hodges, Franklin R.

    2001-01-01

    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.

  12. Pressure suppression system

    DOEpatents

    Gluntz, Douglas M.

    1994-01-01

    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. Pressure suppression system

    DOEpatents

    Gluntz, D.M.

    1994-10-04

    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.

  14. Near-wall serpentine cooled turbine airfoil

    SciTech Connect

    Lee, Ching-Pang

    2014-10-28

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

  15. Radiant vessel auxiliary cooling system

    DOEpatents

    Germer, John H.

    1987-01-01

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

  16. Method of non-destructively inspecting a curved wall portion

    DOEpatents

    Fong, James T.

    1996-01-01

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

  17. Confinement Vessel Assay System: Design and Implementation Report

    SciTech Connect

    Frame, Katherine C.; Bourne, Mark M.; Crooks, William J.; Evans, Louise; Mayo, Douglas R.; Gomez, Cipriano D.; Miko, David K.; Salazar, William R.; Stange, Sy; Vigil, Georgiana M.

    2012-07-18

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

  18. OSS 19.4 Pressure Safety 3/27/95

    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...

  19. Method and apparatus for determining pressure-induced frequency-shifts in shock-compressed materials

    DOEpatents

    Moore, David S.; Schmidt, Stephen C.

    1985-01-01

    A method and an apparatus for conducting coherent anti-Stokes Raman scattering spectroscopy in shock-compressed materials are disclosed. The apparatus includes a sample vessel having an optically transparent wall and an opposing optically reflective wall. Two coherent laser beams, a pump beam and a broadband Stokes beam, are directed through the window and focused on a portion of the sample. In the preferred embodiment, a projectile is fired from a high-pressure gas gun to impact the outside of the reflective wall, generating a planar shock wave which travels through the sample toward the window. The pump and Stokes beams result in the emission from the shock-compressed sample of a coherent anti-Stokes beam, which is emitted toward the approaching reflective wall of the vessel and reflected back through the window. The anti-Stokes beam is folded into a spectrometer for frequency analysis. The results of such analysis are useful for determining chemical and physical phenomena which occur during the shock-compression of the sample.

  20. Method and apparatus for determining pressure-induced frequency-shifts in shock-compressed materials

    DOEpatents

    Moore, D.S.; Schmidt, S.C.

    1983-12-16

    A method and an apparatus for conducting coherent anti-Stokes Raman scattering spectroscopy in shock-compressed materials are disclosed. The apparatus includes a sample vessel having an optically transparent wall and an opposing optically reflective wall. Two coherent laser beams, a pump beam and a broadband Stokes beam, are directed through the window and focused on a portion of the sample. In the preferred embodiment, a projectile is fired from a high-pressure gas gun to impact the outside of the reflective wall, generating a planar shock wave which travels through the sample toward the window. The pump and Stokes beams result in the emission from the shock-compressed sample of a coherent anti-Stokes beam, which is emitted toward the approaching reflective wall of the vessel and reflected back through the window. The anti-Stokes beam is folded into a spectrometer for frequency analysis. The results of such analysis are useful for determining chemical and physical phenomena which occur during the shock-compression of the sample.

  1. High pressure furnace

    DOEpatents

    Morris, D.E.

    1993-09-14

    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.

  2. High pressure furnace

    DOEpatents

    Morris, Donald E.

    1993-01-01

    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.

  3. High pressure liquid level monitor

    DOEpatents

    Bean, Vern E.; Long, Frederick G.

    1984-01-01

    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.

  4. PRESSURE RELIEF DEVICE DATA SHEET FORM PS-5 Pressure System Number: Date:

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    RELIEF DEVICE DATA SHEET FORM PS-5 Pressure System Number: Date: Pressure System Name: Pressure Vessel Number (if Applicable): Device installed directly on vessel?: __Yes __No Code: System Fluid: Code Year: Fluid State: Fluid Category: RELIEF DEVICE DATA Device Type ___Safety Relief Valve ____Rupture Disk ___Other (specify) Certification Type: ___ASME ___CE/PED ___Other (specify) Manufacturer Rated Flow Capacity: Part Number Converted Flow Capacity: Serial Number Set Pressure Inspection/Test

  5. Single module pressurized fuel cell turbine generator system

    DOEpatents

    George, Raymond A.; Veyo, Stephen E.; Dederer, Jeffrey T.

    2001-01-01

    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.

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

    SciTech Connect

    Vinson, Dennis

    2010-06-01

    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.

  7. Start-up control system and vessel for LMFBR

    DOEpatents

    Durrant, Oliver W.; Kakarala, Chandrasekhara R.; Mandel, Sheldon W.

    1987-01-01

    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.

  8. Start-up control system and vessel for LMFBR

    DOEpatents

    Durrant, Oliver W.; Kakarala, Chandrasekhara R.; Mandel, Sheldon W.

    1987-01-01

    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.

  9. Wall surveyor project report

    SciTech Connect

    Mullenhoff, D.J.; Johnston, B.C.; Azevedo, S.G.

    1996-02-22

    A report is made on the demonstration of a first-generation Wall Surveyor that is capable of surveying the interior and thickness of a stone, brick, or cement wall. LLNL`s Micropower Impulse Radar is used, based on emitting and detecting very low amplitude and short microwave impulses (MIR rangefinder). Six test walls were used. While the demonstrator MIR Wall Surveyor is not fieldable yet, it has successfully scanned the test walls and produced real-time images identifying the walls. It is planned to optimize and package the evaluation wall surveyor into a hand held unit.

  10. Metered Evaporator for Tokamak Wall Conditioning --- Inventor(s): Charles

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    H. Skinner, Dennis Mansfield, Henry Kugel, Hans Schneider and Lane Roquemore | Princeton Plasma Physics Lab Metered Evaporator for Tokamak Wall Conditioning --- Inventor(s): Charles H. Skinner, Dennis Mansfield, Henry Kugel, Hans Schneider and Lane Roquemore A novel lithium evaporator for the controlled introduction of lithium into tokamaks for wall conditioning is described. The concept uses a Li granule injector with a heated in-vessel yttrium crucible to evaporate a controlled amount of

  11. Wall loss of atomic nitrogen determined by ionization threshold mass spectrometry

    SciTech Connect

    Sode, M. Schwarz-Selinger, T.; Jacob, W.; Kersten, H.

    2014-11-21

    In the afterglow of an inductively coupled N{sub 2} plasma, relative N atom densities are measured by ionization threshold mass spectrometry as a function of time in order to determine the wall loss time t{sub wN} from the exponential decay curves. The procedure is performed with two mass spectrometers on different positions in the plasma chamber. t{sub wN} is determined for various pressures, i.e., for 3.0, 5.0, 7.5, and 10?Pa. For this conditions also the internal plasma parameters electron density n{sub e} and electron temperature T{sub e} are determined with the Langmuir probe and the rotational temperature T{sub rot}{sup N{sub 2}} of N{sub 2} is determined with the optical emission spectroscopy. For T{sub rot}{sup N{sub 2}}, a procedure is presented to evaluate the spectrum of the transition ?{sup ?}=0??{sup ?}=2 of the second positive system (C{sup 3}?{sub u}?B{sup 3}?{sub g}) of N{sub 2}. With this method, a gas temperature of 610?K is determined. For both mass spectrometers, an increase of the wall loss times of atomic nitrogen with increasing pressure is observed. The wall loss time measured with the first mass spectrometer in the radial center of the cylindrical plasma vessel increases linearly from 0.31?ms for 3?Pa to 0.82?ms for 10?Pa. The wall loss time measured with the second mass spectrometer (further away from the discharge) is about 4 times higher. A model is applied to describe the measured t{sub wN.} The main loss mechanism of atomic nitrogen for the considered pressure is diffusion to the wall. The surface loss probability ?{sub N} of atomic nitrogen on stainless steel was derived from t{sub wN} and is found to be 1 for the present conditions. The difference in wall loss times measured with the mass spectrometers on different positions in the plasma chamber is attributed to the different diffusion lengths.

  12. Flaw preparations for HSST program vessel fracture mechanics testing: mechanical-cyclic pumping and electron-beam weld-hydrogen-charge cracking schemes

    SciTech Connect

    Holz, P.P.

    1980-05-01

    Representative field testing to determine data for potential flaw propagation, fracture behavior, and margin against fracture for high-pressure-, high-temperature-service steel vessels subjected to increasing pressurization and/or thermal shock is premised on the investigators' ability to grow representative sharp cracks of known size, location, and orientation. Gaging for analytical stress and strain procedures and ultrasonic and acoustic emission instrumentation can then be applied to monitor the vessel during testing and to study crack growth. Cracks were grown by two techniques: (1) a mechanical method wherein a premachined notch was sharpened by pressurization; and (2) a method combining electron-beam welds and hydrogen charging to crack the chill zone of a rapidly placed autogenous weld. The mechanical method produces a naturally occurring growth shape controlled primarily by the shape of the machined notch; the weldinging-electrochemical method produces flaws of uniform depth from the surface of a wall or machined notch. Theories, details, discussions, and procedures are covered for both of the flaw-growing schemes. 21 refs., 33 figs., 3 tabs.

  13. Reactor vessel support system. [LMFBR

    DOEpatents

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

    1980-05-09

    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.

  14. Neutron and Gamma Fluxes and dpa Rates for HFIR Vessel Beltline Region (Present and Upgrade Designs)

    SciTech Connect

    Blakeman, E.D.

    2001-01-11

    The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) is currently undergoing an upgrading program, a part of which is to increase the diameters of two of the four radiation beam tubes (HB-2 and HB-4). This change will cause increased neutron and gamma radiation dose rates at and near locations where the tubes penetrate the vessel wall. Consequently, the rate of radiation damage to the reactor vessel wall at those locations will also increase. This report summarizes calculations of the neutron and gamma flux (particles/cm{sup 2}/s) and the dpa rate (displacements/atom/s) in iron at critical locations in the vessel wall. The calculated dpa rate values have been recently incorporated into statistical damage evaluation codes used in the assessment of radiation induced embrittlement. Calculations were performed using models based on the discrete ordinates methodology and utilizing ORNL two-dimensional and three-dimensional discrete ordinates codes. Models for present and proposed beam tube designs are shown and their results are compared. Results show that for HB-2, the dpa rate in the vessel wall where the tube penetrates the vessel will be increased by {approximately}10 by the proposed enlargement. For HB-4, a smaller increase of {approximately}2.6 is calculated.

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

    SciTech Connect

    Fan-Bill Cheung; Joy L. Rempe

    2004-06-01

    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.

  16. Reactor vessel seal service fixture

    DOEpatents

    Ritz, W.C.

    1975-12-01

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

  17. CRAD, Pressurized Systems and Cryogens Assessment Plan

    Office of Energy Efficiency and Renewable Energy (EERE)

    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. Fluidized wall for protecting fusion chamber walls

    DOEpatents

    Maniscalco, James A.; Meier, Wayne R.

    1982-01-01

    Apparatus for protecting the inner wall of a fusion chamber from microexplosion debris, x-rays, neutrons, etc. produced by deuterium-tritium (DT) targets imploded within the fusion chamber. The apparatus utilizes a fluidized wall similar to a waterfall comprising liquid lithium or solid pellets of lithium-ceramic, the waterfall forming a blanket to prevent damage of the structural materials of the chamber.

  19. DISRUPTION MITIGATION WITH HIGH-PRESSURE NOBLE GAS INJECTION

    SciTech Connect

    WHYTE, DG; JERNIGAN, TC; HUMPHREYS, DA; HYATT, AW; LASNIER, CJ; PARKS, PB; EVANS, TE; TAYLOR, PL; KELLMAN, AG; GRAY, DS; HOLLMANN, EM

    2002-10-01

    OAK A271 DISRUPTION MITIGATION WITH HIGH-PRESSURE NOBLE GAS INJECTION. High-pressure gas jets of neon and argon are used to mitigate the three principal damaging effects of tokamak disruptions: thermal loading of the divertor surfaces, vessel stress from poloidal halo currents and the buildup and loss of relativistic electrons to the wall. The gas jet penetrates as a neutral species through to the central plasma at its sonic velocity. The injected gas atoms increase up to 500 times the total electron inventory in the plasma volume, resulting in a relatively benign radiative dissipation of >95% of the plasma stored energy. The rapid cooling and the slow movement of the plasma to the wall reduce poloidal halo currents during the current decay. The thermally collapsed plasma is very cold ({approx} 1-2 eV) and the impurity charge distribution can include > 50% fraction neutral species. If a sufficient quantity of gas is injected, the neutrals inhibit runaway electrons. A physical model of radiative cooling is developed and validated against DIII-D experiments. The model shows that gas jet mitigation, including runaway suppression, extrapolates favorably to burning plasmas where disruption damage will be more severe. Initial results of real-time disruption detection triggering gas jet injection for mitigation are shown.

  20. Larry Wall | Argonne National Laboratory

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Larry Wall Manager, Programmatic Network/Systems Administraton Telephone (630) 252-3718 E-mail wall

  1. Detailed Analysis of a Late-Phase Core-Melt Progression for the Evaluation of In-vessel Corium Retention

    SciTech Connect

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

    2006-12-01

    Detailed analyses of a late-phase melt progression in the advanced power reactor (APR)1400 were completed to identify the melt and the thermal-hydraulic states of the in-vessel materials in the reactor vessel lower plenum at the time of reactor vessel failure to evaluate the candidate strategies for an in-vessel corium retention (IVR). Initiating events considered included high-pressure transients of a total loss of feed water (LOFW) and a station blackout (SBO) and low-pressure transients of a 0.0009-m2 small, 0.0093-m2 medium, and 0.0465-m2 large-break loss-of-coolant accident (LOCA) without safety injection. Best-estimate simulations for these low-probability events with conservative accident progression assumptions that lead to reactor vessel failure were performed by using the SCDAP/RELAP5/MOD3.3 computer code. The SCDAP/RELAP5/MOD3.3 results have shown that the pressurizer surge line failed before the reactor vessel failure, which results in a rapid decrease of the in-vessel pressure and a delay of the reactor vessel failure time of ~40 min in the high-pressure sequences of the total LOFW and the SBO transients. In all the sequences, ~80 to 90% of the core material was melted and relocated to the lower plenum of the reactor vessel at the time of reactor vessel failure. The maximum value of the volumetric heat source in the corium pool was estimated as 1.9 to 3.7 MW/m3. The corium temperature was ~2800 to 3400 K at the time of reactor vessel failure. The highest volumetric heat source sequence is predicted for the 0.0465-m2 large-break LOCA without safety injection in the APR1400, because this sequence leads to an early reactor vessel failure.

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

    SciTech Connect

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

    2000-07-01

    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.

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

    SciTech Connect

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

    2000-01-13

    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.

  4. An Overview Of The ITER In-Vessel Coil Systems

    SciTech Connect

    Heitzenroeder, P J; Chrzanowski, J H; Dahlgren, F; Hawryluk, R J; Loesser, G D; Neumeyer, C; Mansfield, C; Smith, J P; Schaffer, M; Humphreys, D; Cordier, J J; Campbell, D; Johnson, G A; Martin, A; Rebut, P H; Tao, J O; Fogarty, P J; Nelson, B E

    2009-09-24

    ELM mitigation is of particular importance in ITER in order to prevent rapid erosion or melting of the divertor surface, with the consequent risk of water leaks, increased plasma impurity content and disruptivity. Exploitable "natural" small or no ELM regimes might yet be found which extrapolate to ITER but this cannot be depended upon. Resonant Magnetic Perturbation has been added to pellet pacing as a tool for ITER to mitigate ELMs. Both are required, since neither method is fully developed and much work remains to be done. In addition, in-vessel coils enable vertical stabilization and RWM control. For these reasons, in-vessel coils (IVCs) are being designed for ITER to provide control of Edge Localized Modes (ELMs) in addition to providing control of moderately unstable resistive wall modes (RWMs) and the vertical stability (VS) of the plasma.

  5. Near wall cooling for a highly tapered turbine blade

    DOEpatents

    Liang, George

    2011-03-08

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

  6. Lightweight cryogenic-compatible pressure vessels for vehicular fuel

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    storage - Energy Innovation Portal 708,502 Site Map Printable Version Share this resource About Search Categories (15) Advanced Materials Biomass and Biofuels Building Energy Efficiency Electricity Transmission Energy Analysis Energy Storage Geothermal Hydrogen and Fuel Cell Hydropower, Wave and Tidal Industrial Technologies Solar Photovoltaic Solar Thermal Startup America Vehicles and Fuels Wind Energy Partners (27) Visual Patent Search Success Stories Find More Like This Return to Search

  7. International Hydrogen Fuel and Pressure Vessel Forum 2010 Proceedings

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    on testing and certification of storage tanks for compressed hydrogen, CNG, and HCNG ... and certification of Type 3 and Type 4 tanks, PRD testing and validation, tank ...

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

    Energy.gov [DOE] (indexed site)

    Proceedings from the forum, which took place in Beijing, China, on September 27-29, 2010. ... Development and Demonstration of Hydrogen-Compressed Natural Gas Vehicles in China

  9. Evaluation of advanced reactor pressure vessel steels under neutron irradiation

    SciTech Connect

    Davies, L.M.; Ingham, T.; Squires, R.L.

    1982-01-01

    The objective of the work described in this paper is to assess the radiation behavior of improved steels and weldments, produced in various countries. Three of the materials supplied by France and Japan to the IAEA program have been irradiated at 290 and/or 250/sup 0/C in the HERALD light water research reactor to assess the mechanical and upper shelf toughness properties after exposure to neutron fluences of approx. 1 x 10/sup 19/ n.cm/sup 2/ (E > 1 MeV). A fourth material, a weld supplied by the UK, has been included in the program for reference. Supplementary test methods used in the evaluation include hardness tests, chemical analyses, electron microscopy and positron annihilation. Results indicate that the improved steels are more resistant to radiation induced embrittlement compared to both the UK reference material and published data on older steels. The mechanistic understanding of irradiation embrittlement is becoming increasingly important particularly in terms of deriving a damage correlation model. The data from the supplementary tests are making a significant contribution towards this objective.

  10. Lens correction for the implementation of cylindrical vessels in a spectrophotometer

    SciTech Connect

    Spear, J.D.; Russo, R.E. ); Andrews, J.E.; Grant, P.M. )

    1992-10-01

    A simple optical correction system for allowing cylindrical vials to be used as sample vessels in a spectrophotometer is described. Light within the spectrophotometer enters and exits the liquid samples through the curved glass wall of the vials. Absorption spectra can be obtained conveniently without the need for solution transfer into standard rectangular spectrophotometer cuvettes.

  11. High pressure xenon ionization detector

    DOEpatents

    Markey, J.K.

    1989-11-14

    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 to 30 C. 2 figs.

  12. High pressure xenon ionization detector

    DOEpatents

    Markey, John K. (New Haven, CT)

    1989-01-01

    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.

  13. Achieve Continuous Injection of Solid Fuels into Advanced Combustion System Pressures

    SciTech Connect

    Derek L. Aldred; Timothy Saunders

    2007-03-31

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

  14. DIII-D in-vessel port cover and shutter assembly for the phase contrast interferometer

    SciTech Connect

    Phelps, R.D.

    1994-01-01

    The entire outer wall of the DIII-D vacuum vessel interion is covered with a regular array of graphite tiles. Certain of the diagnostic ports through the outer vessel wall contain equipment which is shielded from the plasma by installing port covers designed to withstand energy deposition. If the diagnostic contained in the port must communicate with the vessel volume, a shutter assembly is usually provided. In the ports at 285 degrees, R+1 and R-1, interferometer mirrors have been installed to provide a means for transmitting a large diameter CO-2 laser beam through the edge of the plasma. To protect the mirrors and other hardware contained in these ports, a special protective plate and shutter arrangement has been designed. This report describes the details of design, fabrication, and installation of these protective covers and shutters.

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

    SciTech Connect

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

    2010-09-22

    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

  16. Wood Pulp Digetster Wall Corrosion Investigation

    SciTech Connect

    Giles, GE

    2003-09-18

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

  17. Thin Wall Iron Castings

    SciTech Connect

    J.F. Cuttino; D.M. Stefanescu; T.S. Piwonka

    2001-10-31

    Results of an investigation made to develop methods of making iron castings having wall thicknesses as small as 2.5 mm in green sand molds are presented. It was found that thin wall ductile and compacted graphite iron castings can be made and have properties consistent with heavier castings. Green sand molding variables that affect casting dimensions were also identified.

  18. Electrically conductive containment vessel for molten aluminum

    DOEpatents

    Holcombe, Cressie E.; Scott, Donald G.

    1985-01-01

    The present invention is directed to a containment vessel which is particularly useful in melting aluminum. The vessel of the present invention is a multilayered vessel characterized by being electrically conductive, essentially nonwettable by and nonreactive with molten aluminum. The vessel is formed by coating a tantalum substrate of a suitable configuration with a mixture of yttria and particulate metal borides. The yttria in the coating inhibits the wetting of the coating while the boride particulate material provides the electrical conductivity through the vessel. The vessel of the present invention is particularly suitable for use in melting aluminum by ion bombardment.

  19. Electrically conductive containment vessel for molten aluminum

    DOEpatents

    Holcombe, C.E.; Scott, D.G.

    1984-06-25

    The present invention is directed to a containment vessel which is particularly useful in melting aluminum. The vessel of the present invention is a multilayered vessel characterized by being electrically conductive, essentially nonwettable by and nonreactive with molten aluminum. The vessel is formed by coating a tantalum substrate of a suitable configuration with a mixture of yttria and particulate metal 10 borides. The yttria in the coating inhibits the wetting of the coating while the boride particulate material provides the electrical conductivity through the vessel. The vessel of the present invention is particularly suitable for use in melting aluminum by ion bombardment.

  20. Aqueous Solution Vessel Thermal Model Development II

    SciTech Connect

    Buechler, Cynthia Eileen

    2015-10-28

    The work presented in this report is a continuation of the work described in the May 2015 report, “Aqueous Solution Vessel Thermal Model Development”. This computational fluid dynamics (CFD) model aims to predict the temperature and bubble volume fraction in an aqueous solution of uranium. These values affect the reactivity of the fissile solution, so it is important to be able to calculate them and determine their effects on the reaction. Part A of this report describes some of the parameter comparisons performed on the CFD model using Fluent. Part B describes the coupling of the Fluent model with a Monte-Carlo N-Particle (MCNP) neutron transport model. The fuel tank geometry is the same as it was in the May 2015 report, annular with a thickness-to-height ratio of 0.16. An accelerator-driven neutron source provides the excitation for the reaction, and internal and external water cooling channels remove the heat. The model used in this work incorporates the Eulerian multiphase model with lift, wall lubrication, turbulent dispersion and turbulence interaction. The buoyancy-driven flow is modeled using the Boussinesq approximation, and the flow turbulence is determined using the k-ω Shear-Stress-Transport (SST) model. The dispersed turbulence multiphase model is employed to capture the multiphase turbulence effects.

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

    DOEpatents

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

    1997-12-16

    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.

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

    DOEpatents

    Schreiber, Roger B.; Fero, Arnold H.; Sejvar, James

    1997-01-01

    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.

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

    DOEpatents

    Cambell, Christian X

    2013-09-17

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

  4. Application of the ASME code in designing containment vessels for packages used to transport radioactive materials

    SciTech Connect

    Raske, D.T.; Wang, Z.

    1992-07-01

    The primary concern governing the design of shipping packages containing radioactive materials is public safety during transport. When these shipments are within the regulatory jurisdiction of the US Department of Energy, the recommended design criterion for the primary containment vessel is either Section III or Section VIII, Division 1, of the ASME Boiler and Pressure Vessel Code, depending on the activity of the contents. The objective of this paper is to discuss the design of a prototypic containment vessel representative of a packaging for the transport of high-level radioactive material.

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

    SciTech Connect

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

    2011-06-03

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

  6. MAYmw Wall0

    Office of Legacy Management (LM)

    Wall0 50 l89 NE-23 NE-23 List of California Sites Hattie Carwell. SANNSQA Division Attached for your information is the list of California sites we identified in our search of ...

  7. Development of advanced manufacturing technologies for low cost hydrogen storage vessels

    SciTech Connect

    Leavitt, Mark; Lam, Patrick

    2014-12-29

    The U.S. Department of Energy (DOE) defined a need for low-cost gaseous hydrogen storage vessels at 700 bar to support cost goals aimed at 500,000 units per year. Existing filament winding processes produce a pressure vessel that is structurally inefficient, requiring more carbon fiber for manufacturing reasons, than would otherwise be necessary. Carbon fiber is the greatest cost driver in building a hydrogen pressure vessel. The objective of this project is to develop new methods for manufacturing Type IV pressure vessels for hydrogen storage with the purpose of lowering the overall product cost through an innovative hybrid process of optimizing composite usage by combining traditional filament winding (FW) and advanced fiber placement (AFP) techniques. A numbers of vessels were manufactured in this project. The latest vessel design passed all the critical tests on the hybrid design per European Commission (EC) 79-2009 standard except the extreme temperature cycle test. The tests passed include burst test, cycle test, accelerated stress rupture test and drop test. It was discovered the location where AFP and FW overlap for load transfer could be weakened during hydraulic cycling at 85°C. To design a vessel that passed these tests, the in-house modeling software was updated to add capability to start and stop fiber layers to simulate the AFP process. The original in-house software was developed for filament winding only. Alternative fiber was also investigated in this project, but the added mass impacted the vessel cost negatively due to the lower performance from the alternative fiber. Overall the project was a success to show the hybrid design is a viable solution to reduce fiber usage, thus driving down the cost of fuel storage vessels. Based on DOE’s baseline vessel size of 147.3L and 91kg, the 129L vessel (scaled to DOE baseline) in this project shows a 32% composite savings and 20% cost savings when comparing Vessel 15 hybrid design and the Quantum

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

    SciTech Connect

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

    2006-06-06

    The Department of Energy (DOE) identified the caustic side solvent extraction (CSSX) process as the preferred technology to remove cesium from radioactive waste solutions at the Savannah River Site (SRS). As a result, Washington Savannah River Company (WSRC) began designing and building a Modular CSSX Unit (MCU) in the SRS tank farm to process liquid waste for an interim period until the Salt Waste Processing Facility (SWPF) begins operations. Both the solvent and the strip effluent streams could contain high concentrations of cesium which must be removed from the contactors, process tanks, and piping prior to performing contactor maintenance. When these vessels are drained, thin films or drops will remain on the equipment walls. Following draining, the vessels will be flushed with water and drained to remove the flush water. The draining reduces the cesium concentration in the vessels by reducing the volume of cesium-containing material. The flushing, and subsequent draining, reduces the cesium in the vessels by diluting the cesium that remains in the film or drops on the vessel walls. MCU personnel requested that Savannah River National Laboratory (SRNL) researchers conduct a literature search to identify models to calculate the thickness of the liquid films remaining in the contactors, process tanks, and piping following draining of salt solution, solvent, and strip solution. The conclusions from this work are: (1) The predicted film thickness of the strip effluent is 0.010 mm on vertical walls, 0.57 mm on horizontal walls and 0.081 mm in horizontal pipes. (2) The predicted film thickness of the salt solution is 0.015 mm on vertical walls, 0.74 mm on horizontal walls, and 0.106 mm in horizontal pipes. (3) The predicted film thickness of the solvent is 0.022 mm on vertical walls, 0.91 mm on horizontal walls, and 0.13 mm in horizontal pipes. (4) The calculated film volume following draining is: (a) Salt solution receipt tank--1.6 gallons; (b) Salt solution feed

  9. Effect of fins and repeated-rib roughness on the performance characteristics of a reactor vessel air cooling system for LMFBR shutdown heat removal

    SciTech Connect

    Cheung, F.B.; Chawla, T.C.; Pedersen, D.R.; Tessier, J.H.; Webb, R.L.

    1986-01-01

    The use of a totally passive cooling system for shutdown heat removal that rejects heat from the reactor vessel by radiation to the guard vessel and from the guard vessel to a circulating air stream driven by natural convection is a key feature of the US Department of Energy's liquid-metal reactor advanced design study concepts. General Electric refers to the system as the Reactor Vessel Auxiliary Cooling System (RVACS) and Rockwell International as the Reactor Auxiliary Cooling System (RACS). The circulating air stream is contained in the annular passage formed with guard vessel wall and the duct wall surrounding the guard vessel. Specifically, the RVACS/RACS is designed to assure adequate cooling of the reactor vessel under abnormal operational conditions associated with loss of heat removal through the normal heat transport path via the steam generator system or the DRACS, if available. To enhance the heat transfer, longitudinal radial fins or repeated ribs can be attached to the duct wall and/or the guard vessel. The purpose of the present paper is to summarize the status of the analytical work on the development of an optimum design configuration for the RVACS/RACS.

  10. System for pressure modulation of turbine sidewall cavities

    DOEpatents

    Leone, Sal Albert; Book, Matthew David; Banares, Christopher R.

    2002-01-01

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

  11. Method for pressure modulation of turbine sidewall cavities

    DOEpatents

    Leone, Sal Albert; Book, Matthew David; Banares, Christopher R.

    2002-01-01

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

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

    DOEpatents

    Isaksson, J.M.

    1996-06-18

    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.

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

    DOEpatents

    Isaksson, Juhani M. (Karhula, FI)

    1996-01-01

    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.

  14. High-Pressure Tube Trailers and Tanks

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Gene Berry Salvador M. Aceves Lawrence Livermore National Laboratory (925) 422-0864 saceves@LLNL.GOV DOE Delivery Tech Team Presentation Chicago, Illinois February 8, 2005 Inexpensive delivery of compressed hydrogen with ambient temperature or cryogenic compatible vessels * Pressure vessel research at LLNL Conformable (continuous fiber and replicants) Cryo-compressed * Overview of delivery options * The thermodynamics of compressed and cryo-compressed hydrogen storage * Proposed analysis

  15. Analysis of In-Vessel Late Phase Melt Progression Using SCDAP/RELAP5/MOD3.3

    SciTech Connect

    Park, R.J.; Kim, S.B.; Kim, H.D. [Korea Atomic Energy Research Institute, Yuseong, P.O.Box 105, Daejeon, 305-600 (Korea, Republic of)

    2004-07-01

    High-pressure in-vessel melt progressions of the KSNP (Korean Standard Nuclear Power Plant) have been analyzed using the SCDAP/RELAP5/MOD3.3 computer code. The total loss of feed water (LOFW) to the steam generators with/without intentional RCS depressurization using the safety depressurization system (SDS) and the station blackout (SBO) have been simulated from transient initiation to reactor vessel failure. The SCDAP/RELAP5/MOD3.3 results have shown that the pressure boundary of the reactor coolant system did not fail before reactor vessel failure in the high-pressure sequences of the LOFW and the SBO transients of the KSNP. In all the high-pressure transients, approximately 20-30 % of the core material was melted and relocated to the lower plenum of the reactor vessel at the time of reactor vessel failure. Intentional RCS depressurization using the SDS for the total LOFW delays reactor vessel failure for approximately 5 hours by actuation of the safety injection tanks. At the time of reactor vessel failure, approximately 50-60 % of the fuel rod cladding was oxidized for the total LOFW and the SBO transients of the KSNP. (authors)

  16. Stress and Sealing Performance Analysis of Containment Vessel

    SciTech Connect

    WU, TSU-TE

    2005-05-24

    This paper presents a numerical technique for analyzing the containment vessel subjected to the combined loading of closure-bolt torque and internal pressure. The detailed stress distributions in the O-rings generated by both the torque load and the internal pressure can be evaluated by using this method. Consequently, the sealing performance of the O-rings can be determined. The material of the O-rings can be represented by any available constitutive equation for hyperelastic material. In the numerical calculation of this paper, the form of the Mooney-Rivlin strain energy potential is used. The technique treats both the preloading process of bolt tightening and the application of internal pressure as slow dynamic loads. Consequently, the problem can be evaluated using explicit numerical integration scheme.

  17. Pressure Systems

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Engineering > Pressure Systems Privacy and Security Notice Skip over navigation Search the JLab Site Pressure Systems Please upgrade your browser. This site's design is only ...

  18. Plastic instabilities in statically and dynamically loaded spherical vessels

    SciTech Connect

    Duffey, Thomas A; Rodriguez, Edward A

    2010-01-01

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

  19. Thermal treatment wall

    DOEpatents

    Aines, Roger D.; Newmark, Robin L.; Knauss, Kevin G.

    2000-01-01

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

  20. Axion domain wall baryogenesis

    SciTech Connect

    Daido, Ryuji; Kitajima, Naoya; Takahashi, Fuminobu

    2015-07-28

    We propose a new scenario of baryogenesis, in which annihilation of axion domain walls generates a sizable baryon asymmetry. Successful baryogenesis is possible for a wide range of the axion mass and decay constant, m≃10{sup 8}–10{sup 13} GeV and f≃10{sup 13}–10{sup 16} GeV. Baryonic isocurvature perturbations are significantly suppressed in our model, in contrast to various spontaneous baryogenesis scenarios in the slow-roll regime. In particular, the axion domain wall baryogenesis is consistent with high-scale inflation which generates a large tensor-to-scalar ratio within the reach of future CMB B-mode experiments. We also discuss the gravitational waves produced by the domain wall annihilation and its implications for the future gravitational wave experiments.

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

    SciTech Connect

    Scobel, James H.; Conway, L.E.; Theofanous, T.G.

    2002-07-01

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

  2. Stress corrosion cracking of pressurizer instrumentation nozzles in the French 1300 MWe units

    SciTech Connect

    Alter, D.; Robin, Y.; Pichon, M.; Teissier, A.; Thomeret, B.

    1992-12-31

    The 1300 MWE French PWR pressurizers are equipped with nozzles through which instruments penetrate the pressure vessel. The nozzles are made from forged and bored bars of Inconel 600 mechanically expanded in the pressurizer wall. They are then manually welded with Inconel 182 coated electrodes to the internal stainless steel cladding of the pressuriser. To understand the origin of leaks occurring early in life and to assess the extent of the problem we undertook an analysis of the fabrication conditions. Field investigations were carried out by dye penetrant testing on the nozzle bore. Cracks have been found on 35 percent of the 119 tested penetrations. Destructive examination performed on 3 nozzles showed that the circumferential cracks did not go through the wall thickness. Laboratory investigations of the nozzle pulled from Nogent 1 confirmed that the crack morphology corresponded to that of primary water stress corrosion cracking. No correlation has been found between microstructure of the different heats of Alloy 600 and cracking. Nozzle mock-ups investigations allowed residual stress measurements by X-ray diffraction. Stress corrosion cracking tests, showed that only longitudinal cracks can be through-wall while both longitudinal and circumferential cracks are initiated on the internal surface. As a result, Electricite De France decided to replace the Inconel 600 nozzles by stainless steel ones with austenitic st. st. weld. Furthermore, a full inventory of the Alloy 600 parts contained in the primary circuit has been performed. For each localized parts an assessment of the risk of stress corrosion cracking is under progress by studying material structures, stress level, operating conditions and safety point of view.

  3. Development of Larger Diameter High Pressure CNG Cylinder Manufactured...

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    These slides were presented at the International Hydrogen Fuel and Pressure Vessel Forum on September 27 - 29, 2010, in Beijing, China. ihfpvxma.pdf (3.43 MB) More Documents & ...

  4. Experimental Study on Flow Optimization in Upper Plenum of Reactor Vessel for a Compact Sodium-Cooled Fast Reactor

    SciTech Connect

    Kimura, Nobuyuki; Hayashi, Kenji; Kamide, Hideki; Itoh, Masami; Sekine, Tadashi

    2005-11-15

    An innovative sodium-cooled fast reactor has been investigated in a feasibility study of fast breeder reactor cycle systems in Japan. A compact reactor vessel and a column-type upper inner structure with a radial slit for an arm of a fuel-handling machine (FHM) are adopted. Dipped plates are set in the reactor vessel below the free surface to prevent gas entrainment. We performed a one-tenth-scaled model water experiment for the upper plenum of the reactor vessel. Gas entrainment was not observed in the experiment under the same velocity condition as the reactor. Three vortex cavitations were observed near the hot-leg inlet. A vertical rib on the reactor vessel wall was set to restrict the rotating flow near the hot leg. The vortex cavitation between the reactor vessel wall and the hot leg was suppressed by the rib under the same cavitation factor condition as in the reactor. The cylindrical plug was installed through the hole in the dipped plates for the FHM to reduce the flow toward the free surface. It was effective when the plug was submerged into the middle height in the upper plenum. This combination of two components had a possibility to optimize the flow in the compact reactor vessel.

  5. Moisture Research - Optimizing Wall Assemblies

    SciTech Connect

    Arena, Lois; Mantha, Pallavi

    2013-05-01

    In this project, the Consortium for Advanced Residential Buildings (CARB) team evaluated several different configurations of wall assemblies to determine the accuracy of moisture modeling and make recommendations to ensure durable, efficient assemblies. WUFI and THERM were used to model the hygrothermal and heat transfer characteristics of these walls. Wall assemblies evaluated included code minimum walls using spray foam insulation and fiberglass batts, high R-value walls at least 12 in. thick (R-40 and R-60 assemblies), and brick walls with interior insulation.

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

    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. Fracture toughness test results of thermal aged reactor vessel materials

    SciTech Connect

    DeVan, M.J.; Lowe, A.L. Jr.; Hall, J.B.

    1996-12-31

    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.

  8. High-R Walls for Remodeling. Wall Cavity Moisture Monitoring

    SciTech Connect

    Wiehagen, J.; Kochkin, V.

    2012-12-01

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

  9. High-R Walls for Remodeling: Wall Cavity Moisture Monitoring

    SciTech Connect

    Wiehagen, J.; Kochkin, V.

    2012-12-01

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

  10. Foam vessel for cryogenic fluid storage

    DOEpatents

    Spear, Jonathan D

    2011-07-05

    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.

  11. The Dielectric Wall Accelerator

    SciTech Connect

    Caporaso, George J.; Chen, Yu-Jiuan; Sampayan, Stephen E.

    2009-01-01

    The Dielectric Wall Accelerator (DWA), a class of induction accelerators, employs a novel insulating beam tube to impress a longitudinal electric field on a bunch of charged particles. The surface flashover characteristics of this tube may permit the attainment of accelerating gradients on the order of 100 MV/m for accelerating pulses on the order of a nanosecond in duration. A virtual traveling wave of excitation along the tube is produced at any desired speed by controlling the timing of pulse generating modules that supply a tangential electric field to the tube wall. Because of the ability to control the speed of this virtual wave, the accelerator is capable of handling any charge to mass ratio particle; hence it can be used for electrons, protons and any ion. The accelerator architectures, key technologies and development challenges will be described.

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

    DOEpatents

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

    1996-11-12

    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.

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

    DOEpatents

    Zafred, Paolo R.; Dederer, Jeffrey T.; Gillett, James E.; Basel, Richard A.; Antenucci, Annette B.

    1996-01-01

    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.

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

    SciTech Connect

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

    2012-10-01

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

  15. NREL: Energy Analysis - Anna Wall

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Anna Wall Photo of Anna Wall Anna Wall is a member of the Technology Systems and Sustainability Analysis Group in the Strategic Energy Analysis Center. Energy Technologies Analyst On staff since April 2014 Phone number: 303-384-6887 E-mail: anna.wall@nrel.gov Areas of expertise Geochemistry (aqueous and hard rock), with applications to geothermal resource characterization and mineral carbon sequestration Ratings methodologies and energy resource reporting standards Sustainable equity finance:

  16. Bumper wall for plasma device

    DOEpatents

    Coultas, Thomas A.

    1977-01-01

    Operation of a plasma device such as a reactor for controlled thermonuclear fusion is facilitated by an improved bumper wall enclosing the plasma to smooth the flow of energy from the plasma as the energy impinges upon the bumper wall. The bumper wall is flexible to withstand unequal and severe thermal shocks and it is readily replaced at less expense than the cost of replacing structural material in the first wall and blanket that surround it.

  17. Pressure Safety Program Implementation at ORNL

    SciTech Connect

    Lower, Mark; Etheridge, Tom; Oland, C. Barry

    2013-01-01

    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

  18. Dzyaloshinskii-Moriya domain wall resonance in ferromagnetic nanowires with a spin-transfer torque

    SciTech Connect

    Li, Zai-Dong; Liu, Fei; Li, Qiu-Yan; He, P. B.

    2015-05-07

    We theoretically investigate the current-induced domain wall resonance in ferromagnetic nanowires with a Dzyaloshinskii-Moriya interaction. The adiabatic and nonadiabatic torques distort the wall's internal structure and exert a global pressure on the wall. An effective Newton's equation is obtained analytically for a domain wall moving in one-dimensional potential and subject to a viscous friction and a driving force. Our results demonstrate that the Dzyaloshinskii-Moriya interaction affects the critical current density for depinning the wall, resonance frequency, and amplitude.

  19. Metallurgical evaluation of FMPC Vessel No. 2

    SciTech Connect

    Bagnall, C.; Wise, W.N.

    1989-03-01

    A major purpose of this evaluation program was to accumulate information on the behavior and properties of a vessel at the Feed Materials Production Center, fabricated of Monel 400, after service exposure in a UF/sub 6/--UF/sub 4/ reduction tower. These data will then be used to aid in the formulation of an equation to predict remaining life for the vessels. In addition, data from this destructive evaluation will provide information on the reliability of the reaction vessel surveillance program currently in operation at FMPC. After 1400 h of operation, Vessel No. 2 was removed from service and assigned to this program for extensive study. The report describes an initial survey of the physical condition of the vessel, provides details of the sampling plan, and then proceeds to document information in the various areas of investigation. These include radiography, chemical analysis, and mechanical properties over a temperature range up to 1800/degree/F. Metallographic studies from six key locations of the reaction vessel were conducted; major weld areas and selected tensile specimens were also examined. The report continues with a summary of the findings and a discussion of key aspects in relation to pertinent literature. The final section of the report provides conclusions drawn from evaluation of Vessel No. 2, and sets forth recommendations related to fabrication and extension of its operating life. 12 refs., 43 figs., 16 tabs.

  20. Thermal hydraulics of an external water wall type passive containment cooling system

    SciTech Connect

    Kataoka, Yoshiyuki; Murase, Michio; Fujii, Tadashi; Tominaga, Kenji

    1995-08-01

    An external water wall type containment cooling system is one of the passive containment cooling systems that use no active components and are intended for system simplification in the next generation power reactors. The core decay heat during a postulated loss-of-coolant accident is accumulated in the suppression pool (SP) and transferred to the outer pool, which is a cooling pool located outside and adjacent to the SP, by only natural phenomena such as natural convection, heat conduction, and evaporation. The temperature profiles and the convection heat transfer coefficients in the pools were measured using a 5-m height apparatus. The formation of a thermal stratification boundary at the vent outlets, which restricts the effective heat transfer area between pools, was clarified, and a correlation for natural convection heat transfer coefficients was obtained. Condensation heat transfer coefficients on the containment vessel wall and evaporation heat transfer coefficients on the SP surface under a noncondensable gas presence, which strongly affected the heat removal from the wet well, were evaluated based on the test results, and the correlations were obtained. The heat removal evaluation models, which analyze the trends of the temperatures and pressure, were developed and verified with system tests. As for the improvement of heat removal capability, two methods were proposed. One is a baffle plate to mitigate thermal stratification in the SP and enlarge the effective heat transfer area between pools. The second method is a divided wet well to avoid noncondensable gas effects. The thermal-hydraulic behavior in the SP with a baffle plate was clarified by three-dimensional analysis, and the effectiveness of these methods was experimentally confirmed.

  1. Machining Test Specimens from Harvested Zion RPV Segments for Through Wall Attenuation Studies

    SciTech Connect

    Rosseel, Thomas M; Sokolov, Mikhail A; Nanstad, Randy K

    2015-01-01

    The decommissioning of the Zion Units 1 and 2 Nuclear Generating Station (NGS) in Zion, Illinois presents a special opportunity for developing a better understanding of materials degradation and other issues associated with extending the lifetime of existing Nuclear Power Plants (NPPs) beyond 60 years of service. In support of extended service and current operations of the US nuclear reactor fleet, the Oak Ridge National Laboratory (ORNL), through the Department of Energy (DOE), Light Water Reactor Sustainability (LWRS) Program, is coordinating and contracting with Zion Solutions, LLC, a subsidiary of Energy Solutions, the selective procurement of materials, structures, and components from the decommissioned reactors. In this paper, we will discuss the acquisition of segments of the Zion Unit 2 Reactor Pressure Vessel (RPV), the cutting of these segments into sections and blocks from the beltline and upper vertical welds and plate material, the current status of machining those blocks into mechanical (Charpy, compact tension, and tensile) test specimens and coupons for chemical and microstructural (TEM, APT, SANS, and nano indention) characterization, as well as the current test plans and possible collaborative projects. Access to service-irradiated RPV welds and plate sections will allow through wall attenuation studies to be performed, which will be used to assess current radiation damage models (Rosseel et al. (2012) and Rosseel et al. (2015)).

  2. Modeling and Simulation of the ITER First Wall/Blanket Primary Heat Transfer System

    SciTech Connect

    Ying, Alice; Popov, Emilian L

    2011-01-01

    ITER inductive power operation is modeled and simulated using a thermal-hydraulics system code (RELAP5) integrated with a 3-D CFD (SC-Tetra) code. The Primary Heat Transfer System (PHTS) functions are predicted together with the main parameters operational ranges. The control algorithm strategy and derivation are summarized as well. The First Wall and Blanket modules are the primary components of PHTS, used to remove the major part of the thermal heat from the plasma. The modules represent a set of flow channels in solid metal structure that serve to absorb the radiation heat and nuclear heating from the fusion reactions and to provide shield for the vacuum vessel. The blanket modules are water cooled. The cooling is forced convective with constant blanket inlet temperature and mass flow rate. Three independent water loops supply coolant to the three blanket sectors. The main equipment of each loop consists of a pump, a steam pressurizer and a heat exchanger. A major feature of ITER is the pulsed operation. The plasma does not burn continuously, but on intervals with large periods of no power between them. This specific feature causes design challenges to accommodate the thermal expansion of the coolant during the pulse period and requires active temperature control to maintain a constant blanket inlet temperature.

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

    SciTech Connect

    Jin, Yusuke Konno, Yoshihiro; Nagao, Jiro

    2014-09-01

    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.

  4. Reactor vessel using metal oxide ceramic membranes

    DOEpatents

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

    1992-08-11

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

  5. Thermal wake/vessel detection technique

    DOEpatents

    Roskovensky, John K.; Nandy, Prabal; Post, Brian N

    2012-01-10

    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.

  6. Evaluations of the Coolability Through the Inherent In-Vessel Gap Cooling in the LAVA Experiments

    SciTech Connect

    Kang, K.H.; Park, R.J.; Kim, J.T.; Kim, S.B.; Kim, H.D.

    2002-07-01

    The analysis of the LAVA (Lower-plenum Arrested Vessel Attack) experimental results focused on gap formation and in-vessel gap cooling characteristics have been performed. In the LAVA experiment, Al{sub 2}O{sub 3}/Fe thermite melt (or Al{sub 2}O{sub 3} only) was used as a corium simulant with a 1/8 linear scale mock-up of the reactor vessel lower plenum. The experimental results address the non-adherence of the debris to the lower head vessel and the consequent gap formation in case there was an internal pressure load across the vessel. The thermal behaviors of the lower head vessel during the cooldown period were mainly affected by the heat removal characteristics through this gap, which were mainly determined by the possibilities of the water ingression into the gap. The possibility of heat removal through the gap in the LAVA experiments was confirmed from that the vessel cooled down with the conduction heat flux through the vessel by 70 to 470 kW/m{sup 2}. Also the quantitative evaluations of the in-vessel coolability using gap cooling model based on counter current flow limits (CCFL) have been performed for the LAVA experiments in parallel. It could be inferred from the analysis for the LAVA experiments that the vessel could effectively cooldown via heat removal through the gap cooling even if 2 mm thick gap should form between the interface of the melt and the vessel in the 30 kg of Al{sub 2}O{sub 3} melt tests. In the case of large melt mass of 70 kg of Al{sub 2}O{sub 3} melt, however, the infinite possibility of heat removal through a small size gap such as 1 to 2 mm thick couldn't be guaranteed due to the difficulties of water ingression through the gap into the lower head vessel bottom induced by the CCFL. Synthesized the experimental results and the analytical evaluations using the CCFL model, it could be found that the coolability through gap cooling was affected mainly by the melt composition and mass and also the gap thickness. (authors)

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

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Printer Friendly Engineering Test Reactor (ETR) Vessel Relocated Engineering Test Reactor Vessel Pre-startup 1957 Click on image to enlarge. Image 1 of 5 Gantry jacks attached to ETR vessel. Initial lift starts. - Click on image to enlarge. Image 2 of 5 ETR vessel removed from substructure. Vessel lifted approximately 40 ft. - Click on image to enlarge. On Monday, September 24, 2007 the Engineering Test Reactor (ETR) vessel was removed from its location and delivered to the Idaho CERCLA Disposal

  8. Two-Phase Natural Circulation Flow in AP-1000 In-Vessel Retention-Related ULPU-V Facility Experiments

    SciTech Connect

    Dinh, T.N.; Tu, J.P.; Theofanous, T.G.

    2004-07-01

    This paper is concerned with two-phase flow regimes and characteristics of coolant natural circulation around a reactor pressure vessel (RPV) in in-vessel retention (IVR) scenarios when the external vessel flooding is applied to arrest a hypothetical core melt accident. We focus on the AP1000 advanced plant design, and factors of potential importance to the coolant flow and the limit of coolability in IVR. This paper presents a synthesis of experimental results obtained in the ULPU-V facility, which simulates the AP1000 reactor geometry. We provide an analysis and interpretation of the ULPU-V observations, and discuss their relevant to the IVR performance. (authors)

  9. Development of 230-kV high-pressure, gas-filled, pipe-type cable system: Model test program phase

    SciTech Connect

    Silver, D.A. )

    1990-09-01

    The objective of this project was the development of a 230 kV high-pressure gas-filled (HPGF) pipe-type cable employing paper or laminate of paper-polypropylene-paper (PPP) insulation pressurized with N{sub 2} gas or a blend of 15% SF{sub 6}/85% N{sub 2} gas. Heretofore, HPGF pipe-type cables have been restricted to 138 kV ratings due to technical difficulties in achieving higher voltage ratings. In view of the high cost of manufacturing and testing a large number of full size cables, cable models with 2 mm (80 mils) and 2.5 mm (100 mils) wall thicknesses of insulation enclosed in a test fixture capable of withstanding a test pressure of 2070 kPa (300 psig) and high electrical stresses were employed for dissipation factor versus voltage measurements and for ac and impulse breakdown tests at rated and emergency operating temperatures. In addition, a 36 cm (14 in) full wall cable model enclosed in a pressure vessel was utilized for transient pressure response tests. The results of this investigation attest tot he technical feasibility of the design and manufacture of a 230 kV HPGF pipe-type cable employing paper or PPP insulation pressurized with 100% N{sub 2} gas or a blend of 15% SF{sub 6}/85% N{sub 2} gas for operation under normal and 100 hour emergency conditions at conductor temperatures of 85{degree} and 105{degree}C, respectively. The manufacture of a full size PPP insulated cable pressurized with a blend of 15% SF{sub 6}/85% N{sub 2} gas employing pre-impregnated PPP insulating tapes and an annular conductor based on the design stresses defined in this report is recommended for laboratory evaluation and extended life tests. 11 refs., 45 figs., 11 tabs.

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

    SciTech Connect

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

    2010-07-01

    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.

  11. PRESSURE RELIEF DEVICE TESTING AND INSPECTION DATA SHEET FORM PS-12

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    RELIEF DEVICE TESTING AND INSPECTION DATA SHEET FORM PS-12 Pressure System Number: Date: Pressure System Name: Vessel Number (if Applicable): Device installed directly on vessel?: __Yes __No Code: System Fluid: Code Year: Fluid State: Fluid Category: RELIEF DEVICE DATA Device Type ___Safety Relief Valve ____Rupture Disk ___Other (specify) Certification Type: ___ASME ___CE/PED ___Other (specify) Manufacturer Rated Flow Capacity: Part Number Converted Flow Capacity: Serial Number Set Pressure: Set

  12. Todd Vander Wall | Bioenergy | NREL

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Vander Wall Research Scientist Todd.Vander.Wall@nrel.gov | 303-384-7783 Research Interests Todd Vander Wall comes to the National Renewable Energy Laboratory (NREL) from the environmental and chemical engineering industry, where he focused on environmental chemistry. He originally designed and implemented methanogenesis systems for the growth and collection of biogenic methane from municipal solid waste. Following that, he performed in-situ investigations on thermogenic methane from subsurface

  13. Oven wall panel construction

    DOEpatents

    Ellison, Kenneth; Whike, Alan S.

    1980-04-22

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

  14. disrupting the plant cell wall

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    disrupting the plant cell wall - Sandia Energy Energy Search Icon Sandia Home Locations ... Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power ...

  15. Moisture Research - Optimizing Wall Assemblies

    SciTech Connect

    Arena, L.; Mantha, P.

    2013-05-01

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

  16. Pressure sensor

    SciTech Connect

    Mee, David K.; Ripley, Edward B.; Nienstedt, Zachary C.; Nienstedt, Alex W.; Howell, Jr., Layton N.

    2015-09-29

    Disclosed is a passive, in-situ pressure sensor. The sensor includes a sensing element having a ferromagnetic metal and a tension inducing mechanism coupled to the ferromagnetic metal. The tension inducing mechanism is operable to change a tensile stress upon the ferromagnetic metal based on a change in pressure in the sensing element. Changes in pressure are detected based on changes in the magnetic switching characteristics of the ferromagnetic metal when subjected to an alternating magnetic field caused by the change in the tensile stress. The sensing element is embeddable in a closed system for detecting pressure changes without the need for any penetrations of the system for power or data acquisition by detecting changes in the magnetic switching characteristics of the ferromagnetic metal caused by the tensile stress.

  17. Pressure regulator

    DOEpatents

    Ebeling, Jr., Robert W.; Weaver, Robert B.

    1979-01-01

    The pressure within a pressurized flow reactor operated under harsh environmental conditions is controlled by establishing and maintaining a fluidized bed of uniformly sized granular material of selected density by passing the gas from the reactor upwardly therethrough at a rate sufficient to fluidize the bed and varying the height of the bed by adding granular material thereto or removing granular material therefrom to adjust the backpressure on the flow reactor.

  18. PRESSURE TRANSDUCER

    DOEpatents

    Sander, H.H.

    1959-10-01

    A pressure or mechanical force transducer particularly adaptable to miniature telemetering systems is described. Basically the device consists of a transistor located within a magnetic field adapted to change in response to mechanical force. The conduction characteristics of the transistor in turn vary proportionally with changes in the magnetic flux across the transistor such that the output (either frequency of amplitude) of the transistor circuit is proportional to mechanical force or pressure.

  19. Upflow bioreactor with septum and pressure release mechanism

    DOEpatents

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

    2010-04-20

    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.

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

    SciTech Connect

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

    2012-09-27

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

  1. High pressure ratio turbocharger

    SciTech Connect

    Woollenweber, W.E.

    1991-06-25

    This patent describes a turbocharger system for an internal combustion engine. It comprises means forming a turbine adapted to be driven by exhaust gas from an internal combustion engine comprising: a turbine wheel having a central core and outwardly extending vanes, the turbine wheel being rotatable about a central axis; a meridionally divided volute for exhaust gas surrounding the turbine wheel, the meridionally divided volute including a divider wall defining first and second volute passageways with openings at the turbine wheel; means forming a high-pressure compressor driven by the turbine means, the high-pressure compressor comprising: rotating compressor blades, the compressor blades adapted to be driven in rotation about the central axis by the turbine means to deliver a flow of air at high pressures for an internal combustion engine, and blades being moveable about longitudinal axes generally transverse to the central axis to impart positive or negative pre-whirl motion to the air leaving the stator blades prior to entering the rotating blades of the compressor stage; closure means for providing a flow of engine exhaust gas from one of the first and second volute passageways into the turbine wheel; and a control means for operating the closure means and the stator blades in synchronization.

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

    SciTech Connect

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

    2008-03-01

    In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Advanced 600 MWe Pressurized Water Reactor (PWR) designed by Westinghouse (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing Light Water Reactors (LWRs). However, it is not clear that the ERVC proposed for the AP600 could provide sufficient heat removal for higher-power reactors (up to 1500 MWe) without additional enhancements. This paper reviews efforts made and results reported regarding the enhancement of IVR in LWRs. Where appropriate, the paper identifies what additional data or analyses are needed to demonstrate that there is sufficient margin for successful IVR in high power thermal reactors.

  3. Vessel with filter and method of use

    SciTech Connect

    Morrell, Jonathan S.; Ripley, Edward B.; Cecala, David M.

    2008-01-29

    Chemical processing apparatuses which incorporate a process vessel, such as a crucible or retort, and which include a gas separation or filtration system. Various embodiments incorporate such features as loose filtration material, semi-rigid filtration material, and structured filtration material. The vessel may include material that is a microwave susceptor. Filtration media may be selected so that if it inadvertently mixes with the chemical process or the reaction products of such process, it would not adversely affect the results of the chemical process.

  4. Experiment 2003 – First Pressurization of EE-2

    SciTech Connect

    Murphy, Hugh D.; Matsunaga, Isao; Kuriyagawa, Michio

    1982-01-07

    Water was pumped into EE-2 at a nominal rate of 9gpm, to a final pressure of 2070 psi. The wellbore was exceptionally tight we might just have well pumped into a steel pressure vessel not only was there no evidence of breakdown, but only a total of about 30 gallons of water permeated the rock during the 2-1/2 hour-long pressurization.

  5. Zone separator for multiple zone vessels

    DOEpatents

    Jones, John B.

    1983-02-01

    A solids-gas contact vessel, having two vertically disposed distinct reaction zones, includes a dynamic seal passing solids from an upper to a lower zone and maintaining a gas seal against the transfer of the separate treating gases from one zone to the other, and including a stream of sealing fluid at the seal.

  6. Final Vitrification Melter And Vessels Evaluation Documentation

    Energy.gov [DOE]

    DOE has prepared final evaluations and made waste incidental to reprocessing determinations for the vitrification melter and feed vessels (the concentrator feed makeup tank and the melter feed hold tank), used by DOE’s West Valley Demonstration Project as part of the process to vitrify waste from prior commercial reprocessing of spent nuclear fuel.

  7. Investigation of Cracked Lithium Hydride Reactor Vessels

    SciTech Connect

    bird, e.l.; mustaleski, t.m.

    1999-06-01

    Visual examination of lithium hydride reactor vessels revealed cracks that were adjacent to welds, most of which were circumferentially located in the bottom portion of the vessels. Sections were cut from the vessels containing these cracks and examined by use of the metallograph, scanning electron microscope, and microprobe to determine the cause of cracking. Most of the cracks originated on the outer surface just outside the weld fusion line in the base material and propagated along grain boundaries. Crack depths of those examined sections ranged from {approximately}300 to 500 {micro}m. Other cracks were reported to have reached a maximum depth of 1/8 in. The primary cause of cracking was the creation of high tensile stresses associated with the differences in the coefficients of thermal expansion between the filler metal and the base metal during operation of the vessel in a thermally cyclic environment. This failure mechanism could be described as creep-type fatigue, whereby crack propagation may have been aided by the presence of brittle chromium carbides along the grain boundaries, which indicates a slightly sensitized microstructure.

  8. In-Vessel Torsional Ultrasonic Wave-Based Level Measurement System - Energy

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Innovation Portal Advanced Materials Advanced Materials Find More Like This Return to Search In-Vessel Torsional Ultrasonic Wave-Based Level Measurement System Oak Ridge National Laboratory Contact ORNL About This Technology Technology Marketing Summary At Three Mile Island in 1979, a partial meltdown of the core was caused by a sudden, undetected loss of reactor coolant water. In the past, a reactor's high temperature and pressure environment has complicated the implementation of level

  9. Wall Insulation; BTS Technology Fact Sheet

    SciTech Connect

    Southface Energy Institute; Tromly, K.

    2000-11-07

    Properly sealed, moisture-protected, and insulated walls help increase comfort, reduce noise, and save on energy costs. This fact sheet addresses these topics plus advanced framing techniques, insulation types, wall sheathings, and steps for effective wall construction and insulation.

  10. Pressure Safety of JLAB 12GeV Upgrade Cryomodule

    SciTech Connect

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

    2009-11-01

    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.

  11. Thermo-hydraulic Simulation of Pressurizer in Transient Cases

    SciTech Connect

    Ardeshir, A.T.; Nematollahi, M.; Sepanloo, K.; Daneshvari, F.

    2004-07-01

    This paper describes a simulation of the pressure adjustment in the primary loop of Pressurized Water Reactors (PWR). A mathematical model is developed for the thermo-hydraulic behavior of pressurizer in transient cases (surge in or out) on the basis of concept of conservation of mass and energy in two phases. No restrictive assumptions have been made. A comparison with RELAP5/Mod3.2 data indicates good overall agreement. The model can be used as a good design verification tool for pressurizer vessels and associated pressure control devices. (authors)

  12. Security_Walls_VPP_Award

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Recognized for Outstanding Safety CARLSBAD, N.M., May 10, 2013 - The U.S. Department of Energy (DOE) has awarded Security Walls, LLC, the Waste Isolation Pilot Plant's (WIPP)...

  13. Stephen C. Walls | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Stephen C. Walls About Us Stephen C. Walls - Energy Transition Initiative Program Lead Most Recent Helping Hawaii Chart a Path to 100% Renewable Electricity November 1

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

    DOEpatents

    Herrmann, Steven D.; Mariani, Robert D.

    2002-01-01

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

  15. Jet-wall interaction effects on diesel combustion and soot formation.

    SciTech Connect

    Pickett, Lyle M.; Lopez, J. Javier

    2004-09-01

    The effects of wall interaction on combustion and soot formation processes of a diesel fuel jet were investigated in an optically-accessible constant-volume combustion vessel at experimental conditions typical of a diesel engine. At identical ambient and injector conditions, soot processes were studied in free jets, plane wall jets, and 'confined' wall jets (a box-shaped geometry simulating secondary interaction with adjacent walls and jets in an engine). The investigation showed that soot levels are significantly lower in a plane wall jet compared to a free jet. At some operating conditions, sooting free jets become soot-free as plane wall jets. Possible mechanisms to explain the reduced or delayed soot formation upon wall interaction include an increased fuel-air mixing rate and a wall-jet-cooling effect. However, in a confined-jet configuration, there is an opposite trend in soot formation. Jet confinement causes combustion gases to be redirected towards the incoming jet, causing the lift-off length to shorten and soot to increase. This effect can be avoided by ending fuel injection prior to the time of significant interaction with redirected combustion gases. For a fixed confined-wall geometry, an increase in ambient gas density delays jet interaction, allowing longer injection durations with no increase in soot. Jet interaction with redirected combustion products may also be avoided using reduced ambient oxygen concentration because of an increased ignition delay. Although simplified geometries were employed, the identification of important mechanisms affecting soot formation after the time of wall interaction is expected to be useful for understanding these processes in more complex and realistic diesel engine geometries.

  16. Fast Flux Test Facility Reactor Vessel Removal Study

    SciTech Connect

    BOWMAN, B.R.

    2002-10-23

    This study assesses the feasibility of removing the FFTF reactor vessel from its current location in the reactor cavity inside the Containment vessel to a transporter for relocation to a burial pit in the 200 Area.

  17. Cover Heated, Open Vessels, Energy Tips: STEAM, Steam Tip Sheet...

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    9 Cover Heated, Open Vessels Open vessels that contain heated liquids often have high heat loss due to surface evaporation. Both energy and liquid losses are reduced by covering ...

  18. Pressure transducer

    DOEpatents

    Anderson, T.T.; Roop, C.J.; Schmidt, K.J.; Gunchin, E.R.

    1987-02-13

    A pressure transducer suitable for use in high temperature environments includes two pairs of induction coils, each pair being bifilarly wound together, and each pair of coils connected as opposite arms of a four arm circuit; an electrically conductive target moveably positioned between the coil pairs and connected to a diaphragm such that deflection of the diaphragm causes axial movement of the target and an unbalance in the bridge output. 7 figs.

  19. Pressure transducer

    DOEpatents

    Anderson, Thomas T.; Roop, Conard J.; Schmidt, Kenneth J.; Gunchin, Elmer R.

    1989-01-01

    A pressure transducer suitable for use in high temperature environments includes two pairs of induction coils, each pair being bifilarly wound together, and each pair of coils connected as opposite arms of a four arm circuit; an electrically conductive target moveably positioned between the coil pairs and connected to a diaphragm such that deflection of the diaphragm causes axial movement of the target and an unbalance in the bridge output.

  20. From Cold War to cold vessels

    SciTech Connect

    Melrath, C.

    1996-09-01

    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.

  1. Starting procedure for internal combustion vessels

    DOEpatents

    Harris, Harry A.

    1978-09-26

    A vertical vessel, having a low bed of broken material, having included combustible material, is initially ignited by a plurality of ignitors spaced over the surface of the bed, by adding fresh, broken material onto the bed to buildup the bed to its operating depth and then passing a combustible mixture of gas upwardly through the material, at a rate to prevent back-firing of the gas, while air and recycled gas is passed through the bed to thereby heat the material and commence the desired laterally uniform combustion in the bed. The procedure permits precise control of the air and gaseous fuel mixtures and material rates, and permits the use of the process equipment designed for continuous operation of the vessel.

  2. MHD Electrode and wall constructions

    DOEpatents

    Way, Stewart; Lempert, Joseph

    1984-01-01

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

  3. Turbine airfoil with outer wall thickness indicators

    DOEpatents

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

    2013-08-06

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

  4. Advanced Pressure Boundary Materials

    SciTech Connect

    Santella, Michael L; Shingledecker, John P

    2007-01-01

    Increasing the operating temperatures of fossil power plants is fundamental to improving thermal efficiencies and reducing undesirable emissions such as CO{sub 2}. One group of alloys with the potential to satisfy the conditions required of higher operating temperatures is the advanced ferritic steels such as ASTM Grade 91, 9Cr-2W, and 12Cr-2W. These are Cr-Mo steels containing 9-12 wt% Cr that have martensitic microstructures. Research aimed at increasing the operating temperature limits of the 9-12 wt% Cr steels and optimizing them for specific power plant applications has been actively pursued since the 1970's. As with all of the high strength martensitic steels, specifying upper temperature limits for tempering the alloys and heat treating weldments is a critical issue. To support this aspect of development, thermodynamic analysis was used to estimate how this critical temperature, the A{sub 1} in steel terminology, varies with alloy composition. The results from the thermodynamic analysis were presented to the Strength of Weldments subgroup of the ASME Boiler & Pressure Vessel Code and are being considered in establishing maximum postweld heat treatment temperatures. Experiments are also being planned to verify predictions. This is part of a CRADA project being done with Alstom Power, Inc.

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

    DOEpatents

    Isaksson, J.; Koskinen, J.

    1995-08-22

    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. Treating exhaust gas from a pressurized fluidized bed reaction system

    DOEpatents

    Isaksson, Juhani; Koskinen, Jari

    1995-01-01

    Hot gases from a pressurized fluidized bed reactor system are purified. Under superatmospheric pressure conditions hot exhaust gases are passed through a particle separator, forming a flitrate 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 superatmospheric 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).

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

    DOEpatents

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

    2013-06-04

    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.

  8. Shear wall ultimate drift limits

    SciTech Connect

    Duffey, T.A.; Goldman, A.; Farrar, C.R.

    1994-04-01

    Drift limits for reinforced-concrete shear walls are investigated by reviewing the open literature for appropriate experimental data. Drift values at ultimate are determined for walls with aspect ratios ranging up to a maximum of 3.53 and undergoing different types of lateral loading (cyclic static, monotonic static, and dynamic). Based on the geometry of actual nuclear power plant structures exclusive of containments and concerns regarding their response during seismic (i.e.,cyclic) loading, data are obtained from pertinent references for which the wall aspect ratio is less than or equal to approximately 1, and for which testing is cyclic in nature (typically displacement controlled). In particular, lateral deflections at ultimate load, and at points in the softening region beyond ultimate for which the load has dropped to 90, 80, 70, 60, and 50 percent of its ultimate value, are obtained and converted to drift information. The statistical nature of the data is also investigated. These data are shown to be lognormally distributed, and an analysis of variance is performed. The use of statistics to estimate Probability of Failure for a shear wall structure is illustrated.

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

    DOEpatents

    Yu, Yufeng Phillip; Itzel, Gary Michael; Webbon, Waylon Willard; Bagepalli, Radhakrishna; Burdgick, Steven Sebastian; Kellock, Iain Robertson

    2002-01-01

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

  10. Microsoft Word - Errata for the Pressure and Vacuum Systems Safety Supplement 3-15

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    SYSTEM TURNOVER FORM PS-9 Pressure System Number: Pressure System Name: OPERATING REQUIREMENTS: MAINTENANCE REQUIREMENTS: IN-SERVICE INSPECTION REQUIREMENTS: Piping Vessels Relief Valves Component Component ISI Category ISI Type (VT, UT, RT, etc) ISI Frequency Special ISI Requirements: System Owner name and signature: Date: Design Authority name and signature Date Design Authority shall forward this form to the Pressure Systems Committee Designee for filing and updating the operating pressure

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

    SciTech Connect

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

    2014-09-23

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

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

    DOEpatents

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

    2008-10-07

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

  13. Inservice leak testing of primary pressure isolation valves. Final report

    SciTech Connect

    Livingston, R.A.

    1983-02-01

    This report discusses the inservice leak testing of primary pressure isolation valves in commercial power reactors which was investigated to identify problems with current test procedures and requirements. Nine utilities were surveyed to gather information which is presented in this report. An analysis of the survey information was performed, resulting in recommended changes to improve valve leak testing requirements currently invoked by Section XI of the ASME Boiler and Pressure Vessel Code, Plant Technical Specifications, and Regulatory Guides addressing this subject.

  14. Inservice leak testing of primary pressure isolation valves

    SciTech Connect

    Livingston, R.A.

    1983-02-01

    This report discusses the inservice leak testing of primary pressure isolation valves in commercial power reactors which was investigated to identify problems with current test procedures and requirements. Nine utilities were surveyed to gather information which is presented in this report. An analysis of the survey information was performed, resulting in recommended changes to improve valve leak testing requirements currently invoked by Section XI of the ASME Boiler and Pressure Vessel Code, Plant Technical Specifications, and Regulatory Guides addressing this subject.

  15. DYNA3D Finite Element Analysis of Steam Explosion Loads on a Pedestal Wall Design

    SciTech Connect

    Noble, C R

    2007-01-18

    The objective of this brief report is to document the ESBWR pedestal wall finite element analyses that were performed as a quick turnaround effort in July 2005 at Lawrence Livermore National Laboratory and describe the assumptions and failure criteria used for these analyses [Ref 4]. The analyses described within are for the pedestal wall design that included an internal steel liner. The goal of the finite element analyses was to assist in determining the load carrying capacity of the ESBWR pedestal wall subjected to an impulsive pressure generated by a steam explosion.

  16. CNG Exports by Vessel out of the U.S. | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Vessel out of the U.S. CNG Exports by Vessel out of the U.S. CNG Exports by Vessel Form (Excel) (40.5 KB) CNG Exports by Vessel Form (pdf) (10.88 KB) More Documents & Publications CNG Imports by Vessel into the U.S. Other Imports by Truck into the U.S. Other Exports by Vessel

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

    DOEpatents

    Isaksson, J.

    1996-02-20

    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.

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

    DOEpatents

    Isaksson, Juhani

    1996-01-01

    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.

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

    SciTech Connect

    Raske, D.T.

    1995-06-01

    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.

  20. Method for forming a bladder for fluid storage vessels

    DOEpatents

    Mitlitsky, Fred; Myers, Blake; Magnotta, Frank

    2000-01-01

    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.

  1. ITER Port Interspace Pressure Calculations

    SciTech Connect

    Carbajo, Juan J; Van Hove, Walter A

    2016-01-01

    The ITER Vacuum Vessel (VV) is equipped with 54 access ports. Each of these ports has an opening in the bioshield that communicates with a dedicated port cell. During Tokamak operation, the bioshield opening must be closed with a concrete plug to shield the radiation coming from the plasma. This port plug separates the port cell into a Port Interspace (between VV closure lid and Port Plug) on the inner side and the Port Cell on the outer side. This paper presents calculations of pressures and temperatures in the ITER (Ref. 1) Port Interspace after a double-ended guillotine break (DEGB) of a pipe of the Tokamak Cooling Water System (TCWS) with high temperature water. It is assumed that this DEGB occurs during the worst possible conditions, which are during water baking operation, with water at a temperature of 523 K (250 C) and at a pressure of 4.4 MPa. These conditions are more severe than during normal Tokamak operation, with the water at 398 K (125 C) and 2 MPa. Two computer codes are employed in these calculations: RELAP5-3D Version 4.2.1 (Ref. 2) to calculate the blowdown releases from the pipe break, and MELCOR, Version 1.8.6 (Ref. 3) to calculate the pressures and temperatures in the Port Interspace. A sensitivity study has been performed to optimize some flow areas.

  2. Transpiring wall supercritical water oxidation reactor salt deposition studies

    SciTech Connect

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

    1996-09-01

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

  3. Tube wall thickness measurement apparatus

    DOEpatents

    Lagasse, Paul R.

    1987-01-01

    An apparatus for measuring the thickness of a tube's wall for the tube's entire length and circumference by determining the deviation of the tube wall thickness from the known thickness of a selected standard item. The apparatus comprises a base and a first support member having first and second ends. The first end is connected to the base and the second end is connected to a spherical element. A second support member is connected to the base and spaced apart from the first support member. A positioning element is connected to and movable relative to the second support member. An indicator is connected to the positioning element and is movable to a location proximate the spherical element. The indicator includes a contact ball for first contacting the selected standard item and holding it against the spherical element. The contact ball then contacts the tube when the tube is disposed about the spherical element. The indicator includes a dial having a rotatable needle for indicating the deviation of the tube wall thickness from the thickness of the selected standard item.

  4. Tube wall thickness measurement apparatus

    DOEpatents

    Lagasse, P.R.

    1985-06-21

    An apparatus for measuring the thickness of a tube's wall for the tube's entire length and radius by determining the deviation of the tube wall thickness from the known thickness of a selected standard item. The apparatus comprises a base and a first support member having first and second ends. The first end is connected to the base and the second end is connected to a spherical element. A second support member is connected to the base and spaced apart from the first support member. A positioning element is connected to and movable relative to the second support member. An indicator is connected to the positioning element and is movable to a location proximate the spherical element. The indicator includes a contact ball for first contacting the selected standard item and holding it against the spherical element. The contact ball then contacts the tube when the tube is disposed about the spherical element. The indicator includes a dial having a rotatable needle for indicating the deviation of the tube wall thickness from the thickness of the selected standard item.

  5. Wall System Innovations: Familiar Materials, Better Performance |

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Department of Energy Wall System Innovations: Familiar Materials, Better Performance Wall System Innovations: Familiar Materials, Better Performance This presentation was delivered at the U.S. Department of Energy Building America Technical Update meeting on April 29-30, 2013, in Denver, Colorado. wall_system_innovations_kochkin.pdf (1.48 MB) More Documents & Publications Building America New Homes Case Study: Advanced Extended Plate and Beam Wall System in a Cold-Climate House Building

  6. Mooring system for oil tanker storage vessel or the like

    SciTech Connect

    Hvide, H.J.

    1993-08-24

    A mooring system for an ocean going vessel, said vessel hull having a thickness, said system comprising: (a) a rigid shaft having an upper end and a lower end, said shaft being immovably fixed at said upper end to said vessel and said lower end of said shaft being disposed beneath and external of said hull; and (b) a chain table rotatably mounted on said lower end of said rigid shaft.

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

    SciTech Connect

    Not Available

    1984-11-01

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

  8. An open-walled ionization chamber appropriate to tritium monitoring for glovebox

    SciTech Connect

    Chen Zhilin; Chang Ruiming; Mu Long; Song Guoyang; Wang Heyi; Wu Guanyin; Wei Xiye

    2010-07-15

    An open-walled ionization chamber is developed to monitor the tritium concentration in gloveboxes in tritium processing systems. Two open walls are used to replace the sealed wall in common ionization chambers, through which the tritium gas can diffuse into the chamber without the aid of pumps and pipelines. Some basic properties of the chamber are examined to evaluate its performance. Results turn out that an open-walled chamber of 1 l in volume shows a considerably flat plateau over 700 V for a range of tritium concentration. The chamber also gives a good linear response to gamma fields over 4 decades under a pressure condition of 1 atm. The pressure dependence characteristics show that the ionization current is only sensitive at low pressures. The pressure influence becomes weaker as the pressure increases mainly due to the decrease in the mean free path of {beta} particles produced by tritium decay. The minimum detection limit of the chamber is 3.7x10{sup 5} Bq/m{sup 3}.

  9. Promising Technology: Cool Paints for Exterior Walls

    Energy.gov [DOE]

    Cool Paints increase the solar reflectance of exterior walls. By reflecting more sunlight, the wall surface maintains a cooler temperature. This decrease in temperature leads to less heat transfer through the walls into the building. During the cooling season, the addition of cool paints can decrease the cooling load of the building.

  10. Webinar: Material Characterization of Storage Vessels for Fuel...

    Energy.gov [DOE] (indexed site)

    Above is the video recording for the webinar, "Material Characterization of Storage Vessels for Fuel Cell Forklifts," originally held on August 14, 2012. In addition to this ...

  11. 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 ... The installation of offshore wind farms requires a highly specialized fleet of ...

  12. Method and device for supporting blood vessels during anastomosis

    DOEpatents

    Doss, J.D.

    1985-05-20

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

  13. Experimental evaluation of solids suspension uniformity in canyon process vessels

    SciTech Connect

    Hassan, N.M.

    1996-06-25

    Experimental evaluation of solids suspension in canyon process vessels was performed at several paddle agitator speeds and different volume levels in a geometrically similar vessel. The paddle agitator speeds examined were 280, 370, 528, and 686 rpm and volume levels were 30%, 50%, and 70% fill capacity. Experiments were conducted with simulated solid particles that have particle size range and density similar to plutonium particles and corrosion products typically seen in canyon vessels. Solids suspension took place in baffled cylindrical vessel equipped with two flat-blade agitators and cooling helices.

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

    Energy.gov [DOE]

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

  15. Method for Preparing Nanoporous Cell-Scaled Reaction Vessels...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Return to Search Method for Preparing Nanoporous Cell-Scaled Reaction Vessels Oak Ridge ... reactions at the microscale level. *Cell-mimicking structures can be prepared as ...

  16. Shock wave convergence in water with parabolic wall boundaries

    SciTech Connect

    Yanuka, D.; Shafer, D.; Krasik, Ya.

    2015-04-28

    The convergence of shock waves in water, where the cross section of the boundaries between which the shock wave propagates is either straight or parabolic, was studied. The shock wave was generated by underwater electrical explosions of planar Cu wire arrays using a high-current generator with a peak output current of ?45?kA and rise time of ?80?ns. The boundaries of the walls between which the shock wave propagates were symmetric along the z axis, which is defined by the direction of the exploding wires. It was shown that with walls having a parabolic cross section, the shock waves converge faster and the pressure in the vicinity of the line of convergence, calculated by two-dimensional hydrodynamic simulations coupled with the equations of state of water and copper, is also larger.

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

    DOEpatents

    Davin, J.

    1992-12-01

    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.

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

    DOEpatents

    Davin, James

    1992-01-01

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

  19. Method and apparatus for coupling seismic sensors to a borehole wall

    DOEpatents

    West, Phillip B.

    2005-03-15

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

  20. Mechanism of bubble detachment from vibrating walls

    SciTech Connect

    Kim, Dongjun; Park, Jun Kwon Kang, Kwan Hyoung; Kang, In Seok

    2013-11-15

    We discovered a previously unobserved mechanism by which air bubbles detach from vibrating walls in glasses containing water. Chaotic oscillation and subsequent water jets appeared when a wall vibrated at greater than a critical level. Wave forms were developed at water-air interface of the bubble by the wall vibration, and water jets were formed when sufficiently grown wave-curvatures were collapsing. Droplets were pinched off from the tip of jets and fell to the surface of the glass. When the solid-air interface at the bubble-wall attachment point was completely covered with water, the bubble detached from the wall. The water jets were mainly generated by subharmonic waves and were generated most vigorously when the wall vibrated at the volume resonant frequency of the bubble. Bubbles of specific size can be removed by adjusting the frequency of the wall's vibration.

  1. POROUS WALL, HOLLOW GLASS MICROSPHERES

    SciTech Connect

    Sexton, W.

    2012-06-30

    Hollow Glass Microspheres (HGM) is not a new technology. All one has to do is go to the internet and Google{trademark} HGM. Anyone can buy HGM and they have a wide variety of uses. HGM are usually between 1 to 100 microns in diameter, although their size can range from 100 nanometers to 5 millimeters in diameter. HGM are used as lightweight filler in composite materials such as syntactic foam and lightweight concrete. In 1968 a patent was issued to W. Beck of the 3M{trademark} Company for 'Glass Bubbles Prepared by Reheating Solid Glass Particles'. In 1983 P. Howell was issued a patent for 'Glass Bubbles of Increased Collapse Strength' and in 1988 H. Marshall was issued a patent for 'Glass Microbubbles'. Now Google{trademark}, Porous Wall, Hollow Glass Microspheres (PW-HGMs), the key words here are Porous Wall. Almost every article has its beginning with the research done at the Savannah River National Laboratory (SRNL). The Savannah River Site (SRS) where SRNL is located has a long and successful history of working with hydrogen and its isotopes for national security, energy, waste management and environmental remediation applications. This includes more than 30 years of experience developing, processing, and implementing special ceramics, including glasses for a variety of Department of Energy (DOE) missions. In the case of glasses, SRS and SRNL have been involved in both the science and engineering of vitreous or glass based systems. As a part of this glass experience and expertise, SRNL has developed a number of niches in the glass arena, one of which is the development of porous glass systems for a variety of applications. These porous glass systems include sol gel glasses, which include both xerogels and aerogels, as well as phase separated glass compositions, that can be subsequently treated to produce another unique type of porosity within the glass forms. The porous glasses can increase the surface area compared to 'normal glasses of a 1 to 2 order of

  2. Fluid-solid contact vessel having fluid distributors therein

    DOEpatents

    Jones, Jr., John B.

    1980-09-09

    Rectangularly-shaped fluid distributors for large diameter, vertical vessels include reinforcers for high heat operation, vertical sides with gas distributing orifices and overhanging, sloped roofs. Devices are provided for cleaning the orifices from a buildup of solid deposits resulting from the reactions in the vessel.

  3. 2013 Wall Street Perspective on SMRs | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Wall Street Perspective on SMRs 2013 Wall Street Perspective on SMRs Summary of results of a 2013 survey on Wall Street attitudes toward small modular reactors. PDF icon 2013 Wall ...

  4. SRNL POROUS WALL GLASS MICROSPHERES

    SciTech Connect

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

    2008-04-15

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

  5. Response of a water-filled spherical vessel to an internal explosion

    SciTech Connect

    Lewis, M.W.; Wilson, T.L.

    1997-06-01

    Many problems of interest to the defense community involve fluid-structure interaction (FSI). Such problems include underwater blast loading of structures, bubble dynamics and jetting around structures, and hydrodynamic ram events. These problems may involve gas, fluid, and solid dynamics, nonlinear material behavior, cavitation, reaction kinetics, material failure, and nonlinearity that is due to varying geometry and contact conditions within a structure or between structures. Here, the authors model the response of a water-filled, thick-walled, spherical steel vessel to an internal explosion of 30 grams of C-4 with FSI2D--a two-dimensional coupled finite element and finite volume hydrodynamics code. The gas phase detonation products were modeled with a Becker-Kistiakowsky-Wilson high-explosive equation of state. Predictions from a fully coupled model were compared to experimental results in the form of strain gauge traces. Agreement was reasonably good. Additionally, the calculation was run in an uncoupled mode to understand the importance of fluid-structure interaction in this problem. The uncoupled model results in an accumulation of nonphysical energy in the vessel.

  6. CNG Imports by Vessel into the U.S. | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Vessel into the U.S. CNG Imports by Vessel into the U.S. CNG Imports by Vessel Form (Excel) (41 KB) CNG Imports by Vessel Form (pdf) (14.24 KB) More Documents & Publications Other Imports by Vessel into the U.S. Other Imports by Truck

  7. Other Imports by Vessel into the U.S. | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Vessel into the U.S. Other Imports by Vessel into the U.S. Other Imports by Vessel Form (Excel) (41 KB) Other Imports by Vessel Form (pdf) (14.23 KB) More Documents & Publications CNG Imports by Vessel into the U.S. Other Imports by Truck

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

    DOEpatents

    Hoult, D.P.

    1985-09-10

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

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

    DOEpatents

    Hoult, David P.

    1985-01-01

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

  10. Improvement of the mechanical reliability of monolithic refractory linings for coal gasification process vessels. Final report

    SciTech Connect

    Potter, R.A.

    1981-09-01

    Eighteen heat-up tests were run on nine standard and experimental dual component monolithic refractory concrete linings. These tests were run with a five foot diameter by 14-ft high Pressure Vessel/Test Furnace designed to accommodate a 12-inch thick by 5-ft high refractory lining, heat the hot face to 2000/sup 0/F and expose the lining to air or steam pressures up to 150 psig. Results obtained from standard type linings in the test facility indicated that lining degradation duplicated that observed in field installations. The lining performance was significantly improved due to information gained from a systematic study of the cracking that occurred in the linings; the analysis of the lining strains, shell stresses and acoustic emission results; and the stress analyses performed on the standard and experimental lining designs with the finite element analysis computer programs, REFSAM and RESGAP.

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

    SciTech Connect

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

    2014-09-02

    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.

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

    SciTech Connect

    Yongfeng Zhang; Pritam Chakraborty; S. Bulent Biner

    2013-09-01

    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.

  13. A local criterion for cleavage fracture of a nuclear pressure vessel steel

    SciTech Connect

    Bermin, F.M.

    1983-11-01

    Experiments were performed on three heats of A508 class 3 steel in order to determine the mechanical conditions for cleavage fracture. These tests were carried out on various geometries including 4-point bend specimens and axisymmetric notched tensile bars with different notch radii which have been modelized using the finite element method. In one heat, the temperature range investigated was from 77 K to 233 K. It is shown that the cleavage resistance is increased by tensile straining. Moreover, the probability of fracture obeys the Weibull statistical distribution. All the results can be accounted for in terms of a local criterion based on Weibull theory and which takes into account the effect of plastic strain. In this criterion, the parameters which were experimentally determined are found to be temperature independent over the range 77 K to 170 K. The applicability of the approach proposed for cleavage fracture at the crack tip is also examined. It is shown that the experimental results published in the literature giving the variation of fracture toughness with temperature can be explained by the proposed criterion which predicts reasonably well both the scatter in the experimental results and the K /SUB IC/ temperature dependence.

  14. Studies of stress corrosion cracking in steels used for reactor pressure vessels

    SciTech Connect

    Van Der Sluys, W.A.; Pathania, R.

    1992-12-31

    This paper reviews the state of technology concerning stress corrosion crack growth in LWR environments and reports the results from a series of experiments that attempted to duplicate the results obtained from the literature. These experiments include one conducted in a steam environment representative of the condition in the upper head of a boiling-water reactor (BWR).

  15. Analysis of unclad and sub-clad semi-elliptical flaws in pressure vessel steels

    SciTech Connect

    Irizarry-Quinones, H.; Macdonald, B.D.; McAfee, W.J.

    1996-06-01

    This study was conducted to support warm prestressing experiments on unclad and sub-clad flawed beams loaded in pure bending. Two cladding yield strengths were investigated: 0.6 Sy and 0.8 Sy, where Sy is the yield strength of the base metal. Cladding and base metal were assumed to be stress free at the stress relief temperature for the 3D elastic-plastic finite element analysis used to model the experiments. The model results indicated that when cooled from the stress relief temperature, the cladding was put in tension due to its greater coefficient of thermal expansion. When cooled, the cladding exhibited various amounts of tensile yielding. The degree of yielding depended on the amount of cooling and the strength of the cladding relative to that of the base metal. When subjected to tensile bending stress, the sub-clad flaw elastic-plastic stress intensity factor, K{sub I}(J), was at first dominated by crack closing force due to tensile yielding in the cladding. Thus, imposed loads initially caused no increase in K{sub I}(J) near the clad-base interface. However, K{sub I}(J) at the flaw depth was little affected. When the cladding residual stress was overcome, K{sub I}(J) gradually increased until the cladding began to flow. Thereafter, the rate at which K{sub I}(J) increased with load was the same as that of an unclad beam. A plastic zone corrected K{sub I} approximation for the unclad flaw was found by the superposition of standard Newman and Raju solutions with those due to a cladding crack closure force approximated by the Kaya and Erdogan solution. These elastic estimates of the effect of cladding in reducing the crack driving force were quite in keeping with the 3D elastic-plastic finite element solution for the sub-clad flaw. The results were also compared with the analysis of clad beam experiments by Keeney and the conclusions by Miyazaki, et al. A number of sub-clad flaw specimens not subjected to warm prestressing were thought to have suffered degraded toughness caused by locally intensified strain aging embrittlement (LISAE) due to welding over the preexisting flaw.

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

    Energy.gov [DOE]

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

  17. Welding the AT-400A Containment Vessel

    SciTech Connect

    Brandon, E.

    1998-11-01

    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.

  18. 2015 Wall Street Perspectives on SMRs Update | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    5 Wall Street Perspectives on SMRs Update 2015 Wall Street Perspectives on SMRs Update Summary briefing on the results of a September 2015 survey of Wall Street attitudes toward SMRs. 2015 Financial Survey: Nuclear Energy (415.36 KB) More Documents & Publications 2014 Wall Street Perspective on SMRs 2014 Wall Street Perspectives on SMRs - Report 2013 Wall Street Perspective on SMRs

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

    SciTech Connect

    Klein, J.

    2010-12-14

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

  20. Seismic behavior of geogrid reinforced slag wall

    SciTech Connect

    Edincliler, Ayse; Baykal, Gokhan; Saygili, Altug

    2008-07-08

    Flexible retaining structures are known with their high performance under earthquake loads. In geogrid reinforced walls the performance of the fill material and the interface of the fill and geogrid controls the performance. Geosynthetic reinforced walls in seismic regions must be safe against not only static forces but also seismic forces. The objective of this study is to determine the behavior of a geogrid reinforced slag wall during earthquake by using shaking table experiments. This study is composed of three stages. In the first stage the physical properties of the material to be used were determined. In the second part, a case history involving the use of slag from steel industry in the construction of geogrid reinforced wall is presented. In the third stage, the results of shaking table tests conducted using model geogrid wall with slag are given. From the results, it is seen that slag can be used as fill material for geogrid reinforced walls subjected to earthquake loads.

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

    SciTech Connect

    Krishnan, S.; Bhasin, V.; Mahajan, S.C.

    1997-04-01

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

  2. Pressurized Combustion and Gasification

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    ... (right) of Sandia's pressurized entrained flow reactor (PEFR) for investigating the pressurized combustion and gasification characteristics of solid fuels such as coal and biomass. ...

  3. First wall for polarized fusion reactors

    DOEpatents

    Greenside, Henry S.; Budny, Robert V.; Post, Jr., Douglass E.

    1988-01-01

    Depolarization mechanisms arising from the recycling of the polarized fuel at the limiter and the first-wall of a fusion reactor are greater than those mechanisms in the plasma. Rapid depolarization of the plasma is prevented by providing a first-wall or first-wall coating formed of a low-Z, non-metallic material having a depolarization rate greater than 1 sec.sup.-1.

  4. Textural break foundation wall construction modules

    DOEpatents

    Phillips, Steven J.

    1990-01-01

    Below-grade, textural-break foundation wall structures are provided for inhibiting diffusion and advection of liquids and gases into and out from a surrounding hydrogeologic environment. The foundation wall structure includes a foundation wall having an interior and exterior surface and a porous medium disposed around a portion of the exterior surface. The structure further includes a modular barrier disposed around a portion of the porous medium. The modular barrier is substantially removable from the hydrogeologic environment.

  5. Panelized wall system with foam core insulation

    SciTech Connect

    Kosny, Jan; Gaskin, Sally

    2009-10-20

    A wall system includes a plurality of wall members, the wall members having a first metal panel, a second metal panel, and an insulating core between the first panel and the second panel. At least one of the first panel and the second panel include ridge portions. The insulating core can be a foam, such as a polyurethane foam. The foam can include at least one opacifier to improve the k-factor of the foam.

  6. Diagnostics - Rotating Wall Machine - UW Plasma Physics

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Utilizing shunt resistors on the plasma source and bias capacitors, the amount of current ... Within the vacuum vessel, three rings of ten Btheta coils are mounted at the plasma edge. ...

  7. Virtual gap dielectric wall accelerator

    DOEpatents

    Caporaso, George James; Chen, Yu-Jiuan; Nelson, Scott; Sullivan, Jim; Hawkins, Steven A

    2013-11-05

    A virtual, moving accelerating gap is formed along an insulating tube in a dielectric wall accelerator (DWA) by locally controlling the conductivity of the tube. Localized voltage concentration is thus achieved by sequential activation of a variable resistive tube or stalk down the axis of an inductive voltage adder, producing a "virtual" traveling wave along the tube. The tube conductivity can be controlled at a desired location, which can be moved at a desired rate, by light illumination, or by photoconductive switches, or by other means. As a result, an impressed voltage along the tube appears predominantly over a local region, the virtual gap. By making the length of the tube large in comparison to the virtual gap length, the effective gain of the accelerator can be made very large.

  8. Beetle Kill Wall at NREL

    ScienceCinema

    None

    2013-05-29

    When it comes to designing an interior decorative feature for one of the most energy efficient office buildings in the world, very few would consider bringing in a beetle to do the job. But thats what happened at the U.S. Department of Energy's (DOE) Research Support Facility (RSF) located on the National Renewable Energy Laboratory (NREL) campus.In June, the RSF will become home to more than 800 workers from DOE and NREL and building visitors will be greeted with a soaring, two-story high wall entirely covered with wood harvested from the bark beetle infestation that has killed millions of pine trees in the Western U.S. But, the use of beetle kill wood is just one example of the resources being leveraged to make the RSF a model for sustainability and one more step toward NRELs goal to be a net zero energy campus.

  9. Beetle Kill Wall at NREL

    SciTech Connect

    2010-01-01

    When it comes to designing an interior decorative feature for one of the most energy efficient office buildings in the world, very few would consider bringing in a beetle to do the job. But thats what happened at the U.S. Department of Energy's (DOE) Research Support Facility (RSF) located on the National Renewable Energy Laboratory (NREL) campus.In June, the RSF will become home to more than 800 workers from DOE and NREL and building visitors will be greeted with a soaring, two-story high wall entirely covered with wood harvested from the bark beetle infestation that has killed millions of pine trees in the Western U.S. But, the use of beetle kill wood is just one example of the resources being leveraged to make the RSF a model for sustainability and one more step toward NRELs goal to be a net zero energy campus.

  10. Fillability of Thin-Wall Steel Castings

    SciTech Connect

    Robert C. Voigt; Joseph Bertoletti; Andrew Kaley; Sandi Ricotta; Travis Sunday

    2002-07-30

    The use of steel components is being challenged by lighter nonferrous or cast iron components. The development of techniques for enhancing and ensuring the filability of thin-wall mold cavities is most critical for thinner wall cast steel production. The purpose of this research was to develop thin-wall casting techniques that can be used to reliably produce thin-wall castings from traditional gravity poured sand casting processes. The focus of the research was to enhance the filling behavior to prevent misrunds. Experiments were conducted to investigate the influence of various foundry variables on the filling of thin section steel castings. These variables include casting design, heat transfer, gating design, and metal fluidity. Wall thickness and pouring temperature have the greatest effect on casting fill. As wall thickness increases the volume to surface area of the casting increases, which increases the solidification time, allowing the metal to flow further in thicker sect ions. Pouring time is another significant variable affecting casting fill. Increases or decreases of 20% in the pouring time were found to have a significant effect on the filling of thin-wall production castings. Gating variables, including venting, pouring head height, and mold tilting also significantly affected thin-wall casting fill. Filters offer less turbulent, steadier flow, which is appropriate for thicker castings, but they do not enhance thin-wall casting fill.

  11. Living Walls | OpenEI Community

    OpenEI (Open Energy Information) [EERE & EIA]

    systems and the new field of biomimicry. Biomimicry is the science of imitating nature to solve human design problems. The Living Wall concept takes the principles behind...

  12. Water Wall Turbine | Open Energy Information

    OpenEI (Open Energy Information) [EERE & EIA]

    Jump to: navigation, search Name: Water Wall Turbine Region: Canada Sector: Marine and Hydrokinetic Website: www.wwturbine.com This company is listed in the Marine and Hydrokinetic...

  13. First wall for polarized fusion reactors

    DOEpatents

    Greenside, H.S.; Budny, R.V.; Post, D.E. Jr.

    1985-01-29

    A first-wall or first-wall coating for use in a fusion reactor having polarized fuel may be formed of a low-Z non-metallic material having slow spin relaxation, i.e., a depolarization rate greater than 1 sec/sup -1/. Materials having these properties include hydrogenated and deuterated amorphous semiconductors. A method for preventing the rapid depolarization of a polarized plasma in a fusion device may comprise the step of providing a first-wall or first-wall coating formed of a low-Z, non-metallic material having a depolarization rate greater than 1 sec/sup -1/.

  14. CXD 4606, 9831 Wall Construction Project (4606)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    9831 Wall Construction Project (4606) Y-12 Site Office Oak Ridge, Anderson County, Tennessee The proposed action is to upgrade of the existing contamination area associated with an...

  15. Wall, Pennsylvania: Energy Resources | Open Energy Information

    OpenEI (Open Energy Information) [EERE & EIA]

    Wall, Pennsylvania: Energy Resources Jump to: navigation, search Equivalent URI DBpedia Coordinates 40.3936801, -79.7861577 Show Map Loading map... "minzoom":false,"mappingser...

  16. Multiple moving wall dry coal extrusion pump

    DOEpatents

    Fitzsimmons, Mark Andrew

    2013-05-14

    A pump for transporting particulate material includes a passageway defined on each side between an inlet and an outlet by a moving wall.

  17. new chemistry to break down cell walls

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    new chemistry to break down cell walls - Sandia Energy Energy Search Icon Sandia Home ... Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power ...

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

    DOEpatents

    Powell, J.G.

    1993-11-16

    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.

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

    DOEpatents

    Schabron, John F.; Rovani, Jr., Joseph F.

    2012-09-25

    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.

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

    DOEpatents

    Schabron, John F.; Rovani, Jr., Joseph F.

    2011-01-25

    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.

  1. Nondestructive Technique Survey for Assessing Integrity of Composite Firing Vessel

    SciTech Connect

    Tran, A.

    2000-08-01

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

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

    DOEpatents

    Powell, James G.

    1993-01-01

    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. PERFORMANCE OF A CONTAINMENT VESSEL CLOSURE FOR RADIOACTIVE GAS CONTENTS

    SciTech Connect

    Blanton, P.; Eberl, K.

    2010-07-09

    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.

  4. Moisture Management of High-R Walls (Fact Sheet), Building America Case Study: Technology Solutions for New and Existing Homes, Building Technologies Office (BTO)

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Management of High-R Walls PROJECT APPLICATION Construction: Existing homes with vapor open wall assemblies Type: Residential Climate Zones: All TECHNICAL PARAMETERS Moisture Sources: Water vapor is moisture that flows from areas of high vapor pressure to low vapor pressure. It can accumulate behind vapor impermeable materials. Construction moisture is moisture that is stored in materials when exposed to the environment during construction. Rain is the primary source. This was simulated with a

  5. Pressurizer with a mechanically attached surge nozzle thermal sleeve

    SciTech Connect

    Wepfer, Robert M

    2014-03-25

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

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

    DOEpatents

    Sappok, Alexander; Wong, Victor

    2014-11-18

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

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

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Vessel out of the U.S. Form LNG Exports by Vessel out of the U.S. Form LNG Exports by Vessel Form (Excel) (40.5 KB) LNG Exports by Vessel Form (pdf) (10.9 KB) More Documents & Publications Other Imports by Truck into the U.S. Other Exports by Vessel out of the U.S. CNG Exports by Vessel

  8. LNG Imports by Vessel in ISO Containers into the U.S. | Department of

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Energy Vessel in ISO Containers into the U.S. LNG Imports by Vessel in ISO Containers into the U.S. LNG Imports by Vessel in ISO Containers Form (Excel) (41 KB) LNG Imports by Vessel in ISO Containers Form (pdf) (22.77 KB) More Documents & Publications Complete Set of all Reporting Forms LNG Re-Exports by Vessel out of the U.S. LNG Exports by Vessel

  9. Preliminary development of an integrated approach to the evaluation of pressurized thermal shock as applied to the Oconee Unit 1 Nuclear Power Plant

    SciTech Connect

    Burns, T J; Cheverton, R D; Flanagan, G F; White, J D; Ball, D G; Lamonica, L B; Olson, R

    1986-05-01

    An evaluation of the risk to the Oconee-1 nuclear plant due to pressurized thermal shock (PTS) has been Completed by Oak Ridge National Laboratory (ORNL). This evaluaion was part of a Nuclear Regulatory Commission (NRC) program designed to study the PTS risk to three nuclear plants: Oconee-1, a Babcock and Wilco reactor plant owned and operated by Duke Power Company; Calvert Cliffs-1, a Combustion Engineering reactor plant owned and operated by Baltimore Gas and Electric company; and H.B. Robinson-2, a Westinghouse reactor plant owned and operated by Carolina Power and Light Company. Studies of Calvert Cliffs-1 and H.B. Robinson-2 are still underway. The specific objectives of the Oconee-1 study were to: (1) provide a best estimate of the probability of a through-the-wall crack (TWC) occurring in the reactor pressure vessel as a result of PTS; (2) determine dominant accident sequences, plant features, operator and control actions and uncertainty in the PTS risk; and (3) evaluate effectiveness of potential corrective measures.

  10. External Insulation of Masonry Walls and Wood Framed Walls

    SciTech Connect

    Baker, P.

    2013-01-01

    The use of exterior insulation on a building is an accepted and effective means to increase the overall thermal resistance of the assembly that also has other advantages of improved water management and often increased air tightness of building assemblies. For thin layers of insulation (1” to 1 ½”), the cladding can typically be attached directly through the insulation back to the structure. For thicker insulation layers, furring strips have been added as a cladding attachment location. This approach has been used in the past on numerous Building America test homes and communities (both new and retrofit applications), and has been proven to be an effective and durable means to provide cladding attachment. However, the lack of engineering data has been a problem for many designers, contractors, and code officials. This research project developed baseline engineering analysis to support the installation of thick layers of exterior insulation on existing masonry and frame walls. Furthermore, water management details necessary to integrate windows, doors, decks, balconies and roofs were created to provide guidance on the integration of exterior insulation strategies with other enclosure elements.

  11. External Insulation of Masonry Walls and Wood Framed Walls

    SciTech Connect

    Baker, P.

    2013-01-01

    The use of exterior insulation on a building is an accepted and effective means to increase the overall thermal resistance of the assembly that also has other advantages of improved water management and often increased air tightness of building assemblies. For thin layers of insulation (1" to 1 1/2"), the cladding can typically be attached directly through the insulation back to the structure. For thicker insulation layers, furring strips have been added as a cladding attachment location. This approach has been used in the past on numerous Building America test homes and communities (both new and retrofit applications), and has been proven to be an effective and durable means to provide cladding attachment. However, the lack of engineering data has been a problem for many designers, contractors, and code officials. This research project developed baseline engineering analysis to support the installation of thick layers of exterior insulation on existing masonry and frame walls. Furthermore, water management details necessary to integrate windows, doors, decks, balconies and roofs were created to provide guidance on the integration of exterior insulation strategies with other enclosure elements.

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

    SciTech Connect

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

    2010-07-01

    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.

  13. Furnace Pressure Controllers

    Energy.gov [DOE]

    This tip sheet highlights the benefits of precise furnace pressure control in process heating systems.

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

    SciTech Connect

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

    2012-09-01

    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

  15. 2014 Wall Street Perspectives on SMRs - Report | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    4 Wall Street Perspectives on SMRs - Report 2014 Wall Street Perspectives on SMRs - Report Summary of the results of a 2014 survey of Wall Street attitudes toward small modular reactors. View from Wall Street: Nuclear Energy and Small Modular Reactors (284.42 KB) More Documents & Publications 2013 Wall Street Perspective on SMRs 2014 Wall Street Perspective on SMRs 2015 Wall Street Perspectives on SMRs Update

  16. Comparison of Alternatives to the 2004 Vacuum Vessel Heat Transfer...

    Office of Scientific and Technical Information (OSTI)

    as well as including a small safety-rated pump and HX in parallel to the main circulation pump and HX. The Vacuum Vessel (VV) Primary Heat Transfer System (PHTS) removes heat...

  17. Cover Heated, Open Vessels - Steam Tip Sheet #19

    SciTech Connect

    2012-01-01

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

  18. Analyses of High Pressure Molten Debris Dispersion for a Typical PWR Plant

    SciTech Connect

    Osamu KAawabata; Mitsuhiro Kajimoto [Japan Nuclear Energy Safety Organization (Japan)

    2006-07-01

    In such severe core damage accident, as small LOCAs with no ECCS injection or station blackout, in which the primary reactor system remains pressurized during core melt down, certain modes of vessel failure would lead to a high pressure ejection of molten core material. In case of a local failure of the lower head, the molten materials would initially be ejected into the cavity beneath the pressure vessel may subsequently be swept out from the cavity to the containment atmosphere and it might cause the early containment failure by direct contact of containment steel liner with core debris. When the contribution of a high-pressure scenario in a core damage frequency increases, early conditional containment failure probability may become large. In the present study, the verification analysis of PHOENICS code and the combining analysis with MELCOR and PHOENICS codes were performed to examine the debris dispersion behavior during high pressure melt ejection. The PHOENICS code which can treat thermal hydraulic phenomena, was applied to the verification analysis for melt dispersion experiments conducted by the Purdue university in the United States. A low pressure melt dispersion experiment at initial pressure 1.4 MPas used metal woods as a molten material was simulated. The analytical results with molten debris dispersion mostly from the model reactor cavity compartment showed an agreement with the experimental result, but the analysis result of a volumetric median diameter of the airborne debris droplets was estimated about 1.5 times of the experimental result. The injection rates of molten debris and steam after reactor vessel failure for a typical PWR plant were analyzed using the MELCOR code. In addition, PHOENICS was applied to a 3D analysis for debris dispersion with low primary pressure at the reactor vessel failure. The analysis result showed that almost all the molten debris were dispersed from the reactor vessel cavity compartment by about 45 seconds after the

  19. Risk Assessment of Energy-Efficient Walls

    SciTech Connect

    Pallin, Simon B.; Hun, Diana E.; Jackson, Roderick K.; Kehrer, Manfred

    2014-12-01

    This multi-year project aims to provide the residential construction industry with energy-efficient wall designs that are moisture durable. The present work focused on the initial step of this project, which is to develop a moisture durability protocol that identifies energy efficient wall designs that have a low probability of experiencing moisture problems.

  20. Automotion of domain walls for spintronic interconnects

    SciTech Connect

    Nikonov, Dmitri E.; Manipatruni, Sasikanth; Young, Ian A.

    2014-06-07

    We simulate “automotion,” the transport of a magnetic domain wall under the influence of demagnetization and magnetic anisotropy, in nanoscale spintronic interconnects. In contrast to spin transfer driven magnetic domain wall motion, the proposed interconnects operate without longitudinal charge current transfer, with only a transient current pulse at domain wall creation and have favorable scaling down to the 20 nm dimension. Cases of both in-plane and out-of-plane magnetization are considered. Analytical dependence of the velocity of domain walls on the angle of magnetization are compared with full micromagnetic simulations. Deceleration, attenuation and disappearance, and reflection of domain walls are demonstrated through simulation. Dependences of the magnetization angle on the current pulse parameters are studied. The energy and delay analysis suggests that automotion is an attractive option for spintronic logic interconnects.

  1. Wall current probe: A non-invasive in situ plasma diagnostic for space and time resolved current density distribution measurement

    SciTech Connect

    Baude, R.; Gaboriau, F.; Hagelaar, G. J. M. [Universit de Toulouse, UPS, INPT, LAPLACE (Laboratoire Plasma et Conversion dnergie), 118 route de Narbonne, F-31062 Toulouse Cedex 9, France and CNRS, LAPLACE, F-31062, Toulouse (France)] [Universit de Toulouse, UPS, INPT, LAPLACE (Laboratoire Plasma et Conversion dnergie), 118 route de Narbonne, F-31062 Toulouse Cedex 9, France and CNRS, LAPLACE, F-31062, Toulouse (France)

    2013-08-15

    In the context of low temperature plasma research, we propose a wall current probe to determine the local charged particle fluxes flowing to the chamber walls. This non-intrusive planar probe consists of an array of electrode elements which can be individually biased and for which the current can be measured separately. We detail the probe properties and present the ability of the diagnostic to be used as a space and time resolved measurement of the ion and electron current density at the chamber walls. This diagnostic will be relevant to study the electron transport in magnetized low-pressure plasmas.

  2. Factors affecting coking pressures in tall coke ovens

    SciTech Connect

    Grimley, J.J.; Radley, C.E.

    1995-12-01

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

  3. SIMULATION AND MOCKUP OF SNS JET-FLOW TARGET WITH WALL JET FOR CAVITATION DAMAGE MITIGATION

    SciTech Connect

    Wendel, Mark W; Geoghegan, Patrick J; Felde, David K

    2014-01-01

    Pressure waves created in liquid mercury pulsed spallation targets at the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory induce cavitation damage on the stainless steel target container. The cavitation damage is thought to limit the lifetime of the target for power levels at and above 1 MW. Severe through-wall cavitation damage on an internal wall near the beam entrance window has been observed in spent-targets. Surprisingly though, there is very little damage on the walls that bound an annular mercury channel that wraps around the front and outside of the target. The mercury flow through this channel is characterized by smooth, attached streamlines. One theory to explain this lack of damage is that the uni-directional flow biases the direction of the collapsing cavitation bubble, reducing the impact pressure and subsequent damage. The theory has been reinforced by in-beam separate effects data. For this reason, a second-generation SNS mercury target has been designed with an internal wall jet configuration intended to protect the concave wall where damage has been observed. The wall jet mimics the annular flow channel streamlines, but since the jet is bounded on only one side, the momentum is gradually diffused by the bulk flow interactions as it progresses around the cicular path of the target nose. Numerical simulations of the flow through this jet-flow target have been completed, and a water loop has been assembled with a transparent test target in order to visualize and measure the flow field. This paper presents the wall jet simulation results, as well as early experimental data from the test loop.

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

    SciTech Connect

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

    2013-11-01

    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.

  5. High-pressure microhydraulic actuator

    DOEpatents

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

    2008-06-10

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

  6. Effect of temperature and pressure on the dynamics of nanoconfined propane

    SciTech Connect

    Gautam, Siddharth Liu, Tingting Welch, Susan; Cole, David; Rother, Gernot; Jalarvo, Niina; Mamontov, Eugene

    2014-04-24

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

  7. Navigation and vessel inspection circular No. 2-90. Recommended standards for double hulls to be fitted on new tank vessels or retrofitted on existing tank vessels. Final report

    SciTech Connect

    1990-09-21

    The purpose of the Circular is to provide guidance to the marine industry for the construction of new tank vessels, and the retrofitting of existing tank vessels, with double and as required by the Oil Pollution Act of 1990.

  8. Radial force on the vacuum chamber wall during thermal quench in tokamaks

    SciTech Connect

    Pustovitov, V. D.

    2015-12-15

    The radial force balance during a thermal quench in tokamaks is analyzed. As a rule, the duration τ{sub tp} of such events is much shorter than the resistive time τ{sub w} of the vacuum chamber wall. Therefore, the perturbations of the magnetic field B produced by the evolving plasma cannot penetrate the wall, which makes different the magnetic pressures on its inner and outer sides. The goal of this work is the analytical estimation of the resulting integral radial force on the wall. The plasma is considered axially symmetric; for the description of radial forces on the wall, the results of V.D. Shafranov’s classical work [J. Nucl. Energy C 5, 251 (1963)] are used. Developed for tokamaks, the standard equilibrium theory considers three interacting systems: plasma, poloidal field coils, and toroidal field coils. Here, the wall is additionally incorporated with currents driven by ∂B/∂t≠0 accompanying the fast loss of the plasma thermal energy. It is shown that they essentially affect the force redistribution, thereby leading to large loads on the wall. The estimates prove that these loads have to be accounted for in the disruptive scenarios in large tokamaks.

  9. PNL technical review of pressurized thermal-shock issues. [PWR

    SciTech Connect

    Pedersen, L.T.; Apley, W.J.; Bian, S.H.; Defferding, L.J.; Morgenstern, M.H.; Pelto, P.J.; Simonen, E.P.; Simonen, F.A.; Stevens, D.L.; Taylor, T.T.

    1982-07-01

    Pacific Northwest Laboratory (PNL) was asked to develop and recommend a regulatory position that the Nuclear Regulatory Commission (NRC) should adopt regarding the ability of reactor pressure vessels to withstand the effects of pressurized thermal shock (PTS). Licensees of eight pressurized water reactors provided NRC with estimates of remaining effective full power years before corrective actions would be required to prevent an unsafe operating condition. PNL reviewed these responses and the results of supporting research and concluded that none of the eight reactors would undergo vessel failure from a PTS event before several more years of operation. Operator actions, however, were often required to terminate a PTS event before it deteriorated to the point where failure could occur. Therefore, the near-term (less than one year) recommendation is to upgrade, on a site-specific basis, operational procedures, training, and control room instrumentation. Also, uniform criteria should be developed by NRC for use during future licensee analyses. Finally, it was recommended that NRC upgrade nondestructive inspection techniques used during vessel examinations and become more involved in the evaluation of annealing requirements.

  10. Pressurized solid oxide fuel cell integral air accumular containment

    DOEpatents

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

    2004-02-10

    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.

  11. Price of Liquefied U.S. Natural Gas Exports byVessel to Mexico (Dollars per

    Energy Information Administration (EIA) (indexed site)

    Thousand Cubic Feet) byVessel to Mexico (Dollars per

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

    DOEpatents

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

    2005-03-29

    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.

  13. ITER's Tokamak Cooling Water System and the the Use of ASME Codes to Comply with French Regulations of Nuclear Pressure Equipment

    SciTech Connect

    Berry, Jan; Ferrada, Juan J; Curd, Warren; Dell Orco, Dr. Giovanni; Barabash, Vladimir; Kim, Seokho H

    2011-01-01

    During inductive plasma operation of ITER, fusion power will reach 500 MW with an energy multiplication factor of 10. The heat will be transferred by the Tokamak Cooling Water System (TCWS) to the environment using the secondary cooling system. Plasma operations are inherently safe even under the most severe postulated accident condition a large, in-vessel break that results in a loss-of-coolant accident. A functioning cooling water system is not required to ensure safe shutdown. Even though ITER is inherently safe, TCWS equipment (e.g., heat exchangers, piping, pressurizers) are classified as safety important components. This is because the water is predicted to contain low-levels of radionuclides (e.g., activated corrosion products, tritium) with activity levels high enough to require the design of components to be in accordance with French regulations for nuclear pressure equipment, i.e., the French Order dated 12 December 2005 (ESPN). ESPN has extended the practical application of the methodology established by the Pressure Equipment Directive (97/23/EC) to nuclear pressure equipment, under French Decree 99-1046 dated 13 December 1999, and Order dated 21 December 1999 (ESP). ASME codes and supplementary analyses (e.g., Failure Modes and Effects Analysis) will be used to demonstrate that the TCWS equipment meets these essential safety requirements. TCWS is being designed to provide not only cooling, with a capacity of approximately 1 GW energy removal, but also elevated temperature baking of first-wall/blanket, vacuum vessel, and divertor. Additional TCWS functions include chemical control of water, draining and drying for maintenance, and facilitation of leak detection/localization. The TCWS interfaces with the majority of ITER systems, including the secondary cooling system. U.S. ITER is responsible for design, engineering, and procurement of the TCWS with industry support from an Engineering Services Organization (ESO) (AREVA Federal Services, with support

  14. High Performance Walls in Hot-Dry Climates (Technical Report...

    Office of Scientific and Technical Information (OSTI)

    High Performance Walls in Hot-Dry Climates Citation Details In-Document Search Title: High Performance Walls in Hot-Dry Climates High performance walls represent a high priority...

  15. Pressure-sensitive optrode

    DOEpatents

    Hirschfeld, T.B.

    1986-07-15

    An apparatus is provided for sensing changes in pressure and for generating optical signals related to said changes in pressure. Light from a fiber optic illuminates a fluorescent composition causing it to fluoresce. The fluorescent composition is caused to fluoresce more relative to the end of the fiber optic in response to changes in pressure so that the intensity of fluorescent emissions collected by the same fiber optic used for illumination varies monotonically with pressure. 10 figs.

  16. PRESSURE SYSTEM CONTROL

    DOEpatents

    Esselman, W.H.; Kaplan, G.M.

    1961-06-20

    The control of pressure in pressurized liquid systems, especially a pressurized liquid reactor system, may be achieved by providing a bias circuit or loop across a closed loop having a flow restriction means in the form of an orifice, a storage tank, and a pump connected in series. The subject invention is advantageously utilized where control of a reactor can be achieved by response to the temperature and pressure of the primary cooling system.

  17. In-Vessel Retention - Recent Efforts and Future Needs

    SciTech Connect

    J. L. Rempe

    2004-10-01

    In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. However, it is not clear that the external reactor vessel cooling (ERVC) proposed for existing and some advanced reactors would provide sufficient heat removal for higher-power reactors (up to 1400 MWe) without additional enhancements. This paper summarizes recent efforts to enhance IVR and identifies additional needs to demonstrate that there is sufficient margin for successful IVR in high power reactors.

  18. Measurement of earth pressures on concrete box culverts under highway embankments

    SciTech Connect

    Yang, M.Z.; Drumm, E.C.; Bennett, R.M.; Mauldon, M.

    1999-07-01

    To obtain a better understanding of the stresses acting on cast-in-place concrete box culverts, and to investigate the conditions which resulted in a culvert failure under about 12 meters of backfill, two sections of a new culvert were instrumented. The measured earth pressure distribution was found to depend upon the height of the embankment over the culvert. For low embankment heights (less than one-half the culvert width), the average measured vertical earth pressures, weighted by tributary length, were about 30% greater than the recommended AASHTO pressures. The measured lateral pressures were slightly greater than the AASHTO pressures. As the embankment height increased, the measured weighted average vertical stress exceeded the AASHTO pressures by about 20%. Lateral pressures which exceeded the vertical pressures were recorded at the bottom of the culvert walls, and small lateral pressures were recorded on the upper locations of the wall. The high lateral pressures at the base of the wall are consistent with the results from finite element analyses with high density (modulus) backfill material placed around the culvert.

  19. Modeling and Analysis of Alternative Concept of ITER Vacuum Vessel Primary Heat Transfer System

    SciTech Connect

    Carbajo, Juan J; Yoder Jr, Graydon L; Dell'Orco, Giovanni; Curd, Warren; Kim, Seokho H

    2010-01-01

    A RELAP5-3D model of the ITER (Latin for the way ) vacuum vessel (VV) primary heat transfer system has been developed to evaluate a proposed design change that relocates the heat exchangers (HXs) from the exterior of the tokamak building to the interior. This alternative design protects the HXs from external hazards such as wind, tornado, and aircraft crash. The proposed design integrates the VV HXs into a VV pressure suppression system (VVPSS) tank that contains water to condense vapour in case of a leak into the plasma chamber. The proposal is to also use this water as the ultimate sink when removing decay heat from the VV system. The RELAP5-3D model has been run under normal operating and abnormal (decay heat) conditions. Results indicate that this alternative design is feasible, with no effects on the VVPSS tank under normal operation and with tank temperature and pressure increasing under decay heat conditions resulting in a requirement to remove steam generated if the VVPSS tank low pressure must be maintained.

  20. Control of linear modes in cylindrical resistive magnetohydrodynamics with a resistive wall, plasma rotation, and complex gain

    SciTech Connect

    Brennan, D. P.; Finn, J. M.

    2014-10-15

    Feedback stabilization of magnetohydrodynamic (MHD) modes in a tokamak is studied in a cylindrical model with a resistive wall, plasma resistivity, viscosity, and toroidal rotation. The control is based on a linear combination of the normal and tangential components of the magnetic field just inside the resistive wall. The feedback includes complex gain, for both the normal and for the tangential components, and it is known that the imaginary part of the feedback for the former is equivalent to plasma rotation [J. M. Finn and L. Chacon, Phys. Plasmas 11, 1866 (2004)]. The work includes (1) analysis with a reduced resistive MHD model for a tokamak with finite ? and with stepfunction current density and pressure profiles, and (2) computations with a full compressible visco-resistive MHD model with smooth decreasing profiles of current density and pressure. The equilibria are stable for ??=?0 and the marginal stability values ?{sub rp,rw}?wall; resistive plasma, ideal wall; ideal plasma, resistive wall; and ideal plasma, ideal wall) are computed for both models. The main results are: (a) imaginary gain with normal sensors or plasma rotation stabilizes below ?{sub rp,iw} because rotation suppresses the diffusion of flux from the plasma out through the wall and, more surprisingly, (b) rotation or imaginary gain with normal sensors destabilizes above ?{sub rp,iw} because it prevents the feedback flux from entering the plasma through the resistive wall to form a virtual wall. A method of using complex gain G{sub i} to optimize in the presence of rotation in this regime with ??>??{sub rp,iw} is presented. The effect of imaginary gain with tangential sensors is more complicated but essentially destabilizes above and below ?{sub rp,iw}.