Sample records for fuel assembly components

  1. Compliant fuel cell system

    DOEpatents

    Bourgeois, Richard Scott [Albany, NY; Gudlavalleti, Sauri [Albany, NY

    2009-12-15

    A fuel cell assembly comprising at least one metallic component, at least one ceramic component and a structure disposed between the metallic component and the ceramic component. The structure is configured to have a lower stiffness compared to at least one of the metallic component and the ceramic component, to accommodate a difference in strain between the metallic component and the ceramic component of the fuel cell assembly.

  2. Improved nuclear fuel assembly grid spacer

    DOEpatents

    Marshall, John; Kaplan, Samuel

    1977-01-01

    An improved fuel assembly grid spacer and method of retaining the basic fuel rod support elements in position within the fuel assembly containment channel. The improvement involves attachment of the grids to the hexagonal channel and of forming the basic fuel rod support element into a grid structure, which provides a design which is insensitive to potential channel distortion (ballooning) at high fluence levels. In addition the improved method eliminates problems associated with component fabrication and assembly.

  3. Between-cycle laser system for depressurization and resealing of modified design nuclear fuel assemblies

    DOEpatents

    Bradley, John G.

    1982-01-01

    A laser beam is used to puncture fuel cladding for release of contained pressurized fission gas from plenum sections or irradiated fuel pins. Exhausted fission gases are collected and trapped for safe disposal. The laser beam, adjusted to welding mode, is subsequently used to reseal the puncture holes. The fuel assembly is returned to additional irradiation or, if at end of reactivity lifetime, is routed to reprocess. The fuel assembly design provides graded cladding lengths, by rows or arrays, such that the cladding of each component fuel element of the assembly is accessible to laser beam reception.

  4. Coolant mass flow equalizer for nuclear fuel

    DOEpatents

    Betten, Paul R.

    1978-01-01

    The coolant mass flow distribution in a liquid metal cooled reactor is enhanced by restricting flow in sub-channels defined in part by the peripheral fuel elements of a fuel assembly. This flow restriction, which results in more coolant flow in interior sub-channels, is achieved through the use of a corrugated liner positioned between the bundle of fuel elements and the inner wall of the fuel assembly coolant duct. The corrugated liner is expandable to accommodate irradiation induced growth of fuel assembly components.

  5. Precise calculation of neutron-capture reactions contribution in energy release for different types of VVER-1000 fuel assemblies

    NASA Astrophysics Data System (ADS)

    Tikhomirov, Georgy; Bahdanovich, Rynat; Pham, Phu

    2017-09-01

    Precise calculation of energy release in a nuclear reactor is necessary to obtain the correct spatial power distribution and predict characteristics of burned nuclear fuel. In this work, previously developed method for calculation neutron-capture reactions - capture component - contribution in effective energy release in a fuel core of nuclear reactor is discussed. The method was improved and implemented to the different models of VVER-1000 reactor developed for MCU 5 and MCNP 4 computer codes. Different models of equivalent cell and fuel assembly in the beginning of fuel cycle were calculated. These models differ by the geometry, fuel enrichment and presence of burnable absorbers. It is shown, that capture component depends on fuel enrichment and presence of burnable absorbers. Its value varies for different types of hot fuel assemblies from 3.35% to 3.85% of effective energy release. Average capture component contribution in effective energy release for typical serial fresh fuel of VVER-1000 is 3.5%, which is 7 MeV/fission. The method will be used in future to estimate the dependency of capture energy on fuel density, burn-up, etc.

  6. Advanced techniques for repair of irradiated PWR fuel assemblies

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Knaab, H.; Westphal, M.

    Kraftwerk union has recently designed and built a portable repair unit for use in nuclear power plants for repair of defective fuel assemblies where space limitations do not allow permanent installation of repair equipment. This new equipment is designed to be easily disassembled and decontaminated. The main component of the equipment is the fuel assembly reconstitution unit (FARU) which is placed on the floor of the spent fuel pool. The use of the FARU is described in the paper.

  7. Fuel-air munition and device

    DOEpatents

    Carlson, Gary A.

    1976-01-01

    An aerially delivered fuel-air munition consisting of an impermeable tank filled with a pressurized liquid fuel and joined at its two opposite ends with a nose section and a tail assembly respectively to complete an aerodynamic shape. On impact the tank is explosively ruptured to permit dispersal of the fuel in the form of a fuel-air cloud which is detonated after a preselected time delay by means of high explosive initiators ejected from the tail assembly. The primary component in the fuel is methylacetylene, propadiene, or mixtures thereof to which is added a small mole fraction of a relatively high vapor pressure liquid diluent or a dissolved gas diluent having a low solubility in the primary component.

  8. Burnable absorber arrangement for fuel bundle

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Crowther, R.L.; Townsend, D.B.

    1986-12-16

    This patent describes a boiling water reactor core whose operation is characterized by a substantial proportion of steam voids with concomitantly reduced moderation toward the top of the core when the reactor is in its hot operating condition. The reduced moderation leads to slower burnup and greater conversion ratio in an upper core region so that when the reactor is in its cold shut down condition the resulting relatively increased moderation in the upper core region is accompanied by a reactivity profile that peaks in the upper core region. A fuel assembly is described comprising; a component of fissile materialmore » distributed over a substantial axial extent of the fuel assembly; and a component of neutron absorbing material having an axial distribution characterized by an enhancement in an axial zone of the fuel assembly, designated the cold shutdown control zone, corresponding to at least a portion of the axial region of the core when the cold shutdown reactivity peaks. The aggregate amount of neutron absorbing material in the cold shutdown zone of the fuel assembly is greater than the aggregate amount of neutron absorbing material in the axial zones of the fuel assembly immediately above and immediately below the cold shutdown control zone whereby the cold shutdown reactivity peak is reduced relative to the cold shutdown reactivity in the zones immediately above and immediately below the cold shutdown control zone. The cold shutdown zone has an axial extent measured from the bottom of the fuel assembly in the range between 68-88 percent of the height of the fissile material in the fuel assembly.« less

  9. Electrolyte creepage barrier for liquid electrolyte fuel cells

    DOEpatents

    Li, Jian [Alberta, CA; Farooque, Mohammad [Danbury, CT; Yuh, Chao-Yi [New Milford, CT

    2008-01-22

    A dielectric assembly for electrically insulating a manifold or other component from a liquid electrolyte fuel cell stack wherein the dielectric assembly includes a substantially impermeable dielectric member over which electrolyte is able to flow and a barrier adjacent the dielectric member and having a porosity of less than 50% and greater than 10% so that the barrier is able to measurably absorb and chemically react with the liquid electrolyte flowing on the dielectric member to form solid products which are stable in the liquid electrolyte. In this way, the barrier inhibits flow or creepage of electrolyte from the dielectric member to the manifold or component to be electrically insulated from the fuel cell stack by the dielectric assembly.

  10. Upgraded HFIR Fuel Element Welding System

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Sease, John D

    2010-02-01

    The welding of aluminum-clad fuel plates into aluminum alloy 6061 side plate tubing is a unique design feature of the High Flux Isotope Reactor (HFIR) fuel assemblies as 101 full-penetration circumferential gas metal arc welds (GMAW) are required in the fabrication of each assembly. In a HFIR fuel assembly, 540 aluminum-clad fuel plates are assembled into two nested annular fuel elements 610 mm (24-inches) long. The welding process for the HFIR fuel elements was developed in the early 1960 s and about 450 HFIR fuel assemblies have been successfully welded using the GMAW process qualified in the 1960 s. Inmore » recent years because of the degradation of the electronic and mechanical components in the old HFIR welding system, reportable defects in plate attachment or adapter welds have been present in almost all completed fuel assemblies. In October 2008, a contract was awarded to AMET, Inc., of Rexburg, Idaho, to replace the old welding equipment with standard commercially available welding components to the maximum extent possible while maintaining the qualified HFIR welding process. The upgraded HFIR welding system represents a major improvement in the welding system used in welding HFIR fuel elements for the previous 40 years. In this upgrade, the new inner GMAW torch is a significant advancement over the original inner GMAW torch previously used. The innovative breakthrough in the new inner welding torch design is the way the direction of the cast in the 0.762 mm (0.030-inch) diameter aluminum weld wire is changed so that the weld wire emerging from the contact tip is straight in the plane perpendicular to the welding direction without creating any significant drag resistance in the feeding of the weld wire.« less

  11. Direct methanol feed fuel cell with reduced catalyst loading

    NASA Technical Reports Server (NTRS)

    Kindler, Andrew (Inventor)

    1999-01-01

    Improvements to direct feed methanol fuel cells include new protocols for component formation. Catalyst-water repellent material is applied in formation of electrodes and sintered before application of ionomer. A membrane used in formation of an electrode assembly is specially pre-treated to improve bonding between catalyst and membrane. The improved electrode and the pre-treated membrane are assembled into a membrane electrode assembly.

  12. FUEL ASSEMBLY SHAKER TEST SIMULATION

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Klymyshyn, Nicholas A.; Sanborn, Scott E.; Adkins, Harold E.

    This report describes the modeling of a PWR fuel assembly under dynamic shock loading in support of the Sandia National Laboratories (SNL) shaker test campaign. The focus of the test campaign is on evaluating the response of used fuel to shock and vibration loads that a can occur during highway transport. Modeling began in 2012 using an LS-DYNA fuel assembly model that was first created for modeling impact scenarios. SNL’s proposed test scenario was simulated through analysis and the calculated results helped guide the instrumentation and other aspects of the testing. During FY 2013, the fuel assembly model was refinedmore » to better represent the test surrogate. Analysis of the proposed loads suggested the frequency band needed to be lowered to attempt to excite the lower natural frequencies of the fuel assembly. Despite SNL’s expansion of lower frequency components in their five shock realizations, pretest predictions suggested a very mild dynamic response to the test loading. After testing was completed, one specific shock case was modeled, using recorded accelerometer data to excite the model. Direct comparison of predicted strain in the cladding was made to the recorded strain gauge data. The magnitude of both sets of strain (calculated and recorded) are very low, compared to the expected yield strength of the Zircaloy-4 material. The model was accurate enough to predict that no yielding of the cladding was expected, but its precision at predicting micro strains is questionable. The SNL test data offers some opportunity for validation of the finite element model, but the specific loading conditions of the testing only excite the fuel assembly to respond in a limited manner. For example, the test accelerations were not strong enough to substantially drive the fuel assembly out of contact with the basket. Under this test scenario, the fuel assembly model does a reasonable job of approximating actual fuel assembly response, a claim that can be verified through direct comparison of model results to recorded test results. This does not offer validation for the fuel assembly model in all conceivable cases, such as high kinetic energy shock cases where the fuel assembly might lift off the basket floor to strike to basket ceiling. This type of nonlinear behavior was not witnessed in testing, so the model does not have test data to be validated against.a basis for validation in cases that substantially alter the fuel assembly response range. This leads to a gap in knowledge that is identified through this modeling study. The SNL shaker testing loaded a surrogate fuel assembly with a certain set of artificially-generated time histories. One thing all the shock cases had in common was an elimination of low frequency components, which reduces the rigid body dynamic response of the system. It is not known if the SNL test cases effectively bound all highway transportation scenarios, or if significantly greater rigid body motion than was tested is credible. This knowledge gap could be filled through modeling the vehicle dynamics of a used fuel conveyance, or by collecting acceleration time history data from an actual conveyance under highway conditions.« less

  13. Saturn Apollo Program

    NASA Image and Video Library

    1964-09-01

    This photograph shows the components for the Saturn V S-IC stage fuel tank assembly in the Manufacturing Engineering Laboratory, building 4707, at the Marshall Space Flight Center (MSFC). Left to right are upper head, lower head, and forward skirt assembly. Thirty-three feet in diameter, they will hold a total of 4,400,000 pounds of fuel. Although this tankage was assembled at MSFC, the elements were made by the Boeing Company at Wichita and the Michould Operations at New Orleans.

  14. Fast Neutron Emission Tomography of Used Nuclear Fuel Assemblies

    NASA Astrophysics Data System (ADS)

    Hausladen, Paul; Iyengar, Anagha; Fabris, Lorenzo; Yang, Jinan; Hu, Jianwei; Blackston, Matthew

    2017-09-01

    Oak Ridge National Laboratory is developing a new capability to perform passive fast neutron emission tomography of spent nuclear fuel assemblies for the purpose of verifying their integrity for international safeguards applications. Most of the world's plutonium is contained in spent nuclear fuel, so it is desirable to detect the diversion of irradiated fuel rods from an assembly prior to its transfer to ``difficult to access'' storage, such as a dry cask or permanent repository, where re-verification is practically impossible. Nuclear fuel assemblies typically consist of an array of fuel rods that, depending on exposure in the reactor and consequent ingrowth of 244Cm, are spontaneous sources of as many as 109 neutrons s-1. Neutron emission tomography uses collimation to isolate neutron activity along ``lines of response'' through the assembly and, by combining many collimated views through the object, mathematically extracts the neutron emission from each fuel rod. This technique, by combining the use of fast neutrons -which can penetrate the entire fuel assembly -and computed tomography, is capable of detecting vacancies or substitutions of individual fuel rods. This paper will report on the physics design and component testing of the imaging system. This material is based upon work supported by the U.S. Department of Energy, Office of Defense Nuclear Nonproliferation Research and Development within the National Nuclear Security Administration, under Contract Number DE-AC05-00OR22725.

  15. Process for recycling components of a PEM fuel cell membrane electrode assembly

    DOEpatents

    Shore, Lawrence [Edison, NJ

    2012-02-28

    The membrane electrode assembly (MEA) of a PEM fuel cell can be recycled by contacting the MEA with a lower alkyl alcohol solvent which separates the membrane from the anode and cathode layers of the assembly. The resulting solution containing both the polymer membrane and supported noble metal catalysts can be heated under mild conditions to disperse the polymer membrane as particles and the supported noble metal catalysts and polymer membrane particles separated by known filtration means.

  16. NNSA B-Roll: MOX Facility

    ScienceCinema

    None

    2017-12-09

    In 1999, the National Nuclear Security Administration (NNSA) signed a contract with a consortium, now called Shaw AREVA MOX Services, LLC to design, build, and operate a Mixed Oxide (MOX) Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium. The facility will take surplus weapon-grade plutonium, remove impurities, and mix it with uranium oxide to form MOX fuel pellets for reactor fuel assemblies. These assemblies will be irradiated in commercial nuclear power reactors.

  17. NNSA B-Roll: MOX Facility

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    None

    2010-05-21

    In 1999, the National Nuclear Security Administration (NNSA) signed a contract with a consortium, now called Shaw AREVA MOX Services, LLC to design, build, and operate a Mixed Oxide (MOX) Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium. The facility will take surplus weapon-grade plutonium, remove impurities, and mix it with uranium oxide to form MOX fuel pellets for reactor fuel assemblies. These assemblies will be irradiated in commercial nuclear power reactors.

  18. All About MOX

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    None

    2009-07-29

    In 1999, the Nuclear Nuclear Security Administration (NNSA) signed a contract with a consortium, now called Shaw AREVA MOX Services, LLC to design, build, and operate a Mixed Oxide (MOX) Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium. The facility will take surplus weapon-grade plutonium, remove impurities, and mix it with uranium oxide to form MOX fuel pellets for reactor fuel assemblies. These assemblies will be irradiated in commercial nuclear power reactors.

  19. All About MOX

    ScienceCinema

    None

    2018-01-16

    In 1999, the Nuclear Nuclear Security Administration (NNSA) signed a contract with a consortium, now called Shaw AREVA MOX Services, LLC to design, build, and operate a Mixed Oxide (MOX) Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium. The facility will take surplus weapon-grade plutonium, remove impurities, and mix it with uranium oxide to form MOX fuel pellets for reactor fuel assemblies. These assemblies will be irradiated in commercial nuclear power reactors.

  20. Co-flow anode/cathode supply heat exchanger for a solid-oxide fuel cell assembly

    DOEpatents

    Haltiner, Jr., Karl J.; Kelly, Sean M.

    2005-11-22

    In a solid-oxide fuel cell assembly, a co-flow heat exchanger is provided in the flow paths of the reformate gas and the cathode air ahead of the fuel cell stack, the reformate gas being on one side of the exchanger and the cathode air being on the other. The reformate gas is at a substantially higher temperature than is desired in the stack, and the cathode gas is substantially cooler than desired. In the co-flow heat exchanger, the temperatures of the reformate and cathode streams converge to nearly the same temperature at the outlet of the exchanger. Preferably, the heat exchanger is formed within an integrated component manifold (ICM) for a solid-oxide fuel cell assembly.

  1. Inlet nozzle assembly

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.; Precechtel, Donald R.; Smith, Bob G.; Knight, Ronald C.

    1987-01-01

    An inlet nozzle assembly for directing coolant into the duct tube of a fuel assembly attached thereto. The nozzle assembly includes a shell for housing separable components including an orifice plate assembly, a neutron shield block, a neutron shield plug, and a diffuser block. The orifice plate assembly includes a plurality of stacked plates of differently configurated and sized openings for directing coolant therethrough in a predesigned flow pattern.

  2. Inlet nozzle assembly

    DOEpatents

    Christiansen, D.W.; Karnesky, R.A.; Knight, R.C.; Precechtel, D.R.; Smith, B.G.

    1985-09-09

    An inlet nozzle assembly for directing coolant into the duct tube of a fuel assembly attached thereto. The nozzle assembly includes a shell for housing separable components including an orifice plate assembly, a neutron shield block, a neutron shield plug, and a diffuser block. The orifice plate assembly includes a plurality of stacked plates of differently configurated and sized openings for directing coolant therethrough in a predesigned flow pattern.

  3. Method of making MEA for PEM/SPE fuel cell

    DOEpatents

    Hulett, Jay S.

    2000-01-01

    A method of making a membrane-electrode-assembly (MEA) for a PEM/SPE fuel cell comprising applying a slurry of electrode-forming material directly onto a membrane-electrolyte film. The slurry comprises a liquid vehicle carrying catalyst particles and a binder for the catalyst particles. The membrane-electrolyte is preswollen by contact with the vehicle before the electrode-forming slurry is applied to the membrane-electrolyte. The swollen membrane-electrolyte is constrained against shrinking in the "x" and "y" directions during drying. Following assembly of the fuel cell, the MEA is rehydrated inside the fuel cell such that it swells in the "z" direction for enhanced electrical contact with contiguous electrically conductive components of the fuel cell.

  4. DOE Office of Scientific and Technical Information (OSTI.GOV)

    S.O. Bader

    The purpose of this design analysis is to specify and document the total and respirable fractions for radioactive materials that are released from an accident event at the Monitored Geologic Repository (MGR) involving commercial spent nuclear fuel (CSNF) in a dry environment. The total and respirable release fractions will be used to support the preclosure licensing basis for the MGR. The total release fraction is defined as the fraction of total CSNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. The radionuclides are releasedmore » from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses. This subset of the total release fraction is referred to as the respirable release fraction. Potential accidents may involve waste forms that are characterized as either bare (unconfined) fuel assemblies or confined fuel assemblies. The confined CSNF assemblies at the MGR are contained in shipping casks, canisters, or disposal containers (waste packages). In contrast to the bare fuel assemblies, the container that confines the fuel assemblies has the potential of providing an additional barrier for diminishing the total release fraction should the fuel rod cladding breach during an accident. However, this analysis will not take credit for this additional bamer and will establish only the total release fractions for bare unconfined CSNF assemblies, which may however be conservatively applied to confined CSNF assemblies.« less

  5. Development of an alkaline fuel cell subsystem

    NASA Technical Reports Server (NTRS)

    1987-01-01

    A two task program was initiated to develop advanced fuel cell components which could be assembled into an alkaline power section for the Space Station Prototype (SSP) fuel cell subsystem. The first task was to establish a preliminary SSP power section design to be representative of the 200 cell Space Station power section. The second task was to conduct tooling and fabrication trials and fabrication of selected cell stack components. A lightweight, reliable cell stack design suitable for the SSP regenerative fuel cell power plant was completed. The design meets NASA's preliminary requirements for future multikilowatt Space Station missions. Cell stack component fabrication and tooling trials demonstrated cell components of the SSP stack design of the 1.0 sq ft area can be manufactured using techniques and methods previously evaluated and developed.

  6. DIANA: A multi-phase, multi-component hydrodynamic model for the analysis of severe accidents in heavy water reactors with multiple-tube assemblies

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Tentner, A.M.

    1994-03-01

    A detailed hydrodynamic fuel relocation model has been developed for the analysis of severe accidents in Heavy Water Reactors with multiple-tube Assemblies. This model describes the Fuel Disruption and Relocation inside a nuclear fuel assembly and is designated by the acronym DIANA. DIANA solves the transient hydrodynamic equations for all the moving materials in the core and treats all the relevant flow regimes. The numerical solution techniques and some of the physical models included in DIANA have been developed taking advantage of the extensive experience accumulated in the development and validation of the LEVITATE (1) fuel relocation model of SAS4Amore » [2, 3]. The model is designed to handle the fuel and cladding relocation in both voided and partially voided channels. It is able to treat a wide range of thermal/ hydraulic/neutronic conditions and the presence of various flow regimes at different axial locations within the same hydrodynamic channel.« less

  7. 14 CFR Appendix A to Part 31 - Instructions for Continued Airworthiness

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... that covers details regarding servicing of balloon components, including burner nozzles, fuel tanks... envelope, controls, rigging, basket structure, fuel systems, instruments, and heater assembly that provides... preparation including any storage limits. (i) Instructions for repair on the balloon envelope and its basket...

  8. 14 CFR Appendix A to Part 31 - Instructions for Continued Airworthiness

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... that covers details regarding servicing of balloon components, including burner nozzles, fuel tanks... envelope, controls, rigging, basket structure, fuel systems, instruments, and heater assembly that provides... preparation including any storage limits. (i) Instructions for repair on the balloon envelope and its basket...

  9. 14 CFR Appendix A to Part 31 - Instructions for Continued Airworthiness

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... that covers details regarding servicing of balloon components, including burner nozzles, fuel tanks... envelope, controls, rigging, basket structure, fuel systems, instruments, and heater assembly that provides... preparation including any storage limits. (i) Instructions for repair on the balloon envelope and its basket...

  10. 14 CFR Appendix A to Part 31 - Instructions for Continued Airworthiness

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... that covers details regarding servicing of balloon components, including burner nozzles, fuel tanks... envelope, controls, rigging, basket structure, fuel systems, instruments, and heater assembly that provides... preparation including any storage limits. (i) Instructions for repair on the balloon envelope and its basket...

  11. Posttest examination of Sodium Loop Safety Facility experiments. [LMFBR

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Holland, J.W.

    In-reactor, safety experiments performed in the Sodium Loop Safety Facility (SLSF) rely on comprehensive posttest examinations (PTE) to characterize the postirradiation condition of the cladding, fuel, and other test-subassembly components. PTE information and on-line instrumentation data, are analyzed to identify the sequence of events and the severity of the accident for each experiment. Following in-reactor experimentation, the SLSF loop and test assembly are transported to the Hot Fuel Examination Facility (HFEF) for initial disassembly. Goals of the HFEF-phase of the PTE are to retrieve the fuel bundle by dismantling the loop and withdrawing the test assembly, to assess the macro-conditionmore » of the fuel bundle by nondestructive examination techniques, and to prepare the fuel bundle for shipment to the Alpha-Gamma Hot Cell Facility (AGHCF) at Argonne National Laboratory.« less

  12. Cassettes for solid-oxide fuel cell stacks and methods of making the same

    DOEpatents

    Weil, K. Scott; Meinhardt, Kerry D; Sprenkle, Vincent L

    2012-10-23

    Solid-oxide fuel cell (SOFC) stack assembly designs are consistently investigated to develop an assembly that provides optimal performance, and durability, within desired cost parameters. A new design includes a repeat unit having a SOFC cassette and being characterized by a three-component construct. The three components include an oxidation-resistant, metal window frame hermetically joined to an electrolyte layer of a multi-layer, anode-supported ceramic cell and a pre-cassette including a separator plate having a plurality of vias that provide electrical contact between an anode-side collector within the pre-cassette and a cathode-side current collector of an adjacent cell. The third component is a cathode-side seal, which includes a standoff that supports a cathode channel spacing between each of the cassettes in a stack. Cassettes are formed by joining the pre-cassette and the window frame.

  13. Fuel Processing System for a 5kW Methanol Fuel Cell Power Unit.

    DTIC Science & Technology

    1985-11-27

    report documents the development and design of a 5kW neat methanol reformer for phosphoric acid fuel cell power plants . The reformer design was based...VAPORIZATION OF METHANOL ........... 4.3 REFORMING/SHIFT CATALYST BED ......... 2 5.0 COMPONENT TESTING............... 5.1 COMBUSTION TUBE...69 36 Catalyst Bed Temperature Profile Before and After Transient ................. 70 37 Assembly -5kw Neat Methanol Reformer. ......... 72 Page No

  14. ACEE composite structures technology

    NASA Technical Reports Server (NTRS)

    James, A. M.

    1984-01-01

    Topics addressed include: strength and hygrothermal response of L-1011 fin components; wing fuel containment and damage tolerance development; impact dynamics; acoustic transmission; fuselage structure; composite transport wing technology development; spar/assembly concepts.

  15. Saturn Apollo Program

    NASA Image and Video Library

    1964-10-01

    The components of the Saturn V booster (S-IC stage) fuel tank are shown in this photograph. The liquid oxygen tank bulkhead on the left and both halves of the fuel tank were in the Marshall Space Flight Center (MSFC) Manufacturing Engineering Laboratory, building 4707. These components were used at MSFC in structural testing to prove that they could withstand the forces to which they were subjected in flight. Each S-IC stage has two tanks, one for kerosene and one for liquid oxygen, made from such components as these. Thirty-three feet in diameter, they hold a total of 4,400,000 pounds of fuel. Although this tankage was assembled at MSFC, the elements were made by the Boeing Company at Wichita and the Michoud Operations at New Orleans.

  16. Aerolization During Boron Nanoparticle Multi-Component Fuel Group Burning Studies

    DTIC Science & Technology

    2014-02-03

    Anderson, University of Utah). …………………… 14 Figure 2. Photograph of group burning facility showing benchtop flat flame burner unit with injector nozzle ...and (B) aerosol generator. 16 Figure 6. Diagram of benchtop flat flame burner unit showing injector nozzle assembly with VOAG orifice, fuel and...translation stage, variable fuel and gas supply rates, and injector nozzles that can be configured to investigate diffusion and premixed flames (Fig. 2 & 3

  17. Cerium migration during PEM fuel cell assembly and operation

    DOE PAGES

    Baker, Andrew M.; Torraco, Dennis; Judge, Elizabeth J.; ...

    2015-09-14

    Cerium migration between PEM fuel cell components is influenced by potential-driven mobility, ionic diffusion, and gradients in water content. These factors were investigated in ex situ experiments and in operating fuel cells. Potential-induced migration was measured ex situ in hydrated window cells. Cerium-containing MEAs were also fabricated and tested under ASTs. MEA disassembly and subsequent XRF analysis were used to observe rapid cerium migration during cell assembly and operation. During MEA hot pressing, humidification, and low RH operation at OCV, ionic diffusion causes uniform migration from the membrane into the catalyst layers. During high RH operation at OCV, in-plane ceriummore » gradients arise due to variations in water content. These gradients may diminish the scavenging efficacy of cerium by reducing its proximity to generated radicals.« less

  18. Morphological features (defects) in fuel cell membrane electrode assemblies

    NASA Astrophysics Data System (ADS)

    Kundu, S.; Fowler, M. W.; Simon, L. C.; Grot, S.

    Reliability and durability issues in fuel cells are becoming more important as the technology and the industry matures. Although research in this area has increased, systematic failure analysis, such as a failure modes and effects analysis (FMEA), are very limited in the literature. This paper presents a categorization scheme of causes, modes, and effects related to fuel cell degradation and failure, with particular focus on the role of component quality, that can be used in FMEAs for polymer electrolyte membrane (PEM) fuel cells. The work also identifies component defects imparted on catalyst-coated membranes (CCM) by manufacturing and proposes mechanisms by which they can influence overall degradation and reliability. Six major defects have been identified on fresh CCM materials, i.e., cracks, orientation, delamination, electrolyte clusters, platinum clusters, and thickness variations.

  19. Fuel cell system blower configuration

    DOEpatents

    Patel, Kirtikumar H.; Saito, Kazuo

    2017-11-28

    An exemplary fuel cell system includes a cell stack assembly having a plurality of cathode components and a plurality of anode components. A first reactant blower has an outlet situated to provide a first reactant to the cathode components. A second reactant blower has an outlet situated to provide a second reactant to the anode components. The second reactant blower includes a fan portion that moves the second reactant through the outlet. The second reactant blower also includes a motor portion that drives the fan portion and a bearing portion associated with the fan portion and the motor portion. The motor portion has a motor coolant inlet coupled with the outlet of the first reactant blower to receive some of the first reactant for cooling the motor portion.

  20. Manufacturing of Low Cost, Durable Membrane Electrode Assemblies Engineered for Rapid Conditioning

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Busby, Colin

    Over the past 20 years significant progress in membrane-electrode assembly (MEA) technology development for polymer electrolyte fuel cells (PEMFCs) has resulted in the PEMFC technology approaching a commercial reality for transportation applications. However, there remain two primary technical challenges to be addressed in the MEA. First and foremost is meeting the automotive cost targets: Producing a fuel cell stack cost competitive with today’s internal combustion engine. In addition to the material cost, MEA (and other components) and stack assembly production methods must be amenable for use in low cost, high speed, automotive assembly line. One impediment to this latter goalmore » is that stack components must currently go through a long and tedious conditioning procedure before they produce optimal power. This so-called “break-in” can take many hours, and can involve quite complex voltage, temperature and/or pressure steps. These break-in procedures must be simplified and the time required reduced if fuel cells are to become a viable automotive engine. The second challenge is to achieve the durability targets in real-world automotive duty cycle operations. Significant improvements in cost, break-in time, and durability for the key component of fuel cell stacks, MEAs were achieved in this project. Advanced modeling was used to guide design of the new MEA to maximize performance and durability. A new, innovative process and manufacturing approach utilizing direct in-line coating using scalable, cost-competitive, continuous high volume 3-layer rolled-good manufacturing processes was developed and validated by single cell and short stack testing. In addition, the direct coating methods employed were shown to reduce the cost for sacrificial films. Furthermore, Gore has demonstrated a 10 µm reinforced membrane that is used in the new low-cost process and can meet automotive power density and durability targets. Across a wide range of operating conditions, the direct-coated MEA outperformed the commercial baseline MEA, and did so through a process that delivers MEAs at $92.35/m2 at a volume of 500,000 systems per year, according to Strategic Analysis (SA) estimates.« less

  1. The supersonic molecular beam injector as a reliable tool for plasma fueling and physics experiment on HL-2A.

    PubMed

    Chen, C Y; Yu, D L; Feng, B B; Yao, L H; Song, X M; Zang, L G; Gao, X Y; Yang, Q W; Duan, X R

    2016-09-01

    On HL-2A tokamak, supersonic molecular beam injection (SMBI) has been developed as a routine refueling method. The key components of the system are an electromagnetic valve and a conic nozzle. The valve and conic nozzle are assembled to compose the simplified Laval nozzle for generating the pulsed beam. The appurtenance of the system includes the cooling system serving the cooled SMBI generation and the in situ calibration component for quantitative injection. Compared with the conventional gas puffing, the SMBI features prompt response and larger fueling flux. These merits devote the SMBI a good fueling method, an excellent plasma density feedback control tool, and an edge localized mode mitigation resource.

  2. Fuel rod assembly to manifold attachment

    DOEpatents

    Donck, Harry A.; Veca, Anthony R.; Snyder, Jr., Harold J.

    1980-01-01

    A fuel element is formed with a plurality of fuel rod assemblies detachably connected to an overhead support with each of the fuel rod assemblies having a gas tight seal with the support to allow internal fission gaseous products to flow without leakage from the fuel rod assemblies into a vent manifold passageway system on the support. The upper ends of the fuel rod assemblies are located at vertically extending openings in the support and upper threaded members are threaded to the fuel rod assemblies to connect the latter to the support. The preferred threaded members are cap nuts having a dome wall encircling an upper threaded end on the fuel rod assembly and having an upper sealing surface for sealing contact with the support. Another and lower seal is achieved by abutting a sealing surface on each fuel rod assembly with the support. A deformable portion on the cap nut locks the latter against inadvertent turning off the fuel rod assembly. Orienting means on the fuel rod and support primarily locates the fuel rods azimuthally for reception of a deforming tool for the cap nut. A cross port in the fuel rod end plug discharges into a sealed annulus within the support, which serves as a circumferential chamber, connecting the manifold gas passageways in the support.

  3. Used fuel rail shock and vibration testing options analysis

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Ross, Steven B.; Best, Ralph E.; Klymyshyn, Nicholas A.

    2014-09-25

    The objective of the rail shock and vibration tests is to complete the framework needed to quantify loads of fuel assembly components that are necessary to guide materials research and establish a technical basis for review organizations such as the U.S. Nuclear Regulatory Commission (NRC). A significant body of experimental and numerical modeling data exists to quantify loads and failure limits applicable to normal conditions of transport (NCT) rail transport, but the data are based on assumptions that can only be verified through experimental testing. The test options presented in this report represent possible paths for acquiring the data thatmore » are needed to confirm the assumptions of previous work, validate modeling methods that will be needed for evaluating transported fuel on a case-by-case basis, and inform material test campaigns on the anticipated range of fuel loading. The ultimate goal of this testing is to close all of the existing knowledge gaps related to the loading of used fuel under NCT conditions and inform the experiments and analysis program on specific endpoints for their research. The options include tests that would use an actual railcar, surrogate assemblies, and real or simulated rail transportation casks. The railcar carrying the cradle, cask, and surrogate fuel assembly payload would be moved in a train operating over rail track modified or selected to impart shock and vibration forces that occur during normal rail transportation. Computer modeling would be used to help design surrogates that may be needed for a rail cask, a cask’s internal basket, and a transport cradle. The objective of the design of surrogate components would be to provide a test platform that effectively simulates responses to rail shock and vibration loads that would be exhibited by state-of-the-art rail cask, basket, and/or cradle structures. The computer models would also be used to help determine the placement of instrumentation (accelerometers and strain gauges) on the surrogate fuel assemblies, cask and cradle structures, and the railcar so that forces and deflections that would result in the greatest potential for damage to high burnup and long-cooled UNF can be determined. For purposes of this report we consider testing on controlled track when we have control of the track and speed to facilitate modeling.« less

  4. Advanced Fuels for Chemical Lasers

    DTIC Science & Technology

    1975-05-01

    age that the analysis is not limited to a single distance from the injector. The four curves in Fig. 86 demonstrate the types of reactions that are...annular wall for this concentric-element-type injector. The rear manifold assembly is made of machined nickel 1late and contains the four - bolt-hole...fabrication k] and assembly of the fully cooled laser combustor components, a series of four tests was cordicted to make preliminary assessment of

  5. Fuel injection assembly for use in turbine engines and method of assembling same

    DOEpatents

    Berry, Jonathan Dwight; Johnson, Thomas Edward; York, William David; Uhm, Jong Ho

    2015-12-15

    A fuel injection assembly for use in a turbine engine is provided. The fuel injection assembly includes an end cover, an endcap assembly, a fluid supply chamber, and a plurality of tube assemblies positioned at the endcap assembly. Each of the tube assemblies includes housing having a fuel plenum and a cooling fluid plenum. The cooling fluid plenum is positioned downstream from the fuel plenum and separated from the fuel plenum by an intermediate wall. The plurality of tube assemblies also include a plurality of tubes that extends through the housing. Each of the plurality of tubes is coupled in flow communication with the fluid supply chamber and a combustion chamber positioned downstream from the tube assembly. The plurality of tube assemblies further includes an aft plate at a downstream end of the cooling fluid plenum. The plate includes at least one aperture.

  6. Control assembly for controlling a fuel cell system during shutdown and restart

    DOEpatents

    Venkataraman, Ramki; Berntsen, George; Carlson, Glenn L.; Farooque, Mohammad; Beachy, Dan; Peterhans, Stefan; Bischoff, Manfred

    2010-06-15

    A fuel cell system and method in which the fuel cell system receives and an input oxidant gas and an input fuel gas, and in which a fuel processing assembly is provided and is adapted to at least humidify the input fuel gas which is to be supplied to the anode of the fuel cell of the system whose cathode receives the oxidant input gas via an anode oxidizing assembly which is adapted to couple the output of the anode of the fuel cell to the inlet of the cathode of the fuel cell during normal operation, shutdown and restart of the fuel cell system, and in which a control assembly is further provided and is adapted to respond to shutdown of the fuel cell system during which input fuel gas and input oxidant gas cease to be received by the fuel cell system, the control assembly being further adapted to, when the fuel cell system is shut down: control the fuel cell system so as to enable a purging gas to be able to flow through the fuel processing assembly to remove humidified fuel gas from the processing assembly and to enable a purging gas to be able to flow through the anode of the fuel cell.

  7. Automated assembling of single fuel cell units for use in a fuel cell stack

    NASA Astrophysics Data System (ADS)

    Jalba, C. K.; Muminovic, A.; Barz, C.; Nasui, V.

    2017-05-01

    The manufacturing of PEMFC stacks (POLYMER ELEKTROLYT MEMBRAN Fuel Cell) is nowadays still done by hand. Over hundreds of identical single components have to be placed accurate together for the construction of a fuel cell stack. Beside logistic problems, higher total costs and disadvantages in weight the high number of components produce a higher statistic interference because of faulty erection or material defects and summation of manufacturing tolerances. The saving of costs is about 20 - 25 %. Furthermore, the total weight of the fuel cells will be reduced because of a new sealing technology. Overall a one minute cycle time has to be aimed per cell at the manufacturing of these single components. The change of the existing sealing concept to a bonded sealing is one of the important requisites to get an automated manufacturing of single cell units. One of the important steps for an automated gluing process is the checking of the glue application by using of an image processing system. After bonding the single fuel cell the sealing and electrical function can be checked, so that only functional and high qualitative cells can get into further manufacturing processes.

  8. Fuel injection assembly for gas turbine engine combustor

    NASA Technical Reports Server (NTRS)

    Candy, Anthony J. (Inventor); Glynn, Christopher C. (Inventor); Barrett, John E. (Inventor)

    2002-01-01

    A fuel injection assembly for a gas turbine engine combustor, including at least one fuel stem, a plurality of concentrically disposed tubes positioned within each fuel stem, wherein a cooling supply flow passage, a cooling return flow passage, and a tip fuel flow passage are defined thereby, and at least one fuel tip assembly connected to each fuel stem so as to be in flow communication with the flow passages, wherein an active cooling circuit for each fuel stem and fuel tip assembly is maintained by providing all active fuel through the cooling supply flow passage and the cooling return flow passage during each stage of combustor operation. The fuel flowing through the active cooling circuit is then collected so that a predetermined portion thereof is provided to the tip fuel flow passage for injection by the fuel tip assembly.

  9. 76 FR 12825 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1; Confirmation of...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-09

    ... definitions for Damaged Fuel Assembly and Transfer Operations; add definitions for Fuel Class and Reconstituted Fuel Assembly; add Combustion Engineering 16x16 class fuel assemblies as authorized contents...

  10. Locking support for nuclear fuel assemblies

    DOEpatents

    Ledin, Eric

    1980-01-01

    A locking device for supporting and locking a nuclear fuel assembly within a cylindrical bore formed by a support plate, the locking device including a support and locking sleeve having upwardly extending fingers forming wedge shaped contact portions arranged for interaction between an annular tapered surface on the fuel assembly and the support plate bore as well as downwardly extending fingers having wedge shaped contact portions arranged for interaction between an annularly tapered surface on the support plate bore and the fuel assembly whereby the sleeve tends to support and lock the fuel assembly in place within the bore by its own weight while facilitating removal and/or replacement of the fuel assembly.

  11. 77 FR 70114 - Airworthiness Directives; Cessna Aircraft Company Airplanes

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-11-23

    ... assemblies, which were caused by the fuel return line assembly rubbing against the right steering tube assembly during full rudder pedal actuation. This AD requires you to inspect the fuel return line assembly... the fuel return line assembly and both the right steering tube assembly and the airplane structure...

  12. QUAD+ BWR Fuel Assembly demonstration program at Browns Ferry plant

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Doshi, P.K.; Mayhue, L.T.; Robert, J.T.

    1984-04-01

    The QUAD+ fuel assembly is an improved BWR fuel assembly designed and manufactured by Westinghouse Electric Corporation. The design features a water cross separating four fuel minibundles in an integral channel. A demonstration program for this fuel design is planned for late 1984 in cycle 6 of Browns Ferry 2, a TVA plant. Objectives for the design of the QUAD+ demonstration assemblies are compatibility in performance and transparency in safety analysis with the feed fuel. These objectives are met. Inspections of the QUAD+ demonstration assemblies are planned at each refueling outage.

  13. Alkaline static feed electrolyzer based oxygen generation system

    NASA Technical Reports Server (NTRS)

    Noble, L. D.; Kovach, A. J.; Fortunato, F. A.; Schubert, F. H.; Grigger, D. J.

    1988-01-01

    In preparation for the future deployment of the Space Station, an R and D program was established to demonstrate integrated operation of an alkaline Water Electrolysis System and a fuel cell as an energy storage device. The program's scope was revised when the Space Station Control Board changed the energy storage baseline for the Space Station. The new scope was aimed at the development of an alkaline Static Feed Electrolyzer for use in an Environmental Control/Life Support System as an oxygen generation system. As a result, the program was divided into two phases. The phase 1 effort was directed at the development of the Static Feed Electrolyzer for application in a Regenerative Fuel Cell System. During this phase, the program emphasized incorporation of the Regenerative Fuel Cell System design requirements into the Static Feed Electrolyzer electrochemical module design and the mechanical components design. The mechanical components included a Pressure Control Assembly, a Water Supply Assembly and a Thermal Control Assembly. These designs were completed through manufacturing drawing during Phase 1. The Phase 2 effort was directed at advancing the Alkaline Static Feed Electrolyzer database for an oxygen generation system. This development was aimed at extending the Static Feed Electrolyzer database in areas which may be encountered from initial fabrication through transportation, storage, launch and eventual Space Station startup. During this Phase, the Program emphasized three major areas: materials evaluation, electrochemical module scaling and performance repeatability and Static Feed Electrolyzer operational definition and characterization.

  14. Method for shearing spent nuclear fuel assemblies

    DOEpatents

    Weil, Bradley S.; Watson, Clyde D.

    1977-01-01

    A method is disclosed for shearing spent nuclear fuel assemblies of the type wherein a plurality of long metal tubes packed with ceramic fuel are supported in a spaced apart relationship within an outer metal shell or shroud which provides structural support to the assembly. Spent nuclear fuel assemblies are first compacted in a stepwise manner between specially designed gag-compactors and then sheared into short segments amenable to chemical processing by shear blades contoured to mate with the compacted surface of the fuel assembly.

  15. Fabrication of light water reactor tritium targets

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Pilger, J.P.

    1991-11-01

    The mission of the Fabrication Development Task of the Tritium Target Development Project is: to produce a documented technology basis, including specifications and procedures for target rod fabrication; to demonstrate that light water tritium targets can be manufactured at a rate consistent with tritium production requirements; and to develop quality control methods to evaluate target rod components and assemblies, and establish correlations between evaluated characteristics and target rod performance. Many of the target rod components: cladding tubes, end caps, plenum springs, etc., have similar counterparts in LWR fuel rods. High production rate manufacture and inspection of these components has beenmore » adequately demonstrated by nuclear fuel rod manufacturers. This summary describes the more non-conventional manufacturing processes and inspection techniques developed to fabricate target rod components whose manufacturability at required production rates had not been previously demonstrated.« less

  16. Separator plate for a fuel cell

    DOEpatents

    Petri, R.J.; Meek, J.; Bachta, R.P.; Marianowski, L.G.

    1996-04-02

    A separator plate is described for a fuel cell comprising an anode current collector, a cathode current collector and a main plate, the main plate disposed between the anode current collector and the cathode current collector. The anode current collector forms a flattened peripheral wet seal structure and manifold wet seal structure on the anode side of the separator plate and the cathode current collector forms a flattened peripheral wet seal structure and manifold wet seal structure on the cathode side of the separator plate. In this manner, the number of components required to manufacture and assemble a fuel cell stack is reduced. 9 figs.

  17. Separator plate for a fuel cell

    DOEpatents

    Petri, Randy J.; Meek, John; Bachta, Robert P.; Marianowski, Leonard G.

    1996-01-01

    A separator plate for a fuel cell comprising an anode current collector, a cathode current collector and a main plate, the main plate disposed between the anode current collector and the cathode current collector. The anode current collector forms a flattened peripheral wet seal structure and manifold wet seal structure on the anode side of the separator plate and the cathode current collector forms a flattened peripheral wet seal structure and manifold wet seal structure on the cathode side of the separator plate. In this manner, the number of components required to manufacture and assemble a fuel cell stack is reduced.

  18. Hydrogen-Fuel Engine Component Tests Near Completion

    NASA Technical Reports Server (NTRS)

    2003-01-01

    Gaseous hydrogen is burned off at the E1 Test Stand the night of Oct. 7 during a cold-flow test of the fuel turbopump of the Integrated Powerhead Demonstrator (IPD) at NASA Stennis Space Center (SSC). The gaseous hydrogen spins the pump's turbine during the test, which was conducted to verify the pump's performance. Engineers plan one more test before sending the pump to The Boeing Co. for inspection. It will then be returned to SSC for engine system assembly. The IPD is the first reusable hydrogen-fueled advanced engine in development since the Space Shuttle Main Engine.

  19. Hydrogen-Fuel Engine Component Tests Near Completion

    NASA Image and Video Library

    2003-11-05

    Gaseous hydrogen is burned off at the E1 Test Stand the night of Oct. 7 during a cold-flow test of the fuel turbopump of the Integrated Powerhead Demonstrator (IPD) at NASA Stennis Space Center (SSC). The gaseous hydrogen spins the pump's turbine during the test, which was conducted to verify the pump's performance. Engineers plan one more test before sending the pump to The Boeing Co. for inspection. It will then be returned to SSC for engine system assembly. The IPD is the first reusable hydrogen-fueled advanced engine in development since the Space Shuttle Main Engine.

  20. Two Rotor Stratified Charge Rotary Engine (SCRE) Engine System Technology Evaluation

    NASA Technical Reports Server (NTRS)

    Hoffman, T.; Mack, J.; Mount, R.

    1994-01-01

    This report summarizes results of an evaluation of technology enablement component technologies as integrated into a two rotor Stratified Charge Rotary Engine (SCRE). The work constitutes a demonstration of two rotor engine system technology, utilizing upgraded and refined component technologies derived from prior NASA Contracts NAS3-25945, NAS3-24628 and NAS-23056. Technical objectives included definition of, procurement and assembly of an advanced two rotor core aircraft engine, operation with Jet-A fuel at Take-Off rating of 340 BHP (254kW) and operation at a maximum cruise condition of 255 BHP (190kW), 75% cruise. A fuel consumption objective of 0.435 LBS/BHP-Hr (265 GRS/kW-Hr) was identified for the maximum cruise condition. A critical technology component item, a high speed, unit injector fuel injection system with electronic control was defined, procured and tested in conjunction with this effort. The two rotor engine configuration established herein defines an affordable, advanced, Jet-A fuel capability core engine (not including reduction gear, propeller shaft and some aircraft accessories) for General Aviation of the mid-1990's and beyond.

  1. Lightweight Stacks of Direct Methanol Fuel Cells

    NASA Technical Reports Server (NTRS)

    Narayanan, Sekharipuram; Valdez, Thomas

    2004-01-01

    An improved design concept for direct methanol fuel cells makes it possible to construct fuel-cell stacks that can weigh as little as one-third as much as do conventional bipolar fuel-cell stacks of equal power. The structural-support components of the improved cells and stacks can be made of relatively inexpensive plastics. Moreover, in comparison with conventional bipolar fuel-cell stacks, the improved fuel-cell stacks can be assembled, disassembled, and diagnosed for malfunctions more easily. These improvements are expected to bring portable direct methanol fuel cells and stacks closer to commercialization. In a conventional bipolar fuel-cell stack, the cells are interspersed with bipolar plates (also called biplates), which are structural components that serve to interconnect the cells and distribute the reactants (methanol and air). The cells and biplates are sandwiched between metal end plates. Usually, the stack is held together under pressure by tie rods that clamp the end plates. The bipolar stack configuration offers the advantage of very low internal electrical resistance. However, when the power output of a stack is only a few watts, the very low internal resistance of a bipolar stack is not absolutely necessary for keeping the internal power loss acceptably low.

  2. Fuel cell sub-assembly

    DOEpatents

    Chi, Chang V.

    1983-01-01

    A fuel cell sub-assembly comprising a plurality of fuel cells, a first section of a cooling means disposed at an end of the assembly and means for connecting the fuel cells and first section together to form a unitary structure.

  3. Elucidating energy and electron transfer dynamics within molecular assemblies for solar energy conversion

    NASA Astrophysics Data System (ADS)

    Morseth, Zachary Aaron

    The use of sunlight to make chemical fuels (i.e. solar fuels) is an attractive approach in the quest to develop sustainable energy sources. Using nature as a guide, assemblies for artificial photosynthesis will need to perform multiple functions. They will need to be able to harvest light across a broad region of the solar spectrum, transport excited-state energy to charge-separation sites, and then transport and store redox equivalents for use in the catalytic reactions that produce chemical fuels. This multifunctional behavior will require the assimilation of multiple components into a single macromolecular system. A wide variety of different architectures including porphyrin arrays, peptides, dendrimers, and polymers have been explored, with each design posing unique challenges. Polymer assemblies are attractive due to their relative ease of production and facile synthetic modification. However, their disordered nature gives rise to stochastic dynamics not present in more ordered assemblies. The rational design of assemblies requires a detailed understanding of the energy and electron transfer events that follow light absorption, which can occur on timescales ranging from femtoseconds to hundreds of microseconds, necessitating the use of sophisticated techniques. We have used a combination of time-resolved absorption and emission spectroscopies with observation times that span nine orders of magnitude to follow the excited-state evolution within single-site and polymer-based molecular assemblies. We complement experimental observations with electronic structure calculations, molecular dynamics simulations, and kinetic modeling to develop a microscopic view of these dynamics. This thesis provides an overview of work on single-site molecular assemblies and polymers decorated with pendant chromophores, both in solution and on surfaces. This work was made possible through extensive collaboration with Dr. Kirk Schanze's and Dr. John Reynolds' research groups who synthesized the samples for study.

  4. DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

    DOEpatents

    Harrer, J.M.; Fromm, L.W. Jr.; Kolba, V.M.

    1962-08-14

    A direct-cycle boiling-water nuclear reactor is described that employs a closed vessel and a plurality of fuel assemblies, each comprising an outer tube closed at its lower end, an inner tube, fuel rods in the space between the tubes and within the inner tube. A body of water lying within the pressure vessel and outside the fuel assemblies is converted to saturated steam, which enters each fuel assembly at the top and is converted to superheated steam in the fuel assembly while it is passing therethrough first downward through the space between the inner and outer tubes of the fuel assembly and then upward through the inner tube. (AEC)

  5. Current state of nuclear fuel cycles in nuclear engineering and trends in their development according to the environmental safety requirements

    NASA Astrophysics Data System (ADS)

    Vislov, I. S.; Pischulin, V. P.; Kladiev, S. N.; Slobodyan, S. M.

    2016-08-01

    The state and trends in the development of nuclear fuel cycles in nuclear engineering, taking into account the ecological aspects of using nuclear power plants, are considered. An analysis of advantages and disadvantages of nuclear engineering, compared with thermal engineering based on organic fuel types, was carried out. Spent nuclear fuel (SNF) reprocessing is an important task in the nuclear industry, since fuel unloaded from modern reactors of any type contains a large amount of radioactive elements that are harmful to the environment. On the other hand, the newly generated isotopes of uranium and plutonium should be reused to fabricate new nuclear fuel. The spent nuclear fuel also includes other types of fission products. Conditions for SNF handling are determined by ecological and economic factors. When choosing a certain handling method, one should assess these factors at all stages of its implementation. There are two main methods of SNF handling: open nuclear fuel cycle, with spent nuclear fuel assemblies (NFAs) that are held in storage facilities with their consequent disposal, and closed nuclear fuel cycle, with separation of uranium and plutonium, their purification from fission products, and use for producing new fuel batches. The development of effective closed fuel cycles using mixed uranium-plutonium fuel can provide a successful development of the nuclear industry only under the conditions of implementation of novel effective technological treatment processes that meet strict requirements of environmental safety and reliability of process equipment being applied. The diversity of technological processes is determined by different types of NFA devices and construction materials being used, as well as by the composition that depends on nuclear fuel components and operational conditions for assemblies in the nuclear power reactor. This work provides an overview of technological processes of SNF treatment and methods of handling of nuclear fuel assemblies. Based on analysis of modern engineering solutions on SNF regeneration, it has been concluded that new reprocessing technologies should meet the ecological safety requirements, provide a more extensive use of the resource base of nuclear engineering, allow the production of valuable and trace elements on an industrial scale, and decrease radioactive waste release.

  6. Fabrication Control Plan for ORNL RH-LOCA ATF Test Specimens to be Irradiated in the ATR

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Field, Kevin G.; Howard, Richard; Teague, Michael

    2014-06-01

    The purpose of this fabrication plan is (1) to summarize the design of a set of rodlets that will be fabricated and then irradiated in the Advanced Test Reactor (ATR) and (2) provide requirements for fabrication and acceptance criteria for inspections of the Light Water Reactor (LWR) – Accident Tolerant Fuels (ATF) rodlet components. The functional and operational (F&OR) requirements for the ATF program are identified in the ATF Test Plan. The scope of this document only covers fabrication and inspections of rodlet components detailed in drawings 604496 and 604497. It does not cover the assembly of these items tomore » form a completed test irradiation assembly or the inspection of the final assembly, which will be included in a separate INL final test assembly specification/inspection document. The controls support the requirements that the test irradiations must be performed safely and that subsequent examinations must provide valid results.« less

  7. Application of X-Ray Computer Tomography for Observing the Central Void Formations and the Fuel Pin Deformations of Irradiated FBR Fuel Assemblies

    NASA Astrophysics Data System (ADS)

    Katsuyama, Kozo; Nagamine, Tsuyoshi; Furuya, Hirotaka

    2010-10-01

    In order to observe the structural change in the interior of irradiated fuel assemblies, a non-destructive post-irradiation examination (PIE) technique using X-ray computer tomography (X-ray CT) was developed. This X-ray CT technique was applied to observe the central void formations and fuel pin deformations of fuel assemblies which had been irradiated at high linear heat rating. The central void sizes in all fuel pins were measured on five cross sections of the core fuel column as a parameter for evaluating fuel thermal performance. In addition, the fuel pin deformations were analyzed from X-ray CT images obtained along the axial direction of a fuel assembly at the same separation interval. A dependence of void size on the linear heat rating was seen in the fuel assembly irradiated at high linear heat rating. In addition, significant undulations of the fuel pin were observed along the axial direction, coinciding with the wrapping wire pitch in the core fuel column. Application of the developed technique should provide enhanced resolution of measurements and simplify fuel PIEs.

  8. The Conceptual Design for a Fuel Assembly of a New Research Reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Ryu, J-S.; Cho, Y-G.; Yoon, D-B.

    2004-10-06

    A new Research Reactor (ARR) has been under design by KAERI since 2002. In this work, as a first step for the design of the fuel assembly of the ARR, the conceptual design has been carried out. The vibration characteristics of the tubular fuel model and the locking performance of the preliminary designed locking devices were investigated. In order to investigate the effects of the stiffener on the vibration characteristics of the tubular fuel, a modal analysis was performed for the finite element models of the tubular fuels with stiffeners and without stiffeners. The analysis results show that the vibrationmore » characteristics of the tubular fuel with stiffeners are better than those of the tubular fuel without stiffeners. To investigate the locking performance of the preliminary designed locking devices for the fuel assembly of the ARR, the elements of the locking devices were fabricated. Then the torsional resistance, fixing status and vibration characteristics of the locking devices were tested. The test results show that using the locking device with fins on the bottom guide can prevent the torsional motion of the fuel assembly, and that additional springs or guides on the top of the fuel assembly are needed to suppress the lateral motion of the fuel assembly. Based on the modal analysis and experimental results, the fuel assembly and locking devices of the ARR were designed and its prototype was fabricated. The locking performance, pressure drop characteristics and vibration characteristics of the newly designed fuel assembly will be tested in the near future.« less

  9. Design and fabrication of miniaturized PEM fuel cell combined microreactor with self-regulated hydrogen mechanism

    NASA Astrophysics Data System (ADS)

    Balakrishnan, A.; Frei, M.; Kerzenmacher, S.; Reinecke, H.; Mueller, C.

    2015-12-01

    In this work we present the design and fabrication of the miniaturized PEM fuel cell combined microreactor system with hydrogen regulation mechanism and testing of prototype microreactor. The system consists of two components (i) fuel cell component and (ii) microreactor component. The fuel cell component represents the miniaturized PEM fuel cell system (combination of screen printed fuel cell assembly and an on-board hydrogen storage medium). Hydrogen production based on catalytic hydrolysis of chemical hydride takes place in the microreactor component. The self-regulated hydrogen mechanism based on the gaseous hydrogen produced from the catalytic hydrolysis of sodium borohydride (NaBH4) gets accumulated as bubbles at the vicinity of the hydrophobic coated hydrogen exhaust holes. When the built up hydrogen bubbles pressure exceeds the burst pressure at the hydrogen exhaust holes the bubble collapses. This collapse causes a surge of fresh NaBH4 solution onto the catalyst surface leading to the removal of the reaction by-products formed at the active sites of the catalyst. The catalyst used in the system is platinum deposited on a base substrate. Nickel foam, carbon porous medium (CPM) and ceramic plate were selected as candidates for base substrate for developing a robust catalyst surface. For the first time the platinum layer fabricated by pulsed electrodeposition and dealloying (EPDD) technique is used for hydrolysis of NaBH4. The major advantages of such platinum catalyst layers are its high surface area and their mechanical stability. Prototype microreactor system with self-regulated hydrogen mechanism is demonstrated.

  10. System and method for controlling a combustor assembly

    DOEpatents

    York, William David; Ziminsky, Willy Steve; Johnson, Thomas Edward; Stevenson, Christian Xavier

    2013-03-05

    A system and method for controlling a combustor assembly are disclosed. The system includes a combustor assembly. The combustor assembly includes a combustor and a fuel nozzle assembly. The combustor includes a casing. The fuel nozzle assembly is positioned at least partially within the casing and includes a fuel nozzle. The fuel nozzle assembly further defines a head end. The system further includes a viewing device configured for capturing an image of at least a portion of the head end, and a processor communicatively coupled to the viewing device, the processor configured to compare the image to a standard image for the head end.

  11. Swelling-resistant nuclear fuel

    DOEpatents

    Arsenlis, Athanasios [Hayward, CA; Satcher, Jr., Joe; Kucheyev, Sergei O [Oakland, CA

    2011-12-27

    A nuclear fuel according to one embodiment includes an assembly of nuclear fuel particles; and continuous open channels defined between at least some of the nuclear fuel particles, wherein the channels are characterized as allowing fission gasses produced in an interior of the assembly to escape from the interior of the assembly to an exterior thereof without causing significant swelling of the assembly. Additional embodiments, including methods, are also presented.

  12. 40 CFR 1060.102 - What permeation emission control requirements apply for fuel lines?

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... assemblies as aggregated systems that include multiple sections of fuel line with connectors and fittings. For example, you may certify fuel lines for portable marine fuel tanks as assemblies of fuel hose, primer bulbs, and self-sealing end connections. The length of such an assembly must not be longer than a...

  13. 40 CFR 1060.102 - What permeation emission control requirements apply for fuel lines?

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... assemblies as aggregated systems that include multiple sections of fuel line with connectors and fittings. For example, you may certify fuel lines for portable marine fuel tanks as assemblies of fuel hose, primer bulbs, and self-sealing end connections. The length of such an assembly must not be longer than a...

  14. Fabrication of fuel pin assemblies, phase 3

    NASA Technical Reports Server (NTRS)

    Keeton, A. R.; Stemann, L. G.

    1972-01-01

    Five full size and eight reduced length fuel pins were fabricated for irradiation testing to evaluate design concepts for a fast spectrum lithium cooled compact space power reactor. These assemblies consisted of uranium mononitride fuel pellets encased in a T-111 (Ta-8W-2Hf) clad with a tungsten barrier separating fuel and clad. Fabrication procedures were fully qualified by process development and assembly qualification tests. Detailed specifications and procedures were written for the fabrication and assembly of prototype fuel pins.

  15. ORIGEN-based Nuclear Fuel Inventory Module for Fuel Cycle Assessment: Final Project Report

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Skutnik, Steven E.

    The goal of this project, “ORIGEN-based Nuclear Fuel Depletion Module for Fuel Cycle Assessment" is to create a physics-based reactor depletion and decay module for the Cyclus nuclear fuel cycle simulator in order to assess nuclear fuel inventories over a broad space of reactor operating conditions. The overall goal of this approach is to facilitate evaluations of nuclear fuel inventories for a broad space of scenarios, including extended used nuclear fuel storage and cascading impacts on fuel cycle options such as actinide recovery in used nuclear fuel, particularly for multiple recycle scenarios. The advantages of a physics-based approach (compared tomore » a recipe-based approach which has been typically employed for fuel cycle simulators) is in its inherent flexibility; such an approach can more readily accommodate the broad space of potential isotopic vectors that may be encountered under advanced fuel cycle options. In order to develop this flexible reactor analysis capability, we are leveraging the Origen nuclear fuel depletion and decay module from SCALE to produce a standalone “depletion engine” which will serve as the kernel of a Cyclus-based reactor analysis module. The ORIGEN depletion module is a rigorously benchmarked and extensively validated tool for nuclear fuel analysis and thus its incorporation into the Cyclus framework can bring these capabilities to bear on the problem of evaluating long-term impacts of fuel cycle option choices on relevant metrics of interest, including materials inventories and availability (for multiple recycle scenarios), long-term waste management and repository impacts, etc. Developing this Origen-based analysis capability for Cyclus requires the refinement of the Origen analysis sequence to the point where it can reasonably be compiled as a standalone sequence outside of SCALE; i.e., wherein all of the computational aspects of Origen (including reactor cross-section library processing and interpolation, input and output processing, and depletion/decay solvers) can be self-contained into a single executable sequence. Further, to embed this capability into other software environments (such as the Cyclus fuel cycle simulator) requires that Origen’s capabilities be encapsulated into a portable, self-contained library which other codes can then call directly through function calls, thereby directly accessing the solver and data processing capabilities of Origen. Additional components relevant to this work include modernization of the reactor data libraries used by Origen for conducting nuclear fuel depletion calculations. This work has included the development of new fuel assembly lattices not previously available (such as for CANDU heavy-water reactor assemblies) as well as validation of updated lattices for light-water reactors updated to employ modern nuclear data evaluations. The CyBORG reactor analysis module as-developed under this workscope is fully capable of dynamic calculation of depleted fuel compositions from all commercial U.S. reactor assembly types as well as a number of international fuel types, including MOX, VVER, MAGNOX, and PHWR CANDU fuel assemblies. In addition, the Origen-based depletion engine allows for CyBORG to evaluate novel fuel assembly and reactor design types via creation of Origen reactor data libraries via SCALE. The establishment of this new modeling capability affords fuel cycle modelers a substantially improved ability to model dynamically-changing fuel cycle and reactor conditions, including recycled fuel compositions from fuel cycle scenarios involving material recycle into thermal-spectrum systems.« less

  16. Energy efficient engine: Turbine intermediate case and low-pressure turbine component test hardware detailed design report

    NASA Technical Reports Server (NTRS)

    Leach, K.; Thulin, R. D.; Howe, D. C.

    1982-01-01

    A four stage, low pressure turbine component has been designed to power the fan and low pressure compressor system in the Energy Efficient Engine. Designs for a turbine intermediate case and an exit guide vane assembly also have been established. The components incorporate numerous technology features to enhance efficiency, durability, and performance retention. These designs reflect a positive step towards improving engine fuel efficiency on a component level. The aerodynamic and thermal/mechanical designs of the intermediate case and low pressure turbine components are presented and described. An overview of the predicted performance of the various component designs is given.

  17. Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Pond, R.B.; Matos, J.E.

    1996-05-01

    As part of the Department of Energy`s spent nuclear fuel acceptance criteria, the mass of uranium and transuranic elements in spent research reactor fuel must be specified. These data are, however, not always known or readily determined. It is the purpose of this report to provide estimates of these data for some of the more common research reactor fuel assembly types. The specific types considered here are MTR, TRIGA and DIDO fuel assemblies. The degree of physical protection given to spent fuel assemblies is largely dependent upon the photon dose rate of the spent fuel material. These data also, aremore » not always known or readily determined. Because of a self-protecting dose rate level of radiation (dose rate greater than 100 ren-x/h at I m in air), it is important to know the dose rate of spent fuel assemblies at all time. Estimates of the photon dose rate for spent MTR, TRIGA and DIDO-type fuel assemblies are given in this report.« less

  18. LMFBR fuel assembly design for HCDA fuel dispersal

    DOEpatents

    Lacko, Robert E.; Tilbrook, Roger W.

    1984-01-01

    A fuel assembly for a liquid metal fast breeder reactor having an upper axial blanket region disposed in a plurality of zones within the fuel assembly. The characterization of a zone is dependent on the height of the axial blanket region with respect to the active fuel region. The net effect of having a plurality of zones is to establish a dispersal flow path for the molten materials resulting during a core meltdown accident. Upward flowing molten material can escape from the core region and/or fuel assembly without solidifying on the surface of fuel rods due to the heat sink represented by blanket region pellets.

  19. Fuel of the Bacterial Flagellar Type III Protein Export Apparatus.

    PubMed

    Minamino, Tohru; Kinoshita, Miki; Namba, Keiichi

    2017-01-01

    The flagellar type III export apparatus utilizes ATP and proton motive force (PMF) across the cytoplasmic membrane as the energy sources and transports flagellar component proteins from the cytoplasm to the distal growing end of the growing structure to construct the bacterial flagellum beyond the cellular membranes. The flagellar type III export apparatus coordinates flagellar protein export with assembly by ordered export of substrates to parallel with their order of the assembly. The export apparatus is composed of a PMF-driven transmembrane export gate complex and a cytoplasmic ATPase complex. Since the ATPase complex is dispensable for flagellar protein export, PMF is the primary fuel for protein unfolding and translocation. Interestingly, the export gate complex can also use sodium motive force across the cytoplasmic membrane in addition to PMF when the ATPase complex does not work properly. Here, we describe experimental protocols, which have allowed us to identify the export substrate class and the primary fuel of the flagellar type III protein export apparatus in Salmonella enterica serovar Typhimurium.

  20. Methods of conditioning direct methanol fuel cells

    DOEpatents

    Rice, Cynthia; Ren, Xiaoming; Gottesfeld, Shimshon

    2005-11-08

    Methods for conditioning the membrane electrode assembly of a direct methanol fuel cell ("DMFC") are disclosed. In a first method, an electrical current of polarity opposite to that used in a functioning direct methanol fuel cell is passed through the anode surface of the membrane electrode assembly. In a second method, methanol is supplied to an anode surface of the membrane electrode assembly, allowed to cross over the polymer electrolyte membrane of the membrane electrode assembly to a cathode surface of the membrane electrode assembly, and an electrical current of polarity opposite to that in a functioning direct methanol fuel cell is drawn through the membrane electrode assembly, wherein methanol is oxidized at the cathode surface of the membrane electrode assembly while the catalyst on the anode surface is reduced. Surface oxides on the direct methanol fuel cell anode catalyst of the membrane electrode assembly are thereby reduced.

  1. Flashback resistant pre-mixer assembly

    DOEpatents

    Laster, Walter R [Oviedo, FL; Gambacorta, Domenico [Oviedo, FL

    2012-02-14

    A pre-mixer assembly associated with a fuel supply system for mixing of air and fuel upstream from a main combustion zone in a gas turbine engine. The pre-mixer assembly includes a swirler assembly disposed about a fuel injector of the fuel supply system and a pre-mixer transition member. The swirler assembly includes a forward end defining an air inlet and an opposed aft end. The pre-mixer transition member has a forward end affixed to the aft end of the swirler assembly and an opposed aft end defining an outlet of the pre-mixer assembly. The aft end of the pre-mixer transition member is spaced from a base plate such that a gap is formed between the aft end of the pre-mixer transition member and the base plate for permitting a flow of purge air therethrough to increase a velocity of the air/fuel mixture exiting the pre-mixer assembly.

  2. Reliability prediction of large fuel cell stack based on structure stress analysis

    NASA Astrophysics Data System (ADS)

    Liu, L. F.; Liu, B.; Wu, C. W.

    2017-09-01

    The aim of this paper is to improve the reliability of Proton Electrolyte Membrane Fuel Cell (PEMFC) stack by designing the clamping force and the thickness difference between the membrane electrode assembly (MEA) and the gasket. The stack reliability is directly determined by the component reliability, which is affected by the material property and contact stress. The component contact stress is a random variable because it is usually affected by many uncertain factors in the production and clamping process. We have investigated the influences of parameter variation coefficient on the probability distribution of contact stress using the equivalent stiffness model and the first-order second moment method. The optimal contact stress to make the component stay in the highest level reliability is obtained by the stress-strength interference model. To obtain the optimal contact stress between the contact components, the optimal thickness of the component and the stack clamping force are optimally designed. Finally, a detailed description is given how to design the MEA and gasket dimensions to obtain the highest stack reliability. This work can provide a valuable guidance in the design of stack structure for a high reliability of fuel cell stack.

  3. Nondestructive Crack Detection in a Fuel System Component

    NASA Technical Reports Server (NTRS)

    Koshti, Ajay; Ruffino, Norman; Wincheski, Russell; Prosser, William; Winfree, William; Russell, Richard; Bryson, Craig; Devries, Robert; Engel, James; Landy, James

    2010-01-01

    The presentation examines the background and objective of nondestructive crack detection, flow control valve assembly and poppet post flight evaluation, poppet properties. magnetic property characterization of lab data, NDE, eddy current inspection, simulation, eddy current criteria, poppet cycle testing and NDE criteria, and the use of ultrasonic surface wave for crack detection.

  4. Self-assembling biomolecular catalysts for hydrogen production

    NASA Astrophysics Data System (ADS)

    Jordan, Paul C.; Patterson, Dustin P.; Saboda, Kendall N.; Edwards, Ethan J.; Miettinen, Heini M.; Basu, Gautam; Thielges, Megan C.; Douglas, Trevor

    2016-02-01

    The chemistry of highly evolved protein-based compartments has inspired the design of new catalytically active materials that self-assemble from biological components. A frontier of this biodesign is the potential to contribute new catalytic systems for the production of sustainable fuels, such as hydrogen. Here, we show the encapsulation and protection of an active hydrogen-producing and oxygen-tolerant [NiFe]-hydrogenase, sequestered within the capsid of the bacteriophage P22 through directed self-assembly. We co-opted Escherichia coli for biomolecular synthesis and assembly of this nanomaterial by expressing and maturing the EcHyd-1 hydrogenase prior to expression of the P22 coat protein, which subsequently self assembles. By probing the infrared spectroscopic signatures and catalytic activity of the engineered material, we demonstrate that the capsid provides stability and protection to the hydrogenase cargo. These results illustrate how combining biological function with directed supramolecular self-assembly can be used to create new materials for sustainable catalysis.

  5. Interface ring for gas turbine fuel nozzle assemblies

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Fox, Timothy A.; Schilp, Reinhard

    A gas turbine combustor assembly including a combustor liner and a plurality of fuel nozzle assemblies arranged in an annular array extending within the combustor liner. The fuel nozzle assemblies each include fuel nozzle body integral with a swirler assembly, and the swirler assemblies each include a bellmouth structure to turn air radially inwardly for passage into the swirler assemblies. A radially outer removed portion of each of the bellmouth structures defines a periphery diameter spaced from an inner surface of the combustor liner, and an interface ring is provided extending between the combustor liner and the removed portions ofmore » the bellmouth structures at the periphery diameter.« less

  6. Partial defect verification of spent fuel assemblies by PDET: Principle and field testing in Interim Spent fuel Storage Facility (CLAB) in Sweden

    DOE PAGES

    Ham, Y.; Kerr, P.; Sitaraman, S.; ...

    2016-05-05

    Here, the need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called "difficult-to-access" areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into "difficult-to-access" areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reportedmore » the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17×17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly bunrup levels.« less

  7. Partial Defect Verification of Spent Fuel Assemblies by PDET: Principle and Field Testing in Interim Spent Fuel Storage Facility (CLAB) in Sweden

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Ham, Y.S.; Kerr, P.; Sitaraman, S.

    The need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called 'difficult-to-access' areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into 'difficult-to-access' areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reported themore » successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17x17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly burnup levels. (authors)« less

  8. Partial defect verification of spent fuel assemblies by PDET: Principle and field testing in Interim Spent fuel Storage Facility (CLAB) in Sweden

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Ham, Y.; Kerr, P.; Sitaraman, S.

    Here, the need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called "difficult-to-access" areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into "difficult-to-access" areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reportedmore » the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17×17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly bunrup levels.« less

  9. Inactive end cell assembly for fuel cells for improved electrolyte management and electrical contact

    DOEpatents

    Yuh, Chao-Yi [New Milford, CT; Farooque, Mohammad [Danbury, CT; Johnsen, Richard [New Fairfield, CT

    2007-04-10

    An assembly for storing electrolyte in a carbonate fuel cell is provided. The combination of a soft, compliant and resilient cathode current collector and an inactive anode part including a foam anode in each assembly mitigates electrical contact loss during operation of the fuel cell stack. In addition, an electrode reservoir in the positive end assembly and an electrode sink in the negative end assembly are provided, by which ribbed and flat cathode members inhibit electrolyte migration in the fuel cell stack.

  10. Spent nuclear fuel assembly inspection using neutron computed tomography

    NASA Astrophysics Data System (ADS)

    Pope, Chad Lee

    The research presented here focuses on spent nuclear fuel assembly inspection using neutron computed tomography. Experimental measurements involving neutron beam transmission through a spent nuclear fuel assembly serve as benchmark measurements for an MCNP simulation model. Comparison of measured results to simulation results shows good agreement. Generation of tomography images from MCNP tally results was accomplished using adapted versions of built in MATLAB algorithms. Multiple fuel assembly models were examined to provide a broad set of conclusions. Tomography images revealing assembly geometric information including the fuel element lattice structure and missing elements can be obtained using high energy neutrons. A projection difference technique was developed which reveals the substitution of unirradiated fuel elements for irradiated fuel elements, using high energy neutrons. More subtle material differences such as altering the burnup of individual elements can be identified with lower energy neutrons provided the scattered neutron contribution to the image is limited. The research results show that neutron computed tomography can be used to inspect spent nuclear fuel assemblies for the purpose of identifying anomalies such as missing elements or substituted elements. The ability to identify anomalies in spent fuel assemblies can be used to deter diversion of material by increasing the risk of early detection as well as improve reprocessing facility operations by confirming the spent fuel configuration is as expected or allowing segregation if anomalies are detected.

  11. Cell module and fuel conditioner

    NASA Technical Reports Server (NTRS)

    Hoover, D. Q., Jr.

    1980-01-01

    The computer code for the detailed analytical model of the MK-2 stacks is described. An ERC proprietary matrix is incorporated in the stacks. The mechanical behavior of the stack during thermal cycles under compression was determined. A 5 cell stack of the MK-2 design was fabricated and tested. Designs for the next three stacks were selected and component fabrication initiated. A 3 cell stack which verified the use of wet assembly and a new acid fill procedure were fabricated and tested. Components for the 2 kW test facility were received or fabricated and construction of the facility is underway. The definition of fuel and water is used in a study of the fuel conditioning subsystem. Kinetic data on several catalysts, both crushed and pellets, was obtained in the differential reactor. A preliminary definition of the equipment requirements for treating tap and recovered water was developed.

  12. Spent nuclear fuel system dynamic stability under normal conditions of transportation

    DOE PAGES

    Jiang, Hao; Wang, Jy-An John

    2016-10-14

    In a horizontal layout of a spent nuclear fuel (SNF) assembly under normal conditions of transportation (NCT), the fuel assembly’s skeleton formed by guide tubes and spacer grids is the primary load bearing structure for carrying and transferring the vibration loads within an SNF assembly. Therefore, the integrity of guide tubes and spacer grids will dictate the vibration amplitude/intensity of the fuel assembly during transport, and must be considered when designing multipurpose purpose canister (MPC) for safe SNF transport. This paper investigates the SNF assembly deformation dynamics during normal vibration mode, as well as the transient shock mode inside themore » cask during NCT. In conclusion, dynamic analyses were performed in the frequency domain to study frequency characteristic of the fuel assembly system and in the time domain to simulate the transient dynamic response of the fuel assembly.« less

  13. Spent nuclear fuel system dynamic stability under normal conditions of transportation

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Jiang, Hao; Wang, Jy-An John

    In a horizontal layout of a spent nuclear fuel (SNF) assembly under normal conditions of transportation (NCT), the fuel assembly’s skeleton formed by guide tubes and spacer grids is the primary load bearing structure for carrying and transferring the vibration loads within an SNF assembly. Therefore, the integrity of guide tubes and spacer grids will dictate the vibration amplitude/intensity of the fuel assembly during transport, and must be considered when designing multipurpose purpose canister (MPC) for safe SNF transport. This paper investigates the SNF assembly deformation dynamics during normal vibration mode, as well as the transient shock mode inside themore » cask during NCT. In conclusion, dynamic analyses were performed in the frequency domain to study frequency characteristic of the fuel assembly system and in the time domain to simulate the transient dynamic response of the fuel assembly.« less

  14. System and method for injecting fuel

    DOEpatents

    Uhm, Jong Ho; Johnson, Thomas Edward

    2012-12-04

    According to various embodiments, a system includes a staggered multi-nozzle assembly. The staggered multi-nozzle assembly includes a first fuel nozzle having a first axis and a first flow path extending to a first downstream end portion, wherein the first fuel nozzle has a first non-circular perimeter at the first downstream end portion. The staggered multi-nozzle assembly also includes a second fuel nozzle having a second axis and a second flow path extending to a second downstream end portion, wherein the first and second downstream end portions are axially offset from one another relative to the first and second axes. The staggered multi-nozzle assembly further includes a cap member disposed circumferentially about at least the first and second fuel nozzles to assemble the staggered multi-nozzle assembly.

  15. Fuel assembly for nuclear reactors

    DOEpatents

    Creagan, Robert J.; Frisch, Erling

    1977-01-01

    A new and improved fuel assembly is formed to minimize the amount of parasitic structural material wherein a plurality of hollow tubular members are juxtaposed to the fuel elements of the assembly. The tubular members may serve as guide tubes for control elements and are secured to a number of longitudinally spaced grid members along the fuel assembly. The grid members include means thereon engaging each of the fuel elements to laterally position the fuel elements in a predetermined array. Openings in the bottom of each hollow member serve as a shock absorber to cushion shock transmitted to the structure when the control elements are rapidly inserted in their corresponding tubular members.

  16. Spent fuel burnup estimation by Cerenkov glow intensity measurement

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Kuribara, Masayuki

    1994-10-01

    The Cerenkov glow images from irradiated fuel assemblies of boiling-water reactors (BWR) and pressurized-water reactors (PWR) are generally used for inspections. For this purpose, a new UV-I.I. CVD (ultra-violet light image intensifier Cerenkov viewing device), has been developed. This new device can measure the intensity of the Cerenkov glow from a spent fuel assembly, thus making it possible to estimate the burnup of the fuel assembly by comparing the Cerenkov glow intensity to the reference intensity. The experiment was carried out on BWR spent fuel assemblies and the results show that burnups are estimated within 20% accuracy compared to themore » declared burnups for the tested spent fuel assemblies for cooling times ranging from 900--2.000 d.« less

  17. DOE Office of Scientific and Technical Information (OSTI.GOV)

    Wiebe, David J; Fox, Timothy A

    A fuel nozzle assembly for use in a combustor apparatus of a gas turbine engine. An outer housing of the fuel nozzle assembly includes an inner volume and provides a direct structural connection between a duct structure and a fuel manifold. The duct structure defines a flow passage for combustion gases flowing within the combustor apparatus. The fuel manifold defines a fuel supply channel therein in fluid communication with a source of fuel. A fuel injector of the fuel nozzle assembly is provided in the inner volume of the outer housing and defines a fuel passage therein. The fuel passagemore » is in fluid communication with the fuel supply channel of the fuel manifold for distributing the fuel from the fuel supply channel into the flow passage of the duct structure.« less

  18. MOX fuel arrangement for nuclear core

    DOEpatents

    Kantrowitz, M.L.; Rosenstein, R.G.

    1998-10-13

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly. 38 figs.

  19. Mox fuel arrangement for nuclear core

    DOEpatents

    Kantrowitz, Mark L.; Rosenstein, Richard G.

    2001-05-15

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion. characteristics of the assembly.

  20. MOX fuel arrangement for nuclear core

    DOEpatents

    Kantrowitz, Mark L.; Rosenstein, Richard G.

    2001-07-17

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

  1. MOX fuel arrangement for nuclear core

    DOEpatents

    Kantrowitz, Mark L.; Rosenstein, Richard G.

    1998-01-01

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

  2. Cell module and fuel conditioner development

    NASA Technical Reports Server (NTRS)

    Hoover, D. Q., Jr.

    1981-01-01

    The results of pretesting and performance testing of Stack 564 are reported. The design features, progress in fabrication and plans for assembly of Stack 800 are given. The status of endurance testing of Stack 560 is reported. The design, fabrication, test procedures and preliminary tests of the 10 kW double counterflow reformer and the reformer test stand are described. Results of vendor contacts to define the performance and cost of fuel conditioning system components are reported. The results of burner tests and continuing development of the BOLTAR program are reported.

  3. Viper cabin-fuselage structural design concept with engine installation and wing structural design

    NASA Technical Reports Server (NTRS)

    Marchesseault, B.; Carr, D.; Mccorkle, T.; Stevens, C.; Turner, D.

    1993-01-01

    This report describes the process and considerations in designing the cabin, nose, drive shaft, and wing assemblies for the 'Viper' concept aircraft. Interfaces of these assemblies, as well as interfaces with the sections of the aircraft aft of the cabin, are also discussed. The results of the design process are included. The goal of this project is to provide a structural design which complies with FAR 23 requirements regarding occupant safety, emergency landing loads, and maneuvering loads. The design must also address the interfaces of the various systems in the cabin, nose, and wing, including the drive shaft, venting, vacuum, electrical, fuel, and control systems. Interfaces between the cabin assembly and the wing carrythrough and empennage assemblies were required, as well. In the design of the wing assemblies, consistency with the existing cabin design was required. The major areas considered in this report are materials and construction, loading, maintenance, environmental considerations, wing assembly fatigue, and weight. The first three areas are developed separately for the nose, cabin, drive shaft, and wing assemblies, while the last three are discussed for the entire design. For each assembly, loading calculations were performed to determine the proper sizing of major load carrying components. Table 1.0 lists the resulting margins of safety for these key components, along with the types of the loads involved, and the page number upon which they are discussed.

  4. Nuclear reactor

    DOEpatents

    Yant, Howard W.; Stinebiser, Karl W.; Anzur, Gregory C.

    1977-01-01

    A nuclear reactor, particularly a liquid-metal breeder reactor, whose upper internals include outlet modules for channeling the liquid-metal coolant from selected areas of the outlet of the core vertically to the outlet plenum. The modules are composed of a highly-refractory, high corrosion-resistant alloy, for example, INCONEL-718. Each module is disposed to confine and channel generally vertically the coolant emitted from a subplurality of core-component assemblies. Each module has a grid with openings, each opening disposed to receive the coolant from an assembly of the subplurality. The grid in addition serves as a holdown for the assemblies of the corresponding subplurality preventing their excessive ejection upwardly from the core. In the region directly over the core the outlet modules are of such peripheral form that they nest forming a continuum over the core-component assemblies whose outlet coolant they confine. Each subassembly includes a chimney which confines the coolant emitted by its corresponding subassemblies to generally vertical flow between the outlet of the core and the outlet plenum. Each subplurality of assemblies whose emitted coolant is confined by an outlet module includes assemblies which emit lower-temperature coolant, for example, a control-rod assembly, or fertile assemblies, and assemblies which emit coolant of substantially higher temperature, for example, fuel-rod assemblies. The coolants of different temperatures are mixed in the chimneys reducing the effect of stripping (hot-cold temperature fluctuations) on the remainder of the upper internals which are composed typically of AISI-304 or AISI-316 stainless steel.

  5. Reduced size fuel cell for portable applications

    NASA Technical Reports Server (NTRS)

    Narayanan, Sekharipuram R. (Inventor); Valdez, Thomas I. (Inventor); Clara, Filiberto (Inventor); Frank, Harvey A. (Inventor)

    2004-01-01

    A flat pack type fuel cell includes a plurality of membrane electrode assemblies. Each membrane electrode assembly is formed of an anode, an electrolyte, and an cathode with appropriate catalysts thereon. The anode is directly into contact with fuel via a wicking element. The fuel reservoir may extend along the same axis as the membrane electrode assemblies, so that fuel can be applied to each of the anodes. Each of the fuel cell elements is interconnected together to provide the voltage outputs in series.

  6. Status of Fuel Development and Manufacturing for Space Nuclear Reactors at BWX Technologies

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Carmack, W.J.; Husser, D.L.; Mohr, T.C.

    2004-02-04

    New advanced nuclear space propulsion systems will soon seek a high temperature, stable fuel form. BWX Technologies Inc (BWXT) has a long history of fuel manufacturing. UO2, UCO, and UCx have been fabricated at BWXT for various US and international programs. Recent efforts at BWXT have focused on establishing the manufacturing techniques and analysis capabilities needed to provide a high quality, high power, compact nuclear reactor for use in space nuclear powered missions. To support the production of a space nuclear reactor, uranium nitride has recently been manufactured by BWXT. In addition, analytical chemistry and analysis techniques have been developedmore » to provide verification and qualification of the uranium nitride production process. The fabrication of a space nuclear reactor will require the ability to place an unclad fuel form into a clad structure for assembly into a reactor core configuration. To this end, BWX Technologies has reestablished its capability for machining, GTA welding, and EB welding of refractory metals. Specifically, BWX Technologies has demonstrated GTA welding of niobium flat plate and EB welding of niobium and Nb-1Zr tubing. In performing these demonstration activities, BWX Technologies has established the necessary infrastructure to manufacture UO2, UCx, or UNx fuel, components, and complete reactor assemblies in support of space nuclear programs.« less

  7. Shock ignition of thermonuclear fuel with high areal density.

    PubMed

    Betti, R; Zhou, C D; Anderson, K S; Perkins, L J; Theobald, W; Solodov, A A

    2007-04-13

    A novel method by C. Zhou and R. Betti [Bull. Am. Phys. Soc. 50, 140 (2005)] to assemble and ignite thermonuclear fuel is presented. Massive cryogenic shells are first imploded by direct laser light with a low implosion velocity and on a low adiabat leading to fuel assemblies with large areal densities. The assembled fuel is ignited from a central hot spot heated by the collision of a spherically convergent ignitor shock and the return shock. The resulting fuel assembly features a hot-spot pressure greater than the surrounding dense fuel pressure. Such a nonisobaric assembly requires a lower energy threshold for ignition than the conventional isobaric one. The ignitor shock can be launched by a spike in the laser power or by particle beams. The thermonuclear gain can be significantly larger than in conventional isobaric ignition for equal driver energy.

  8. Method of preparing gas tags for identification of single and multiple failures of nuclear reactor fuel assemblies

    DOEpatents

    McCormick, Norman J.

    1976-01-01

    For use in the identification of failed fuel assemblies in a nuclear reactor, the ratios of the tag gas isotopic concentrations are located on curved surfaces to enable the ratios corresponding to failure of a single fuel assembly to be distinguished from those formed from any combination of two or more failed assemblies.

  9. Turbine combustor with fuel nozzles having inner and outer fuel circuits

    DOEpatents

    Uhm, Jong Ho; Johnson, Thomas Edward; Kim, Kwanwoo

    2013-12-24

    A combustor cap assembly for a turbine engine includes a combustor cap and a plurality of fuel nozzles mounted on the combustor cap. One or more of the fuel nozzles would include two separate fuel circuits which are individually controllable. The combustor cap assembly would be controlled so that individual fuel circuits of the fuel nozzles are operated or deliberately shut off to provide for physical separation between the flow of fuel delivered by adjacent fuel nozzles and/or so that adjacent fuel nozzles operate at different pressure differentials. Operating a combustor cap assembly in this fashion helps to reduce or eliminate the generation of undesirable and potentially harmful noise.

  10. Self-actuated nuclear reactor shutdown system using induction pump to facilitate sensing of core coolant temperature

    DOEpatents

    Sievers, Robert K.; Cooper, Martin H.; Tupper, Robert B.

    1987-01-01

    A self-actuated shutdown system incorporated into a reactivity control assembly in a nuclear reactor includes pumping means for creating an auxiliary downward flow of a portion of the heated coolant exiting from the fuel assemblies disposed adjacent to the control assembly. The shutdown system includes a hollow tubular member which extends through the outlet of the control assembly top nozzle so as to define an outer annular flow channel through the top nozzle outlet separate from an inner flow channel for primary coolant flow through the control assembly. Also, a latching mechanism is disposed in an inner duct of the control assembly and is operable for holding absorber bundles in a raised position in the control assembly and for releasing them to drop them into the core of the reactor for shutdown purposes. The latching mechanism has an inner flow passage extending between and in flow communication with the absorber bundles and the inner flow channel of the top nozzle for accommodating primary coolant flow upwardly through the control assembly. Also, an outer flow passage separate from the inner flow passage extends through the latching mechanism between and in flow communication with the inner duct and the outer flow channel of the top nozzle for accommodating inflow of a portion of the heated coolant from the adjacent fuel assemblies. The latching mechanism contains a magnetic material sensitive to temperature and operable to cause mating or latching together of the components of the latching mechanism when the temperature sensed is below a known temperature and unmating or unlatching thereof when the temperature sensed is above a given temperature. The temperature sensitive magnetic material is positioned in communication with the heated coolant flow through the outer flow passage for directly sensing the temperature thereof. Finally, the pumping means includes a jet induction pump nozzle and diffuser disposed adjacent the bottom nozzle of the control assembly and in flow communication with the inlet thereof. The pump nozzle is operable to create an upward driving flow of primary coolant through the pump diffuser and then to the absorber bundles. The upward driving flow of primary coolant, in turn, creates a suction head within the outer flow channel of the top nozzle and thereby an auxiliary downward flow of the heated coolant portion exiting from the upper end of the adjacent fuel assemblies through the outer flow channel to the pump nozzle via the outer flow passage of the latching mechanism and an annular space between the outer and inner spaced ducts of the control assembly housing. The temperature of the heated coolant exiting from the adjacent fuel assemblies can thereby be sensed directly by the temperature sensitive magnetic material in the latching mechanism.

  11. Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Mertyurek, Ugur; Gauld, Ian C.

    In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup.more » The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.« less

  12. Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

    DOE PAGES

    Mertyurek, Ugur; Gauld, Ian C.

    2015-12-24

    In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup.more » The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.« less

  13. Fuel and oxidizer valve assembly employs single solenoid actuator

    NASA Technical Reports Server (NTRS)

    1966-01-01

    Valve assembly simultaneously starts or stops the flow of oxidizer and fuel from separate inlet channels to reaction control motors. The assembly combines an oxidizer shutoff valve and a fuel shutoff valve which are mechanically linked and operated by a single high-speed solenoid actuator.

  14. Modular fuel-cell stack assembly

    DOEpatents

    Patel, Pinakin

    2010-07-13

    A fuel cell assembly having a plurality of fuel cells arranged in a stack. An end plate assembly abuts the fuel cell at an end of said stack. The end plate assembly has an inlet area adapted to receive an exhaust gas from the stack, an outlet area and a passage connecting the inlet area and outlet area and adapted to carry the exhaust gas received at the inlet area from the inlet area to the outlet area. A further end plate assembly abuts the fuel cell at a further opposing end of the stack. The further end plate assembly has a further inlet area adapted to receive a further exhaust gas from the stack, a further outlet area and a further passage connecting the further inlet area and further outlet area and adapted to carry the further exhaust gas received at the further inlet area from the further inlet area to the further outlet area.

  15. Method and apparatus for close packing of nuclear fuel assemblies

    DOEpatents

    Newman, Darrell F.

    1993-01-01

    The apparatus of the present invention is a plate of neutron absorbing material. The plate may have a releasable locking feature permitting the plate to be secured within a nuclear fuel assembly between nuclear fuel rods during storage or transportation then removed for further use or destruction. The method of the present invention has the step of placing a plate of neutron absorbing material between nuclear fuel rods within a nuclear fuel assembly, preferably between the two outermost columns of nuclear fuel rods. Additionally, the plate may be releasably locked in place.

  16. Method and apparatus for close packing of nuclear fuel assemblies

    DOEpatents

    Newman, D.F.

    1993-03-30

    The apparatus of the present invention is a plate of neutron absorbing material. The plate may have a releasable locking feature permitting the plate to be secured within a nuclear fuel assembly between nuclear fuel rods during storage or transportation then removed for further use or destruction. The method of the present invention has the step of placing a plate of neutron absorbing material between nuclear fuel rods within a nuclear fuel assembly, preferably between the two outermost columns of nuclear fuel rods. Additionally, the plate may be releasably locked in place.

  17. Pool-site fuel inspection and examination techniques applied by the Kraftwerk Union Aktiengesellschaft Fuel Service. [PWR; BWR

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Knaab, H.; Knecht, K.

    The need for pool-site inspection and examination of fuel assemblies was recognized by Kraftwerk Union Aktiengesellschaft with the commissioning of the first nuclear power stations. A wet sipping method has demonstrated high reliability in detection of leaking fuel assemblies. The visual inspection system is a versatile tool. It can be supplemented by attaching devices for oxide thickness measurement or surface replication. Repair of leaking pressurized water reactor fuel assemblies has improved fuel utilization. Applied methods and typical results are described.

  18. Piloted rich-catalytic lean-burn hybrid combustor

    DOEpatents

    Newburry, Donald Maurice

    2002-01-01

    A catalytic combustor assembly which includes, an air source, a fuel delivery means, a catalytic reactor assembly, a mixing chamber, and a means for igniting a fuel/air mixture. The catalytic reactor assembly is in fluid communication with the air source and fuel delivery means and has a fuel/air plenum which is coated with a catalytic material. The fuel/air plenum has cooling air conduits passing therethrough which have an upstream end. The upstream end of the cooling conduits is in fluid communication with the air source but not the fuel delivery means.

  19. Some methods for achieving more efficient performance of fuel assemblies

    NASA Astrophysics Data System (ADS)

    Boltenko, E. A.

    2014-07-01

    More efficient operation of reactor plant fuel assemblies can be achieved through the use of new technical solutions aimed at obtaining more uniform distribution of coolant over the fuel assembly section, more intense heat removal on convex heat-transfer surfaces, and higher values of departure from nucleate boiling ratio (DNBR). Technical solutions using which it is possible to obtain more intense heat removal on convex heat-transfer surfaces and higher DNBR values in reactor plant fuel assemblies are considered. An alternative heat removal arrangement is described using which it is possible to obtain a significantly higher power density in a reactor plant and essentially lower maximal fuel rod temperature.

  20. Assess How Changes in Fuel Cycle Operation Impact Safeguards

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Tobin, Stephen Joseph; Adigun, Babatunde John; Fugate, Michael Lynn

    Since the beginning of commercial nuclear power generation in the 1960s, the ability of researchers to understand and control the isotopic content of spent fuel has improved. It is therefore not surprising that both fuel assembly design and fuel assembly irradiation optimization have improved over the past 50+ years. It is anticipated that the burnup and isotopics of the spent fuel should exhibit less variation over the decades as reactor operators irradiate each assembly to the optimum amount. In contrast, older spent fuel is anticipated to vary more in burnup and resulting isotopics for a given initial enrichment. Modern fuelmore » therefore should be more uniform in composition, and thus, measured safeguards results should be easier to interpret than results from older spent fuel. With spent fuel ponds filling up, interim and long-­term storage of spent fuel will need to be addressed. Additionally after long periods of storage, spent fuel is no longer self-­protecting and, as such, the IAEA will categorize it as more attractive; in approximately 20 years many of the assemblies from early commercial cores will no longer be considered self-­protecting. This study will assess how more recent changes in the reactor operation could impact the interpretation of safeguards measurements. The status quo for spent fuel assay in the safeguards context is that the overwhelming majority of spent fuel assemblies are not measured in a quantitative way except for those assemblies about to be loaded into a difficult or impossible to access location (dry storage or, in the future, a repository). In other words, when the assembly is still accessible to a state actor, or an insider, when it is cooling in a pool, the inspectorate does not have a measurement database that could assist them in re-­verifying the integrity of that assembly. The spent fuel safeguards regime would be strengthened if spent fuel assemblies were measured from discharge to loading into a difficult or impossible to access location. The primary driver for suggesting this shift in approach is the change in robotic technology and information technology in general. It should be possible, with minimal impact to the facility, to measure each assembly every time that it is moved in the pool, with the first measurements being made at discharge. The following conclusions were reached: The total neutron count rate can be accurately predicted at any future moment in time based upon the measured count rate at discharge, provided the initial enrichment and burnup of the assembly is known at discharge. It is expected that the total neutron count rate measured at discharge will be indicative of the initial enrichment and burnup of that assembly. If the automated robot were to focus on measuring the assemblies in the rack without moving them, the time available would increase immensely.« less

  1. Analysis of gamma ray dose for dried up pond storing low enriched UO2 fuel

    NASA Astrophysics Data System (ADS)

    Nauchi, Yasushi; Suzuki, Motomu

    2017-09-01

    Gamma ray dose is calculated for loss of coolant accident in spent fuel pond (SFP) storing irradiated fuels used in light water reactors. Influence of modelling of fuel assemblies, source distributions, and loading fraction of fuel assemblies in the fuel rack on the dose are investigated.

  2. Risk Assessment of Structural Integrity of Transportation Casks after Extended Storage

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Ibarra, Luis; Medina, Ricardo; Yang, Haori

    This study assessed the risk of loss of structural integrity of transportation casks and fuel cladding after extended storage. Although it is known that fuel rods discharged from NPPs have a small percentage of rod cladding defects, the behavior of fuel cladding and the structural elements of assemblies during transportation after long-term storage is not well understood. If the fuel degrades during extended storage, it could be susceptible to damage from vibration and impact loads during transport operations, releasing fission-product gases into the canister or the cask interior (NWTRB 2010). Degradation of cladding may occur due to mechanisms associated withmore » hydrogen embrittlement, delayed hydride cracking, low temperature creep, and stress corrosion cracking (SCC) that may affect fuel cladding and canister components after extended storage of hundreds of years. Over extended periods at low temperatures, these mechanisms affect the ductility, strength, and fracture toughness of the fuel cladding, which becomes brittle. For transportation purposes, the fuel may be transferred from storage to shipping casks, or dual-purpose casks may be used for storage and transportation. Currently, most of the transportation casks will be the former case. A risk assessment evaluation is conducted based on results from experimental tests and simulations with advanced numerical models. A novel contribution of this study is the evaluation of the combined effect of component aging and vibration/impact loads in transportation scenarios. The expected levels of deterioration will be obtained from previous and current studies on the effect of aging on fuel and cask components. The emphasis of the study is placed on the structural integrity of fuel cladding and canisters.« less

  3. Optimum concentration gradient of the electrocatalyst, Nafion® and poly(tetrafluoroethylene) in a membrane-electrode-assembly for enhanced performance of direct methanol fuel cells.

    PubMed

    Liu, Jing Hua; Jeon, Min Ku; Lee, Ki Rak; Woo, Seong Ihl

    2010-12-14

    A combinatorial library of membrane-electrode-assemblies (MEAs) which consisted of 27 different compositions was fabricated to optimize the multilayer structure of direct methanol fuel cells. Each spot consisted of three layers of ink and a gradient was generated by employing different concentrations of the three components (Pt catalyst, Nafion® and polytetrafluoroethylene (PTFE)) of each layer. For quick evaluation of the library, a high-throughput optical screening technique was employed for methanol electro-oxidation reaction (MOR) activity. The screening results revealed that gradient layers could lead to higher MOR activity than uniform layers. It was found that the MOR activity was higher when the concentrations of Pt catalyst and Nafion ionomer decreased downward from the top layer to the bottom layer. On the other hand, higher MOR activity was observed when PTFE concentration increased downward from the top to the bottom layer.

  4. A mathematical model for predicting the life of polymer electrolyte fuel cell membranes subjected to hydration cycling

    NASA Astrophysics Data System (ADS)

    Burlatsky, S. F.; Gummalla, M.; O'Neill, J.; Atrazhev, V. V.; Varyukhin, A. N.; Dmitriev, D. V.; Erikhman, N. S.

    2012-10-01

    Under typical Polymer Electrolyte Membrane Fuel Cell (PEMFC) fuel cell operating conditions, part of the membrane electrode assembly is subjected to humidity cycling due to variation of inlet gas RH and/or flow rate. Cyclic membrane hydration/dehydration would cause cyclic swelling/shrinking of the unconstrained membrane. In a constrained membrane, it causes cyclic stress resulting in mechanical failure in the area adjacent to the gas inlet. A mathematical modeling framework for prediction of the lifetime of a PEMFC membrane subjected to hydration cycling is developed in this paper. The model predicts membrane lifetime as a function of RH cycling amplitude and membrane mechanical properties. The modeling framework consists of three model components: a fuel cell RH distribution model, a hydration/dehydration induced stress model that predicts stress distribution in the membrane, and a damage accrual model that predicts membrane lifetime. Short descriptions of the model components along with overall framework are presented in the paper. The model was used for lifetime prediction of a GORE-SELECT membrane.

  5. Isotopic Details of the Spent Catawba-1 MOX Fuel Rods at ORNL

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Ellis, Ronald James

    The United States Department of Energy funded Shaw/AREVA MOX Services LLC to fabricate four MOX Lead Test Assemblies (LTA) from weapons-grade plutonium. A total of four MOX LTAs (including MX03) were irradiated in the Catawba Nuclear Station (Unit 1) Catawba-1 PWR which operated at a total thermal power of 3411 MWt and had a core with 193 total fuel assemblies. The MOX LTAs were irradiated along with Duke Energy s irradiation of eight Westinghouse Next Generation Fuel (NGF) LEU LTAs (ref.1) and the remaining 181 LEU fuel assemblies. The MX03 LTA was irradiated in the Catawba-1 PWR core (refs.2,3) duringmore » cycles C-16 and C-17. C-16 began on June 5, 2005, and ended on November 11, 2006, after 499 effective full power days (EFPDs). C-17 started on December 29, 2006, (after a shutdown of 48 days) and continued for 485 EFPDs. The MX03 and three other MOX LTAs (and other fuel assemblies) were discharged at the end of C-17 on May 3, 2008. The design of the MOX LTAs was based on the (Framatome ANP, Inc.) Mark-BW/MOX1 17 17 fuel assembly design (refs. 4,5,6) for use in Westinghouse PWRs, but with MOX fuel rods with three Pu loading ranges: the nominal Pu loadings are 4.94 wt%, 3.30 wt%, and 2.40 wt%, respectively, for high, medium, and low Pu content. The Mark-BW/MOX1 (MOX LTA) fuel assembly design is the same as the Advanced Mark-BW fuel assembly design but with the LEU fuel rods replaced by MOX fuel rods (ref. 5). The fabrication of the fuel pellets and fuel rods for the MOX LTAs was performed at the Cadarache facility in France, with the fabrication of the LTAs performed at the MELOX facility, also in France.« less

  6. Development of Polybenzimidazole-Based High-Temperature Membrane and Electrode Assemblies for Stationary and Automotive Applications

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Vogel, John A.

    The program began on August 1, 2003 and ended on July 31, 2007. The goal of the project was to optimize a high-temperature polybenzimidazole (PBI) membrane to meet the performance, durability, and cost targets required for stationary fuel cell applications. These targets were identified in the Fuel Cell section (3.4) of DOE’s Hydrogen, Fuel Cells and Infrastructure Technologies Program Multi-Year Research, Development and Demonstration Plan. A membrane that operates at high temperatures is important to the fuel cell industry because it is insensitive to carbon monoxide (a poison to low-temperature fuel cells), and does not require complex water management strategies.more » Together, these two benefits greatly simplify the fuel cell system. As a result, the high-temperature fuel cell system realizes a cost benefit as the number of components is reduced by nearly 30%. There is also an inherent reliability benefit as components such as humidifiers and pumps for water management are unnecessary. Furthermore, combined heat and power (CHP) systems may be the best solution for a commercial, grid-connected, stationary product that must offer a cost benefit to the end user. For a low-temperature system, the quality of the heat supplied is insufficient to meet consumer needs and comfort requirements, so peak heaters or supplemental boilers are required. The higher operating temperature of PBI technology allows the fuel cell to meet the heat and comfort demand without the additional equipment. Plug Power, working with the Rensselaer Polytechnic Institute (RPI) Polymer Science Laboratory, made significant advances in optimizing the PBI membrane material for operation at temperatures greater than 160oC with a lifetime of 40,000 hours. Supporting hardware such as flow field plates and a novel sealing concept were explored to yield the lower-cost stack assembly and corresponding manufacturing process. Additional work was conducted on acid loss, flow field design and cathode electrode development. Membranes and MEAs were supplied by team member BASF Fuel Cell (formerly PEMEAS), a manufacturer of polymer and fiber. Additional subcontractors Entegris, the University of South Carolina (USC) Fuel Cell Center, and RPI’s Fuel Cell Center conducted activities with regard to stack sealing, acid modeling, and electrode development.« less

  7. In situ metal ion contamination and the effects on proton exchange membrane fuel cell performance

    NASA Astrophysics Data System (ADS)

    Sulek, Mark; Adams, Jim; Kaberline, Steve; Ricketts, Mark; Waldecker, James R.

    Automotive fuel cell technology has made considerable progress, and hydrogen fuel cell vehicles are regarded as a possible long-term solution to reduce carbon dioxide emissions, reduce fossil fuel dependency and increase energy efficiency. Even though great strides have been made, durability is still an issue. One key challenge is controlling MEA contamination. Metal ion contamination within the membrane and the effects on fuel cell performance were investigated. Given the possible benefits of using stainless steel or aluminum for balance-of-plant components or bipolar plates, cations of Al, Fe, Ni and Cr were studied. Membranes were immersed in metal sulfide solutions of varying concentration and then assembled into fuel cell MEAs tested in situ. The ranking of the four transition metals tested in terms of the greatest reduction in fuel cell performance was: Al 3+ ≫ Fe 2+ > Ni 2+, Cr 3+. For iron-contaminated membranes, no change in cell performance was detected until the membrane conductivity loss was greater than approximately 15%.

  8. Kerosene space heaters--combustion technology and kerosene characteristics

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Kubayashi, k.; I Wasaki, N.

    1984-07-01

    This paper describes kerosene combustion technology. Unvented wick-type kerosene space heaters are very popular in Japan because of their economy and convenience. In recent years new vaporized kerosene burners having premixed combustion systems have been developed to solve some of the problems encountered in the older portable type. Some of the features of the new burners are instantaneous ignition, no vaporizing deposit on the burner and a wide range heating capacity. These new kerosene heaters have four major components: an air supply fan, a fuel supply assembly, a burner assembly and a control assembly. These heaters are designed to bemore » highly reliable, have stable combustion characteristics, yield minimum carbon deposit. Finally, they are simple and inexpensive to operate.« less

  9. A preliminary evaluation of the ability of from-reactor casks to geometrically accommodate commercial LWR spent nuclear fuel

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Andress, D.; Joy, D.S.; McLeod, N.B.

    The Department of Energy has sponsored a number of cask design efforts to define several transportation casks to accommodate the various assemblies expected to be accepted by the Federal Waste Management System. At this time, three preliminary cask designs have been selected for the final design--the GA-4 and GA-9 truck casks and the BR-100 rail cask. In total, this assessment indicates that the current Initiative I cask designs can be expected to dimensionally accommodate 100% of the PWR fuel assemblies (other than the extra-long South Texas Fuel) with control elements removed, and >90% of the assemblies having the control elementsmore » as an integral part of the fuel assembly. For BWR assemblies, >99% of the assemblies can be accommodated with fuel channels removed. This paper summarizes preliminary results of one part of that evaluation related to the ability of the From-Reactor Initiative I casks to accommodate the physical and radiological characteristics of the Spent Nuclear Fuel projected to be accepted into the Federal Waste Management System. 3 refs., 5 tabs.« less

  10. Fuel injection assembly for use in turbine engines and method of assembling same

    DOEpatents

    Uhm, Jong Ho; Johnson, Thomas Edward

    2015-03-24

    A fuel injection assembly for use in a turbine engine is provided. The fuel injection assembly includes a plurality of tube assemblies, wherein each of the tube assemblies includes an upstream portion and a downstream portion. Each tube assembly includes a plurality of tubes that extend from the upstream portion to the downstream portion or from the upstream portion through the downstream portion. At least one injection system is coupled to at least one tube assembly of the plurality of tube assemblies. The injection system includes a fluid supply member that extends from a fluid source to the downstream portion of the tube assembly. The fluid supply member includes a first end portion located in the downstream portion of the tube assembly, wherein the first end portion has at least one first opening for channeling fluid through the tube assembly to facilitate reducing a temperature therein.

  11. Sputter-deposited fuel cell membranes and electrodes

    NASA Technical Reports Server (NTRS)

    Narayanan, Sekharipuram R. (Inventor); Jeffries-Nakamura, Barbara (Inventor); Chun, William (Inventor); Ruiz, Ron P. (Inventor); Valdez, Thomas I. (Inventor)

    2001-01-01

    A method for preparing a membrane for use in a fuel cell membrane electrode assembly includes the steps of providing an electrolyte membrane, and sputter-depositing a catalyst onto the electrolyte membrane. The sputter-deposited catalyst may be applied to multiple sides of the electrolyte membrane. A method for forming an electrode for use in a fuel cell membrane electrode assembly includes the steps of obtaining a catalyst, obtaining a backing, and sputter-depositing the catalyst onto the backing. The membranes and electrodes are useful for assembling fuel cells that include an anode electrode, a cathode electrode, a fuel supply, and an electrolyte membrane, wherein the electrolyte membrane includes a sputter-deposited catalyst, and the sputter-deposited catalyst is effective for sustaining a voltage across a membrane electrode assembly in the fuel cell.

  12. Nondestructive Assay Data Integration with the SKB-50 Assemblies - FY16 Update

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Tobin, Stephen Joseph; Fugate, Michael Lynn; Trellue, Holly Renee

    2016-10-28

    A project to research the application of non-destructive assay (NDA) techniques for spent fuel assemblies is underway at the Central Interim Storage Facility for Spent Nuclear Fuel (for which the Swedish acronym is Clab) in Oskarshamn, Sweden. The research goals of this project contain both safeguards and non-safeguards interests. These nondestructive assay (NDA) technologies are designed to strengthen the technical toolkit of safeguard inspectors and others to determine the following technical goals more accurately; Verify initial enrichment, burnup, and cooling time of facility declaration for spent fuel assemblies; Detect replaced or missing pins from a given spent fuel assembly tomore » confirm its integrity; and Estimate plutonium mass and related plutonium and uranium fissile mass parameters in spent fuel assemblies. Estimate heat content, and measure reactivity (multiplication).« less

  13. Integrated main rail, feed rail, and current collector

    DOEpatents

    Petri, Randy J.; Meek, John; Bachta, Robert P.; Marianowski, Leonard G.

    1994-01-01

    A separator plate for a fuel cell comprising an anode current collector, a cathode current collector and a main plate, the main plate disposed between the anode current collector and the cathode current collector. The anode current collector forms a flattened peripheral wet seal structure and manifold wet seal structure on the anode side of the separator plate and the cathode current collector forms a flattened peripheral wet seal structure and manifold wet seal structure on the cathode side of the separator plate. In this manner, the number of components required to manufacture and assemble a fuel cell stack is reduced.

  14. Constitutive and damage material modeling in a high pressure hydrogen environment

    NASA Technical Reports Server (NTRS)

    Russell, D. A.; Fritzemeier, L. G.

    1991-01-01

    Numerous components in reusable space propulsion systems such as the SSME are exposed to high pressure gaseous hydrogen environments. Flow areas and passages in the fuel turbopump, fuel and oxidizer preburners, main combustion chamber, and injector assembly contain high pressure hydrogen either high in purity or as hydrogen rich steam. Accurate constitutive and damage material models applicable to high pressure hydrogen environments are therefore needed for engine design and analysis. Existing constitutive and cyclic crack initiation models were evaluated only for conditions of oxidizing environments. The main objective is to evaluate these models for applicability to high pressure hydrogen environments.

  15. Fabrication of capsule assemblies, phase 3

    NASA Technical Reports Server (NTRS)

    Keeton, A. R.; Stemann, L. G.

    1973-01-01

    Thirteen capsule assemblies were fabricated for evaluation of fuel pin design concepts for a fast spectrum lithium cooled compact space power reactor. These instrumented assemblies were designed for real time test of prototype fuel pins. Uranium mononitride fuel pins were encased in AISI 304L stainless steel capsules. Fabrication procedures were fully qualified by process development and assembly qualification tests. Instrumentation reliability was achieved utilizing specially processed and closely controlled thermocouple hot zone fabrication and by thermal screening tests. Overall capsule reliability was achieved with an all electron beam welded assembly.

  16. Nuclear core and fuel assemblies

    DOEpatents

    Downs, Robert E.

    1981-01-01

    A fast flux nuclear core of a plurality of rodded, open-lattice assemblies having a rod pattern rotated relative to a rod support structure pattern. Elongated fuel rods are oriented on a triangular array and laterally supported by grid structures positioned along the length of the assembly. Initial inter-assembly contact is through strongbacks at the corners of the support pattern and peripheral fuel rods between adjacent assemblies are nested so as to maintain a triangular pitch across a clearance gap between the other portions of adjacent assemblies. The rod pattern is rotated relative to the strongback support pattern by an angle .alpha. equal to sin .sup.-1 (p/2c), where p is the intra-assembly rod pitch and c is the center-to-center spacing among adjacent assemblies.

  17. Fuel assembly for the production of tritium in light water reactors

    DOEpatents

    Cawley, W.E.; Trapp, T.J.

    1983-06-10

    A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

  18. Fuel assembly for the production of tritium in light water reactors

    DOEpatents

    Cawley, William E.; Trapp, Turner J.

    1985-01-01

    A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

  19. Thermal Analysis of a TREAT Fuel Assembly

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Papadias, Dionissios; Wright, Arthur E.

    2014-07-09

    The objective of this study was to explore options as to reduce peak cladding temperatures despite an increase in peak fuel temperatures. A 3D thermal-hydraulic model for a single TREAT fuel assembly was benchmarked to reproduce results obtained with previous thermal models developed for a TREAT HEU fuel assembly. In exercising this model, and variants thereof depending on the scope of analysis, various options were explored to reduce the peak cladding temperatures.

  20. Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly †

    PubMed Central

    De Pauw, Ben; Lamberti, Alfredo; Ertveldt, Julien; Rezayat, Ali; van Tichelen, Katrien; Vanlanduit, Steve; Berghmans, Francis

    2016-01-01

    Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation. PMID:27110782

  1. Fuel cell system including a unit for electrical isolation of a fuel cell stack from a manifold assembly and method therefor

    DOEpatents

    Kelley; Dana A. , Farooque; Mohammad , Davis; Keith

    2007-10-02

    A fuel cell system with improved electrical isolation having a fuel cell stack with a positive potential end and a negative potential, a manifold for use in coupling gases to and from a face of the fuel cell stack, an electrical isolating assembly for electrically isolating the manifold from the stack, and a unit for adjusting an electrical potential of the manifold such as to impede the flow of electrolyte from the stack across the isolating assembly.

  2. Separator assembly for use in spent nuclear fuel shipping cask

    DOEpatents

    Bucholz, James A.

    1983-01-01

    A separator assembly for use in a spent nuclear fuel shipping cask has a honeycomb-type wall structure defining parallel cavities for holding nuclear fuel assemblies. Tubes formed of an effective neutron-absorbing material are embedded in the wall structure around each of the cavities and provide neutron flux traps when filled with water.

  3. Temperature measuring analysis of the nuclear reactor fuel assembly

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Urban, F., E-mail: jozef.bereznai@stuba.sk, E-mail: zdenko.zavodny@stuba.sk; Kučák, L., E-mail: jozef.bereznai@stuba.sk, E-mail: zdenko.zavodny@stuba.sk; Bereznai, J., E-mail: jozef.bereznai@stuba.sk, E-mail: zdenko.zavodny@stuba.sk

    2014-08-06

    Study was based on rapid changes of measured temperature values from the thermocouple in the VVER 440 nuclear reactor fuel assembly. Task was to determine origin of fluctuations of the temperature values by experiments on physical model of the fuel assembly. During an experiment, heated water was circulating in the system and cold water inlet through central tube to record sensitivity of the temperature sensor. Two positions of the sensor was used. First, just above the central tube in the physical model fuel assembly axis and second at the position of the thermocouple in the VVER 440 nuclear reactor fuelmore » assembly. Dependency of the temperature values on time are presented in the diagram form in the paper.« less

  4. Current conducting end plate of fuel cell assembly

    DOEpatents

    Walsh, Michael M.

    1999-01-01

    A fuel cell assembly has a current conducting end plate with a conductive body formed integrally with isolating material. The conductive body has a first surface, a second surface opposite the first surface, and an electrical connector. The first surface has an exposed portion for conducting current between a working section of the fuel cell assembly and the electrical connector. The isolating material is positioned on at least a portion of the second surface. The conductive body can have support passage(s) extending therethrough for receiving structural member(s) of the fuel cell assembly. Isolating material can electrically isolate the conductive body from the structural member(s). The conductive body can have service passage(s) extending therethrough for servicing one or more fluids for the fuel cell assembly. Isolating material can chemically isolate the one or more fluids from the conductive body. The isolating material can also electrically isolate the conductive body from the one or more fluids.

  5. Characterization of thermal-hydraulic and ignition phenomena in prototypic, full-length boiling water reactor spent fuel pool assemblies after a complete loss-of-coolant accident.

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Lindgren, Eric Richard; Durbin, Samuel G

    2007-04-01

    The objective of this project was to provide basic thermal-hydraulic data associated with a SFP complete loss-of-coolant accident. The accident conditions of interest for the SFP were simulated in a full-scale prototypic fashion (electrically-heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate code validation (primarily MELCOR) and reduce questions associated with interpretation of the experimental results. It was necessary to simulate a cluster of assemblies to represent a higher decay (younger) assembly surrounded by older, lower-power assemblies. Specifically, this program providedmore » data and analysis confirming: (1) MELCOR modeling of inter-assembly radiant heat transfer, (2) flow resistance modeling and the natural convective flow induced in a fuel assembly as it heats up in air, (3) the potential for and nature of thermal transient (i.e., Zircaloy fire) propagation, and (4) mitigation strategies concerning fuel assembly management.« less

  6. Chimney for enhancing flow of coolant water in natural circulation boiling water reactor

    DOEpatents

    Oosterkamp, Willem Jan; Marquino, Wayne

    1999-01-05

    A chimney which can be reconfigured or removed during refueling to allow vertical removal of the fuel assemblies. The chimney is designed to be collapsed or dismantled. Collapse or dismantlement of the chimney reduces the volume required for chimney storage during the refueling operation. Alternatively, the chimney has movable parts which allow reconfiguration of its structure. In a first configuration suitable for normal reactor operation, the chimney is radially constricted such that the chimney obstructs vertical removal of the fuel assemblies. In a second configuration suitable for refueling or maintenance of the fuel core, the parts of the chimney which obstruct access to the fuel assemblies are moved radially outward to positions whereat access to the fuel assemblies is not obstructed.

  7. PWR and BWR spent fuel assembly gamma spectra measurements

    NASA Astrophysics Data System (ADS)

    Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

    2016-10-01

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  8. PWR and BWR spent fuel assembly gamma spectra measurements

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

  9. PWR and BWR spent fuel assembly gamma spectra measurements

    DOE PAGES

    Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea; ...

    2016-07-17

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

  10. SMART Structures User's Guide - Version 3.0

    NASA Technical Reports Server (NTRS)

    Spangler, Jan L.

    1996-01-01

    Version 3.0 of the Solid Modeling Aerospace Research Tool (SMART Structures) is used to generate structural models for conceptual and preliminary-level aerospace designs. Features include the generation of structural elements for wings and fuselages, the integration of wing and fuselage structural assemblies, and the integration of fuselage and tail structural assemblies. The highly interactive nature of this software allows the structural engineer to move quickly from a geometry that defines a vehicle's external shape to one that has both external components and internal components which may include ribs, spars, longerons, variable depth ringframes, a floor, a keel, and fuel tanks. The geometry that is output is consistent with FEA requirements and includes integrated wing and empennage carry-through and frame attachments. This report provides a comprehensive description of SMART Structures and how to use it.

  11. Uncertainty analysis on reactivity and discharged inventory for a pressurized water reactor fuel assembly due to {sup 235,238}U nuclear data uncertainties

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Da Cruz, D. F.; Rochman, D.; Koning, A. J.

    2012-07-01

    This paper discusses the uncertainty analysis on reactivity and inventory for a typical PWR fuel element as a result of uncertainties in {sup 235,238}U nuclear data. A typical Westinghouse 3-loop fuel assembly fuelled with UO{sub 2} fuel with 4.8% enrichment has been selected. The Total Monte-Carlo method has been applied using the deterministic transport code DRAGON. This code allows the generation of the few-groups nuclear data libraries by directly using data contained in the nuclear data evaluation files. The nuclear data used in this study is from the JEFF3.1 evaluation, and the nuclear data files for {sup 238}U and {supmore » 235}U (randomized for the generation of the various DRAGON libraries) are taken from the nuclear data library TENDL. The total uncertainty (obtained by randomizing all {sup 238}U and {sup 235}U nuclear data in the ENDF files) on the reactor parameters has been split into different components (different nuclear reaction channels). Results show that the TMC method in combination with a deterministic transport code constitutes a powerful tool for performing uncertainty and sensitivity analysis of reactor physics parameters. (authors)« less

  12. Surrogate fuel assembly multi-axis shaker tests to simulate normal conditions of rail and truck transport

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    McConnell, Paul E.; Koenig, Greg John; Uncapher, William Leonard

    2016-05-01

    This report describes the third set of tests (the “DCLa shaker tests”) of an instrumented surrogate PWR fuel assembly. The purpose of this set of tests was to measure strains and accelerations on Zircaloy-4 fuel rods when the PWR assembly was subjected to rail and truck loadings simulating normal conditions of transport when affixed to a multi-axis shaker. This is the first set of tests of the assembly simulating rail normal conditions of transport.

  13. Surrogate fuel assembly multi-axis shaker tests to simulate normal conditions of rail and truck transport

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    McConnell, Paul E.; Koenig, Greg John; Uncapher, William Leonard

    2016-05-12

    This report describes the third set of tests (the “DCL a shaker tests”) of an instrumented surrogate PWR fuel assembly. The purpose of this set of tests was to measure strains and accelerations on Zircaloy-4 fuel rods when the PWR assembly was subjected to rail and truck loadings simulating normal conditions of transport when affixed to a multi-axis shaker. This is the first set of tests of the assembly simulating rail normal conditions of transport.

  14. Simplified process for leaching precious metals from fuel cell membrane electrode assemblies

    DOEpatents

    Shore, Lawrence [Edison, NJ; Matlin, Ramail [Berkeley Heights, NJ

    2009-12-22

    The membrane electrode assemblies of fuel cells are recycled to recover the catalyst precious metals from the assemblies. The assemblies are cryogenically embrittled and pulverized to form a powder. The pulverized assemblies are then mixed with a surfactant to form a paste which is contacted with an acid solution to leach precious metals from the pulverized membranes.

  15. Phosphoric and electric utility fuel cell technology development

    NASA Astrophysics Data System (ADS)

    Breault, R. D.; Briggs, T. A.; Congdon, J. V.; Gelting, R. L.; Goller, G. J.; Luoma, W. L.; McCloskey, M. W.; Mientek, A. P.; Obrien, J. J.; Randall, S. A.

    1985-01-01

    A subscale cell containing GSB-18, dry mix catalyst has accumulated over 6500 hours with performance 10 mV above E-line at 120 psia and 400 F. Over 150 thick separator plates were molded for use in cooler assemblies. The full-size 10-ft, 460 cell structural work-up is completed. All repeat components for the next 10-ft short stack are formed and processed.

  16. Modeling of Water-Breathing Propulsion Systems Utilizing the Aluminum-Seawater Reaction and Solid-Oxide Fuel Cells

    DTIC Science & Technology

    2011-01-01

    ABSTRACT Title of Document: MODELING OF WATER-BREATHING PROPULSION SYSTEMS UTILIZING THE ALUMINUM-SEAWATER REACTION AND SOLID...Hybrid Aluminum Combustor (HAC): a novel underwater power system based on the exothermic reaction of aluminum with seawater. The system is modeled ...using a NASA-developed framework called Numerical Propulsion System Simulation (NPSS) by assembling thermodynamic models developed for each component

  17. Fuel cell design and assembly

    DOEpatents

    Myerhoff, Alfred

    1984-01-01

    The present invention is directed to a novel bipolar cooling plate, fuel cell design and method of assembly of fuel cells. The bipolar cooling plate used in the fuel cell design and method of assembly has discrete opposite edge and means carried by the plate defining a plurality of channels extending along the surface of the plate toward the opposite edges. At least one edge of the channels terminates short of the edge of the plate defining a recess for receiving a fastener.

  18. Fuel nozzle assembly

    DOEpatents

    Johnson, Thomas Edward [Greer, SC; Ziminsky, Willy Steve [Simpsonville, SC; Lacey, Benjamin Paul [Greer, SC; York, William David [Greer, SC; Stevenson, Christian Xavier [Inman, SC

    2011-08-30

    A fuel nozzle assembly is provided. The assembly includes an outer nozzle body having a first end and a second end and at least one inner nozzle tube having a first end and a second end. One of the nozzle body or nozzle tube includes a fuel plenum and a fuel passage extending therefrom, while the other of the nozzle body or nozzle tube includes a fuel injection hole slidably aligned with the fuel passage to form a fuel flow path therebetween at an interface between the body and the tube. The nozzle body and the nozzle tube are fixed against relative movement at the first ends of the nozzle body and nozzle tube, enabling the fuel flow path to close at the interface due to thermal growth after a flame enters the nozzle tube.

  19. Cell module and fuel conditioner

    NASA Technical Reports Server (NTRS)

    Hoover, D. Q., Jr.

    1980-01-01

    Stack tests indicate that the discrepancies between calculated and measured temperature profiles are due to reactant cross-over and a lower than expected thermal conductivity of cells. Preliminary results indicate that acceptable contact resistance between cooling plane halves can be achieved without the use of paper. The preliminary design of the enclosure, definition of required labor and equipment for manufacturing repeating components, and the assembly procedures for the benchwork design were developed. Fabrication of components for a second 5-cell stack of the MK-2 design and a second 23-cell stack of the MK-1 design was started. The definition of water and fuel for the reforming subsystem was developed along with a preliminary definition of the control system for the subsystem. The construction and shakedown of the differential catalytic reactor was completed and testing of the first catalyst initiated.

  20. Internal reforming fuel cell assembly with simplified fuel feed

    DOEpatents

    Farooque, Mohammad; Novacco, Lawrence J.; Allen, Jeffrey P.

    2001-01-01

    A fuel cell assembly in which fuel cells adapted to internally reform fuel and fuel reformers for reforming fuel are arranged in a fuel cell stack. The fuel inlet ports of the fuel cells and the fuel inlet ports and reformed fuel outlet ports of the fuel reformers are arranged on one face of the fuel cell stack. A manifold sealing encloses this face of the stack and a reformer fuel delivery system is arranged entirely within the region between the manifold and the one face of the stack. The fuel reformer has a foil wrapping and a cover member forming with the foil wrapping an enclosed structure.

  1. Combustor assembly for use in a turbine engine and methods of assembling same

    DOEpatents

    Uhm, Jong Ho; Johnson, Thomas Edward

    2013-05-14

    A fuel nozzle assembly for use with a turbine engine is described herein. The fuel nozzle assembly includes a plurality of fuel nozzles positioned within an air plenum defined by a casing. Each of the plurality of fuel nozzles is coupled to a combustion liner defining a combustion chamber. Each of the plurality of fuel nozzles includes a housing that includes an inner surface that defines a cooling fluid plenum and a fuel plenum therein, and a plurality of mixing tubes extending through the housing. Each of the mixing tubes includes an inner surface defining a flow channel extending between the air plenum and the combustion chamber. At least one mixing tube of the plurality of mixing tubes including at least one cooling fluid aperture for channeling a flow of cooling fluid from the cooling fluid plenum to the flow channel.

  2. Apparatus for in situ determination of burnup, cooling time and fissile content of an irradiated nuclear fuel assembly in a fuel storage pond

    DOEpatents

    Phillips, John R.; Halbig, James K.; Menlove, Howard O.; Klosterbuer, Shirley F.

    1985-01-01

    A detector head for in situ inspection of irradiated nuclear fuel assemblies submerged in a water-filled nuclear fuel storage pond. The detector head includes two parallel arms which extend from a housing and which are spaced apart so as to be positionable on opposite sides of a submerged fuel assembly. Each arm includes an ionization chamber and two fission chambers. One fission chamber in each arm is enclosed in a cadmium shield and the other fission chamber is unshielded. The ratio of the outputs of the shielded and unshielded fission chambers is used to determine the boron content of the pond water. Correcting for the boron content, the neutron flux and gamma ray intensity are then used to verify the declared exposure, cooling time and fissile material content of the irradiated fuel assembly.

  3. Chimney for enhancing flow of coolant water in natural circulation boiling water reactor

    DOEpatents

    Oosterkamp, W.J.; Marquino, W.

    1999-01-05

    A chimney which can be reconfigured or removed during refueling to allow vertical removal of the fuel assemblies is disclosed. The chimney is designed to be collapsed or dismantled. Collapse or dismantlement of the chimney reduces the volume required for chimney storage during the refueling operation. Alternatively, the chimney has movable parts which allow reconfiguration of its structure. In a first configuration suitable for normal reactor operation, the chimney is radially constricted such that the chimney obstructs vertical removal of the fuel assemblies. In a second configuration suitable for refueling or maintenance of the fuel core, the parts of the chimney which obstruct access to the fuel assemblies are moved radially outward to positions whereas access to the fuel assemblies is not obstructed. 11 figs.

  4. Hybrid Gama Emission Tomography (HGET): FY16 Annual Report

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Miller, Erin A.; Smith, Leon E.; Wittman, Richard S.

    2017-02-01

    Current International Atomic Energy Agency (IAEA) methodologies for the verification of fresh low-enriched uranium (LEU) and mixed oxide (MOX) fuel assemblies are volume-averaging methods that lack sensitivity to individual pins. Further, as fresh fuel assemblies become more and more complex (e.g., heavy gadolinium loading, high degrees of axial and radial variation in fissile concentration), the accuracy of current IAEA instruments degrades and measurement time increases. Particularly in light of the fact that no special tooling is required to remove individual pins from modern fuel assemblies, the IAEA needs new capabilities for the verification of unirradiated (i.e., fresh LEU and MOX)more » assemblies to ensure that fissile material has not been diverted. Passive gamma emission tomography has demonstrated potential to provide pin-level verification of spent fuel, but gamma-ray emission rates from unirradiated fuel emissions are significantly lower, precluding purely passive tomography methods. The work presented here introduces the concept of Hybrid Gamma Emission Tomography (HGET) for verification of unirradiated fuels, in which a neutron source is used to actively interrogate the fuel assembly and the resulting gamma-ray emissions are imaged using tomographic methods to provide pin-level verification of fissile material concentration.« less

  5. KSC-2009-5826

    NASA Image and Video Library

    2009-10-24

    CAPE CANAVERAL, Fla. – External tank 134 has arrived in the transfer aisle of the Vehicle Assembly Building, or VAB, at NASA's Kennedy Space Center in Florida. The Pegasus barge, carrying the fuel tank, arrived in Florida after an ocean voyage towed by a solid rocket booster retrieval ship from NASA's Michoud Assembly Facility near New Orleans. ET-134 will be used to launch space shuttle Endeavour on the STS-130 mission to the International Space Station. Launch is targeted for Feb. 4, 2010. For information on the components of the space shuttle and the STS-130 mission, visit http://www.nasa.gov/shuttle. Photo credit: NASA/Jack Pfaller

  6. Fuel-Mediated Transient Clustering of Colloidal Building Blocks.

    PubMed

    van Ravensteijn, Bas G P; Hendriksen, Wouter E; Eelkema, Rienk; van Esch, Jan H; Kegel, Willem K

    2017-07-26

    Fuel-driven assembly operates under the continuous influx of energy and results in superstructures that exist out of equilibrium. Such dissipative processes provide a route toward structures and transient behavior unreachable by conventional equilibrium self-assembly. Although perfected in biological systems like microtubules, this class of assembly is only sparsely used in synthetic or colloidal analogues. Here, we present a novel colloidal system that shows transient clustering driven by a chemical fuel. Addition of fuel causes an increase in hydrophobicity of the building blocks by actively removing surface charges, thereby driving their aggregation. Depletion of fuel causes reappearance of the charged moieties and leads to disassembly of the formed clusters. This reassures that the system returns to its initial, equilibrium state. By taking advantage of the cyclic nature of our system, we show that clustering can be induced several times by simple injection of new fuel. The fuel-mediated assembly of colloidal building blocks presented here opens new avenues to the complex landscape of nonequilibrium colloidal structures, guided by biological design principles.

  7. SOLID GAS SUSPENSION NUCLEAR FUEL ASSEMBLY

    DOEpatents

    Schluderberg, D.C.; Ryon, J.W.

    1962-05-01

    A fuel assembly is designed for use in a gas-suspension cooled nuclear fuel reactor. The coolant fluid is an inert gas such as nitrogen or helium with particles such as carbon suspended therein. The fuel assembly is contained within an elongated pressure vessel extending down into the reactor. The fuel portion is at the lower end of the vessel and is constructed of cylindrical segments through which the coolant passes. Turbulence promotors within the passageways maintain the particles in agitation to increase its ability to transfer heat away from the outer walls. Shielding sections and alternating passageways above the fueled portion limit the escape of radiation out of the top of the vessel. (AEC)

  8. Air breathing direct methanol fuel cell

    DOEpatents

    Ren, Xiaoming; Gottesfeld, Shimshon

    2002-01-01

    An air breathing direct methanol fuel cell is provided with a membrane electrode assembly, a conductive anode assembly that is permeable to air and directly open to atmospheric air, and a conductive cathode assembly that is permeable to methanol and directly contacting a liquid methanol source. Water loss from the cell is minimized by making the conductive cathode assembly hydrophobic and the conductive anode assembly hydrophilic.

  9. Nuclear imaging of the fuel assembly in ignition experimentsa)

    NASA Astrophysics Data System (ADS)

    Grim, G. P.; Guler, N.; Merrill, F. E.; Morgan, G. L.; Danly, C. R.; Volegov, P. L.; Wilde, C. H.; Wilson, D. C.; Clark, D. S.; Hinkel, D. E.; Jones, O. S.; Raman, K. S.; Izumi, N.; Fittinghoff, D. N.; Drury, O. B.; Alger, E. T.; Arnold, P. A.; Ashabranner, R. C.; Atherton, L. J.; Barrios, M. A.; Batha, S.; Bell, P. M.; Benedetti, L. R.; Berger, R. L.; Bernstein, L. A.; Berzins, L. V.; Betti, R.; Bhandarkar, S. D.; Bionta, R. M.; Bleuel, D. L.; Boehly, T. R.; Bond, E. J.; Bowers, M. W.; Bradley, D. K.; Brunton, G. K.; Buckles, R. A.; Burkhart, S. C.; Burr, R. F.; Caggiano, J. A.; Callahan, D. A.; Casey, D. T.; Castro, C.; Celliers, P. M.; Cerjan, C. J.; Chandler, G. A.; Choate, C.; Cohen, S. J.; Collins, G. W.; Cooper, G. W.; Cox, J. R.; Cradick, J. R.; Datte, P. S.; Dewald, E. L.; Di Nicola, P.; Di Nicola, J. M.; Divol, L.; Dixit, S. N.; Dylla-Spears, R.; Dzenitis, E. G.; Eckart, M. J.; Eder, D. C.; Edgell, D. H.; Edwards, M. J.; Eggert, J. H.; Ehrlich, R. B.; Erbert, G. V.; Fair, J.; Farley, D. R.; Felker, B.; Fortner, R. J.; Frenje, J. A.; Frieders, G.; Friedrich, S.; Gatu-Johnson, M.; Gibson, C. R.; Giraldez, E.; Glebov, V. Y.; Glenn, S. M.; Glenzer, S. H.; Gururangan, G.; Haan, S. W.; Hahn, K. D.; Hammel, B. A.; Hamza, A. V.; Hartouni, E. P.; Hatarik, R.; Hatchett, S. P.; Haynam, C.; Hermann, M. R.; Herrmann, H. W.; Hicks, D. G.; Holder, J. P.; Holunga, D. M.; Horner, J. B.; Hsing, W. W.; Huang, H.; Jackson, M. C.; Jancaitis, K. S.; Kalantar, D. H.; Kauffman, R. L.; Kauffman, M. I.; Khan, S. F.; Kilkenny, J. D.; Kimbrough, J. R.; Kirkwood, R.; Kline, J. L.; Knauer, J. P.; Knittel, K. M.; Koch, J. A.; Kohut, T. R.; Kozioziemski, B. J.; Krauter, K.; Krauter, G. W.; Kritcher, A. L.; Kroll, J.; Kyrala, G. A.; Fortune, K. N. La; LaCaille, G.; Lagin, L. J.; Land, T. A.; Landen, O. L.; Larson, D. W.; Latray, D. A.; Leeper, R. J.; Lewis, T. L.; LePape, S.; Lindl, J. D.; Lowe-Webb, R. R.; Ma, T.; MacGowan, B. J.; MacKinnon, A. J.; MacPhee, A. G.; Malone, R. M.; Malsbury, T. N.; Mapoles, E.; Marshall, C. D.; Mathisen, D. G.; McKenty, P.; McNaney, J. M.; Meezan, N. B.; Michel, P.; Milovich, J. L.; Moody, J. D.; Moore, A. S.; Moran, M. J.; Moreno, K.; Moses, E. I.; Munro, D. H.; Nathan, B. R.; Nelson, A. J.; Nikroo, A.; Olson, R. E.; Orth, C.; Pak, A. E.; Palma, E. S.; Parham, T. G.; Patel, P. K.; Patterson, R. W.; Petrasso, R. D.; Prasad, R.; Ralph, J. E.; Regan, S. P.; Rinderknecht, H.; Robey, H. F.; Ross, G. F.; Ruiz, C. L.; Séguin, F. H.; Salmonson, J. D.; Sangster, T. C.; Sater, J. D.; Saunders, R. L.; Schneider, M. B.; Schneider, D. H.; Shaw, M. J.; Simanovskaia, N.; Spears, B. K.; Springer, P. T.; Stoeckl, C.; Stoeffl, W.; Suter, L. J.; Thomas, C. A.; Tommasini, R.; Town, R. P.; Traille, A. J.; Wonterghem, B. Van; Wallace, R. J.; Weaver, S.; Weber, S. V.; Wegner, P. J.; Whitman, P. K.; Widmann, K.; Widmayer, C. C.; Wood, R. D.; Young, B. K.; Zacharias, R. A.; Zylstra, A.

    2013-05-01

    First results from the analysis of neutron image data collected on implosions of cryogenically layered deuterium-tritium capsules during the 2011-2012 National Ignition Campaign are reported. The data span a variety of experimental designs aimed at increasing the stagnation pressure of the central hotspot and areal density of the surrounding fuel assembly. Images of neutrons produced by deuterium-tritium fusion reactions in the hotspot are presented, as well as images of neutrons that scatter in the surrounding dense fuel assembly. The image data are compared with 1D and 2D model predictions, and consistency checked using other diagnostic data. The results indicate that the size of the fusing hotspot is consistent with the model predictions, as well as other imaging data, while the overall size of the fuel assembly, inferred from the scattered neutron images, is systematically smaller than models' prediction. Preliminary studies indicate these differences are consistent with a significant fraction (20%-25%) of the initial deuterium-tritium fuel mass outside the compact fuel assembly, due either to low mode mass asymmetry or high mode 3D mix effects at the ablator-ice interface.

  10. AHTR Mechanical, Structural, And Neutronic Preconceptual Design

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Varma, Venugopal Koikal; Holcomb, David Eugene; Peretz, Fred J

    2012-10-01

    This report provides an overview of the mechanical, structural, and neutronic aspects of the Advanced High Temperature Reactor (AHTR) design concept. The AHTR is a design concept for a large output Fluoride salt cooled High-temperature Reactor (FHR) that is being developed to enable evaluation of the technology hurdles remaining to be overcome prior to FHRs becoming a commercial reactor class. This report documents the incremental AHTR design maturation performed over the past year and is focused on advancing the design concept to a level of a functional, self-consistent system. The AHTR employs plate type coated particle fuel assemblies with rapid,more » off-line refueling. Neutronic analysis of the core has confirmed the viability of a 6-month 2-batch cycle with 9 weight-percent enriched uranium fuel. Refueling is intended to be performed automatically under visual guidance using dedicated robotic manipulators. The present design intent is for used fuel to be stored inside of containment for at least 6 months and then transferred to local dry wells for intermediate term, on-site storage. The mechanical and structural concept development effort has included an emphasis on transportation and constructability to minimize construction costs and schedule. The design intent is that all components be factory fabricated into rail transportable modules that are assembled into subsystems at an on-site workshop prior to being lifted into position using a heavy-lift crane in an open-top style construction. While detailed accident identification and response sequence analysis has yet to be performed, the design concept incorporates multiple levels of radioactive material containment including fully passive responses to all identified design basis or non-very-low frequency beyond design basis accidents. Key building design elements include: 1) below grade siting to minimize vulnerability to aircraft impact, 2) multiple natural circulation decay heat rejection chimneys, 3) seismic base isolation, and 4) decay heat powered back-up electricity generation. The report provides a preconceptual design of the manipulators, the fuel transfer system, and the salt transfer loops. The mechanical handling of the fuel and how it is accomplished without instrumentation inside the salt is described within the report. All drives for the manipulators reside outside the reactor top flange. The design has also taken into account the transportability of major components and how they will be assembled on site« less

  11. U.S. Commercial Spent Nuclear Fuel Assembly Characteristics - 1968-2013

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Hu, Jianwei; Peterson, Joshua L.; Gauld, Ian C.

    2016-09-01

    Activities related to management of spent nuclear fuel (SNF) are increasing in the US and many other countries. Over 240,000 SNF assemblies have been discharged from US commercial reactors since the late 1960s. The enrichment and burnup of SNF have changed significantly over the past 40 years, and fuel assembly designs have also evolved. Understanding the general characteristics of SNF helps regulators and other stakeholders form overall strategies towards the final disposal of US SNF. This report documents a survey of all US commercial SNF assemblies in the GC-859 database and provides reference SNF source terms (e.g., nuclide inventories, decaymore » heat, and neutron/photon emission) at various cooling times up to 200 years after fuel discharge. This study reviews the distribution and evolution of fuel parameters of all SNF assemblies discharged over the past 40 years. Assemblies were categorized into three groups based on discharge year, and the median burnups and enrichments of each group were used to establish representative cases. An extended burnup case was created for boiling water reactor (BWR) fuels, and another was created for the pressurized water reactor (PWR) fuels. Two additional cases were developed to represent the eight mixed oxide (MOX) fuel assemblies in the database. Burnup calculations were performed for each representative case. Realistic parameters for fuel design and operations were used to model the SNF and to provide reference fuel characteristics representative of the current inventory. Burnup calculations were performed using the ORIGEN code, which is part of the SCALE nuclear modeling and simulation code system. Results include total activity, decay heat, photon emission, neutron flux, gamma heat, and plutonium content, as well as concentrations for 115 significant nuclides. These quantities are important in the design, regulation, and operations of SNF storage, transportation, and disposal systems.« less

  12. A User’s Guide to the PLTEMP/ANL Code

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Olson, A. P.; Kalimullah, M.; Feldman, E. E.

    2016-07-25

    PLTEMP/ANL V4.2 is a program that obtains a steady-state flow and temperature solution for a nuclear reactor core, or for a single fuel assembly. It is based on an evolutionary sequence of codes originally used for plate temperatures, hence “PLTEMP”, developed at Argonne National Laboratory over several decades. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each withmore » its own axial power shape. The temperature solution is effectively 2-dimensional. It begins with a one-dimensional solution across all coolant channels and fuel plates or tubes within a given fuel assembly, at the entrance to the assembly. The temperature solution is repeated for each axial node along the length of the fuel assembly. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available with which to determine safety margins such as onset-of-nucleate boiling ratio(ONBR), departure from nucleate boiling ratio (DNBR), and onset of flow instability ratio (OFIR). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability is available (Appendix XII) to greatly reduce the reactor analyst’s time.« less

  13. 75 FR 901 - Airworthiness Directives; Honeywell International Inc. ALF502 Series and LF507 Series Turbofan...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-01-07

    .... ALF502 series and LF507 series turbofan engines with certain fuel manifold assemblies installed. That AD... of certain part number (P/N) fuel manifold assemblies for cracks, and replacement of cracked fuel manifolds with serviceable manifolds. This AD continues to require inspecting those fuel manifolds for...

  14. 77 FR 72200 - Airworthiness Directives; The Boeing Company Airplanes

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-12-05

    ... correctly on the engine fuel feed manifold couplings. This AD also requires inspecting the assembly of the engine fuel feed manifold rigid and full flexible couplings. This AD was prompted by reports of fuel leaks due to improperly assembled engine fuel feed manifold couplings. We are issuing this AD to detect...

  15. Fuel cell cooler-humidifier plate

    DOEpatents

    Vitale, Nicholas G.; Jones, Daniel O.

    2000-01-01

    A cooler-humidifier plate for use in a proton exchange membrane (PEM) fuel cell stack assembly is provided. The cooler-humidifier plate combines functions of cooling and humidification within the fuel cell stack assembly, thereby providing a more compact structure, simpler manifolding, and reduced reject heat from the fuel cell. Coolant on the cooler side of the plate removes heat generated within the fuel cell assembly. Heat is also removed by the humidifier side of the plate for use in evaporating the humidification water. On the humidifier side of the plate, evaporating water humidifies reactant gas flowing over a moistened wick. After exiting the humidifier side of the plate, humidified reactant gas provides needed moisture to the proton exchange membranes used in the fuel cell stack assembly. The invention also provides a fuel cell plate that maximizes structural support within the fuel cell by ensuring that the ribs that form the boundaries of channels on one side of the plate have ends at locations that substantially correspond to the locations of ribs on the opposite side of the plate.

  16. Automatically closing swing gate closure assembly

    DOEpatents

    Chang, Shih-Chih; Schuck, William J.; Gilmore, Richard F.

    1988-01-01

    A swing gate closure assembly for nuclear reactor tipoff assembly wherein the swing gate is cammed open by a fuel element or spacer but is reliably closed at a desired closing rate primarily by hydraulic forces in the absence of a fuel charge.

  17. End plate assembly having a two-phase fluid-filled bladder and method for compressing a fuel cell stack

    DOEpatents

    Carlstrom, Jr., Charles M.

    2001-01-01

    An end plate assembly is disclosed for use in a fuel cell assembly in which the end plate assembly includes a housing having a cavity, and a bladder receivable in the cavity and engageable with the fuel cell stack. The bladder includes a two-phase fluid having a liquid portion and a vapor portion. Desirably, the two-phase fluid has a vapor pressure between about 100 psi and about 600 psi at a temperature between about 70 degrees C. to about 110 degrees C.

  18. FY 2016 Status Report: Documentation of All CIRFT Data including Hydride Reorientation Tests (Draft M2)

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Wang, Jy-An John; Wang, Hong; Jiang, Hao

    The first portion of this report provides a detailed description of fiscal year (FY) 2015 test result corrections and analysis updates based on FY 2016 updates to the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) program methodology, which is used to evaluate the vibration integrity of spent nuclear fuel (SNF) under normal conditions of transport (NCT). The CIRFT consists of a U-frame test setup and a real-time curvature measurement method. The three-component U-frame setup of the CIRFT has two rigid arms and linkages connecting to a universal testing machine. The curvature SNF rod bending is obtained through a three-point deflection measurementmore » method. Three linear variable differential transformers (LVDTs) are clamped to the side connecting plates of the U-frame and used to capture deformation of the rod. The second portion of this report provides the latest CIRFT data, including data for the hydride reorientation test. The variations in fatigue life are provided in terms of moment, equivalent stress, curvature, and equivalent strain for the tested SNFs. The equivalent stress plot collapsed the data points from all of the SNF samples into a single zone. A detailed examination revealed that, at the same stress level, fatigue lives display a descending order as follows: H. B. Robinson Nuclear Power Station (HBR), LMK, and mixed uranium-plutonium oxide (MOX). Just looking at the strain, LMK fuel has a slightly longer fatigue life than HBR fuel, but the difference is subtle. The third portion of this report provides finite element analysis (FEA) dynamic deformation simulation of SNF assemblies . In a horizontal layout under NCT, the fuel assembly’s skeleton, which is formed by guide tubes and spacer grids, is the primary load bearing apparatus carrying and transferring vibration loads within an SNF assembly. These vibration loads include interaction forces between the SNF assembly and the canister basket walls. Therefore, the integrity of the guide tubes and spacer grids critically affects the vibration intensity of the fuel assembly during transport and must be considered when developing the multipurpose purpose canister (MPC) design for safe SNF transport.« less

  19. Differential Die-Away Instrument: Report on Fuel Assembly Mock-up Measurements with Neutron Generator

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Goodsell, Alison Victoria; Swinhoe, Martyn Thomas; Henzl, Vladimir

    2014-09-18

    Fresh fuel experiments for the differential die-away (DDA) project were performed using a DT neutron generator, a 15x15 PWR fuel assembly, and nine 3He detectors in a water tank inside of a shielded cell at Los Alamos National Laboratory (LANL). Eight different fuel enrichments were created using low enriched (LEU) and depleted uranium (DU) dioxide fuel rods. A list-mode data acquisition system recorded the time-dependent signal and analysis of the DDA signal die-away time was performed. The die-away time depended on the amount of fissile material in the fuel assembly and the position of the detector. These experiments were performedmore » in support of the spent nuclear fuel Next Generation Safeguards Initiative DDA project. Lessons learned from the fresh fuel DDA instrument experiments and simulations will provide useful information to the spent fuel project.« less

  20. On the development of radiation tolerant surveillance camera from consumer-grade components

    NASA Astrophysics Data System (ADS)

    Klemen, Ambrožič; Luka, Snoj; Lars, Öhlin; Jan, Gunnarsson; Niklas, Barringer

    2017-09-01

    In this paper an overview on the process of designing a radiation tolerant surveillance camera from consumer grade components and commercially available particle shielding materials is given. This involves utilization of Monte-Carlo particle transport code MCNP6 and ENDF/B-VII.0 nuclear data libraries, as well as testing the physical electrical systems against γ radiation, utilizing JSI TRIGA mk. II fuel elements as a γ-ray sources. A new, aluminum, 20 cm × 20 cm × 30 cm irradiation facility with electrical power and signal wire guide-tube to the reactor platform, was designed and constructed and used for irradiation of large electronic and optical components assemblies with activated fuel elements. Electronic components to be used in the camera were tested against γ-radiation in an independent manner, to determine their radiation tolerance. Several camera designs were proposed and simulated using MCNP, to determine incident particle and dose attenuation factors. Data obtained from the measurements and MCNP simulations will be used to finalize the design of 3 surveillance camera models, with different radiation tolerances.

  1. Dependency of the Reynolds number on the water flow through the perforated tube

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Závodný, Zdenko, E-mail: zdenko.zavodny@stuba.sk; Bereznai, Jozef, E-mail: jozef.bereznai@stuba.sk; Urban, František

    Safe and effective loading of nuclear reactor fuel assemblies demands qualitative and quantitative analysis of the relationship between the coolant temperature in the fuel assembly outlet, measured by the thermocouple, and the mean coolant temperature profile in the thermocouple plane position. It is not possible to perform the analysis directly in the reactor, so it is carried out using measurements on the physical model, and the CFD fuel assembly coolant flow models. The CFD models have to be verified and validated in line with the temperature and velocity profile obtained from the measurements of the cooling water flowing in themore » physical model of the fuel assembly. Simplified physical model with perforated central tube and its validated CFD model serve to design of the second physical model of the fuel assembly of the nuclear reactor VVER 440. Physical model will be manufactured and installed in the laboratory of the Institute of Energy Machines, Faculty of Mechanical Engineering of the Slovak University of Technology in Bratislava.« less

  2. Integrated main rail, feed rail, and current collector

    DOEpatents

    Petri, R.J.; Meek, J.; Bachta, R.P.; Marianowski, L.G.

    1994-11-08

    A separator plate is described for a fuel cell comprising an anode current collector, a cathode current collector and a main plate, the main plate disposed between the anode current collector and the cathode current collector. The anode current collector forms a flattened peripheral wet seal structure and manifold wet seal structure on the anode side of the separator plate and the cathode current collector forms a flattened peripheral wet seal structure and manifold wet seal structure on the cathode side of the separator plate. In this manner, the number of components required to manufacture and assemble a fuel cell stack is reduced. 9 figs.

  3. Method of preparing thin porous sheets of ceramic material

    DOEpatents

    Swarr, Thomas E.; Nickols, Richard C.; Krasij, Myron

    1987-03-24

    A method of forming thin porous sheets of ceramic material for use as electrodes or other components in a molten carbonate fuel cell is disclosed. The method involves spray drying a slurry of fine ceramic particles in liquid carrier to produce generally spherical agglomerates of high porosity and a rough surface texture. The ceramic particles may include the electrode catalyst and the agglomerates can be calcined to improve mechanical strength. After slurrying with suitable volatile material and binder tape casting is used to form sheets that are sufficiently strong for further processing and handling in the assembly of a high temperature fuel cell.

  4. Method of preparing thin porous sheets of ceramic material

    DOEpatents

    Swarr, T.E.; Nickols, R.C.; Krasij, M.

    1984-05-23

    A method of forming thin porous sheets of ceramic material for use as electrodes or other components in a molten carbonate fuel cell is disclosed. The method involves spray drying a slurry of fine ceramic particles in liquid carrier to produce generally spherical agglomerates of high porosity and a rough surface texture. The ceramic particles may include the electrode catalyst and the agglomerates can be calcined to improve mechanical strength. After slurrying with suitable volatile material and binder tape casting is used to form sheets that are sufficiently strong for further processing and handling in the assembly of a high temperature fuel cell.

  5. Miniature ceramic fuel cell

    DOEpatents

    Lessing, Paul A.; Zuppero, Anthony C.

    1997-06-24

    A miniature power source assembly capable of providing portable electricity is provided. A preferred embodiment of the power source assembly employing a fuel tank, fuel pump and control, air pump, heat management system, power chamber, power conditioning and power storage. The power chamber utilizes a ceramic fuel cell to produce the electricity. Incoming hydro carbon fuel is automatically reformed within the power chamber. Electrochemical combustion of hydrogen then produces electricity.

  6. 78 FR 76328 - Application for Renewal of Special Nuclear Material License SNM-2014 From Tennessee Valley...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-12-17

    ... SNM in the form of fully-assembled fuel assemblies that would later form the initial reactor core of WBN2. The SNM in the fuel assemblies is enriched up to 5% in the isotope U-235. The fresh fuel... received the initial core for WBN2. The NRC has not yet issued the OL for the Unit 2 reactor. The...

  7. Impact of Reactor Operating Parameters on Cask Reactivity in BWR Burnup Credit

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Ilas, Germina; Betzler, Benjamin R; Ade, Brian J

    This paper discusses the effect of reactor operating parameters used in fuel depletion calculations on spent fuel cask reactivity, with relevance for boiling-water reactor (BWR) burnup credit (BUC) applications. Assessments that used generic BWR fuel assembly and spent fuel cask configurations are presented. The considered operating parameters, which were independently varied in the depletion simulations for the assembly, included fuel temperature, bypass water density, specific power, and operating history. Different operating history scenarios were considered for the assembly depletion to determine the effect of relative power distribution during the irradiation cycles, as well as the downtime between cycles. Depletion, decay,more » and criticality simulations were performed using computer codes and associated nuclear data within the SCALE code system. Results quantifying the dependence of cask reactivity on the assembly depletion parameters are presented herein.« less

  8. Fresh Fuel Measurements With the Differential Die-Away Self-Interrogation Instrument

    NASA Astrophysics Data System (ADS)

    Trahan, Alexis C.; Belian, Anthony P.; Swinhoe, Martyn T.; Menlove, Howard O.; Flaska, Marek; Pozzi, Sara A.

    2017-07-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) Project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: 1) verify the initial enrichment, burnup, and cooling time of facility declaration; 2) detect the diversion or replacement of pins; 3) estimate the plutonium mass; 4) estimate decay heat; and 5) determine the reactivity of spent fuel assemblies. The differential die-away self-interrogation (DDSI) instrument is one instrument that was assessed for years regarding its feasibility for robust, timely verification of spent fuel assemblies. The instrument was recently built and was tested using fresh fuel assemblies in a variety of configurations, including varying enrichment, neutron absorber content, and symmetry. The early die-away method, a multiplication determination method developed in simulation space, was successfully tested on the fresh fuel assembly data and determined multiplication with a root-mean-square (RMS) error of 2.9%. The experimental results were compared with MCNP simulations of the instrument as well. Low multiplication assemblies had agreement with an average RMS error of 0.2% in the singles count rate (i.e., total neutrons detected per second) and 3.4% in the doubles count rates (i.e., neutrons detected in coincidence per second). High-multiplication assemblies had agreement with an average RMS error of 4.1% in the singles and 13.3% in the doubles count rates.

  9. Fresh Fuel Measurements With the Differential Die-Away Self-Interrogation Instrument

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Trahan, Alexis C.; Belian, Anthony P.; Swinhoe, Martyn T.

    The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) Project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. Thus the NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: 1) verify the initial enrichment, burnup, and cooling time of facility declaration; 2) detect the diversion or replacement of pins; 3) estimate the plutonium mass; 4) estimate decay heat; and 5) determine the reactivity of spent fuel assemblies. The differential die-away self-interrogation (DDSI) instrument is one instrumentmore » that was assessed for years regarding its feasibility for robust, timely verification of spent fuel assemblies. The instrument was recently built and was tested using fresh fuel assemblies in a variety of configurations, including varying enrichment, neutron absorber content, and symmetry. The early die-away method, a multiplication determination method developed in simulation space, was successfully tested on the fresh fuel assembly data and determined multiplication with a root-mean-square (RMS) error of 2.9%. The experimental results were compared with MCNP simulations of the instrument as well. Low multiplication assemblies had agreement with an average RMS error of 0.2% in the singles count rate (i.e., total neutrons detected per second) and 3.4% in the doubles count rates (i.e., neutrons detected in coincidence per second). High-multiplication assemblies had agreement with an average RMS error of 4.1% in the singles and 13.3% in the doubles count rates.« less

  10. Fresh Fuel Measurements With the Differential Die-Away Self-Interrogation Instrument

    DOE PAGES

    Trahan, Alexis C.; Belian, Anthony P.; Swinhoe, Martyn T.; ...

    2017-01-05

    The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) Project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. Thus the NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: 1) verify the initial enrichment, burnup, and cooling time of facility declaration; 2) detect the diversion or replacement of pins; 3) estimate the plutonium mass; 4) estimate decay heat; and 5) determine the reactivity of spent fuel assemblies. The differential die-away self-interrogation (DDSI) instrument is one instrumentmore » that was assessed for years regarding its feasibility for robust, timely verification of spent fuel assemblies. The instrument was recently built and was tested using fresh fuel assemblies in a variety of configurations, including varying enrichment, neutron absorber content, and symmetry. The early die-away method, a multiplication determination method developed in simulation space, was successfully tested on the fresh fuel assembly data and determined multiplication with a root-mean-square (RMS) error of 2.9%. The experimental results were compared with MCNP simulations of the instrument as well. Low multiplication assemblies had agreement with an average RMS error of 0.2% in the singles count rate (i.e., total neutrons detected per second) and 3.4% in the doubles count rates (i.e., neutrons detected in coincidence per second). High-multiplication assemblies had agreement with an average RMS error of 4.1% in the singles and 13.3% in the doubles count rates.« less

  11. Hierarchical Inorganic Assemblies for Artificial Photosynthesis

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Kim, Wooyul; Edri, Eran; Frei, Heinz

    Artificial photosynthesis is an attractive approach for renewable fuel generation because it offers the prospect of a technology suitable for deployment on highly abundant, non-arable land. Recent leaps forward in the development of efficient and durable light absorbers and catalysts for oxygen evolution and the growing attention to catalysts for carbon dioxide activation brings into focus the tasks of hierarchically integrating the components into assemblies for closing of the photosynthetic cycle. A particular challenge is the efficient coupling of the multi-electron processes of CO 2 reduction and H 2O oxidation. Among the most important requirements for a complete integrated systemmore » are catalytic rates that match the solar flux, efficient charge transport between the various components, and scalability of the photosynthetic assembly on the unprecedented scale of terawatts in order to have impact on fuel consumption. To address these challenges, we have developed in this paper a heterogeneous inorganic materials approach with molecularly precise control of light absorption and charge transport pathways. Oxo-bridged heterobinuclear units with metal-to-metal charge-transfer transitions absorbing deep in the visible act as single photon, single charge transfer pumps for driving multi-electron catalysts. A photodeposition method has been introduced for the spatially directed assembly of nanoparticle catalysts for selective coupling to the donor or acceptor metal of the light absorber. For CO 2 reduction, a Cu oxide cluster is coupled to the Zr center of a ZrOCo light absorber, while coupling of an Ir nanoparticle catalyst for water oxidation to the Co donor affords closing of the photosynthetic cycle of CO 2 conversion by H 2O to CO and O 2. Optical, vibrational, and X-ray spectroscopy provide detailed structural knowledge of the polynuclear assemblies. Time resolved visible and rapid-scan FT-IR studies reveal charge transfer mechanisms and transient surface intermediates under photocatalytic conditions for guiding performance improvements. Separation of the water oxidation and carbon dioxide reduction half reactions by a membrane is essential for efficient photoreduction of CO 2 by H 2O to liquid fuel products. A concept of a macroscale artificial photosystem consisting of arrays of Co oxide–silica core–shell nanotubes is introduced in which each tube operates as a complete, independent photosynthetic unit with built-in membrane separation. The ultrathin amorphous silica shell with embedded molecular wires functions as a proton conducting, molecule impermeable membrane. Photoelectrochemical and transient optical measurements confirm tight control of charge transport through the membrane by the orbital energetics of the wire molecules. Finally, hierarchical arrangement of the components is accomplished by a combination of photodeposition, controlled anchoring, and atomic layer deposition methods.« less

  12. METHOD AND MEANS FOR SUPPORTING REACTOR FUEL CONTAINERS IN AN ASSEMBLY

    DOEpatents

    Currier, E.L. Jr.; Nicklas, J.H.; Coombs, C.A.

    1962-12-11

    This patent relates to means for supporting fuelcontaining tubes in an assembly which include grid means at either end of the fuel element assembly antl improved grid means intermediate of the ends to provide support against lateral displacement. (AEC)

  13. Solid oxide fuel cell generator with removable modular fuel cell stack configurations

    DOEpatents

    Gillett, J.E.; Dederer, J.T.; Zafred, P.R.; Collie, J.C.

    1998-04-21

    A high temperature solid oxide fuel cell generator produces electrical power from oxidation of hydrocarbon fuel gases such as natural gas, or conditioned fuel gases, such as carbon monoxide or hydrogen, with oxidant gases, such as air or oxygen. This electrochemical reaction occurs in a plurality of electrically connected solid oxide fuel cells bundled and arrayed in a unitary modular fuel cell stack disposed in a compartment in the generator container. The use of a unitary modular fuel cell stack in a generator is similar in concept to that of a removable battery. The fuel cell stack is provided in a pre-assembled self-supporting configuration where the fuel cells are mounted to a common structural base having surrounding side walls defining a chamber. Associated generator equipment may also be mounted to the fuel cell stack configuration to be integral therewith, such as a fuel and oxidant supply and distribution systems, fuel reformation systems, fuel cell support systems, combustion, exhaust and spent fuel recirculation systems, and the like. The pre-assembled self-supporting fuel cell stack arrangement allows for easier assembly, installation, maintenance, better structural support and longer life of the fuel cells contained in the fuel cell stack. 8 figs.

  14. Solid oxide fuel cell generator with removable modular fuel cell stack configurations

    DOEpatents

    Gillett, James E.; Dederer, Jeffrey T.; Zafred, Paolo R.; Collie, Jeffrey C.

    1998-01-01

    A high temperature solid oxide fuel cell generator produces electrical power from oxidation of hydrocarbon fuel gases such as natural gas, or conditioned fuel gases, such as carbon monoxide or hydrogen, with oxidant gases, such as air or oxygen. This electrochemical reaction occurs in a plurality of electrically connected solid oxide fuel cells bundled and arrayed in a unitary modular fuel cell stack disposed in a compartment in the generator container. The use of a unitary modular fuel cell stack in a generator is similar in concept to that of a removable battery. The fuel cell stack is provided in a pre-assembled self-supporting configuration where the fuel cells are mounted to a common structural base having surrounding side walls defining a chamber. Associated generator equipment may also be mounted to the fuel cell stack configuration to be integral therewith, such as a fuel and oxidant supply and distribution systems, fuel reformation systems, fuel cell support systems, combustion, exhaust and spent fuel recirculation systems, and the like. The pre-assembled self-supporting fuel cell stack arrangement allows for easier assembly, installation, maintenance, better structural support and longer life of the fuel cells contained in the fuel cell stack.

  15. Small, modular, low-cost coal-fired power plants for the international market

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Zauderer, B.; Frain, B.; Borck, B.

    1997-12-31

    This paper presents recent operating results of Coal Tech`s second generation, air cooled, slagging coal combustor, and its application to power plants in the 1 to 20 MW range. This 20 MMBtu/hour combustor was installed in a new demonstration plant in Philadelphia, PA in 1995. It contains the combustion components of a 1 MWe coal fired power plant, a 17,500 lb/hour steam boiler, coal storage and feed components, and stack gas cleanup components. The plant`s design incorporates improvements resulting from 2,000 hours of testing between 1987 and 1993 on a first generation, commercial scale, air cooled combustor of equal thermalmore » rating. Since operations began in early 1996, a total of 51 days of testing have been successfully completed. Major results include durability of the combustor`s refractory wall, excellent combustion with high ash concentration in the fuel, removal of 95% to 100% of the slag in the combustor, very little ash deposition in the boiler, major reduction of in-plant parasitic power, and simplified power system control through the use of modular designs of sub-systems and computer control. Rapid fuel switching between oil, gas, and coal and turndown of up to a factor of three was accomplished. All these features have been incorporated in advanced coal fired plant designs in the 1 to 20 MWe range. Incremental capital costs are only $100 to $200/kW higher than comparable rated gas or oil fired steam generating systems. Most of its components and subsystems can be factory assembled for very rapid field installation. The low capital, low operating costs, fuel flexibility, and compatibility with very high ash fuels, make this power system very attractive in regions of the world having domestic supplies of these fuels.« less

  16. ORIGAMI Automator Primer. Automated ORIGEN Source Terms and Spent Fuel Storage Pool Analysis

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Wieselquist, William A.; Thompson, Adam B.; Bowman, Stephen M.

    2016-04-01

    Source terms and spent nuclear fuel (SNF) storage pool decay heat load analyses for operating nuclear power plants require a large number of Oak Ridge Isotope Generation and Depletion (ORIGEN) calculations. SNF source term calculations also require a significant amount of bookkeeping to track quantities such as core and assembly operating histories, spent fuel pool (SFP) residence times, heavy metal masses, and enrichments. The ORIGEN Assembly Isotopics (ORIGAMI) module in the SCALE code system provides a simple scheme for entering these data. However, given the large scope of the analysis, extensive scripting is necessary to convert formats and process datamore » to create thousands of ORIGAMI input files (one per assembly) and to process the results into formats readily usable by follow-on analysis tools. This primer describes a project within the SCALE Fulcrum graphical user interface (GUI) called ORIGAMI Automator that was developed to automate the scripting and bookkeeping in large-scale source term analyses. The ORIGAMI Automator enables the analyst to (1) easily create, view, and edit the reactor site and assembly information, (2) automatically create and run ORIGAMI inputs, and (3) analyze the results from ORIGAMI. ORIGAMI Automator uses the standard ORIGEN binary concentrations files produced by ORIGAMI, with concentrations available at all time points in each assembly’s life. The GUI plots results such as mass, concentration, activity, and decay heat using a powerful new ORIGEN Post-Processing Utility for SCALE (OPUS) GUI component. This document includes a description and user guide for the GUI, a step-by-step tutorial for a simplified scenario, and appendices that document the file structures used.« less

  17. 96. SEED 1 FUEL ASSEMBLY FROM LOCATION L9 BEING REMOVED ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    96. SEED 1 FUEL ASSEMBLY FROM LOCATION L-9 BEING REMOVED FROM REACTOR VESSEL BY MEANS OF FUEL EXTRACTION CRANE, JANUARY 7, 1960 - Shippingport Atomic Power Station, On Ohio River, 25 miles Northwest of Pittsburgh, Shippingport, Beaver County, PA

  18. INTERIOR VIEW OF FUEL STORAGE BUILDING (CPP603) SHOWING CRANE ASSEMBLY ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    INTERIOR VIEW OF FUEL STORAGE BUILDING (CPP-603) SHOWING CRANE ASSEMBLY FOR TRANSFER PIT. INL PHOTO NUMBER NRTS-51-2404. Unknown Photographer, 5/31/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  19. DOE Office of Scientific and Technical Information (OSTI.GOV)

    Not Available

    AFP no. 28 (General Electric Lynn Manufacturing dept). is located in the City of Everett, Mass. The facility is composed of 10 buildings having 344,342 square feet of floor space on a 43-acre tract. The plant is engaged in the manufacture of large jet engine components and sub-assemblies. AFT 29 (General Electric River Works Facility) is located in the City of Lynn, Mass. AFT No. 29 is part of the General Electric Aircraft Engine Business Group and the facilities are used for testing and assembly of jet engines. The following conclusions have been developed based on the results of themore » project team's field inspection, review of plant records and files, and interviews with plant personnel. Each of the sites listed below was ranked using the HARM system and was determined to have a sufficient potential for environmental contamination to warrant some degree of follow-on investigation. AFB no. 28: Waste sump and chip storage area; and AFT no. 29: Underground fuel line leaks and underground fuel storage tank leak.« less

  20. Nuclear reactor composite fuel assembly

    DOEpatents

    Burgess, Donn M.; Marr, Duane R.; Cappiello, Michael W.; Omberg, Ronald P.

    1980-01-01

    A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

  1. Passive Safety Features Evaluation of KIPT Neutron Source Facility

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Zhong, Zhaopeng; Gohar, Yousry

    2016-06-01

    Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have cooperated on the development, design, and construction of a neutron source facility. The facility was constructed at Kharkov, Ukraine and its commissioning process is underway. It will be used to conduct basic and applied nuclear research, produce medical isotopes, and train young nuclear specialists. The facility has an electron accelerator-driven subcritical assembly. The electron beam power is 100 kW using 100 MeV electrons. Tungsten or natural uranium is the target material for generating neutrons driving the subcritical assembly. The subcritical assemblymore » is composed of WWR-M2 - Russian fuel assemblies with U-235 enrichment of 19.7 wt%, surrounded by beryllium reflector assembles and graphite blocks. The subcritical assembly is seated in a water tank, which is a part of the primary cooling loop. During normal operation, the water coolant operates at room temperature and the total facility power is ~300 KW. The passive safety features of the facility are discussed in in this study. Monte Carlo computer code MCNPX was utilized in the analyses with ENDF/B-VII.0 nuclear data libraries. Negative reactivity temperature feedback was consistently observed, which is important for the facility safety performance. Due to the design of WWR-M2 fuel assemblies, slight water temperature increase and the corresponding water density decrease produce large reactivity drop, which offset the reactivity gain by mistakenly loading an additional fuel assembly. The increase of fuel temperature also causes sufficiently large reactivity decrease. This enhances the facility safety performance because fuel temperature increase provides prompt negative reactivity feedback. The reactivity variation due to an empty fuel position filled by water during the fuel loading process is examined. Also, the loading mistakes of removing beryllium reflector assemblies and replacing them with dummy assemblies were analyzed. In all these circumstances, the reactivity change results do not cause any safety concerns.« less

  2. Air breathing direct methanol fuel cell

    DOEpatents

    Ren, Xiaoming

    2002-01-01

    An air breathing direct methanol fuel cell is provided with a membrane electrode assembly, a conductive anode assembly that is permeable to air and directly open to atmospheric air, and a conductive cathode assembly that is permeable to methanol and directly contacting a liquid methanol source.

  3. Steady-State Thermal-Hydraulics Analyses for the Conversion of BR2 to Low Enriched Uranium Fuel

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Licht, J.; Bergeron, A.; Dionne, B.

    The code PLTEMP/ANL version 4.2 was used to perform the steady-state thermal-hydraulic analyses of the BR2 research reactor for conversion from Highly-Enriched to Low Enriched Uranium fuel (HEU and LEU, respectively). Calculations were performed to evaluate different fuel assemblies with respect to the onset of nucleate boiling (ONB), flow instability (FI), critical heat flux (CHF) and fuel temperature at beginning of cycle conditions. The fuel assemblies were characteristic of fresh fuel (0% burnup), highest heat flux (16% burnup), highest power (32% burnup) and highest burnup (46% burnup). Results show that the high heat flux fuel element is limiting for ONB,more » FI, and CHF, for both HEU and LEU fuel, but that the high power fuel element produces similar margin in a few cases. The maximum fuel temperature similarly occurs in both the high heat flux and high power fuel assemblies for both HEU and LEU fuel. A sensitivity study was also performed to evaluate the variation in fuel temperature due to uncertainties in the thermal conductivity degradation associated with burnup.« less

  4. Comparison of fresh fuel experimental measurements to MCNPX calculations using self-interrogation neutron resonance densitometry

    NASA Astrophysics Data System (ADS)

    LaFleur, Adrienne M.; Charlton, William S.; Menlove, Howard O.; Swinhoe, Martyn T.

    2012-07-01

    A new non-destructive assay technique called Self-Interrogation Neutron Resonance Densitometry (SINRD) is currently being developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for Light Water Reactor (LWR) fuel assemblies. SINRD consists of four 235U fission chambers (FCs): bare FC, boron carbide shielded FC, Gd covered FC, and Cd covered FC. Ratios of different FCs are used to determine the amount of resonance absorption from 235U in the fuel assembly. The sensitivity of this technique is based on using the same fissile materials in the FCs as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n,f) reaction peaks in the fission chamber. In this work, experimental measurements were performed in air with SINRD using a reference Pressurized Water Reactor (PWR) 15×15 low enriched uranium (LEU) fresh fuel assembly at LANL. The purpose of this experiment was to assess the following capabilities of SINRD: (1) ability to measure the effective 235U enrichment of the PWR fresh LEU fuel assembly and (2) sensitivity and penetrability to the removal of fuel pins from an assembly. These measurements were compared to Monte Carlo N-Particle eXtended transport code (MCNPX) simulations to verify the accuracy of the MCNPX model of SINRD. The reproducibility of experimental measurements via MCNPX simulations is essential to validating the results and conclusions obtained from the simulations of SINRD for LWR spent fuel assemblies.

  5. “A System for Automatically Maintaining Pressure in a Commercial Truck Tire”

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Maloney, John

    2017-07-07

    Under-inflated tires significantly reduce a vehicle’s fuel efficiency by increasing rolling resistance (drag force). The Air Maintenance Technology (“AMT”) system developed through this project replenishes lost air and maintains optimal tire cavity pressure whenever the tire is rolling in service, thus improving overall fuel economy by reducing the tire’s rolling resistance. The system consists of an inlet air filter, an air pump driven by tire deformation during rotation, and a pressure regulating device. Pressurized air in the tire cavity naturally escapes by diffusion through the tire and wheel, leaks in tire seating, and through the filler valve and its seating.more » As a result, tires require constant maintenance to replenish lost air. Since manual tire inflation maintenance is both labor intensive and time consuming, it is frequently overlooked or ignored. By automating the maintenance of optimal tire pressure, the tire’s contribution to the vehicle’s overall fuel economy can be maximized. The work was divided into three phases. The objectives of Phase 1, Planning and Initial Design, resulted in an effective project plan and to create a baseline design. The objectives for Phase 2, Design and Process Optimization, were: to identify finalized design for the pump, regulator and filter components; identify a process to build prototype tires; assemble prototype tires; test prototype tires and document results. The objectives of Phase 3, Design Release and Industrialization, were to finalize system tire assembly, perform release testing and industrialize the assembly process.« less

  6. Radial flow nuclear thermal rocket (RFNTR)

    DOEpatents

    Leyse, Carl F.

    1995-11-07

    A radial flow nuclear thermal rocket fuel assembly includes a substantially conical fuel element having an inlet side and an outlet side. An annular channel is disposed in the element for receiving a nuclear propellant, and a second, conical, channel is disposed in the element for discharging the propellant. The first channel is located radially outward from the second channel, and separated from the second channel by an annular fuel bed volume. This fuel bed volume can include a packed bed of loose fuel beads confined by a cold porous inlet frit and a hot porous exit frit. The loose fuel beads include ZrC coated ZrC-UC beads. In this manner, nuclear propellant enters the fuel assembly axially into the first channel at the inlet side of the element, flows axially across the fuel bed volume, and is discharged from the assembly by flowing radially outward from the second channel at the outlet side of the element.

  7. Radial flow nuclear thermal rocket (RFNTR)

    DOEpatents

    Leyse, Carl F.

    1995-01-01

    A radial flow nuclear thermal rocket fuel assembly includes a substantially conical fuel element having an inlet side and an outlet side. An annular channel is disposed in the element for receiving a nuclear propellant, and a second, conical, channel is disposed in the element for discharging the propellant. The first channel is located radially outward from the second channel, and separated from the second channel by an annular fuel bed volume. This fuel bed volume can include a packed bed of loose fuel beads confined by a cold porous inlet frit and a hot porous exit frit. The loose fuel beads include ZrC coated ZrC-UC beads. In this manner, nuclear propellant enters the fuel assembly axially into the first channel at the inlet side of the element, flows axially across the fuel bed volume, and is discharged from the assembly by flowing radially outward from the second channel at the outlet side of the element.

  8. Square lattice honeycomb reactor for space power and propulsion

    NASA Astrophysics Data System (ADS)

    Gouw, Reza; Anghaie, Samim

    2000-01-01

    The most recent nuclear design study at the Innovative Nuclear Space Power and Propulsion Institute (INSPI) is the Moderated Square-Lattice Honeycomb (M-SLHC) reactor design utilizing the solid solution of ternary carbide fuels. The reactor is fueled with solid solution of 93% enriched (U,Zr,Nb)C. The square-lattice honeycomb design provides high strength and is amenable to the processing complexities of these ultrahigh temperature fuels. The optimum core configuration requires a balance between high specific impulse and thrust level performance, and maintaining the temperature and strength limits of the fuel. The M-SLHC design is based on a cylindrical core that has critical radius and length of 37 cm and 50 cm, respectively. This design utilized zirconium hydrate to act as moderator. The fuel sub-assemblies are designed as cylindrical tubes with 12 cm in diameter and 10 cm in length. Five fuel subassemblies are stacked up axially to form one complete fuel assembly. These fuel assemblies are then arranged in the circular arrangement to form two fuel regions. The first fuel region consists of six fuel assemblies, and 18 fuel assemblies for the second fuel region. A 10-cm radial beryllium reflector in addition to 10-cm top axial beryllium reflector is used to reduce neutron leakage from the system. To perform nuclear design analysis of the M-SLHC design, a series of neutron transport and diffusion codes are used. To optimize the system design, five axial regions are specified. In each axial region, temperature and fuel density are varied. The axial and radial power distributions for the system are calculated, as well as the axial and radial flux distributions. Temperature coefficients of the system are also calculated. A water submersion accident scenario is also analyzed for these systems. Results of the nuclear design analysis indicate that a compact core can be designed based on ternary uranium carbide square-lattice honeycomb fuel, which provides a relatively high thrust to weight ratio. .

  9. Nuclear imaging of the fuel assembly in ignition experiments

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Grim, G. P.; Guler, N.; Merrill, F. E.

    First results from the analysis of neutron image data collected on implosions of cryogenically layered deuterium-tritium capsules during the 2011-2012 National Ignition Campaign are reported. The data span a variety of experimental designs aimed at increasing the stagnation pressure of the central hotspot and areal density of the surrounding fuel assembly. Images of neutrons produced by deuterium–tritium fusion reactions in the hotspot are presented, as well as images of neutrons that scatter in the surrounding dense fuel assembly. The image data are compared with 1D and 2D model predictions, and consistency checked using other diagnostic data. The results indicate thatmore » the size of the fusing hotspot is consistent with the model predictions, as well as other imaging data, while the overall size of the fuel assembly, inferred from the scattered neutron images, is systematically smaller than models’ prediction. Preliminary studies indicate these differences are consistent with a significant fraction (20%–25%) of the initial deuterium-tritium fuel mass outside the compact fuel assembly, due either to low mode mass asymmetry or high mode 3D mix effects at the ablator-ice interface.« less

  10. A custom-tailored FAMOS burn-up meter for VVER 440 fuel assemblies

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Simon, G.G.; Golochtchapov, S.; Glazov, A.G.

    1995-12-31

    The FAMOS fuel assembly monitoring system had been originally developed for monitoring irradiated fuel assemblies of the Karlsruhe Nuclear Research Center concentrating on neutron detection systems for special applications.The measurements in the past had demonstrated that FAMOS can perform precise measurements to control or measure with accuracy the main physical parameters of spent fuel. The FAMOS 3 system is specialized for burn-up determination of fuel assemblies. Thus it is possible to take into account the burn-up for the purposes of storage and transportation. The Kola NPP VVER 440 requirements necessitated developing an especially adopted FAMOS 3 system. In addition tomore » the passive neutron measurement, a gross gamma detection and a boron concentration monitoring system are implemented. The new system was constructed as well as tested in laboratory experiments. The monitoring system has been delivered to the customer and is ready for use.« less

  11. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Ragusa, Jean; Vierow, Karen

    2011-09-01

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzedmore » advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.« less

  12. Analysis and Test of a Proton Exchange Membrane Fuel Cell Power System for Space Power Applications

    NASA Technical Reports Server (NTRS)

    Vasquez, Arturo; Varanauski, Donald; Clark, Robert, Jr.

    2000-01-01

    An effort is underway to develop a prototype Proton Exchange Membrane (PEM) Fuel Cell breadboard system for fuhlre space applications. This prototype will be used to develop a comprehensive design basis for a space-rated PEM fuel cell powerplant. The prototype system includes reactant pressure regulators, ejector-based reactant pumps, a 4-kW fuel cell stack and cooling system, and a passive, membranebased oxygen / water separator. A computer model is being developed concurrently to analytically predict fluid flow in the oxidant reactant system. Fuel cells have historically played an important role in human-rated spacecraft. The Gemini and Apollo spacecraft used fuel cells for vehicle electrical power. The Space Shuttle currently uses three Alkaline Fuel Cell Powerplants (AFCP) to generate all of the vehicle's 15-20kW electrical power. Engineers at the Johnson Space Center have leveraged off the development effort ongoing in the commercial arena to develop PEM fuel cel ls for terrestrial uses. The prototype design originated from efforts to develop a PEM fuel cell replacement for the current Space Shuttle AFCP' s. In order to improve on the life and an already excellent hi storical record of reliability and safety, three subsystems were focused on. These were the fuel cell stack itself, the reactant circulation devices, and reactant / product water separator. PEM fuel cell stack performance is already demonstrating the potential for greater than four times the useful life of the current Shuttle's AFCP. Reactant pumping for product water removal has historically been accomplished with mechanical pumps. Ejectors offer an effective means of reactant pumping as well as the potential for weight reduction, control simplification, and long life. Centrifugal water separation is used on the current AFCP. A passive, membrane-based water separator offers compatibility with the micro-gravity environment of space, and the potential for control simplification, elimination of moving parts in an oxygen environment, and long life. The prototype system has been assembled from components that have previously been tested and evaluated at the component level. Preliminary data obtained from tests performed with the prototype system, as well as other published data, has been used to validate the analytical component models. These components have been incorporated into an integrated oxidant fluid system model. Results obtained from both the performance tests and the analytical model are presented.

  13. Transfer of fuel assemblies

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Vuckovich, M.; Burkett, J. P.; Sallustio, J.

    1984-12-11

    Fuel assemblies of a nuclear reactor are transferred during fueling or refueling or the like by a crane. The work-engaging fixture of the crane picks up an assembly, removes it from this slot, transfers it to the deposit site and deposits it in its slot at the deposit site. The control for the crane includes a strain gauge connected to the crane line which raises and lowers the load. The strain gauge senses the load on the crane. The signal from the strain gauge is compared with setpoints; a high-level setpoint, a low-level setpoint and a slack-line setpoint. If themore » strain gauge signal exceeds the high-level setpoint, the line drive is disabled. This event may occur during raising of a fuel assembly which encounters resistance. The high-level setpoint may be overridden under proper precautions. The line drive is also disabled if the strain gauge signal is less than the low-level setpoint. This event occurs when a fuel assembly being deposited contacts the bottom of its slot or an obstruction in, or at the entry to the slot. To preclude lateral movement and possible damage to a fuel assembly suspended from the crane line, the traverse drive of the crane is disabled once the strain-gauge exceets the lov-level setpoint. The traverse drive can only be enabled after the strain-gauge signal is less than the slack-line set-point. This occurs when the lines has been set in slack-line setting. When the line is tensioned after slack-li ne setting, the traverse drive remains enabled only if the line has been disconnected from the fuel assembly.« less

  14. Anti-air pollution & energy conservation system for automobiles using leaded or unleaded gasoline, diesel or alternate fuel

    DOEpatents

    Bose, Ranendra K.

    2002-06-04

    Exhaust gases from an internal combustion engine operating with leaded or unleaded gasoline or diesel or natural gas, are used for energizing a high-speed gas turbine. The convoluting gas discharge causes a first separation stage by stratifying of heavier and lighter exhaust gas components that exit from the turbine in opposite directions, the heavier components having a second stratifying separation in a vortex tube to separate combustible pollutants from non-combustible components. The non-combustible components exit a vortex tube open end to atmosphere. The lighter combustible, pollutants effected in the first separation are bubbled through a sodium hydroxide solution for dissolving the nitric oxide, formaldehyde impurities in this gas stream before being piped to the engine air intake for re-combustion, thereby reducing the engine's exhaust pollution and improving its fuel economy. The combustible, heavier pollutants from the second separation stage are piped to air filter assemblies. This gas stream convoluting at a high-speed through the top stator-vanes of the air filters, centrifugally separates the coalescent water, aldehydes, nitrogen dioxides, sulfates, sulfur, lead particles which collect at the bottom of the bowl, wherein it is periodically released to the roadway. Whereas, the heavier hydrocarbon, carbon particles are piped through the air filter's porous element to the engine air intake for re-combustion, further reducing the engine's exhaust pollution and improving its fuel economy.

  15. Polymer electrolyte membrane assembly for fuel cells

    NASA Technical Reports Server (NTRS)

    Yen, Shiao-Ping S. (Inventor); Kindler, Andrew (Inventor); Yavrouian, Andre (Inventor); Halpert, Gerald (Inventor)

    2002-01-01

    An electrolyte membrane for use in a fuel cell can contain sulfonated polyphenylether sulfones. The membrane can contain a first sulfonated polyphenylether sulfone and a second sulfonated polyphenylether sulfone, wherein the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone have equivalent weights greater than about 560, and the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone also have different equivalent weights. Also, a membrane for use in a fuel cell can contain a sulfonated polyphenylether sulfone and an unsulfonated polyphenylether sulfone. Methods for manufacturing a membrane electrode assemblies for use in fuel cells can include roughening a membrane surface. Electrodes and methods for fabricating such electrodes for use in a chemical fuel cell can include sintering an electrode. Such membranes and electrodes can be assembled into chemical fuel cells.

  16. Polymer electrolyte membrane assembly for fuel cells

    NASA Technical Reports Server (NTRS)

    Yen, Shiao-Ping S. (Inventor); Kindler, Andrew (Inventor); Yavrouian, Andre (Inventor); Halpert, Gerald (Inventor)

    2000-01-01

    An electrolyte membrane for use in a fuel cell can contain sulfonated polyphenylether sulfones. The membrane can contain a first sulfonated polyphenylether sulfone and a second sulfonated polyphenylether sulfone, wherein the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone have equivalent weights greater than about 560, and the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone also have different equivalent weights. Also, a membrane for use in a fuel cell can contain a sulfonated polyphenylether sulfone and an unsulfonated polyphenylether sulfone. Methods for manufacturing a membrane electrode assemblies for use in fuel cells can include roughening a membrane surface. Electrodes and methods for fabricating such electrodes for use in a chemical fuel cell can include sintering an electrode. Such membranes and electrodes can be assembled into chemical fuel cells.

  17. 78 FR 9796 - Airworthiness Directives; Cessna Aircraft Company Airplanes

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-02-12

    ... against the right steering tube assembly during rudder pedal actuation. This AD requires you to install... between the fuel return line assembly and the steering tube assembly and clearance between the fuel return...://www.cessnasupport.com . You may review copies of the referenced service information at the FAA, Small...

  18. 77 FR 50054 - Airworthiness Directives; Cessna Aircraft Company Airplanes

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-08-20

    ... rubbing against the right steering tube assembly during full rudder pedal actuation. This proposed AD would require you to inspect the fuel return line assembly for chafing; replace the fuel return line... right steering tube assembly and the airplane structure; and adjustment as necessary. We are proposing...

  19. 77 FR 72250 - Airworthiness Directives; Cessna Aircraft Company Airplanes

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-12-05

    ... rubbing against the right steering tube assembly during rudder pedal actuation. This proposed AD would require you to install the forward and aft fuel return line support clamps and brackets; inspect for a minimum clearance between the fuel return line assembly and the steering tube assembly and clearance...

  20. CONSTRUCTION OF NUCLEAR FUEL ELEMENTS

    DOEpatents

    Weems, S.J.

    1963-09-24

    >A rib arrangement and an end construction for nuclearfuel elements laid end to end in a coolant tube are described. The rib arrangement is such that each fuel element, when separated from other fuel elements, fits loosely in the coolant tube and so can easily be inserted or withdrawn from the tube. The end construction of the fuel elements is such that the fuel elements when assembled end to end are keyed against relative rotation, and the ribs of each fuel element cooperate with the ribs of the adjacent fuel elements to give the assembled fuel elements a tight fit with the coolant tube. (AEC)

  1. Monte Carlo characterization of PWR spent fuel assemblies to determine the detectability of pin diversion

    NASA Astrophysics Data System (ADS)

    Burdo, James S.

    This research is based on the concept that the diversion of nuclear fuel pins from Light Water Reactor (LWR) spent fuel assemblies is feasible by a careful comparison of spontaneous fission neutron and gamma levels in the guide tube locations of the fuel assemblies. The goal is to be able to determine whether some of the assembly fuel pins are either missing or have been replaced with dummy or fresh fuel pins. It is known that for typical commercial power spent fuel assemblies, the dominant spontaneous neutron emissions come from Cm-242 and Cm-244. Because of the shorter half-life of Cm-242 (0.45 yr) relative to that of Cm-244 (18.1 yr), Cm-244 is practically the only neutron source contributing to the neutron source term after the spent fuel assemblies are more than two years old. Initially, this research focused upon developing MCNP5 models of PWR fuel assemblies, modeling their depletion using the MONTEBURNS code, and by carrying out a preliminary depletion of a ¼ model 17x17 assembly from the TAKAHAMA-3 PWR. Later, the depletion and more accurate isotopic distribution in the pins at discharge was modeled using the TRITON depletion module of the SCALE computer code. Benchmarking comparisons were performed with the MONTEBURNS and TRITON results. Subsequently, the neutron flux in each of the guide tubes of the TAKAHAMA-3 PWR assembly at two years after discharge as calculated by the MCNP5 computer code was determined for various scenarios. Cases were considered for all spent fuel pins present and for replacement of a single pin at a position near the center of the assembly (10,9) and at the corner (17,1). Some scenarios were duplicated with a gamma flux calculation for high energies associated with Cm-244. For each case, the difference between the flux (neutron or gamma) for all spent fuel pins and with a pin removed or replaced is calculated for each guide tube. Different detection criteria were established. The first was whether the relative error of the difference was less than 1.00, allowing for the existence of the difference within the margin of error. The second was whether the difference between the two values was big enough to prevent their error bars from overlapping. Error analysis was performed both using a one second count and pseudo-Maxwell statistics for a projected 60 second count, giving four criteria for detection. The number of guide tubes meeting these criteria was compared and graphed for each case. Further analysis at extremes of high and low enrichment and long and short burnup times was done using data from assemblies at the Beaver Valley 1 and 2 PWR. In all neutron flux cases, at least two guide tube locations meet all the criteria for detection of pin diversion. At least one location in almost all of the gamma flux cases does. These results show that placing detectors in the empty guide tubes of spent fuel bundles to identify possible pin diversion is feasible.

  2. Closeup side view of Space Shuttle Main Engine (SSME) 2059 ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    Close-up side view of Space Shuttle Main Engine (SSME) 2059 mounted in a SSME Engine Handler near the Drying Area in the High Bay section of the SSME Processing Facility. The prominent features of the SSME in this view are the hot-gas expansion nozzle extending from the approximate image center toward the image right. The main-engine components extend from the approximate image center toward image right until it meets up with the mount for the SSME Engine Handler. The engine is rotated to a position where the major components in the view are the Low-Pressure Fuel Turbopump Discharge Duct with reflective foil insulation on the upper side of the engine, the Low-Pressure Oxidizer Turbopump and its Discharge Duct on the right side of the engine assembly extending itself down and wrapping under the bottom side of the assembly to the High-Pressure Oxidizer Turbopump pump. The High-Pressure Oxidizer Turbopump Discharge Duct exists the turbopump and extends up to the top side of the assembly where it enters the main oxidizer valve. The sphere on the lower side of the engine assembly is an accumulator that is part of the SSMEs POGO suppression system. - Space Transportation System, Space Shuttle Main Engine, Lyndon B. Johnson Space Center, 2101 NASA Parkway, Houston, Harris County, TX

  3. KSC-2009-5823

    NASA Image and Video Library

    2009-10-24

    CAPE CANAVERAL, Fla. – External tank 134 is towed toward the Vehicle Assembly Building, or VAB, at NASA's Kennedy Space Center in Florida. The Pegasus barge, carrying the fuel tank, arrived in Florida after an ocean voyage towed by a solid rocket booster retrieval ship from NASA's Michoud Assembly Facility near New Orleans. Next, the tank will be transported into the VAB where it will be stored until needed. ET-134 will be used to launch space shuttle Endeavour on the STS-130 mission to the International Space Station. Launch is targeted for Feb. 4, 2010. For information on the components of the space shuttle and the STS-130 mission, visit http://www.nasa.gov/shuttle. Photo credit: NASA/Jack Pfaller

  4. KSC-2009-5822

    NASA Image and Video Library

    2009-10-24

    CAPE CANAVERAL, Fla. – External tank 134 emerges from the Pegasus barge docked in the turn basin near the Vehicle Assembly Building, or VAB, at NASA's Kennedy Space Center in Florida. Pegasus arrived in Florida after an ocean voyage towed by a solid rocket booster retrieval ship from NASA's Michoud Assembly Facility near New Orleans. The fuel tank next will be transported into the VAB where it will be stored until needed. ET-134 will be used to launch space shuttle Endeavour on the STS-130 mission to the International Space Station. Launch is targeted for Feb. 4, 2010. For information on the components of the space shuttle and the STS-130 mission, visit http://www.nasa.gov/shuttle. Photo credit: NASA/Jack Pfaller

  5. Passive gamma analysis of the boiling-water-reactor assemblies

    NASA Astrophysics Data System (ADS)

    Vo, D.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S.; Trellue, H.; Vaccaro, S.

    2016-09-01

    This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative-Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

  6. Passive gamma analysis of the boiling-water-reactor assemblies

    DOE PAGES

    Vo, D.; Favalli, A.; Grogan, B.; ...

    2016-09-01

    This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden’s Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative–Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in themore » past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.« less

  7. Two-phase pressure drop reduction BWR assembly design

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Dix, G.E.; Crowther, R.L.; Colby, M.J.

    1991-05-21

    This patent describes an improved fuel assembly for a boiling water reactor. It comprises: a fuel channel; a lower tie plate; an upper tie plate; the lower tie plate and the upper tie plate defining a two-dimensional matrix; at least one water rod the fuel rods being partial length rods.

  8. 75 FR 3876 - Mark Edward Leyse; Receipt of Petition for Rulemaking

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-01-25

    ... (assembly) severe fuel damage experiments. The petitioner also requests that the NRC promulgate a regulation... aware that data from multi-rod (assembly) severe fuel damage experiments indicates that the current... fuel damage experiments indicates that the current peak cladding temperature limit contained in 10 CFR...

  9. General view of a Space Shuttle Main Engine (SSME) mounted ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    General view of a Space Shuttle Main Engine (SSME) mounted on an SSME engine handler, taken in the SSME Processing Facility at Kennedy Space Center. The most prominent features of the engine assembly in this view are the Low-Pressure Fuel Turbopump Discharge Duct looping around the right side and underneath the assembly, the High-Pressure Fuel Turbopump located on the lower left portion of the assembly, the Engine Controller and Main Fuel Valve Hydraulic Actuator located on the upper portion of the assembly and the Low-Pressure Oxidizer Turbopump Discharge Duct at the top of the engine assembly in this view. - Space Transportation System, Space Shuttle Main Engine, Lyndon B. Johnson Space Center, 2101 NASA Parkway, Houston, Harris County, TX

  10. PWR integral tie plate and locking mechanism

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Flora, B.S.; Osborne, J.L.

    1980-08-26

    A locking mechanism for securing an upper tie plate to the tie rods of a nuclear fuel bundle is described. The mechanism includes an upper tie plate assembly and locking sleeves fixed to the ends of the tie rods. The tie plate is part of the upper tie plate assembly and is secured to the fuel bundle by securing the entire upper tie plate assembly to the locking sleeves fixed to the tie rods. The assembly includes, in addition to the tie plate, locking nuts for engaging the locking sleeves, retaining sleeves to operably connect the locking nuts to themore » assembly, a spring biased reaction plate to restrain the locking nuts in the locked position and a means to facilitate the removal of the entire assembly as a unit from the fuel bundle.« less

  11. Mixed Oxide Fresh Fuel Package Auxiliary Equipment

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Yapuncich, F.; Ross, A.; Clark, R.H.

    2008-07-01

    The United States Department of Energy's National Nuclear Security Administration (NNSA) is overseeing the construction the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF) on the Savannah River Site. The new facility, being constructed by NNSA's contractor Shaw AREVA MOX Services, will fabricate fuel assemblies utilizing surplus plutonium as feedstock. The fuel will be used in designated commercial nuclear reactors. The MOX Fresh Fuel Package (MFFP), which has recently been licensed by the Nuclear Regulatory Commission (NRC) as a type B package (USA/9295/B(U)F-96), will be utilized to transport the fabricated fuel assemblies from the MFFF to the nuclear reactors. It wasmore » necessary to develop auxiliary equipment that would be able to efficiently handle the high precision fuel assemblies. Also, the physical constraints of the MFFF and the nuclear power plants require that the equipment be capable of loading and unloading the fuel assemblies both vertically and horizontally. The ability to reconfigure the load/unload evolution builds in a large degree of flexibility for the MFFP for the handling of many types of both fuel and non fuel payloads. The design and analysis met various technical specifications including dynamic and static seismic criteria. The fabrication was completed by three major fabrication facilities within the United States. The testing was conducted by Sandia National Laboratories. The unique design specifications and successful testing sequences will be discussed. (authors)« less

  12. Ultrafast dynamics in multifunctional Ru(II)-loaded polymers for solar energy conversion.

    PubMed

    Morseth, Zachary A; Wang, Li; Puodziukynaite, Egle; Leem, Gyu; Gilligan, Alexander T; Meyer, Thomas J; Schanze, Kirk S; Reynolds, John R; Papanikolas, John M

    2015-03-17

    The use of sunlight to make chemical fuels (i.e., solar fuels) is an attractive approach in the quest to develop sustainable energy sources. Using nature as a guide, assemblies for artificial photosynthesis will need to perform multiple functions. They will need to be able to harvest light across a broad region of the solar spectrum, transport excited-state energy to charge-separation sites, and then transport and store redox equivalents for use in the catalytic reactions that produce chemical fuels. This multifunctional behavior will require the assimilation of multiple components into a single macromolecular system. A wide variety of different architectures including porphyrin arrays, peptides, dendrimers, and polymers have been explored, with each design posing unique challenges. Polymer assemblies are attractive due to their relative ease of production and facile synthetic modification. However, their disordered nature gives rise to stochastic dynamics not present in more ordered assemblies. The rational design of assemblies requires a detailed understanding of the energy and electron transfer events that follow light absorption, which can occur on time scales ranging from femtoseconds to hundreds of microseconds, necessitating the use of sophisticated techniques. We have used a combination of time-resolved absorption and emission spectroscopies with observation times that span 9 orders of magnitude to follow the excited-state evolution within polymer-based molecular assemblies. We complement experimental observations with molecular dynamics simulations to develop a microscopic view of these dynamics. This Account provides an overview of our work on polymers decorated with pendant Ru(II) chromophores, both in solution and on surfaces. We have examined site-to-site energy transport among the Ru(II) complexes, and in systems incorporating π-conjugated polymers, we have observed ultrafast formation of a long-lived charge-separated state. When attached to TiO2, these assemblies exhibit multifunctional behavior in which photon absorption is followed by energy transport to the surface and electron injection to produce an oxidized metal complex. The oxidizing equivalent is then transferred to the conjugated polymer, giving rise to a long-lived charge-separated state.

  13. Evaluation of the magnitude and effects of bundle duct interaction in fuel assemblies at developmental plant conditions

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Serell, D.C.; Kaplan, S.

    1980-09-01

    Purpose of this evaluation is to estimate the magnitude and effects of irradiation and creep induced fuel bundle deformations in the developmental plant. This report focuses on the trends of the results and the ability of present models to evaluate the assembly temperatures in the presence of bundle deformation. Although this analysis focuses on the developmental plant, the conclusions are applicable to LMFBR fuel assemblies in general if they have wire spacers.

  14. Portable instrument for inspecting irradiated nuclear fuel assemblies

    DOEpatents

    Nicholson, Nicholas; Dowdy, Edward J.; Holt, David M.; Stump, Jr., Charles J.

    1985-01-01

    A portable instrument for measuring induced Cerenkov radiation associated with irradiated nuclear fuel assemblies in a water-filled storage pond is disclosed. The instrument includes a photomultiplier tube and an image intensifier which are operable in parallel and simultaneously by means of a field lens assembly and an associated beam splitter. The image intensifier permits an operator to aim and focus the apparatus on a submerged fuel assembly. Once the instrument is aimed and focused, an illumination reading can be obtained with the photomultiplier tube. The instrument includes a lens cap with a carbon-14/phosphor light source for calibrating the apparatus in the field.

  15. Cell module and fuel conditioner

    NASA Technical Reports Server (NTRS)

    Hoover, D. Q., Jr.

    1980-01-01

    Measurements of stack height changes with temperature and cell material characteristics were made. Stack 559 was assembled and components were fabricated for 560, 561, and 562. Stack 425 was transferred from the parallel DOE program and installed in the OS/IES simulation loop for mechanical and electrical testing. Construction and preliminary checkout of the 2 kW test facility was completed and design and procurement of the 8 kW test facility was initiated. The fuel conditioning subsystem design continued to evolve and the state points for the current design were calculated at full and part load conditions. Steam reforming catalyst activity tests were essentially completed and aging tests and CO shift converter tests were initiated.

  16. Robotic Manufacturing of 18-ft (5.5m) Diameter Cryogenic Fuel Tank Dome Assemblies for the NASA Ares I Rocket

    NASA Technical Reports Server (NTRS)

    Jones, Ronald E.; Carter, Robert W.

    2012-01-01

    The Ares I rocket was the first launch vehicle scheduled for manufacture under the National Aeronautic and Space Administration's Constellation program. A series of full-scale Ares I development articles were constructed on the Robotic Weld Tool at the NASA George C. Marshall Space Flight Center in Huntsville, Alabama. The Robotic Weld Tool is a 100 ton, 7- axis, robotic manufacturing system capable of machining and friction stir welding large-scale space hardware. This paper will focus on the friction stir welding of 18-ft (5.5m) diameter cryogenic fuel tank components; specifically, the liquid hydrogen forward dome and two common bulkhead manufacturing development articles.

  17. Method and apparatus for assembling solid oxide fuel cells

    DOEpatents

    Szreders, B.E.; Campanella, N.

    1988-05-11

    This invention relates generally to solid oxide fuel power generators and is particularly directed to improvements in the assembly and coupling of solid oxide fuel cell modules. A plurality of jet air tubes are supported and maintained in a spaced matrix array by a positioning/insertion assembly for insertion in respective tubes of a solid oxide fuel cell (SOFC) in the assembly of an SOFC module. The positioning/insertion assembly includes a plurality of generally planar, elongated, linear vanes which are pivotally mounted at each end thereof to a support frame. A rectangular compression assembly of adjustable size is adapted to receive and squeeze a matrix of SOFC tubes so as to compress the inter-tube nickel felt conductive pads which provide series/parallel electrical connection between adjacent SOFCs, with a series of increasingly larger retainer frames used to maintain larger matrices of SOFC tubes in position. Expansion of the SOFC module housing at the high operating temperatures of the SOFC is accommodated by conductive, flexible, resilient expansion, connector bars which provide support and electrical coupling at the top and bottom of the SOFC module housing. 17 figs.

  18. PLATES WITH OXIDE INSERTS

    DOEpatents

    West, J.M.; Schumar, J.F.

    1958-06-10

    Planar-type fuel assemblies for nuclear reactors are described, particularly those comprising fuel in the oxide form such as thoria and urania. The fuel assembly consists of a plurality of parallel spaced fuel plate mennbers having their longitudinal side edges attached to two parallel supporting side plates, thereby providing coolant flow channels between the opposite faces of adjacent fuel plates. The fuel plates are comprised of a plurality of longitudinally extending tubular sections connected by web portions, the tubular sections being filled with a plurality of pellets of the fuel material and the pellets being thermally bonded to the inside of the tubular section by lead.

  19. Hatch assembly

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Marshall, J.R.; Hardin, R.T. Jr.

    1987-07-07

    This patent describes a nuclear reactor installation including means defining a fuel handling area and means defining a containment area separated from the fuel handling area and including a refuelling cavity; the improvement comprising: (a) a fuel transfer tube connecting the refuelling cavity with the fuel handling area; the fuel transfer tube having a first end in the fuel handling area and a second end in the refueling cavity; (b) valve means for opening and closing the first end; and (c) a hatch assembly mounted on the second end; the hatch assembly including (1) a hatch ring affixed to themore » fuel transfer tube at the second end the hatch ring has an integral annular seat surrounded by the hatch ring and defines a hatch opening in the second end of the fuel transfer tube; (2) a hatch cover adapts to be positioned on the annular seat for covering the hatch opening; (3) latching units are supported on the hatch ring about the hatch opening, each latching unit.« less

  20. Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Marshall, William BJ J; Ade, Brian J; Bowman, Stephen M

    2015-01-01

    Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (k eff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of latticemore » design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup credit at peak reactivity requires a different set of experiments than for pressurized-water reactor burnup credit analysis because of differences in actinide compositions, presence of residual gadolinium absorber, and lower fission product concentrations. A survey of available critical experiments is presented along with a sample criticality code validation and determination of undercoverage penalties for some nuclides. The validation of depleted fuel compositions at peak reactivity presents many challenges which largely result from a lack of radiochemical assay data applicable to BWR fuel in this burnup range. In addition, none of the existing low burnup measurement data include residual gadolinium measurements. An example bias and uncertainty associated with validation of actinide-only fuel compositions is presented.« less

  1. NEUTRONIC REACTOR WITH ACCESSIBLE THIMBLE AND EMERGENCY COOLING FEATURES

    DOEpatents

    McCorkle, W.H.

    1960-02-23

    BS>A safety system for a water-moderated reactor is described. The invention comprises a reservoir system for spraying the fuel elements within a fuel assembly with coolant and keeping them in a continuous bath even if the coolant moderator is lost from the reactor vessel. A reservoir gravity feeds one or more nozzels positioned within each fuel assembly which continually forces water past the fuel elements.

  2. High efficiency gas burner

    DOEpatents

    Schuetz, Mark A.

    1983-01-01

    A burner assembly provides for 100% premixing of fuel and air by drawing the air into at least one high velocity stream of fuel without power assist. Specifically, the nozzle assembly for injecting the fuel into a throat comprises a plurality of nozzles in a generally circular array. Preferably, swirl is imparted to the air/fuel mixture by angling the nozzles. The diffuser comprises a conical primary diffuser followed by a cusp diffuser.

  3. Sobol's sensitivity analysis for a fuel cell stack assembly model with the aid of structure-selection techniques

    NASA Astrophysics Data System (ADS)

    Zhang, Wei; Cho, Chongdu; Piao, Changhao; Choi, Hojoon

    2016-01-01

    This paper presents a novel method for identifying the main parameters affecting the stress distribution of the components used in assembly modeling of proton exchange membrane fuel cell (PEMFC) stack. This method is a combination of an approximation model and Sobol's method, which allows a fast global sensitivity analysis for a set of uncertain parameters using only a limited number of calculations. Seven major parameters, i.e., Young's modulus of the end plate and the membrane electrode assembly (MEA), the contact stiffness between the MEA and bipolar plate (BPP), the X and Y positions of the bolts, the pressure of each bolt, and the thickness of the end plate, are investigated regarding their effect on four metrics, i.e., the maximum stresses of the MEA, BPP, and end plate, and the stress distribution percentage of the MEA. The analysis reveals the individual effects of each parameter and its interactions with the other parameters. The results show that the X position of a bolt has a major influence on the maximum stresses of the BPP and end plate, whereas the thickness of the end plate has the strongest effect on both the maximum stress and the stress distribution percentage of the MEA.

  4. Detecting pin diversion from pressurized water reactors spent fuel assemblies

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Ham, Young S.; Sitaraman, Shivakumar

    Detecting diversion of spent fuel from Pressurized Water Reactors (PWR) by determining possible diversion including the steps of providing a detector cluster containing gamma ray and neutron detectors, inserting the detector cluster containing the gamma ray and neutron detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring gamma ray and neutron radiation responses of the gamma ray and neutron detectors in the guide tube holes, processing the gamma ray and neutron radiation responses at the guide tube locations by normalizing them to the maximum value among each set of responses and takingmore » the ratio of the gamma ray and neutron responses at the guide tube locations and normalizing the ratios to the maximum value among them and producing three signatures, gamma, neutron, and gamma-neutron ratio, based on these normalized values, and producing an output that consists of these signatures that can indicate possible diversion of the pins from the spent fuel assembly.« less

  5. Silicon carbide composite for light water reactor fuel assembly applications

    NASA Astrophysics Data System (ADS)

    Yueh, Ken; Terrani, Kurt A.

    2014-05-01

    The feasibility of using SiCf-SiCm composites in light water reactor (LWR) fuel designs was evaluated. The evaluation was motivated by the desire to improve fuel performance under normal and accident conditions. The Fukushima accident once again highlighted the need for improved fuel materials that can maintain fuel integrity to higher temperatures for longer periods of time. The review identified many benefits as well as issues in using the material. Issues perceived as presenting the biggest challenges to the concept were identified to be flux gradient induced differential volumetric swelling, fragmentation and thermal shock resistance. The oxidation of silicon and its release into the coolant as silica has been identified as an issue because existing plant systems have limited ability for its removal. Detailed evaluation using available literature data and testing as part of this evaluation effort have eliminated most of the major concerns. The evaluation identified Boiling Water Reactor (BWR) channel, BWR fuel water tube, and Pressurized Water Reactor (PWR) guide tube as feasible applications for SiC composite. A program has been initiated to resolve some of the remaining issues and to generate physical property data to support the design of commercial fuel components.

  6. Combustor with two stage primary fuel tube with concentric members and flow regulating

    DOEpatents

    Parker, David Marchant; Whidden, Graydon Lane; Zolyomi, Wendel

    1999-01-01

    A combustor for a gas turbine having a centrally located fuel nozzle and inner, middle and outer concentric cylindrical liners, the inner liner enclosing a primary combustion zone. The combustor has an air inlet that forms two passages for pre-mixing primary fuel and air to be supplied to the primary combustion zone. Each of the pre-mixing passages has a circumferential array of swirl vanes. A plurality of primary fuel tube assemblies extend through both pre-mixing passages, with each primary fuel tube assembly located between a pair of swirl vanes. Each primary fuel tube assembly is comprised of two tubular members. The first member supplies fuel to the first pre-mixing passage, while the second member, which extends through the first member, supplies fuel to the second pre-mixing passage. An annular fuel manifold is divided into first and second chambers by a circumferentially extending baffle. The proximal end of the first member is attached to the manifold itself while the proximal end of the second member is attached to the baffle. The distal end of the first member is attached directly to the second member at around its mid-point. The inlets of the first and second members are in flow communication with the first and second manifold chambers, respectively. Control valves separately regulate the flow of fuel to the two chambers and, therefore, to the two members of the fuel tube assemblies, thereby allowing the flow of fuel to the first and second pre-mixing passages to be separately controlled.

  7. Tunable Gas Sensing Gels by Cooperative Assembly

    PubMed Central

    Hussain, Abid; Semeano, Ana T. S.; Palma, Susana I. C. J.; Pina, Ana S.; Almeida, José; Medrado, Bárbara F.; Pádua, Ana C. C. S.; Carvalho, Ana L.; Dionísio, Madalena; Li, Rosamaria W. C.; Gamboa, Hugo; Ulijn, Rein V.; Gruber, Jonas; Roque, Ana C. A.

    2017-01-01

    The cooperative assembly of biopolymers and small molecules can yield functional materials with precisely tunable properties. Here, the fabrication, characterization, and use of multicomponent hybrid gels as selective gas sensors are reported. The gels are composed of liquid crystal droplets self-assembled in the presence of ionic liquids, which further coassemble with biopolymers to form stable matrices. Each individual component can be varied and acts cooperatively to tune gels’ structure and function. The unique molecular environment in hybrid gels is explored for supramolecular recognition of volatile compounds. Gels with distinct compositions are used as optical and electrical gas sensors, yielding a combinatorial response conceptually mimicking olfactory biological systems, and tested to distinguish volatile organic compounds and to quantify ethanol in automotive fuel. The gel response is rapid, reversible, and reproducible. These robust, versatile, modular, pliant electro-optical soft materials possess new possibilities in sensing triggered by chemical and physical stimuli. PMID:28747856

  8. Requirements to the procedure and stages of innovative fuel development

    NASA Astrophysics Data System (ADS)

    Troyanov, V.; Zabudko, L.; Grachyov, A.; Zhdanova, O.

    2016-04-01

    According to the accepted current understanding under the nuclear fuel we will consider the assembled active zone unit (Fuel assembly) with its structural elements, fuel rods, pellet column, structural materials of fuel rods and fuel assemblies. The licensing process includes justification of safe application of the proposed modifications, including design-basis and experimental justification of the modified items under normal operating conditions and in violation of normal conditions, including accidents as well. Besides the justification of modified units itself, it is required to show the influence of modifications on the performance and safety of the other Reactor Unit’ and Nuclear Plant’ elements (e.g. burst can detection system, transportation and processing operations during fuel handling), as well as to justify the new standards of fuel storage etc. Finally, the modified fuel should comply with the applicable regulations, which often becomes a very difficult task, if only because those regulations, such as the NP-082-07, are not covered modification issues. Making amendments into regulations can be considered as the only solution, but the process is complicated and requires deep grounds for amendments. Some aspects of licensing new nuclear fuel are considered the example of mixed nitride uranium -plutonium fuel application for the BREST reactor unit.

  9. Storage, transportation and disposal system for used nuclear fuel assemblies

    DOEpatents

    Scaglione, John M.; Wagner, John C.

    2017-01-10

    An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

  10. Simulation of Assembly Line Balancing in Automotive Component Manufacturing

    NASA Astrophysics Data System (ADS)

    Jamil, Muthanna; Mohd Razali, Noraini

    2016-02-01

    This study focuses on the simulation of assembly line balancing in an automotive component in a vendor manufacturing company. A mixed-model assembly line of charcoal canister product that is used in an engine system as fuel's vapour filter was observed and found that the current production rate of the line does not achieve customer demand even though the company practices buffer stock for two days in advance. This study was carried out by performing detailed process flow and time studies along the line. To set up a model of the line by simulation, real data was taken from a factory floor and tested for distribution fit. The data gathered was then transformed into a simulation model. After verification of the model by comparing it with the actual system, it was found that the current line efficiency is not at its optimum condition due to blockage and idle time. Various what-if analysis were applied to eliminate the cause. Proposed layout shows that the line is balanced by adding buffer to avoid the blockage. Whereas, manpower is added the stations to reduce process time therefore reducing idling time. The simulation study was carried out using ProModel software.

  11. TSTA Piping and Flame Arrestor Operating Experience Data

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Cadwallader, Lee C.; Willms, R. Scott

    The Tritium Systems Test Assembly (TSTA) was a facility dedicated to tritium handling technology and experiment research at the Los Alamos National Laboratory. The facility operated from 1984 to 2001, running a prototype fusion fuel processing loop with ~100 grams of tritium as well as small experiments. There have been several operating experience reports written on this facility’s operation and maintenance experience. This paper describes analysis of two additional components from TSTA, small diameter gas piping that handled small amounts of tritium in a nitrogen carrier gas, and the flame arrestor used in this piping system. The operating experiences andmore » the component failure rates for these components are discussed in this paper. Comparison data from other applications are also presented.« less

  12. Portable instrument for inspecting irradiated nuclear-fuel assemblies in a water-filled storage pond by measurement of induced Cerenkov radiation

    DOEpatents

    Nicholson, N.; Dowdy, E.J.; Holt, D.M.; Stump, C.J. Jr.

    1982-05-13

    A portable instrument for measuring induced Cerenkov radiation associated with irradiated nuclear fuel assemblies in a water-filled storage pond is disclosed. The instrument includes a photomultiplier tube and an image intensifier which are operable in parallel and simultaneously by means of a field lens assembly and an associated beam splitter. The image intensifier permits an operator to aim and focus the apparatus on a submerged fuel assembly. Once the instrument is aimed and focused, an illumination reading can be obtained with the photomultiplier tube. The instrument includes a lens cap with a carbon-14/phosphor light source for calibrating the apparatus in the field.

  13. Corrosion and hydrogen pick-up behaviors of cladding and structural components in BWR high burnup 9x9 lead use assemblies

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Miyashita, Toshiyasu; Nakae, Nobuo; Ogata, Keizo

    The high burnup BWR 9x9 lead use fuel assemblies, which have been designed for maximum assembly burnup of 55 GWd/t in Japan, have been examined after irradiations to confirm the reliability of the current safety evaluation methodology, and to accumulate data to judge the adequacy to apply it to the future higher burnup fuel. After 3 and 5 cycle irradiations, post irradiation examinations were performed for both 9x9 Type-A and Type-B fuel assemblies. Both Type LUAs utilize Zry-2 claddings, while there are deviation in the contents of impurity and alloying elements between Type-A and Type-B, especially in Fe and Simore » concentration. Measured oxide thicknesses of fuel rods showed no significant difference between after 3 and 5 cycle irradiation except for some rods at corner position in Type B LUA. The axial profile of hydrogen concentration and oxide thickness for the corner rods in Type B LUA after 5 cycle irradiation had peaks at the second lowest span from the bottom. The maximum oxide thickness is about 50 {mu}m on the surface facing the bundle outside at the second lowest span and dense hydrides layer (Hydride rim) is observed in peripheral region of cladding showing unexpected high hydrogen concentration. The results of calculated thermal-hydraulic conditions show that the thermal neutron flux at the corner position was higher than the other position. On the other hand, the void fraction and the mass flux were relatively lower at the corner position. The oxide thickness on spacer band and spacer cell of Zry-2 increases from 3 to 5 cycle irradiations. Spacer band of Zry-4 showed significantly thick oxide after 5 cycle irradiations but Hydrogen concentration was relatively small in contrast its obviously thick oxide in comparison with Zry-2 spacer bands. The large increase in hydrogen concentration was measured in Zry-2 spacers after 5 cycle irradiations and the evaluated hydrogen pick-up rate also increased remarkably. (authors)« less

  14. KSC-2009-5819

    NASA Image and Video Library

    2009-10-24

    CAPE CANAVERAL, Fla. – Tugboats safely deliver the Pegasus barge, carrying external tank 134, to the dock in the turn basin near the Vehicle Assembly Building, or VAB, at NASA's Kennedy Space Center in Florida. Pegasus arrived in Florida after an ocean voyage towed by a solid rocket booster retrieval ship from NASA's Michoud Assembly Facility near New Orleans. After Pegasus docks in the turn basin, the fuel tank will be offloaded and transported into the VAB. ET-134 will be used to launch space shuttle Endeavour on the STS-130 mission to the International Space Station. Launch is targeted for Feb. 4, 2010. For information on the components of the space shuttle and the STS-130 mission, visit http://www.nasa.gov/shuttle. Photo credit: NASA/Jack Pfaller

  15. KSC-2009-5818

    NASA Image and Video Library

    2009-10-24

    CAPE CANAVERAL, Fla. – Tugboats safely deliver the Pegasus barge, carrying external tank 134, to the dock in the turn basin near the Vehicle Assembly Building, or VAB, at NASA's Kennedy Space Center in Florida. Pegasus arrived in Florida after an ocean voyage towed by a solid rocket booster retrieval ship from NASA's Michoud Assembly Facility near New Orleans. After Pegasus docks in the turn basin, the fuel tank will be offloaded and transported into the VAB. ET-134 will be used to launch space shuttle Endeavour on the STS-130 mission to the International Space Station. Launch is targeted for Feb. 4, 2010. For information on the components of the space shuttle and the STS-130 mission, visit http://www.nasa.gov/shuttle. Photo credit: NASA/Jack Pfaller

  16. KSC-2009-5825

    NASA Image and Video Library

    2009-10-24

    CAPE CANAVERAL, Fla. – External tank 134 is towed toward the open door of the transfer aisle of the Vehicle Assembly Building, or VAB, at NASA's Kennedy Space Center in Florida. The Pegasus barge, carrying the fuel tank, arrived in Florida after an ocean voyage towed by a solid rocket booster retrieval ship from NASA's Michoud Assembly Facility near New Orleans. Next, the tank will be transported into the VAB where it will be stored until needed. ET-134 will be used to launch space shuttle Endeavour on the STS-130 mission to the International Space Station. Launch is targeted for Feb. 4, 2010. For information on the components of the space shuttle and the STS-130 mission, visit http://www.nasa.gov/shuttle. Photo credit: NASA/Jack Pfaller

  17. KSC-2009-5821

    NASA Image and Video Library

    2009-10-24

    CAPE CANAVERAL, Fla. – A worker tows external tank 134 off the Pegasus barge docked in the turn basin near the Vehicle Assembly Building, or VAB, at NASA's Kennedy Space Center in Florida. Pegasus arrived in Florida after an ocean voyage towed by a solid rocket booster retrieval ship from NASA's Michoud Assembly Facility near New Orleans. After the fuel tank is offloaded, it will be transported into the VAB where it will be stored until needed. ET-134 will be used to launch space shuttle Endeavour on the STS-130 mission to the International Space Station. Launch is targeted for Feb. 4, 2010. For information on the components of the space shuttle and the STS-130 mission, visit http://www.nasa.gov/shuttle. Photo credit: NASA/Jack Pfaller

  18. KSC-2009-5817

    NASA Image and Video Library

    2009-10-24

    CAPE CANAVERAL, Fla. – Tugboats safely deliver the Pegasus barge, carrying external tank 134, to the dock in the turn basin near the Vehicle Assembly Building, or VAB, at NASA's Kennedy Space Center in Florida. Pegasus arrived in Florida after an ocean voyage towed by a solid rocket booster retrieval ship from NASA's Michoud Assembly Facility near New Orleans. After Pegasus docks in the turn basin, the fuel tank will be offloaded and transported into the VAB. ET-134 will be used to launch space shuttle Endeavour on the STS-130 mission to the International Space Station. Launch is targeted for Feb. 4, 2010. For information on the components of the space shuttle and the STS-130 mission, visit http://www.nasa.gov/shuttle. Photo credit: NASA/Jack Pfaller

  19. KSC-2009-5820

    NASA Image and Video Library

    2009-10-24

    CAPE CANAVERAL, Fla. – Workers prepare to offload external tank 134 from the Pegasus barge docked in the turn basin near the Vehicle Assembly Building, or VAB, at NASA's Kennedy Space Center in Florida. Pegasus arrived in Florida after an ocean voyage towed by a solid rocket booster retrieval ship from NASA's Michoud Assembly Facility near New Orleans. After Pegasus docks in the turn basin, the fuel tank will be offloaded and transported into the VAB. ET-134 will be used to launch space shuttle Endeavour on the STS-130 mission to the International Space Station. Launch is targeted for Feb. 4, 2010. For information on the components of the space shuttle and the STS-130 mission, visit http://www.nasa.gov/shuttle. Photo credit: NASA/Jack Pfaller

  20. KSC-2009-5824

    NASA Image and Video Library

    2009-10-24

    CAPE CANAVERAL, Fla. – External tank 134 is towed toward the looming 525-foot-tall Vehicle Assembly Building, or VAB, at NASA's Kennedy Space Center in Florida. The Pegasus barge, carrying the fuel tank, arrived in Florida after an ocean voyage towed by a solid rocket booster retrieval ship from NASA's Michoud Assembly Facility near New Orleans. Next, the tank will be transported into the VAB where it will be stored until needed. ET-134 will be used to launch space shuttle Endeavour on the STS-130 mission to the International Space Station. Launch is targeted for Feb. 4, 2010. For information on the components of the space shuttle and the STS-130 mission, visit http://www.nasa.gov/shuttle. Photo credit: NASA/Jack Pfaller

  1. Closeup view of the top of Space Shuttle Main Engine ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    Close-up view of the top of Space Shuttle Main Engine (SSME) 2057 mounted in a SSME Engine Handler in the Vertical Processing area of the SSME Processing Facility at Kennedy Space Center. The most prominent components in this view is the large Low-Pressure Oxidizer Turbopump (LPOTP) Discharge Duct wrapping itself around the right side of the engine assembly. The smaller tube to the left of LPOTP Discharge Duct is the High-Pressure Oxidizer Duct used to supply the turbine of the LPOTP. The other major feature in this view is the Low-Pressure Fuel Turbopump at the top of the engine assembly. - Space Transportation System, Space Shuttle Main Engine, Lyndon B. Johnson Space Center, 2101 NASA Parkway, Houston, Harris County, TX

  2. Fuel cell with electrolyte matrix assembly

    DOEpatents

    Kaufman, Arthur; Pudick, Sheldon; Wang, Chiu L.

    1988-01-01

    This invention is directed to a fuel cell employing a substantially immobilized electrolyte imbedded therein and having a laminated matrix assembly disposed between the electrodes of the cell for holding and distributing the electrolyte. The matrix assembly comprises a non-conducting fibrous material such as silicon carbide whiskers having a relatively large void-fraction and a layer of material having a relatively small void-fraction.

  3. Independent Orbiter Assessment (IOA): Analysis of the electrical power generation/fuel cell powerplant subsystem

    NASA Technical Reports Server (NTRS)

    Brown, K. L.; Bertsch, P. J.

    1986-01-01

    Results of the Independent Orbiter Assessment (IOA) of the Failure Modes and Effects Analysis (FMEA) and Critical Items List (CIL) are presented. The IOA approach features a top-down analysis of the hardware to determine failure modes, criticality, and potential critical items. To preserve independence, this analysis was accomplished without reliance upon the results contained within the NASA FMEA/CIL documentation. This report documents the independent analysis results corresponding to the Orbiter Electrical Power Generation (EPG)/Fuel Cell Powerplant (FCP) hardware. The EPG/FCP hardware is required for performing functions of electrical power generation and product water distribution in the Orbiter. Specifically, the EPG/FCP hardware consists of the following divisions: (1) Power Section Assembly (PSA); (2) Reactant Control Subsystem (RCS); (3) Thermal Control Subsystem (TCS); and (4) Water Removal Subsystem (WRS). The IOA analysis process utilized available EPG/FCP hardware drawings and schematics for defining hardware assemblies, components, and hardware items. Each level of hardware was evaluated and analyzed for possible failure modes and effects. Criticality was assigned based upon the severity of the effect for each failure mode.

  4. Modular fuel-cell stack assembly

    DOEpatents

    Patel, Pinakin [Danbury, CT; Urko, Willam [West Granby, CT

    2008-01-29

    A modular multi-stack fuel-cell assembly in which the fuel-cell stacks are situated within a containment structure and in which a gas distributor is provided in the structure and distributes received fuel and oxidant gases to the stacks and receives exhausted fuel and oxidant gas from the stacks so as to realize a desired gas flow distribution and gas pressure differential through the stacks. The gas distributor is centrally and symmetrically arranged relative to the stacks so that it itself promotes realization of the desired gas flow distribution and pressure differential.

  5. Creating NDA working standards through high-fidelity spent fuel modeling

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Skutnik, Steven E; Gauld, Ian C; Romano, Catherine E

    2012-01-01

    The Next Generation Safeguards Initiative (NGSI) is developing advanced non-destructive assay (NDA) techniques for spent nuclear fuel assemblies to advance the state-of-the-art in safeguards measurements. These measurements aim beyond the capabilities of existing methods to include the evaluation of plutonium and fissile material inventory, independent of operator declarations. Testing and evaluation of advanced NDA performance will require reference assemblies with well-characterized compositions to serve as working standards against which the NDA methods can be benchmarked and for uncertainty quantification. To support the development of standards for the NGSI spent fuel NDA project, high-fidelity modeling of irradiated fuel assemblies is beingmore » performed to characterize fuel compositions and radiation emission data. The assembly depletion simulations apply detailed operating history information and core simulation data as it is available to perform high fidelity axial and pin-by-pin fuel characterization for more than 1600 nuclides. The resulting pin-by-pin isotopic inventories are used to optimize the NDA measurements and provide information necessary to unfold and interpret the measurement data, e.g., passive gamma emitters, neutron emitters, neutron absorbers, and fissile content. A key requirement of this study is the analysis of uncertainties associated with the calculated compositions and signatures for the standard assemblies; uncertainties introduced by the calculation methods, nuclear data, and operating information. An integral part of this assessment involves the application of experimental data from destructive radiochemical assay to assess the uncertainty and bias in computed inventories, the impact of parameters such as assembly burnup gradients and burnable poisons, and the influence of neighboring assemblies on periphery rods. This paper will present the results of high fidelity assembly depletion modeling and uncertainty analysis from independent calculations performed using SCALE and MCNP. This work is supported by the Next Generation Safeguards Initiative, Office of Nuclear Safeguards and Security, National Nuclear Security Administration.« less

  6. Investigation of Coating and Corrosion Mitigation Strategies in Magnesium/Mixed Metal Assemblies

    NASA Astrophysics Data System (ADS)

    Forsmark, Joy H.; McCune, Robert C.; Giles, Terry; Audette, Michelle; Snowden, Jasmine; Stalker, Jeff; Morey, Matthew; O'Keefe, Matt; Castano, Carlos

    The US Automotive Materials Partnership through the Magnesium-Intensive Front End Development Project (MFERD) is currently investigating a number of joining, coating and corrosion mitigation strategies to incorporate magnesium components into the automotive body-in-white with the ultimate goal of decreasing vehicle curb weight, thus improving fuel economy. Because Mg is anodic to all other structural metals, this is a key hurdle to Mg component implementation in vehicles. This paper will discuss the results of a study to examine the effectiveness of different corrosion mitigation strategies in joined plate assemblies and provide some insight into the systems challenges of incorporation of Mg parts into a vehicle. Details of a statistically-designed experiment developed to explore the interaction of several materials of construction (magnesium, steel and aluminum), pretreatment and topcoatings, joining methods and standardized test protocols including SAE J-2334 and ASTM B-117 are discussed. A number of avenues have emerged from this study as potential strategies for corrosion mitigation.

  7. Decay heat uncertainty for BWR used fuel due to modeling and nuclear data uncertainties

    DOE PAGES

    Ilas, Germina; Liljenfeldt, Henrik

    2017-05-19

    Characterization of the energy released from radionuclide decay in nuclear fuel discharged from reactors is essential for the design, safety, and licensing analyses of used nuclear fuel storage, transportation, and repository systems. There are a limited number of decay heat measurements available for commercial used fuel applications. Because decay heat measurements can be expensive or impractical for covering the multitude of existing fuel designs, operating conditions, and specific application purposes, decay heat estimation relies heavily on computer code prediction. Uncertainty evaluation for calculated decay heat is an important aspect when assessing code prediction and a key factor supporting decision makingmore » for used fuel applications. While previous studies have largely focused on uncertainties in code predictions due to nuclear data uncertainties, this study discusses uncertainties in calculated decay heat due to uncertainties in assembly modeling parameters as well as in nuclear data. Capabilities in the SCALE nuclear analysis code system were used to quantify the effect on calculated decay heat of uncertainties in nuclear data and selected manufacturing and operation parameters for a typical boiling water reactor (BWR) fuel assembly. Furthermore, the BWR fuel assembly used as the reference case for this study was selected from a set of assemblies for which high-quality decay heat measurements are available, to assess the significance of the results through comparison with calculated and measured decay heat data.« less

  8. Decay heat uncertainty for BWR used fuel due to modeling and nuclear data uncertainties

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Ilas, Germina; Liljenfeldt, Henrik

    Characterization of the energy released from radionuclide decay in nuclear fuel discharged from reactors is essential for the design, safety, and licensing analyses of used nuclear fuel storage, transportation, and repository systems. There are a limited number of decay heat measurements available for commercial used fuel applications. Because decay heat measurements can be expensive or impractical for covering the multitude of existing fuel designs, operating conditions, and specific application purposes, decay heat estimation relies heavily on computer code prediction. Uncertainty evaluation for calculated decay heat is an important aspect when assessing code prediction and a key factor supporting decision makingmore » for used fuel applications. While previous studies have largely focused on uncertainties in code predictions due to nuclear data uncertainties, this study discusses uncertainties in calculated decay heat due to uncertainties in assembly modeling parameters as well as in nuclear data. Capabilities in the SCALE nuclear analysis code system were used to quantify the effect on calculated decay heat of uncertainties in nuclear data and selected manufacturing and operation parameters for a typical boiling water reactor (BWR) fuel assembly. Furthermore, the BWR fuel assembly used as the reference case for this study was selected from a set of assemblies for which high-quality decay heat measurements are available, to assess the significance of the results through comparison with calculated and measured decay heat data.« less

  9. General view of the Space Shuttle Main Engine (SSME) assembly ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    General view of the Space Shuttle Main Engine (SSME) assembly with the expansion nozzle removed and resting on a cushioned mat on the floor of the SSME Processing Facility. The most prominent features in this view are the Low-pressure oxidizer Turbopump discharge Duct looping from the upper left side of the engine assembly to the lower left side of the assembly, the Low-Pressure Fuel Turbopump (LPFTP) is on the upper left of the assembly in this view and the LPFTP Discharge Duct loops from the upper left to upper right then turns back and down the assembly to the High-Pressure Fuel Turbopump on the lower right of the assembly. The Engine Controller and the Main fuel Valve Hydraulic Actuator are on the lower left portion of the assembly. The vertical rod that is in the approximate center of the engine assembly is a piece of ground support equipment call a Gimbal Actuator Replacement Strut which are used on the SSMEs when they are not installed in an orbiter. - Space Transportation System, Space Shuttle Main Engine, Lyndon B. Johnson Space Center, 2101 NASA Parkway, Houston, Harris County, TX

  10. PBF Reactor Building (PER620). Fuel rod test assembly is on ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    PBF Reactor Building (PER-620). Fuel rod test assembly is on display at PBF. Date: 1982. INEEL negative no. 82-4893 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  11. Membrane electrode assembly for a fuel cell

    NASA Technical Reports Server (NTRS)

    Prakash, Surya (Inventor); Narayanan, Sekharipuram R. (Inventor); Atti, Anthony (Inventor); Olah, George (Inventor); Smart, Marshall C. (Inventor)

    2006-01-01

    A catalyst ink for a fuel cell including a catalytic material and poly(vinylidene fluoride). The ink may be applied to a substrate to form an electrode, or bonded with other electrode layers to form a membrane electrode assembly (MEA).

  12. Method of bonding

    DOEpatents

    Saller, deceased, Henry A.; Hodge, Edwin S.; Paprocki, Stanley J.; Dayton, Russell W.

    1987-12-01

    1. A method of making a fuel-containing structure for nuclear reactors, comprising providing an assembly comprising a plurality of fuel units; each fuel unit consisting of a core plate containing thermal-neutron-fissionable material, sheets of cladding metal on its bottom and top surfaces, said cladding sheets being of greater width and length than said core plates whereby recesses are formed at the ends and sides of said core plate, and end pieces and first side pieces of cladding metal of the same thickness as the core plate positioned in said recesses, the assembly further comprising a plurality of second side pieces of cladding metal engaging the cladding sheets so as to space the fuel units from one another, and a plurality of filler plates of an acid-dissolvable nonresilient material whose melting point is above 2000.degree. F., each filler plate being arranged between a pair of said second side pieces and the cladding plates of two adjacent fuel units, the filler plates having the same thickness as the second side pieces; the method further comprising enclosing the entire assembly in an envelope; evacuating the interior of the entire assembly through said envelope; applying inert gas under a pressure of about 10,000 psi to the outside of said envelope while at the same time heating the assembly to a temperature above the flow point of the cladding metal but below the melting point of any material of the assembly, whereby the envelope is pressed against the assembly and integral bonds are formed between plates, sheets, first side pieces, and end pieces and between the sheets and the second side pieces; slowly cooling the assembly to room temperature; removing the envelope; and dissolving the filler plates without attacking the cladding metal.

  13. Storage, transportation and disposal system for used nuclear fuel assemblies

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Scaglione, John M.; Wagner, John C.

    An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. Themore » system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.« less

  14. Square lattice honeycomb tri-carbide fuels for 50 to 250 KN variable thrust NTP design

    NASA Astrophysics Data System (ADS)

    Anghaie, Samim; Knight, Travis; Gouw, Reza; Furman, Eric

    2001-02-01

    Ultrahigh temperature solid solution of tri-carbide fuels are used to design an ultracompact nuclear thermal rocket generating 950 seconds of specific impulse with scalable thrust level in range of 50 to 250 kilo Newtons. Solid solutions of tri-carbide nuclear fuels such as uranium-zirconium-niobium carbide. UZrNbC, are processed to contain certain mixing ratio between uranium carbide and two stabilizing carbides. Zirconium or niobium in the tri-carbide could be replaced by tantalum or hafnium to provide higher chemical stability in hot hydrogen environment or to provide different nuclear design characteristics. Recent studies have demonstrated the chemical compatibility of tri-carbide fuels with hydrogen propellant for a few to tens of hours of operation at temperatures ranging from 2800 K to 3300 K, respectively. Fuel elements are fabricated from thin tri-carbide wafers that are grooved and locked into a square-lattice honeycomb (SLHC) shape. The hockey puck shaped SLHC fuel elements are stacked up in a grooved graphite tube to form a SLHC fuel assembly. A total of 18 fuel assemblies are arranged circumferentially to form two concentric rings of fuel assemblies with zirconium hydride filling the space between assemblies. For 50 to 250 kilo Newtons thrust operations, the reactor diameter and length including reflectors are 57 cm and 60 cm, respectively. Results of the nuclear design and thermal fluid analyses of the SLHC nuclear thermal propulsion system are presented. .

  15. Non-precious metal catalysts prepared from precursors comprising cyanamide and polyaniline

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Chung, Hoon Taek; Zelenay, Piotr

    A catalyst for oxygen reduction reaction (ORR) for a fuel cell was prepared by pyrolyzing a mixture of polyaniline, cyanamide, carbon black, and a non-precious metal salt under an inert atmosphere. The pyrolyzed product was treated to remove acid soluble components and then pyrolyzed again. The resulting powder was used to prepare a cathode for a membrane electrode assembly that was used in a fuel cell. When iron(III) chloride was used as the salt, the resulting catalyst was porous with a web-shaped structure. It displayed a maximum power density of 0.79 W/cm at 0.4 V in H.sub.2/O.sub.2 at 1.0 barmore » back pressure.« less

  16. In-Field Performance Testing of the Fork Detector for Quantitative Spent Fuel Verification

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Gauld, Ian C.; Hu, Jianwei; De Baere, P.

    Expanding spent fuel dry storage activities worldwide are increasing demands on safeguards authorities that perform inspections. The European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) require measurements to verify declarations when spent fuel is transferred to difficult-to-access locations, such as dry storage casks and the repositories planned in Finland and Sweden. EURATOM makes routine use of the Fork detector to obtain gross gamma and total neutron measurements during spent fuel inspections. Data analysis is performed by modules in the integrated Review and Analysis Program (iRAP) software, developed jointly by EURATOM and the IAEA. Under the frameworkmore » of the US Department of Energy–EURATOM cooperation agreement, a module for automated Fork detector data analysis has been developed by Oak Ridge National Laboratory (ORNL) using the ORIGEN code from the SCALE code system and implemented in iRAP. EURATOM and ORNL recently performed measurements on 30 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel (Clab), operated by the Swedish Nuclear Fuel and Waste Management Company (SKB). The measured assemblies represent a broad range of fuel characteristics. Neutron count rates for 15 measured pressurized water reactor assemblies are predicted with an average relative standard deviation of 4.6%, and gamma signals are predicted on average within 2.6% of the measurement. The 15 measured boiling water reactor assemblies exhibit slightly larger deviations of 5.2% for the gamma signals and 5.7% for the neutron count rates, compared to measurements. These findings suggest that with improved analysis of the measurement data, existing instruments can provide increased verification of operator declarations of the spent fuel and thereby also provide greater ability to confirm integrity of an assembly. These results support the application of the Fork detector as a fully quantitative spent fuel verification technique.« less

  17. Fuel burner and combustor assembly for a gas turbine engine

    DOEpatents

    Leto, Anthony

    1983-01-01

    A fuel burner and combustor assembly for a gas turbine engine has a housing within the casing of the gas turbine engine which housing defines a combustion chamber and at least one fuel burner secured to one end of the housing and extending into the combustion chamber. The other end of the fuel burner is arranged to slidably engage a fuel inlet connector extending radially inwardly from the engine casing so that fuel is supplied, from a source thereof, to the fuel burner. The fuel inlet connector and fuel burner coact to anchor the housing against axial movement relative to the engine casing while allowing relative radial movement between the engine casing and the fuel burner and, at the same time, providing fuel flow to the fuel burner. For dual fuel capability, a fuel injector is provided in said fuel burner with a flexible fuel supply pipe so that the fuel injector and fuel burner form a unitary structure which moves with the fuel burner.

  18. Determining initial enrichment, burnup, and cooling time of pressurized-water reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Favalli, Andrea; Vo, D.; Grogan, Brandon R.

    The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuelmore » assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/ 137Cs, 134Cs/ 137Cs, 106Ru/ 137Cs, and 144Ce/ 137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. Furthermore, the results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.« less

  19. Determining initial enrichment, burnup, and cooling time of pressurized-water reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    DOE PAGES

    Favalli, Andrea; Vo, D.; Grogan, Brandon R.; ...

    2016-02-26

    The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuelmore » assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/ 137Cs, 134Cs/ 137Cs, 106Ru/ 137Cs, and 144Ce/ 137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. Furthermore, the results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.« less

  20. Temporal switching of an amphiphilic self-assembly by a chemical fuel-driven conformational response† †Electronic supplementary information (ESI) available: Experimental section, synthetic procedures and supporting figures. See DOI: 10.1039/c7sc01730h Click here for additional data file.

    PubMed Central

    Jalani, Krishnendu; Dhiman, Shikha

    2017-01-01

    The spatial and temporal control of self-assemblies is the latest scientific hurdle in supramolecular chemistry which is inspired by the functioning of biological systems fueled by chemical signals. In this study, we work towards alleviating this scenario by employing a unique amphiphilic foldamer that operates under the effect of a chemical fuel. The conformational changes in the foldamer amplify into observable morphological changes in its amphiphilic assembly that are controlled by external molecular cues (fuel). We take advantage of this redox responsive foldamer to affect its conformation in a temporal manner by an enzymatic pathway. The temporal characteristics of the transient conformation/assembly can be modulated by varying the concentrations of the fuel and enzyme. We believe that such a design strategy can have positive consequences in designing molecular and supramolecular systems for future active, adaptive and autonomous materials. PMID:28989632

  1. Performance of Transuranic-Loaded Fully Ceramic Micro-Encapsulated Fuel in LWRs Final Report, Including Void Reactivity Evaluation

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Michael A. Pope; R. Sonat Sen; Brian Boer

    2011-09-01

    The current focus of the Deep Burn Project is on once-through burning of transuranics (TRU) in light-water reactors (LWRs). The fuel form is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the tri-isotropic (TRISO) fuel particle design from high-temperature reactor technology. In the Deep Burn LWR (DB-LWR) concept, these fuel particles are pressed into compacts using SiC matrix material and loaded into fuel pins for use in conventional LWRs. The TRU loading comes from the spent fuel of a conventional LWR after 5 years of cooling. Unit cell and assembly calculations have been performed using the DRAGON-4 code tomore » assess the physics attributes of TRU-only FCM fuel in an LWR lattice. Depletion calculations assuming an infinite lattice condition were performed with calculations of various reactivity coefficients performed at each step. Unit cells and assemblies containing typical UO2 and mixed oxide (MOX) fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Then, assembly calculations were performed evaluating the performance of heterogeneous arrangements of TRU-only FCM fuel pins along with UO2 pins.« less

  2. Assessment of Silicon Carbide Composites for Advanced Salt-Cooled Reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Katoh, Yutai; Wilson, Dane F; Forsberg, Charles W

    2007-09-01

    The Advanced High-Temperature Reactor (AHTR) is a new reactor concept that uses a liquid fluoride salt coolant and a solid high-temperature fuel. Several alternative fuel types are being considered for this reactor. One set of fuel options is the use of pin-type fuel assemblies with silicon carbide (SiC) cladding. This report provides (1) an initial viability assessment of using SiC as fuel cladding and other in-core components of the AHTR, (2) the current status of SiC technology, and (3) recommendations on the path forward. Based on the analysis of requirements, continuous SiC fiber-reinforced, chemically vapor-infiltrated SiC matrix (CVI SiC/SiC) compositesmore » are recommended as the primary option for further study on AHTR fuel cladding among various industrially available forms of SiC. Critical feasibility issues for the SiC-based AHTR fuel cladding are identified to be (1) corrosion of SiC in the candidate liquid salts, (2) high dose neutron radiation effects, (3) static fatigue failure of SiC/SiC, (4) long-term radiation effects including irradiation creep and radiation-enhanced static fatigue, and (5) fabrication technology of hermetic wall and sealing end caps. Considering the results of the issues analysis and the prospects of ongoing SiC research and development in other nuclear programs, recommendations on the path forward is provided in the order or priority as: (1) thermodynamic analysis and experimental examination of SiC corrosion in the candidate liquid salts, (2) assessment of long-term mechanical integrity issues using prototypical component sections, and (3) assessment of high dose radiation effects relevant to the anticipated operating condition.« less

  3. PBF Reactor Building (PER620). PBF crane holds fuel test assembly ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    PBF Reactor Building (PER-620). PBF crane holds fuel test assembly aloft prior to lowering into reactor for test. Date: 1982. INEEL negative no. 82-4909 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  4. System for fuel rod removal from a reactor module

    DOEpatents

    Matchett, R.L.; Fodor, G.; Kikta, T.J.; Bacvinsicas, W.S.; Roof, D.R.; Nilsen, R.J.; Wilczynski, R.

    1988-07-28

    A robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system. 7 figs.

  5. System for fuel rod removal from a reactor module

    DOEpatents

    Matchett, Richard L.; Roof, David R.; Kikta, Thomas J.; Wilczynski, Rosemarie; Nilsen, Roy J.; Bacvinskas, William S.; Fodor, George

    1990-01-01

    A robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system.

  6. Method for monitoring irradiated fuel using Cerenkov radiation

    DOEpatents

    Dowdy, E.J.; Nicholson, N.; Caldwell, J.T.

    1980-05-21

    A method is provided for monitoring irradiated nuclear fuel inventories located in a water-filled storage pond wherein the intensity of the Cerenkov radiation emitted from the water in the vicinity of the nuclear fuel is measured. This intensity is then compared with the expected intensity for nuclear fuel having a corresponding degree of irradiation exposure and time period after removal from a reactor core. Where the nuclear fuel inventory is located in an assembly having fuel pins or rods with intervening voids, the Cerenkov light intensity measurement is taken at selected bright sports corresponding to the water-filled interstices of the assembly in the water storage, the water-filled interstices acting as Cerenkov light channels so as to reduce cross-talk. On-line digital analysis of an analog video signal is possible, or video tapes may be used for later measurement using a video editor and an electrometer. Direct measurement of the Cerenkov radiation intensity also is possible using spot photometers pointed at the assembly.

  7. Pulverized fuel-oxygen burner

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Taylor, Curtis; Patterson, Brad; Perdue, Jayson

    A burner assembly combines oxygen and fuel to produce a flame. The burner assembly includes an oxygen supply tube adapted to receive a stream of oxygen and a solid fuel conduit arranged to extend through the oxygen tube to convey a stream of fluidized, pulverized, solid fuel into a flame chamber. Oxygen flowing through the oxygen supply tube passes generally tangentially through a first set of oxygen-injection holes formed in the solid fuel conduit and off-tangentially from a second set of oxygen-injection holes formed in the solid fuel conduit and then mixes with fluidized, pulverized, solid fuel passing through themore » solid fuel conduit to create an oxygen-fuel mixture in a downstream portion of the solid fuel conduit. This mixture is discharged into a flame chamber and ignited in the flame chamber to produce a flame.« less

  8. A NEUTRONIC REACTOR

    DOEpatents

    Luebke, E.A.; Vandenberg, L.B.

    1959-09-01

    A nuclear reactor for producing thermoelectric power is described. The reactor core comprises a series of thermoelectric assemblies, each assembly including fissionable fuel as an active element to form a hot junction and a thermocouple. The assemblies are disposed parallel to each other to form spaces and means are included for Introducing an electrically conductive coolant between the assemblies to form cold junctions of the thermocouples. An electromotive force is developed across the entire series of the thermoelectric assemblies due to fission heat generated in the fuel causing a current to flow perpendicular to the flow of coolant and is distributed to a load outside of the reactor by means of bus bars electrically connected to the outermost thermoelectric assembly.

  9. Development of an extended-burnup Mark B design. First semi-annual progress report, July-December 1978. Report BAW-1532-1. [PWR

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    None

    1979-10-01

    The primary objective of this program is to develop and demonstrate an improved PWR fuel assembly design capable of batch average burnups of 45,000-50,000 MWd/mtU. To accomplish this, a number of technical areas must be investigated to verify acceptable extended-burnup fuel performance. This report is the first semi-annual progress report for the program, and it describes work performed during the July-December 1978 time period. Efforts during this period included the definition of a preliminary design for a high-burnup fuel rod, physics analyses of extended-burnup fuel cycles, studies of the physics characteristics of changes in fuel assembly metal-to-water ratios, and developmentmore » of a design concept for post-irradiation examination equipment to be utilized in examining high-burnup lead-test assemblies.« less

  10. Two-phase pressure drop reduction BWR assembly design

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Dix, G.E.; Crowther, R.L.; Colby, M.J.

    1992-05-12

    This patent describes a boiling water reactor having discrete bundles of fuel rods confined within channel enclosed fuel assemblies, an improvement to a fuel bundle assembly for placement in the reactor. It comprises a fuel channel having vertically extending walls forming a continuous channel around a fuel assembly volume, the channel being open at the bottom end for engagement to a lower tie plate and open at the upper end for engagement to an upper tie plate; rods for placement within the chamber, each the rod containing fissile material for producing nuclear reaction when in the presence of sufficient moderatedmore » neutron flux; a lower tie plate for supporting the bundle of rods within the channel, the lower tie plate for supporting the bundle of rods within the channel, the lower tie plate joining the bottom of the channel to close the bottom end of the channel, the lower tie plate providing defined apertures for the inflow of water in the channel between the rods for the generating of steam during the nuclear reaction; the plurality of fuel rods extending from the lower tie plate wherein a single phase region of the water in the bundle is defined to an upward portion of the bundle wherein a two phase region of the water and steam in the bundle is defined during nuclear steam generating reaction in the fuel bundle.« less

  11. Air Breathing Direct Methanol Fuel Cell

    DOEpatents

    Ren; Xiaoming

    2003-07-22

    A method for activating a membrane electrode assembly for a direct methanol fuel cell is disclosed. The method comprises operating the fuel cell with humidified hydrogen as the fuel followed by running the fuel cell with methanol as the fuel.

  12. Interim status report on lead-cooled fast reactor (LFR) research and development.

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Tzanos, C. P.; Sienicki, J. J.; Moisseytsev, A.

    2008-03-31

    This report discusses the status of Lead-Cooled Fast Reactor (LFR) research and development carried out during the first half of FY 2008 under the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. Lead-Cooled Fast Reactor research and development has recently been transferred from Generation IV to the Reactor Campaign of the Global Nuclear Energy Partnership (GNEP). Another status report shall be issued at the end of FY 2008 covering all of the LFR activities carried out in FY 2008 for both Generation IV and GNEP. The focus of research and development in FY 2008 is an initial investigationmore » of a concept for a LFR Advanced Recycling Reactor (ARR) Technology Pilot Plant (TPP)/demonstration test reactor (demo) incorporating features and operating conditions of the European Lead-cooled SYstem (ELSY) {approx} 600 MWe lead (Pb)-cooled LFR preconceptual design for the transmutation of waste and central station power generation, and which would enable irradiation testing of advanced fuels and structural materials. Initial scoping core concept development analyses have been carried out for a 100 MWt core composed of sixteen open-lattice 20 by 20 fuel assemblies largely similar to those of the ELSY preconceptual fuel assembly design incorporating fuel pins with mixed oxide (MOX) fuel, central control rods in each fuel assembly, and cooled with Pb coolant. For a cycle length of three years, the core is calculated to have a conversion ratio of 0.79, an average discharge burnup of 108 MWd/kg of heavy metal, and a burnup reactivity swing of about 13 dollars. With a control rod in each fuel assembly, the reactivity worth of an individual rod would need to be significantly greater than one dollar which is undesirable for postulated rod withdrawal reactivity insertion events. A peak neutron fast flux of 2.0 x 10{sup 15} (n/cm{sup 2}-s) is calculated. For comparison, the 400 MWt Fast Flux Test Facility (FFTF) achieved a peak neutron fast flux of 7.2 x 10{sup 15} (n/cm{sup 2}-s) and the initially 563 MWt PHENIX reactor attained 2.0 x 10{sup 15} (n/cm{sup 2}-s) before one of three intermediate cooling loops was shut down due to concerns about potential steam generator tube failures. The calculations do not assume a test assembly location for advanced fuels and materials irradiation in place of a fuel assembly (e.g., at the center of the core); the calculations have not examined whether it would be feasible to replace the central assembly by a test assembly location. However, having only fifteen driver assemblies implies a significant effect due to perturbations introduced by the test assembly. The peak neutron fast flux is low compared with the fast fluxes previously achieved in FFTF and PHENIX. Furthermore, the peak neutron fluence is only about half of the limiting value (4 x 10{sup 23} n/cm{sup 2}) typically used for ferritic steels. The results thus suggest that a larger power level (e.g., 400 MWt) and a larger core would be better for a TPP based upon the ELSY fuel assembly design and which can also perform irradiation testing of advanced fuels and materials. In particular, a core having a higher power level and larger dimensions would achieve a suitable average discharge burnup, peak fast flux, peak fluence, and would support the inclusion of one or more test assembly locations. Participation in the Generation IV International Forum Provisional System Steering Committee for the LFR is being maintained throughout FY 2008. Results from the analysis of samples previously exposed to flowing lead-bismuth eutectic (LBE) in the DELTA loop are summarized and a model for the oxidation/corrosion kinetics of steels in heavy liquid metal coolants was applied to systematically compare the calculated long-term (i.e., following several years of growth) oxide layer thicknesses of several steels.« less

  13. Advancing the Fork detector for quantitative spent nuclear fuel verification

    DOE PAGES

    Vaccaro, S.; Gauld, I. C.; Hu, J.; ...

    2018-01-31

    The Fork detector is widely used by the safeguards inspectorate of the European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) to verify spent nuclear fuel. Fork measurements are routinely performed for safeguards prior to dry storage cask loading. Additionally, spent fuel verification will be required at the facilities where encapsulation is performed for acceptance in the final repositories planned in Sweden and Finland. The use of the Fork detector as a quantitative instrument has not been prevalent due to the complexity of correlating the measured neutron and gamma ray signals with fuel inventories and operator declarations.more » A spent fuel data analysis module based on the ORIGEN burnup code was recently implemented to provide automated real-time analysis of Fork detector data. This module allows quantitative predictions of expected neutron count rates and gamma units as measured by the Fork detectors using safeguards declarations and available reactor operating data. This study describes field testing of the Fork data analysis module using data acquired from 339 assemblies measured during routine dry cask loading inspection campaigns in Europe. Assemblies include both uranium oxide and mixed-oxide fuel assemblies. More recent measurements of 50 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel are also analyzed. An evaluation of uncertainties in the Fork measurement data is performed to quantify the ability of the data analysis module to verify operator declarations and to develop quantitative go/no-go criteria for safeguards verification measurements during cask loading or encapsulation operations. The goal of this approach is to provide safeguards inspectors with reliable real-time data analysis tools to rapidly identify discrepancies in operator declarations and to detect potential partial defects in spent fuel assemblies with improved reliability and minimal false positive alarms. Finally, the results are summarized, and sources and magnitudes of uncertainties are identified, and the impact of analysis uncertainties on the ability to confirm operator declarations is quantified.« less

  14. Advancing the Fork detector for quantitative spent nuclear fuel verification

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Vaccaro, S.; Gauld, I. C.; Hu, J.

    The Fork detector is widely used by the safeguards inspectorate of the European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) to verify spent nuclear fuel. Fork measurements are routinely performed for safeguards prior to dry storage cask loading. Additionally, spent fuel verification will be required at the facilities where encapsulation is performed for acceptance in the final repositories planned in Sweden and Finland. The use of the Fork detector as a quantitative instrument has not been prevalent due to the complexity of correlating the measured neutron and gamma ray signals with fuel inventories and operator declarations.more » A spent fuel data analysis module based on the ORIGEN burnup code was recently implemented to provide automated real-time analysis of Fork detector data. This module allows quantitative predictions of expected neutron count rates and gamma units as measured by the Fork detectors using safeguards declarations and available reactor operating data. This study describes field testing of the Fork data analysis module using data acquired from 339 assemblies measured during routine dry cask loading inspection campaigns in Europe. Assemblies include both uranium oxide and mixed-oxide fuel assemblies. More recent measurements of 50 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel are also analyzed. An evaluation of uncertainties in the Fork measurement data is performed to quantify the ability of the data analysis module to verify operator declarations and to develop quantitative go/no-go criteria for safeguards verification measurements during cask loading or encapsulation operations. The goal of this approach is to provide safeguards inspectors with reliable real-time data analysis tools to rapidly identify discrepancies in operator declarations and to detect potential partial defects in spent fuel assemblies with improved reliability and minimal false positive alarms. Finally, the results are summarized, and sources and magnitudes of uncertainties are identified, and the impact of analysis uncertainties on the ability to confirm operator declarations is quantified.« less

  15. Advancing the Fork detector for quantitative spent nuclear fuel verification

    NASA Astrophysics Data System (ADS)

    Vaccaro, S.; Gauld, I. C.; Hu, J.; De Baere, P.; Peterson, J.; Schwalbach, P.; Smejkal, A.; Tomanin, A.; Sjöland, A.; Tobin, S.; Wiarda, D.

    2018-04-01

    The Fork detector is widely used by the safeguards inspectorate of the European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) to verify spent nuclear fuel. Fork measurements are routinely performed for safeguards prior to dry storage cask loading. Additionally, spent fuel verification will be required at the facilities where encapsulation is performed for acceptance in the final repositories planned in Sweden and Finland. The use of the Fork detector as a quantitative instrument has not been prevalent due to the complexity of correlating the measured neutron and gamma ray signals with fuel inventories and operator declarations. A spent fuel data analysis module based on the ORIGEN burnup code was recently implemented to provide automated real-time analysis of Fork detector data. This module allows quantitative predictions of expected neutron count rates and gamma units as measured by the Fork detectors using safeguards declarations and available reactor operating data. This paper describes field testing of the Fork data analysis module using data acquired from 339 assemblies measured during routine dry cask loading inspection campaigns in Europe. Assemblies include both uranium oxide and mixed-oxide fuel assemblies. More recent measurements of 50 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel are also analyzed. An evaluation of uncertainties in the Fork measurement data is performed to quantify the ability of the data analysis module to verify operator declarations and to develop quantitative go/no-go criteria for safeguards verification measurements during cask loading or encapsulation operations. The goal of this approach is to provide safeguards inspectors with reliable real-time data analysis tools to rapidly identify discrepancies in operator declarations and to detect potential partial defects in spent fuel assemblies with improved reliability and minimal false positive alarms. The results are summarized, and sources and magnitudes of uncertainties are identified, and the impact of analysis uncertainties on the ability to confirm operator declarations is quantified.

  16. Proton Exchange Membrane (PEM) Fuel Cell Status and Remaining Challenges for Manned Space-Flight Applications

    NASA Technical Reports Server (NTRS)

    Reaves, Will F.; Hoberecht, Mark A.

    2003-01-01

    The Fuel Cell has been used for manned space flight since the Gemini program. Its power output and water production capability over long durations for the mass and volume are critical for manned space-flight requirements. The alkaline fuel cell used on the Shuttle, while very reliable and capable for it s application, has operational sensitivities, limited life, and an expensive recycle cost. The PEM fuel cell offers many potential improvements in those areas. NASA Glenn Research Center is currently leading a PEM fuel cell development and test program intended to move the technology closer to the point required for manned space-flight consideration. This paper will address the advantages of PEM fuel cell technology and its potential for future space flight as compared to existing alkaline fuel cells. It will also cover the technical hurdles that must be overcome. In addition, a description of the NASA PEM fuel cell development program will be presented, and the current status of this effort discussed. The effort is a combination of stack and ancillary component hardware development, culminating in breadboard and engineering model unit assembly and test. Finally, a detailed roadmap for proceeding fiom engineering model hardware to qualification and flight hardware will be proposed. Innovative test engineering and potential payload manifesting may be required to actually validate/certify a PEM fuel cell for manned space flight.

  17. REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Nichols, T.; Beals, D.; Sternat, M.

    2011-07-18

    Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Manymore » research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical analysis. The fuel assembly was modeled using MONTEBURNS(MCNP5/ ORIGEN2.2) and MCNPX/CINDER90. The results from the models have been compared to each other and to the measured data.« less

  18. Method and apparatus for selectively controlling the speed of an engine

    DOEpatents

    Davis, Roy Inge

    2001-02-27

    A control assembly 12 for use within a vehicle 10 having an engine 14 and which selectively controls the speed of the engine 14 in order to increase fuel efficiency and to effect relatively smooth starting and stopping of the engine. Particularly, in one embodiment, control assembly 12 cooperatively operates with a starter/alternator assembly 20 and is adapted for use with hybrid vehicles employing a start/stop powertrain assembly, wherein fuel efficiency is increased by selectively stopping engine operation when the vehicle has stopped.

  19. Thorium Fuel Options for Sustained Transuranic Burning in Pressurized Water Reactors - 12381

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Rahman, Fariz Abdul; Lee, John C.; Franceschini, Fausto

    2012-07-01

    As described in companion papers, Westinghouse is proposing the adoption of a thorium-based fuel cycle to burn the transuranics (TRU) contained in the current Used Nuclear Fuel (UNF) and transition towards a less radio-toxic high level waste. A combination of both light water reactors (LWR) and fast reactors (FR) is envisaged for the task, with the emphasis initially posed on their TRU burning capability and eventually to their self-sufficiency. Given the many technical challenges and development times related to the deployment of TRU burners fast reactors, an interim solution making best use of the current resources to initiate burning themore » legacy TRU inventory while developing and testing some technologies of later use is desirable. In this perspective, a portion of the LWR fleet can be used to start burning the legacy TRUs using Th-based fuels compatible with the current plants and operational features. This analysis focuses on a typical 4-loop PWR, with 17x17 fuel assembly design and TRUs (or Pu) admixed with Th (similar to U-MOX fuel, but with Th instead of U). Global calculations of the core were represented with unit assembly simulations using the Linear Reactivity Model (LRM). Several assembly configurations have been developed to offer two options that can be attractive during the TRU transmutation campaign: maximization of the TRU transmutation rate and capability for TRU multi-recycling, to extend the option of TRU recycling in LWR until the FR is available. Homogeneous as well as heterogeneous assembly configurations have been developed with various recycling schemes (Pu recycle, TRU recycle, TRU and in-bred U recycle etc.). Oxide as well as nitride fuels have been examined. This enabled an assessment of the potential for burning and multi-recycling TRU in a Th-based fuel PWR to compare against other more typical alternatives (U-MOX and variations thereof). Results will be shown indicating that Th-based PWR fuel is a promising option to multi-recycle and burn TRU in a thermal spectrum, while satisfying top-level operational and safety constraints. Various assembly designs have been proposed to assess the TRU burning potential of Th-based fuel in PWRs. In addition to typical homogeneous loading patterns, heterogeneous configurations exploiting the breeding potential of thorium to enable multiple cycles of TRU irradiation and burning have been devised. The homogeneous assembly design, with all pins featuring TRU in Th, has the benefit of a simple loading pattern and the highest rate of TRU transmutation, but it can be used only for a few cycles due to the rapid rise in the TRU content of the recycled fuel, which challenges reactivity control, safety coefficients and fuel handling. Due to its simple loading pattern, such assembly design can be used as the first step of Th implementation, achieving up to 3 times larger TRU transmutation rate than conventional U-MOX, assuming same fraction of MOX assemblies in the core. As the next step in thorium implementation, heterogeneous assemblies featuring a mixed array of Th-U and Th-U-TRU pins, where the U is in-bred from Th, have been proposed. These designs have the potential to enable burning an external supply of TRU through multiple cycles of irradiation, recovery (via reprocessing) and recycling of the residual actinides at the end of each irradiation cycle. This is achieved thanks to a larger breeding of U from Th in the heterogeneous assemblies, which reduces the TRU supply and thus mitigates the increase in the TRU core inventory for the multi-recycled fuel. While on an individual cycle basis the amount of TRU burned in the heterogeneous assembly is reduced with respect to the homogeneous design, TRU burning rates higher than single-pass U-MOX fuel can still be achieved, with the additional benefits of a multi-cycle transmutation campaign recycling all TRU isotopes. Nitride fuel, due its higher density and U breeding potential, together with its better thermal properties, ideally suits the objectives and constraints of the heterogeneous assemblies. However, significant technological advancements must be made before nitride fuels can be employed in an LWR: its water resistance needs to be improved and a viable technology to enrich N in N-15 must be devised. Moreover, for the nitride heterogeneous configurations examined in this study, the enhancement in TRU burning performance is achieved not only by replacing oxide with nitride fuel, but also by increasing the fuel rod size. This latter modification, allowed by the high thermal conductivity of nitride fuel, leads however to a very tight lattice, which may challenge reactor coolant pumps and assembly hold-down mechanisms, the former through an increase in core pressure drop and the latter through an increase in assembly lift-off forces. To alleviate these issues, while still achieving the large fuel-to-moderator ratios resulting from using tight lattices, wire wraps could be used in place of grid spacers. For tight lattices, typical grid spacers are hard to manufacture and their replacement with wire wraps is known to allow for a pressure drop reduction by at least 2 times. The studies, while certainly very preliminary, provide a starting point to devise an optimum strategy for TRU transmutation in Th-based PWR fuel. The viability of the scheme proposed depends on the timely phasing in of the associated technologies, with proper lead time and to solve the many challenges. These challenges are certainly substantial, and make the current once-through U-based scheme pursued in the US by far a more practical (and cheaper) option. However, when compared to other transmutation schemes, the proposed one has arguably similar challenges and unknowns with potentially bigger rewards. (authors)« less

  20. CHF considerations for highly moderated 100% MOX fuels PWRs

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Saphier, D.; Raymond, P.

    1995-09-01

    A feasibility study on using 100% MOX fuel in a PWR with increased moderating ratio, RMA, was initiated. In the proposed design all the parameters were chosen identical to the French 1450MW PWR, except the fuel pin diameter which was reduced to achieve higher moderating ratios, V{sub M}/V{sub F}, where V{sub M} and V{sub F} are the moderator and fuel volume respectively. Moderating ratios from 2 to 4 were considered. In the present study the thermal-hydraulic feasibility of using fuel assemblies with smaller diameter fuel pins was investigated. The major design constrain in this study was the critical heat fluxmore » (CHF). In order to maintain the fuel pin integrity under nominal operating and transient conditions, the minimum DNBR, (Departure from Nucleate Boiling Ratio given by CHF/q{close_quotes}{sub local}, where q{close_quotes}{sub local} is the local heat flux), has to be above a given value. The limitations of the existing CHF correlations for the present study are outlined. Two designs based on the conventional 17x17 fuel assembly and on the advanced 19x19 assembly meeting the MDNBR criteria and satisfying the control margin requirements, are proposed.« less

  1. Fuel cell with electrolyte feed system

    DOEpatents

    Feigenbaum, Haim

    1984-01-01

    A fuel cell having a pair of electrodes at the sites of electrochemical reactions of hydrogen and oxygen and a phosphoric acid electrolyte provided with an electrolyte supporting structure in the form of a laminated matrix assembly disposed between the electrodes. The matrix assembly is formed of a central layer disposed between two outer layers, each being permeable to the flow of the electrolyte. The central layer is provided with relatively large pores while the outer layers are provided with relatively small pores. An external reservoir supplies electrolyte via a feed means to the central layer to compensate for changes in electrolyte volume in the matrix assembly during the operation of fuel cell.

  2. KSC-2009-5811

    NASA Image and Video Library

    2009-10-24

    CAPE CANAVERAL, Fla. – At Port Canaveral, a tugboat relieves the Liberty Star for the move of the Pegasus barge, carrying external tank 134, through the Banana River toward the turn basin near the Vehicle Assembly Building, or VAB, at NASA's Kennedy Space Center in Florida. Pegasus arrived in Florida after an ocean voyage towed by the solid rocket booster retrieval ship from NASA's Michoud Assembly Facility near New Orleans. After Pegasus docks in the turn basin, the fuel tank will be offloaded and transported into the VAB. ET-134 will be used to launch space shuttle Endeavour on the STS-130 mission to the International Space Station. Launch is targeted for Feb. 4, 2010. For information on the components of the space shuttle and the STS-130 mission, visit http://www.nasa.gov/shuttle. Photo credit: NASA/Jack Pfaller

  3. KSC-2009-5812

    NASA Image and Video Library

    2009-10-24

    CAPE CANAVERAL, Fla. – At Port Canaveral, a tugboat begins to push the Pegasus barge, carrying external tank 134, through the Banana River toward the turn basin near the Vehicle Assembly Building, or VAB, at NASA's Kennedy Space Center in Florida. Pegasus arrived in Florida after an ocean voyage towed by a solid rocket booster retrieval ship from NASA's Michoud Assembly Facility near New Orleans. After Pegasus docks in the turn basin, the fuel tank will be offloaded and transported into the VAB. ET-134 will be used to launch space shuttle Endeavour on the STS-130 mission to the International Space Station. Launch is targeted for Feb. 4, 2010. For information on the components of the space shuttle and the STS-130 mission, visit http://www.nasa.gov/shuttle. Photo credit: NASA/Jack Pfaller

  4. KSC-2009-5813

    NASA Image and Video Library

    2009-10-24

    CAPE CANAVERAL, Fla. – At Port Canaveral, a tugboat begins to move the Pegasus barge, carrying external tank 134, through the Banana River toward the turn basin near the Vehicle Assembly Building, or VAB, at NASA's Kennedy Space Center in Florida. Pegasus arrived in Florida after an ocean voyage towed by a solid rocket booster retrieval ship from NASA's Michoud Assembly Facility near New Orleans. After Pegasus docks in the turn basin, the fuel tank will be offloaded and transported into the VAB. ET-134 will be used to launch space shuttle Endeavour on the STS-130 mission to the International Space Station. Launch is targeted for Feb. 4, 2010. For information on the components of the space shuttle and the STS-130 mission, visit http://www.nasa.gov/shuttle. Photo credit: NASA/Jack Pfaller

  5. Solar receiver heliostat reflector having a linear drive and position information system

    DOEpatents

    Horton, Richard H.

    1980-01-01

    A heliostat for a solar receiver system comprises an improved drive and control system for the heliostat reflector assembly. The heliostat reflector assembly is controllably driven in a predetermined way by a light-weight drive system so as to be angularly adjustable in both elevation and azimuth to track the sun and efficiently continuously reflect the sun's rays to a focal zone, i.e., heat receiver, which forms part of a solar energy utilization system, such as a solar energy fueled electrical power generation system. The improved drive system includes linear stepping motors which comprise low weight, low cost, electronic pulse driven components. One embodiment comprises linear stepping motors controlled by a programmed, electronic microprocessor. Another embodiment comprises a tape driven system controlled by a position control magnetic tape.

  6. HEU Holdup Measurements in 321-M B and Spare U-Al Casting Furnaces

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Salaymeh, S.R.

    The Analytical Development Section of Savannah River Technology Center (SRTC) was requested by the Facilities Decontamination Division (FDD) to determine the holdup of enriched uranium in the 321-M facility as part of an overall deactivation project of the facility. The 321-M facility was used to fabricate enriched uranium fuel assemblies, lithium-aluminum target tubes, neptunium assemblies, and miscellaneous components for the production reactors. This report covers holdup measurements in two uranium aluminum alloy (U-Al) casting furnaces. Our results indicate an upper limit of 235U content for the B and Spare furnaces of 51 and 67 g respectively. This report discusses themore » methodology, non-destructive assay (NDA) measurements, and results of the uranium holdup on the two furnaces.« less

  7. A new method to measure the U-235 content in fresh LWR fuel assemblies via fast-neutron passive self-interrogation

    DOE PAGES

    Menlove, Howard Olsen; Belian, Anthony P.; Geist, William H.; ...

    2017-10-07

    The purpose of this paper is to provide a solution to a decades old safeguards problem in the verification of the fissile concentration in fresh light water reactor (LWR) fuel assemblies. The problem is that the burnable poison (e.g. Gd 2O 3) addition to the fuel rods decreases the active neutron assay for the fuel assemblies. This paper presents a new innovative method for the verification of the 235U linear mass density in fresh LEU fuel assemblies that is insensitive to the burnable poison content. The technique makes use of the 238U atoms in the fuel rods to self-interrogate themore » 235U mass. The innovation for the new approach is that the 238U spontaneous fission (SF) neutrons from the rods induces fission reactions (IF) in the 235U that are time correlated with the SF source neutrons. Thus, the coincidence gate counting rate benefits from both the nu-bar of the 238U SF (2.07) and the 235U IF (2.44) for a fraction of the IF reactions. Whereas, the 238U SF background has no time-correlation boost. The higher the detection efficiency, the higher the correlated boost because background neutron counts from the SF are being converted to signal doubles. This time-correlation in the IF signal increases signal/background ratio that provides a good precision for the net signal from the 235U mass. The hard neutron energy spectrum makes the technique insensitive to the burnable poison loading where a Cd or Gd liner on the detector walls is used to prevent thermal-neutron reflection back into the fuel assembly from the detector. Here, we have named the system the fast-neutron passive collar (FNPC).« less

  8. A new method to measure the U-235 content in fresh LWR fuel assemblies via fast-neutron passive self-interrogation

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Menlove, Howard Olsen; Belian, Anthony P.; Geist, William H.

    The purpose of this paper is to provide a solution to a decades old safeguards problem in the verification of the fissile concentration in fresh light water reactor (LWR) fuel assemblies. The problem is that the burnable poison (e.g. Gd 2O 3) addition to the fuel rods decreases the active neutron assay for the fuel assemblies. This paper presents a new innovative method for the verification of the 235U linear mass density in fresh LEU fuel assemblies that is insensitive to the burnable poison content. The technique makes use of the 238U atoms in the fuel rods to self-interrogate themore » 235U mass. The innovation for the new approach is that the 238U spontaneous fission (SF) neutrons from the rods induces fission reactions (IF) in the 235U that are time correlated with the SF source neutrons. Thus, the coincidence gate counting rate benefits from both the nu-bar of the 238U SF (2.07) and the 235U IF (2.44) for a fraction of the IF reactions. Whereas, the 238U SF background has no time-correlation boost. The higher the detection efficiency, the higher the correlated boost because background neutron counts from the SF are being converted to signal doubles. This time-correlation in the IF signal increases signal/background ratio that provides a good precision for the net signal from the 235U mass. The hard neutron energy spectrum makes the technique insensitive to the burnable poison loading where a Cd or Gd liner on the detector walls is used to prevent thermal-neutron reflection back into the fuel assembly from the detector. Here, we have named the system the fast-neutron passive collar (FNPC).« less

  9. A new method to measure the U-235 content in fresh LWR fuel assemblies via fast-neutron passive self-interrogation

    NASA Astrophysics Data System (ADS)

    Menlove, Howard; Belian, Anthony; Geist, William; Rael, Carlos

    2018-01-01

    The purpose of this paper is to provide a solution to a decades old safeguards problem in the verification of the fissile concentration in fresh light water reactor (LWR) fuel assemblies. The problem is that the burnable poison (e.g. Gd2O3) addition to the fuel rods decreases the active neutron assay for the fuel assemblies. This paper presents a new innovative method for the verification of the 235U linear mass density in fresh LEU fuel assemblies that is insensitive to the burnable poison content. The technique makes use of the 238U atoms in the fuel rods to self-interrogate the 235U mass. The innovation for the new approach is that the 238U spontaneous fission (SF) neutrons from the rods induces fission reactions (IF) in the 235U that are time correlated with the SF source neutrons. Thus, the coincidence gate counting rate benefits from both the nu-bar of the 238U SF (2.07) and the 235U IF (2.44) for a fraction of the IF reactions. Whereas, the 238U SF background has no time-correlation boost. The higher the detection efficiency, the higher the correlated boost because background neutron counts from the SF are being converted to signal doubles. This time-correlation in the IF signal increases signal/background ratio that provides a good precision for the net signal from the 235U mass. The hard neutron energy spectrum makes the technique insensitive to the burnable poison loading where a Cd or Gd liner on the detector walls is used to prevent thermal-neutron reflection back into the fuel assembly from the detector. We have named the system the fast-neutron passive collar (FNPC).

  10. Fuel cell system with separating structure bonded to electrolyte

    DOEpatents

    Bourgeois, Richard Scott; Gudlavalleti, Sauri; Quek, Shu Ching; Hasz, Wayne Charles; Powers, James Daniel

    2010-09-28

    A fuel cell assembly comprises a separating structure configured for separating a first reactant and a second reactant wherein the separating structure has an opening therein. The fuel cell assembly further comprises a fuel cell comprising a first electrode, a second electrode, and an electrolyte interposed between the first and second electrodes, and a passage configured to introduce the second reactant to the second electrode. The electrolyte is bonded to the separating structure with the first electrode being situated within the opening, and the second electrode being situated within the passage.

  11. Integral edge seals for phosphoric acid fuel cells

    DOEpatents

    Granata, Jr., Samuel J.; Woodle, Boyd M.; Dunyak, Thomas J.

    1992-01-01

    A phosphoric acid fuel cell having integral edge seals formed by an elastomer permeating an outer peripheral band contiguous with the outer peripheral edges of the cathode and anode assemblies and the matrix to form an integral edge seal which is reliable, easy to manufacture and has creep characteristics similar to the anode, cathode and matrix assemblies inboard of the seals to assure good electrical contact throughout the life of the fuel cell.

  12. Vibrating fuel grapple

    DOEpatents

    Chertock, deceased, Alan J.; Fox, Jack N.; Weissinger, Robert B.

    1982-01-01

    A reactor refueling method utilizing a vibrating fuel grapple for removing spent fuel assemblies from a reactor core which incorporates a pneumatic vibrator in the grapple head, enabling additional withdrawal capability without exceeding the allowable axial force limit. The only moving part in the vibrator is a steel ball, pneumatically driven by a gas, such as argon, around a track, with centrifugal force created by the ball being transmitted through the grapple to the assembly handling socket.

  13. Improved FCG-1 cell technology

    NASA Astrophysics Data System (ADS)

    Breault, R. D.; Congdon, J. V.; Coykendall, R. D.; Luoma, W. L.

    1980-10-01

    Fuel cell performance in the ribbed substrate cell configuration consistent with that projected for a commercial power plant is demonstrated. Tests were conducted on subscale cells and on two 20 cell stacks of 4.8 MW demonstrator size cell components. These tests evaluated cell stack materials, processes, components, and assembly configurations. The first task was to conduct a component development effort to introduce improvements in 3.7 square foot, ribbed substrate acid cell repeating parts which represented advances in performance, function, life, and lower cost for application in higher pressure and temperature power plants. Specific areas of change were the electrode substrate, catalyst, matrix, seals, separator plates, and coolers. Full sized ribbed substrate stack components incorporating more stable materials were evaluated at increased pressure (93 psia) and temperature (405 F) conditions. Two 20 cell stacks with a 3.7 square feet, ribbed substrate cell configuration were tested.

  14. Measuring the Multiplication of Spent Fuel Assemblies – It’s easier than you think!

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Tobin, Stephen Joseph

    This is a set of eight slides which advertise how easy it can be to measure the multiplication of a spent fuel assembly. A robust (fission chambers), rapid (under 15 minutes), direct (multiplication is measured, not photons from fission fragments) measurement of multiplication is possible.

  15. Validation of the U.S. NRC NGNP evaluation model with the HTTR

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Saller, T.; Seker, V.; Downar, T.

    2012-07-01

    The High Temperature Test Reactor (HTTR) was modeled with TRITON/PARCS. Traditional light water reactor (LWR) homogenization methods rely on the short mean free paths of neutrons in LWR. In gas-cooled, graphite-moderated reactors like the HTTR neutrons have much longer mean free paths and penetrate further into neighboring assemblies than in LWRs. Because of this, conventional lattice calculations with a single assembly may not be valid. In addition to difficulties caused by the longer mean free paths, the HTTR presents unique axial and radial heterogeneities that require additional modifications to the single assembly homogenization method. To handle these challenges, the homogenizationmore » domain is decreased while the computational domain is increased. Instead of homogenizing a single hexagonal fuel assembly, the assembly is split into six triangles on the radial plane and five blocks axially in order to account for the placement of burnable poisons. Furthermore, the radial domain is increased beyond a single fuel assembly to account for spectrum effects from neighboring fuel, reflector, and control rod assemblies. A series of five two-dimensional cases, each closer to the full core, were calculated to evaluate the effectiveness of the homogenization method and cross-sections. (authors)« less

  16. Development and applications of retro-reflective surfaces for ultrasound in LBE

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    NONE

    The Belgian Nuclear Research Centre SCK-CEN is in the process of developing MYRRHA - a fast flux research reactor to replace the aging BR2. MYRRHA is conceptualized as an accelerator driven system cooled by lead bismuth eutectic mixture (LBE). As LBE is opaque to visual light, ultrasonic measurement techniques are employed as the main technology to provide feedback to submerged operations when needed. Conceptually, MYRRHA is a pool type reactor divided in a cold lower plenum and hot upper plenum separated by a diaphragm that forces the main flow through the core. The main flow is cooled by four heatmore » exchangers and driven by two liquid metal pumps. One of the tasks tackled using ultrasound is locating a potentially lost fuel assembly to assist a recovery operation. As all fuel manipulations in MYRRHA are performed in the lower plenum, a potentially lost fuel assembly is located in the lower plenum. Buoyancy will force the lost fuel assembly to float against the diaphragm unless it is still partially inserted in the core. Because of the latter situation, an ultrasonic scan localizing the fuel assembly should be performed from a large distance to avoid a collision with such a partially inserted fuel assembly. Unfortunately, standard machined stainless steel objects, such as a fuel assembly, reflect an ultrasonic pulse in a specular way which induces stringent requirements on the alignment of the ultrasonic sensor with respect to the fuel assembly as we cannot rely on diffuse reflections and/or scattering of the ultrasonic pulse. Moreover, increasing the distance also increases geometric spreading and absorption of the pulse weakening the signal amplitude even faster to noise levels when deviating from perfect alignment. An alternative approach consists in relying on reflections from the known surroundings: a lost fuel assembly will block the line-of-sight to the diaphragm resulting in an anomaly in the reflection - either a shorter than expected time-of-flight of the pulse or a complete absence of a reflection like a shadow. In that way, it suffices to align the sensor with the diaphragm instead of the fuel assembly which is much easier to achieve as the robotics on which the sensor is mounted move parallel with the diaphragm. The alignment requirement in the latter approach can be further relaxed by using a tiling of retro-reflectors on the lower surface of the diaphragm. In that way, alignment becomes less vital and the main source of acoustic energy loss - geometric spread of the beam - is almost completely removed, leaving only absorption losses. In this paper, we present the first results in developing a retro reflectance surface for ultrasound in LBE. We present experimental results for different designs of retro-reflectors in both water and LBE. We discuss both linear and array retro-reflectors of different sizes and investigate the influence of the main relevant ultrasonic parameters such as wavelength and spot size on the strength of the received reflection under different alignment angles. We also demonstrate how retro-reflective surfaces can be exploited when localizing objects using linear and rotating scanning methods. (authors)« less

  17. Self-assembling hydrogel scaffolds for photocatalytic hydrogen production

    DOE PAGES

    Weingarten, Adam S.; Kazantsev, Roman V.; Palmer, Liam C.; ...

    2014-10-05

    Integration into a soft material of all the molecular components necessary to generate storable fuels is an interesting target in supramolecular chemistry. The concept is inspired by the internal structure of photosynthetic organelles, such as plant chloroplasts, which colocalize molecules involved in light absorption, charge transport and catalysis to create chemical bonds using light energy. We report in this paper on the light-driven production of hydrogen inside a hydrogel scaffold built by the supramolecular self-assembly of a perylene monoimide amphiphile. The charged ribbons formed can electrostatically attract a nickel-based catalyst, and electrolyte screening promotes gelation. We found the emergent phenomenonmore » that screening by the catalyst or the electrolytes led to two-dimensional crystallization of the chromophore assemblies and enhanced the electronic coupling among the molecules. Finally, photocatalytic production of hydrogen is observed in the three-dimensional environment of the hydrogel scaffold and the material is easily placed on surfaces or in the pores of solid supports.« less

  18. Development of molten carbonate fuel cell technology at M-C Power Corporation

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Dilger, D.

    1996-04-01

    M-C Power Corporation was founded in 1987 with the mission to further develop and subsequently commercialize molten carbonate fuel cells (MCFC). The technology chosen for commercialization was initially developed by the Institute of Gas technology (IGT). At the center of this MCFC technology is the Internally Manifolded Heat EXchange (IMHEX) separator plate design. The IMHEX technology design provides several functions within one component assembly. These functions include integrating the gas manifold structure into the fuel cell stack, separating the fuel gas stream from the oxidant gas stream, providing the required electrical contact between cells to achieve desired power output, andmore » removing excess heat generated in the electrochemical process. Development of this MCFC technology from lab-scale sizes too a commercial area size of 1m{sup 2} has focused our efforts an demonstrating feasibility and evolutionary progress. The development effort will culminate in a proof-of-concept- 250kW power plant demonstration in 1996. The remainder of our commercialization program focuses upon lowering the costs associated with the MCFC power plant system in low production volumes.« less

  19. Design and simulation of novel flow field plate geometry for proton exchange membrane fuel cells

    NASA Astrophysics Data System (ADS)

    Ruan, Hanxia; Wu, Chaoqun; Liu, Shuliang; Chen, Tao

    2016-10-01

    Bipolar plate is one of the many important components of proton exchange membrane fuel cell (PEMFC) stacks as it supplies fuel and oxidant to the membrane-electrode assembly (MEA), removes water, collects produced current and provides mechanical support for the single cells in the stack. The flow field design of a bipolar plate greatly affects the performance of a PEMFC. It must uniformly distribute the reactant gases over the MEA and prevent product water flooding. This paper aims at improving the fuel cell performance by optimizing flow field designs and flow channel configurations. To achieve this, a novel biomimetic flow channel for flow field designs is proposed based on Murray's Law. Computational fluid dynamics based simulations were performed to compare three different designs (parallel, serpentine and biomimetic channel, respectively) in terms of current density distribution, power density distribution, pressure distribution, temperature distribution, and hydrogen mass fraction distribution. It was found that flow field designs with biomimetic flow channel perform better than that with convectional flow channel under the same operating conditions.

  20. Fibre optic sensor based measurements of flow-induced vibration in a liquid metal cooled nuclear reactor set-up

    NASA Astrophysics Data System (ADS)

    De Pauw, B.; Vanlanduit, S.; Van Tichelen, K.; Geernaert, T.; Thienpont, H.; Berghmans, F.

    2017-04-01

    Fuel assembly vibrations in nuclear reactor cores should not be excessive as these can compromise the lifetime of the assembly and lead to safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants. We therefore demonstrate accurate measurements of the vibrations of a fuel assembly in a lead-bismuth eutectic cooled installation with fibre Bragg grating (FBG) based sensors. The use of FBGs in combination with a dedicated sensor integration approach allows accounting for the severe geometrical constraints and providing for the required minimal intrusiveness of the instrumentation, identifying the vibration modes with required accuracy and observing the differences between the vibration amplitudes of the individual fuel pins as well as evidencing a low frequency fuel pin vibration mode resulting from the supports.

  1. Consistent criticality and radiation studies of Swiss spent nuclear fuel: The CS2M approach.

    PubMed

    Rochman, D; Vasiliev, A; Ferroukhi, H; Pecchia, M

    2018-06-15

    In this paper, a new method is proposed to systematically calculate at the same time canister loading curves and radiation sources, based on the inventory information from an in-core fuel management system. As a demonstration, the isotopic contents of the assemblies come from a Swiss PWR, considering more than 6000 cases from 34 reactor cycles. The CS 2 M approach consists in combining four codes: CASMO and SIMULATE to extract the assembly characteristics (based on validated models), the SNF code for source emission and MCNP for criticality calculations for specific canister loadings. The considered cases cover enrichments from 1.9 to 5.0% for the UO 2 assemblies and 4.8% for the MOX, with assembly burnup values from 7 to 74 MWd/kgU. Because such a study is based on the individual fuel assembly history, it opens the possibility to optimize canister loadings from the point-of-view of criticality, decay heat and emission sources. Copyright © 2018 Elsevier B.V. All rights reserved.

  2. Closeup view of the bottom area of Space Shuttle Main ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    Close-up view of the bottom area of Space Shuttle Main Engine (SSME) 2052 engine assembly mounted in a SSME Engine Handler in the Horizontal Processing area of the SSME Processing Facility at Kennedy Space Center. The most prominent features in this view are the Low-Pressure Oxidizer Discharge Duct toward the bottom of the assembly, the SSME Engine Controller and the Main Fuel Valve Hydraulic Actuator are in the approximate center of the assembly in this view, the Low-Pressure Fuel Turbopump (LPFTP), the LPFTP Discharge Duct are to the left on the assembly in this view and the High-Pressure Fuel Turbopump is located toward the top of the engine assembly in this view. The ring of tabs in the right side of the image, at the approximate location of the Nozzle and the Coolant Outlet Manifold interface is the Heat Shield Support Ring. - Space Transportation System, Space Shuttle Main Engine, Lyndon B. Johnson Space Center, 2101 NASA Parkway, Houston, Harris County, TX

  3. Testing of uranium nitride fuel in T-111 cladding at 1200 K cladding temperature

    NASA Technical Reports Server (NTRS)

    Rohal, R. G.; Tambling, T. N.; Smith, R. L.

    1973-01-01

    Two groups of six fuel pins each were assembled, encapsulated, and irradiated in the Plum Brook Reactor. The fuel pins employed uranium mononitride (UN) in a tantalum alloy clad. The first group of fuel pins was irradiated for 1500 hours to a maximum burnup of 0.7-atom-percent uranium. The second group of fuel pins was irradiated for about 3000 hours to a maximum burnup of 1.0-atom-percent uranium. The average clad surface temperature during irradiation of both groups of fuel pins was approximately 1200 K. The postirradiation examination revealed the following: no clad failures or fuel swelling occurred; less than 1 percent of the fission gases escaped from the fuel; and the clad of the first group of fuel pins experienced clad embrittlement whereas the second group, which had modified assembly and fabrication procedures to minimize contamination, had a ductile clad after irradiation.

  4. Gamma and fast neutron radiation monitoring inside spent reactor fuel assemblies

    NASA Astrophysics Data System (ADS)

    Lakosi, L.; Tam Nguyen, C.

    2007-09-01

    Gamma and neutron signatures of spent reactor fuel were monitored by small-size silicon diode and track etch detectors, respectively, in a nuclear power plant (NPP). These signatures, reflecting gross gamma intensity and the 242,244Cm content, contain information on the burn-up (BU) and cooling time (CT) of the fuel. The small size of the detectors allows close access to inside parts of the assemblies out of reach of other methods. A commercial Si diode was encapsulated in a cylindrical steel case and was used for gross γ monitoring. CR-39 detectors were used for neutron measurements. Irradiation exposures at the NPP were implemented in the central dosimetric channel of spent fuel assemblies (SFAs) stored in borated water. Gross γ and neutron axial profiles were taken up by scanning with the aid of a long steel guide tube, lowered down to the spent fuel pond by crane and fitted to the headpiece of the fuel assemblies. Gamma measurements were performed using a long cable introduced in this tube, with the Si diode at the end. A long steel wire was also led through the guide tube, to which a chain of 15 sample holder capsules was attached, each containing a track detector. Gamma dose rates of 0.1-10 kGy h -1, while neutron fluxes in a range of (0.25-26) 10 4 cm -2 s -1 were recorded. The results are in good correlation with those of a calculation for spent fuel neutron yield.

  5. Outcomes of the JNT 1955 Phase I Viability Study of Gamma Emission Tomography for Spent Fuel Verification

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Jacobsson-Svard, Staffan; Smith, Leon E.; White, Timothy

    The potential for gamma emission tomography (GET) to detect partial defects within a spent nuclear fuel assembly has been assessed within the IAEA Support Program project JNT 1955, phase I, which was completed and reported to the IAEA in October 2016. Two safeguards verification objectives were identified in the project; (1) independent determination of the number of active pins that are present in a measured assembly, in the absence of a priori information about the assembly; and (2) quantitative assessment of pin-by-pin properties, for example the activity of key isotopes or pin attributes such as cooling time and relative burnup,more » under the assumption that basic fuel parameters (e.g., assembly type and nominal fuel composition) are known. The efficacy of GET to meet these two verification objectives was evaluated across a range of fuel types, burnups and cooling times, while targeting a total interrogation time of less than 60 minutes. The evaluations were founded on a modelling and analysis framework applied to existing and emerging GET instrument designs. Monte Carlo models of different fuel types were used to produce simulated tomographer responses to large populations of “virtual” fuel assemblies. The simulated instrument response data were then processed using a variety of tomographic-reconstruction and image-processing methods, and scoring metrics were defined and used to evaluate the performance of the methods.This paper describes the analysis framework and metrics used to predict tomographer performance. It also presents the design of a “universal” GET (UGET) instrument intended to support the full range of verification scenarios envisioned by the IAEA. Finally, it gives examples of the expected partial-defect detection capabilities for some fuels and diversion scenarios, and it provides a comparison of predicted performance for the notional UGET design and an optimized variant of an existing IAEA instrument.« less

  6. Cooling assembly for fuel cells

    DOEpatents

    Kaufman, Arthur; Werth, John

    1990-01-01

    A cooling assembly for fuel cells having a simplified construction whereby coolant is efficiently circulated through a conduit arranged in serpentine fashion in a channel within a member of such assembly. The channel is adapted to cradle a flexible, chemically inert, conformable conduit capable of manipulation into a variety of cooling patterns without crimping or otherwise restricting of coolant flow. The conduit, when assembled with the member, conforms into intimate contact with the member for good thermal conductivity. The conduit is non-corrodible and can be constructed as a single, manifold-free, continuous coolant passage means having only one inlet and one outlet.

  7. Flexible nitrogen-doped graphene/carbon nanotube/Co3O4 paper and its oxygen reduction activity

    NASA Astrophysics Data System (ADS)

    Li, Shan-Shan; Cong, Huai-Ping; Wang, Ping; Yu, Shu-Hong

    2014-06-01

    Due to the demand of an efficient, inexpensive and scalable synthesis of oxygen reduction reaction (ORR) catalyst for practical application in fuel cell, we demonstrate a facile strategy to fabricate the flexible nitrogen-doped graphene/carbon nanotube/Co3O4 (NG/CNT/Co3O4) paper catalyst. In the hydrothermal process, the in situ formation of Co3O4 nanoparticles, reduction of GO and doping of nitrogen species occur simultaneously in the assembled paper in ammonia solution. Because of the synergistic effects of three active components and the spacing effect of CNTs and Co3O4 nanoparticles on avoiding the re-aggregation of assembled graphene nanosheets, the free-standing NG/CNT/Co3O4 paper exhibits an enhanced ORR catalytic performance with stable durability and strong methanol-tolerant capability, indicating promising potential as ORR electrocatalyst in practical applications.Due to the demand of an efficient, inexpensive and scalable synthesis of oxygen reduction reaction (ORR) catalyst for practical application in fuel cell, we demonstrate a facile strategy to fabricate the flexible nitrogen-doped graphene/carbon nanotube/Co3O4 (NG/CNT/Co3O4) paper catalyst. In the hydrothermal process, the in situ formation of Co3O4 nanoparticles, reduction of GO and doping of nitrogen species occur simultaneously in the assembled paper in ammonia solution. Because of the synergistic effects of three active components and the spacing effect of CNTs and Co3O4 nanoparticles on avoiding the re-aggregation of assembled graphene nanosheets, the free-standing NG/CNT/Co3O4 paper exhibits an enhanced ORR catalytic performance with stable durability and strong methanol-tolerant capability, indicating promising potential as ORR electrocatalyst in practical applications. Electronic supplementary information (ESI) available: XRD patterns, elemental mapping images, DSC-TGA curves and XPS spectrum of NG/CNT/Co3O4 paper; SEM images and XPS spectra of NG/Co3O4 paper; RDE curves and corresponding K-L plots of different catalysts. See DOI: 10.1039/c4nr02101k

  8. System Regulates the Water Contents of Fuel-Cell Streams

    NASA Technical Reports Server (NTRS)

    Vasquez, Arturo; Lazaroff, Scott

    2005-01-01

    An assembly of devices provides for both humidification of the reactant gas streams of a fuel cell and removal of the product water (the water generated by operation of the fuel cell). The assembly includes externally-sensing forward-pressure regulators that supply reactant gases (fuel and oxygen) at variable pressures to ejector reactant pumps. The ejector supply pressures depend on the consumption flows. The ejectors develop differential pressures approximately proportional to the consumption flow rates at constant system pressure and with constant flow restriction between the mixer-outlet and suction ports of the ejectors. For removal of product water from the circulating oxygen stream, the assembly includes a water/gas separator that contains hydrophobic and hydrophilic membranes. The water separator imposes an approximately constant flow restriction, regardless of the quality of the two-phase flow that enters it from the fuel cell. The gas leaving the water separator is nearly 100 percent humid. This gas is returned to the inlet of the fuel cell along with a quantity of dry incoming oxygen, via the oxygen ejector, thereby providing some humidification.

  9. Apparatus for inspecting fuel elements

    DOEpatents

    Oakley, David J.; Groves, Oliver J.; Kaiser, Bruce J.

    1986-01-01

    Disclosed is an alpha monitor usable in an automated nuclear fuel pin loading and processing unit. Fuel pins or other elongated pins are fed laterally into the alpha monitor in a singular fashion and are translated by a first roller assembly into a weld flare machining and decontamination substation not forming a part of the invention. Pins return and are lifted upwardly and transferred across to a combined pin lifting and electrode operating means which lifts the pins upwardly into a clamshell electrode assembly which is spread open by a combined pin lifting and electrode operating means. Once inserted the clamshell type electrode arrangement closes around the fuel pins so that inspection can occur. Fuel pins are inspected by charging electrodes to a negative potential and measuring the change in charge occurring when positively charged alpha particles strike the negatively charged electrodes. After inspection, the fuel pins are lowered by the pin lifting and electrode operating means into a second roller assembly which longitudinally conveys approved pins from the airtight enclosure in which the alpha monitor is mounted. If the fuel pins are rejected then they are moved laterally by a second transfer means and onto another system for further processing.

  10. Apparatus for inspecting fuel elements

    DOEpatents

    Kaiser, B.J.; Oakley, D.J.; Groves, O.J.

    1984-12-21

    This disclosure describes an alpha monitor usable in an automated nuclear fuel pin loading and processing unit. Fuel pins or other elongated pins are fed laterally into the alpha monitor in a singular fashion and are translated by a first roller assembly into a weld flare machining and decontamination substation not forming a part of the invention. Pins return and are lifted upwardly and transferred across to a combined pin lifting and electrode operating means which lifts the pins upwardly into a clamshell electrode assembly which is spread open by a combined pin lifting and electrode operating means. Once inserted the clamshell type electrode arrangement closes around the fuel pins so that inspection can occur. Fuel pins are inspected by charging electrodes to a negative potential and measuring the change in charge occurring when positively charged alpha particles strike the negatively charged electrodes. After inspection, the fuel pins are lowered by the pin lifting and electrode operating means into a second roller assembly which longitudinally conveys approved pins from the airtight enclosure in which the alpha monitor is mounted. If the fuel pins are rejected then they are moved laterally by a second transfer means and onto another system for further processing.

  11. CFD Analysis of Coolant Flow in VVER-440 Fuel Assemblies with the Code ANSYS CFX 10.0

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Toth, Sandor; Legradi, Gabor; Aszodi, Attila

    2006-07-01

    From the aspect of planning the power upgrading of nuclear reactors - including the VVER-440 type reactor - it is essential to get to know the flow field in the fuel assembly. For this purpose we have developed models of the fuel assembly of the VVER-440 reactor using the ANSYS CFX 10.0 CFD code. At first a 240 mm long part of a 60 degrees segment of the fuel pin bundle was modelled. Implementing this model a sensitivity study on the appropriate meshing was performed. Based on the development of the above described model, further models were developed: a 960more » mm long part of a 60-degree-segment and a full length part (2420 mm) of the fuel pin bundle segment. The calculations were run using constant coolant properties and several turbulence models. The impacts of choosing different turbulence models were investigated. The results of the above-mentioned investigations are presented in this paper. (authors)« less

  12. Fuel cell assembly with electrolyte transport

    DOEpatents

    Chi, Chang V.

    1983-01-01

    A fuel cell assembly wherein electrolyte for filling the fuel cell matrix is carried via a transport system comprising a first passage means for conveying electrolyte through a first plate and communicating with a groove in a second plate at a first point, the first and second plates together sandwiching the matrix, and second passage means acting to carry electrolyte exclusively through the second plate and communicating with the groove at a second point exclusive of the first point.

  13. MELCOR model for an experimental 17x17 spent fuel PWR assembly.

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Cardoni, Jeffrey

    2010-11-01

    A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

  14. Shaping of the axial power density distribution in the core to minimize the vapor volume fraction at the outlet of the VVER-1200 fuel assemblies

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Savander, V. I.; Shumskiy, B. E., E-mail: borisshumskij@yandex.ru; Pinegin, A. A.

    The possibility of decreasing the vapor fraction at the VVER-1200 fuel assembly outlet by shaping the axial power density field is considered. The power density field was shaped by axial redistribution of the concentration of the burnable gadolinium poison in the Gd-containing fuel rods. The mathematical modeling of the VVER-1200 core was performed using the NOSTRA computer code.

  15. Vibrating fuel grapple. [LMFBR

    DOEpatents

    Chertock, A.J.; Fox, J.N.; Weissinger, R.B.

    A reactor refueling method is described which utilizes a vibrating fuel grapple for removing spent fuel assemblies from a reactor core. It incorporates a pneumatic vibrator in the grapple head which allows additional withdrawal capability without exceeding the allowable axial force limit. The only moving part in the vibrator is a steel ball, pneumatically driven by a gas, such as argon, around a track, with centrifugal force created by the ball being transmitted through the grapple to the assembly handling socket.

  16. TMI-2 (Three Mile Island Unit 2) core region defueling

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Rodabaugh, J.M.; Cowser, D.K.

    1988-01-01

    In July of 1982, a video camera was inserted into the Three Mile Island Unit 2 reactor vessel providing the first visual evidence of core damage. This inspection, and numerous subsequent data acquisition tasks, revealed a central void /approx/1.5 m (5 ft) deep. This void region was surrounded by partial length fuel assemblies and ringed on the periphery by /approx/40 full-length, but partial cross-section, fuel assemblies. All of the original 177 fuel assemblies exhibited signs of damage. The bottom of the void cavity was covered with a bed of granular rubble, fuel assembly upper end fittings, control rod spiders, fuelmore » rod fragments, and fuel pellets. It was obvious that the normal plant refueling system not suitable for removing the damaged core. A new system of defueling tools and equipment was necessary to perform this task. Design of the new system was started immediately, followed by >1 yr of fabrication. Delivery and checkout of the defueling system occurred in mid-1985. Actual defueling was initiated in late 1985 with removal of the debris bed at the bottom of the core void. Obstructions to the debris, such as end fittings and fuel rod fragments ere removed first; then /approx/23,000 kg (50,000lb) of granular debris was quickly loaded into canisters. Core region defueling was completed in late 1987, /approx/2 yr after it was initiated.« less

  17. Finite element simulation of gap opening between cladding tube and spacer grid in a fuel rod assembly using crystallographic models of irradiation growth and creep

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Patra, Anirban; Tomé, Carlos N.

    A physically-based crystal plasticity framework for modeling irradiation growth and creep is interfaced with the finite element code ABAQUS in order to study the contact forces and the gap evolution between the spacer grid and the cladding tube as a function of irradiation in a representative section of a fuel rod assembly. Deformation mechanisms governing the gap opening are identified and correlated to the texture-dependent material response. Thus, in the absence of coolant flow-induced vibrations, these simulations predict the contribution of irradiation growth and creep to the gap opening between the cladding tube and the springs and dimples on themore » spacer grid. The simulated contact forces on the springs and dimples are compared to available experimental and modeling data. Various combinations of external loads are applied on the springs and dimples to simulate fuel rods in the interior and at the periphery of the fuel rod assembly. Furthermore, we found that loading conditions representative (to a first order approximation) of fuel rods at the periphery show higher gap opening. This is in agreement with in-reactor data, where rod leakages due to the synergistic effects of gap opening and coolant flow-induced vibrations were generally found to occur at the periphery of the fuel rod assembly.« less

  18. Finite element simulation of gap opening between cladding tube and spacer grid in a fuel rod assembly using crystallographic models of irradiation growth and creep

    DOE PAGES

    Patra, Anirban; Tomé, Carlos N.

    2017-03-06

    A physically-based crystal plasticity framework for modeling irradiation growth and creep is interfaced with the finite element code ABAQUS in order to study the contact forces and the gap evolution between the spacer grid and the cladding tube as a function of irradiation in a representative section of a fuel rod assembly. Deformation mechanisms governing the gap opening are identified and correlated to the texture-dependent material response. Thus, in the absence of coolant flow-induced vibrations, these simulations predict the contribution of irradiation growth and creep to the gap opening between the cladding tube and the springs and dimples on themore » spacer grid. The simulated contact forces on the springs and dimples are compared to available experimental and modeling data. Various combinations of external loads are applied on the springs and dimples to simulate fuel rods in the interior and at the periphery of the fuel rod assembly. Furthermore, we found that loading conditions representative (to a first order approximation) of fuel rods at the periphery show higher gap opening. This is in agreement with in-reactor data, where rod leakages due to the synergistic effects of gap opening and coolant flow-induced vibrations were generally found to occur at the periphery of the fuel rod assembly.« less

  19. Reactor physics behavior of transuranic-bearing TRISO-particle fuel in a pressurized water reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Pope, M. A.; Sen, R. S.; Ougouag, A. M.

    2012-07-01

    Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU) - only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space availablemore » for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is retained. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint. (authors)« less

  20. Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag

    2012-04-01

    Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU)-only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel,more » the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.« less

  1. Chalcogen catalysts for polymer electrolyte fuel cell

    DOEpatents

    Alonso-Vante, Nicolas [Buxerolles, FR; Zelenay, Piotr [Los Alamos, NM; Choi, Jong-Ho [Los Alamos, NM; Wieckowski, Andrzej [Champaign, IL; Cao, Dianxue [Urbana, IL

    2009-09-15

    A methanol-tolerant cathode catalyst and a membrane electrode assembly for fuel cells that includes such a cathode catalyst. The cathode catalyst includes a support having at least one transition metal in elemental form and a chalcogen disposed on the support. Methods of making the cathode catalyst and membrane electrode assembly are also described.

  2. Mixer Assembly for a Gas Turbine Engine

    NASA Technical Reports Server (NTRS)

    Smith, Lance L. (Inventor); Fotache, Catalin G. (Inventor); Dai, Zhongtao (Inventor); Cohen, Jeffrey M. (Inventor); Hautman, Donald J. (Inventor)

    2015-01-01

    A mixer assembly for a gas turbine engine is provided, including a main mixer with fuel injection holes located between at least one radial swirler and at least one axial swirler, wherein the fuel injected into the main mixer is atomized and dispersed by the air flowing through the radial swirler and the axial swirler.

  3. Mixer Assembly for a Gas Turbine Engine

    NASA Technical Reports Server (NTRS)

    Dai, Zhongtao (Inventor); Cohen, Jeffrey M. (Inventor); Fotache, Catalin G. (Inventor); Hautman, Donald J. (Inventor); Smith, Lance L. (Inventor)

    2018-01-01

    A mixer assembly for a gas turbine engine is provided, including a main mixer with fuel injection holes located between at least one radial swirler and at least one axial swirler, wherein the fuel injected into the main mixer is atomized and dispersed by the air flowing through the radial swirler and the axial swirler.

  4. Chalcogen catalysts for polymer electrolyte fuel cell

    DOEpatents

    Zelenay, Piotr; Choi, Jong-Ho; Alonso-Vante, Nicolas; Wieckowski, Andrzej; Cao, Dianxue

    2010-08-24

    A methanol-tolerant cathode catalyst and a membrane electrode assembly for fuel cells that includes such a cathode catalyst. The cathode catalyst includes a support having at least one transition metal in elemental form and a chalcogen disposed on the support. Methods of making the cathode catalyst and membrane electrode assembly are also described.

  5. Utilization of hydrolysate from lignocellulosic biomass pretreatment to generate electricity by enzymatic fuel cell system.

    PubMed

    Kim, Sung Bong; Kim, Dong Sup; Yang, Ji Hyun; Lee, Junyoung; Kim, Seung Wook

    2016-04-01

    The waste hydrolysate after dilute acid pretreatment (DAP) of lignocellulosic biomass was utilized to generate electricity using an enzymatic fuel cell (EFC) system. During DAP, the components of biomass containing hemicellulose and other compounds are hydrolyzed, and glucose is solubilized into the dilute acid solution, called as the hydrolysate liquid. Glucose oxidase (GOD) and laccase (Lac) were assembled on the electrode of the anode and cathode, respectively. Cyclic voltammetry (CV) and electrochemical impedance spectroscopy (EIS) were measured, and the maximum power density was found to be 1.254×10(3) μW/cm(2). The results indicate that the hydrolysate from DAP is a reliable electrolyte containing the fuel of EFC. Moreover, the impurities in the hydrolysate such as phenols and furans slightly affected the charge transfer on the surface of the electrode, but did not affect the power generation of the EFC system in principal. Copyright © 2016 Elsevier Inc. All rights reserved.

  6. High-Flow Jet Exit Rig Designed and Fabricated

    NASA Technical Reports Server (NTRS)

    Buehrle, Robert J.; Trimarchi, Paul A.

    2003-01-01

    The High-Flow Jet Exit Rig at the NASA Glenn Research Center is designed to test single flow jet nozzles and to measure the appropriate thrust and noise levels. The rig has been designed for the maximum hot condition of 16 lbm/sec of combustion air at 1960 R (maximum) and to produce a maximum thrust of 2000 lb. It was designed for cold flow of 29.1 lbm/sec of air at 530 R. In addition, it can test dual-flow nozzles (nozzles with bypass flow in addition to core flow) with independent control of each flow. The High- Flow Jet Exit Rig was successfully fabricated in late 2001 and is being readied for checkout tests. The rig will be installed in Glenn's Aeroacoustic Propulsion Laboratory. The High-Flow Jet Exit Rig consists of the following major components: a single component force balance, the natural-gas-fueled J-79 combustor assembly, the plenum and manifold assembly, an acoustic/instrumentation/seeding (A/I/S) section, a table, and the research nozzles. The rig will be unique in that it is designed to operate uncooled. The structure survives the 1960 R test condition because it uses carefully selected high temperature alloy materials such as Hastelloy-X. The lower plenum assembly was designed to operate at pressures to 450 psig at 1960 R, in accordance with the ASME B31.3 piping code. The natural gas-fueled combustor fires directly into the lower manifold. The hot air is directed through eight 1-1/2-in. supply pipes that supply the upper plenum. The flow is conditioned in the upper plenum prior to flowing to the research nozzle. The 1-1/2-in. supply lines are arranged in a U-shaped design to provide for a flexible piping system. The combustor assembly checkout was successfully conducted in Glenn's Engine Component Research Laboratory in the spring of 2001. The combustor is a low-smoke version of the J79 combustor used to power the F4 Phantom military aircraft. The natural gas-fueled combustor demonstrated high-efficiency combustion over a wide range of operating conditions. This wide operating envelope is required to support the testing of both single- and dual-flow nozzles. Key research goals include providing simultaneous, highly accurate acoustic, flow, and thrust measurements on jet nozzle models in realistic flight conditions, as well as providing scaleable acoustic results. The High-Flow Jet Exit Rig is a second-generation high-flow test rig. Improvements include cleaner flow with reduced levels of particulate, soot, and odor. Choked-flow metering is required with plus or minus 0.25-percent accuracy. Thrust measurements from 0 to 2000 lbf are required with plus or minus 0.25-percent accuracy. Improved acoustics will be achieved by minimizing noise through large pipe bend radii, lower internal flow velocities, and microdrilled choke plates with thousands of 0.040-in.- diameter holes.

  7. Detachable connection for a nuclear reactor fuel assembly

    DOEpatents

    Christiansen, D.W.; Karnesky, R.A.

    1983-08-29

    A locking connection for releasably attaching a handling socket to the duct tube of a fuel assembly for a nuclear reactor. The connection comprises a load pad housing mechanically attached to the duct tube and a handling socket threadably secured within the housing. A retaining ring is interposed between the housing and the handling socket and is formed with a projection and depression engagable within a cavity and groove of the housing and handling socket, respectively, to form a detachable interlocked connection assembly.

  8. Detachable connection for a nuclear reactor fuel assembly

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.

    1986-01-01

    A locking connection for releasably attaching a handling socket to the duct tube of a fuel assembly for a nuclear reactor. The connection comprises a load pad housing mechanically attached to the duct tube and a handling socket threadably secured within the housing. A retaining ring is interposed between the housing and the handling socket and is formed with a projection and depression engageable within a cavity and groove of the housing and handling socket, respectively, to form a detachable interlocked connection assembly.

  9. Measuring spent fuel assembly multiplication in borated water with a passive neutron albedo reactivity instrument

    NASA Astrophysics Data System (ADS)

    Tobin, Stephen J.; Peura, Pauli; Bélanger-Champagne, Camille; Moring, Mikael; Dendooven, Peter; Honkamaa, Tapani

    2018-07-01

    The performance of a passive neutron albedo reactivity (PNAR) instrument to measure neutron multiplication of spent nuclear fuel in borated water is investigated as part of an integrated non-destructive assay safeguards system. To measure the PNAR Ratio, which is proportional to the neutron multiplication, the total neutron count rate is measured in high- and low-multiplying environments by the PNAR instrument. The integrated system also contains a load cell and a passive gamma emission tomograph, and as such meets all the recommendations of the IAEA's recent ASTOR Experts Group report. A virtual spent fuel library for VVER-440 fuel was used in conjunction with MCNP simulations of the PNAR instrument to estimate the measurement uncertainties from (1) variation in the water boron content, (2) assembly positioning in the detector and (3) counting statistics. The estimated aggregate measurement uncertainty on the PNAR Ratio measurement is 0.008, to put this uncertainty in context, the difference in the PNAR Ratio between a fully irradiated assembly and this same assembly when fissile isotopes only absorb neutrons, but do not emit neutrons, is 0.106, a 13-sigma effect. The 1-sigma variation of 0.008 in the PNAR Ratio is estimated to correspond to a 3.2 GWd/tU change in assembly burnup.

  10. Compact propane fuel processor for auxiliary power unit application

    NASA Astrophysics Data System (ADS)

    Dokupil, M.; Spitta, C.; Mathiak, J.; Beckhaus, P.; Heinzel, A.

    With focus on mobile applications a fuel cell auxiliary power unit (APU) using liquefied petroleum gas (LPG) is currently being developed at the Centre for Fuel Cell Technology (Zentrum für BrennstoffzellenTechnik, ZBT gGmbH). The system is consisting of an integrated compact and lightweight fuel processor and a low temperature PEM fuel cell for an electric power output of 300 W. This article is presenting the current status of development of the fuel processor which is designed for a nominal hydrogen output of 1 k Wth,H2 within a load range from 50 to 120%. A modular setup was chosen defining a reformer/burner module and a CO-purification module. Based on the performance specifications, thermodynamic simulations, benchmarking and selection of catalysts the modules have been developed and characterised simultaneously and then assembled to the complete fuel processor. Automated operation results in a cold startup time of about 25 min for nominal load and carbon monoxide output concentrations below 50 ppm for steady state and dynamic operation. Also fast transient response of the fuel processor at load changes with low fluctuations of the reformate gas composition have been achieved. Beside the development of the main reactors the transfer of the fuel processor to an autonomous system is of major concern. Hence, concepts for packaging have been developed resulting in a volume of 7 l and a weight of 3 kg. Further a selection of peripheral components has been tested and evaluated regarding to the substitution of the laboratory equipment.

  11. Depleted uranium as a backfill for nuclear fuel waste package

    DOEpatents

    Forsberg, Charles W.

    1998-01-01

    A method for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package.

  12. Depleted uranium as a backfill for nuclear fuel waste package

    DOEpatents

    Forsberg, C.W.

    1998-11-03

    A method is described for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package. 6 figs.

  13. Initial Neutronics Analyses for HEU to LEU Fuel Conversion of the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Kontogeorgakos, D.; Derstine, K.; Wright, A.

    2013-06-01

    The purpose of the TREAT reactor is to generate large transient neutron pulses in test samples without over-heating the core to simulate fuel assembly accident conditions. The power transients in the present HEU core are inherently self-limiting such that the core prevents itself from overheating even in the event of a reactivity insertion accident. The objective of this study was to support the assessment of the feasibility of the TREAT core conversion based on the present reactor performance metrics and the technical specifications of the HEU core. The LEU fuel assembly studied had the same overall design, materials (UO 2more » particles finely dispersed in graphite) and impurities content as the HEU fuel assembly. The Monte Carlo N–Particle code (MCNP) and the point kinetics code TREKIN were used in the analyses.« less

  14. FUEL ASSEMBLY FOR A NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1958-04-29

    A fuel assembly for a nuclear reactor of the type wherein liquid coolant is circulated through the core of the reactor in contact with the external surface of the fuel elements is described. In this design a plurality of parallel plates containing fissionable material are spaced about one-tenth of an inch apart and are supported between a pair of spaced parallel side members generally perpendicular to the plates. The plates all have a small continuous and equal curvature in the same direction between the side members.

  15. Fabrication Method for Laboratory-Scale High-Performance Membrane Electrode Assemblies for Fuel Cells.

    PubMed

    Sassin, Megan B; Garsany, Yannick; Gould, Benjamin D; Swider-Lyons, Karen E

    2017-01-03

    Custom catalyst-coated membranes (CCMs) and membrane electrode assemblies (MEAs) are necessary for the evaluation of advanced electrocatalysts, gas diffusion media (GDM), ionomers, polymer electrolyte membranes (PEMs), and electrode structures designed for use in next-generation fuel cells, electrolyzers, or flow batteries. This Feature provides a reliable and reproducible fabrication protocol for laboratory scale (10 cm 2 ) fuel cells based on ultrasonic spray deposition of a standard Pt/carbon electrocatalyst directly onto a perfluorosulfonic acid PEM.

  16. Fuel inspection and reconstitution experience at Surry Power Station

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Brookmire, T.A.

    Surry Power Station, located on the James River near Williamsburg, Virginia, has two Westinghouse pressurized water reactors. Unit 2 consistently sets a high standard of fuel performance (no indication of fuel failures in recent cycles), while unit 1, since cycle 6, has been plagued with numerous fuel failures. Both Surry units operate with Westinghouse standard 15 x 15 fuel. Virginia Power management set goals to reduce the coolant activity, thus reducing person-rem exposure and the associated costs of high coolant activity. To achieve this goal, extensive fuel examination campaigns were undertaken that included high-magnification video inspectionsa, debris cleaning, wet andmore » vacuum fuel sipping, fuel rod ultrasonic testing, and eddy current examination. In the summer of 1985, during cycle 8 operation, Kraftwerk Union reconstituted (repaired) the damage, once-burned assemblies from cycles 6 and 7 by replacing failed fuel rods with solid Zircaloy-4 rods. Currently, cycle 9 has operated for 5 months without any indication of fuel failure (the cycle 9 core has two reconstituted assemblies).« less

  17. Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

    DOE PAGES

    Carmack, W. Jon; Chichester, Heather M.; Porter, Douglas L.; ...

    2016-02-27

    The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important potential comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The irradiations were the beginning tests to qualify U-10wt%Zr as a driver fuel for FFTF. The FFTF core, with a 91.4 cm tall fuel column and a chopped cosine neutron flux profile, operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This then places the peakmore » fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in previous EBR-II experiments that had a 32-cm height core. The MFF-3 and MFF-5 qualification assemblies operated in FFTF to >10 at% burnup, and performed very well with no cladding breaches. The MFF-3 assembly operated to 13.8 at% burnup with a peak inner cladding temperature of 643°C, and the MFF-5 assembly operated to 10.1 at% burnup with a peak inner cladding temperature of 651°C. Because of the very high operating temperatures for both the fuel and the cladding, data from the MFF assemblies are most comparable to the data obtained from the EBR-II X447 experiment, which experienced two pin breaches. The X447 breaches were strongly influenced by a large amount of fuel/cladding chemical interaction (FCCI). The MFF pins benefitted from different axial locations of high burnup and peak cladding temperature, which helped to reduce interdiffusion between rare earth fission products and stainless steel cladding. Post-irradiation examination evidence illustrates this advantage. After comparing other performance data of the long MFF pins to prior EBR-II test data, the MFF fuel inside the cladding grew less axially, and the gas release data did not reveal a definitive difference.« less

  18. Analysis of key safety metrics of thorium utilization in LWRs

    DOE PAGES

    Ade, Brian J.; Bowman, Stephen M.; Worrall, Andrew; ...

    2016-04-08

    Here, thorium has great potential to stretch nuclear fuel reserves because of its natural abundance and because it is possible to breed the 232Th isotope into a fissile fuel ( 233U). Various scenarios exist for utilization of thorium in the nuclear fuel cycle, including use in different nuclear reactor types (e.g., light water, high-temperature gas-cooled, fast spectrum sodium, and molten salt reactors), along with use in advanced accelerator-driven systems and even in fission-fusion hybrid systems. The most likely near-term application of thorium in the United States is in currently operating light water reactors (LWRs). This use is primarily based onmore » concepts that mix thorium with uranium (UO 2 + ThO 2) or that add fertile thorium (ThO 2) fuel pins to typical LWR fuel assemblies. Utilization of mixed fuel assemblies (PuO 2 + ThO 2) is also possible. The addition of thorium to currently operating LWRs would result in a number of different phenomenological impacts to the nuclear fuel. Thorium and its irradiation products have different nuclear characteristics from those of uranium and its irradiation products. ThO 2, alone or mixed with UO 2 fuel, leads to different chemical and physical properties of the fuel. These key reactor safety–related issues have been studied at Oak Ridge National Laboratory and documented in “Safety and Regulatory Issues of the Thorium Fuel Cycle” (NUREG/CR-7176, U.S. Nuclear Regulatory Commission, 2014). Various reactor analyses were performed using the SCALE code system for comparison of key performance parameters of both ThO 2 + UO 2 and ThO 2 + PuO 2 against those of UO 2 and typical UO 2 + PuO 2 mixed oxide fuels, including reactivity coefficients and power sharing between surrounding UO 2 assemblies and the assembly of interest. The decay heat and radiological source terms for spent fuel after its discharge from the reactor are also presented. Based on this evaluation, potential impacts on safety requirements and identification of knowledge gaps that require additional analysis or research to develop a technical basis for the licensing of thorium fuel are identified.« less

  19. Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Carmack, W. J.; Chichester, H. M.; Porter, D. L.

    2016-05-01

    Abstract The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important potential comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The irradiations were the beginning tests to qualify U-10wt%Zr as a driver fuel for FFTF. The FFTF core, with a 91.4 cm tall fuel column and a chopped cosine neutron flux profile, operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peakmore » fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in previous EBR-II experiments that had a 32-cm height core. The MFF-3 and MFF-5 qualification assemblies operated in FFTF to >10 at% burnup, and performed very well with no cladding breaches. The MFF-3 assembly operated to 13.8 at% burnup with a peak inner cladding temperature of 643°C, and the MFF-5 assembly operated to 10.1 at% burnup with a peak inner cladding temperature of 651°C. Because of the very high operating temperatures for both the fuel and the cladding, data from the MFF assemblies are most comparable to the data obtained from the EBR-II X447 experiment, which experienced two pin breaches. The X447 breaches were strongly influenced by a large amount of fuel/cladding chemical interaction (FCCI). The MFF pins benefitted from different axial locations of high burnup and peak cladding temperature, which helped to reduce interdiffusion between rare earth fission products and stainless steel cladding. Post-irradiation examination evidence illustrates this advantage. Comparing other performance data of the long MFF pins to prior EBR-II test data, the MFF fuel inside the cladding grew less axially, and the gas release data did not reveal a definitive difference.« less

  20. KSC-2009-5814

    NASA Image and Video Library

    2009-10-24

    CAPE CANAVERAL, Fla. – A bright sun and blue sky herald the passage of the Pegasus barge, carrying external tank 134, through Port Canaveral. The tugboat will move the barge through the Banana River toward the turn basin near the Vehicle Assembly Building, or VAB, at NASA's Kennedy Space Center in Florida. Pegasus arrived in Florida after an ocean voyage towed by a solid rocket booster retrieval ship from NASA's Michoud Assembly Facility near New Orleans. After Pegasus docks in the turn basin, the fuel tank will be offloaded and transported into the VAB. ET-134 will be used to launch space shuttle Endeavour on the STS-130 mission to the International Space Station. Launch is targeted for Feb. 4, 2010. For information on the components of the space shuttle and the STS-130 mission, visit http://www.nasa.gov/shuttle. Photo credit: NASA/Jack Pfaller

  1. KSC-2009-5816

    NASA Image and Video Library

    2009-10-24

    CAPE CANAVERAL, Fla. – A tugboat moves the Pegasus barge, carrying external tank 134, through the Banana River toward the turn basin near the Vehicle Assembly Building, or VAB, at NASA's Kennedy Space Center in Florida. Space Shuttle Atlantis awaits launch on Launch Pad 39A in the background. Pegasus arrived in Florida after an ocean voyage towed by a solid rocket booster retrieval ship from NASA's Michoud Assembly Facility near New Orleans. After Pegasus docks in the turn basin, the fuel tank will be offloaded and transported into the VAB. ET-134 will be used to launch space shuttle Endeavour on the STS-130 mission to the International Space Station. Launch is targeted for Feb. 4, 2010. For information on the components of the space shuttle and the STS-130 mission, visit http://www.nasa.gov/shuttle. Photo credit: NASA/Jack Pfaller

  2. KSC-2009-5815

    NASA Image and Video Library

    2009-10-24

    CAPE CANAVERAL, Fla. – The Pegasus barge, carrying external tank 134, passes through a bridge into the river near Port Canaveral. The tugboat will move the barge through the Banana River toward the turn basin near the Vehicle Assembly Building, or VAB, at NASA's Kennedy Space Center in Florida. Pegasus arrived in Florida after an ocean voyage towed by a solid rocket booster retrieval ship from NASA's Michoud Assembly Facility near New Orleans. After Pegasus docks in the turn basin, the fuel tank will be offloaded and transported into the VAB. ET-134 will be used to launch space shuttle Endeavour on the STS-130 mission to the International Space Station. Launch is targeted for Feb. 4, 2010. For information on the components of the space shuttle and the STS-130 mission, visit http://www.nasa.gov/shuttle. Photo credit: NASA/Jack Pfaller

  3. 235U Holdup Measurements in the 321-M Exhaust Elbows

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Salaymeh, S.R.

    The Analytical Development Section of Savannah River Technology Center (SRTC) was requested by the Facilities Disposition Division (FDD) to determine the holdup of enriched uranium in the 321-M facility as part of an overall deactivation project of the facility. The 321-M facility was used to fabricate enriched uranium fuel assemblies, lithium-aluminum target tubes, neptunium assemblies, and miscellaneous components for the production reactors. The facility also includes the 324-M storage building and the passageway connecting it to 321-M. The results of the holdup assays are essential for determining compliance with the Waste Acceptance Criteria, Material Control and Accountability, and to meetmore » criticality safety controls. This report covers holdup measurements of uranium residue in the exhaust piping elbows removed from the roof the 321-M facility.« less

  4. L-Area STS MTR/NRU/NRX Grapple Assembly Closure Mechanics Review

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Huizenga, D. J.

    2016-06-08

    A review of the closure mechanics associated with the Shielded Transfer System (STS) MTR/NRU/NRX grapple assembly utilized at the Savannah River Site (SRS) was performed. This review was prompted by an operational event which occurred at the Canadian Nuclear Laboratories (CNL) utilizing a DTS-XL grapple assembly which is essentially identical to the STS MTR/NRU/NRX grapple assembly used at the SRS. The CNL operational event occurred when a NRU/NRX fuel basket containing spent nuclear fuel assemblies was inadvertently released by the DTS-XL grapple assembly during a transfer. The SM review of the STS MTR/NRU/NRX grapple assembly will examine the operational aspectsmore » of the STS and the engineered features of the STS which prevent such an event at the SRS. The design requirements for the STS NRU/NRX modifications and the overall layout of the STS are provided in other documents.« less

  5. Proton-Fueled, Reversible DNA Hybridization Chain Assembly for pH Sensing and Imaging.

    PubMed

    Liu, Lan; Liu, Jin-Wen; Huang, Zhi-Mei; Wu, Han; Li, Na; Tang, Li-Juan; Jiang, Jian-Hui

    2017-07-05

    Design of DNA self-assembly with reversible responsiveness to external stimuli is of great interest for diverse applications. We for the first time develop a pH-responsive, fully reversible hybridization chain reaction (HCR) assembly that allows sensitive sensing and imaging of pH in living cells. Our design relies on the triplex forming sequences that form DNA triplex with toehold regions under acidic conditions and then induce a cascade of strand displacement and DNA assembly. The HCR assembly has shown dynamic responses in physiological pH ranges with excellent reversibility and demonstrated the potential for in vitro detection and live-cell imaging of pH. Moreover, this method affords HCR assemblies with highly localized fluorescence responses, offering advantages of improving sensitivity and better selectivity. The proton-fueled, reversible HCR assembly may provide a useful approach for pH-related cell biology study and disease diagnostics.

  6. Integral gas seal for fuel cell gas distribution assemblies and method of fabrication

    DOEpatents

    Dettling, Charles J.; Terry, Peter L.

    1985-03-19

    A porous gas distribution plate assembly for a fuel cell, such as a bipolar assembly, includes an inner impervious region wherein the bipolar assembly has good surface porosity but no through-plane porosity and wherein electrical conductivity through the impervious region is maintained. A hot-pressing process for forming the bipolar assembly includes placing a layer of thermoplastic sealant material between a pair of porous, electrically conductive plates, applying pressure to the assembly at elevated temperature, and allowing the assembly to cool before removing the pressure whereby the layer of sealant material is melted and diffused into the porous plates to form an impervious bond along a common interface between the plates holding the porous plates together. The distribution of sealant within the pores along the surface of the plates provides an effective barrier at their common interface against through-plane transmission of gas.

  7. Method of fabricating an integral gas seal for fuel cell gas distribution assemblies

    DOEpatents

    Dettling, Charles J.; Terry, Peter L.

    1988-03-22

    A porous gas distribution plate assembly for a fuel cell, such as a bipolar assembly, includes an inner impervious region wherein the bipolar assembly has good surface porosity but no through-plane porosity and wherein electrical conductivity through the impervious region is maintained. A hot-pressing process for forming the bipolar assembly includes placing a layer of thermoplastic sealant material between a pair of porous, electrically conductive plates, applying pressure to the assembly at elevated temperature, and allowing the assembly to cool before removing the pressure whereby the layer of sealant material is melted and diffused into the porous plates to form an impervious bond along a common interface between the plates holding the porous plates together. The distribution of sealant within the pores along the surface of the plates provides an effective barrier at their common interface against through-plane transmission of gas.

  8. Development of Neutron Energy Spectral Signatures for Passive Monitoring of Spent Nuclear Fuels in Dry Cask Storage

    NASA Astrophysics Data System (ADS)

    Harkness, Ira; Zhu, Ting; Liang, Yinong; Rauch, Eric; Enqvist, Andreas; Jordan, Kelly A.

    2018-01-01

    Demand for spent nuclear fuel dry casks as an interim storage solution has increased globally and the IAEA has expressed a need for robust safeguards and verification technologies for ensuring the continuity of knowledge and the integrity of radioactive materials inside spent fuel casks. Existing research has been focusing on "fingerprinting" casks based on count rate statistics to represent radiation emission signatures. The current research aims to expand to include neutron energy spectral information as part of the fuel characteristics. First, spent fuel composition data are taken from the Next Generation Safeguards Initiative Spent Fuel Libraries, representative for Westinghouse 17ˣ17 PWR assemblies. The ORIGEN-S code then calculates the spontaneous fission and (α,n) emissions for individual fuel rods, followed by detailed MCNP simulations of neutrons transported through the fuel assemblies. A comprehensive database of neutron energy spectral profiles is to be constructed, with different enrichment, burn-up, and cooling time conditions. The end goal is to utilize the computational spent fuel library, predictive algorithm, and a pressurized 4He scintillator to verify the spent fuel assemblies inside a cask. This work identifies neutron spectral signatures that correlate with the cooling time of spent fuel. Both the total and relative contributions from spontaneous fission and (α,n) change noticeably with respect to cooling time, due to the relatively short half-life (18 years) of the major neutron source 244Cm. Identification of this and other neutron spectral signatures allows the characterization of spent nuclear fuels in dry cask storage.

  9. Mechanical and transport properties of layer-by-layer electrospun composite proton exchange membranes for fuel cell applications.

    PubMed

    Mannarino, Matthew M; Liu, David S; Hammond, Paula T; Rutledge, Gregory C

    2013-08-28

    Composite membranes composed of highly conductive and selective layer-by-layer (LbL) films and electrospun fiber mats were fabricated and characterized for mechanical strength and electrochemical selectivity. The LbL component consists of a proton-conducting, methanol-blocking poly(diallyl dimethyl ammonium chloride)/sulfonated poly(2,6-dimethyl-1,4-phenylene oxide) (PDAC/sPPO) thin film. The electrospun fiber component consists of poly(trimethyl hexamethylene terephthalamide) (PA 6(3)T) fibers in a nonwoven mat of 60-90% porosity. The bare mats were annealed to improve their mechanical properties, which improvements are shown to be retained in the composite membranes. Spray LbL assembly was used as a means for the rapid formation of proton-conducting films that fill the void space throughout the porous electrospun matrix and create a fuel-blocking layer. Coated mats as thin as 15 μm were fabricated, and viable composite membranes with methanol permeabilities 20 times lower than Nafion and through-plane proton selectivity five and a half times greater than Nafion are demonstrated. The mechanical properties of the spray coated electrospun mats are shown to be superior to the LbL-only system and possess intrinsically greater dimensional stability and lower mechanical hysteresis than Nafion under hydrated conditions. The composite proton exchange membranes fabricated here were tested in an operational direct methanol fuel cell. The results show the potential for higher open circuit voltages (OCV) and comparable cell resistances when compared to fuel cells based on Nafion.

  10. System design of a large fuel cell hybrid locomotive

    NASA Astrophysics Data System (ADS)

    Miller, A. R.; Hess, K. S.; Barnes, D. L.; Erickson, T. L.

    Fuel cell power for locomotives combines the environmental benefits of a catenary-electric locomotive with the higher overall energy efficiency and lower infrastructure costs of a diesel-electric. A North American consortium, a public-private partnership, is developing a prototype hydrogen-fueled fuel cell-battery hybrid switcher locomotive for urban and military-base rail applications. Switcher locomotives are used in rail yards for assembling and disassembling trains and moving trains from one point to another. At 127 tonnes (280,000 lb), continuous power of 250 kW from its (proton exchange membrane) PEM fuel cell prime mover, and transient power well in excess of 1 MW, the hybrid locomotive will be the heaviest and most powerful fuel cell land vehicle yet. This fast-paced project calls for completion of the vehicle itself near the end of 2007. Several technical challenges not found in the development of smaller vehicles arise when designing and developing such a large fuel cell vehicle. Weight, center of gravity, packaging, and safety were design factors leading to, among other features, the roof location of the lightweight 350 bar compressed hydrogen storage system. Harsh operating conditions, especially shock loads during coupling to railcars, require component mounting systems capable of absorbing high energy. Vehicle scale-up by increasing mass, density, or power presents new challenges primarily related to issues of system layout, hydrogen storage, heat transfer, and shock loads.

  11. Analysis of features of hydrodynamics and heat transfer in the fuel assembly of prospective sodium reactor with a high rate of reproduction in the uranium-plutonium fuel cycle

    NASA Astrophysics Data System (ADS)

    Lubina, A. S.; Subbotin, A. S.; Sedov, A. A.; Frolov, A. A.

    2016-12-01

    The fast sodium reactor fuel assembly (FA) with U-Pu-Zr metallic fuel is described. In comparison with a "classical" fast reactor, this FA contains thin fuel rods and a wider fuel rod grid. Studies of the fluid dynamics and the heat transfer were carried out for such a new FA design. The verification of the ANSYS CFX code was provided for determination of the velocity, pressure, and temperature fields in the different channels. The calculations in the cells and in the FA were carried out using the model of shear stress transport (SST) selected at the stage of verification. The results of the hydrodynamics and heat transfer calculations have been analyzed.

  12. Electric utility acid fuel cell stack technology advancement

    NASA Astrophysics Data System (ADS)

    Congdon, J. V.; Goller, G. J.; Greising, G. J.; Obrien, J. J.; Randall, S. A.; Sandelli, G. J.; Breault, R. D.; Austin, G. W.; Bopse, S.; Coykendall, R. D.

    1984-11-01

    The principal effort under this program was directed at the fuel cell stack technology required to accomplish the initial feasibility demonstrations of increased cell stack operating pressures and temperatures, increased cell active area, incorporation of the ribbed substrate cell configuration at the bove conditions, and the introduction of higher performance electrocatalysts. The program results were successful with the primary accomplishments being: (1) fabrication of 10 sq ft ribbed substrate, cell components including higher performing electrocatalysts; (2) assembly of a 10 sq ft, 30-cell short stack; and (3) initial test of this stack at 120 psia and 405 F. These accomplishments demonstrate the feasibility of fabricating and handling large area cells using materials and processes that are oriented to low cost manufacture. An additional accomplishment under the program was the testing of two 3.7 sq ft short stacks at 12 psia/405 F to 5400 and 4500 hours respectively. These tests demonstrate the durability of the components and the cell stack configuration to a nominal 5000 hours at the higher pressure and temperature condition planned for the next electric utility power plant.

  13. Electric utility acid fuel cell stack technology advancement

    NASA Technical Reports Server (NTRS)

    Congdon, J. V.; Goller, G. J.; Greising, G. J.; Obrien, J. J.; Randall, S. A.; Sandelli, G. J.; Breault, R. D.; Austin, G. W.; Bopse, S.; Coykendall, R. D.

    1984-01-01

    The principal effort under this program was directed at the fuel cell stack technology required to accomplish the initial feasibility demonstrations of increased cell stack operating pressures and temperatures, increased cell active area, incorporation of the ribbed substrate cell configuration at the bove conditions, and the introduction of higher performance electrocatalysts. The program results were successful with the primary accomplishments being: (1) fabrication of 10 sq ft ribbed substrate, cell components including higher performing electrocatalysts; (2) assembly of a 10 sq ft, 30-cell short stack; and (3) initial test of this stack at 120 psia and 405 F. These accomplishments demonstrate the feasibility of fabricating and handling large area cells using materials and processes that are oriented to low cost manufacture. An additional accomplishment under the program was the testing of two 3.7 sq ft short stacks at 12 psia/405 F to 5400 and 4500 hours respectively. These tests demonstrate the durability of the components and the cell stack configuration to a nominal 5000 hours at the higher pressure and temperature condition planned for the next electric utility power plant.

  14. Modular fabrication and characterization of complex silicon carbide composite structures Advanced Reactor Technologies (ART) Research Final Report (Feb 2015 – May 2017)

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Khalifa, Hesham

    Advanced ceramic materials exhibit properties that enable safety and fuel cycle efficiency improvements in advanced nuclear reactors. In order to fully exploit these desirable properties, new processing techniques are required to produce the complex geometries inherent to nuclear fuel assemblies and support structures. Through this project, the state of complex SiC-SiC composite fabrication for nuclear components has advanced significantly. New methods to produce complex SiC-SiC composite structures have been demonstrated in the form factors needed for in-core structural components in advanced high temperature nuclear reactors. Advanced characterization techniques have been employed to demonstrate that these complex SiC-SiC composite structures providemore » the strength, toughness and hermeticity required for service in harsh reactor conditions. The complex structures produced in this project represent a significant step forward in leveraging the excellent high temperature strength, resistance to neutron induced damage, and low neutron cross section of silicon carbide in nuclear applications.« less

  15. Hidden Figures To Modern Figures- Students See SLS Rocket at Michoud

    NASA Image and Video Library

    2017-01-09

    New Orleans teacher Katherine Michelle Sanders of St. Peter Claver School, takes her 7th grade science on a tour of nearby NASA’s Michoud Assembly Facility to see where the Space Launch System - the world’s most powerful rocket - is being built. Sanders is the granddaughter of famed NASA scientist Katherine Johnson who was featured in the book and movie, Hidden Figures. Sanders discusses how her grandmother’s work inspired her and why she wanted her students to learn about NASA’s plan to explore deep space. Sanders followed her grandmother's footsteps, as Katherine Johnson was also a teacher before going to work for the National Advisory Committee for Aeronautics (NACA) and later at NASA's Langley Research Center as a human computer. Due to Johnson’s historical role as one of the first African-American women to work as a NASA scientist, she was awarded the Presidential Medal of Freedom by President Barack Obama in 2015. In this video, Michoud Director Bobby Watkins and NASA engineer Renee Horton speak with the students about key components building the SLS rocket, including the Vertical Assembly Center - a large robotic tool for manufacturing hardware - and a liquid hydrogen tank that will hold fuel for the first mission, and a liquid hydrogen tank that will hold fuel for the first SLS mission. For more information on SLS, visit www.nasa.gov/sls. For more information on Michoud Assembly Facility, visit www.nasa.gov/centers/marshall/michoud/index.html

  16. Implications of Zircaloy creep and growth to light water reactor performance

    NASA Astrophysics Data System (ADS)

    Franklin, David G.; Adamson, Ronald B.

    1988-10-01

    Deformation of zirconium alloy components in nuclear reactors has been a concern since the decision of Admiral Rickover to use them in the US Navy submarine reactors. With the exception of the first few light water reactors (LWRs) most of the core structural materials have been fabricated from either Zircaloy-2 or Zircaloy-4. Performance of these alloys has been extremely good, even though the effects of irradiation on deformation magnitudes and mechanisms were not fully appreciated until extensive service and in-reactor tests were accomplished. Since the reactor components are designed to operate at stress levels well below yield for normal conditions, the only significant deformation is time dependent. Although creep was anticipated, the enhancement by neutron irradiation and the stress-free, nearly constant-volume shape change known as irradiation growth were not known prior to materials testing in reactors under controlled conditions. Both of these phenomena have significant impact on performance and must be accounted for properly in design. Although irradiation creep and growth have resulted in only one significant performance problem (creep collapse of fuel cladding, which has been eliminated), deformation magnitudes and, particularly, differentials in strain magnitudes, are a continuing source of interest. Factors that affect dimensional stability due to both creep and growth include temperature, fluence, residual stress, texture, and microstructure. The first two are reactor variables and the others are related to component fabrication history. This paper includes a review of the applications of Zircaloy creep and growth to LWR fuel designs, a review of the impact of in-reactor creep and growth on fuel rod and fuel assembly performance, and comments on potential improvements. Since the reactor design, fuel design and the core environment in BWRs and PWRs are quite different, appropriate separation of the application of effects are made; of course, the basic phenomena are the same in both systems.

  17. Dry transfer system for spent fuel: Project report, A system designed to achieve the dry transfer of bare spent fuel between two casks. Final report

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Dawson, D.M.; Guerra, G.; Neider, T.

    1995-12-01

    This report describes the system developed by EPRI/DOE for the dry transfer of spent fuel assemblies outside the reactor spent fuel pool. The system is designed to allow spent fuel assemblies to be removed from a spent fuel pool in a small cask, transported to the transfer facility, and transferred to a larger cask, either for off-site transportation or on-site storage. With design modifications, this design is capable of transferring single spent fuel assemblies from dry storage casks to transportation casks or visa versa. One incentive for the development of this design is that utilities with limited lifting capacity ormore » other physical or regulatory constraints are limited in their ability to utilize the current, more efficient transportation and storage cask designs. In addition, DOE, in planning to develop and implement the multi-purpose canister (MPC) system for the Civilian Radioactive Waste Management System, included the concept of an on-site dry transfer system to support the implementation of the MPC system at reactors with limitations that preclude the handling of the MPC system transfer casks. This Dry Transfer System can also be used at reactors wi decommissioned spent fuel pools and fuel in dry storage in non-MPC systems to transfer fuel into transportation casks. It can also be used at off-reactor site interim storage facilities for the same purpose.« less

  18. PBF Reactor Building (PER620). Detail of fuel test assembly in ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    PBF Reactor Building (PER-620). Detail of fuel test assembly in preparation for test. When complete, it will fit into in-pile tube. The maximum outside diameter of which must be about 8.25 inches. Date: 1982. INEEL negative no. 82-4908 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  19. Towards neat methanol operation of direct methanol fuel cells: a novel self-assembled proton exchange membrane.

    PubMed

    Li, Jing; Cai, Weiwei; Ma, Liying; Zhang, Yunfeng; Chen, Zhangxian; Cheng, Hansong

    2015-04-18

    We report here a novel proton exchange membrane with remarkably high methanol-permeation resistivity and excellent proton conductivity enabled by carefully designed self-assembled ionic conductive channels. A direct methanol fuel cell utilizing the membrane performs well with a 20 M methanol solution, very close to the concentration of neat methanol.

  20. Development and Application of a Sample Holder for In Situ Gaseous TEM Studies of Membrane Electrode Assemblies for Polymer Electrolyte Fuel Cells.

    PubMed

    Kamino, Takeo; Yaguchi, Toshie; Shimizu, Takahiro

    2017-10-01

    Polymer electrolyte fuel cells hold great potential for stationary and mobile applications due to high power density and low operating temperature. However, the structural changes during electrochemical reactions are not well understood. In this article, we detail the development of the sample holder equipped with gas injectors and electric conductors and its application to a membrane electrode assembly of a polymer electrolyte fuel cell. Hydrogen and oxygen gases were simultaneously sprayed on the surfaces of the anode and cathode catalysts of the membrane electrode assembly sample, respectively, and observation of the structural changes in the catalysts were simultaneously carried out along with measurement of the generated voltages.

  1. The J-2X Fuel Turbopump - Design, Development, and Test

    NASA Technical Reports Server (NTRS)

    Tellier, James G.; Hawkins, Lakiesha V.; Shinguchi, Brian H.; Marsh, Matthew W.

    2011-01-01

    Pratt and Whitney Rocketdyne (PWR), a NASA subcontractor, is executing the design, development, test, and evaluation (DDT&E) of a liquid oxygen, liquid hydrogen two hundred ninety four thousand pound thrust rocket engine initially intended for the Upper Stage (US) and Earth Departure Stage (EDS) of the Constellation Program Ares-I Crew Launch Vehicle (CLV). A key element of the design approach was to base the new J-2X engine on the heritage J-2S engine with the intent of uprating the engine and incorporating SSME and RS-68 lessons learned. The J-2S engine was a design upgrade of the flight proven J-2 configuration used to put American astronauts on the moon. The J-2S Fuel Turbopump (FTP) was the first Rocketdyne-designed liquid hydrogen centrifugal pump and provided many of the early lessons learned for the Space Shuttle Main Engine High Pressure Fuel Turbopumps. This paper will discuss the design trades and analyses performed for the current J-2X FTP to increase turbine life; increase structural margins, facilitate component fabrication; expedite turbopump assembly; and increase rotordynamic stability margins. Risk mitigation tests including inducer water tests, whirligig turbine blade tests, turbine air rig tests, and workhorse gas generator tests characterized operating environments, drove design modifications, or identified performance impact. Engineering design, fabrication, analysis, and assembly activities support FTP readiness for the first J-2X engine test scheduled for July 2011.

  2. Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Licht, J. R.; Bergeron, A.; Dionne, B.

    2015-12-01

    BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water (Figure 1). The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux ofmore » 470 W/cm2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident.« less

  3. Neutronics Investigations for the Lower Part of a Westinghouse SVEA-96+ Assembly

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Murphy, M.F.; Luethi, A.; Seiler, R.

    2002-05-15

    Accurate critical experiments have been performed for the validation of total fission (F{sub tot}) and {sup 238}U-capture (C{sub 8}) reaction rate distributions obtained with CASMO-4, HELIOS, BOXER, and MCNP4B for the lower axial region of a real Westinghouse SVEA-96+ fuel assembly. The assembly comprised fresh fuel with an average {sup 235}U enrichment of 4.02 wt%, a maximum enrichment of 4.74 wt%, 14 burnable-absorber fuel pins, and full-density water moderation. The experimental configuration investigated was core 1A of the LWR-PROTEUS Phase I project, where 61 different fuel pins, representing {approx}64% of the assembly, were gamma-scanned individually. Calculated (C) and measured (E)more » values have been compared in terms of C/E distributions. For F{sub tot}, the standard deviations are 1.2% for HELIOS, 0.9% for CASMO-4, 0.8% for MCNP4B, and 1.7% for BOXER. Standard deviations of 1.1% for HELIOS, CASMO-4, and MCNP4B and 1.2% for BOXER were obtained in the case of C{sub 8}. Despite the high degree of accuracy observed on the average, it was found that the five burnable-absorber fuel pins investigated showed a noticeable underprediction of F{sub tot}, quite systematically, for the deterministic codes evaluated (average C/E for the burnable-absorber fuel pins in the range 0.974 to 0.988, depending on the code)« less

  4. Thermal Hydraulic Computational Fluid Dynamics Simulations and Experimental Investigation of Deformed Fuel Assemblies

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Mays, Brian; Jackson, R. Brian

    2017-03-08

    The project, Toward a Longer Life Core: Thermal Hydraulic CFD Simulations and Experimental Investigation of Deformed Fuel Assemblies, DOE Project code DE-NE0008321, was a verification and validation project for flow and heat transfer through wire wrapped simulated liquid metal fuel assemblies that included both experiments and computational fluid dynamics simulations of those experiments. This project was a two year collaboration between AREVA, TerraPower, Argonne National Laboratory and Texas A&M University. Experiments were performed by AREVA and Texas A&M University. Numerical simulations of these experiments were performed by TerraPower and Argonne National Lab. Project management was performed by AREVA Federal Services.more » The first of a kind project resulted in the production of both local point temperature measurements and local flow mixing experiment data paired with numerical simulation benchmarking of the experiments. The project experiments included the largest wire-wrapped pin assembly Mass Index of Refraction (MIR) experiment in the world, the first known wire-wrapped assembly experiment with deformed duct geometries and the largest numerical simulations ever produced for wire-wrapped bundles.« less

  5. 24 CFR 3285.601 - Field assembly.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 24 Housing and Urban Development 5 2011-04-01 2011-04-01 false Field assembly. 3285.601 Section... § 3285.601 Field assembly. Home manufacturers must provide specific installation instructions for the proper field assembly of manufacturer-supplied and shipped loose ducts, plumbing, and fuel supply system...

  6. 24 CFR 3285.601 - Field assembly.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... 24 Housing and Urban Development 5 2013-04-01 2013-04-01 false Field assembly. 3285.601 Section... § 3285.601 Field assembly. Home manufacturers must provide specific installation instructions for the proper field assembly of manufacturer-supplied and shipped loose ducts, plumbing, and fuel supply system...

  7. 24 CFR 3285.601 - Field assembly.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... 24 Housing and Urban Development 5 2014-04-01 2014-04-01 false Field assembly. 3285.601 Section... § 3285.601 Field assembly. Home manufacturers must provide specific installation instructions for the proper field assembly of manufacturer-supplied and shipped loose ducts, plumbing, and fuel supply system...

  8. 24 CFR 3285.601 - Field assembly.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... 24 Housing and Urban Development 5 2012-04-01 2012-04-01 false Field assembly. 3285.601 Section... § 3285.601 Field assembly. Home manufacturers must provide specific installation instructions for the proper field assembly of manufacturer-supplied and shipped loose ducts, plumbing, and fuel supply system...

  9. DOE Office of Scientific and Technical Information (OSTI.GOV)

    Zhang, Tianfu; Ma, Zhuang; Li, Guoping

    Electrostatic self-assembly in organic solvent without intensively oxidative or corrosive environments, was adopted to prepare Al/Fe{sub 2}O{sub 3}/MWCNT nanostructured energetic materials as an energy generating material. The negatively charged MWCNT was used as a glue-like agent to direct the self-assembly of the well dispersed positively charged Al (fuel) and Fe{sub 2}O{sub 3} (oxide) nanoparticles. This spontaneous assembly method without any surfactant chemistry or other chemical and biological moieties decreased the aggregation of the same nanoparticles largely, moreover, the poor interfacial contact between the Al (fuel) and Fe{sub 2}O{sub 3} (oxide) nanoparticles was improved significantly, which was the key characteristic ofmore » high performance nanostructured energetic materials. In addition, the assembly process was confirmed as Diffusion-Limited Aggregation. The assembled Al/Fe{sub 2}O{sub 3}/MWCNT nanostructured energetic materials showed excellent performance with heat release of 2400 J/g, peak pressure of 0.42 MPa and pressurization rate of 105.71 MPa/s, superior to that in the control group Al/Fe{sub 2}O{sub 3} nanostructured energetic materials prepared by sonication with heat release of 1326 J/g, peak pressure of 0.19 MPa and pressurization rate of 33.33 MPa/s. Therefore, the approach, which is facile, opens a promising route to the high performance nanostructured energetic materials. - Graphical abstract: The negatively charged MWCNT was used as a glue-like agent to direct the self-assembly of the well dispersed positively charged Al (fuel) and Fe{sub 2}O{sub 3} (oxide) nanoparticles. - Highlights: • A facile spontaneous electrostatic assembly strategy without surfactant was adopted. • The fuels and oxidizers assembled into densely packed nanostructured composites. • The assembled nanostructured energetic materials have excellent performance. • This high performance energetic material can be scaled up for practical application. • This strategy can be applied into other nanostructured energetic material systems.« less

  10. Heat transfer enhancement for spent nuclear fuel assembly disposal packages using metallic void fillers: A prevention technique for solidification shrinkage-induced interfacial gaps

    NASA Astrophysics Data System (ADS)

    Park, Yongsoo; McKrell, Thomas J.; Driscoll, Michael J.

    2017-06-01

    This study considers replacing the externally accessible void spaces inside a disposal package containing a spent nuclear fuel assembly (SNFA) with high heat conducting metal to increase the effective thermal conductivity of the package and simplify the heat transfer mechanism inside the package by reducing it to a conduction dominant problem. The focus of the study is on preventing the gaps adjacent to the walls of the package components, produced by solidification shrinkage of poured liquid metal. We approached the problem by providing a temporary coating layer on the components to avoid direct build-up of thick metal oxides on their surface to promote metallic bonding at the interfaces under a non-inert environment. Laboratory scale experiments without SNFA were performed with Zn coated low carbon steel canisters and Zamak-3 void filler under two different filling temperature conditions - below and above the melting point of Zn (designated BMP and AMP respectively). Gap formation was successfully prevented in both cases while we confirmed an open gap in a control experiment, which used an uncoated canister. Minor growth of Al-Fe intermetallic phases was observed at the canister/filler interface of the sample produced under the BMP condition while their growth was significant and showed irregularly distributed morphology in the sample produced under the AMP condition, which has a potential to mitigate excessive residual stresses caused by shrinkage prevention. A procedure for the full-scale application was specified based on the results.

  11. Using the Time-Correlated Induced Fission Method to Simultaneously Measure the 235U Content and the Burnable Poison Content in LWR Fuel Assemblies

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Root, M. A.; Menlove, H. O.; Lanza, R. C.

    The uranium neutron coincidence collar uses thermal neutron interrogation to verify the 235U mass in low-enriched uranium (LEU) fuel assemblies in fuel fabrication facilities. Burnable poisons are commonly added to nuclear fuel to increase the lifetime of the fuel. The high thermal neutron absorption by these poisons reduces the active neutron signal produced by the fuel. Burnable poison correction factors or fast-mode runs with Cd liners can help compensate for this effect, but the correction factors rely on operator declarations of burnable poison content, and fast-mode runs are time-consuming. Finally, this paper describes a new analysis method to measure themore » 235U mass and burnable poison content in LEU nuclear fuel simultaneously in a timely manner, without requiring additional hardware.« less

  12. Using the Time-Correlated Induced Fission Method to Simultaneously Measure the 235U Content and the Burnable Poison Content in LWR Fuel Assemblies

    DOE PAGES

    Root, M. A.; Menlove, H. O.; Lanza, R. C.; ...

    2018-03-21

    The uranium neutron coincidence collar uses thermal neutron interrogation to verify the 235U mass in low-enriched uranium (LEU) fuel assemblies in fuel fabrication facilities. Burnable poisons are commonly added to nuclear fuel to increase the lifetime of the fuel. The high thermal neutron absorption by these poisons reduces the active neutron signal produced by the fuel. Burnable poison correction factors or fast-mode runs with Cd liners can help compensate for this effect, but the correction factors rely on operator declarations of burnable poison content, and fast-mode runs are time-consuming. Finally, this paper describes a new analysis method to measure themore » 235U mass and burnable poison content in LEU nuclear fuel simultaneously in a timely manner, without requiring additional hardware.« less

  13. Molybdenum-99 production calculation analysis of SAMOP reactor based on thorium nitrate fuel

    NASA Astrophysics Data System (ADS)

    Syarip; Togatorop, E.; Yassar

    2018-03-01

    SAMOP (Subcritical Assembly for Molybdenum-99 Production) has the potential to use thorium as fuel to produce 99Mo after modifying the design, but the production performance has not been discovered yet. A study needs to be done to obtain the correlation between 99Mo production with the mixed fuel composition of uranium and with SAMOP power on the modified SAMOP design. The study aims to obtain the production of 99Mo based thorium nitrate fuel on SAMOP’s modified designs. Monte Carlo N-Particle eXtended (MCNPX) is required to simulate the operation of the assembly by varying the composition of the uranium-thorium nitrate mixed fuel, geometry and power fraction on the SAMOP modified designs. The burnup command on the MCNPX is used to confirm the 99Mo production result. The assembly is simulated to operate for 6 days with subcritical neutron multiplication factor (keff = 0.97-0.99). The neutron multiplication factor of the modified design (keff) is 0.97, the activity obtained from 99Mo is 18.58 Ci at 1 kW power operation.

  14. Premixer assembly for mixing air and fuel for combustion

    DOEpatents

    York, William David; Johnson, Thomas Edward; Keener, Christopher Paul

    2016-12-13

    A premixer assembly for mixing air and fuel for combustion includes a plurality of tubes disposed at a head end of a combustor assembly. Also included is a tube of the plurality of tubes, the tube including an inlet end and an outlet end. Further included is at least one non-circular portion of the tube extending along a length of the tube, the at least one non-circular portion having a non-circular cross-section, and the tube having a substantially constant cross-sectional area along its length

  15. A neutronics feasibility study for the LEU conversion of Poland's Maria research reactor.

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Bretscher, M. M.

    1998-10-14

    The MARIA reactor is a high-flux multipurpose research reactor which is water-cooled and moderated with both beryllium and water. Standard HEU (80% {sup 235}U)fuel assemblies consist of six concentric fuel tubes of a U-Al alloy clad in aluminum. Although the inventory of HEU (80%) fuel is nearly exhausted, a supply of highly-loaded 36%-enriched fuel assemblies is available at the reactor site. Neutronic equilibrium studies have been made to determine the relative performance of fuels with enrichments of 80%, 36% and 19.7%. These studies indicate that LEU (19.7%) densities of about 2.5 gU/cm{sup 3} and 3.8 gU/cm{sup 3} are required tomore » match the performance of the MARIA reactor with 80%-enriched and with 36%-enriched fuels, respectively.« less

  16. Analysis of features of hydrodynamics and heat transfer in the fuel assembly of prospective sodium reactor with a high rate of reproduction in the uranium-plutonium fuel cycle

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Lubina, A. S., E-mail: lubina-as@nrcki.ru; Subbotin, A. S.; Sedov, A. A.

    2016-12-15

    The fast sodium reactor fuel assembly (FA) with U–Pu–Zr metallic fuel is described. In comparison with a “classical” fast reactor, this FA contains thin fuel rods and a wider fuel rod grid. Studies of the fluid dynamics and the heat transfer were carried out for such a new FA design. The verification of the ANSYS CFX code was provided for determination of the velocity, pressure, and temperature fields in the different channels. The calculations in the cells and in the FA were carried out using the model of shear stress transport (SST) selected at the stage of verification. The resultsmore » of the hydrodynamics and heat transfer calculations have been analyzed.« less

  17. Aspects of remote maintenance in an FRG reprocessing plant from the manufacturer's viewpoint

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Zeitzchel, G.; Tennie, M.; Saal, G.

    In April 1986 a consortium led by Kraftwerk Union AG was commissioned by the German society for nuclear fuel reprocessing (DWK) to build the first West German commercial reprocessing plant for spent fuel assemblies. The main result of the planning efforts regarding remote maintenance operations inside the main process building was the introduction of FEMO technology (FEMO is an acronym based on German for remote handling modular technique). According to this technology the two cells in which the actual reprocessing (which is based on the PUREX technique) takes place are provided with frames to accommodate the process components (tanks, pumps,more » agitators, etc.), each frame together with the components which it supports forming one module. The two cells are inaccessible and windowless. For handling operations each cell is equipped with an overhead crane and a crane-like manipulator carrier system (MTS) with power manipulator. Viewing of the operations from outside the cells is made possible by television (TV) cameras installed at the crane, the MTS, and the manipulator. This paper addresses some examples of problems that still need to be solved in connection with FEMO handling. In particular, the need for close cooperation between the equipment operator, the component designer, the process engineer, the planning engineer, and the licensing authorities will be demonstrated.« less

  18. HSPES membrane electrode assembly

    NASA Technical Reports Server (NTRS)

    Kindler, Andrew (Inventor); Yen, Shiao-Ping (Inventor)

    2000-01-01

    An improved fuel cell electrode, as well as fuel cells and membrane electrode assemblies that include such an electrode, in which the electrode includes a backing layer having a sintered layer thereon, and a non-sintered free-catalyst layer. The invention also features a method of forming the electrode by sintering a backing material with a catalyst material and then applying a free-catalyst layer.

  19. 5. CONSTRUCTION PROGRESS VIEW OF ASSEMBLY USED TO RAISE AND ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    5. CONSTRUCTION PROGRESS VIEW OF ASSEMBLY USED TO RAISE AND LOWER FUEL ELEMENTS. TAKEN FROM TOP OF SHIELDING TANK WITH CAMERA POINTING TOWARDS BOTTOM OF TANK. SHOWS LADDER, SQUARE LIFTING FRAME, FUEL ELEMENT HOLDERS, AND CABLE CYLINDERS. INEL PHOTO NUMBER 65-5434, TAKEN OCTOBER 20, 1965. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

  20. Catalyst-layer ionomer imaging of fuel cells

    DOE PAGES

    Guetaz, Laure; Lopez-Haro, M.; Escribano, S.; ...

    2015-09-14

    Investigation of membrane/electrode assembly (MEA) microstructure has become an essential step to optimize the MEA components and manufacturing processes or to study the MEA degradation. For these investigations, transmission electron microscopy (TEM) is a tool of choice as it provides direct imaging of the different components. TEM is then widely used for analyzing the catalyst nanoparticles and their carbon support. However, the ionomer inside the electrode is more difficult to be imaged. The difficulties come from the fact that the ionomer forms an ultrathin layer surrounding the carbon particles and in addition, these two components, having similar density, present nomore » difference in contrast. In this paper, we show how the recent progresses in TEM techniques as spherical aberration (Cs) corrected HRTEM, electron tomography and X-EDS elemental mapping provide new possibilities for imaging this ionomer network and consequently to study its degradation.« less

  1. Ultrastable cellulosome-adhesion complex tightens under load.

    PubMed

    Schoeler, Constantin; Malinowska, Klara H; Bernardi, Rafael C; Milles, Lukas F; Jobst, Markus A; Durner, Ellis; Ott, Wolfgang; Fried, Daniel B; Bayer, Edward A; Schulten, Klaus; Gaub, Hermann E; Nash, Michael A

    2014-12-08

    Challenging environments have guided nature in the development of ultrastable protein complexes. Specialized bacteria produce discrete multi-component protein networks called cellulosomes to effectively digest lignocellulosic biomass. While network assembly is enabled by protein interactions with commonplace affinities, we show that certain cellulosomal ligand-receptor interactions exhibit extreme resistance to applied force. Here, we characterize the ligand-receptor complex responsible for substrate anchoring in the Ruminococcus flavefaciens cellulosome using single-molecule force spectroscopy and steered molecular dynamics simulations. The complex withstands forces of 600-750 pN, making it one of the strongest bimolecular interactions reported, equivalent to half the mechanical strength of a covalent bond. Our findings demonstrate force activation and inter-domain stabilization of the complex, and suggest that certain network components serve as mechanical effectors for maintaining network integrity. This detailed understanding of cellulosomal network components may help in the development of biocatalysts for production of fuels and chemicals from renewable plant-derived biomass.

  2. CERCA LEU fuel assemblies testing in Maria Reactor - safety analysis summary and testing program scope.

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Pytel, K.; Mieleszczenko, W.; Lechniak, J.

    2010-03-01

    The presented paper contains neutronic and thermal-hydraulic (for steady and unsteady states) calculation results prepared to support annex to Safety Analysis Report for MARIA reactor in order to obtain approval for program of testing low-enriched uranium (LEU) lead test fuel assemblies (LTFA) manufactured by CERCA. This includes presentation of the limits and operational constraints to be in effect during the fuel testing investigations. Also, the scope of testing program (which began in August 2009), including additional measurements and monitoring procedures, is described.

  3. Methods and systems to thermally protect fuel nozzles in combustion systems

    DOEpatents

    Helmick, David Andrew; Johnson, Thomas Edward; York, William David; Lacy, Benjamin Paul

    2013-12-17

    A method of assembling a gas turbine engine is provided. The method includes coupling a combustor in flow communication with a compressor such that the combustor receives at least some of the air discharged by the compressor. A fuel nozzle assembly is coupled to the combustor and includes at least one fuel nozzle that includes a plurality of interior surfaces, wherein a thermal barrier coating is applied across at least one of the plurality of interior surfaces to facilitate shielding the interior surfaces from combustion gases.

  4. Green-fuel-mediated synthesis of self-assembled NiO nano-sticks for dual applications—photocatalytic activity on Rose Bengal dye and antimicrobial action on bacterial strains

    NASA Astrophysics Data System (ADS)

    Iyyappa Rajan, P.; Vijaya, J. Judith; Jesudoss, S. K.; Kaviyarasu, K.; Kennedy, L. John; Jothiramalingam, R.; Al-Lohedan, Hamad A.; Vaali-Mohammed, Mansoor-Ali

    2017-08-01

    With aim of promoting the employability of green fuels in the synthesis of nano-scaled materials with new kinds of morphologies for multiple applications, successful synthesis of self-assembled NiO nano-sticks was achieved through a 100% green-fuel-mediated hot-plate combustion reaction. The synthesized NiO nano-sticks show excellent photocatalytic activity on Rose Bengal dye and superior antibacterial potential towards both Gram-positive and Gram-negative bacteria.

  5. Advanced membrane electrode assemblies for fuel cells

    DOEpatents

    Kim, Yu Seung; Pivovar, Bryan S.

    2012-07-24

    A method of preparing advanced membrane electrode assemblies (MEA) for use in fuel cells. A base polymer is selected for a base membrane. An electrode composition is selected to optimize properties exhibited by the membrane electrode assembly based on the selection of the base polymer. A property-tuning coating layer composition is selected based on compatibility with the base polymer and the electrode composition. A solvent is selected based on the interaction of the solvent with the base polymer and the property-tuning coating layer composition. The MEA is assembled by preparing the base membrane and then applying the property-tuning coating layer to form a composite membrane. Finally, a catalyst is applied to the composite membrane.

  6. General view of a Space Shuttle Main Engine (SSME) mounted ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    General view of a Space Shuttle Main Engine (SSME) mounted on an SSME engine handler, taken in the SSME Processing Facility at Kennedy Space Center. The most prominent features of the engine assembly in this view are the Low-Pressure Fuel Turbopump Discharge Duct looping diagonally across the top of the assembly and connecting to the High-Pressure Fuel Turbopump, the Low-Pressure Oxidizer Turbopump (LPOTP) located center right of the assembly and the LPOTP Discharge Duct looping around from the pump to the underside of the engine assembly in this view. - Space Transportation System, Space Shuttle Main Engine, Lyndon B. Johnson Space Center, 2101 NASA Parkway, Houston, Harris County, TX

  7. Advanced membrane electrode assemblies for fuel cells

    DOEpatents

    Kim, Yu Seung; Pivovar, Bryan S

    2014-02-25

    A method of preparing advanced membrane electrode assemblies (MEA) for use in fuel cells. A base polymer is selected for a base membrane. An electrode composition is selected to optimize properties exhibited by the membrane electrode assembly based on the selection of the base polymer. A property-tuning coating layer composition is selected based on compatibility with the base polymer and the electrode composition. A solvent is selected based on the interaction of the solvent with the base polymer and the property-tuning coating layer composition. The MEA is assembled by preparing the base membrane and then applying the property-tuning coating layer to form a composite membrane. Finally, a catalyst is applied to the composite membrane.

  8. Validating the technological feasibility of yttria-stabilized zirconia-based semiconducting-ionic composite in intermediate-temperature solid oxide fuel cells

    NASA Astrophysics Data System (ADS)

    Cai, Yixiao; Wang, Baoyuan; Wang, Yi; Xia, Chen; Qiao, Jinli; van Aken, Peter A.; Zhu, Bin; Lund, Peter

    2018-04-01

    YSZ as the electrolyte of choice has dominated the progressive development of solid oxide fuel cell (SOFC) technologies for many years. To enable SOFCs operating at intermediate temperatures of 600 °C or below, major technical advances were built on a foundation of a thin-film YSZ electrolyte, NiO anode, and perovskite cathode, e.g. La0.6Sr0.4Co0.8Fe0.2O3-δ (LSCF). Inspired by functionalities in engineered heterostructure interfaces, the present work uses the components from state-of-the-art SOFCs, i.e, the anode NiO-YSZ and the cathode LSCF-YSZ, or the convergence of all three components, i.e., NiO-YSZ-LSCF, to fabricate semiconductor-ionic membranes (SIMs) and devices. A series of proof-of-concept fuel cell devices are designed by using each of the above SIMs sandwiched between two semiconducting Ni0.8Co0.15Al0.05LiO2-δ (NCAL) layers. We systematically compare these novel designs at 600 °C with two reference fuel cells: a commercial product of anode-supported YSZ electrolyte thin-film cell, and a lab-assembled fuel cell with a conventional configuration of NiO-YSZ (anode)/YSZ (electrolyte)/LSCF-YSZ (cathode). In comparison to the reference cells, the SIM device in a configuration of NCAL/NiO-YSZ-LSCF/NCAL reaches more than 3-fold enhancement of the maximum power output. By using spherical aberration-corrected transmission electron microscopy and spectroscopy approaches, this work offers insight into the mechanisms underlying SIM-associated SOFC performance enhancement.

  9. Th and U fuel photofission study by NTD for AD-MSR subcritical assembly

    NASA Astrophysics Data System (ADS)

    Sajo-Bohus, Laszlo; Greaves, Eduardo D.; Davila, Jesus; Barros, Haydn; Pino, Felix; Barrera, Maria T.; Farina, Fulvio

    2015-07-01

    During the last decade a considerable effort has been devoted for developing energy generating systems based on advanced nuclear technology within the design concepts of GEN-IV. Thorium base fuel systems such as accelerator driven nuclear reactors are one of the often mentioned attractive and affordable options. Several radiotherapy linear accelerators are on the market and due to their reliability, they could be employed as drivers for subcritical liquid fuel assemblies. Bremsstrahlung photons with energies above 5.5MeV, induce (γ,n) and (e,e'n) reactions in the W-target. Resulting gamma radiation and photo or fission neutrons may be absorbed in target materials such as thorium and uranium isotopes to induce sustained fission or nuclear transmutation in waste radioactive materials. Relevant photo driven and photo-fission reaction cross sections are important for actinides 232Th, 238U and 237Np in the radiotherapy machines energy range of 10-20 MV. In this study we employ passive nuclear track detectors (NTD) to determine fission rates and neutron production rates with the aim to establish the feasibility for gamma and photo-neutron driven subcritical assemblies. To cope with these objectives a 20 MV radiotherapy machine has been employed with a mixed fuel target. Results will support further development for a subcritical assembly employing a thorium containing liquid fuel. It is expected that acquired technological knowledge will contribute to the Venezuelan nuclear energy program.

  10. Neutronic fuel element fabrication

    DOEpatents

    Korton, George

    2004-02-24

    This disclosure describes a method for metallurgically bonding a complete leak-tight enclosure to a matrix-type fuel element penetrated longitudinally by a multiplicity of coolant channels. Coolant tubes containing solid filler pins are disposed in the coolant channels. A leak-tight metal enclosure is then formed about the entire assembly of fuel matrix, coolant tubes and pins. The completely enclosed and sealed assembly is exposed to a high temperature and pressure gas environment to effect a metallurgical bond between all contacting surfaces therein. The ends of the assembly are then machined away to expose the pin ends which are chemically leached from the coolant tubes to leave the coolant tubes with internal coolant passageways. The invention described herein was made in the course of, or under, a contract with the U.S. Atomic Energy Commission. It relates generally to fuel elements for neutronic reactors and more particularly to a method for providing a leak-tight metal enclosure for a high-performance matrix-type fuel element penetrated longitudinally by a multiplicity of coolant tubes. The planned utilization of nuclear energy in high-performance, compact-propulsion and mobile power-generation systems has necessitated the development of fuel elements capable of operating at high power densities. High power densities in turn require fuel elements having high thermal conductivities and good fuel retention capabilities at high temperatures. A metal clad fuel element containing a ceramic phase of fuel intimately mixed with and bonded to a continuous refractory metal matrix has been found to satisfy the above requirements. Metal coolant tubes penetrate the matrix to afford internal cooling to the fuel element while providing positive fuel retention and containment of fission products generated within the fuel matrix. Metal header plates are bonded to the coolant tubes at each end of the fuel element and a metal cladding or can completes the fuel-matrix enclosure by encompassing the sides of the fuel element between the header plates.

  11. Sustained Recycle in Light Water and Sodium-Cooled Reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Steven J. Piet; Samuel E. Bays; Michael A. Pope

    2010-11-01

    From a physics standpoint, it is feasible to sustain recycle of used fuel in either thermal or fast reactors. This paper examines multi-recycle potential performance by considering three recycling approaches and calculating several fuel cycle parameters, including heat, gamma, and neutron emission of fresh fuel; radiotoxicity of waste; and uranium utilization. The first recycle approach is homogeneous mixed oxide (MOX) fuel assemblies in a light water reactor (LWR). The transuranic portion of the MOX was varied among Pu, NpPu, NpPuAm, or all-TRU. (All-TRU means all isotopes through Cf-252.) The Pu case was allowed to go to 10% Pu in freshmore » fuel, but when the minor actinides were included, the transuranic enrichment was kept below 8% to satisfy the expected void reactivity constraint. The uranium portion of the MOX was enriched uranium. That enrichment was increased (to as much as 6.5%) to keep the fuel critical for a typical LWR irradiation. The second approach uses heterogeneous inert matrix fuel (IMF) assemblies in an LWR - a mix of IMF and traditional UOX pins. The uranium-free IMF fuel pins were Pu, NpPu, NpPuAm, or all-TRU. The UOX pins were limited to 4.95% U-235 enrichment. The number of IMF pins was set so that the amount of TRU in discharged fuel from recycle N (from both IMF and UOX pins) was made into the new IMF pins for recycle N+1. Up to 60 of the 264 pins in a fuel assembly were IMF. The assembly-average TRU content was 1-6%. The third approach uses fast reactor oxide fuel in a sodium-cooled fast reactor with transuranic conversion ratio of 0.50 and 1.00. The transuranic conversion ratio is the production of transuranics divided by destruction of transuranics. The FR at CR=0.50 is similar to the CR for the MOX case. The fast reactor cases had a transuranic content of 33-38%, higher than IMF or MOX.« less

  12. Nuclear reactor fuel assembly duct-tube-to-inlet-nozzle attachment system

    DOEpatents

    Christiansen, David W.; Smith, Bob G.

    1982-01-01

    A reusable system for removably attaching the lower end 21 of a nuclear reactor fuel assembly duct tube to an upper end 11 of a nuclear reactor fuel assembly inlet nozzle. The duct tube's lower end 21 has sides terminating in locking tabs 22 which end in inwardly-extending flanges 23. The flanges 23 engage recesses 13 in the top section 12 of the inlet nozzle's upper end 11. A retaining collar 30 slides over the inlet nozzle's upper end 11 to restrain the flanges 23 in the recesses 13. A locking nut 40 has an inside threaded portion 41 which engages an outside threaded portion 15 of the inlet nozzle's upper end 11 to secure the retaining collar 30 against protrusions 24 on the duct tube's sides.

  13. DOE Office of Scientific and Technical Information (OSTI.GOV)

    Lafleur, Adrienne M.; Ulrich, Timothy J. II; Menlove, Howard O.

    Objective is to investigate the use of Passive Neutron Albedo Reactivity (PNAR) and Self-Interrogation Neutron Resonance Densitometry (SINRD) to quantify fissile content in FUGEN spent fuel assemblies (FAs). Methodology used is: (1) Detector was designed using fission chambers (FCs); (2) Optimized design via MCNPX simulations; and (3) Plan to build and field test instrument in FY13. Significance was to improve safeguards verification of spent fuel assemblies in water and increase sensitivity to partial defects. MCNPX simulations were performed to optimize the design of the SINRD+PNAR detector. PNAR ratio was less sensitive to FA positioning than SINRD and SINRD ratio wasmore » more sensitive to Pu fissile mass than PNAR. Significance was that the integration of these techniques can be used to improve verification of spent fuel assemblies in water.« less

  14. Engine-integrated solid oxide fuel cells for efficient electrical power generation on aircraft

    NASA Astrophysics Data System (ADS)

    Waters, Daniel F.; Cadou, Christopher P.

    2015-06-01

    This work investigates the use of engine-integrated catalytic partial oxidation (CPOx) reactors and solid oxide fuel cells (SOFCs) to reduce fuel burn in vehicles with large electrical loads like sensor-laden unmanned air vehicles. Thermodynamic models of SOFCs, CPOx reactors, and three gas turbine (GT) engine types (turbojet, combined exhaust turbofan, separate exhaust turbofan) are developed and checked against relevant data and source material. Fuel efficiency is increased by 4% and 8% in the 50 kW and 90 kW separate exhaust turbofan systems respectively at only modest cost in specific power (8% and 13% reductions respectively). Similar results are achieved in other engine types. An additional benefit of hybridization is the ability to provide more electric power (factors of 3 or more in some cases) than generator-based systems before encountering turbine inlet temperature limits. A sensitivity analysis shows that the most important parameters affecting the system's performance are operating voltage, percent fuel oxidation, and SOFC assembly air flows. Taken together, this study shows that it is possible to create a GT-SOFC hybrid where the GT mitigates balance of plant losses and the SOFC raises overall system efficiency. The result is a synergistic system with better overall performance than stand-alone components.

  15. Next Generation Safeguards Initiative research to determine the Pu mass in spent fuel assemblies: Purpose, approach, constraints, implementation, and calibration

    NASA Astrophysics Data System (ADS)

    Tobin, S. J.; Menlove, H. O.; Swinhoe, M. T.; Schear, M. A.

    2011-10-01

    The Next Generation Safeguards Initiative (NGSI) of the U.S. Department of Energy has funded a multi-lab/multi-university collaboration to quantify the plutonium mass in spent nuclear fuel assemblies and to detect the diversion of pins from them. The goal of this research effort is to quantify the capability of various non-destructive assay (NDA) technologies as well as to train a future generation of safeguards practitioners. This research is "technology driven" in the sense that we will quantify the capabilities of a wide range of safeguards technologies of interest to regulators and policy makers; a key benefit to this approach is that the techniques are being tested in a unified manner. When the results of the Monte Carlo modeling are evaluated and integrated, practical constraints are part of defining the potential context in which a given technology might be applied. This paper organizes the commercial spent fuel safeguard needs into four facility types in order to identify any constraints on the NDA system design. These four facility types are the following: future reprocessing plants, current reprocessing plants, once-through spent fuel repositories, and any other sites that store individual spent fuel assemblies (reactor sites are the most common facility type in this category). Dry storage is not of interest since individual assemblies are not accessible. This paper will overview the purpose and approach of the NGSI spent fuel effort and describe the constraints inherent in commercial fuel facilities. It will conclude by discussing implementation and calibration of measurement systems. This report will also provide some motivation for considering a couple of other safeguards concepts (base measurement and fingerprinting) that might meet the safeguards need but not require the determination of plutonium mass.

  16. Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor [Novel Methodology and Software for Spent Fuel Gross Defect Detection at the Atucha-I Reactor

    DOE PAGES

    Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek; ...

    2017-03-27

    Here, fuel assemblies in the spent fuel pool are stored by suspending them in two vertically stacked layers at the Atucha Unit 1 nuclear power plant (Atucha-I). This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 wt% 235U and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A gross defect detection tool, the spent fuel neutron counter (SFNC), has been used at themore » site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups, the existing signal processing software of the tool was found to fail due to nonlinearity of the source term with burnup.« less

  17. Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor [Novel Methodology and Software for Spent Fuel Gross Defect Detection at the Atucha-I Reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek

    Here, fuel assemblies in the spent fuel pool are stored by suspending them in two vertically stacked layers at the Atucha Unit 1 nuclear power plant (Atucha-I). This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 wt% 235U and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A gross defect detection tool, the spent fuel neutron counter (SFNC), has been used at themore » site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups, the existing signal processing software of the tool was found to fail due to nonlinearity of the source term with burnup.« less

  18. Materials and Fuels Complex Tour

    ScienceCinema

    Miley, Don

    2017-12-11

    The Materials and Fuels Complex at Idaho National Laboratory is home to several facilities used for the research and development of nuclear fuels. Stops include the Fuel Conditioning Facility, the Hot Fuel Examination Facility (post-irradiation examination), and the Space and Security Power System Facility, where radioisotope thermoelectric generators (RTGs) are assembled for deep space missions.

  19. 30 CFR 36.27 - Fuel-supply system.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... tank by a chain or other fastening to prevent loss. (2) The fuel tank shall have a definite position in the equipment assembly, and no provision shall be made for attachment of separate or auxiliary fuel... continuously at full load for approximately four hours. (b) Fuel lines. All fuel lines shall be installed to...

  20. 30 CFR 36.27 - Fuel-supply system.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... tank by a chain or other fastening to prevent loss. (2) The fuel tank shall have a definite position in the equipment assembly, and no provision shall be made for attachment of separate or auxiliary fuel... continuously at full load for approximately four hours. (b) Fuel lines. All fuel lines shall be installed to...

  1. 30 CFR 36.27 - Fuel-supply system.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... tank by a chain or other fastening to prevent loss. (2) The fuel tank shall have a definite position in the equipment assembly, and no provision shall be made for attachment of separate or auxiliary fuel... continuously at full load for approximately four hours. (b) Fuel lines. All fuel lines shall be installed to...

  2. 30 CFR 36.27 - Fuel-supply system.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... tank by a chain or other fastening to prevent loss. (2) The fuel tank shall have a definite position in the equipment assembly, and no provision shall be made for attachment of separate or auxiliary fuel... continuously at full load for approximately four hours. (b) Fuel lines. All fuel lines shall be installed to...

  3. 30 CFR 36.27 - Fuel-supply system.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... tank by a chain or other fastening to prevent loss. (2) The fuel tank shall have a definite position in the equipment assembly, and no provision shall be made for attachment of separate or auxiliary fuel... continuously at full load for approximately four hours. (b) Fuel lines. All fuel lines shall be installed to...

  4. In-field Calibration of a Fast Neutron Collar for the Measurement of Fresh PWR Fuel Assemblies

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Swinhoe, Martyn Thomas; De Baere, Paul

    2015-04-17

    A new neutron collar has been designed for the measurement of fresh LEU fuel assemblies. This collar uses “fast mode” measurement to reduce the effect of burnable poison rods on the assay and thus reduce the dependence on the operator’s declaration. The new collar design reduces effect of poison rods considerably. Instead of 12 pins of 5.2% Gd causing a 20.4% effect, as in the standard thermal mode collar, they only cause a 3.2% effect in the new collar. However it has higher efficiency so that reasonably precise measurements can be made in 25 minutes, rather than the 1 hourmore » of previous collars. The new collar is fully compatible with the use of the standard data collection and analysis code INCC. This report describes the calibration that was made with a mock-up assembly at Los Alamos National Laboratory and with actual assemblies at the AREVA Fuel fabrication Plant in Lingen, Germany.« less

  5. Californium interrogation prompt neutron (CIPN) instrument for non-destructive assay of spent nuclear fuel – design concept and experimental demonstration

    DOE PAGES

    Henzlova, Daniela; Menlove, Howard Olsen; Rael, Carlos D.; ...

    2015-10-09

    Our paper presents results of the first experimental demonstration of the Californium Interrogation Prompt Neutron (CIPN) instrument developed within a multi-year effort launched by the Next Generation Safeguards Initiative Spent Fuel Project of the United States Department of Energy. The goals of this project focused on developing viable non-destructive assay techniques with capabilities to improve an independent verification of spent fuel assembly characteristics. For this purpose, the CIPN instrument combines active and passive neutron interrogation, along with passive gamma-ray measurements, to provide three independent observables. We describe the initial feasibility demonstration of the CIPN instrument, which involved measurements of fourmore » pressurized-water-reactor spent fuel assemblies with different levels of burnup and two initial enrichments. The measurements were performed at the Post-Irradiation Examination Facility at the Korea Atomic Energy Institute in the Republic of Korea. The key aim of the demonstration was to evaluate CIPN instrument performance under realistic deployment conditions, with the focus on a detailed assessment of systematic uncertainties that are best evaluated experimentally. The measurements revealed good positioning reproducibility, as well as a high degree of insensitivity of the CIPN instrument's response to irregularities in a radial burnup profile. Systematic uncertainty of individual CIPN instrument signals due to assembly rotation was found to be <4.5%, even for assemblies with fairly extreme gradients in the radial burnup profile. Lastly, these features suggest that the CIPN instrument is capable of providing a good representation of assembly average characteristics, independent of assembly orientation in the instrument.« less

  6. Californium interrogation prompt neutron (CIPN) instrument for non-destructive assay of spent nuclear fuel – design concept and experimental demonstration

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Henzlova, Daniela; Menlove, Howard Olsen; Rael, Carlos D.

    Our paper presents results of the first experimental demonstration of the Californium Interrogation Prompt Neutron (CIPN) instrument developed within a multi-year effort launched by the Next Generation Safeguards Initiative Spent Fuel Project of the United States Department of Energy. The goals of this project focused on developing viable non-destructive assay techniques with capabilities to improve an independent verification of spent fuel assembly characteristics. For this purpose, the CIPN instrument combines active and passive neutron interrogation, along with passive gamma-ray measurements, to provide three independent observables. We describe the initial feasibility demonstration of the CIPN instrument, which involved measurements of fourmore » pressurized-water-reactor spent fuel assemblies with different levels of burnup and two initial enrichments. The measurements were performed at the Post-Irradiation Examination Facility at the Korea Atomic Energy Institute in the Republic of Korea. The key aim of the demonstration was to evaluate CIPN instrument performance under realistic deployment conditions, with the focus on a detailed assessment of systematic uncertainties that are best evaluated experimentally. The measurements revealed good positioning reproducibility, as well as a high degree of insensitivity of the CIPN instrument's response to irregularities in a radial burnup profile. Systematic uncertainty of individual CIPN instrument signals due to assembly rotation was found to be <4.5%, even for assemblies with fairly extreme gradients in the radial burnup profile. Lastly, these features suggest that the CIPN instrument is capable of providing a good representation of assembly average characteristics, independent of assembly orientation in the instrument.« less

  7. Finite Element Stress Analysis of Spent Nuclear Fuel Disposal Canister in a Deep Geological Repository

    NASA Astrophysics Data System (ADS)

    Kwon, Young Joo; Choi, Jong Won

    This paper presents the finite element stress analysis of a spent nuclear fuel disposal canister to provide basic information for dimensioning the canister and configuration of canister components and consequently to suggest the structural analysis methodology for the disposal canister in a deep geological repository which is nowadays very important in the environmental waste treatment technology. Because of big differences in the pressurized water reactor (PWR) and the Canadian deuterium and uranium reactor (CANDU) fuel properties, two types of canisters are conceived. For manufacturing, operational reasons and standardization, however, both canisters have the same outer diameter and length. The construction type of canisters introduced here is a solid structure with a cast insert and a corrosion resistant overpack. The structural stress analysis is carried out using a finite element analysis code, NISA, and focused on the structural strength of the canister against the expected external pressures due to the swelling of the bentonite buffer and the hydrostatic head. The canister must withstand these large pressure loads. Consequently, canisters presented here contain 4 PWR fuel assemblies and 33×9 CANDU fuel bundles. The outside diameter of the canister for both fuels is 122cm and the cast insert diameter is 112cm. The total length of the canister is 483cm with the lid/bottom and the outer shell of 5cm.

  8. NUCLEAR REACTOR FUEL ELEMENT ASSEMBLY

    DOEpatents

    Stengel, F.G.

    1963-12-24

    A method of fabricating nuclear reactor fuel element assemblies having a plurality of longitudinally extending flat fuel elements in spaced parallel relation to each other to form channels is presented. One side of a flat side plate is held contiguous to the ends of the elements and a welding means is passed along the other side of the platertransverse to the direction of the longitudinal extension of the elements. The setting and speed of travel of the welding means is set to cause penetration of the side plate with welds at bridge the gap in each channel between adjacent fuel elements with a weld-through bubble of predetermined size. The fabrication of a high strength, dependable fuel element is provided, and the reduction of distortion and high production costs are facilitated by this method. (AEC)

  9. VERA and VERA-EDU 3.5 Release Notes

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Sieger, Matt; Salko, Robert K.; Kochunas, Brendan M.

    The Virtual Environment for Reactor Applications components included in this distribution include selected computational tools and supporting infrastructure that solve neutronics, thermal-hydraulics, fuel performance, and coupled neutronics-thermal hydraulics problems. The infrastructure components provide a simplified common user input capability and provide for the physics integration with data transfer and coupled-physics iterative solution algorithms. Neutronics analysis can be performed for 2D lattices, 2D core and 3D core problems for pressurized water reactor geometries that can be used to calculate criticality and fission rate distributions by pin for input fuel compositions. MPACT uses the Method of Characteristics transport approach for 2D problems.more » For 3D problems, MPACT uses the 2D/1D method which uses 2D MOC in a radial plane and diffusion or SPn in the axial direction. MPACT includes integrated cross section capabilities that provide problem-specific cross sections generated using the subgroup methodology. The code can be executed both 2D and 3D problems in parallel to reduce overall run time. A thermal-hydraulics capability is provided with CTF (an updated version of COBRA-TF) that allows thermal-hydraulics analyses for single and multiple assemblies using the simplified VERA common input. This distribution also includes coupled neutronics/thermal-hydraulics capabilities to allow calculations using MPACT coupled with CTF. The VERA fuel rod performance component BISON calculates, on a 2D or 3D basis, fuel rod temperature, fuel rod internal pressure, free gas volume, clad integrity and fuel rod waterside diameter. These capabilities allow simulation of power cycling, fuel conditioning and deconditioning, high burnup performance, power uprate scoping studies, and accident performance. Input/Output capabilities include the VERA Common Input (VERAIn) script which converts the ASCII common input file to the intermediate XML used to drive all of the physics codes in the VERA Core Simulator (VERA-CS). VERA component codes either input the VERA XML format directly, or provide a preprocessor which can convert the XML into native input. VERAView is an interactive graphical interface for the visualization and engineering analyses of output data from VERA. The python-based software is easy to install and intuitive to use, and provides instantaneous 2D and 3D images, 1D plots, and alpha-numeric data from VERA multi-physics simulations. Testing within CASL has focused primarily on Westinghouse four-loop reactor geometries and conditions with example problems included in the distribution.« less

  10. High performance internal reforming unit for high temperature fuel cells

    DOEpatents

    Ma, Zhiwen [Sandy Hook, CT; Venkataraman, Ramakrishnan [New Milford, CT; Novacco, Lawrence J [Brookfield, CT

    2008-10-07

    A fuel reformer having an enclosure with first and second opposing surfaces, a sidewall connecting the first and second opposing surfaces and an inlet port and an outlet port in the sidewall. A plate assembly supporting a catalyst and baffles are also disposed in the enclosure. A main baffle extends into the enclosure from a point of the sidewall between the inlet and outlet ports. The main baffle cooperates with the enclosure and the plate assembly to establish a path for the flow of fuel gas through the reformer from the inlet port to the outlet port. At least a first directing baffle extends in the enclosure from one of the sidewall and the main baffle and cooperates with the plate assembly and the enclosure to alter the gas flow path. Desired graded catalyst loading pattern has been defined for optimized thermal management for the internal reforming high temperature fuel cells so as to achieve high cell performance.

  11. Design of a nuclear isotope heat source assembly for a spaceborne mini-Brayton power module.

    NASA Technical Reports Server (NTRS)

    Wein, D.; Gorland, S. H.

    1973-01-01

    Results of a study to develop a feasible design definition of a heat source assembly (HSA) for use in nominal 500-, 1200-, or 2000-W(e) mini-Brayton spacecraft power systems. The HSA is a modular design which is used either as a single unit to provide thermal energy to the 500-W(e) mini-Brayton power module or in parallel with one or two additional HSAs for the 1200- or 2000-W(e) power module systems. Principal components consist of a multihundred watt RTG isotope heat source, a heat source heat exchanger which transfers the thermal energy from the heat source to the mini-Brayton power conversion system, an auxiliary cooling system which provides requisite cooling during nonoperation of the power conversion module and an emergency cooling system which precludes accidental release of isotope fuel in the event of system failure.

  12. Development of high strength, high temperature ceramics

    NASA Technical Reports Server (NTRS)

    Hall, W. B.

    1982-01-01

    Improvement in the high-pressure turbopumps, both fuel and oxidizer, in the Space Shuttle main engine were considered. The operation of these pumps is limited by temperature restrictions of the metallic components used in these pumps. Ceramic materials that retain strength at high temperatures and appear to be promising candidates for use as turbine blades and impellers are discussed. These high strength materials are sensitive to many related processing parameters such as impurities, sintering aids, reaction aids, particle size, processing temperature, and post thermal treatment. The specific objectives of the study were to: (1) identify and define the processing parameters that affect the properties of Si3N4 ceramic materials, (2) design and assembly equipment required for processing high strength ceramics, (3) design and assemble test apparatus for evaluating the high temperature properties of Si3N4, and (4) conduct a research program of manufacturing and evaluating Si3N4 materials as applicable to rocket engine applications.

  13. Modeling Energy Efficiency As A Green Logistics Component In Vehicle Assembly Line

    NASA Astrophysics Data System (ADS)

    Oumer, Abduaziz; Mekbib Atnaw, Samson; Kie Cheng, Jack; Singh, Lakveer

    2016-11-01

    This paper uses System Dynamics (SD) simulation to investigate the concept green logistics in terms of energy efficiency in automotive industry. The car manufacturing industry is considered to be one of the highest energy consuming industries. An efficient decision making model is proposed that capture the impacts of strategic decisions on energy consumption and environmental sustainability. The sources of energy considered in this research are electricity and fuel; which are the two main types of energy sources used in a typical vehicle assembly plant. The model depicts the performance measurement for process- specific energy measures of painting, welding, and assembling processes. SD is the chosen simulation method and the main green logistics issues considered are Carbon Dioxide (CO2) emission and energy utilization. The model will assist decision makers acquire an in-depth understanding of relationship between high level planning and low level operation activities on production, environmental impacts and costs associated. The results of the SD model signify the existence of positive trade-offs between green practices of energy efficiency and the reduction of CO2 emission.

  14. 14 CFR 23.994 - Fuel system components.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel system components. 23.994 Section 23... AIRWORTHINESS STANDARDS: NORMAL, UTILITY, ACROBATIC, AND COMMUTER CATEGORY AIRPLANES Powerplant Fuel System Components § 23.994 Fuel system components. Fuel system components in an engine nacelle or in the fuselage...

  15. 14 CFR 25.994 - Fuel system components.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 14 Aeronautics and Space 1 2013-01-01 2013-01-01 false Fuel system components. 25.994 Section 25... AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System Components § 25.994 Fuel system components. Fuel system components in an engine nacelle or in the fuselage must be protected from damage...

  16. 14 CFR 23.994 - Fuel system components.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 14 Aeronautics and Space 1 2012-01-01 2012-01-01 false Fuel system components. 23.994 Section 23... AIRWORTHINESS STANDARDS: NORMAL, UTILITY, ACROBATIC, AND COMMUTER CATEGORY AIRPLANES Powerplant Fuel System Components § 23.994 Fuel system components. Fuel system components in an engine nacelle or in the fuselage...

  17. 14 CFR 23.994 - Fuel system components.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 14 Aeronautics and Space 1 2013-01-01 2013-01-01 false Fuel system components. 23.994 Section 23... AIRWORTHINESS STANDARDS: NORMAL, UTILITY, ACROBATIC, AND COMMUTER CATEGORY AIRPLANES Powerplant Fuel System Components § 23.994 Fuel system components. Fuel system components in an engine nacelle or in the fuselage...

  18. 14 CFR 23.994 - Fuel system components.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 14 Aeronautics and Space 1 2011-01-01 2011-01-01 false Fuel system components. 23.994 Section 23... AIRWORTHINESS STANDARDS: NORMAL, UTILITY, ACROBATIC, AND COMMUTER CATEGORY AIRPLANES Powerplant Fuel System Components § 23.994 Fuel system components. Fuel system components in an engine nacelle or in the fuselage...

  19. 14 CFR 25.994 - Fuel system components.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 14 Aeronautics and Space 1 2011-01-01 2011-01-01 false Fuel system components. 25.994 Section 25... AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System Components § 25.994 Fuel system components. Fuel system components in an engine nacelle or in the fuselage must be protected from damage...

  20. 14 CFR 25.994 - Fuel system components.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 14 Aeronautics and Space 1 2012-01-01 2012-01-01 false Fuel system components. 25.994 Section 25... AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System Components § 25.994 Fuel system components. Fuel system components in an engine nacelle or in the fuselage must be protected from damage...

  1. 14 CFR 25.994 - Fuel system components.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 14 Aeronautics and Space 1 2014-01-01 2014-01-01 false Fuel system components. 25.994 Section 25... AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System Components § 25.994 Fuel system components. Fuel system components in an engine nacelle or in the fuselage must be protected from damage...

  2. 14 CFR 23.994 - Fuel system components.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 14 Aeronautics and Space 1 2014-01-01 2014-01-01 false Fuel system components. 23.994 Section 23... AIRWORTHINESS STANDARDS: NORMAL, UTILITY, ACROBATIC, AND COMMUTER CATEGORY AIRPLANES Powerplant Fuel System Components § 23.994 Fuel system components. Fuel system components in an engine nacelle or in the fuselage...

  3. 14 CFR 25.994 - Fuel system components.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel system components. 25.994 Section 25... AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System Components § 25.994 Fuel system components. Fuel system components in an engine nacelle or in the fuselage must be protected from damage...

  4. SSME component assembly and life management expert system

    NASA Technical Reports Server (NTRS)

    Ali, M.; Dietz, W. E.; Ferber, H. J.

    1989-01-01

    The space shuttle utilizes several rocket engine systems, all of which must function with a high degree of reliability for successful mission completion. The space shuttle main engine (SSME) is by far the most complex of the rocket engine systems and is designed to be reusable. The reusability of spacecraft systems introduces many problems related to testing, reliability, and logistics. Components must be assembled from parts inventories in a manner which will most effectively utilize the available parts. Assembly must be scheduled to efficiently utilize available assembly benches while still maintaining flight schedules. Assembled components must be assigned to as many contiguous flights as possible, to minimize component changes. Each component must undergo a rigorous testing program prior to flight. In addition, testing and assembly of flight engines and components must be done in conjunction with the assembly and testing of developmental engines and components. The development, testing, manufacture, and flight assignments of the engine fleet involves the satisfaction of many logistical and operational requirements, subject to many constraints. The purpose of the SSME Component Assembly and Life Management Expert System (CALMES) is to assist the engine assembly and scheduling process, and to insure that these activities utilize available resources as efficiently as possible.

  5. Method for distributed object communications based on dynamically acquired and assembled software components

    NASA Technical Reports Server (NTRS)

    Sundermier, Amy (Inventor)

    2002-01-01

    A method for acquiring and assembling software components at execution time into a client program, where the components may be acquired from remote networked servers is disclosed. The acquired components are assembled according to knowledge represented within one or more acquired mediating components. A mediating component implements knowledge of an object model. A mediating component uses its implemented object model knowledge, acquired component class information and polymorphism to assemble components into an interacting program at execution time. The interactions or abstract relationships between components in the object model may be implemented by the mediating component as direct invocations or indirect events or software bus exchanges. The acquired components may establish communications with remote servers. The acquired components may also present a user interface representing data to be exchanged with the remote servers. The mediating components may be assembled into layers, allowing arbitrarily complex programs to be constructed at execution time.

  6. Road user fee task force report to the 72nd Oregon Legislative Assembly on the possible alternatives of the current system of taxing highway use through motor vehicle fuel taxes

    DOT National Transportation Integrated Search

    2003-03-01

    Recognizing that the fuel tax is a declining revenue source for Oregon's road system, the 2001 Oregon Legislative Assembly passed House Bill 3946, mandating the formation of the Road User Fee Task Force with the mission to develop a design for revenu...

  7. Manifold seal for fuel cell stack assembly

    DOEpatents

    Schmitten, Phillip F.; Wright, Maynard K.

    1989-01-01

    An assembly for sealing a manifold to a stack of fuel cells includes a first resilient member for providing a first sealing barrier between the manifold and the stack. A second resilient member provides a second sealing barrier between the manifold and the stack. The first and second resilient members are retained in such a manner as to define an area therebetween adapted for retaining a sealing composition.

  8. Estimation of carbon 14 inventory in hull and end-piece wastes from Japanese commercial reprocessing operation

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Tomofumi Sakuragi; Hiromi Tanabe; Emiko Hirose

    2013-07-01

    Hull and end-piece wastes generated from reprocessing plant operations are expected to be disposed of in a deep underground repository as Group 2 TRU wastes under the Japanese classification system. The activated metals that compose the spent fuel assemblies such as Zircaloy claddings and stainless steel nozzles are mixed and compressed after fuel dissolution, and then stuffed into stainless steel canisters. Carbon 14 is a typical activated product in the hulls and end-pieces and is mainly generated by the {sup 14}N(n,p){sup 14}C reaction. In the previous safety assessment of the TRU waste in Japan, the radionuclides inventory was calculated bymore » ORIGEN-2 code. Some conservative assumptions and preliminary estimates were used in this calculation. For example, total radionuclides generated from a single type of fuel assembly (45 GWd/tU for a PWR unit), and the thickness of the Zircaloy oxide film on the hulls (80 μm) were both overestimated. The second assumption in particular has a large effect on exposure dose evaluation. Therefore, it is essential to have a realistic source term evaluation regarding such items as the C-14 inventory and its distribution to waste parts. In the present study, a C-14 inventory of the hull and end-piece wastes from the operation of a commercial reprocessing plant in Japan corresponding to 32,000 tU (16,000 tU in each BWR and PWR) was calculated. Analysis using individual irradiation conditions and fuel characteristics was conducted on 6 types of fuel assemblies for BWRs and 12 types for PWRs (4 pile types x 3 burnup limits). The oxide film thickness data for each fuel type cladding were obtained from the published literature. Activation calculations were performed by using ORIGEN-2 code. For the amount of spent assembly and other waste characteristics, representative values were assumed based on the published literature. As a preliminary experiment, C-14 in irradiated BWR claddings was measured and found to be consistent with the calculated activation. The total C-14 inventory was estimated as 4.46x10{sup 14} Bq, consisting of 2.58x10{sup 14} Bq for BWRs and 1.87x10{sup 14} Bq for PWRs, and is consistent with the safety assessment of 4.4x10{sup 14} Bq. However, the distribution of the C-14 inventory to hull oxide, which was estimated under the assumption of instantaneous radionuclide release in the safety assessment, decreased from 5.72x10{sup 13} Bq (13% of the total) in the previous assessment to 1.30x10{sup 13} Bq (2.9% of the total; consisting of 1.48x10{sup 12} for BWRs and 1.15x10{sup 13} for PWRs). In other words, the exposure dose peak is reduced to approximate 25% of its previous value due to the use of detailed oxide film data that the BWR cladding has a thin oxide film. Other instantaneous release components for C-14 such as the fuel residual were negligible. (authors)« less

  9. Fuel assembly shaker and truck test simulation

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Klymyshyn, Nicholas A.; Jensen, Philip J.; Sanborn, Scott E.

    2014-09-30

    This study continues the modeling support of the SNL shaker table task from 2013 and includes analysis of the SNL 2014 truck test campaign. Detailed finite element models of the fuel assembly surrogate used by SNL during testing form the basis of the modeling effort. Additional analysis was performed to characterize and filter the accelerometer data collected during the SNL testing. The detailed fuel assembly finite element model was modified to improve the performance and accuracy of the original surrogate fuel assembly model in an attempt to achieve a closer agreement with the low strains measured during testing. The revisedmore » model was used to recalculate the shaker table load response from the 2013 test campaign. As it happened, the results remained comparable to the values calculated with the original fuel assembly model. From this it is concluded that the original model was suitable for the task and the improvements to the model were not able to bring the calculated strain values down to the extremely low level recorded during testing. The model needs more precision to calculate strains that are so close to zero. The truck test load case had an even lower magnitude than the shaker table case. Strain gage data from the test was compared directly to locations on the model. Truck test strains were lower than the shaker table case, but the model achieved a better relative agreement of 100-200 microstrains (or 0.0001-0.0002 mm/mm). The truck test data included a number of accelerometers at various locations on the truck bed, surrogate basket, and surrogate fuel assembly. This set of accelerometers allowed an evaluation of the dynamics of the conveyance system used in testing. It was discovered that the dynamic load transference through the conveyance has a strong frequency-range dependency. This suggests that different conveyance configurations could behave differently and transmit different magnitudes of loads to the fuel even when traveling down the same road at the same speed. It is recommended that the SNL conveyance system used in testing be characterized through modal analysis and frequency response analysis to provide context and assist in the interpretation of the strain data that was collected during the truck test campaign.« less

  10. Microscale X-ray tomographic investigation of the interfacial morphology between the catalyst and micro porous layers in proton exchange membrane fuel cells

    NASA Astrophysics Data System (ADS)

    Prass, Sebastian; Hasanpour, Sadegh; Sow, Pradeep Kumar; Phillion, André B.; Mérida, Walter

    2016-07-01

    The interfacial morphology between the catalyst layer (CL) and micro porous layer (MPL) influences the performance of proton exchange membrane fuel cells (PEMFCs). Here we report a direct method to investigate the CL-MPL interfacial morphology of stacked and compressed gas diffusion layer (GDL with MPL)-catalyst coated membrane (CCM) assemblies. The area, origin and dimensions of interfacial gaps are studied with high-resolution X-ray micro computed tomography (X-μCT). The projected gap area (fraction of the CL-MPL interface separated by gaps) is higher for GDL-CCM assemblies with large differences in the surface roughness between CL and MPL but reduces with increasing compression and similarity in roughness. Relatively large continuous gaps are found in proximity to cracks in the MPL. These are hypothesized to form due to the presence of large pores on the surface of the GDL. Smaller gaps are induced by the surface roughness features throughout the CL-MPL interface. By modification of the pore sizes on the GDL surface serving as substrate for the MPL, the number and dimension of MPL crack induced gaps can be manipulated. Moreover, adjusting the CL and MPL surface roughness parameters to achieve similar orders of roughness can improve the surface mating characteristics of these two components.

  11. The underwater coincidence counter (UWCC) for plutonium measurements in mixed oxide fuels

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Eccleston, G.W.; Menlove, H.O.; Abhold, M.

    1998-12-31

    The use of fresh uranium-plutonium mixed oxide (MOX) fuel in light-water reactors (LWR) is increasing in Europe and Japan and it is necessary to verify the plutonium content in the fuel for international safeguards purposes. The UWCC is a new instrument that has been designed to operate underwater and nondestructively measure the plutonium in unirradiated MOX fuel assemblies. The UWCC can be quickly configured to measure either boiling-water reactor (BWR) or pressurized-water reactor (PWR) fuel assemblies. The plutonium loading per unit length is measured using the UWCC to precisions of less than 1% in a measurement time of 2 tomore » 3 minutes. Initial calibrations of the UWCC were completed on measurements of MOX fuel in Mol, Belgium. The MCNP-REN Monte Carlo simulation code is being benchmarked to the calibration measurements to allow accurate simulations for extended calibrations of the UWCC.« less

  12. Method of using infrared radiation for assembling a first component with a second component

    DOEpatents

    Sikka, Vinod K.; Whitson, Barry G.; Blue, Craig A.

    1999-01-01

    A method of assembling a first component for assembly with a second component involves a heating device which includes an enclosure having a cavity for inserting a first component. An array of infrared energy generators is disposed within the enclosure. At least a portion of the first component is inserted into the cavity, exposed to infrared energy and thereby heated to a temperature wherein the portion of the first component is sufficiently softened and/or expanded for assembly with a second component.

  13. Measurement of the 235U Induced Fission Gamma-ray Spectrum as an Active Non-destructive Assay of Fresh Nucleear Fuel

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Sarnoski, Sarah E.; Fast, James E.; Fulsom, Bryan G.

    2017-07-17

    Non-destructive assay is a powerful tool the International Atomic Energy Agency (IAEA) employs to verify adherence to safeguards agreements. Current IAEA veri- cation techniques for fresh nuclear fuel include passive gamma-ray spectroscopy to determine fuel enrichment. This technique suers from self-shielding and lakes the percision to detect diversion of central fuel rods. The aim of this research is to develop a new, more capable non-destructive analysis technique using active neutron interroga- tion of fuel assemblies and determining the yields of short-lived ssion products from high-resolution gamma-ray spectroscopy using high-purity germanium (HPGe). This paper reports results from irradiation of a onemore » meter tall mock fresh fuel assembly with low enriched uranium (LEU) or depleted uranium (DU) rods using a down-scattered deuterium-tritium (D-T) neutron source. Both prompt and delayed gamma-ray spec- tra were collected as time-stamped list-mode data in a coax detector and without list mode data in a planar strip detector. No dierentiating signatures were observed in the prompt spectra in either detector; however, both detectors observed several short-lived ssion product signatures in LEU and not DU fuel, indicating that this technique has potential for determination of enrichment of fresh fuel assemblies. There were eight unique ssion products observed in the LEU spectra with the coax detector spectra, and three ssion products were observed in the LEU spectra with the strip detector.« less

  14. Downhole steam generator with improved preheating/cooling features

    DOEpatents

    Donaldson, A. Burl; Hoke, Donald E.; Mulac, Anthony J.

    1983-01-01

    An apparatus for downhole steam generation employing dual-stage preheaters for liquid fuel and for the water. A first heat exchange jacket for the fuel surrounds the fuel/oxidant mixing section of the combustor assembly downstream of the fuel nozzle and contacts the top of the combustor unit of the combustor assembly, thereby receiving heat directly from the combustion of the fuel/oxidant. A second stage heat exchange jacket surrounds an upper portion of the oxidant supply line adjacent the fuel nozzle receiving further heat from the compression heat which results from pressurization of the oxidant. The combustor unit includes an inner combustor sleeve whose inner wall defines the combustion zone. The inner combustor sleeve is surrounded by two concentric water channels, one defined by the space between the inner combustor sleeve and an intermediate sleeve, and the second defined by the space between the intermediate sleeve and an outer cylindrical housing. The channels are connected by an annular passage adjacent the top of the combustor assembly and the countercurrent nature of the water flow provides efficient cooling of the inner combustor sleeve. An annular water ejector with a plurality of nozzles is provided to direct water downwardly into the combustor unit at the boundary of the combustion zone and along the lower section of the intermediate sleeve.

  15. Downhole steam generator with improved preheating/cooling features. [Patent application

    DOEpatents

    Donaldson, A.B.; Hoke, D.E.; Mulac, A.J.

    1980-10-10

    An apparatus is described for downhole steam generation employing dual-stage preheaters for liquid fuel and for the water. A first heat exchange jacket for the fuel surrounds the fuel/oxidant mixing section of the combustor assembly downstream of the fuel nozzle and contacts the top of the combustor unit of the combustor assembly, thereby receiving heat directly from the combustion of the fuel/oxidant. A second stage heat exchange jacket surrounds an upper portion of the oxidant supply line adjacent the fuel nozzle receiving further heat from the compression heat which results from pressurization of the oxidant. The combustor unit includes an inner combustor sleeve whose inner wall defines the combustion zone. The inner combustor sleeve is surrounded by two concentric water channels, one defined by the space between the inner combustor sleeve and an intermediate sleeve, and the second defined by the space between the intermediate sleeve and an outer cylindrical housing. The channels are connected by an annular passage adjacent the top of the combustor assembly and the countercurrent nature of the water flow provides efficient cooling of the inner combustor sleeve. An annular water ejector with a plurality of nozzles is provided to direct water downwardly into the combustor unit at the boundary of the combustion zone and along the lower section of the intermediate sleeve.

  16. High energy X-ray CT study on the central void formations and the fuel pin deformations of FBR fuel assemblies

    NASA Astrophysics Data System (ADS)

    Katsuyama, Kozo; Nagamine, Tsuyoshi; Matsumoto, Shin-ichiro; Sato, Seichi

    2007-02-01

    The central void formations and deformations of fuel pins were investigated in fuel assemblies irradiated to high burn-up, using a non-destructive X-ray CT (computer tomography) technique. In this X-ray CT, the effect of strong gamma ray activity could be reduced to a negligible degree by using the pulse of a high energy X-ray source and detecting the intensity of the transmitted X-rays in synchronization with the generated X-rays. Clear cross-sectional images of fuel assemblies irradiated to high burn-up in a fast breeder reactor were successively obtained, in which the wrapping wires, cladding, pellets and central voids could be distinctly seen. The diameter of a typical central void measured by X-ray CT agreed with the one obtained by ceramography within an error of 0.1 mm. Based on this result, the dependence of the central void diameter on the linear heating rate was analyzed. In addition, the deformation behavior of a fuel pin along its axial direction could be analyzed from 20 stepwise X-ray cross-sectional images obtained in a small interval, and the results obtained showed a good agreement with the predictions calculated by two computer codes.

  17. Fuel cell with metal screen flow-field

    DOEpatents

    Wilson, M.S.; Zawodzinski, C.

    1998-08-25

    A polymer electrolyte membrane (PEM) fuel cell is provided with electrodes supplied with a reactant on each side of a catalyzed membrane assembly (CMA). The fuel cell includes a metal mesh defining a rectangular flow-field pattern having an inlet at a first corner and an outlet at a second corner located on a diagonal from the first corner, wherein all flow paths from the inlet to the outlet through the square flow field pattern are equivalent to uniformly distribute the reactant over the CMA. In a preferred form of metal mesh, a square weave screen forms the flow-field pattern. In a particular characterization of the present invention, a bipolar plate electrically connects adjacent fuel cells, where the bipolar plate includes a thin metal foil having an anode side and a cathode side; a first metal mesh on the anode side of the thin metal foil; and a second metal mesh on the cathode side of the thin metal foil. In another characterization of the present invention, a cooling plate assembly cools adjacent fuel cells, where the cooling plate assembly includes an anode electrode and a cathode electrode formed of thin conducting foils; and a metal mesh flow field there between for distributing cooling water flow over the electrodes to remove heat generated by the fuel cells. 11 figs.

  18. Fuel cell with metal screen flow-field

    DOEpatents

    Wilson, Mahlon S.; Zawodzinski, Christine

    2001-01-01

    A polymer electrolyte membrane (PEM) fuel cell is provided with electrodes supplied with a reactant on each side of a catalyzed membrane assembly (CMA). The fuel cell includes a metal mesh defining a rectangular flow-field pattern having an inlet at a first corner and an outlet at a second corner located on a diagonal from the first corner, wherein all flow paths from the inlet to the outlet through the square flow field pattern are equivalent to uniformly distribute the reactant over the CMA. In a preferred form of metal mesh, a square weave screen forms the flow-field pattern. In a particular characterization of the present invention, a bipolar plate electrically connects adjacent fuel cells, where the bipolar plate includes a thin metal foil having an anode side and a cathode side; a first metal mesh on the anode side of the thin metal foil; and a second metal mesh on the cathode side of the thin metal foil. In another characterization of the present invention, a cooling plate assembly cools adjacent fuel cells, where the cooling plate assembly includes an anode electrode and a cathode electrode formed of thin conducting foils; and a metal mesh flow field therebetween for distributing cooling water flow over the electrodes to remove heat generated by the fuel cells.

  19. Fuel cell with metal screen flow-field

    DOEpatents

    Wilson, Mahlon S.; Zawodzinski, Christine

    1998-01-01

    A polymer electrolyte membrane (PEM) fuel cell is provided with electrodes supplied with a reactant on each side of a catalyzed membrane assembly (CMA). The fuel cell includes a metal mesh defining a rectangular flow-field pattern having an inlet at a first corner and an outlet at a second corner located on a diagonal from the first corner, wherein all flow paths from the inlet to the outlet through the square flow field pattern are equivalent to uniformly distribute the reactant over the CMA. In a preferred form of metal mesh, a square weave screen forms the flow-field pattern. In a particular characterization of the present invention, a bipolar plate electrically connects adjacent fuel cells, where the bipolar plate includes a thin metal foil having an anode side and a cathode side; a first metal mesh on the anode side of the thin metal foil; and a second metal mesh on the cathode side of the thin metal foil. In another characterization of the present invention, a cooling plate assembly cools adjacent fuel cells, where the cooling plate assembly includes an anode electrode and a cathode electrode formed of thin conducting foils; and a metal mesh flow field therebetween for distributing cooling water flow over the electrodes to remove heat generated by the fuel cells.

  20. Nuclear reactor

    DOEpatents

    Pennell, William E.; Rowan, William J.

    1977-01-01

    A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

  1. Dismantlement of the TSF-SNAP Reactor Assembly

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Peretz, Fred J

    2009-01-01

    This paper describes the dismantlement of the Tower Shielding Facility (TSF)?Systems for Nuclear Auxiliary Power (SNAP) reactor, a SNAP-10A reactor used to validate radiation source terms and shield performance models at Oak Ridge National Laboratory (ORNL) from 1967 through 1973. After shutdown, it was placed in storage at the Y-12 National Security Complex (Y-12), eventually falling under the auspices of the Highly Enriched Uranium (HEU) Disposition Program. To facilitate downblending of the HEU present in the fuel elements, the TSF-SNAP was moved to ORNL on June 24, 2006. The reactor assembly was removed from its packaging, inspected, and the sodium-potassiummore » (NaK) coolant was drained. A superheated steam process was used to chemically react the residual NaK inside the reactor assembly. The heat exchanger assembly was removed from the top of the reactor vessel, and the criticality safety sleeve was exchanged for a new safety sleeve that allowed for the removal of the vessel lid. A chain-mounted tubing cutter was used to separate the lid from the vessel, and the 36 fuel elements were removed and packaged in four U.S. Department of Transportation 2R/6M containers. The fuel elements were returned to Y-12 on July 13, 2006. The return of the fuel elements and disposal of all other reactor materials accomplished the formal objectives of the dismantlement project. In addition, a project model was established for the handling of a fully fueled liquid-metal?cooled reactor assembly. Current criticality safety codes have been benchmarked against experiments performed by Atomics International in the 1950s and 1960s. Execution of this project provides valuable experience applicable to future projects addressing space and liquid-metal-cooled reactors.« less

  2. Generator module architecture for a large solid oxide fuel cell power plant

    DOEpatents

    Gillett, James E.; Zafred, Paolo R.; Riggle, Matthew W.; Litzinger, Kevin P.

    2013-06-11

    A solid oxide fuel cell module contains a plurality of integral bundle assemblies, the module containing a top portion with an inlet fuel plenum and a bottom portion receiving air inlet feed and containing a base support, the base supports dense, ceramic exhaust manifolds which are below and connect to air feed tubes located in a recuperator zone, the air feed tubes passing into the center of inverted, tubular, elongated, hollow electrically connected solid oxide fuel cells having an open end above a combustion zone into which the air feed tubes pass and a closed end near the inlet fuel plenum, where the fuel cells comprise a fuel cell stack bundle all surrounded within an outer module enclosure having top power leads to provide electrical output from the stack bundle, where the fuel cells operate in the fuel cell mode and where the base support and bottom ceramic air exhaust manifolds carry from 85% to all 100% of the weight of the stack, and each bundle assembly has its own control for vertical and horizontal thermal expansion control.

  3. Fuel cell assembly unit for promoting fluid service and electrical conductivity

    DOEpatents

    Jones, Daniel O.

    1999-01-01

    Fluid service and/or electrical conductivity for a fuel cell assembly is promoted. Open-faced flow channel(s) are formed in a flow field plate face, and extend in the flow field plate face between entry and exit fluid manifolds. A resilient gas diffusion layer is located between the flow field plate face and a membrane electrode assembly, fluidly serviced with the open-faced flow channel(s). The resilient gas diffusion layer is restrained against entering the open-faced flow channel(s) under a compressive force applied to the fuel cell assembly. In particular, a first side of a support member abuts the flow field plate face, and a second side of the support member abuts the resilient gas diffusion layer. The support member is formed with a plurality of openings extending between the first and second sides of the support member. In addition, a clamping pressure is maintained for an interface between the resilient gas diffusion layer and a portion of the membrane electrode assembly. Preferably, the support member is spikeless and/or substantially flat. Further, the support member is formed with an electrical path for conducting current between the resilient gas diffusion layer and position(s) on the flow field plate face.

  4. CASMO5/TSUNAMI-3D spent nuclear fuel reactivity uncertainty analysis

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Ferrer, R.; Rhodes, J.; Smith, K.

    2012-07-01

    The CASMO5 lattice physics code is used in conjunction with the TSUNAMI-3D sequence in ORNL's SCALE 6 code system to estimate the uncertainties in hot-to-cold reactivity changes due to cross-section uncertainty for PWR assemblies at various burnup points. The goal of the analysis is to establish the multiplication factor uncertainty similarity between various fuel assemblies at different conditions in a quantifiable manner and to obtain a bound on the hot-to-cold reactivity uncertainty over the various assembly types and burnup attributed to fundamental cross-section data uncertainty. (authors)

  5. Method for recovering catalytic elements from fuel cell membrane electrode assemblies

    DOEpatents

    Shore, Lawrence [Edison, NJ; Matlin, Ramail [Berkeley Heights, NJ; Heinz, Robert [Ludwigshafen, DE

    2012-06-26

    A method for recovering catalytic elements from a fuel cell membrane electrode assembly is provided. The method includes converting the membrane electrode assembly into a particulate material, wetting the particulate material, forming a slurry comprising the wetted particulate material and an acid leachate adapted to dissolve at least one of the catalytic elements into a soluble catalytic element salt, separating the slurry into a depleted particulate material and a supernatant containing the catalytic element salt, and washing the depleted particulate material to remove any catalytic element salt retained within pores in the depleted particulate material.

  6. Fail-safe storage rack for irradiated fuel rod assemblies

    DOEpatents

    Lewis, D.R.

    1993-03-23

    A fail-safe storage rack is provided for interim storage of spent but radioactive nuclear fuel rod assemblies. The rack consists of a checkerboard array of substantially square, elongate receiving tubes fully enclosed by a double walled container, the outer wall of which is imperforate for liquid containment and the inner wall of which is provided with perforations for admitting moderator liquid flow to the elongate receiving tubes, the liquid serving to take up waste heat from the stored nuclear assemblies and dissipate same to the ambient liquid reservoir. A perforated cover sealing the rack facilitates cooling liquid entry and dissipation.

  7. Fail-safe storage rack for irradiated fuel rod assemblies

    DOEpatents

    Lewis, Donald R.

    1993-01-01

    A fail-safe storage rack is provided for interim storage of spent but radioactive nuclear fuel rod assemblies. The rack consists of a checkerboard array of substantially square, elongate receiving tubes fully enclosed by a double walled container, the outer wall of which is imperforate for liquid containment and the inner wall of which is provided with perforations for admitting moderator liquid flow to the elongate receiving tubes, the liquid serving to take up waste heat from the stored nuclear assemblies and dissipate same to the ambient liquid reservoir. A perforated cover sealing the rack facilitates cooling liquid entry and dissipation.

  8. Management of thermal peaking factors in CONFU-B PWR assemblies using neutron poisons and tailored enrichment

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Visosky, M.; Hejzlar, P.; Kazimi, M.

    2006-07-01

    CONFU-B assemblies are PWR assemblies containing standard Uranium fuel rods and TRU bearing inert material fuel rods and are designed to achieve net TRU destruction over a 4.5-year irradiation. These highly heterogeneous assemblies tend to exhibit large intra-assembly power peaking factors (IAPPF). Neutronic strategies to reduce IAPPF are developed. The IAPPF are calculated at the assembly level using CASMO4, and these are used to calculate the most restrictive thermal margin (the Minimum Departure from Nucleate Boiling Ratio, MDNBR) using a whole-core VIPRE-01 model. This paper examines two strategies to manage the thermal margin of a CONFU-B assembly while retaining themore » TRU destruction performance: use of neutron poisons and tailored enrichment schemes. Burnable poisons can be used to suppress BOL reactivity of fresh CONFU-B assemblies with only minor impact on MDNBR and TRU destruction performance. Tailored enrichment, along with the use of soluble boron, can achieve significant improvements in MDNBR, but at some cost to TRU destruction performance. (authors)« less

  9. Optical fuel pin scanner. [Patent application; for reading identifications

    DOEpatents

    Kirchner, T.L.; Powers, H.G.

    1980-12-09

    This patent relates to an optical identification system developed for post-irradiation disassembly and analysis of fuel bundle assemblies. The apparatus is designed to be lowered onto a stationary fuel pin to read identification numbers or letters imprinted on the circumference of the top fuel pin and cap. (DLC)

  10. Method and apparatus for measuring reactivity of fissile material

    DOEpatents

    Lee, D.M.; Lindquist, L.O.

    1982-09-07

    Given are a method and apparatus for measuring nondestructively and noninvasively (i.e., using no internal probing) the burnup, reactivity, or fissile content of any material which emits neutrons and which has fissionable components. The assay is accomplished by altering the return flux of neutrons into the fuel assembly by means of changing the reflecting material. The existing passive neutron emissions in the material being assayed are used as the source of interrogating neutrons. Two measurements of either emitted neutron or emitted gamma-ray count rates are made and are then correlated to either reactivity, burnup, or fissionable content of the material being assayed, thus providing a measurement of either reactivity, burnup, or fissionable content of the material being assayed. Spent fuel which has been freshly discharged from a reactor can be assayed using this method and apparatus. Precisions of 1000 MWd/tU appear to be feasible.

  11. The Development of Expansion Plug Wedge Test for Clad Tubing Structure Mechanical Property Evaluation

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Wang, Jy-An John; Jiang, Hao

    2016-01-12

    To determine the tensile properties of irradiated fuel cladding in a hot cell, a simple test was developed at the Oak Ridge National Laboratory (ORNL) and is described fully in US Patent Application 20060070455, “Expanded plug method for developing circumferential mechanical properties of tubular materials.” This method is designed for testing fuel rod cladding ductility in a hot cell using an expandable plug to stretch a small ring of irradiated cladding material. The specimen strain is determined using the measured diametrical expansion of the ring. This method removes many complexities associated with specimen preparation and testing. The advantages are themore » simplicity of measuring the test component assembly in the hot cell and the direct measurement of the specimen’s strain. It was also found that cladding strength could be determined from the test results.« less

  12. Nuclear fuel element nut retainer cup. [PWR

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Walton, L.A.

    1977-07-19

    A typical embodiment has an end fitting for a nuclear reactor fuel element that is joined to the control rod guide tubes by means of a nut plate assembly. The nut plate assembly has an array of nuts, each engaging the respective threaded end of the control rod guide tubes. The nuts, moreover, are retained on the plate during handling and before fuel element assembly by means of hollow cylindrical locking cups that are brazed to the plate and loosely circumscribe the individual enclosed nuts. After the nuts are threaded onto the respective guide tube ends, the locking cups aremore » partially deformed to prevent one or more of the nuts from working loose during reactor operation. The locking cups also prevent loose or broken end fitting parts from becoming entrained in the reactor coolant.« less

  13. Systems and methods for preventing flashback in a combustor assembly

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Johnson, Thomas Edward; Ziminsky, Willy Steve; Stevenson, Christian Xavier

    2016-04-05

    Embodiments of the present application include a combustor assembly. The combustor assembly may include a combustion chamber, a first plenum, a second plenum, and one or more elongate air/fuel premixing injection tubes. Each of the elongate air/fuel premixing injection tubes may include a first length at least partially disposed within the first plenum and configured to receive a first fluid from the first plenum. Moreover, each of the elongate air/fuel premixing injection tubes may include a second length disposed downstream of the first length and at least partially disposed within the second plenum. The second length may be formed ofmore » a porous wall configured to allow a second fluid from the second plenum to enter the second length and create a boundary layer about the porous wall.« less

  14. Membrane electrode gasket assembly (MEGA) technology for polymer electrolyte fuel cells

    NASA Astrophysics Data System (ADS)

    Pozio, A.; Giorgi, L.; De Francesco, M.; Silva, R. F.; Lo Presti, R.; Danzi, A.

    A new technology for the production of a membrane electrode gasket assembly (MEGA) for polymer electrolyte fuel cells (PEFCs) is defined. The MEGA system was prepared by sealing a previously prepared membrane electrode assembly (MEA) in a moulded gasket. For this aim, a proprietary silicone based liquid mixture was injected directly into the MEA borders. Gaskets obtained in different shapes and hardness grades are stable in a wide temperature range. The MEGA technology shows several advantages with respect to traditional PEFCs stack assembling systems: effective membrane saving, reduced fabrication time, possibility of quality control and failed elements substitution. This technology was successfully tested at the ENEA laboratories and the results were acquired in laboratory scale, but industrial production appears to be simple and cheap.

  15. Integral Fast Reactor fuel pin processor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Levinskas, D.

    1993-01-01

    This report discusses the pin processor which receives metal alloy pins cast from recycled Integral Fast Reactor (IFR) fuel and prepares them for assembly into new IFR fuel elements. Either full length as-cast or precut pins are fed to the machine from a magazine, cut if necessary, and measured for length, weight, diameter and deviation from straightness. Accepted pins are loaded into cladding jackets located in a magazine, while rejects and cutting scraps are separated into trays. The magazines, trays, and the individual modules that perform the different machine functions are assembled and removed using remote manipulators and master-slaves.

  16. Integral Fast Reactor fuel pin processor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Levinskas, D.

    1993-03-01

    This report discusses the pin processor which receives metal alloy pins cast from recycled Integral Fast Reactor (IFR) fuel and prepares them for assembly into new IFR fuel elements. Either full length as-cast or precut pins are fed to the machine from a magazine, cut if necessary, and measured for length, weight, diameter and deviation from straightness. Accepted pins are loaded into cladding jackets located in a magazine, while rejects and cutting scraps are separated into trays. The magazines, trays, and the individual modules that perform the different machine functions are assembled and removed using remote manipulators and master-slaves.

  17. Thermal analysis of the FSP-1 fuel pin irradiation test. [for SP-100 space power reactor

    NASA Technical Reports Server (NTRS)

    Lyon, William F., III

    1991-01-01

    Thermal analysis of a pin from the FSP-1 fuels irradiation test has been completed. The purpose of the analysis was to provide predictions of fuel pin temperatures, determine the flow regime within the lithium annulus of the test assembly, and provide a standardized model for a consistent basis of comparison between pins within the test assembly. The calculations have predicted that the pin is operating at slightly above the test design temperatures and that the flow regime within the lithium annulus is a laminar buoyancy driven flow.

  18. In situ observations of water production and distribution in an operating H2/O2 PEM fuel cell assembly using 1H NMR microscopy.

    PubMed

    Feindel, Kirk W; LaRocque, Logan P-A; Starke, Dieter; Bergens, Steven H; Wasylishen, Roderick E

    2004-09-22

    Proton NMR imaging was used to investigate in situ the distribution of water in a polymer electrolyte membrane fuel cell operating on H2 and O2. In a single experiment, water was monitored in the gas flow channels, the membrane electrode assembly, and in the membrane surrounding the catalysts. Radial gradient diffusion removes water from the catalysts into the surrounding membrane. This research demonstrates the strength of 1H NMR microscopy as an aid for designing fuel cells to optimize water management.

  19. Critical experiments at Sandia National Laboratories : technical meeting on low-power critical facilities and small reactors.

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Harms, Gary A.; Ford, John T.; Barber, Allison Delo

    2010-11-01

    Sandia National Laboratories (SNL) has conducted radiation effects testing for the Department of Energy (DOE) and other contractors supporting the DOE since the 1960's. Over this period, the research reactor facilities at Sandia have had a primary mission to provide appropriate nuclear radiation environments for radiation testing and qualification of electronic components and other devices. The current generation of reactors includes the Annular Core Research Reactor (ACRR), a water-moderated pool-type reactor, fueled by elements constructed from UO2-BeO ceramic fuel pellets, and the Sandia Pulse Reactor III (SPR-III), a bare metal fast burst reactor utilizing a uranium-molybdenum alloy fuel. The SPR-IIImore » is currently defueled. The SPR Facility (SPRF) has hosted a series of critical experiments. A purpose-built critical experiment was first operated at the SPRF in the late 1980's. This experiment, called the Space Nuclear Thermal Propulsion Critical Experiment (CX), was designed to explore the reactor physics of a nuclear thermal rocket motor. This experiment was fueled with highly-enriched uranium carbide fuel in annular water-moderated fuel elements. The experiment program was completed and the fuel for the experiment was moved off-site. A second critical experiment, the Burnup Credit Critical Experiment (BUCCX) was operated at Sandia in 2002. The critical assembly for this experiment was based on the assembly used in the CX modified to accommodate low-enriched pin-type fuel in water moderator. This experiment was designed as a platform in which the reactivity effects of specific fission product poisons could be measured. Experiments were carried out on rhodium, an important fission product poison. The fuel and assembly hardware for the BUCCX remains at Sandia and is available for future experimentation. The critical experiment currently in operation at the SPRF is the Seven Percent Critical Experiment (7uPCX). This experiment is designed to provide benchmark reactor physics data to support validation of the reactor physics codes used to design commercial reactor fuel elements in an enrichment range above the current 5% enrichment cap. A first set of critical experiments in the 7uPCX has been completed. More experiments are planned in the 7uPCX series. The critical experiments at Sandia National Laboratories are currently funded by the US Department of Energy Nuclear Criticality Safety Program (NCSP). The NCSP has committed to maintain the critical experiment capability at Sandia and to support the development of a critical experiments training course at the facility. The training course is intended to provide hands-on experiment experience for the training of new and re-training of practicing Nuclear Criticality Safety Engineers. The current plans are for the development of the course to continue through the first part of fiscal year 2011 with the development culminating is the delivery of a prototype of the course in the latter part of the fiscal year. The course will be available in fiscal year 2012.« less

  20. DOE Office of Scientific and Technical Information (OSTI.GOV)

    Hanan, N. A.; Matos, J. E.

    At The request of the Czech Technical University in Prague, ANL has performed independent verification calculations using the MCNP Monte Carlo code for three core configurations of the VR-1 reactor: a current core configuration B1 with HEU (36%) IRT-3M fuel assemblies and planned core configurations C1 and C2 with LEU (19.7%) IRT-4M fuel assemblies. Details of these configurations were provided to ANL by CTU. For core configuration B1, criticality calculations were performed for two sets of control rod positions provided to ANL by CTU. For core configurations C1 and C2, criticality calculations were done for cases with all control rodsmore » at the top positions, all control rods at the bottom positions, and two critical states of the reactor for different control rod positions. In addition, sensitivity studies for variation of the {sup 235}U mass in each fuel assembly and variation of the fuel meat and cladding thicknesses in each of the fuel tubes were done for the C1 core configuration. Finally the reactivity worth of the individual control rods was calculated for the B1, C1, and C2 core configurations.« less

  1. Dry compliant seal for phosphoric acid fuel cell

    DOEpatents

    Granata, Jr., Samuel J.; Woodle, Boyd M.

    1990-01-01

    A dry compliant overlapping seal for a phosphoric acid fuel cell preformed f non-compliant Teflon to make an anode seal frame that encircles an anode assembly, a cathode seal frame that encircles a cathode assembly and a compliant seal frame made of expanded Teflon, generally encircling a matrix assembly. Each frame has a thickness selected to accommodate various tolerances of the fuel cell elements and are either bonded to one of the other frames or to a bipolar or end plate. One of the non-compliant frames is wider than the other frames forming an overlap of the matrix over the wider seal frame, which cooperates with electrolyte permeating the matrix to form a wet seal within the fuel cell that prevents process gases from intermixing at the periphery of the fuel cell and a dry seal surrounding the cell to keep electrolyte from the periphery thereof. The frames may be made in one piece, in L-shaped portions or in strips and have an outer perimeter which registers with the outer perimeter of bipolar or end plates to form surfaces upon which flanges of pan shaped, gas manifolds can be sealed.

  2. Rail-Cask Tests: Normal-Conditionsof- Transport Tests of Surrogate PWR Fuel Assemblies in an ENSA ENUN 32P Cask.

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    McConnell, Paul E.; Ross, Steven; Grey, Carissa Ann

    This report describes tests conducted using a full-size rail cask, the ENSA ENUN 32P, involving handling of the cask and transport of the cask via truck, ships, and rail. The purpose of the tests was to measure strains and accelerations on surrogate pressurized water reactor fuel rods when the fuel assemblies were subjected to Normal Conditions of Transport within the rail cask. In addition, accelerations were measured on the transport platform, the cask cradle, the cask, and the basket within the cask holding the assemblies. These tests were an international collaboration that included Equipos Nucleares S.A., Sandia National Laboratories, Pacificmore » Northwest National Laboratory, Coordinadora Internacional de Cargas S.A., the Transportation Technology Center, Inc., the Korea Radioactive Waste Agency, and the Korea Atomic Energy Research Institute. All test results in this report are PRELIMINARY – complete analyses of test data will be completed and reported in FY18. However, preliminarily: The strains were exceedingly low on the surrogate fuel rods during the rail-cask tests for all the transport and handling modes. The test results provide a compelling technical basis for the safe transport of spent fuel.« less

  3. Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

    NASA Technical Reports Server (NTRS)

    Clement, J. D.; Rust, J. H.

    1977-01-01

    Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

  4. A critical review on gas diffusion micro and macroporous layers degradations for improved membrane fuel cell durability

    NASA Astrophysics Data System (ADS)

    Lapicque, Francois; Belhadj, Mariem; Bonnet, Caroline; Pauchet, Joël; Thomas, Yohann

    2016-12-01

    Formerly considered as a secondary component of fuel cell, gas diffusion layers (GDLs) have been shown to have a key role in gas transport to the catalyst layers and in water management: in particular, the microporous layer (MPL) deposited on the diffusion substrate has an active part in water distribution in the membrane electrode assembly and in its efficient removal from the cell. In addition to its perfect design for the targeted application and in combination with the macroporous substrate (MPS), the MPL structure and physicochemical properties have to contribute to the cell durability, which is still considered as insufficient for larger, massive commercialisation of this energy conversion system. The paper is aimed at reviewing the main knowledge gained on the role of the MPL on GDL operation and durability, with investigation of degradation phenomena of both MPL and MPS, together with the role played by the MPL in mitigating the occurrence of degradation phenomena that can occur in the whole fuel cell. In addition to the reviewing purpose, original data on ex-situ degradation of GDL are presented.

  5. Preliminary Thermal Modeling of HI-Storm 100S-218 Version B Storage Modules at Hope Creek Cuclear Power Station ISFSI

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Cuta, Judith M.; Adkins, Harold E.

    2013-08-30

    As part of the Used Fuel Disposition Campaign of the U. S. Department of Energy, Office of Nuclear Energy (DOE-NE) Fuel Cycle Research and Development, a consortium of national laboratories and industry is performing visual inspections and temperature measurements of selected storage modules at various locations around the United States. This report documents thermal analyses in in support of the inspections at the Hope Creek Nuclear Generating Station ISFSI. This site utilizes the HI-STORM100 vertical storage system developed by Holtec International. This is a vertical storage module design, and the thermal models are being developed using COBRA-SFS (Michener, et al.,more » 1987), a code developed by PNNL for thermal-hydraulic analyses of multi assembly spent fuel storage and transportation systems. This report describes the COBRA-SFS model in detail, and presents pre-inspection predictions of component temperatures and temperature distributions. The final report will include evaluation of inspection results, and if required, additional post-test calculations, with appropriate discussion of results.« less

  6. One-pot synthesis of transition metal ion-chelating ordered mesoporous carbon/carbon nanotube composites for active and durable fuel cell catalysts

    NASA Astrophysics Data System (ADS)

    Dombrovskis, Johanna K.; Palmqvist, Anders E. C.

    2017-07-01

    Development of non-precious metal catalysts for the oxygen reduction reaction (ORR) in proton exchange membrane (PEM) fuel cells with high activity and durability and with optimal water management properties is of outmost technological importance and highly challenging. Here we study the possibilities offered through judicious selection of small molecular precursors used for the formation of ordered mesoporous carbon-based non-precious metal ORR catalysts. By combining two complementary precursors, we present a one-pot synthesis that leads to a composite material consisting of transition metal ion-chelating ordered mesoporous carbon and multi-walled carbon nanotubes (TM-OMC/CNT). The resulting composite materials show high specific surface areas and a carbon structure that exhibits graphitic signatures. The synthesis procedure allows for tuning of the carbon structure, the surface area, the pore volume and the ratio of the two components of the composite. The TM-OMC/CNT composites were processed into membrane electrode assemblies and evaluated in single cell fuel cell measurements where they showed a combination of good ORR activity and very high durability.

  7. IMPROVED TYPE OF FUEL ELEMENT

    DOEpatents

    Monson, H.O.

    1961-01-24

    A radiator-type fuel block assembly is described. It has a hexagonal body of neutron fissionable material having a plurality of longitudinal equal- spaced coolant channels therein aligned in rows parallel to each face of the hexagonal body. Each of these coolant channels is hexagonally shaped with the corners rounded and enlarged and the assembly has a maximum temperature isothermal line around each channel which is approximately straight and equidistant between adjacent channels.

  8. Horizontal modular dry irradiated fuel storage system

    DOEpatents

    Fischer, Larry E.; McInnes, Ian D.; Massey, John V.

    1988-01-01

    A horizontal, modular, dry, irradiated fuel storage system (10) includes a thin-walled canister (12) for containing irradiated fuel assemblies (20), which canister (12) can be positioned in a transfer cask (14) and transported in a horizontal manner from a fuel storage pool (18), to an intermediate-term storage facility. The storage system (10) includes a plurality of dry storage modules (26) which accept the canister (12) from the transfer cask (14) and provide for appropriate shielding about the canister (12). Each module (26) also provides for air cooling of the canister (12) to remove the decay heat of the irradiated fuel assemblies (20). The modules (26) can be interlocked so that each module (26) gains additional shielding from the next adjacent module (26). Hydraulic rams (30) are provided for inserting and removing the canisters (12) from the modules (26).

  9. Development and Assessment of CFD Models Including a Supplemental Program Code for Analyzing Buoyancy-Driven Flows Through BWR Fuel Assemblies in SFP Complete LOCA Scenarios

    NASA Astrophysics Data System (ADS)

    Artnak, Edward Joseph, III

    This work seeks to illustrate the potential benefits afforded by implementing aspects of fluid dynamics, especially the latest computational fluid dynamics (CFD) modeling approach, through numerical experimentation and the traditional discipline of physical experimentation to improve the calibration of the severe reactor accident analysis code, MELCOR, in one of several spent fuel pool (SFP) complete loss-ofcoolant accident (LOCA) scenarios. While the scope of experimental work performed by Sandia National Laboratories (SNL) extends well beyond that which is reasonably addressed by our allotted resources and computational time in accordance with initial project allocations to complete the report, these simulated case trials produced a significant array of supplementary high-fidelity solutions and hydraulic flow-field data in support of SNL research objectives. Results contained herein show FLUENT CFD model representations of a 9x9 BWR fuel assembly in conditions corresponding to a complete loss-of-coolant accident scenario. In addition to the CFD model developments, a MATLAB based controlvolume model was constructed to independently assess the 9x9 BWR fuel assembly under similar accident scenarios. The data produced from this work show that FLUENT CFD models are capable of resolving complex flow fields within a BWR fuel assembly in the realm of buoyancy-induced mass flow rates and that characteristic hydraulic parameters from such CFD simulations (or physical experiments) are reasonably employed in corresponding constitutive correlations for developing simplified numerical models of comparable solution accuracy.

  10. DOE Office of Scientific and Technical Information (OSTI.GOV)

    Gottesfeld, S.

    The fuel cell is the most efficient device for the conversion of hydrogen fuel to electric power. As such, the fuel cell represents a key element in efforts to demonstrate and implement hydrogen fuel utilization for electric power generation. The low temperature, polymer electrolyte membrane fuel cell (PEMFC) has recently been identified as an attractive option for stationary power generation, based on the relatively simple and benign materials employed, the zero-emission character of the device, and the expected high power density, high reliability and low cost. However, a PEMFC stack fueled by hydrogen with the combined properties of low cost,more » high performance and high reliability has not yet been demonstrated. Demonstration of such a stack will remove a significant barrier to implementation of this advanced technology for electric power generation from hydrogen. Work done in the past at LANL on the development of components and materials, particularly on advanced membrane/electrode assemblies (MEAs), has contributed significantly to the capability to demonstrate in the foreseeable future a PEMFC stack with the combined characteristics described above. A joint effort between LANL and an industrial stack manufacturer will result in the demonstration of such a fuel cell stack for stationary power generation. The stack could operate on hydrogen fuel derived from either natural gas or from renewable sources. The technical plan includes collaboration with a stack manufacturer (CRADA). It stresses the special requirements from a PEMFC in stationary power generation, particularly maximization of the energy conversion efficiency, extension of useful life to the 10 hours time scale and tolerance to impurities from the reforming of natural gas.« less

  11. Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek

    At the Atucha-I pressurized heavy water reactor in Argentina, fuel assemblies in the spent fuel pools are stored by suspending them in two vertically stacked layers. This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Since much of the fuel is very old, Cerenkov viewing devices are often not very useful even for the top layer. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 w% {sup 235}U, and has been in operation since 1974, a wide range of burnups and cooling times canmore » exist in any given pool. A spent fuel neutron counting tool consisting of a fission chamber, SFNC, has been used at the site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups to levels up 11,000 MWd/t, the existing signal processing software of the tool was found to fail due to non-linearity of the source term with burnup. A new Graphical User Interface software package based on the LabVIEW platform was developed to predict expected neutron signals covering all ranges of burnups and cooling times and establish maps of expected signals at various pool locations. The algorithm employed in the software uses a set of transfer functions in a 47-energy group structure which are coupled with a 47-energy group neutron source spectrum based on various cooling times and burnups for each of the two enrichment levels. The database of the software consists of these transfer functions for the three different inter-assembly pitches that the fuel is stored in at the site. The transfer functions were developed for a 6 by 6 matrix of fuel assemblies with the detector placed at the center surrounded by four near neighbors, eight next nearest neighbors and so on for the 36 assemblies. These calculations were performed using Monte Carlo radiation transport methods. The basic methodology consisted of starting sources in each of the assemblies and tallying the contribution to the detector by a single neutron in each of the 47 energy groups used. Thus for the single existing symmetric pitch in the pools, where the vertical and horizontal separations are equal, only 6 sets of transfer functions are required. For the two asymmetrical pitches, nine sets of transfer functions are stored. In addition, source spectra at burnups ranging from 4000 to 20000 MWd/t and cooling times up to 40 years are stored. These source terms were established based on CANDU 37-rod fuel that is very similar to the Atucha fuel. Linear interpolation is used by the software for both burnup and cooling time to establish source terms at any intermediate condition. Using the burnup, cooling time and initial enrichment of the surrounding assemblies a set of source strengths in the 47-group structure for each of the 36 assemblies is established and multiplied group-wise with the appropriate transfer function set. The grand total over the 47 groups for all 36 assemblies is the predicted signal at the detector. The software was initially calibrated against a set of typically 5-6 measurements chosen from among the measured data at each level of the six pools and calibration factors were established. The set used for calibration is chosen such that it is fairly representative of the range of spent fuel assembly characteristics present in each level. Once established, these calibration factors can be repeatedly used for verification purposes. Recalibration will be required if the hardware or pool configurations has changed. It will also be required if a long enough time has elapsed since they were established thus making a cooling time correction necessary. The objective of the inspection is to detect missing fuel from one or more nearest neighbors of the detector. During the verification mode of the software, the predicted and measured signals are compared and the inspector is alerted if the difference between the two signals is beyond a set tolerance limit. Based on the uncertainties associated with both the calculations and measurements, a lower limit of the tolerance will be 15% with an upper limit of 20%. For the most part a 20% tolerance limit will be able to detect a missing assembly since in the vast majority of cases the drop in signal due to a single missing nearest neighbor assembly will be in the range 24-27%. The software was benchmarked against an extensive set of measured data taken at the site in 2004. Overall, 326 data points were examined and the prediction of the calibrated software was compared to the measurements within a set tolerance of ±20%. Of these, 283 of the predicted signals representing 87% of the total matched the measured data within ±10%. A further 27 or 8% were in the range of ±10-15% and 8 or 2.5% were in the range of ±15-20%. Thus, 97.5% of the data matched the measurements within the set tolerance limit of 20%, with 95% matching measured data with the lowest allowed tolerance limit of ±15%. The remaining 2.5% had measured signals that were very different from those at locations with very similar surrounding assemblies and the cause of these discrepancies could not be ascertained from the measurement logs. In summary, 97.5% of the predictions matched the measurements within the set 20% tolerance limit providing proof of the robustness of the software. This software package linked to SFNC will be deployed at the site and will enhance the capability of gross defect verification for the whole range of burnup, cooling time and initial enrichments of the spent fuel being discharged into the various pools at the Atucha-I reactor site.« less

  12. Robotic Manufacturing of 5.5 Meter Cryogenic Fuel Tank Dome Assemblies for the NASA Ares I Rocket

    NASA Technical Reports Server (NTRS)

    Jones, Ronald E.

    2012-01-01

    The Ares I rocket is the first launch vehicle scheduled for manufacture under the National Aeronautic and Space Administration's (NASA's) Constellation program. A series of full-scale Ares I development articles have been constructed on the Robotic Weld Tool at the NASA George C. Marshall Space Flight Center in Huntsville, Alabama. The Robotic Weld Tool is a 100 ton, 7-axis, robotic manufacturing system capable of machining and friction stir welding large-scale space hardware. This presentation will focus on the friction stir welding of 5.5m diameter cryogenic fuel tank components; specifically, the liquid hydrogen forward dome (LH2 MDA), the common bulkhead manufacturing development articles (CBMDA) and the thermal protection system demonstration dome (TPS Dome). The LH2 MDA was the first full-scale, flight-like Ares I hardware produced under the Constellation Program. It is a 5.5m diameter elliptical dome assembly consisting of eight gore panels, a y-ring stiffener and a manhole fitting. All components are made from aluminumlithium alloy 2195. Conventional and self-reacting friction stir welding was used on this article. An overview of the manufacturing processes will be discussed. The LH2 MDA is the first known fully friction stir welded dome ever produced. The completion of four Common Bulkhead Manufacturing Development Articles (CBMDA) and the TPS Dome will also be highlighted. Each CBMDA and the TPS Dome consists of a 5.5m diameter spun-formed dome friction stir welded to a y-ring stiffener. The domes and y-rings are made of aluminum 2014 and 2219 respectively. The TPS Dome has an additional aluminum alloy 2195 barrel section welded to the y-ring. Manufacturing solutions will be discussed including "fixtureless" welding with self reacting friction stir welding.

  13. FUEL ROD ASSEMBLY

    DOEpatents

    Hutter, E.

    1959-09-01

    A cluster of nuclear fuel rods aod a tubular casing through which a coolant flows in heat-change contact with the ruel rods are described. The casting is of trefoil section and carries the fuel rods, each of which has two fin engaging the serrated fins of the other two fuel rods, whereby the fuel rods are held in the casing and are interlocked against relative longitudinal movement.

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Triplett, C.E.

    1996-12-01

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

  15. 77 FR 2658 - Airworthiness Directives; Bombardier, Inc. Airplanes

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-01-19

    ... high pressure fuel lines due to improper installation of an expandable pin on the lower cowl assembly... chafing of the high pressure fuel lines, which if not corrected, could cause fuel leakage in a fire zone... on a high pressure (HP) fuel line. The source of chafing was related to the improper installation of...

  16. Fuel cell integral bundle assembly including ceramic open end seal and vertical and horizontal thermal expansion control

    DOEpatents

    Zafred, Paolo R [Murrysville, PA; Gillett, James E [Greensburg, PA

    2012-04-24

    A plurality of integral bundle assemblies contain a top portion with an inlet fuel plenum and a bottom portion containing a base support, the base supports a dense, ceramic air exhaust manifold having four supporting legs, the manifold is below and connects to air feed tubes located in a recuperator zone, the air feed tubes passing into the center of inverted, tubular, elongated, hollow electrically connected solid oxide fuel cells having an open end above a combustion zone into which the air feed tubes pass and a closed end near the inlet fuel plenum, where the open end of the fuel cells rest upon and within a separate combination ceramic seal and bundle support contained in a ceramic support casting, where at least one flexible cushion ceramic band seal located between the recuperator and fuel cells protects and controls horizontal thermal expansion, and where the fuel cells operate in the fuel cell mode and where the base support and bottom ceramic air exhaust manifolds carry from 85% to all of the weight of the generator.

  17. Nuclear reactors built, being built, or planned, 1991

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Simpson, B.

    1992-07-01

    This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1991. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor ismore » an American company -- working either independently or in cooperation with a foreign company (Part 4, in each section). Critical assembly refers to an assembly of fuel and assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).« less

  18. Improving Energy Efficiency for the Vehicle Assembly Industry: A Discrete Event Simulation Approach

    NASA Astrophysics Data System (ADS)

    Oumer, Abduaziz; Mekbib Atnaw, Samson; Kie Cheng, Jack; Singh, Lakveer

    2016-11-01

    This paper presented a Discrete Event Simulation (DES) model for investigating and improving energy efficiency in vehicle assembly line. The car manufacturing industry is one of the highest energy consuming industries. Using Rockwell Arena DES package; a detailed model was constructed for an actual vehicle assembly plant. The sources of energy considered in this research are electricity and fuel; which are the two main types of energy sources used in a typical vehicle assembly plant. The model depicts the performance measurement for process- specific energy measures of painting, welding, and assembling processes. Sound energy efficiency model within this industry has two-fold advantage: reducing CO2 emission and cost reduction associated with fuel and electricity consumption. The paper starts with an overview of challenges in energy consumption within the facilities of automotive assembly line and highlights the parameters for energy efficiency. The results of the simulation model indicated improvements for energy saving objectives and reduced costs.

  19. Turbulent Burning Velocities of Two-Component Fuel Mixtures of Methane, Propane and Hydrogen

    NASA Astrophysics Data System (ADS)

    Kido, Hiroyuki; Nakahara, Masaya; Hashimoto, Jun; Barat, Dilmurat

    In order to clarify the turbulent burning velocity of multi-component fuel mixtures, both lean and rich two-component fuel mixtures, in which methane, propane and hydrogen were used as fuels, were prepared while maintaining the laminar burning velocity approximately constant. A distinct difference in the measured turbulent burning velocity at the same turbulence intensity is observed for two-component fuel mixtures having different addition rates of fuel, even the laminar burning velocities are approximately the same. The burning velocities of lean mixtures change almost constantly as the rate of addition changes, whereas the burning velocities of the rich mixtures show no such tendency. This trend can be explained qualitatively based on the mean local burning velocity, which is estimated by taking into account the preferential diffusion effect for each fuel component. In addition, a model of turbulent burning velocity proposed for single-component fuel mixtures may be applied to two-component fuel mixtures by considering the estimated mean local burning velocity of each fuel.

  20. Combustor assembly in a gas turbine engine

    DOEpatents

    Wiebe, David J; Fox, Timothy A

    2013-02-19

    A combustor assembly in a gas turbine engine. The combustor assembly includes a combustor device coupled to a main engine casing, a first fuel injection system, a transition duct, and an intermediate duct. The combustor device includes a flow sleeve for receiving pressurized air and a liner disposed radially inwardly from the flow sleeve. The first fuel injection system provides fuel that is ignited with the pressurized air creating first working gases. The intermediate duct is disposed between the liner and the transition duct and defines a path for the first working gases to flow from the liner to the transition duct. An intermediate duct inlet portion is associated with a liner outlet and allows movement between the intermediate duct and the liner. An intermediate duct outlet portion is associated with a transition duct inlet section and allows movement between the intermediate duct and the transition duct.

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