National Library of Energy BETA

Sample records for test assembly irradiation

  1. Lead test assembly irradiation and analysis Watts Bar Nuclear Plant, Tennessee and Hanford Site, Richland, Washington

    SciTech Connect (OSTI)

    1997-07-01

    The U.S. Department of Energy (DOE) needs to confirm the viability of using a commercial light water reactor (CLWR) as a potential source for maintaining the nation`s supply of tritium. The Proposed Action discussed in this environmental assessment is a limited scale confirmatory test that would provide DOE with information needed to assess that option. This document contains the environmental assessment results for the Lead test assembly irradiation and analysis for the Watts Bar Nuclear Plant, Tennessee, and the Hanford Site in Richland, Washington.

  2. Lead Test Assembly Irradiation and Analysis Watts Bar Nuclear Plant, Tennessee and Hanford Site, Richland, Washington

    Office of Energy Efficiency and Renewable Energy (EERE)

    This EA evaluates the environmental impacts associated with the U.S. Department of Energy proposed action to conduct a lead test assembly program to confirm the viability of using a commercial...

  3. Production decladding of irradiated fuel assemblies using a YAG laser

    SciTech Connect (OSTI)

    Frazier, R.S.; Campbell, D.C.

    1987-01-01

    Three hundred fifty-seven irradiated blanket assemblies containing 17 tonnes of depleted uranium (DU) from the Experimental Breeder Reactor II (EBR-II) were shipped to the Rockwell International Santa Susana Field Test Laboratory for decladding. Each assembly contained 19 fuel pins with approx. 2.5 kg of DU and inbred plutonium ranging from 10 to 450 g per assembly. The Rockwell International Hot Laboratory (RIHL) disassembled the fuel pins, removed the sodium bond material, and shipped the fuel slugs to the Savannah River Plant for reprocessing and plutonium recovery. In all, 6783 fuel pins containing 76.181 kg of plutonium were declad for reprocessing. To process the large number of fuel pins in an expedient manner, a 400-W neodymium-doped yttrium aluminum garnet (YAG) laser was used to cut the cladding. The laser resulted in a high-capacity process.

  4. FUEL ASSEMBLY SHAKER TEST SIMULATION

    SciTech Connect (OSTI)

    Klymyshyn, Nicholas A.; Sanborn, Scott E.; Adkins, Harold E.; Hanson, Brady D.

    2013-05-30

    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 refined 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.

  5. Safety evaluation report related to the Department of Energy`s proposal for the irradiation of lead test assemblies containing tritium-producing burnable absorber rods in commercial light-water reactors. Project Number 697

    SciTech Connect (OSTI)

    1997-05-01

    The NRC staff has reviewed a report, submitted by DOE to determine whether the use of a commercial light-water reactor (CLWR) to irradiate a limited number of tritium-producing burnable absorber rods (TPBARs) in lead test assemblies (LTAs) raises generic issues involving an unreviewed safety question. The staff has prepared this safety evaluation to address the acceptability of these LTAs in accordance with the provision of 10 CFR 50.59 without NRC licensing action. As summarized in Section 10 of this safety evaluation, the staff has identified issues that require NRC review. The staff has also identified a number of areas in which an individual licensee undertaking irradiation of TPBAR LTAs will have to supplement the information in the DOE report before the staff can determine whether the proposed irradiation is acceptable at a particular facility. The staff concludes that a licensee undertaking irradiation of TPBAR LTAs in a CLWR will have to submit an application for amendment to its facility operating license before inserting the LTAs into the reactor.

  6. Irradiated Materials Examination and Testing Facility (IMET)...

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

    in a laboratory hood or modified glove boxes in the SPL. Applications Physical and mechanical properties testing Examination of irradiated materials Irradiated specimen storage...

  7. ALSEP CASK ASSEMBLY GEARBOX THERMAL VACUUM TEST

    E-Print Network [OSTI]

    Rathbun, Julie A.

    ATM-835 ALSEP CASK ASSEMBLY GEARBOX · THERMAL VACUUM TEST 1 July 1969 Prepared by: Approved by: L. R. Lewis #12;: :, t · ALSEP Cask Assembly Gearbox Thermal Vacuum Test NO. ATM-835 PAGE 1 Aerospace Assembly CARR of 24 and 25 June 1969 the following summary of test results is submitted. This summary deals

  8. Low-voltage switchgear and controlgear assemblies: part 1: type-tested and partially type-tested assemblies

    E-Print Network [OSTI]

    International Electrotechnical Commission. Geneva

    2004-01-01

    Low-voltage switchgear and controlgear assemblies: part 1: type-tested and partially type-tested assemblies

  9. TEST RESULTS FROM GAMMA IRRADIATION OF ALUMINUM OXYHYDROXIDES

    SciTech Connect (OSTI)

    Fisher, D.; Westbrook, M.; Sindelar, R.

    2012-02-01

    Hydrated metal oxides or oxyhydroxides boehmite and gibbsite that can form on spent aluminum-clad nuclear fuel assemblies during in-core and post-discharge wet storage were exposed as granular powders to gamma irradiation in a {sup 60}Co irradiator in closed laboratory test vessels with air and with argon as separate cover gases. The results show that boehmite readily evolves hydrogen with exposure up to a dose of 1.8 x 10{sup 8} rad, the maximum tested, in both a full-dried and moist condition of the powder, whereas only a very small measurable quantity of hydrogen was generated from the granular powder of gibbsite. Specific information on the test setup, sample characteristics, sample preparation, irradiation, and gas analysis are described.

  10. Tritium systems test assembly stabilization

    SciTech Connect (OSTI)

    Jasen, W. G. (William G.); Michelotti, R. A. (Roy A.); Anast, K. R. (Kurt R.); Tesch, Charles

    2004-01-01

    The Tritium Systems Test Assembly (TSTA) was a facility dedicated to tritium technology Research and Development (R&D) primarily for future fusion power reactors. The facility was conceived in mid 1970's, operations commenced in early 1980's, stabilization and deactivation began in 2000 and were completed in 2003. The facility will remain in a Surveillance and Maintenance (S&M) mode until the Department of Energy (DOE) funds demolition of the facility, tentatively in 2009. A safe and stable end state was achieved by the TSTA Facility Stabilization Project (TFSP) in anticipation of long term S&M. At the start of the stabilization project, with an inventory of approximately 140 grams of tritium, the facility was designated a Hazard Category (HC) 2 Non-Reactor Nuclear facility as defined by US Department of Energy standard DOE-STD-1027-92 (1997). The TSTA facility comprises a laboratory area, supporting rooms, offices and associated laboratory space that included more than 20 major tritium handling systems. The project's focus was to reduce the tritium inventory by removing bulk tritium, tritiated water wastes, and tritium-contaminated high-inventory components. Any equipment that remained in the facility was stabilized in place. All of the gloveboxes and piping were rendered inoperative and vented to atmosphere. All equipment, and inventoried tritium contamination, remaining in the facility was left in a safe-and-stable state. The project used the End Points process as defined by the DOE Office of Environmental Management (web page http://www.em.doe.- gov/deact/epman.htmtlo) document and define the end state required for the stabilization of TSTA Facility. The End Points process added structure that was beneficial through virtually all phases of the project. At completion of the facility stabilization project the residual tritium inventory was approximately 3,000 curies, considerably less than the 1.6-gram threshold for a HC 3 facility. TSTA is now designated as a Radiological Facility. Innovative approaches were employed for characterization and removal of legacy wastes and high inventory components. Major accomplishments included: (1) Reduction of tritium inventory, elimination of chemical hazards, and identification and posting of remaining hazards. (2) Removal of legacy wastes. (3) Transferred equipment for reuse in other DOE projects, including some at other DOE facilities. (4) Transferred facility in a safe and stable condition to the S&M organization. The project successfully completed all project goals and the TSTA facility was transferred into S&M on August 1,2003. This project demonstrates the benefit of radiological inventory reduction and the removal of legacy wastes to achieve a safe and stable end state that protects workers and the environment pending eventual demolition of the facility.

  11. Recent Accomplishments in the Irradiation Testing of Engineering-Scale Monolithic Fuel Specimens

    SciTech Connect (OSTI)

    N.E. Woolstenhulme; D.M. Wachs; M.K. Meyer; H.W. Glunz; R.B. Nielson

    2012-10-01

    The US fuel development team is focused on qualification and demonstration of the uranium-molybdenum monolithic fuel including irradiation testing of engineering-scale specimens. The team has recently accomplished the successful irradiation of the first monolithic multi-plate fuel element assembly within the AFIP-7 campaign. The AFIP-6 MKII campaign, while somewhat truncated by hardware challenges, exhibited successful irradiation of a large-scale monolithic specimen under extreme irradiation conditions. The channel gap and ultrasonic data are presented for AFIP-7 and AFIP-6 MKII, respectively. Finally, design concepts are summarized for future irradiations such as the base fuel demonstration and design demonstration experiment campaigns.

  12. FUEL ASSEMBLY SHAKER AND TRUCK TEST SIMULATION

    SciTech Connect (OSTI)

    Klymyshyn, Nicholas A.; Jensen, Philip J.; Sanborn, Scott E.; Hanson, Brady D.

    2014-09-25

    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 revised 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 travelling 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.

  13. Report on the evaluation of the tritium producing burnable absorber rod lead test assembly. Revision 1

    SciTech Connect (OSTI)

    1997-03-01

    This report describes the design and fabrication requirements for a tritium-producing burnable absorber rod lead test assembly and evaluates the safety issues associated with tritium-producing burnable absorber rod irradiation on the operation of a commercial light water reactor. The report provides an evaluation of the tritium-producing burnable absorber rod design and concludes that irradiation can be performed within U.S. Nuclear Regulatory Commission regulations applicable to a commercial pressurized light water reactor.

  14. Irradiation test of electrical insulation materials performed at

    E-Print Network [OSTI]

    McDonald, Kirk

    Irradiation test of electrical insulation materials performed at NCBJ, Poland RESMM12- 10 #12;Outlet · EuCARD insulators certification irradiation requirements · Selection of the irradiation source for the sample irradiation purpose · Irradiation cryostat and set-up at NCBJ, Swierk, POLAND

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

    SciTech Connect (OSTI)

    Kevin G. Field; Richard Howard; Michael Teague

    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 to 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.

  16. Irradiation Environment of the Materials Test Station

    SciTech Connect (OSTI)

    Pitcher, Eric John

    2012-06-21

    Conceptual design of the proposed Materials Test Station (MTS) at the Los Alamos Neutron Science Center (LANSCE) is now complete. The principal mission is the irradiation testing of advanced fuels and materials for fast-spectrum nuclear reactor applications. The neutron spectrum in the fuel irradiation region of MTS is sufficiently close to that of fast reactor that MTS can match the fast reactor fuel centerline temperature and temperature profile across a fuel pellet. This is an important characteristic since temperature and temperature gradients drive many phenomena related to fuel performance, such as phase stability, stoichiometry, and fission product transport. The MTS irradiation environment is also suitable in many respects for fusion materials testing. In particular, the rate of helium production relative to atomic displacements at the peak flux position in MTS matches well that of fusion reactor first wall. Nuclear transmutation of the elemental composition of the fusion alloy EUROFER97 in MTS is similar to that expected in the first wall of a fusion reactor.

  17. Structural analysis of fuel assembly clads for the Upgraded Transient Reactor Test Facility (TREAT Upgrade)

    SciTech Connect (OSTI)

    Ewing, T.F.; Wu, T.S.

    1986-01-01

    The Upgraded Transient Reactor Test Facility (TREAT Upgrade) is designed to test full-length, pre-irradiated fuel pins of the type used in large LMFBRs under accident conditions, such as severe transient overpower and loss-of-coolant accidents. In TREAT Upgrade, the central core region is to contain new fuel assemblies of higher fissile loadings to maximize the energy deposition to the test fuel. These fuel assemblies must withstand normal peak clad temperatures of 850/sup 0/C for hundreds of test transients. Due to high temperatures and gradients predicted in the clad, creep and plastic strain effects are significant, and the clad structural behavior cannot be analyzed by conventional linear techniques. Instead, the detailed elastic-plastic-creep behavior must be followed along the time-dependent load history. This paper presents details of the structural evaluations of the conceptual TREAT Upgrade fuel assembly clads.

  18. LWRS ATR Irradiation Testing Readiness Status

    SciTech Connect (OSTI)

    Kristine Barrett

    2012-09-01

    The Light Water Reactor Sustainability (LWRS) Program was established by the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors. The LWRS Program is divided into four R&D Pathways: (1) Materials Aging and Degradation; (2) Advanced Light Water Reactor Nuclear Fuels; (3) Advanced Instrumentation, Information and Control Systems; and (4) Risk-Informed Safety Margin Characterization. This report describes an irradiation testing readiness analysis in preparation of LWRS experiments for irradiation testing at the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) under Pathway (2). The focus of the Advanced LWR Nuclear Fuels Pathway is to improve the scientific knowledge basis for understanding and predicting fundamental performance of advanced nuclear fuel and cladding in nuclear power plants during both nominal and off-nominal conditions. This information will be applied in the design and development of high-performance, high burn-up fuels with improved safety, cladding integrity, and improved nuclear fuel cycle economics

  19. Instrumentation to Enhance Advanced Test Reactor Irradiations

    SciTech Connect (OSTI)

    J. L. Rempe; D. L. Knudson; K. G. Condie; J. E. Daw; S. C. Taylor

    2009-09-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  20. Fail-safe storage rack for irradiated fuel rod assemblies

    DOE Patents [OSTI]

    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.

  1. Furnace assembly

    DOE Patents [OSTI]

    Panayotou, Nicholas F. (Kennewick, WA); Green, Donald R. (Richland, WA); Price, Larry S. (Pittsburg, CA)

    1985-01-01

    A method of and apparatus for heating test specimens to desired elevated temperatures for irradiation by a high energy neutron source. A furnace assembly is provided for heating two separate groups of specimens to substantially different, elevated, isothermal temperatures in a high vacuum environment while positioning the two specimen groups symmetrically at equivalent neutron irradiating positions.

  2. Evaluation of FSV-1 cask for the transport of LWR irradiated fuel assemblies

    SciTech Connect (OSTI)

    Not Available

    1980-05-01

    The Model FSV-1 spent fuel shipping cask was designed by General Atomic Company (GA) to service the Fort St. Vrain (FSV) nuclear generating station, a High Temperature Gas Reactor (HTGR) owned and operated by Public Service Company of Colorado (PSC). This report presents an evaluation of the suitability of the FSV-1 cask for the transport of irradiated Light Water Reactor (LWR) fuel assemblies from both Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). The FSV-1 cask evaluation parameters covered a wide spectrum of LWR fuel assemblies, based on burnup in Megawatt Days/Metric Ton of Heavy Metal (MWD/MTHM) and years of decay since irradiation. The criteria for suitability included allowable radiation dose rates, cask surface and interior temperatures and the Gross Vehicle Weight (GVW) of the complete shipping system.

  3. Dose rate estimates from irradiated light-water-reactor fuel assemblies in air

    SciTech Connect (OSTI)

    Lloyd, W.R.; Sheaffer, M.K.; Sutcliffe, W.G.

    1994-01-31

    It is generally considered that irradiated spent fuel is so radioactive (self-protecting) that it can only be moved and processed with specialized equipment and facilities. However, a small, possibly subnational, group acting in secret with no concern for the environment (other than the reduction of signatures) and willing to incur substantial but not lethal radiation doses, could obtain plutonium by stealing and processing irradiated spent fuel that has cooled for several years. In this paper, we estimate the dose rate at various distances and directions from typical pressurized-water reactor (PWR) and boiling-water reactor (BWR) spent-fuel assemblies as a function of cooling time. Our results show that the dose rate is reduced rapidly for the first ten years after exposure in the reactor, and that it is reduced by a factor of {approx}10 (from the one year dose rate) after 15 years. Even for fuel that has cooled for 15 years, a lethal dose (LD50) of 450 rem would be received at 1 m from the center of the fuel assembly after several minutes. However, moving from 1 to 5 m reduces the dose rate by over a factor of 10, and moving from 1 to 10 m reduces the dose rate by about a factor of 50. The dose rates 1 m from the top or bottom of the assembly are considerably less (about 10 and 22%, respectively) than 1 m from the center of the assembly, which is the direction of the maximum dose rate.

  4. Robotic assembly and testing of rotary solenoids: Final report

    SciTech Connect (OSTI)

    Lembke, J.R.

    1987-02-01

    Robotic automation of rotary solenoid assembly and testing operations was investigated. A robot was used to evaluate the automation of solenoid torque measurements. A measurement arrangement was devised in which the robot connects the solenoid shaft to a torque sensor, rotates the solenoid housing, and correlates rotational data with digitized torque readings. Results are typically in good agreement with those from the manually operated universal tension tester. Stack assembly of the rotary solenoid was attempted using another robot. Starting with parts in a component tray, the robot assembles the solenoid vertically in a precision cavity and installs retaining rings on the shaft. Special tooling and a remote center compliance device in the robot wrist have enabled the prototype assembly process to be largely successful, in spite of extremely small clearances between mating parts.

  5. Monte-Carlo Code (MCNP) Modeling of the Advanced Test Reactor Applicable to the Mixed Oxide (MOX) Test Irradiation

    SciTech Connect (OSTI)

    G. S. Chang; R. C. Pederson

    2005-07-01

    Mixed oxide (MOX) test capsules prepared with weapons-derived plutonium have been irradiated to a burnup of 50 GWd/t. The MOX fuel was fabricated at Los Alamos National Laboratory by a master-mix process and has been irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Previous withdrawals of the same fuel have occurred at 9, 21, 30, and 40 GWd/t. Oak Ridge National Laboratory (ORNL) manages this test series for the Department of Energy’s Fissile Materials Disposition Program (FMDP). The fuel burnup analyses presented in this study were performed using MCWO, a welldeveloped tool that couples the Monte Carlo transport code MCNP with the isotope depletion and buildup code ORIGEN-2. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations for the ATR small I-irradiation test position. The purpose of this report is to validate both the Weapons-Grade Mixed Oxide (WG-MOX) test assembly model and the new fuel burnup analysis methodology by comparing the computed results against the neutron monitor measurements.

  6. Weapons-Grade MOX Fuel Burnup Characteristics in Advanced Test Reactor Irradiation

    SciTech Connect (OSTI)

    G. S. Chang

    2006-07-01

    Mixed oxide (MOX) test capsules prepared with weapons-derived plutonium have been irradiated to a burnup of 50 GWd/t. The MOX fuel was fabricated at Los Alamos National Laboratory (LANL) by a master-mix process and has been irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Previous withdrawals of the same fuel have occurred at 9, 21, 30, 40, and 50 GWd/t. Oak Ridge National Laboratory (ORNL) manages this test series for the Department of Energy’s Fissile Materials Disposition Program (FMDP). A UNIX BASH (Bourne Again SHell) script CMO has been written and validated at the Idaho National Laboratory (INL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN-2 (CMO). The new Monte Carlo burnup analysis methodology in this paper consists of MCNP coupling through CMO with ORIGEN-2(MCWO). MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. The fuel burnup analyses presented in this study were performed using MCWO. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations for the ATR small I-irradiation test position. The purpose of this report is to validate both the Weapons-Grade Mixed Oxide (WG-MOX) test assembly model and the new fuel burnup analysis methodology by comparing the computed results against the neutron monitor measurements and the irradiated WG-MOX post irradiation examination (PIE) data.

  7. Reliability Testing the Die-Attach of CPV Cell Assemblies

    SciTech Connect (OSTI)

    Bosco, N.; Sweet, C.; Kurtz, S.

    2011-02-01

    Results and progress are reported for a course of work to establish an efficient reliability test for the die-attach of CPV cell assemblies. Test vehicle design consists of a ~1 cm2 multijunction cell attached to a substrate via several processes. A thermal cycling sequence is developed in a test-to-failure protocol. Methods of detecting a failed or failing joint are prerequisite for this work; therefore both in-situ and non-destructive methods, including infrared imaging techniques, are being explored as a method to quickly detect non-ideal or failing bonds.

  8. Controlled synthesis of snowflake-like self-assemblies palladium nanostructures under microwave irradiation

    SciTech Connect (OSTI)

    Xie, Ting; Ma, Yue; Yang, Hanmin, E-mail: yanghanmin@hotmail.com; Li, Jinlin

    2013-08-01

    Graphical abstract: - Highlights: • We demonstrated the synthesis of snowflake-like palladium nanostructures for the first time. • We discussed the influencing factors on the synthesis of snowflake-like Pd nanostructures. • The molar ratio of H{sub 2}Pd{sub 4} to PVP at 5 is the optimal selection. • The growth process was discussed. - Abstract: Self-assembly snowflake-like palladium nanostructures were synthesized under microwave irradiation using H{sub 2}PdCl{sub 4} as precursor, benzyl alcohol as both solvent and reducing agent, and PVP as stabilizer. The Pd snowflake-like nanostructures were formed and then characterized by transmission electron microscopy (TEM) and X-ray powder diffraction. The TEM images showed that the Pd nano-snowflakes were self-assemblies organized by hundreds of small spherical nanoparticles. Pd snowflake-like nanostructures with well-defined shape and uniform size can be obtained by tuning the concentration of palladium precursor, the molar ratio of H{sub 2}PdCl{sub 4}/PVP, as well as the heating time by microwave irradiation. The possible growing process of the snowflake-like Pd structures was also proposed on the basis of investigating the properties of as-synthesized Pd nanostructures under different conditions.

  9. Small Specimen Mechanical Property Testing of Irradiated

    E-Print Network [OSTI]

    McDonald, Kirk

    Brittle fracture of ferritic/martensitic steel fracture toughness specimen #12;Fracture Toughness Testing to smallest gauge dimension must be 10. Typical stainless steel has ~25 µm grain size, so gauge thickness

  10. AGR-2 IRRADIATION TEST FINAL AS-RUN REPORT

    SciTech Connect (OSTI)

    Blaise, Collin

    2014-07-01

    This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a half calendar years. The test contained six independently controlled and monitored capsules. Each U.S. capsule contained 12 compacts of either UCO or UO2 AGR coated fuel. No fuel particles failed during the AGR-2 irradiation. Final burnup values on a per compact basis ranged from 7.26 to 13.15% FIMA (fissions per initial heavy-metal atom) for UCO fuel, and 9.01 to 10.69% FIMA for UO2 fuel, while fast fluence values ranged from 1.94 to 3.47´1025 n/m2 (E >0.18 MeV) for UCO fuel, and from 3.05 to 3.53´1025 n/m2 (E >0.18 MeV) for UO2 fuel. Time-average volume-average (TAVA) temperatures on a capsule basis at the end of irradiation ranged from 987°C in Capsule 6 to 1296°C in Capsule 2 for UCO, and from 996 to 1062°C in UO2-fueled Capsule 3. By the end of the irradiation, all of the installed thermocouples (TCs) had failed. Fission product release-to-birth (R/B) ratios were quite low. In the UCO capsules, R/B values during the first three cycles were below 10-6 with the exception of the hotter Capsule 2, in which the R/Bs reached 2´10-6. In the UO2 capsule (Capsule 3), the R/B values during the first three cycles were below 10-7. R/B values for all following cycles are not reliable due to gas flow and cross talk issues.

  11. AGR-1 Irradiation Test Final As-Run Report

    SciTech Connect (OSTI)

    Blaise P. Collin

    2012-06-01

    This document presents the as-run analysis of the AGR-1 irradiation experiment. AGR-1 is the first of eight planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the US Department of Energy (DOE) as part of the Next-Generation Nuclear Plant (NGNP) project. The objectives of the AGR-1 experiment are: 1. To gain experience with multi-capsule test train design, fabrication, and operation with the intent to reduce the probability of capsule or test train failure in subsequent irradiation tests. 2. To irradiate fuel produced in conjunction with the AGR fuel process development effort. 3. To provide data that will support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. In order to achieve the test objectives, the AGR-1 experiment was irradiated in the B-10 position of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) for a total duration of 620 effective full power days of irradiation. Irradiation began on December 24, 2006 and ended on November 6, 2009 spanning 13 ATR cycles and approximately three calendar years. The test contained six independently controlled and monitored capsules. Each capsule contained 12 compacts of a single type, or variant, of the AGR coated fuel. No fuel particles failed during the AGR-1 irradiation. Final burnup values on a per compact basis ranged from 11.5 to 19.6 %FIMA, while fast fluence values ranged from 2.21 to 4.39 ?1025 n/m2 (E >0.18 MeV). We’ll say something here about temperatures once thermal recalc is done. Thermocouples performed well, failing at a lower rate than expected. At the end of the irradiation, nine of the originally-planned 19 TCs were considered functional. Fission product release-to-birth (R/B) ratios were quite low. In most capsules, R/B values at the end of the irradiation were at or below 10-7 with only one capsule significantly exceeding this value. A maximum R/B of around 2?10-7 was reached at the end of the irradiation in Capsule 5. Several shakedown issues were encountered and resolved during the first three cycles. These include the repair of minor gas line leaks; repair of faulty gas line valves; the need to position moisture monitors in regions of low radiation fields for proper functioning; the enforcement of proper on-line data storage and backup, the need to monitor thermocouple performance, correcting for detector spectral gain shift, and a change in the mass flow rate range of the neon flow controllers.

  12. Gamma Irradiation of sLHC Prototype Silicon Strip Detectors and Test Structures of the SMART

    E-Print Network [OSTI]

    California at Santa Cruz, University of

    Gamma Irradiation of sLHC Prototype Silicon Strip Detectors and Test Structures of the SMART funded SMART project. They were irradiated with 60 Co to test total dose (TID) effects, in order to study

  13. AGR 3/4 Irradiation Test Final As Run Report

    SciTech Connect (OSTI)

    Collin, Blaise P.

    2015-06-01

    Several fuel and material irradiation experiments have been planned for the Idaho National Laboratory Advanced Reactor Technologies Technology Development Office Advanced Gas Reactor Fuel Development and Qualification Program (referred to as the INL ART TDO/AGR fuel program hereafter), which supports the development and qualification of tristructural-isotropic (TRISO) coated particle fuel for use in HTGRs. The goals of these experiments are to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support development and validation of fuel performance and fission product transport models and codes, and provide irradiated fuel and materials for post irradiation examination and safety testing (INL 05/2015). AGR-3/4 combined the third and fourth in this series of planned experiments to test TRISO coated low enriched uranium (LEU) oxycarbide fuel. This combined experiment was intended to support the refinement of fission product transport models and to assess the effects of sweep gas impurities on fuel performance and fission product transport by irradiating designed-to-fail fuel particles and by measuring subsequent fission metal transport in fuel-compact matrix material and fuel-element graphite. The AGR 3/4 fuel test was successful in irradiating the fuel compacts to the burnup and fast fluence target ranges, considering the experiment was terminated short of its initial 400 EFPD target (Collin 2015). Out of the 48 AGR-3/4 compacts, 42 achieved the specified burnup of at least 6% fissions per initial heavy-metal atom (FIMA). Three capsules had a maximum fuel compact average burnup < 10% FIMA, one more than originally specified, and the maximum fuel compact average burnup was <19% FIMA for the remaining capsules, as specified. Fast neutron fluence fell in the expected range of 1.0 to 5.5×1025 n/m2 (E >0.18 MeV) for all compacts. In addition, the AGR-3/4 experiment was globally successful in keeping the temperature in the twelve capsules relatively flat in a range of temperatures suitable for the measurement of fission product diffusion in compact matrix and structural graphite materials.

  14. Updated FY12 Ceramic Fuels Irradiation Test Plan

    SciTech Connect (OSTI)

    Nelson, Andrew T.

    2012-05-24

    The Fuel Cycle Research and Development program is currently devoting resources to study of numerous fuel types with the aim of furthering understanding applicable to a range of reactors and fuel cycles. In FY11, effort within the ceramic fuels campaign focused on planning and preparation for a series of rabbit irradiations to be conducted at the High Flux Isotope Reactor located at Oak Ridge National Laboratory. The emphasis of these planned tests was to study the evolution of thermal conductivity in uranium dioxide and derivative compositions as a function of damage induced by neutron damage. Current fiscal realities have resulted in a scenario where completion of the planned rabbit irradiations is unlikely. Possibilities for execution of irradiation testing within the ceramic fuels campaign in the next several years will thus likely be restricted to avenues where strong synergies exist both within and outside the Fuel Cycle Research and Development program. Opportunities to augment the interests and needs of modeling, advanced characterization, and other campaigns present the most likely avenues for further work. These possibilities will be pursued with the hope of securing future funding. Utilization of synthetic microstructures prepared to better understand the most relevant actors encountered during irradiation of ceramic fuels thus represents the ceramic fuel campaign's most efficient means to enhance understanding of fuel response to burnup. This approach offers many of the favorable attributes embraced by the Separate Effects Testing paradigm, namely production of samples suitable to study specific, isolated phenomena. The recent success of xenon-imbedded thick films is representative of this approach. In the coming years, this strategy will be expanded to address a wider range of problems in conjunction with use of national user facilities novel characterization techniques to best utilize programmatic resources to support a science-based research program.

  15. Nov. 21, 1999 Neutron Irradiation Tests of an S-LINK-over-G-link System

    E-Print Network [OSTI]

    Nov. 21, 1999 Neutron Irradiation Tests of an S-LINK-over-G-link System K. Anderson, J. Pilcher, H and was carefully monitored. 2. Neutron Source Characteristics The test was performed at the National Institute, MD 1. Objective This note describes neutron irradiation tests of an S-LINK [1] source card (LSC

  16. Irradiation Test of Advanced PWR Fuel in Fuel Test Loop at HANARO

    SciTech Connect (OSTI)

    Yang, Yong Sik; Bang, Je Geon; Kim, Sun Ki; Song, Kun Woo; Park, Su Ki; Seo, Chul Gyo

    2007-07-01

    A new fuel test loop has been constructed in the research reactor HANARO at KAERI. The main objective of the FTL (Fuel Test Loop) is an irradiation test of a newly developed LWR fuel under PWR or Candu simulated conditions. The first test rod will be loaded within 2007 and its irradiation test will be continued until a rod average their of 62 MWd/kgU. A total of five test rods can be loaded into the IPS (In-Pile Section) and fuel centerline temperature, rod internal pressure and fuel stack elongation can be measured by an on-line real time system. A newly developed advanced PWR fuel which consists of a HANA{sup TM} alloy cladding and a large grain UO{sub 2} pellet was selected as the first test fuel in the FTL. The fuel cladding, the HANA{sup TM} alloy, is an Nb containing Zirconium alloy that has shown better corrosion and creep resistance properties than the current Zircaloy-4 cladding. A total of six types of HANA{sup TM} alloy were developed and two or three of these candidate alloys will be used as test rod cladding, which have shown a superior performance to the others. A large-grain UO{sub 2} pellet has a 14{approx}16 micron 2D diameter grain size for a reduction of a fission gas release at a high burnup. In this paper, characteristics of the FTL and IPS are introduced and the expected operation and irradiation conditions are summarized for the test periods. Also the preliminary fuel performance analysis results, such as the cladding oxide thickness, fission gas release and rod internal pressure, are evaluated from the test rod safety analysis aspects. (authors)

  17. Assembly and method for testing the integrity of stuffing tubes

    DOE Patents [OSTI]

    Morrison, Edward Francis (Burnt Hills, NY)

    1997-01-01

    A stuffing tube integrity checking assembly includes first and second annular seals, with each seal adapted to be positioned about a stuffing tube penetration component. An annular inflation bladder is provided, the bladder having a slot extending longitudinally therealong and including a separator for sealing the slot. A first valve is in fluid communication with the bladder for introducing pressurized fluid to the space defined by the bladder when mounted about the tube. First and second releasible clamps are provided. Each clamp assembly is positioned about the bladder for securing the bladder to one of the seals for thereby establishing a fluid-tight chamber about the tube.

  18. Report of Survey of the Los Alamos Tritium Systems Test Assembly Facility

    Broader source: Energy.gov [DOE]

    The purpose of this document is to report the results of a survey conducted at the Los Alamos Tritium Systems Test Assembly (TSTA Facility). The survey was conducted during the week of 3/20/00.

  19. AFCI Fuel Irradiation Test Plan, Test Specimens AFC-1Æ and AFC-1F

    SciTech Connect (OSTI)

    D. C. Crawford; S. L. Hayes; B. A. Hilton; M. K. Meyer; R. G. Ambrosek; G. S. Chang; D. J. Utterbeck

    2003-11-01

    The U. S. Advanced Fuel Cycle Initiative (AFCI) seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products, thereby dramatically decreasing the volume of material requiring disposition and the long-term radiotoxicity and heat load of high-level waste sent to a geologic repository (DOE, 2003). One important component of the technology development is actinide-bearing transmutation fuel forms containing plutonium, neptunium, americium (and possibly curium) isotopes. There are little irradiation performance data available on non-fertile fuel forms, which would maximize the destruction rate of plutonium, and low-fertile (i.e., uranium-bearing) fuel forms, which would support a sustainable nuclear energy option. Initial scoping level irradiation tests on a variety of candidate fuel forms are needed to establish a transmutation fuel form design and evaluate deployment of transmutation fuels.

  20. Ion irradiation testing of Improved Accident Tolerant Cladding Materials

    SciTech Connect (OSTI)

    Anderoglu, Osman [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tesmer, Joseph R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Maloy, Stuart A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-01-14

    This report summarizes the results of ion irradiations conducted on two FeCrAl alloys (named as ORNL A&B) for improving the accident tolerance of LWR nuclear fuel cladding. After irradiation with 1.5 MeV protons to ~0.5 to ~1 dpa and 300°C nanoindentations were performed on the cross-sections along the ion range. An increase in hardness was observed in both alloys. Microstructural analysis shows radiation induced defects.

  1. Initiate test loop irradiations of ALSEP process solvent

    SciTech Connect (OSTI)

    Peterman, Dean R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Olson, Lonnie G. [Idaho National Lab. (INL), Idaho Falls, ID (United States); McDowell, Rocklan G. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    This report describes the initial results of the study of the impacts of gamma radiolysis upon the efficacy of the ALSEP process and is written in completion of milestone M3FT-14IN030202. Initial irradiations, up to 100 kGy absorbed dose, of the extraction section of the ALSEP process have been completed. The organic solvent used for these experiments contained 0.05 M TODGA and 0.75 M HEH[EHP] dissolved in n-dodecane. The ALSEP solvent was irradiated while in contact with 3 M nitric acid and the solutions were sparged with compressed air in order to maintain aerated conditions. The irradiated phases were used for the determination of americium and europium distribution ratios as a function of absorbed dose for the extraction and stripping conditions. Analysis of the irradiated phases in order to determine solvent composition as a function of absorbed dose is ongoing. Unfortunately, the failure of analytical equipment necessary for the analysis of the irradiated samples has made the consistent interpretation of the analytical results difficult. Continuing work will include study of the impacts of gamma radiolysis upon the extraction of actinides and lanthanides by the ALSEP solvent and the stripping of the extracted metals from the loaded solvent. The irradiated aqueous and organic phases will be analyzed in order to determine the variation in concentration of solvent components with absorbed gamma dose. Where possible, radiolysis degradation product will be identified.

  2. Subtask 1: Total systems analysis, assembly and testing | Center for

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home RoomPreservationBio-Inspired Solar Fuel Production 1: Total systems analysis, assembly and

  3. Summary of irradiation tests of mixed oxide fuel prepared with weapons-derived plutonium

    SciTech Connect (OSTI)

    Ott, Larry J [ORNL; Spellman, Donald J [ORNL

    2008-01-01

    Mixed oxide (MOX) test capsules prepared with weapons-derived plutonium were irradiated to a burnup of 50 GWd/MT. The MOX fuel was fabricated at Los Alamos National Laboratory by a master-mix process and irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Previous withdrawals of the same fuel have occurred at 9, 21, 30, and 40 GWd/MT. Oak Ridge National Laboratory managed this test series for the Department of Energy s Fissile Materials Disposition Program (FMDP). This paper describes the preparation of the MOX fuel, the equipment design, and the irradiation history of the test capsules and discussesthe significance of the more important observations of the post-irradiation examinations (PIEs). Fuel performance has been excellent and consistent with code predictions and with existing U.S. and European experience.

  4. Summary of Irradiation Tests of Mixed-Oxide Fuel Prepared with Weapons-Derived Plutonium

    SciTech Connect (OSTI)

    Ott, Larry J [ORNL; Spellman, Donald J [ORNL

    2008-01-01

    Mixed oxide (MOX) test capsules prepared with weapons-derived plutonium have been irradiated to a burnup of 50 GWd/MT. The MOX fuel was fabricated at Los Alamos National Laboratory by a master-mix process and irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Previous withdrawals of the same fuel have occurred at 9, 21, 30, and 50 GWd/MT. Oak Ridge National Laboratory managed this test series for the Department of Energy's Fissile Materials Disposition Program (FMDP). This paper describes the preparation of the MOX fuel, the equipment desig, and the irradiation history of the test capsules and discusses the significance of the more important observations of the post-irradiation examinations (PIEs). Fuel performance has been excellent and consistent with code predictions and with existing U.S. and Europen experience.

  5. EIS-0017: Fusion Materials Irradiation Testing Facility, Hanford Reservation, Richland, Washington

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy developed this statement to evaluate the environmental impacts associated with proposed construction and operation of an irradiation test facility, the Deuterium-Lithium High Flux Neutron Source Facility, at the Hanford Reservation.

  6. The Next Generation Nuclear Plant Graphite Creep Experiment Irradiation in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Blaine Grover

    2010-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will have differing compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during initial start-up of the experiment. The first experiment was inserted in the ATR in August 2009 and started its irradiation in September 2009. It is anticipated to complete its irradiation in early calendar 2011. This paper will discuss the design of the experiment including the test train and the temperature and compressive load monitoring, control, and the irradiation experience to date.

  7. Irradiation Programs and Test Plans to Assess High-Fluence Irradiation Assisted Stress Corrosion Cracking Susceptibility.

    SciTech Connect (OSTI)

    Teysseyre, Sebastien

    2015-03-01

    . Irradiation assisted stress corrosion cracking (IASCC) is a known issue in current reactors. In a 60 year lifetime, reactor core internals may experience fluence levels up to 15 dpa for boiling water reactors (BWR) and 100+ dpa for pressurized water reactors (PWR). To support a safe operation of our fleet of reactors and maintain their economic viability it is important to be able to predict any evolution of material behaviors as reactors age and therefore fluence accumulated by reactor core component increases. For PWR reactors, the difficulty to predict high fluence behavior comes from the fact that there is not a consensus of the mechanism of IASCC and that little data is available. It is however possible to use the current state of knowledge on the evolution of irradiated microstructure and on the processes that influences IASCC to emit hypotheses. This report identifies several potential changes in microstructure and proposes to identify their potential impact of IASCC. The susceptibility of a component to high fluence IASCC is considered to not only depends on the intrinsic IASCC susceptibility of the component due to radiation effects on the material but to also be related to the evolution of the loading history of the material and interaction with the environment as total fluence increases. Single variation type experiments are proposed to be performed with materials that are representative of PWR condition and with materials irradiated in other conditions. To address the lack of IASCC propagation and initiation data generated with material irradiated in PWR condition, it is proposed to investigate the effect of spectrum and flux rate on the evolution of microstructure. A long term irradiation, aimed to generate a well-controlled irradiation history on a set on selected materials is also proposed for consideration. For BWR, the study of available data permitted to identify an area of concern for long term performance of component. The efficiency of hydrogen water chemistry mitigation technology may decrease as fluence increases for high-stress intensity factors. This report describes a program plan to determine the efficiency of hydrogen water chemistry as a function of the stress intensity factor applied and fluence. The use of existing, available, materials and the generation of additional materials via irradiation in a research reactor are considered.

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

    DOE Patents [OSTI]

    Phillips, J.R.; Halbig, J.K.; Menlove, H.O.; Klosterbuer, S.F.

    1984-01-01

    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.

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

    DOE Patents [OSTI]

    Phillips, John R. (Los Alamos, NM); Halbig, James K. (Los Alamos, NM); Menlove, Howard O. (Los Alamos, NM); Klosterbuer, Shirley F. (Los Alamos, NM)

    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.

  10. Status of the NGNP fuel experiment AGR-2 irradiated in the advanced test reactor

    SciTech Connect (OSTI)

    S. Blaine Grover; David A. Petti

    2014-05-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also undergo on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and sup

  11. Status of the NGNP Fuel Experiment AGR-2 Irradiated in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and support systems will be briefly discussed, followed by the progress and status of the experiment to date.

  12. JOYO-1 Irradiation Test Campaign Technical Close-out, For Information

    SciTech Connect (OSTI)

    G. Borges

    2006-01-31

    The JOYO-1 irradiation testing was designed to screen the irradiation performance of candidate cladding, structural and reflector materials in support of space reactor development. The JOYO-1 designation refers to the first of four planned irradiation tests in the JOYO reactor. Limited irradiated material performance data for the candidate materials exists for the expected Prometheus-1 duration, fluences and temperatures. Materials of interest include fuel element cladding and core materials (refractory metal alloys and silicon carbide (Sic)), vessel and plant structural materials (refractory metal alloys and nickel-base superalloys), and control and reflector materials (BeO). Key issues to be evaluated were long term microstructure and material property stability. The JOYO-1 test campaign was initiated to irradiate a matrix of specimens at prototypical temperatures and fluences anticipated for the Prometheus-1 reactor [Reference (1)]. Enclosures 1 through 9 describe the specimen and temperature monitors/dosimetry fabrication efforts, capsule design, disposition of structural material irradiation rigs, and plans for post-irradiation examination. These enclosures provide a detailed overview of Naval Reactors Prime Contractor Team (NRPCT) progress in specific areas; however, efforts were in various states of completion at the termination of NRPCT involvement with and restructuring of Project Prometheus.

  13. Linear variable differential transformer (LVDT)-based elongation measurements in Advanced Test Reactor high temperature irradiation testing

    SciTech Connect (OSTI)

    D. L. Knudson; J. L. Rempe

    2012-02-01

    New materials are being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. These materials can undergo significant dimensional and physical changes during high temperature irradiations. Currently, such changes are determined by repeatedly irradiating a specimen for a specified period of time in the Advanced Test Reactor (ATR) and then removing it from the reactor for evaluation. The labor and time to remove, examine, and return irradiated samples for each measurement makes this approach very expensive. In addition, such techniques provide limited data and may disturb the phenomena of interest. To resolve these issues, an instrumented creep testing capability is being developed for specimens irradiated in pressurized water reactor (PWR) coolant conditions in the ATR at the Idaho National Laboratory (INL). This paper reports the status of INL efforts to develop this testing capability. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper focuses on efforts to design and evaluate a prototype test rig in an autoclave at INL's High Temperature Test Laboratory (HTTL).

  14. Linear variable differential transformer (LVDT)-based elongation measurements in Advanced Test Reactor high temperature irradiation testing

    SciTech Connect (OSTI)

    D. L. Knudson; J. L. Rempe

    2012-02-01

    New materials are being considered for fuel, cladding and structures in next generation and existing nuclear reactors. These materials can undergo significant dimensional and physical changes during high temperature irradiations. Currently, such changes are determined by repeatedly irradiating a specimen for a specified period of time in the Advanced Test Reactor (ATR) and then removing it from the reactor for evaluation. The labor and time to remove, examine and return irradiated samples for each measurement make this approach very expensive. In addition, such techniques provide limited data and may disturb the phenomena of interest. To resolve these issues, an instrumented creep testing capability is being developed for specimens irradiated under pressurized water reactor coolant conditions in the ATR at the Idaho National Laboratory (INL). This paper reports the status of INL efforts to develop this testing capability. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper focuses on efforts to design and evaluate a prototype test rig in an autoclave at INL’s High Temperature Test Laboratory.

  15. Testing a Novel Self-Assembling Data Paradigm in the Context of IACT Data

    E-Print Network [OSTI]

    Amanda Weinstein; Lucy Fortson; Thomas Brantseg; Cameron Rulten; Robyn Lutz; Jarvis Haupt; Mojtaba Kakhodaie Elyaderani; John Quinn

    2015-09-07

    The process of gathering and associating data from multiple sensors or sub-detectors due to a common physical event (the process of event-building) is used in many fields, including high-energy physics and $\\gamma$-ray astronomy. Fault tolerance in event-building is a challenging problem that increases in difficulty with higher data throughput rates and increasing numbers of sub-detectors. We draw on biological self-assembly models in the development of a novel event-building paradigm that treats each packet of data from an individual sensor or sub-detector as if it were a molecule in solution. Just as molecules are capable of forming chemical bonds, "bonds" can be defined between data packets using metadata-based discriminants. A database -- which plays the role of a beaker of solution -- continually selects pairs of assemblies at random to test for bonds, which allows single packets and small assemblies to aggregate into larger assemblies. During this process higher-quality associations supersede spurious ones. The database thereby becomes fluid, dynamic, and self-correcting rather than static. We will describe tests of the self-assembly paradigm using our first fluid database prototype and data from the VERITAS $\\gamma$-ray telescope.

  16. SUMMARY OF ‘AFIP’ FULL SIZED PLATE IRRADIATIONS IN THE ADVANCED TEST REACTOR

    SciTech Connect (OSTI)

    Robinson, Adam B; Wachs, Daniel M

    2010-03-01

    Recent testing at the Idaho National Laboratory has included four AFIP (ATR Full Size plate In center flux trap Position) experiments. These experiments included both dispersion plates and monolithic plates fabricated by both hot isostatic pressing and friction bonding utilizing both thermally sprayed inter-layers and zirconium barriers. These plates were tested between 100 and 350 w/cm2 at low temperatures and high burn-ups. The post irradiation exams performed have indicated good performance under the conditions tested and a summary of the findings and irradiation history are included herein.

  17. Neutron-Irradiated Samples as Test Materials for MPEX

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Ellis, Ronald James; Rapp, Juergen

    2015-10-09

    Plasma Material Interaction (PMI) is a major concern in fusion reactor design and analysis. The Material-Plasma Exposure eXperiment (MPEX) will explore PMI under fusion reactor plasma conditions. Samples with accumulated displacements per atom (DPA) damage produced by fast neutron irradiations in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) will be studied in the MPEX facility. This paper presents assessments of the calculated induced radioactivity and resulting radiation dose rates of a variety of potential fusion reactor plasma-facing materials (such as tungsten). The scientific code packages MCNP and SCALE were used to simulate irradiation of themore »samples in HFIR including the generation and depletion of nuclides in the material and the subsequent composition, activity levels, gamma radiation fields, and resultant dose rates as a function of cooling time. A challenge of the MPEX project is to minimize the radioactive inventory in the preparation of the samples and the sample dose rates for inclusion in the MPEX facility.« less

  18. Neutron-Irradiated Samples as Test Materials for MPEX

    SciTech Connect (OSTI)

    Ellis, Ronald James; Rapp, Juergen

    2015-10-09

    Plasma Material Interaction (PMI) is a major concern in fusion reactor design and analysis. The Material-Plasma Exposure eXperiment (MPEX) will explore PMI under fusion reactor plasma conditions. Samples with accumulated displacements per atom (DPA) damage produced by fast neutron irradiations in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) will be studied in the MPEX facility. This paper presents assessments of the calculated induced radioactivity and resulting radiation dose rates of a variety of potential fusion reactor plasma-facing materials (such as tungsten). The scientific code packages MCNP and SCALE were used to simulate irradiation of the samples in HFIR including the generation and depletion of nuclides in the material and the subsequent composition, activity levels, gamma radiation fields, and resultant dose rates as a function of cooling time. A challenge of the MPEX project is to minimize the radioactive inventory in the preparation of the samples and the sample dose rates for inclusion in the MPEX facility.

  19. Advanced Gas Reactor (AGR)-5/6/7 Fuel Irradiation Experiments in the Advanced Test Reactor

    SciTech Connect (OSTI)

    A. Joseph Palmer; David A. Petti; S. Blaine Grover

    2014-04-01

    The United States Department of Energy’s Very High Temperature Reactor (VHTR) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which each consist of at least five separate capsules, are being irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gases also have on-line fission product monitoring the effluent from each capsule to track performance of the fuel during irradiation. The first two experiments (designated AGR-1 and AGR-2), have been completed. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. The design of the fuel qualification experiment, designated AGR-5/6/7, is well underway and incorporates lessons learned from the three previous experiments. Various design issues will be discussed with particular details related to selection of thermometry.

  20. Test plan for the irradiation of nonmetallic materials.

    SciTech Connect (OSTI)

    Brush, Laurence H.; Farnum, Cathy Ottinger; Dahl, M.; Joslyn, C. C.; Venetz, T. J.

    2013-05-01

    A comprehensive test program to evaluate nonmetallic materials use in the Hanford tank farms is described in detail. This test program determines the effects of simultaneous multiple stressors at reasonable conditions on in-service configuration components by engineering performance testing.

  1. Test plan for the irradiation of nonmetallic materials.

    SciTech Connect (OSTI)

    Brush, Laurence H.; Farnum, Cathy Ottinger; Gelbard, Fred; Dahl, M.; Joslyn, C. C.; Venetz, T. J.

    2013-03-01

    A comprehensive test program to evaluate nonmetallic materials use in the Hanford Tank Farms is described in detail. This test program determines the effects of simultaneous multiple stressors at reasonable conditions on in-service configuration components by engineering performance testing.

  2. Thermal cycling tests on surface-mount assemblies

    SciTech Connect (OSTI)

    Jennings, C.W.

    1988-03-01

    The capability of surface-mount (SM) solder joints to withstand various thermal cycle stresses was evaluated through electrical circuit resistance changes of a test pattern and by visual examination for cracks in the solder after exposure to thermal cycling. The joints connected different electrical components, primarily leadless-chip carriers (LCCs), and printed wiring-board (PWB) pads on different laminate substrates. Laminate compositions were epoxy-glass and polyimide-glass with and without copper/Invar/copper (CIC) inner layers, polyimide-quartz, epoxy-Kevlar, and polyimide-Kevlar. The most resistant joints were between small LCCs (24 and 48 pins) and polyimide-glass laminate with CIC inner layers. Processing in joint formation was found to be an important part of joint resistant. Thermal cycling was varied with respect to both time and temperature. A few resistors, capacitors, and inductors showed opens after 500 30-min cycles between -65/degree/C and 125/degree/C. Appreciable moisture contents were measured for laminate materials, especially those of polyimide-Kevlar after equilibration in 100/percent/ relative humidity at room temperature. If not removed or reduced, moisture can cause delamination in vapor-phase soldering. 17 refs, 12 figs.,10 tabs.

  3. Fuel Accident Condition Simulator (FACS) Furnace for Post-Irradiation Heating Tests of VHTR Fuel Compacts

    SciTech Connect (OSTI)

    Paul A Demkowicz; Paul Demkowicz; David V Laug

    2010-10-01

    Abstract –Fuel irradiation testing and post-irradiation examination are currently in progress as part of the Next Generation Nuclear Plant Fuels Development and Qualification Program. The PIE campaign will include extensive accident testing of irradiated very high temperature reactor fuel compacts to verify fission product retention characteristics at high temperatures. This work will be carried out at both the Idaho National Laboratory (INL) and the Oak Ridge National Laboratory, beginning with accident tests on irradiated fuel from the AGR-1 experiment in 2010. A new furnace system has been designed, built, and tested at INL to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000°C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, Eu, and I) and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator (FACS) furnace system, as well as preliminary system calibration results.

  4. AGR-2 irradiation test final as-run report, Rev. 1

    SciTech Connect (OSTI)

    Collin, Blaise

    2014-08-01

    This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities; (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing; and, (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a half calendar years. The test contained six independently controlled and monitored capsules. Each U.S. capsule contained 12 compacts of either UCO or UO2 AGR coated fuel. No fuel particles failed during the AGR-2 irradiation. Final burnup values on a per compact basis ranged from 7.26 to 13.15% FIMA (fissions per initial heavy-metal atom) for UCO fuel, and 9.01 to 10.69% FIMA for UO2 fuel, while fast fluence values ranged from 1.94 to 3.47´1025 n/m2 (E >0.18 MeV) for UCO fuel, and from 3.05 to 3.53´1025 n/m2 (E >0.18 MeV) for UO2 fuel. Time-average volume-average (TAVA) temperatures on a capsule basis at the end of irradiation ranged from 987°C in Capsule 6 to 1296°C in Capsule 2 for UCO, and from 996 to 1062°C in UO2-fueled Capsule 3. By the end of the irradiation, all of the installed thermocouples (TCs) had failed. Fission product release-to-birth (R/B) ratios were quite low. In the UCO capsules, R/B values during the first three cycles were below 10-6 with the exception of the hotter Capsule 2, in which the R/Bs reached 2´10-6. In the UO2 capsule (Capsule 3), the R/B values during the first three cycles were below 10-7. R/B values for all following cycles are not reliable due to gas flow and cross talk issues.

  5. AGR-2 Irradiation Test Final As-Run Report, Rev 2

    SciTech Connect (OSTI)

    Blaise Collin

    2014-08-01

    This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a half calendar years. The test contained six independently controlled and monitored capsules. Each U.S. capsule contained 12 compacts of either UCO or UO2 AGR coated fuel. No fuel particles failed during the AGR-2 irradiation. Final burnup values on a per compact basis ranged from 7.26 to 13.15% FIMA (fissions per initial heavy-metal atom) for UCO fuel, and 9.01 to 10.69% FIMA for UO2 fuel, while fast fluence values ranged from 1.94 to 3.47´1025 n/m2 (E >0.18 MeV) for UCO fuel, and from 3.05 to 3.53´1025 n/m2 (E >0.18 MeV) for UO2 fuel. Time-average volume-average (TAVA) temperatures on a capsule basis at the end of irradiation ranged from 987°C in Capsule 6 to 1296°C in Capsule 2 for UCO, and from 996 to 1062°C in UO2-fueled Capsule 3. By the end of the irradiation, all of the installed thermocouples (TCs) had failed. Fission product release-to-birth (R/B) ratios were quite low. In the UCO capsules, R/B values during the first three cycles were below 10-6 with the exception of the hotter Capsule 2, in which the R/Bs reached 2´10-6. In the UO2 capsule (Capsule 3), the R/B values during the first three cycles were below 10-7. R/B values for all following cycles are not reliable due to gas flow and cross talk issues.

  6. Status of the Combined Third and Fourth NGNP Fuel Irradiations In the Advanced Test Reactor

    SciTech Connect (OSTI)

    S. Blaine Grover; David A. Petti; Michael E. Davenport

    2013-07-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in September 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. Since the purpose of this combined experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is significantly different from the first two experiments, though the control and monitoring systems are extremely similar. The design of the experiment will be discussed followed by its progress and status to date.

  7. Fabrication, assembly, bench and drilling tests of two prototype downhole pneumatic turbine motors: Final technical report

    SciTech Connect (OSTI)

    Bookwalter, R.; Duettra, P.D.; Johnson, P.; Lyons, W.C.; Miska, S.

    1987-04-01

    The first and second prototype downhole pneumatic turbine motors have been fabricated, assembled and tested. All bench tests showed that the motor will produce horsepower and bit speeds approximating the predicted values. Specifically, the downhole pneumatic turbine motor produced approximately 50 horsepower at 100 rpm, while being supplied with about 3600 SCFM of compressed air. The first prototype was used in a drilling test from a depth of 389 feet to a depth of 789 feet in the Kirtland formation. This first prototype motor drilled at a rate exceeding 180 ft/hr, utilizing only 3000 SCFM of compressed air. High temperature tests (at approximately 460/sup 0/F) were carried out on the thrust assembly and the gearboxes for the two prototypes. These components operated successfully at these temperatures. Although the bench and drilling tests were successful, the tests revealed design changes that should be made before drilling tests are carried out in geothermal boreholes at the Geysers area, near Santa Rosa, California.

  8. AGR-1 Irradiation Test Final As-Run Report, Rev. 3

    SciTech Connect (OSTI)

    Collin, Blaise P.

    2015-01-01

    This document presents the as-run analysis of the AGR-1 irradiation experiment. AGR-1 is the first of eight planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the US Department of Energy (DOE) as part of the Next-Generation Nuclear Plant (NGNP) project. The objectives of the AGR-1 experiment are: 1. To gain experience with multi-capsule test train design, fabrication, and operation with the intent to reduce the probability of capsule or test train failure in subsequent irradiation tests. 2. To irradiate fuel produced in conjunction with the AGR fuel process development effort. 3. To provide data that will support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. In order to achieve the test objectives, the AGR-1 experiment was irradiated in the B-10 position of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) for a total duration of 620 effective full power days of irradiation. Irradiation began on December 24, 2006 and ended on November 6, 2009 spanning 13 ATR cycles and approximately three calendar years. The test contained six independently controlled and monitored capsules. Each capsule contained 12 compacts of a single type, or variant, of the AGR coated fuel. No fuel particles failed during the AGR-1 irradiation. Final burnup values on a per compact basis ranged from 11.5 to 19.6 %FIMA, while fast fluence values ranged from 2.21 to 4.39 x 1025 n/m2 (E >0.18 MeV). We’ll say something here about temperatures once thermal recalc is done. Thermocouples performed well, failing at a lower rate than expected. At the end of the irradiation, nine of the originally-planned 19 TCs were considered functional. Fission product release-to-birth (R/B) ratios were quite low. In most capsules, R/B values at the end of the irradiation were at or below 10-7 with only one capsule significantly exceeding this value. A maximum R/B of around 2 x 10-7 was reached at the end of the irradiation in Capsule 5. Several shakedown issues were encountered and resolved during the first three cycles. These include the repair of minor gas line leaks; repair of faulty gas line valves; the need to position moisture monitors in regions of low radiation fields for proper functioning; the enforcement of proper on-line data storage and backup, the need to monitor thermocouple performance, correcting for detector spectral gain shift, and a change in the mass flow rate range of the neon flow controllers.

  9. Testing a Novel Self-Assembling Data Paradigm in the Context of IACT Data

    E-Print Network [OSTI]

    Weinstein, Amanda; Brantseg, Thomas; Rulten, Cameron; Lutz, Robyn; Haupt, Jarvis; Elyaderani, Mojtaba Kakhodaie; Quinn, John

    2015-01-01

    The process of gathering and associating data from multiple sensors or sub-detectors due to a common physical event (the process of event-building) is used in many fields, including high-energy physics and $\\gamma$-ray astronomy. Fault tolerance in event-building is a challenging problem that increases in difficulty with higher data throughput rates and increasing numbers of sub-detectors. We draw on biological self-assembly models in the development of a novel event-building paradigm that treats each packet of data from an individual sensor or sub-detector as if it were a molecule in solution. Just as molecules are capable of forming chemical bonds, "bonds" can be defined between data packets using metadata-based discriminants. A database -- which plays the role of a beaker of solution -- continually selects pairs of assemblies at random to test for bonds, which allows single packets and small assemblies to aggregate into larger assemblies. During this process higher-quality associations supersede spurious o...

  10. Irradiation test on FD-SOI Readout ASIC of Pair-monitor

    E-Print Network [OSTI]

    Yutaro Sato; Hirokazu Ikeda; Akiya Miyamoto; Yosuke Takubo; Toshiaki Tauchi; Hitoshi Yamamoto

    2010-06-22

    We fabricated a readout ASIC with the fully depleted silicon-on-insulator (FD-SOI) technology for the pair-monitor. The pair-monitor is a silicon pixel device that measures the beam profile of the international linear collider. It utilizes the directional distribution of a large number of electron-positron pairs created by collision of bunches, and is required to tolerate radiation dose of about a few Mrad/year. The irradiation might cause the buried oxide layer of SOI to accumulate charges which interfere with intended functions. We thus performed extensive irradiation tests on the prototype ASIC, and the results are described in this paper.

  11. Status of the Norwegian thorium light water reactor (LWR) fuel development and irradiation test program

    SciTech Connect (OSTI)

    Drera, S.S.; Bjork, K.I.; Kelly, J.F.; Asphjell, O. [Thor Energy AS: Sommerrogaten 13-15, Oslo, NO255 (Norway)

    2013-07-01

    Thorium based fuels offer several benefits compared to uranium based fuels and should thus be an attractive alternative to conventional fuel types. In order for thorium based fuel to be licensed for use in current LWRs, material properties must be well known for fresh as well as irradiated fuel, and accurate prediction of fuel behavior must be possible to make for both normal operation and transient scenarios. Important parameters are known for fresh material but the behaviour of the fuel under irradiation is unknown particularly for low Th content. The irradiation campaign aims to widen the experience base to irradiated (Th,Pu)O{sub 2} fuel and (Th,U)O{sub 2} with low Th content and to confirm existing data for fresh fuel. The assumptions with respect to improved in-core fuel performance are confirmed by our preliminary irradiation test results, and our fuel manufacture trials so far indicate that both (Th,U)O{sub 2} and (Th,Pu)O{sub 2} fuels can be fabricated with existing technologies, which are possible to upscale to commercial volumes.

  12. Independent Review of AFC 2A, 2B, and 2E ATR Irradiation Tests

    SciTech Connect (OSTI)

    M. Cappiello; R. Hobbins; K. Penny; L. Walters

    2014-01-01

    As part of the Department of Energy Advanced Fuel Cycle program, a series of fuels development irradiation tests have been performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. These tests are providing excellent data for advanced fuels development. The program is focused on the transmutation of higher actinides which best can be accomplished in a sodium-cooled fast reactor. Because a fast test reactor is no longer available in the US, a special test vehicle is used to achieve near-prototypic fast reactor conditions (neutron spectra and temperature) for use in ATR (a water-cooled thermal reactor). As part of the testing program, there were many successful tests of advanced fuels including metals and ceramics. Recently however, there have been three experimental campaigns using metal fuels that experienced failure during irradiation. At the request of the program, an independent review committee was convened to review the post-test analyses performed by the fuels development team, to assess the conclusions of the team for the cause of the failures, to assess the adequacy and completeness of the analyses, to identify issues that were missed, and to make recommendations for improvements in the design and operation of future tests. Although there is some difference of opinion, the review committee largely agreed with the conclusions of the fuel development team regarding the cause of the failures. For the most part, the analyses that support the conclusions are sufficient.

  13. AGR-2: The first irradiation of French HTR fuel in Advanced Test Reactor

    SciTech Connect (OSTI)

    T. Lambert; B. Grover; P. Guillermier; D. Moulinier; F. Imbault Huart

    2012-10-01

    AGR-2, the second irradiation of the US program for qualification of the NGNP fuel, is open to international participation within the scope of the Generation IV International Forum. In this frame, it includes in its multi-capsule irradiation rig an irradiation of French HTR fuel manufactured in the CAPRI line (GAIA facility at CEA/Cadarache and AREVA/CERCA compacting line at Romans). The AGR-2 irradiation is designed to place our first fabrications of HTR particles under operating conditions that are representative of ANTARES project while keeping close to the test range of the German fuel as much as possible, which is the reference in terms of irradiation behavior. A few batches of particles and 12 fuel compacts were produced and characterized in 2009 by CEA and CERCA. The fuel main characteristics are in conformity with our specifications and in compliance with INL requirements. The AGR-2 experiment is based on the design and devices used in the first experiment of the AGR program. The design makes it possible to monitor the irradiation conditions and in particular, the temperature, the power and the fission products released from fuel particles. The in pile equipment consists of a multi-capsule device designed to simultaneously irradiate six independent capsules with temperature control. The out-of-core part consists of the equipment for actively controlling temperature and measuring the fission products release on-line. The target conditions for the irradiation experiment were defined with the aim of comparing the results obtained under irradiation with German particles along with the objectives of reaching burn-up and fluence targets to validate the behavior of our fuel in a significant range (15% FIMA – 5 × 1025 n/m2 at 600 EFPD with centerline fuel temperature about 1100 degrees C). These conditions have to be representative of ANTARES project characteristics. These target conditions were compared with final results from neutron and thermal design studies performed by INL team, and preliminary thermal mechanical ATLAS calculations were carried out by CEA from this pre-design. Despite the mean burn-up achieved in approximately 600 EFPD being a little high (16.3% FIMA max. associated with a low fluence up to 2.85 × 1025 n/m2), this irradiation will nevertheless encompass the range of irradiation effects covered in our experimental objectives (maximum stress peak at start of irradiation then sign inversion of the stress in the SiC layer). In addition, the fluence and burn-up acceleration factors are very similar to those of the German reference experiments. This experimental irradiation began in July 2010 in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) and first results have been acquired.

  14. Irradiation tests of ITER candidate Hall sensors using two types of neutron spectra

    SciTech Connect (OSTI)

    Duran, I. [Institute of Plasma Physics AS CR, v. v. i., Association EURATOM/IPP.CR, 182 00 Prague 8 (Czech Republic); Bolshakova, I.; Holyaka, R. [Magnetic Sensor Laboratory, Lviv Polytechnic National University, 790 31 Lviv (Ukraine); Viererbl, L.; Lahodova, Z. [Nuclear Research Institute plc., 250 68 Husinec-Rez (Czech Republic); Sentkerestiova, J. [Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, 115 19 Prague 1 (Czech Republic); Bem, P. [Nuclear Physics Institute AS CR, v. v. i., 250 68 Husinec-Rez (Czech Republic)

    2010-10-15

    We report on irradiation tests of InSb based Hall sensors at two irradiation facilities with two distinct types of neutron spectra. One was a fission reactor neutron spectrum with a significant presence of thermal neutrons, while another one was purely fast neutron field. Total neutron fluence of the order of 10{sup 16} cm{sup -2} was accumulated in both cases, leading to significant drop of Hall sensor sensitivity in case of fission reactor spectrum, while stable performance was observed at purely fast neutron spectrum. This finding suggests that performance of this particular type of Hall sensors is governed dominantly by transmutation. Additionally, it further stresses the need to test ITER candidate Hall sensors under neutron flux with ITER relevant spectrum.

  15. Experiments on a ceramic electrolysis cell and a palladium diffuser at the tritium systems test assembly

    SciTech Connect (OSTI)

    Konishi, Satoshi; Yoshida, Hiroshi; Ohno, Hideo; Naruse, Yuji; Coffin, D.O.; Walthers, C.R.; Binning, K.E.

    1985-01-01

    A ceramic electrolysis cell and a palladium diffuser are developed in Japan and is tested with tritium in Tritium Systems Test Assembly (TSTA) of the Los Alamos National Laboratory, in order to confirm the feasibility as possible upgrades for the fuel cleanup system (PCU). The ceramic electrolysis cell made of stabilized zirconia was operated at 630/sup 0/C for an extended period with a mixture of 3% T/sub 2/O in He carrier gas in the circulation system with oxidizing catalyst bed. The palladium diffuser was tested with circulated pure tritium gas at 280/sup 0/C to verify the compatibility of the alloy with tritium, since the /sup 3/He produced in the metal could cause a degradation. The isotopic effects were also measured for both devices.

  16. Status of the NGNP Graphite Creep Experiments AGC-1 and AGC-2 Irradiated in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six peripheral stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six peripheral stacks will have different compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during irradiation of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. This paper will briefly discuss the design of the experiment and control systems, and then present the irradiation results for each experiment to date.

  17. Results of crack-arrest tests on two irradiated high-copper welds

    SciTech Connect (OSTI)

    Iskander, S.K.; Corwin, W.R.; Nanstead, R.K. (Oak Ridge National Lab., TN (USA))

    1990-12-01

    The objective of this study was to determine the effect of neutron irradiation on the shift and shape of the lower-bound curve to crack-arrest data. Two submerged-arc welds with copper contents of 0.23 and 0.31 wt % were commercially fabricated in 220-mm-thick plate. Crack-arrest specimens fabricated from these welds were irradiated at a nominal temperature of 288{degree}C to an average fluence of 1.9 {times} 10{sup 19} neutrons/cm{sup 2} (>1 MeV). Evaluation of the results shows that the neutron-irradiation-induced crack-arrest toughness temperature shift is about the same as the Charpy V-notch impact temperature shift at the 41-J energy level. The shape of the lower-bound curves (for the range of test temperatures covered) did not seem to have been altered by irradiation compared to those of the ASME K{sub Ia} curve. 9 refs., 21 figs., 10 tabs.

  18. Facility for fast neutron irradiation tests of electronics at the ISIS spallation neutron source

    SciTech Connect (OSTI)

    Andreani, C.; Pietropaolo, A.; Salsano, A.; Gorini, G.; Tardocchi, M.; Paccagnella, A.; Gerardin, S.; Frost, C. D.; Ansell, S.; Platt, S. P.

    2008-03-17

    The VESUVIO beam line at the ISIS spallation neutron source was set up for neutron irradiation tests in the neutron energy range above 10 MeV. The neutron flux and energy spectrum were shown, in benchmark activation measurements, to provide a neutron spectrum similar to the ambient one at sea level, but with an enhancement in intensity of a factor of 10{sup 7}. Such conditions are suitable for accelerated testing of electronic components, as was demonstrated here by measurements of soft error rates in recent technology field programable gate arrays.

  19. Tritium handling experience in vacuum systems at TSTA (Tritium Systems Test Assembly)

    SciTech Connect (OSTI)

    Anderson, J.L.; Jenkins, E.M.; Walthers, C.R.; Yoshida, H.; Fukui, H.; Naruse, Y. (Los Alamos National Lab., NM (USA); Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan))

    1989-01-01

    Compound cryopumps have been added to the Tritium Systems Test Assembly (TSTA) integrated fusion fuel loop. Operations have been performed which closely simulate an actual fusion reactor pumping scenario. In addition, performance data have been taken that support the concept of using coconut charcoal as a sorbent at 4K for pumping helium. Later tests show that coconut charcoal may be used to co-pump D,T and He mixtures on a single 4K panel. Rotary spiral pumps have been used successfully in several applications at TSTA and have acquired more than 9000 hours of maintenance-free operation. Metal bellows pumps have been used to back the spiral pumps and have been relatively trouble free in loop operations. Bellows pumps also have more than 9000 hours of maintenance-free operation. 5 refs., 6 figs.

  20. Ion irradiation testing and characterization of FeCrAl candidate alloys

    SciTech Connect (OSTI)

    Anderoglu, Osman; Aydogan, Eda; Maloy, Stuart Andrew; Wang, Yongqiang

    2014-10-29

    The Fuel Cycle Research and Development program’s Advanced Fuels Campaign has initiated a multifold effort aimed at facilitating development of accident tolerant fuels. This effort involves development of fuel cladding materials that will be resistant to oxidizing environments for extended period of time such as loss of coolant accident. Ferritic FeCrAl alloys are among the promising candidates due to formation of a stable Al?O? oxide scale. In addition to being oxidation resistant, these promising alloys need to be radiation tolerant under LWR conditions (maximum dose of 10-15 dpa at 250 – 350°C). Thus, in addition to a number of commercially available alloys, nuclear grade FeCrAl alloys developed at ORNL were tested using high energy proton irradiations and subsequent characterization of irradiation hardening and damage microstructure. This report summarizes ion irradiation testing and characterization of three nuclear grade FeCrAl cladding materials developed at ORNL and four commercially available Kanthal series FeCrAl alloys in FY14 toward satisfying FCRD campaign goals.

  1. Assembly of Sub-10-nm Block Copolymer Patterns with Mixed Morphology and Period Using Electron Irradiation and Solvent Annealing

    E-Print Network [OSTI]

    Son, Jeong Gon

    Block copolymer self-assembly generates patterns with periodicity in the ?10–100 nm range and is increasingly recognized as a route to lithographic patterning beyond the resolution of photolithography. Block copolymers ...

  2. Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

    SciTech Connect (OSTI)

    M. Chen; CM Regan; D. Noe

    2006-01-09

    Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas release and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.

  3. A spallation-based irradiation test facility for fusion and future fission materials

    E-Print Network [OSTI]

    Samec, K.; Kadi, Y.; Luis, R.; Romanets, Y.; Behzad, M.; Aleksan, R.; Bousson, S.

    2014-01-01

    The EU’s FP7 TIARA program for developing accelerator-based facilities has recently demonstrated the unique capabilities of a compact and powerful spallation source for irradiating advanced nuclear materials. The spectrum and intensity of the neutron flux produced in the proposed facility fulfils the requirements of the DEMO fusion reactor for ITER, ADS reactors and also Gen III / IV reactors. Test conditions can be modulated, covering temperature from 400 to 550°C, liquid metal corrosion, cyclical or static stress up to 500 MPa and neutron/proton irradiation damage of up to 25 DPA per annum. The entire “TMIF” facility fits inside a cube 2 metres on a side, and is dimensioned for an accelerator beam power of 100 kW, thus reducing costs and offering great versatility and flexibility.

  4. Recent palladium membrane reactor development at the tritium systems test assembly

    SciTech Connect (OSTI)

    Willms, R.S.; Birdsell, S.A.; Wilhelm, R.C.

    1995-07-01

    The palladium membrane reactor (PMR) is proving to be a simple and effective means for recovering hydrogen isotopes from fusion fuel impurities such as methane and water. This device directly combines two techniques which have long been utilized for hydrogen processing, namely catalytic shift reactions and palladium/silver permeators. A proof-of-principle (PMR) has been constructed and tested at the Tritium Systems Test Assembly of Los Alamos National Laboratory. The first tests with this device showed that is was effective for the proposed purpose. Initial work concluded that a nickel catalyst was an appropriate choice for use in a PMR. More detailed testing of the PMR with such a catalyst was performed and reported in other works. It was shown that a nickel catalyst-packed PMR did, indeed, recover hydrogen from water and methane with efficiencies approaching 100% in a single processing pass. These experiments were conducted over an extended period of time and no failure or need for regeneration was encountered. These positive results have prompted further PMR development. Topics addressed include alternate PMR geometries and initial testing of the PMR with tritium. These are the subjects of this paper.

  5. Mechanical behavior of AISI 304SS determined by miniature test methods after neutron irradiation to 28 dpa

    SciTech Connect (OSTI)

    Ellen M. Rabenberg; Brian J. Jaques; Bulent H. Sencer; Frank A. Garner; Paula D. Freyer; Taira Okita; Darryl P. Butt

    2014-05-01

    The mechanical properties of AISI 304 stainless steel irradiated for over a decade in the Experimental Breeder Reactor (EBR-II) were measured using miniature mechanical testing methods. The shear punch method was used to evaluate the shear strengths of the neutron-irradiated steel and a correlation factor was empirically determined to predict its tensile strength. The strength of the stainless steel slightly decreased with increasing irradiation temperature, and significantly increased with increasing dose until it saturated above approximately 5 dpa. Ferromagnetic measurements were used to observe and deduce the effects of the stress-induced austenite to martensite transformation as a result of shear punch testing.

  6. Performance of AGR-1 High-Temperature Reactor Fuel During Post-Irradiation Heating Tests

    SciTech Connect (OSTI)

    Morris, Robert Noel [ORNL; Baldwin, Charles A [ORNL; Hunn, John D [ORNL; Demkowicz, Paul [Idaho National Laboratory (INL); Reber, Edward [Idaho National Laboratory (INL)

    2014-01-01

    The fission product retention of irradiated low-enriched uranium oxide/uranium carbide TRISO fuel compacts from the AGR-1 experiment has been evaluated at temperatures of 1600 1800 C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4 to 19.1% FIMA have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium, and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 10-6 after 300 h at 1600 C or 100 h at 1800 C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 C, and 85Kr release was very low during the tests (particles with breached SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ~210 h at 1800 C in one compact. Post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating layers to understand particle behavior.

  7. Status of the NGNP graphite creep experiments AGC-1 and AGC-2 irradiated in the advanced test reactor

    SciTech Connect (OSTI)

    S. Blaine Grover

    2014-05-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the next generation nuclear plant (NGNP) very high temperature gas reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six peripheral stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six peripheral stacks will have three different compressive loads applied to the top half of three diametrically opposite pairs of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during irradiation of the experiment.

  8. Results of crack-arrest tests on irradiated a 508 class 3 steel

    SciTech Connect (OSTI)

    Iskander, S.K.; Milella, P.P.; Pini, M.A.

    1998-02-01

    Ten crack-arrest toughness values for irradiated specimens of A 508 class 3 forging steel have been obtained. The tests were performed according to the American Society for Testing and Materials (ASTM) Standard Test Method for Determining Plane-Strain Crack-Arrest Fracture Toughness, K{sub la} of Ferritic Steels, E 1221-88. None of these values are strictly valid in all five ASTM E 1221-88 validity criteria. However, they are useful when compared to unirradiated crack-arrest specimen toughness values since they show the small (averaging approximately 10{degrees}C) shifts in the mean and lower-bound crack-arrest toughness curves. This confirms that a low copper content in ASTM A 508 class 3 forging material can be expected to result in small shifts of the transition toughness curve. The shifts due to neutron irradiation of the lower bound and mean toughness curves are approximately the same as the Charpy V-notch (CVN) 41-J temperature shift. The nine crack-arrest specimens were irradiated at temperatures varying from 243 to 280{degrees}C, and to a fluence varying from 1.7 to 2.7 x 10{sup 19} neutrons/cm{sup 2} (> 1 MeV). The test results were normalized to reference values that correspond to those of CVN specimens irradiated at 284{degrees}C to a fluence of 3.2 x 10{sup 19} neutrons/cm{sup 2} (> 1 MeV) in the same capsule as the crack-arrest specimens. This adjustment resulted in a shift to lower temperatures of all the data, and in particular moved two data points that appeared to lie close to or lower than the American Society of Mechanical Engineers K{sub la} curve to positions that seemed more reasonable with respect to the remaining data. A special fixture was designed, fabricated, and successfully used in the testing. For reasons explained in the text, special blocks to receive the Oak Ridge National Laboratory clip gage were designed, and greater-than-standard crack-mouth opening displacements measured were accounted for. 24 refs., 13 figs., 12 tabs.

  9. AGR-2 Irradiated Test Train Preliminary Inspection and Disassembly First Look

    SciTech Connect (OSTI)

    Ploger, Scott; Demkowciz, Paul; Harp, Jason

    2015-05-01

    The AGR 2 irradiation experiment began in June 2010 and was completed in October 2013. The test train was shipped to the Materials and Fuels Complex in July 2014 for post-irradiation examination (PIE). The first PIE activities included nondestructive examination of the test train, followed by disassembly of the test train and individual capsules and detailed inspection of the capsule contents, including the fuel compacts and their graphite fuel holders. Dimensional metrology was then performed on the compacts, graphite holders, and steel capsule shells. AGR 2 disassembly and metrology were performed with the same equipment used successfully on AGR 1 test train components. Gamma spectrometry of the intact test train gave a preliminary look at the condition of the interior components. No evidence of damage to compacts or graphite components was evident from the isotopic and gross gamma scans. Disassembly of the AGR 2 test train and its capsules was conducted rapidly and efficiently by employing techniques refined during the AGR 1 disassembly campaign. Only one major difficulty was encountered while separating the test train into capsules when thermocouples (of larger diameter than used in AGR 1) and gas lines jammed inside the through tubes of the upper capsules, which required new tooling for extraction. Disassembly of individual capsules was straightforward with only a few minor complications. On the whole, AGR 2 capsule structural components appeared less embrittled than their AGR 1 counterparts. Compacts from AGR 2 Capsules 2, 3, 5, and 6 were in very good condition upon removal. Only relatively minor damage or markings were visible using high resolution photographic inspection. Compact dimensional measurements indicated radial shrinkage between 0.8 to 1.7%, with the greatest shrinkage observed on Capsule 2 compacts that were irradiated at higher temperature. Length shrinkage ranged from 0.1 to 0.9%, with by far the lowest axial shrinkage on Capsule 3 compacts—possibly as a consequence of lower packing fraction or larger particle size. Differences in fast neutron fluence among compacts from these four capsules had no obvious effect on radial and axial shrinkage. (The AGR 2 experiment included Capsule 1 containing French compacts and Capsule 4 with compacts made at Oak Ridge National Laboratory using South African fuel particles. Information on these two batches of AGR 2 fuel compacts is confined to restricted Appendices A and B because of proprietary information limitations.)

  10. Assessment of the integrity of spent fuel assemblies used in dry storage demonstrations at the Nevada Test Site

    SciTech Connect (OSTI)

    Johnson, A.B. Jr.; Dobbins, J.C.; Zaloudek, F.R.

    1987-07-01

    This report summarizes the histories of 17 Zircaloy-clad spent fuel assemblies used in dry storage tests and demonstrations at the Engine Maintenance and Disassembly (EMAD) and Climax facilities at the Nevada Test Site (NTS). The 18th assembly was shipped to the Battelle Columbus Laboratory (BCL) and remained there for extensive characterization and as a source of specimens for whole-rod and rod-segment dry storage tests. The report traces the history of the assemblies after discharge from the Turkey Point Unit 3 pressurized-water reactor (1975 and 1977) through shipment (first arrival at EMAD in December 1978), dry storage tests and demonstrations, and shipment by truck cask from EMAD to the Idaho National Engineering Laboratory (INEL) in May/June 1986. The principal objectives of this report are to assess and document the integrity of the fuel during the extensive dry storage activities at NTS and BCL, and to briefly summarize the dry storage technologies and procedures demonstrated in this program. The dry storage tests and demonstrations involved the following concepts and facilities: (1) surface drywells (EMAD); (2) deep drywells (425 m underground in the Climax granite formation); (3) concrete silo (EMAD); (4) air-cooled vault (EMAD); (5) electrically-heated module for fuel assembly thermal calibration and testing (EMAD/FAITM). 20 refs., 43 figs., 9 tabs.

  11. ASSEMBLY AND TEST OF A 120 MM BORE 15 T NB3SN QUADRUPOLE FOR THE LHC UPGRADE

    E-Print Network [OSTI]

    Felice, H.

    2011-01-01

    ASSEMBLY AND TEST OF A 120 MM BORE 15 T NB SN QUADRUPOLE FORdeveloping a 1-meter long, 120 mm bore N b S n IR quadrupoleH Q is a 1-meter long 120 mm aperture cos29 quadrupole. The

  12. Assembly and Test of HD2, a 36 mm bore high field Nb3Sn Dipole Magnet

    E-Print Network [OSTI]

    Ferracin, P.

    2010-01-01

    Assembly and Test of HD2, a 36 mm bore high field Nb 3 Sna 15 T Nb 3 Sn dipole with a 35 mm bore”, IEEE Trans. Appl.Nb 3 Sn dipole magnet with a 35 mm bore”, IEEE Trans. Appl.

  13. Crack-arrest tests on two irradiated high-copper welds. Phase 2: Results of duplex-type experiments

    SciTech Connect (OSTI)

    Iskander, S.K.; Corwin, W.R.; Nanstad, R.K. [Oak Ridge National Lab., TN (United States)

    1994-03-01

    The objective of the Heavy-Section Steel Irradiation Program Sixth Irradiation Series is to determine the effect of neutron irradiation on the shift and shape of the lower-bound curve to crack-arrest toughness data. Two submerged-arc welds with copper contents of 0.23 and 0.31 wt % were commercially fabricated in 220-mm-thick plate. Crack-arrest specimens fabricated from these welds were irradiated at a nominal temperature of 288{degrees}C to an average fluence of 1.9 {times} 10{sup 19} neutrons/cm{sup 2} (>1 MeV). This is the second report giving the results of the tests on irradiated duplex-type crack-arrest specimens. A previous report gave results of tests on irradiated weld-embrittled-type specimens. Charpy V-notch (CVN) specimens irradiated in the same capsules as the crack-arrest specimens were also tested, and a 41-J transition temperature shift was determined from these specimens. {open_quotes}Mean{close_quote} curves of the same form as the American Society of Mechanical Engineers (ASME) K{sub la} curve were fit to the data with only the {open_quotes}reference temperature{close_quotes} as a parameter. The shift between the mean curves agrees well with the 41-J transition temperature shift obtained from the CVN specimen tests. Moreover, the four data points resulting from tests on the duplex crack-arrest specimens of the present study did not make a significant change to mean curve fits to either the previously obtained data or all the data combined.

  14. Self-aligning hydraulic piston assembly for tensile testing of ceramic

    DOE Patents [OSTI]

    Liu, K.C.

    1987-08-18

    The present invention is directed to a self-aligning grip housing assembly that can transmit an uniaxial load to a tensile specimen without introducing bending stresses into the specimen. Disposed inside said grip housing assembly are a multiplicity of supporting pistons connected to a common source of pressurized oil that carry equal shares of the load applied to the specimen regardless whether there is initial misalignment between the specimen load column assembly and housing axis. 4 figs.

  15. Design and Status of the NGNP Fuel Experiment AGR-3/4 Irradiated in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in November 2013. Since the purpose of this experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is significantly different from the first two experiments, though the control and monitoring systems are very similar. The purpose and design of this experiment will be discussed followed by its progress and status to date.

  16. Steep-Slope Assembly Testing of Clay and Concrete Tile With and Without Cool Pigmented Colors

    SciTech Connect (OSTI)

    Miller, William A [ORNL

    2005-11-01

    Cool color pigments and sub-tile venting of clay and concrete tile roofs significantly impact the heat flow crossing the roof deck of a steep-slope roof. Field measures for the tile roofs revealed a 70% drop in the peak heat flow crossing the deck as compared to a direct-nailed asphalt shingle roof. The Tile Roofing Institute (TRI) and its affiliate members are keenly interested in documenting the magnitude of the drop for obtaining solar reflectance credits with state and federal "cool roof" building efficiency standards. Tile roofs are direct-nailed or are attached to a deck with batten or batten and counter-batten construction. S-Misson clay and concrete tile roofs, a medium-profile concrete tile roof, and a flat slate tile roof were installed on fully nstrumented attic test assemblies. Temperature measures of the roof, deck, attic, and ceiling, heat flows, solar reflectance, thermal emittance, and the ambient weather were recorded for each of the tile roofs and also on an adjacent attic cavity covered with a conventional pigmented and directnailed asphalt shingle roof. ORNL measured the tile's underside temperature and the bulk air temperature and heat flows just underneath the tile for batten and counter-batten tile systems and compared the results to the conventional asphalt shingle.

  17. First elevated-temperature performance testing of coated particle fuel compacts from the AGR-1 irradiation experiment

    SciTech Connect (OSTI)

    Charles A. Baldwin; John D. Hunn; Robert N. Morris; Fred C. Montgomery; Chinthaka M. Silva; Paul A. Demkowicz

    2014-05-01

    In the AGR-1 irradiation experiment, 72 coated-particle fuel compacts were taken to a peak burnup of 19.5% fissions per initial metal atom with no in-pile failures. This paper discusses the first post-irradiation test of these mixed uranium oxide/uranium carbide fuel compacts at elevated temperature to examine the fuel performance under a simulated depressurized conduction cooldown event. A compact was heated for 400 h at 1600 degrees C. Release of 85Kr was monitored throughout the furnace test as an indicator of coating failure, while other fission product releases from the compact were periodically measured by capturing them on exchangeable, water-cooled deposition cups. No coating failure was detected during the furnace test, and this result was verified by subsequent electrolytic deconsolidation and acid leaching of the compact, which showed that all SiC layers were still intact. However, the deposition cups recovered significant quantities of silver, europium, and strontium. Based on comparison of calculated compact inventories at the end of irradiation versus analysis of these fission products released to the deposition cups and furnace internals, the minimum estimated fractional losses from the compact during the furnace test were 1.9 x 10-2 for silver, 1.4 x 10-3 for europium, and 1.1 x 10-5 for strontium. Other post-irradiation examination of AGR-1 compacts indicates that similar fractions of europium and silver may have already been released by the intact coated particles during irradiation, and it is therefore likely that the detected fission products released from the compact in this 1600 degrees C furnace test were from residual fission products in the matrix. Gamma analysis of coated particles deconsolidated from the compact after the heating test revealed that silver content within each particle varied considerably; a result that is probably not related to the furnace test, because it has also been observed in other as-irradiated AGR-1 compacts. X-ray imaging of selected particles was performed to examine the internal microstructure. This examination revealed variable irradiation performance of the coating layers, but sufficient statistical sampling is not yet available to identify any possible correlation to variation in individual particle fission product retention.

  18. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    SciTech Connect (OSTI)

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

    2011-03-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this review, recommendations were made with respect to what instrumentation is needed at the ATR; and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. In 2009, a report was issued documenting this instrumentation development strategy and initial progress toward accomplishing instrumentation development program objectives. This document reports progress toward implementing this strategy in 2010.

  19. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    SciTech Connect (OSTI)

    J. Rempe; D. Knudson; J. Daw; T. Unruh; B. Chase; R. Schley; J. Palmer; K. Condie

    2014-01-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support the growth of nuclear science and technology in the United States (US). By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort at the Idaho National Laboratory (INL) is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this initial review, recommendations were made with respect to what instrumentation is needed at the ATR, and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. Since 2009, annual reports have been issued to provide updates on the program strategy and the progress made on implementing the strategy. This report provides an update reflecting progress as of January 2014.

  20. Hardware assembly and prototype testing for the development of a dedicated liquefied propane gas ultra low emission vehicle

    SciTech Connect (OSTI)

    1995-07-01

    On February 3, 1994, IMPCO Technologies, Inc. started the development of a dedicated LPG Ultra Low Emissions Vehicle (ULEV) under contract to the Midwest Research Institute National Renewable Energy Laboratory Division (NREL). The objective was to develop a dedicated propane vehicle that would meet or exceed the California ULEV emissions standards. The project is broken into four phases to be performed over a two year period. The four phases of the project include: (Phase 1) system design, (Phase 2) prototype hardware assembly and testing, (Phase 3) full-scale systems testing and integration, (Phase 4) vehicle demonstration. This report describes the approach taken for the development of the vehicle and the work performed through the completion of Phase II dynamometer test results. Work was started on Phase 2 (Hardware Assembly and Prototype Testing) in May 1994 prior to completion of Phase 1 to ensure that long lead items would be available in a timely fashion for the Phase 2 work. In addition, the construction and testing of the interim electronic control module (ECM), which was used to test components, was begun prior to the formal start of Phase 2. This was done so that the shortened revised schedule for the project (24 months) could be met. In this report, a brief summary of the activities of each combined Phase 1 and 2 tasks will be presented, as well as project management activities. A technical review of the system is also given, along with test results and analysis. During the course of Phase 2 activities, IMPCO staff also had the opportunity to conduct cold start performance tests of the injectors. The additional test data was most positive and will be briefly summarized in this report.

  1. The 1993 baseline biological studies and proposed monitoring plan for the Device Assembly Facility at the Nevada Test Site

    SciTech Connect (OSTI)

    Woodward, B.D.; Hunter, R.B.; Greger, P.D.; Saethre, M.B.

    1995-02-01

    This report contains baseline data and recommendations for future monitoring of plants and animals near the new Device Assembly Facility (DAF) on the Nevada Test Site (NTS). The facility is a large structure designed for safely assembling nuclear weapons. Baseline data was collected in 1993, prior to the scheduled beginning of DAF operations in early 1995. Studies were not performed prior to construction and part of the task of monitoring operational effects will be to distinguish those effects from the extensive disturbance effects resulting from construction. Baseline information on species abundances and distributions was collected on ephemeral and perennial plants, mammals, reptiles, and birds in the desert ecosystems within three kilometers (km) of the DAF. Particular attention was paid to effects of selected disturbances, such as the paved road, sewage pond, and the flood-control dike, associated with the facility. Radiological monitoring of areas surrounding the DAF is not included in this report.

  2. Assessment of Initial Test Conditions for Experiments to Assess Irradiation Assisted Stress Corrosion Cracking Mechanisms

    Broader source: Energy.gov [DOE]

    Irradiation-assisted stress corrosion cracking is a key materials degradation issue in today’s nuclear power reactor fleet and affects critical structural components within the reactor core. The...

  3. THE MECHANICAL AND SHIELDING DESIGN OF A PORTABLE SPECTROMETER AND BEAM DUMP ASSEMBLY AT BNLS ACCELERATOR TEST FACILITY.

    SciTech Connect (OSTI)

    HU,J.P.; CASEY,W.R.; HARDER,D.A.; PJEROV,S.; RAKOWSKY,G.; SKARITKA,J.R.

    2002-09-05

    A portable assembly containing a vertical-bend dipole magnet has been designed and installed immediately down-beam of the Compton electron-laser interaction chamber on beamline 1 of the Accelerator Test Facility (ATF) at Brookhaven National Laboratory (BNL). The water-cooled magnet designed with field strength of up to 0.7 Tesla will be used as a spectrometer in the Thompson scattering and vacuum acceleration experiments, where field-dependent electron scattering, beam focusing and energy spread will be analyzed. This magnet will deflect the ATF's 60 MeV electron-beam 90{sup o} downward, as a vertical beam dump for the Compton scattering experiment. The dipole magnet assembly is portable, and can be relocated to other beamlines at the ATF or other accelerator facilities to be used as a spectrometer or a beam dump. The mechanical and shielding calculations are presented in this paper. The structural rigidity and stability of the assembly were studied. A square lead shield surrounding the assembly's Faraday Cup was designed to attenuate the radiation emerging from the 1 inch-copper beam stop. All photons produced were assumed to be sufficiently energetic to generate photoneutrons. A safety evaluation of groundwater tritium contamination due to the thermal neutron capturing by the deuterium in water was performed, using updated Monte Carlo neutron-photon coupled transport code (MCNP). High-energy neutron spallation, which is a potential source to directly generate radioactive tritium and sodium-22 in soil, was conservatively assessed in verifying personal and environmental safety.

  4. High-Heat Flux Testing of Irradiated Tungsten based Materials for Fusion Applications using Infrared Plasma Arc Lamps

    SciTech Connect (OSTI)

    Sabau, Adrian S; Ohriner, Evan Keith; Kiggans Jr, James O; Schaich, Charles Ross; Ueda, Yoshio; Harper, David C; Katoh, Yutai; Snead, Lance Lewis; Byun, Thak Sang

    2014-01-01

    Testing of advanced materials and component mock-ups under prototypical fusion high-heat flux conditions, while historically a mainstay of fusion research has proved challenging, especially for irradiated materials. A new high-heat flux testing facility based on water-wall Plasma Arc Lamps (PALs) is now being used for materials and small component testing. Two PAL systems, utilizing a 12,000 C plasma arc contained in a quartz tube cooled by a spiral water flow over the inside tube surface, are currently in use. The first PAL system provides a maximum incident heat flux of 4.2 MW/m2 over an area of 9x12 cm2. The second PAL available at ORNL provides a maximum incident heat flux of 27 MW/m2 over an area of 1x10 cm2. The absorbed heat fluxes into a tungsten target for the two PALs are approximately 1.97 and 12.7 MW/m2, respectively. This paper will present the overall design of the new PAL facilities as well as the design and implementation of the Irradiated Material Target Station (IMTS). The IMTS is primarily designed for testing the effects of heat flux or thermal cycling on material coupons of interested, such as those for plasma facing components. Moreover, IMTS designs are underway to extend the testing of small mock-ups for assessing the combined heating and thermomechanical effects of cooled, irradiated components. For the testing of material coupons , the specimens are placed in a shallow recess within the molybdenum holder that is attached to a water-cooled copper alloy rod. As the measurement of the specimen temperature for PAL is historically challenging since traditional approaches of temperature measurement cannot be employed due to the infrared heating and proximity of the PAL reflector to the specimen that does not allow a direct line of site, experiments for temperature calibration are presented. Finally, results for the high-heat flux testing of tungsten-based materials using the PAL are presented. As a demonstration of the system, results will be shown of thermal fatigue and high-heat flux testing of tungsten coupon specimens that were neutron irradiated in the HFIR reactor to neutron dose consistent to ITER lifetime.

  5. Facility for high heat flux testing of irradiated fusion materials and components using infrared plasma arc lamps

    SciTech Connect (OSTI)

    Sabau, Adrian S; Ohriner, Evan Keith; Kiggans, Jim; Harper, David C; Snead, Lance Lewis; Schaich, Charles Ross

    2014-01-01

    A new high-heat flux testing facility using water-wall stabilized high-power high-pressure argon Plasma Arc Lamps (PALs) has been developed for fusion applications. It can handle irradiated plasma facing component materials and mock-up divertor components. Two PALs currently available at ORNL can provide maximum incident heat fluxes of 4.2 and 27 MW/m2 over a heated area of 9x12 and 1x10 cm2, respectively, which are fusion-prototypical steady state heat flux conditions. The facility will be described and the main differences between the photon-based high-heat flux testing facilities, such as PALs, and the e-beam and particle beam facilities more commonly used for fusion HHF testing are discussed. The components of the test chamber were designed to accommodate radiation safety and materials compatibility requirements posed by high-temperature exposure of low levels irradiated tungsten articles. Issues related to the operation and temperature measurements during testing are presented and discussed.

  6. Irradiation Tests and Expected Performance of Readout Electronics of the ATLAS Hadronic Endcap Calorimeter for the HL-LHC

    E-Print Network [OSTI]

    Martin Nagel

    2013-09-03

    The readout electronics of the ATLAS Hadronic Endcap Calorimeter (HEC) will have to withstand an about 3-5 times larger radiation environment at the future high-luminosity LHC (HLLHC) compared to their design values. The preamplifier and summing boards (PSBs), which are equipped with GaAs ASICs and comprise the heart of the readout electronics, were irradiated with neutrons and protons with fluences surpassing several times ten years of operation of the HL-LHC. Neutron tests were performed at the NPI in Rez, Czech Republic, where a 36 MeV proton beam was directed on a thick heavy water target to produce neutrons. The proton irradiation was done with 200 MeV protons at the PROSCAN area of the Proton Irradiation Facility at the PSI in Villigen, Switzerland. In-situ measurements of S-parameters in both tests allow the evaluation of frequency dependent performance parameters, like gain and input impedance, as a function of fluence. The linearity of the ASIC response was measured directly in the neutron tests with a triangular input pulse of varying amplitude. The results obtained allow an estimation of the expected performance degradation of the HEC. For a possible replacement of the PSB chips, alternative technologies were investigated and exposed to similar neutron radiation levels. In particular, IHP 250 nm Si CMOS technology has turned out to show good performance and match the specifications required. The performance measurements of the current PSB devices, the expected performance degradations under HL-LHC conditions, and results from alternative technologies will be presented.

  7. Design and Nuclear-Safety Related Simulations of Bare-Pellet Test Irradiations for the Production of Pu-238 in the High Flux Isotope Reactor using COMSOL

    SciTech Connect (OSTI)

    Freels, James D; Jain, Prashant K; Hobbs, Randy W

    2012-01-01

    The Oak Ridge National Laboratory (ORNL)is developing technology to produce plutonium-238 for the National Aeronautics and Space Administration (NASA) as a power source material for powering vehicles while in deep-space[1]. The High Flux Isotope Reactor (HFIR) of ORNL has been utilized to perform test irradiations of incapsulated neptunium oxide (NpO2) and aluminum powder bare pellets for purposes of understanding the performance of the pellets during irradiation[2]. Post irradiation examinations (PIE) are currently underway to assess the effect of temperature, thermal expansion, swelling due to gas production, fission products, and other phenomena

  8. Simulated dry storage test of a spent PWR nuclear fuel assembly in air

    SciTech Connect (OSTI)

    Johnson, A.B. Jr.; Gilbert, E.R.; Oden, D.R.; Stidham, D.L.; Garnier, J.E.; Weeks, D.L.; Dobbins, J.C.

    1985-02-01

    The purpose of the dry storage test was to investigate the behavior of Zircaloy-clad spent fuel in air between 200 and 275/sup 0/C. Atmospheric air was used for the cover gas because of the interest in establishing regimes where air inleakage into an initially inert system would not cause potential fuel degradation. Samples of the cover gas atmosphere were extracted monthly to determine fission gas concentrations as a function of time. The oxygen concentration was monitored to detect oxygen depletion, which would signal oxidation of the fuel. The gas analyses indicated very low but detectable levels of /sup 85/Kr during the first month of the test. A large increase (five orders of magnitude) in /sup 85/Kr and the appearance of helium in the cover gas indicated that a fuel rod had breached during the second month of the test. Stress rupture calculations showed that the stresses and temperatures were too low to expect breaches to form in defect-free cladding. It is theorized that the breach occurred in a fuel rod weakened by an existing cladding or end cap defect. Calculations based on the rate of /sup 85/Kr release suggest that the diameter of the initial breach was about 25 microns. A post-test fuel examination will be performed to locate and investigate the cause of the cladding breach and to determine if detectable fuel degradation progressed after the breach occurred. The post-test evaluation will define the consequences of a fuel rod breach occurring in an air cover gas at 270/sup 0/C, followed by subsequent exposure to air at a prototypic descending temperature.

  9. Finishing, assembling, and testing the empire, demand limiting meter-style A-Ottawa, Kansas

    E-Print Network [OSTI]

    Tait, Ralph Shannon

    1922-01-01

    ZZj Y F/GURE 23. T/?AY EOR CAKRY/MG YQSSEFIO&T-SRO F7ETERG. 46 At stations 5 the leakage tests were madej which showed the leakage in Cubic Feet per hour past the valve and disc. This leakage varied from perhaps ten to twenty cubic feet per... hour. Table 3 shows the leakage past various sizes of valves and discs. The leakage tester was told what capacity to make the devices. From a table similar to the one given below he filled in on the record slip the number and size of oufice Holes...

  10. Status of steady-state irradiation testing of mixed-carbide fuel designs. [LMFBR

    SciTech Connect (OSTI)

    Harry, G.R.

    1983-01-01

    The steady-state irradiation program of mixed-carbide fuels has demonstrated clearly the ability of carbide fuel pins to attain peak burnup greater than 12 at.% and peak fluences of 1.4 x 10/sup 23/ n/cm/sup 2/ (E > 0.1 MeV). Helium-bonded fuel pins in 316SS cladding have achieved peak burnups of 20.7 at.% (192 MWd/kg), and no breaches have occurred in pins of this design. Sodium-bonded fuel pins in 316SS cladding have achieved peak burnups of 15.8 at.% (146 MWd/kg). Breaches have occurred in helium-bonded fuel pins in PE-16 cladding (approx. 5 at.% burnup) and in D21 cladding (approx. 4 at.% burnup). Sodium-bonded fuel pins achieved burnups over 11 at.% in PE-16 cladding and over 6 at.% in D9 and D21 cladding.

  11. Testing of Performance of Optical Fibers Under Irradiation in Intense Radiation Fields, When Subjected to Very High Temperatures

    SciTech Connect (OSTI)

    Blue, Thomas; Windl, Wolfgang; Dickerson, Bryan

    2013-01-03

    The primary objective of this project is to measure and model the performance of optical fibers in intense radiation fields when subjected to very high temperatures. This research will pave the way for fiber optic and optically based sensors under conditions expected in future high-temperature gas-cooled reactors. Sensor life and signal-to-noise ratios are susceptible to attenuation of the light signal due to scattering and absorbance in the fibers. This project will provide an experimental and theoretical study of the darkening of optical fibers in high-radiation and high-temperature environments. Although optical fibers have been studied for moderate radiation fluence and flux levels, the results of irradiation at very high temperatures have not been published for extended in-core exposures. Several previous multi-scale modeling efforts have studied irradiation effects on the mechanical properties of materials. However, model-based prediction of irradiation-induced changes in silica�s optical transport properties has only recently started to receive attention due to possible applications as optical transmission components in fusion reactors. Nearly all damage-modeling studies have been performed in the molecular-dynamics domain, limited to very short times and small systems. Extended-time modeling, however, is crucial to predicting the long-term effects of irradiation at high temperatures, since the experimental testing may not encompass the displacement rate that the fibers will encounter if they are deployed in the VHTR. The project team will pursue such extended-time modeling, including the effects of the ambient and recrystallization. The process will be based on kinetic MC modeling using the concept of amorphous material consisting of building blocks of defect-pairs or clusters, which has been successfully applied to kinetic modeling in amorphized and recrystallized silicon. Using this procedure, the team will model compensation for rate effects, and the interplay of rate effects with the effects of annealing, to accurately predict the fibers� reliability and expected lifetime

  12. Recent palladium membrane reactor development at the tritium systems test assembly

    SciTech Connect (OSTI)

    Scott, W.R.; Birdsell, S.A.; Wilhelm, R.C. [Los Alamos National Lab., NM (United States)

    1995-10-01

    The palladium membrane reactor (PMR) is being investigated as a means for recovering hydrogen isotopes (including tritium) from compounds such as water and methane. Previous work with protiated water and methane showed that this device can be used to obtain high hydrogen recovery efficiencies using a single processing pass and with essentially no waste production. With these successful proof-of-principle results completed, recent work has focused on PMR development. This included studies of various geometries and testing with tritium. The results, which are reported here, have led to a better understanding of the PMR and will lead to the ultimate goal of building a production PMR and putting it into practical tritium processing service. 3 refs., 5 figs., 1 tab.

  13. Micro-Pocket Fission Detectors (MPFD) For Fuel Assembly Analysis

    SciTech Connect (OSTI)

    Troy Unruh; Michael Reichenberger; Phillip Ugorowski

    2013-09-01

    Neutron sensors capable of real-time measurement of thermal flux, fast flux, and temperature in a single miniaturized probe are needed in irradiation tests required to demonstrate the performance of candidate new fuels, and cladding materials. In-core ceramic-based miniature neutron detectors or “Micro-Pocket Fission Detectors” (MPFDs) have been studied at Kansas State University (KSU). The first MPFD prototypes were tested in various neutron fields at the KSU TRIGA research reactor with successful results. Currently, a United States Department of Energy-sponsored joint KSU/Idaho National Laboratory (INL) effort is underway to develop a high-temperature, high-pressure version of the MPFD using radiation-resistant, high temperature materials, which would be capable of withstanding irradiation test conditions in high performance material and test reactors (MTRs). Ultimately, this more compact, more accurate, and longer lifetime flux sensor for critical mock-ups, existing and advanced reactor designs, high performance MTRs, and transient test reactors has the potential to lead to higher accuracy and resolution data from irradiation testing, more detailed core flux measurements and enhanced fuel assembly processing. Prior evaluations by KSU indicate that these sensors could also be used to monitor burn-up of nuclear fuel. If integrated into nuclear fuel assemblies, MPFDs offer several advantages to current spent fuel management systems.

  14. High speed door assembly

    DOE Patents [OSTI]

    Shapiro, C.

    1993-04-27

    A high speed door assembly is described, comprising an actuator cylinder and piston rods, a pressure supply cylinder and fittings, an electrically detonated explosive bolt, a honeycomb structured door, a honeycomb structured decelerator, and a structural steel frame encasing the assembly to close over a 3 foot diameter opening within 50 milliseconds of actuation, to contain hazardous materials and vapors within a test fixture.

  15. High speed door assembly

    DOE Patents [OSTI]

    Shapiro, Carolyn (Idaho Falls, ID)

    1993-01-01

    A high speed door assembly, comprising an actuator cylinder and piston rods, a pressure supply cylinder and fittings, an electrically detonated explosive bolt, a honeycomb structured door, a honeycomb structured decelerator, and a structural steel frame encasing the assembly to close over a 3 foot diameter opening within 50 milliseconds of actuation, to contain hazardous materials and vapors within a test fixture.

  16. Irradiation of insulators for Workshop on Accelerator Magnet, Superconductor, Design and Optimization

    E-Print Network [OSTI]

    McDonald, Kirk

    Irradiation of insulators for EuCARD Workshop on Accelerator Magnet, Superconductor, Design #12;Outline WAMSDO ­ 14.11.2011 CERN · Motivation of launching EUCARD irradiation task · Irradiation methodology · Post irradiation tests ­ Electrical ­ Thermal ­ Mechanical · Irradiation cryostat · Conclusions

  17. ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments

    SciTech Connect (OSTI)

    G. Palmiotti

    2011-12-01

    The ENDF/B-VII.1 library is the latest revision to the United States' Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 418 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [1]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Project's International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unmoderated and uranium reflected 235U and 239Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also confirmed for selected actinide reaction rates such as 236U capture. Other deficiencies, such as the overprediction of Pu solution system critical eigenvalues and a decreasing trend in calculated eigenvalue for 233U fueled systems as a function of Above-Thermal Fission Fraction remain. The comprehensive nature of this critical benchmark suite and the generally accurate calculated eigenvalues obtained with ENDF/B-VII.1 neutron cross sections support the conclusion that this is the most accurate general purpose ENDF/B cross section library yet released to the technical community.

  18. Monte Carlo testing of new cross section data sets for thermal and intermediate highly enriched uranium critical assemblies

    SciTech Connect (OSTI)

    Weinman, J.P. [Lockheed Martin Corp., Schenectady, NY (United States)

    1998-06-01

    The purpose of this study is to investigate the eigenvalue sensitivity to new {sup 235}U, hydrogen, and oxygen cross section data sets by comparing RACER Monte Carlo calculations for several thermal and intermediate spectrum critical experiments. The new {sup 235}U library (Version 107) was derived by L. Leal and H. Derrien by fitting differential experimental data for {sup 235}U while constraining the fit to match experimental capture and fission resonance integrals and Maxwellian averaged thermal K1 (v fission minus absorption). The new hydrogen library (Version 45) consists of the ENDF/B-VI release 3 data with a 332.0 mb 2,200 m/s cross section which replaces the value of 332.6 mb in the current library. The new oxygen library (Version 39) is based on a recent evaluation of {sup 16}O by E. Caro. Nineteen Oak Ridge and Rocky Flats thermal solution benchmark critical assemblies that span a range of hydrogen-to-{sup 235}U (H/U) concentrations (2,052 to 27.1) and above-thermal neutron leakage fractions (0.555 to 0.011) were analyzed. In addition, three intermediate spectrum critical assemblies (UH3-UR, UH3-NI, and HISS-HUG) were studied.

  19. IRRADIATION EXPERIMENTS &

    E-Print Network [OSTI]

    McDonald, Kirk

    IRRADIATION EXPERIMENTS & FACILITIES AT BNL: BLIP & NSLS II Peter Wanderer Superconducting MagnetV). Simultaneous irradiation and isotope production by increasing beam energy, placing irradiation target ahead of isotope production target · 117 Mev 140, 160, 180, 202 MeV · Irradiation at room temperature Alternative

  20. International Fusion Materials Irradiation Facility injector acceptance tests at CEA/Saclay: 140 mA/100 keV deuteron beam characterization

    SciTech Connect (OSTI)

    Gobin, R., E-mail: rjgobin@cea.fr; Bogard, D.; Chauvin, N.; Chel, S.; Delferrière, O.; Harrault, F.; Mattei, P.; Senée, F. [Commissariat à l’Energie Atomique et aux Energies Alternatives, CEA/Saclay, DSM/IRFU, 91191-Gif/Yvette (France)] [Commissariat à l’Energie Atomique et aux Energies Alternatives, CEA/Saclay, DSM/IRFU, 91191-Gif/Yvette (France); Cara, P. [Fusion for Energy, BFD Department, Garching (Germany)] [Fusion for Energy, BFD Department, Garching (Germany); Mosnier, A. [Commissariat à l’Energie Atomique et aux Energies Alternatives, CEA/Saclay, DSM/IRFU, 91191-Gif/Yvette (France) [Commissariat à l’Energie Atomique et aux Energies Alternatives, CEA/Saclay, DSM/IRFU, 91191-Gif/Yvette (France); Fusion for Energy, BFD Department, Garching (Germany); Shidara, H. [IFMIF/EVEDA Project Team, Obuchi-Omotedate 2-166, Rokkasho, Aomori (Japan)] [IFMIF/EVEDA Project Team, Obuchi-Omotedate 2-166, Rokkasho, Aomori (Japan); Okumura, Y. [JAEA, Division of Rokkasho BA Project, Obuchi-Omotedate 2-166, Rokkasho, Aomori (Japan)] [JAEA, Division of Rokkasho BA Project, Obuchi-Omotedate 2-166, Rokkasho, Aomori (Japan)

    2014-02-15

    In the framework of the ITER broader approach, the International Fusion Materials Irradiation Facility (IFMIF) deuteron accelerator (2 × 125 mA at 40 MeV) is an irradiation tool dedicated to high neutron flux production for future nuclear plant material studies. During the validation phase, the Linear IFMIF Prototype Accelerator (LIPAc) machine will be tested on the Rokkasho site in Japan. This demonstrator aims to produce 125 mA/9 MeV deuteron beam. Involved in the LIPAc project for several years, specialists from CEA/Saclay designed the injector based on a SILHI type ECR source operating at 2.45 GHz and a 2 solenoid low energy beam line to produce such high intensity beam. The whole injector, equipped with its dedicated diagnostics, has been then installed and tested on the Saclay site. Before shipment from Europe to Japan, acceptance tests have been performed in November 2012 with 100 keV deuteron beam and intensity as high as 140 mA in continuous and pulsed mode. In this paper, the emittance measurements done for different duty cycles and different beam intensities will be presented as well as beam species fraction analysis. Then the reinstallation in Japan and commissioning plan on site will be reported.

  1. Methods for fatigue testing off-road bicycle handlebars based on assembly effects using two different stem designs

    E-Print Network [OSTI]

    McKenna, Sean P; Hill, Michael R; Hull, Maury L

    2003-01-01

    AND EVALUATION [2] “Trek Bicycle Corp. Announces Recall ofSafety Requirements for Bicycles, International Organizationfor Fatigue Testing Off-Road Bicycle Handlebars Based on

  2. Streamlined Approach for Environmental Restoration Plan for Corrective Action Unit 113: Reactor Maintenance, Assembly, and Disassembly Building Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    J. L. Smith

    2001-01-01

    This Streamlined Approach for Environmental Restoration (SAFER) Plan addresses the action necessary for the closure in place of Corrective Action Unit (CAU) 113 Area 25 Reactor Maintenance, Assembly, and Disassembly Facility (R-MAD). CAU 113 is currently listed in Appendix III of the Federal Facility Agreement and Consent Order (FFACO) (NDEP, 1996). The CAU is located in Area 25 of the Nevada Test Site (NTS) and consists of Corrective Action Site (CAS) 25-04-01, R-MAD Facility (Figures 1-2). This plan provides the methodology for closure in place of CAU 113. The site contains radiologically impacted and hazardous material. Based on preassessment field work, there is sufficient process knowledge to close in place CAU 113 using the SAFER process. At a future date when funding becomes available, the R-MAD Building (25-3110) will be demolished and inaccessible radiologic waste will be properly disposed in the Area 3 Radiological Waste Management Site (RWMS).

  3. Irradiation Creep in Graphite

    SciTech Connect (OSTI)

    Ubic, Rick; Butt, Darryl; Windes, William

    2014-03-13

    An understanding of the underlying mechanisms of irradiation creep in graphite material is required to correctly interpret experimental data, explain micromechanical modeling results, and predict whole-core behavior. This project will focus on experimental microscopic data to demonstrate the mechanism of irradiation creep. High-resolution transmission electron microscopy should be able to image both the dislocations in graphite and the irradiation-induced interstitial clusters that pin those dislocations. The team will first prepare and characterize nanoscale samples of virgin nuclear graphite in a transmission electron microscope. Additional samples will be irradiated to varying degrees at the Advanced Test Reactor (ATR) facility and similarly characterized. Researchers will record microstructures and crystal defects and suggest a mechanism for irradiation creep based on the results. In addition, the purchase of a tensile holder for a transmission electron microscope will allow, for the first time, in situ observation of creep behavior on the microstructure and crystallographic defects.

  4. Impacts Analyses Supporting the National Environmental Policy Act Environmental Assessment for the Resumption of Transient Testing Program

    SciTech Connect (OSTI)

    Annette L. Schafer; Lloyd C. Brown; David C. Carathers; Boyd D. Christensen; James J. Dahl; Mark L. Miller; Cathy Ottinger Farnum; Steven Peterson; A. Jeffrey Sondrup; Peter V. Subaiya; Daniel M. Wachs; Ruth F. Weiner

    2013-11-01

    Environmental and health impacts are presented for activities associated with transient testing of nuclear fuel and material using two candidate test reactors. Transient testing involves irradiation of nuclear fuel or materials for short time-periods under high neutron flux rates. The transient testing process includes transportation of nuclear fuel or materials inside a robust shipping cask to a hot cell, removal from the shipping cask, pre-irradiation examination of the nuclear materials, assembly of an experiment assembly, transportation of the experiment assembly to the test reactor, irradiation in the test reactor, transport back to the hot cell, and post-irradiation examination of the nuclear fuel or material. The potential for environmental or health consequences during the transportation, examination, and irradiation actions are assessed for normal operations, off-normal (accident) scenarios, and transportation. Impacts to the environment (air, soil, and groundwater), are assessed during each phase of the transient testing process. This report documents the evaluation of potential consequences to the general public. This document supports the Environmental Assessment (EA) required by the U.S. National Environmental Policy Act (NEPA) (42 USC Subsection 4321 et seq.).

  5. Simulated Irradiation of Samples in HFIR for use as Possible Test Materials in the MPEX (Material Plasma Exposure Experiment) Facility

    SciTech Connect (OSTI)

    Ellis, Ronald James; Rapp, Juergen

    2014-01-01

    The importance of Plasma Material Interaction (PMI) is a major concern in fusion reactor design and analysis. The Material-Plasma Exposure eXperiment (MPEX) facility will explore PMI under fusion reactor plasma conditions. Samples with accumulated displacements per atom (DPA) damage produced by irradiations in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) will be studied in the MPEX facility. The project presented in this paper involved performing assessments of the induced radioactivity and resulting radiation fields of a variety of potential fusion reactor materials. The scientific code packages MCNP and SCALE were used to simulate irradiation of the samples in HFIR; generation and depletion of nuclides in the material and the subsequent composition, activity levels, gamma radiation fields, and resultant dose rates as a function of cooling time. These state-of-the-art simulation methods were used in addressing the challenge of the MPEX project to minimize the radioactive inventory in the preparation of the samples for inclusion in the MPEX facility.

  6. DISSOLUTION OF IRRADIATED MURR FUEL ASSEMBLIES

    SciTech Connect (OSTI)

    Kyser, E.

    2010-06-17

    A literature survey on the dissolution of spent nuclear fuel from the University of Missouri Research Reactor (MURR) has been performed. This survey encompassed both internal and external literature sources for the dissolution of aluminum-clad uranium alloy fuels. The most limiting aspect of dissolution in the current facility configuration involves issues related to the control of the flammability of the off-gas from this process. The primary conclusion of this work is that based on past dissolution of this fuel in H-Canyon, four bundles of this fuel (initial charge) may be safely dissolved in a nitric acid flowsheet catalyzed with 0.002 M mercuric nitrate using a 40 scfm purge to control off-gas flammability. The initial charge may be followed by a second charge of up to five bundles to the same dissolver batch depending on volume and concentration constraints. The safety of this flowsheet relies on composite lower flammability limits (LFL) estimated from prior literature, pilot-scale work on the dissolution of site fuels, and the proposed processing flowsheet. Equipment modifications or improved LFL data offer the potential for improved processing rates. The fuel charging sequence, as well as the acid and catalyst concentrations, will control the dissolution rate during the initial portion of the cycle. These parameters directly impact the hydrogen and off-gas generation and, along with the purge flowrate determine the number of bundles that may be charged. The calculation approach within provides Engineering a means to determine optimal charging patterns. Downstream processing of this material should be similar to that of recent processing of site fuels requiring only minor adjustments of the existing flowsheet parameters.

  7. 1. Large Scale Climate Simulator (Building 3144) The LSCS tests roof and/or attic assemblies weighing up to

    E-Print Network [OSTI]

    Oak Ridge National Laboratory

    Envelope 1. Large Scale Climate Simulator (Building 3144) The LSCS tests roof and/or attic accommodates specimens of up to 606020 cm 24248 in.). 5. Roof Thermal Research Apparatus (Apparatus 3138) #12;RTRA measures effects of long-term natural weather exposure on small, low-slope roof specimens

  8. Crash hit frequency analysis of aircraft overflights of the Nevada Test Site (NTS) and the Device Assembly Facility (DAF)

    SciTech Connect (OSTI)

    Kimura, C Y; Sanzo, D L; Sharirli, M

    1998-07-09

    Aircraft crashes are an element of external events required to be analyzed and documented in Facility Safety Analysis Reports (SARs) and Nuclear Explosive Safety Studies (NESS). Aircraft crashes into DOE facilities are of concern due to effects related to impact and fire that can potentially lead to penetration of the facility, disruption of operations, and the potential of release of radioactive and/or hazardous materials subsequent to the aircraft impact. Recent changes in the control of the airspace were not considered in previous safety studies of aircraft flights over the NTS [Refs. 4,5,6]. The Airspace changes have warranted review of the effects of the issued MOU on the Device Assembly Facility (DAF) Authorization Basis Documents [Refs. 4,5], the underlying analysis assumptions, and results relevant to aircraft crash. This report documents the review and analysis of aircraft crash hit frequency on the DAF within NTS. It focuses on the impact of airspace changes based on the MOU. The frequency of an aircraft crashing and hitting the DAF is in the 1 E-7 to E-8 range. While this is considered to be acceptably small, it should not be considered an upper bound. This conclusion should not be interpreted to mean that no further work need be done. The results of the analysis are highly dependent on the assumptions made and the available data. There is considerable uncertainty in the number of overflights which are taking place over the NTS and restricted airspace R-4808N. To reduce this uncertainty, additional follow-on work should be done to activate the monitor in the CP at NTS which is to receive information from the Nellis Range control station, to monitor the level of air activity in R-4808N and to recalculate the aircraft crash hit frequency on the DAF when better overflight estimates are obtained. Finally, to reduce the human error component, the process by which the DOE notifies the USAF of ?no-fly? periods for R-4808N during which SNM is present in the DAF should be formalized either by modifying an existing operational procedure or instituting a new procedure. Additionally, the process for DOE notification by DAF personnel of the presence of SNM should be formalized.

  9. Crash hit frequency analysis of aircraft overflights of the Nevada Test Site (NTS) and the Device Assembly Facility (DAF)

    SciTech Connect (OSTI)

    Kimura, C. Y.; Sanzo, D. L.; Sharirli, M.

    1998-12-16

    Aircraft crashes are an element of external events required to be analyzed and documented in Facility Safety Analysis Reports (SARs) and Nuclear Explosive Safety Studies (NESS). Aircraft crashes into DOE facilities are of concern due to effects related to impact and fire that can potentially lead to penetration of the facility, disruption of operations, and the potential of release of radioactive and/or hazardous materials subsequent to the aircraft impact. Recent changes in the control of the airspace were not considered in previous safety studies of aircraft flights over the NTS [Refs. 4,5,6]. The Airspace changes have warranted review of the effects of the issued MOU on the Device Assembly Facility (DAF) Authorization Basis Documents [Refs. 4,5], the underlying analysis assumptions, and results relevant to aircraft crash. This report documents the review and analysis of aircraft crash hit frequency on the DAF within NTS. It focuses on the impact of airspace changes based on the MOU. The frequency of an aircraft crashing and hitting the DAF is in the 1 E-7 to E-8 range. While this is considered to be acceptably small, it should not be considered an upper bound. This conclusion should not be interpreted to mean that no further work need be done. The results of the analysis are highly dependent on the assumptions made and the available data. There is considerable uncertainty in the number of overflights which are taking place over the NTS and restricted airspace R-4808N. To reduce this uncertainty, additional follow-on work should be done to activate the monitor in the CP at NTS which is to receive information from the Nellis Range control station, to monitor the level of air activity in R-4808N and to recalculate the aircraft crash hit frequency on the DAF when better overflight estimates are obtained. Finally, to reduce the human error component, the process by which the DOE notifies the USAF of ?no-fly? periods for R-4808N during which SNM is present in the DAF should be formalized either by modifying an existing operational procedure or instituting a new procedure. Additionally, the process for DOE notification by DAF personnel of the presence of SNM should be formalized.

  10. Structured Penumbral Irradiance Computation George Drettakis, Eugene Fiume

    E-Print Network [OSTI]

    Toronto, University of

    Structured Penumbral Irradiance Computation George Drettakis, Eugene Fiume Abstract-- A definitive understanding of irradiance behavior in penumbral regions has been hard to come by, mainly due irradiance varies non-monotonically are characterized empirically, and numerical tests are performed

  11. Design, assembly, and testing of a high-resolution relay lens used for holography with operation at both doubled and tripled Nd:YAG laser wavelengths

    SciTech Connect (OSTI)

    Sorenson, Danny S [Los Alamos National Laboratory; Pazuchanics, Peter D [Los Alamos National Laboratory; Malone, Robert M [NSTEC; Cox, Brian C [NSTEC; Frogget, Brent C [NSTEC; Kaufman, Morris I [NSTEC; Capelle, Gene A [NSTEC/SB; Grover, M [NSTEC/SB; Stevens, Gerald D [NSTEC/SB; Turley, William D [NSTEC/SB

    2009-01-01

    The design and assembly of a nine-element lens that achieves >2000 1p/mm resolution at a 355-nm wavelength (ultraviolet) has been completed. By adding a doublet to this lens system, operation at a 532-nm wavelength (green) with > 1100 1p/mm resolution is achieved. This lens is used with high-power laser light to record holograms of fast-moving ejecta particles from a shocked metal surface located inside a test package. Part of the lens and the entire test package are under vacuum with a 1-cm air gap separation. Holograms have been recorded with both doubled and tripled Nd:YAG laser light. The UV operation is very sensitive to the package window's tilt. If this window is tilted by more than 0.1 degrees, the green operation performs with better resolution than that of the UV operation. The setup and alignment are performed with green light, but the dynamic recording can be done with either UV light or green light. A resolution plate can be temporarily placed inside the test package so that a television microscope located beyond the hologram position can archive images of resolution patterns that prove that the calibration wires., interference filter, holographic plate, and relay lenses are in their correct positions. Part of this lens is under vacuum, at the point where the laser illumination passes through a focus. Alignment and tolerancing of this high-resolution lens are presented. Resolution variation across the 12-mm field of view and throughout the 5-mm depth of field is discussed for both wavelengths.

  12. Testing the Modern Merger Hypothesis via the Assembly of Massive Blue Elliptical Galaxies in the Local Universe

    E-Print Network [OSTI]

    Haines, Tim; Sánchez, S F; Tremonti, C; Rudnick, G

    2015-01-01

    The modern merger hypothesis offers a method of forming a new elliptical galaxy through merging two equal-mass, gas-rich disk galaxies fuelling a nuclear starburst followed by efficient quenching and dynamical stabilization. A key prediction of this scenario is a central concentration of young stars during the brief phase of morphological transformation from highly-disturbed remnant to new elliptical galaxy. To test this aspect of the merger hypothesis, we use integral field spectroscopy to track the stellar Balmer absorption and 4000\\AA\\ break strength indices as a function of galactic radius for 12 massive (${\\rm M_{*}}\\ge10^{10}{\\rm M_{\\odot}}$), nearby (${\\rm z}\\le0.03$), visually-selected plausible new ellipticals with blue-cloud optical colours and varying degrees of morphological peculiarities. We find that these index values and their radial dependence correlate with specific morphological features such that the most disturbed galaxies have the smallest 4000\\AA\\ break strengths and the largest Balmer ...

  13. Hinge assembly

    DOE Patents [OSTI]

    Vandergriff, D.H.

    1999-08-31

    A hinge assembly is disclosed having a first leaf, a second leaf and linking member. The first leaf has a contact surface. The second leaf has a first contact surface and a second contact surface. The linking member pivotally connects to the first leaf and to the second leaf. The hinge assembly is capable of moving from a closed position to an open position. In the closed position, the contact surface of the first leaf merges with the first contact surface of the second leaf. In the open position, the contact surface of the first leaf merges with the second contact surface of the second leaf. The hinge assembly can include a seal on the contact surface of the first leaf. 8 figs.

  14. Latch assembly

    DOE Patents [OSTI]

    Frederickson, J.R.; Harper, W.H.; Perez, R.

    1984-08-17

    A latch assembly for releasably securing an article in the form of a canister within a container housing. The assembly includes a cam pivotally mounted on the housing wall and biased into the housing interior. The cam is urged into a disabled position by the canister as it enters the housing and a latch release plate maintains the cam disabled when the canister is properly seated in the housing. Upon displacement of the release plate, the cam snaps into latching engagement against the canister for securing the same within the housing. 2 figs.

  15. IRRADIATION PERFORMANCE OF U-Mo MONOLITHIC FUEL

    SciTech Connect (OSTI)

    M.K. Meyer; J. Gan; J.-F. Jue; D.D. Keiser; E. Perez; A. Robinson; D.M. Wachs; N. Woolstenhulme; G.L. Hofman; Y.-S. Kim

    2014-04-01

    High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. UMo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

  16. Summary of Post Irradiation Examination Results of the AFIP-6 Failure

    SciTech Connect (OSTI)

    Adam Robinson; Daniel M. Wachs; Francine Rice; Danielle Perez

    2011-10-01

    The AFIP-6 test assembly was irradiated for one cycle in the Advanced Test Reactor at Idaho National Laboratory. The experiment was designed to test two monolithic fuel plates at power and burn-ups which bounded the operating conditions of both ATR and HFIR driver fuel. Both plates contain a solid U-Mo fuel foil with a zirconium diffusion barrier between 6061-aluminum cladding plates bonded by hot isostatic pressing. The experiment was designed with an orifice to restrict the coolant flow in order to obtain prototypic coolant temperature conditions. While these coolant temperatures were obtained, flow restriction resulted in low heat transfer coefficients and the failure of the fuel plates. The results from the post irradiation examinations and some observations of the failure mechanisms are outlined herein.

  17. Sensor assembly

    DOE Patents [OSTI]

    Bennett, Thomas E.; Nelson, Drew V.

    2004-04-13

    A ribbon-like sensor assembly is described wherein a length of an optical fiber embedded within a similar lengths of a prepreg tow. The fiber is ""sandwiched"" by two layers of the prepreg tow which are merged to form a single consolidated ribbon. The consolidated ribbon achieving a generally uniform distribution of composite filaments near the embedded fiber such that excess resin does not ""pool"" around the periphery of the embedded fiber.

  18. Detailed Destructive Post-Irradiation Examinations of Mixed Uranium and Plutonium Oxide Fuel

    SciTech Connect (OSTI)

    Delashmitt, Jeffrey {Jeff} S [ORNL; Keever, Tamara {Tammy} Jo [ORNL; Smith, Rob R [ORNL] [ORNL; Hexel, Cole R [ORNL] [ORNL; Ilgner, Ralph H [ORNL] [ORNL

    2010-01-01

    The United States Department of Energy (DOE) Fissile Materials Disposition Program (FMDP) is pursuing disposal of surplus weapons-usable plutonium by reactor irradiation as the fissile constituent of MOX fuel. Lead test assemblies (LTAs) have been irradiated for approximately 36 months in Duke Energy's Catawba-1 nuclear power plant (NPP). Per the mixed oxide (MOX) fuel topical report, approved by the U.S. Nuclear Regulatory Commission (NRC), destructive post-irradiation examinations (PIEs) are to be performed on second cycle rods (irradiated to an average burnup of approximately 45 GWd/MTHM). The Radiochemical Analysis Group (RAG) at Oak Ridge National Laboratory (ORNL) is currently performing the detailed destructive post-irradiation examinations (PIE) on four of the mixed uranium and plutonium oxide fuel rods. The analytical process involves dissolution of designated fuel segments in a shielded hot cell for high precision quantification of select fission products and actinide isotopes employing isotope dilution mass spectrometry (IDMS) among other analyses. The hot cell dissolution protocol to include the collection and subsequent alkaline fusion digestion of the fuel's acid resistant metallic particulates will be presented. Although the IDMS measurements of the fission products and actinide isotopes will not be completed by the time of the 51st INMM meeting, the setup and testing of the HPLC chromatographic separations in preparation for these measurements will be discussed.

  19. Shingle assembly

    DOE Patents [OSTI]

    Dinwoodie, Thomas L.

    2007-02-20

    A barrier, such as a PV module, is secured to a base by a support to create a shingle assembly with a venting region defined between the barrier and base for temperature regulation. The first edge of one base may be interengageable with the second edge of an adjacent base to be capable of resisting first and second disengaging forces oriented perpendicular to the edges and along planes oriented parallel to and perpendicular to the base. A deflector may be used to help reduce wind uplift forces.

  20. Pushrod assembly

    DOE Patents [OSTI]

    Potter, J.D.

    1984-03-30

    A pushrod assembly including a carriage mounted on a shaft for movement therealong and carrying a pushrod engageable with a load to be moved is described. A magnet is mounted on a supporting bracket for movement along such shaft. Means are provided for adjustably spacing magnet away from the carriage to obtain a selected magnetic attractive or coupling force therebetween. Movement of the supporting bracket and the magnet carried thereby pulls the carriage along with it until the selected magnetic force is exceeded by a resistance load acting on the carriage.

  1. Dump assembly

    DOE Patents [OSTI]

    Goldmann, L.H.

    1984-12-06

    This is a claim for a dump assembly having a fixed conduit and a rotatable conduit provided with overlapping plates, respectively, at their adjacent ends. The plates are formed with openings, respectively, normally offset from each other to block flow. The other end of the rotatable conduit is provided with means for securing the open end of a filled container thereto. Rotation of the rotatable conduit raises and inverts the container to empty the contents while concurrently aligning the conduit openings to permit flow of material therethrough. 4 figs.

  2. Pushrod assembly

    DOE Patents [OSTI]

    Potter, Jerry D. (Kennewick, WA)

    1987-01-01

    A pushrod assembly including a carriage mounted on a shaft for movement therealong and carrying a pushrod engageable with a load to be moved. A magnet is mounted on a supporting bracket for movement along such shaft. Means are provided for adjustably spacing said magnet away from said carriage to obtain a selected magnetic attractive or coupling force therebetween. Movement of the supporting bracket and the magnet carried thereby pulls the carriage along with it until the selected magnetic force is exceeded by a resistance load acting on the carriage.

  3. Authorized Limits for the Release of a 25 Ton Locomotive, Serial Number 21547, at the Area 25 Engine Maintenance, Assembly, and Disassembly Facility, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    Jeremy Gwin and Douglas Frenette

    2010-04-08

    This document contains process knowledge and radiological data and analysis to support approval for release of the 25-ton locomotive, Serial Number 21547, at the Area 25 Engine Maintenance, Assembly, and Disassembly (EMAD) Facility, located on the Nevada Test Site (NTS). The 25-ton locomotive is a small, one-of-a-kind locomotive used to move railcars in support of the Nuclear Engine for Rocket Vehicle Application project. This locomotive was identified as having significant historical value by the Nevada State Railroad Museum in Boulder City, Nevada, where it will be used as a display piece. A substantial effort to characterize the radiological conditions of the locomotive was undertaken by the NTS Management and Operations Contractor, National Security Technologies, LLC (NSTec). During this characterization process, seven small areas on the locomotive had contamination levels that exceeded the NTS release criteria (limits consistent with U.S. Department of Energy [DOE] Order DOE O 5400.5, “Radiation Protection of the Public and the Environment”). The decision was made to perform radiological decontamination of these known accessible impacted areas to further the release process. On February 9, 2010, NSTec personnel completed decontamination of these seven areas to within the NTS release criteria. Although all accessible areas of the locomotive had been successfully decontaminated to within NTS release criteria, it was plausible that inaccessible areas of the locomotive (i.e., those areas on the locomotive where it was not possible to perform radiological surveys) could potentially have contamination above unrestricted release limits. To access the majority of these inaccessible areas, the locomotive would have to be disassembled. A complete disassembly for a full radiological survey could have permanently destroyed parts and would have ruined the historical value of the locomotive. Complete disassembly would also add an unreasonable financial burden for the contractor. A decision was reached between the NTS regulator and NSTec, opting for alternative authorized limits from DOE Headquarters. In doing so, NSTec personnel performed a dose model using the DOE-approved modeling code RESRAD-BUILD v3.5 to evaluate scenarios. The parameters used in the dose model were conservative. NSTec’s Radiological Engineering Calculation, REC-2010-001, “Public Dose Estimate from the EMAD 25 Ton Locomotive,” concluded that the four scenarios evaluated were below the 25-millirem per year limit, the “likely” dose scenarios met the “few millirem in a year” criteria, and that the EMAD 25-ton locomotive met the radiological requirements to be released with residual radioactivity to the public.

  4. Thermocouple assembly

    DOE Patents [OSTI]

    Thermos, Anthony Constantine (Greer, SC); Rahal, Fadi Elias (Easley, SC)

    2002-01-01

    A thermocouple assembly includes a thermocouple; a plurality of lead wires extending from the thermocouple; an insulating jacket extending along and enclosing the plurality of leads; and at least one internally sealed area within the insulating jacket to prevent fluid leakage along and within the insulating jacket. The invention also provides a method of preventing leakage of a fluid along and through an insulating jacket of a thermocouple including the steps of a) attaching a plurality of lead wires to a thermocouple; b) adding a heat sensitive pseudo-wire to extend along the plurality of lead wires; c) enclosing the lead wires and pseudo-wire inside an insulating jacket; d) locally heating axially spaced portions of the insulating jacket to a temperature which melts the pseudo-wire and fuses it with an interior surface of the jacket.

  5. Swivel assembly

    DOE Patents [OSTI]

    Hall, David R.; Pixton, David S.; Briscoe, Michael; Bradford, Kline; Rawle, Michael; Bartholomew, David B.; McPherson, James

    2007-03-20

    A swivel assembly for a downhole tool string comprises a first and second coaxial housing cooperatively arranged. The first housing comprises a first transmission element in communication with surface equipment. The second housing comprises a second transmission element in communication with the first transmission element. The second housing further comprises a third transmission element adapted for communication with a network integrated into the downhole tool string. The second housing may be rotational and adapted to transmit a signal between the downhole network and the first housing. Electronic circuitry is in communication with at least one of the transmission elements. The electronic circuitry may be externally mounted to the first or second housing. Further, the electronic circuitry may be internally mounted in the second housing. The electronic circuitry may be disposed in a recess in either first or second housing of the swivel.

  6. Horizontal modular dry irradiated fuel storage system

    DOE Patents [OSTI]

    Fischer, Larry E. (Los Gatos, CA); McInnes, Ian D. (San Jose, CA); Massey, John V. (San Jose, CA)

    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).

  7. Key Differences in the Fabrication, Irradiation, and Safety Testing of U.S. and German TRISO-coated Particle Fuel and Their Implications on Fuel Performance

    SciTech Connect (OSTI)

    Petti, David Andrew; Maki, John Thomas; Buongiorno, Jacopo; Hobbins, Richard Redfield

    2002-06-01

    High temperature gas reactor technology is achieving a renaissance around the world. This technology relies on high quality production and performance of coated particle fuel. Historically, the irradiation performance of TRISO-coated gas reactor particle fuel in Germany has been superior to that in the United States. German fuel generally displayed in-pile gas release values that were three orders of magnitude lower than U.S. fuel. Thus, we have critically examined the TRISO-coated fuel fabrication processes in the U.S. and Germany and the associated irradiation database with a goal of understanding why the German fuel behaves acceptably, why the U.S. fuel has not faired as well, and what process/ production parameters impart the reliable performance to this fuel form. The postirradiation examination results are also reviewed to identify failure mechanisms that may be the cause of the poorer U.S. irradiation performance. This comparison will help determine the roles that particle fuel process/product attributes and irradiation conditions (burnup, fast neutron fluence, temperature, and degree of acceleration) have on the behavior of the fuel during irradiation and provide a more quantitative linkage between acceptable processing parameters, as-fabricated fuel properties and subsequent in-reactor performance.

  8. Sequence Assembly Validation by Restriction Digest Fingerprint

    E-Print Network [OSTI]

    Rouchka, Eric

    Sequence Assembly Validation by Restriction Digest Fingerprint Comparison Eric C. Rouchka and David examines the use of restriction digest analysis as a method for testing the fidelity of sequence assembly. Restriction digest fingerprint matching is an established technology for high resolution physical map

  9. Latching relay switch assembly

    DOE Patents [OSTI]

    Duimstra, Frederick A. (Anaheim Hills, CA)

    1991-01-01

    A latching relay switch assembly which includes a coil section and a switch or contact section. The coil section includes at least one permanent magnet and at least one electromagnet. The respective sections are, generally, arranged in separate locations or cavities in the assembly. The switch is latched by a permanent magnet assembly and selectively switched by an overriding electromagnetic assembly.

  10. Inlet nozzle assembly

    DOE Patents [OSTI]

    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.

  11. Tilt assembly for tracking solar collector assembly

    DOE Patents [OSTI]

    Almy, Charles; Peurach, John; Sandler, Reuben

    2012-01-24

    A tilt assembly is used with a solar collector assembly of the type comprising a frame, supporting a solar collector, for movement about a tilt axis by pivoting a drive element between first and second orientations. The tilt assembly comprises a drive element coupler connected to the drive element and a driver, the driver comprising a drive frame, a drive arm and a drive arm driver. The drive arm is mounted to the drive frame for pivotal movement about a drive arm axis. Movement on the drive arm mimics movement of the drive element. Drive element couplers can extend in opposite directions from the outer portion of the drive arm, whereby the assembly can be used between adjacent solar collector assemblies in a row of solar collector assemblies.

  12. Adaptation of Crack Growth Detection Techniques to US Material Test Reactors

    SciTech Connect (OSTI)

    A. Joseph Palmer; Sebastien P. Teysseyre; Kurt L. Davis; Joy L. Rempe; Gordon Kohse; Yakov Ostrovsky; David M. Carpenter

    2014-04-01

    A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some materials testing reactors (MTRs) outside the U.S., such as the Halden Boiling Water Reactor (HBWR), have deployed a technique to measure crack growth propagation during irradiation. This technique incorporates a compact loading mechanism to stress the specimen during irradiation. A crack in the specimen is monitored using the Direct Current Potential Drop (DCPD) method. A project is underway to develop and demonstrate the performance of a similar type of test rig for use in U.S. MTRs. The first year of this three year project was devoted to designing, analyzing, fabricating, and bench top testing a mechanism capable of applying a controlled stress to specimens while they are irradiated in a pressurized water loop (simulating PWR reactor conditions). During the second year, the mechanism will be tested in autoclaves containing high pressure, high temperature water with representative water chemistries. In addition, necessary documentation and safety reviews for testing in a reactor environment will be completed. In the third year, the assembly will be tested in the Massachusetts Institute of Technology Reactor (MITR) and Post Irradiation Examinations (PIE) will be performed.

  13. Phase Stability of an HT-9 Duct Irradiated in FFTF

    SciTech Connect (OSTI)

    O. Anderoglu; J. Van den Bosch; B. H. Sencer; E. Stergar; D. Bhattacharya; P. Dickerson; M. Hartl; S.A. Maloy; P. Hosemann

    2012-11-01

    A fuel test assembly known as ACO-3 duct made out of a fully tempered ferritic/martensitic steel (HT-9) was previously irradiated in the Fast Flux Test Reactor Facility (FFTF) up to 155 dpa at a temperature range of 380-504°C. The microstructures of the samples from 5 different zones along the face of the duct were analyzed using a combination of TEM based techniques, SANS and APT. A high density of Cr rich a' precipitates together with moderate density G-phase precipitates with an average sizes of 5 and 11 nm respectively were found at 20 dpa, 380°C zone. It was found that the precipitations of the second phases are more sensitive to temperature then the dose. In general, the density of both precipitates decreases with increasing irradiation temperature. No significant change is observed in average size of a' while the average size of G-phase precipitates increases up to 27 nm at 440°C. Voids are seen at 100 (410°C) and 155 (440°C) dpa zones but none was detected at 96 dpa (466°C) zone. In contrast to what is reported in the literature, no laves or Chi phases were found in any of the zones.

  14. Autonomous electrochromic assembly

    DOE Patents [OSTI]

    Berland, Brian Spencer; Lanning, Bruce Roy; Stowell, Jr., Michael Wayne

    2015-03-10

    This disclosure describes system and methods for creating an autonomous electrochromic assembly, and systems and methods for use of the autonomous electrochromic assembly in combination with a window. Embodiments described herein include an electrochromic assembly that has an electrochromic device, an energy storage device, an energy collection device, and an electrochromic controller device. These devices may be combined into a unitary electrochromic insert assembly. The electrochromic assembly may have the capability of generating power sufficient to operate and control an electrochromic device. This control may occur through the application of a voltage to an electrochromic device to change its opacity state. The electrochromic assembly may be used in combination with a window.

  15. Supplemental information for a notice of construction for the Fueled Clad Fabrication System, the Radioisotope Power Systems Facility, and the Fuel Assembly Area

    SciTech Connect (OSTI)

    Not Available

    1989-08-01

    This ''Notice of Construction'' has been submitted by the US Department of Energy-Richland Operations Office (P.O. Box 550, Richland, Washington 99352), pursuant to WAC 402-80-070, for three new sources of radionuclide emissions at the Hanford Site in Washington State (Figure 1). The three new sources, the Fueled Clad Fabrication System (FCFS) the Radioisotope Power Systems Facility (RPSF) and the Fuel Assembly Area (FAA) will be located in one facility, the Fuels and materials Examination Facility (FMEF) of the 400 Area. The FMEF was originally designed to provide for post- irradiation examination and fabrication of breeder reactor fuels. These FMEF missions were cancelled before the introduction of any fuel materials or any irradiated material. The current plans are to use the facility to fabricate power supplies to be used in space applications and to produce Fast Flux Test Facility (FFTF) fuel and target assemblies. The FCFS and the RPSF will produce materials and assemblies for application in space. The FAA project will produce FFTF fuel and target assemblies. The FCFS and the RPSF will share the same building, stack, and, in certain cases, the same floor space. Given this relationship, to the extent possible, these systems will be dealt with separately. The FAA is a comparatively independent operation though it will share the FMEF complex.

  16. Prevention of significant deterioration permit application for the Fueled Clad Fabrication System, the Radioisotope Power Systems Facility, and the Fuel Assembly Area

    SciTech Connect (OSTI)

    Not Available

    1989-08-01

    This New Source Review'' has been submitted by the US Department of Energy-Richland Operations Office (PO Box 550, Richland, Washington 99352), pursuant to WAC 173-403-050 and in compliance with the Department of Ecology Guide to Processing A Prevention Of Significant Deterioration (PSD) Permit'' for three new sources of radionuclide emissions at the Hanford Site in Washington State. The three new sources, the Fueled Clad Fabrication System (FCFS), the Radioisotope Power Systems Facility (RPSF), and the Fuel Assembly Area (FAA), will be located in one facility, the Fuels and Materials Examination Facility (FMEF) of the 400 Area. The FMEF was originally designed to provide for post-irradiation examination and fabrication of breeder reactor fuels. These FMEF missions were cancelled before the introduction of any fuel materials or any irradiated material. The current plans are to use the facility to fabricate power supplies for use in space applications and to produce Fast Flux Test Facility (FFTF) fuel and target assemblies. The FCFS and the RPSF will produce materials and assemblies for application in space. The FAA project will produce FFTF fuel and target assemblies. The FCFS and the RPSF will share the same building, stack, and, in certain cases, the same floor space. Given this relationship, these systems will be dealt with separately to the extent possible. The FAA is a comparatively independent operation though it will share the FMEF complex.

  17. Unified irradiance equations

    E-Print Network [OSTI]

    Preisendorfer, Rudolph W

    1957-01-01

    dealt vdth a pair of irradiance functions representing twoHjC^.n^)^ which i s the irradiance a t time t on a unit areaCalifornia UNIFIED IRRADIANCE EQUATIONS R. W. Preisendorfer

  18. Proton irradiation effect on SCDs

    E-Print Network [OSTI]

    Yan-Ji Yang; Jing-Bin Lu; Yu-Sa Wang; Yong Chen; Yu-Peng Xu; Wei-Wei Cui; Wei Li; Zheng-Wei Li; Mao-Shun Li; Xiao-Yan Liu; Juan Wang; Da-Wei Han; Tian-Xiang Chen; Cheng-Kui Li; Jia Huo; Wei Hu; Yi Zhang; Bo Lu; Yue Zhu; Ke-Yan Ma; Di Wu; Yan Liu; Zi-Liang Zhang; Guo-He Yin; Yu Wang

    2014-04-19

    The Low Energy X-ray Telescope is a main payload on the Hard X-ray Modulation Telescope satellite. The swept charge device is selected for the Low Energy X-ray Telescope. As swept charge devices are sensitive to proton irradiation, irradiation test was carried out on the HI-13 accelerator at the China Institute of Atomic Energy. The beam energy was measured to be 10 MeV at the SCD. The proton fluence delivered to the SCD was $3\\times10^{8}\\mathrm{protons}/\\mathrm{cm}^{2}$ over two hours. It is concluded that the proton irradiation affects both the dark current and the charge transfer inefficiency of the SCD through comparing the performance both before and after the irradiation. The energy resolution of the proton-irradiated SCD is 212 eV@5.9 keV at $-60\\,^{\\circ}\\mathrm{C}$, while it before irradiated is 134 eV. Moreover, better performance can be reached by lowering the operating temperature of the SCD on orbit.

  19. Proton irradiation effect on SCDs

    E-Print Network [OSTI]

    Yang, Yan-Ji; Wang, Yu-Sa; Chen, Yong; Xu, Yu-Peng; Cui, Wei-Wei; Li, Wei; Li, Zheng-Wei; Li, Mao-Shun; Liu, Xiao-Yan; Wang, Juan; Han, Da-Wei; Chen, Tian-Xiang; Li, Cheng-Kui; Huo, Jia; Hu, Wei; Zhang, Yi; Lu, Bo; Zhu, Yue; Ma, Ke-Yan; Wu, Di; Liu, Yan; Zhang, Zi-Liang; Yin, Guo-He; Wang, Yu

    2014-01-01

    The Low Energy X-ray Telescope is a main payload on the Hard X-ray Modulation Telescope satellite. The swept charge device is selected for the Low Energy X-ray Telescope. As swept charge devices are sensitive to proton irradiation, irradiation test was carried out on the HI-13 accelerator at the China Institute of Atomic Energy. The beam energy was measured to be 10 MeV at the SCD. The proton fluence delivered to the SCD was $3\\times10^{8}\\mathrm{protons}/\\mathrm{cm}^{2}$ over two hours. It is concluded that the proton irradiation affects both the dark current and the charge transfer inefficiency of the SCD through comparing the performance both before and after the irradiation. The energy resolution of the proton-irradiated SCD is 212 eV@5.9 keV at $-60\\,^{\\circ}\\mathrm{C}$, while it before irradiated is 134 eV. Moreover, better performance can be reached by lowering the operating temperature of the SCD on orbit.

  20. Membrane module assembly

    DOE Patents [OSTI]

    Kaschemekat, J.

    1994-03-15

    A membrane module assembly is described which is adapted to provide a flow path for the incoming feed stream that forces it into prolonged heat-exchanging contact with a heating or cooling mechanism. Membrane separation processes employing the module assembly are also disclosed. The assembly is particularly useful for gas separation or pervaporation. 2 figures.

  1. Membrane module assembly

    DOE Patents [OSTI]

    Kaschemekat, Jurgen (Palo Alto, CA)

    1994-01-01

    A membrane module assembly adapted to provide a flow path for the incoming feed stream that forces it into prolonged heat-exchanging contact with a heating or cooling mechanism. Membrane separation processes employing the module assembly are also disclosed. The assembly is particularly useful for gas separation or pervaporation.

  2. Laterally Mobile, Functionalized Self-Assembled Monolayers at the Fluorous?Aqueous Interface in a Plug-Based Microfluidic System: Characterization and Testing with Membrane Protein Crystallization

    SciTech Connect (OSTI)

    Kreutz, Jason E.; Li, Liang; Roach, L. Spencer; Hatakeyama, Takuji; Ismagilov, Rustem F.; (UC)

    2009-11-04

    This paper describes a method to generate functionalizable, mobile self-assembled monolayers (SAMs) in plug-based microfluidics. Control of interfaces is advancing studies of biological interfaces, heterogeneous reactions, and nanotechnology. SAMs have been useful for such studies, but they are not laterally mobile. Lipid-based methods, though mobile, are not easily amenable to setting up the hundreds of experiments necessary for crystallization screening. Here we demonstrate a method, complementary to current SAM and lipid methods, for rapidly generating mobile, functionalized SAMs. This method relies on plugs, droplets surrounded by a fluorous carrier fluid, to rapidly explore chemical space. Specifically, we implemented his-tag binding chemistry to design a new fluorinated amphiphile, RfNTA, using an improved one-step synthesis of RfOEG under Mitsunobu conditions. RfNTA introduces specific binding of protein at the fluorous-aqueous interface, which concentrates and orients proteins at the interface, even in the presence of other surfactants. We then applied this approach to the crystallization of a his-tagged membrane protein, Reaction Center from Rhodobacter sphaeroides, performed 2400 crystallization trials, and showed that this approach can increase the range of crystal-producing conditions, the success rate at a given condition, the rate of nucleation, and the quality of the crystal formed.

  3. (Irradiation embrittlement of reactor pressure vessels)

    SciTech Connect (OSTI)

    Corwin, W.R.

    1990-09-24

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

  4. Sensor mount assemblies and sensor assemblies

    DOE Patents [OSTI]

    Miller, David H. (Redondo Beach, CA)

    2012-04-10

    Sensor mount assemblies and sensor assemblies are provided. In an embodiment, by way of example only, a sensor mount assembly includes a busbar, a main body, a backing surface, and a first finger. The busbar has a first end and a second end. The main body is overmolded onto the busbar. The backing surface extends radially outwardly relative to the main body. The first finger extends axially from the backing surface, and the first finger has a first end, a second end, and a tooth. The first end of the first finger is disposed on the backing surface, and the tooth is formed on the second end of the first finger.

  5. ATMOSPHERIC TURBIDITY DETERMINATION FROM IRRADIANCE RATIOS Chris Gueymard Frank Vignola

    E-Print Network [OSTI]

    Oregon, University of

    ATMOSPHERIC TURBIDITY DETERMINATION FROM IRRADIANCE RATIOS Chris Gueymard Frank Vignola Florida irradiance measurements and other atmo- spheric parameters. An error analysis and various tests against cloudless and the diffuse irradiance data are very accurate. Yet, this method is insensitive to errors

  6. RERTR-13 Irradiation Summary Report

    SciTech Connect (OSTI)

    D. M. Perez; M. A. Lillo; G. S. Chang; D. M. Wachs; G. A. Roth; N. E. Woolstenhulme

    2012-09-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-13 was designed to assess performance of different types of neutron absorbers that can be potentially used as burnable poisons in the low enriched uranium-molybdenum based dispersion and monolithic fuels.1 The following report summarizes the life of the RERTR-13 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

  7. LUNAR SEISMIC PROFILING EXPERIMENT HOUSING AND CHARGE ASSEMBLY

    E-Print Network [OSTI]

    Rathbun, Julie A.

    ATM-1094f LUNAR SEISMIC PROFILING EXPERIMENT HOUSING AND CHARGE ASSEMBLY FOAM TEST REPORT PAGE DATE. R. Lewis, Manager Lunar Seismic Profiling Experiment #12;ATM-1094 I LUNAR SEISMIC PROFILING'MNAR SEISMIC PROFILING EXPERIMENT HOUSING AND CHARGE ASSEMBLY FOAM TEST REPORT .C"l. j_ ..LV~ .L v 7"'% I PAGE

  8. Irradiation Effects on Human Cortical Bone Fracture Behavior

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

    on different size scales within bone, as well as the role of sustained irradiation damage. Combining in situ mechanical testing with synchrotron x-ray diffraction imaging and...

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

    DOE Patents [OSTI]

    Bradley, John G. (Richland, WA)

    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.

  10. Phased Startup Initiative Phases 3 and 4 Test Plan and Test Specification (OCRWM)

    SciTech Connect (OSTI)

    PITNER, A.L.

    2000-02-28

    Construction for the Spent Nuclear Fuel (SNF) Project facilities is continuing per the Level III Baseline Schedule, and installation of the Fuel Retrieval System (FRS) and Integrated Water Treatment System (IWTS) in K West Basin is now complete. In order to accelerate the project, a phased start up strategy to initiate testing of the FRS and IWTS early in the overall project schedule was proposed (Williams 1999). Wilkinson (1999) expands the definition of the original proposal into four functional testing phases of the Phased Startup Initiative (PSI). Phases 1 and 2 are based on performing functional tests using dummy fuel. These tests are described in separate planning documents. This test plan provides overall guidance for Phase 3 and 4 tests, which are performed using actual irradiated N fuel assemblies. The overall objective of the Phase 3 and 4 testing is to verify how the FRS and IWTS respond while processing actual fuel. Conducting these tests early in the project schedule will allow identification and resolution of equipment and process problems before they become activities on the start-up critical path. The specific objectives of this test plan are to: (1) Define the test scope for the FRS and IWTS; (2) Provide detailed test requirements that can be used to write the specific test procedures; (3) Define data required and measurements to be taken. Where existing methods to obtain these do not exist, enough detail will be provided to define required additional equipment; and (4) Define specific test objectives and acceptance criteria.

  11. Composite turbine bucket assembly

    DOE Patents [OSTI]

    Liotta, Gary Charles; Garcia-Crespo, Andres

    2014-05-20

    A composite turbine blade assembly includes a ceramic blade including an airfoil portion, a shank portion and an attachment portion; and a transition assembly adapted to attach the ceramic blade to a turbine disk or rotor, the transition assembly including first and second transition components clamped together, trapping said ceramic airfoil therebetween. Interior surfaces of the first and second transition portions are formed to mate with the shank portion and the attachment portion of the ceramic blade, and exterior surfaces of said first and second transition components are formed to include an attachment feature enabling the transition assembly to be attached to the turbine rotor or disk.

  12. Public Assembly Buildings

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    as "primary energy," which includes the energy consumed during the generation and transmission of electricity. Public assembly buildings used 577 trillion Btu of primary...

  13. Phase Startup Initiative Phases 3 and 4 Test Plan and Test Specification ( OCRWM)

    SciTech Connect (OSTI)

    PAJUNEN, A.L.; LANGEVIN, M.J.

    2000-08-07

    Construction for the Spent Nuclear Fuel (SNF) Project facilities is continuing per the Level III Baseline Schedule, and installation of the Fuel Retrieval System (FRS) and Integrated Water Treatment System (IWTS) in K West Basin is now complete. In order to accelerate the project, a phased start up strategy to initiate testing of the FRS and IWTS early in the overall project schedule was proposed (Williams 1999). Wilkinson (1999) expands the definition of the original proposal into four functional testing phases of the Phased Startup Initiative (PSI). Phases 1 and 2 are based on performing functional tests using dummy fuel. This test plan provides overall guidance for Phase 3 and 4 tests, which are performed using actual irradiated N fuel assemblies. The overall objective of the Phase 3 and 4 testing is to verify how the FRS and IWTS respond while processing actual fuel. Conducting these tests early in the project schedule will allow identification and resolution of equipment and process problems before they become activities on the start-up critical path. The specific objectives of this test plan are to: Define the Phase 3 and 4 test scope for the FRS and IWTS; Provide detailed test requirements that can be used to write the specific test procedures; Define data required and measurements to be taken. Where existing methods to obtain these do not exist, enough detail will be provided to define required additional equipment; and Define specific test objectives and acceptance criteria.

  14. Laser bottom hole assembly

    DOE Patents [OSTI]

    Underwood, Lance D; Norton, Ryan J; McKay, Ryan P; Mesnard, David R; Fraze, Jason D; Zediker, Mark S; Faircloth, Brian O

    2014-01-14

    There is provided for laser bottom hole assembly for providing a high power laser beam having greater than 5 kW of power for a laser mechanical drilling process to advance a borehole. This assembly utilizes a reverse Moineau motor type power section and provides a self-regulating system that addresses fluid flows relating to motive force, cooling and removal of cuttings.

  15. Turbine disc sealing assembly

    DOE Patents [OSTI]

    Diakunchak, Ihor S.

    2013-03-05

    A disc seal assembly for use in a turbine engine. The disc seal assembly includes a plurality of outwardly extending sealing flange members that define a plurality of fluid pockets. The sealing flange members define a labyrinth flow path therebetween to limit leakage between a hot gas path and a disc cavity in the turbine engine.

  16. Automated Assembly Using Feature Localization

    E-Print Network [OSTI]

    Gordon, Steven Jeffrey

    1986-12-01

    Automated assembly of mechanical devices is studies by researching methods of operating assembly equipment in a variable manner; that is, systems which may be configured to perform many different assembly operations ...

  17. Advanced Test Reactor - A National Scientific User Facility

    SciTech Connect (OSTI)

    Clifford J. Stanley

    2008-05-01

    The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected nuclear research reactor with a maximum operating power of 250 MWth. The unique serpentine configuration of the fuel elements creates five main reactor power lobes (regions) and nine flux traps. In addition to these nine flux traps there are 68 additional irradiation positions in the reactor core reflector tank. There are also 34 low-flux irradiation positions in the irradiation tanks outside the core reflector tank. The ATR is designed to provide a test environment for the evaluation of the effects of intense radiation (neutron and gamma). Due to the unique serpentine core design each of the five lobes can be operated at different powers and controlled independently. Options exist for the individual test trains and assemblies to be either cooled by the ATR coolant (i.e., exposed to ATR coolant flow rates, pressures, temperatures, and neutron flux) or to be installed in their own independent test loops where such parameters as temperature, pressure, flow rate, neutron flux, and energy can be controlled per experimenter specifications. The full-power maximum thermal neutron flux is ~1.0 x1015 n/cm2-sec with a maximum fast flux of ~5.0 x1014 n/cm2-sec. The Advanced Test Reactor, now a National Scientific User Facility, is a versatile tool in which a variety of nuclear reactor, nuclear physics, reactor fuel, and structural material irradiation experiments can be conducted. The cumulative effects of years of irradiation in a normal power reactor can be duplicated in a few weeks or months in the ATR due to its unique design, power density, and operating flexibility.

  18. CAD Assemblies Academic Resource Center

    E-Print Network [OSTI]

    Heller, Barbara

    Desk Inventor, which is free to students #12;Why are Assemblies Important? Assemblies are collections of parts for the Inventor software Prepared by Adin Goings #12;

  19. DC source assemblies

    DOE Patents [OSTI]

    Campbell, Jeremy B; Newson, Steve

    2013-02-26

    Embodiments of DC source assemblies of power inverter systems of the type suitable for deployment in a vehicle having an electrically grounded chassis are provided. An embodiment of a DC source assembly comprises a housing, a DC source disposed within the housing, a first terminal, and a second terminal. The DC source also comprises a first capacitor having a first electrode electrically coupled to the housing, and a second electrode electrically coupled to the first terminal. The DC source assembly further comprises a second capacitor having a first electrode electrically coupled to the housing, and a second electrode electrically coupled to the second terminal.

  20. Sequence Assembly Validation by Multiple Restriction Digest Fragment Coverage Analysis

    E-Print Network [OSTI]

    Rouchka, Eric

    Sequence Assembly Validation by Multiple Restriction Digest Fragment Coverage Analysis Eric C the possibility of analyzing multiple, independent restriction digests as a method for testing the fidelity of the fragment assembly is highly desirable. Comparison of predicted and experimental restriction digests has

  1. Steam separator latch assembly

    DOE Patents [OSTI]

    Challberg, R.C.; Kobsa, I.R.

    1994-02-01

    A latch assembly removably joins a steam separator assembly to a support flange disposed at a top end of a tubular shroud in a nuclear reactor pressure vessel. The assembly includes an annular head having a central portion for supporting the steam separator assembly thereon, and an annular head flange extending around a perimeter thereof for supporting the head to the support flange. A plurality of latches are circumferentially spaced apart around the head flange with each latch having a top end, a latch hook at a bottom end thereof, and a pivot support disposed at an intermediate portion therebetween and pivotally joined to the head flange. The latches are pivoted about the pivot supports for selectively engaging and disengaging the latch hooks with the support flange for fixedly joining the head to the shroud or for allowing removal thereof. 12 figures.

  2. Core assembly storage structure

    DOE Patents [OSTI]

    Jones, Jr., Charles E. (Northridge, CA); Brunings, Jay E. (Chatsworth, CA)

    1988-01-01

    A structure for the storage of core assemblies from a liquid metal-cooled nuclear reactor. The structure comprises an enclosed housing having a substantially flat horizontal top plate, a bottom plate and substantially vertical wall members extending therebetween. A plurality of thimble members extend downwardly through the top plate. Each thimble member is closed at its bottom end and has an open end adjacent said top plate. Each thimble member has a length and diameter greater than that of the core assembly to be stored therein. The housing is provided with an inlet duct for the admission of cooling air and an exhaust duct for the discharge of air therefrom, such that when hot core assemblies are placed in the thimbles, the heat generated will by convection cause air to flow from the inlet duct around the thimbles and out the exhaust duct maintaining the core assemblies at a safe temperature without the necessity of auxiliary powered cooling equipment.

  3. Rnnotator Assembly Pipeline

    SciTech Connect (OSTI)

    Martin, Jeff

    2010-06-03

    Jeff Martin of the DOE Joint Genome Institute discusses a de novo transcriptome assembly pipeline from short RNA-Seq reads on June 3, 2010 at the "Sequencing, Finishing, Analysis in the Future" meeting in Santa Fe, NM

  4. Magnetic assisted statistical assembly

    E-Print Network [OSTI]

    Cheng, Diana I

    2008-01-01

    The objective of this thesis is to develop a process using magnetic forces to assemble micro-components into recesses on silicon based integrated circuits. Patterned SmCo magnetic thin films at the bottom of recesses are ...

  5. Swipe transfer assembly

    DOE Patents [OSTI]

    Christiansen, Robert M. (Blackfoot, ID); Mills, William C. (McKeesport, PA)

    1992-01-01

    The swipe transfer assembly is a mechanical assembly which is used in conjunction with glove boxes and other sealed containments. It is used to pass small samples into or out of glove boxes without an open breach of the containment, and includes a rotational cylinder inside a fixed cylinder, the inside cylinder being rotatable through an arc of approximately 240.degree. relative to the outer cylinder. An offset of 120.degree. from end to end allows only one port to be opened at a time. The assembly is made of stainless steel or aluminum and clear acrylic plastic to enable visual observation. The assembly allows transfer of swipes and smears from radiological and other specially controlled environments.

  6. Design of an Online Fission Gas Monitoring System for Post-irradiation Examination Heating Tests of Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

    SciTech Connect (OSTI)

    Dawn Scates

    2010-10-01

    A new Fission Gas Monitoring System (FGMS) has been designed at the Idaho National Laboratory (INL) for use of monitoring online fission gas-released during fuel heating tests. The FGMS will be used with the Fuel Accident Condition Simulator (FACS) at the Hot Fuels Examination Facility (HFEF) located at the Materials and Fuels Complex (MFC) within the INL campus. Preselected Advanced Gas Reactor (AGR) TRISO (Tri-isotropic) fuel compacts will undergo testing to assess the fission product retention characteristics under high temperature accident conditions. The FACS furnace will heat the fuel to temperatures up to 2,000ºC in a helium atmosphere. Released fission products such as Kr and Xe isotopes will be transported downstream to the FGMS where they will accumulate in cryogenically cooledcollection traps and monitored with High Purity Germanium (HPGe) detectors during the heating process. Special INL developed software will be used to monitor the accumulated fission products and will report data in near real-time. These data will then be reported in a form that can be readily available to the INL reporting database. This paper describes the details of the FGMS design, the control and acqusition software, system calibration, and the expected performance of the FGMS. Preliminary online data may be available for presentation at the High Temperature Reactor (HTR) conference.

  7. Recuperator assembly and procedures

    DOE Patents [OSTI]

    Kang, Yungmo; McKeirnan, Jr., Robert D.

    2006-06-27

    A construction of recuperator core segments is provided which insures proper assembly of the components of the recuperator core segment, and of a plurality of recuperator core segments. Each recuperator core segment must be constructed so as to prevent nesting of fin folds of the adjacent heat exchanger foils of the recuperator core segment. A plurality of recuperator core segments must be assembled together so as to prevent nesting of adjacent fin folds of adjacent recuperator core segments.

  8. Recuperator assembly and procedures

    DOE Patents [OSTI]

    Kang, Yungmo (La Canada Flintridge, CA); McKeirnan, Jr., Robert D. (Westlake Village, CA)

    2008-08-26

    A construction of recuperator core segments is provided which insures proper assembly of the components of the recuperator core segment, and of a plurality of recuperator core segments. Each recuperator core segment must be constructed so as to prevent nesting of fin folds of the adjacent heat exchanger foils of the recuperator core segment. A plurality of recuperator core segments must be assembled together so as to prevent nesting of adjacent fin folds of adjacent recuperator core segments.

  9. Hg System Assembly and Testing Status

    E-Print Network [OSTI]

    McDonald, Kirk

    NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY Wiring Started · Sensor wiring inside the hydraulic control #12;2VRVS Meeting 1 Nov 2006 OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY Welding OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY Cart Modifications · Secondary box center

  10. Hg System Assembly and Testing Status

    E-Print Network [OSTI]

    McDonald, Kirk

    ;4VRVS Meeting 8 Nov 2006 OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY Hydraulic Hose Issue #12;2VRVS Meeting 8 Nov 2006 OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY Hg Drain & Spill Meeting 8 Nov 2006 OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY Local Hg Vapor Filters · Hg

  11. Hg System Assembly and Testing Status

    E-Print Network [OSTI]

    McDonald, Kirk

    U. S. DEPARTMENT OF ENERGY Hydraulic Hose Issue Corrected · Replaced flex hoses with rigid tubing #12;2VRVS Meeting 22 Nov 2006 OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY Hg System baseplate #12;3VRVS Meeting 22 Nov 2006 OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY Basic

  12. Hg System Assembly and Testing Status

    E-Print Network [OSTI]

    McDonald, Kirk

    · Secondary beam windows received and installed · Problem with bolt pattern on downstream window was corrected tech support #12;5VRVS Meeting 20 Dec 2006 OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY

  13. RERTR-7 Irradiation Summary Report

    SciTech Connect (OSTI)

    D. M. Perez; M. A. Lillo; G. S. Chang; G. A. Roth; N. E. Woolstenhulme; D. M. Wachs

    2011-12-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-7A, was designed to test several modified fuel designs to target fission densities representative of a peak low enriched uranium (LEU) burnup in excess of 90% U-235 at peak experiment power sufficient to generate a peak surface heat flux of approximately 300 W/cm2. The RERTR-7B experiment was designed as a high power test of 'second generation' dispersion fuels at peak experiment power sufficient to generate a surface heat flux on the order of 230 W/cm2.1 The following report summarizes the life of the RERTR-7A and RERTR-7B experiments through end of irradiation, including as-run neutronic analyses, thermal analyses and hydraulic testing results.

  14. Neutron Irradiation Resistance of RAFM Steels

    SciTech Connect (OSTI)

    Gaganidze, Ermile; Dafferner, Bernhard; Aktaa, Jarir

    2008-07-01

    The neutron irradiation resistance of the reduced-activation ferritic/martensitic (RAFM) steel EUROFER97 and international reference steels (F82H-mod, OPTIFER-Ia, GA3X and MANET-I) have been investigated after irradiation in the Petten High Flux Reactor up to 16.3 dpa at different irradiation temperatures (250-450 deg. C). The embrittlement behavior and hardening are investigated by instrumented Charpy-V tests with sub-size specimens. Neutron irradiation-induced embrittlement and hardening of EUROFER97 was studied under different heat treatment conditions. Embrittlement and hardening of as-delivered EUROFER97 steel are comparable to those of reference steels. Heat treatment of EUROFER97 at a higher austenitizing temperature substantially improves the embrittlement behaviour at low irradiation temperatures. Analysis of embrittlement vs. hardening behavior of RAFM steels within a proper model in terms of the parameter C={delta}DBTT/{delta}{sigma} indicates hardening-dominated embrittlement at irradiation temperatures below 350 deg. C with 0.17 {<=} C {<=} 0.53 deg. C/MPa. Scattering of C at irradiation temperatures above 400 deg. C indicates non hardening embrittlement. A role of He in a process of embrittlement is investigated in EUROFER97 based steels, that are doped with different contents of natural B and the separated {sup 10}B-isotope (0.008-0.112 wt.%). Testing on small scale fracture mechanical specimens for determination of quasi-static fracture toughness will be also presented in a view of future irradiation campaigns. (authors)

  15. Superconducting radiofrequency window assembly

    DOE Patents [OSTI]

    Phillips, Harry L. (Seaford, VA); Elliott, Thomas S. (Yorktown, VA)

    1997-01-01

    The present invention is a superconducting radiofrequency window assembly for use in an electron beam accelerator. The srf window assembly (20) has a superconducting metal-ceramic design. The srf window assembly (20) comprises a superconducting frame (30), a ceramic plate (40) having a superconducting metallized area, and a superconducting eyelet (50) for sealing plate (40) into frame (30). The plate (40) is brazed to eyelet (50) which is then electron beam welded to frame (30). A method for providing a ceramic object mounted in a metal member to withstand cryogenic temperatures is also provided. The method involves a new metallization process for coating a selected area of a ceramic object with a thin film of a superconducting material. Finally, a method for assembling an electron beam accelerator cavity utilizing the srf window assembly is provided. The procedure is carried out within an ultra clean room to minimize exposure to particulates which adversely affect the performance of the cavity within the electron beam accelerator.

  16. Superconductive radiofrequency window assembly

    DOE Patents [OSTI]

    Phillips, Harry Lawrence (Seaford, VA); Elliott, Thomas S. (Yorktown, VA)

    1998-01-01

    The present invention is a superconducting radiofrequency window assembly for use in an electron beam accelerator. The srf window assembly (20) has a superconducting metal-ceramic design. The srf window assembly (20) comprises a superconducting frame (30), a ceramic plate (40) having a superconducting metallized area, and a superconducting eyelet (50) for sealing plate (40) into frame (30). The plate (40) is brazed to eyelet (50) which is then electron beam welded to frame (30). A method for providing a ceramic object mounted in a metal member to withstand cryogenic temperatures is also provided. The method involves a new metallization process for coating a selected area of a ceramic object with a thin film of a superconducting material. Finally, a method for assembling an electron beam accelerator cavity utilizing the srf window assembly is provided. The procedure is carried out within an ultra clean room to minimize exposure to particulates which adversely affect the performance of the cavity within the electron beam accelerator.

  17. Photovoltaic self-assembly.

    SciTech Connect (OSTI)

    Lavin, Judith; Kemp, Richard Alan; Stewart, Constantine A.

    2010-10-01

    This late-start LDRD was focused on the application of chemical principles of self-assembly on the ordering and placement of photovoltaic cells in a module. The drive for this chemical-based self-assembly stems from the escalating prices in the 'pick-and-place' technology currently used in the MEMS industries as the size of chips decreases. The chemical self-assembly principles are well-known on a molecular scale in other material science systems but to date had not been applied to the assembly of cells in a photovoltaic array or module. We explored several types of chemical-based self-assembly techniques, including gold-thiol interactions, liquid polymer binding, and hydrophobic-hydrophilic interactions designed to array both Si and GaAs PV chips onto a substrate. Additional research was focused on the modification of PV cells in an effort to gain control over the facial directionality of the cells in a solvent-based environment. Despite being a small footprint research project worked on for only a short time, the technical results and scientific accomplishments were significant and could prove to be enabling technology in the disruptive advancement of the microelectronic photovoltaics industry.

  18. Constrained space camera assembly

    DOE Patents [OSTI]

    Heckendorn, F.M.; Anderson, E.K.; Robinson, C.W.; Haynes, H.B.

    1999-05-11

    A constrained space camera assembly which is intended to be lowered through a hole into a tank, a borehole or another cavity is disclosed. The assembly includes a generally cylindrical chamber comprising a head and a body and a wiring-carrying conduit extending from the chamber. Means are included in the chamber for rotating the body about the head without breaking an airtight seal formed therebetween. The assembly may be pressurized and accompanied with a pressure sensing means for sensing if a breach has occurred in the assembly. In one embodiment, two cameras, separated from their respective lenses, are installed on a mounting apparatus disposed in the chamber. The mounting apparatus includes means allowing both longitudinal and lateral movement of the cameras. Moving the cameras longitudinally focuses the cameras, and moving the cameras laterally away from one another effectively converges the cameras so that close objects can be viewed. The assembly further includes means for moving lenses of different magnification forward of the cameras. 17 figs.

  19. Fire resistant PV shingle assembly

    SciTech Connect (OSTI)

    Lenox, Carl J.

    2012-10-02

    A fire resistant PV shingle assembly includes a PV assembly, including PV body, a fire shield and a connection member connecting the fire shield below the PV body, and a support and inter-engagement assembly. The support and inter-engagement assembly is mounted to the PV assembly and comprises a vertical support element, supporting the PV assembly above a support surface, an upper interlock element, positioned towards the upper PV edge, and a lower interlock element, positioned towards the lower PV edge. The upper interlock element of one PV shingle assembly is inter-engageable with the lower interlock element of an adjacent PV shingle assembly. In some embodiments the PV shingle assembly may comprise a ventilation path below the PV body. The PV body may be slidably mounted to the connection member to facilitate removal of the PV body.

  20. Power module assembly

    DOE Patents [OSTI]

    Campbell, Jeremy B. (Torrance, CA); Newson, Steve (Redondo Beach, CA)

    2011-11-15

    A power module assembly of the type suitable for deployment in a vehicular power inverter, wherein the power inverter has a grounded chassis, is provided. The power module assembly comprises a conductive base layer electrically coupled to the chassis, an insulating layer disposed on the conductive base layer, a first conductive node disposed on the insulating layer, a second conductive node disposed on the insulating layer, wherein the first and second conductive nodes are electrically isolated from each other. The power module assembly also comprises a first capacitor having a first electrode electrically connected to the conductive base layer, and a second electrode electrically connected to the first conductive node, and further comprises a second capacitor having a first electrode electrically connected to the conductive base layer, and a second electrode electrically connected to the second conductive node.

  1. Supported PV module assembly

    SciTech Connect (OSTI)

    Mascolo, Gianluigi; Taggart, David F.; Botkin, Jonathan D.; Edgett, Christopher S.

    2013-10-15

    A supported PV assembly may include a PV module comprising a PV panel and PV module supports including module supports having a support surface supporting the module, a module registration member engaging the PV module to properly position the PV module on the module support, and a mounting element. In some embodiments the PV module registration members engage only the external surfaces of the PV modules at the corners. In some embodiments the assembly includes a wind deflector with ballast secured to a least one of the PV module supports and the wind deflector. An array of the assemblies can be secured to one another at their corners to prevent horizontal separation of the adjacent corners while permitting the PV modules to flex relative to one another so to permit the array of PV modules to follow a contour of the support surface.

  2. Lightweight, self-ballasting photovoltaic roofing assembly

    DOE Patents [OSTI]

    Dinwoodie, Thomas L.

    2006-02-28

    A photovoltaic roofing assembly comprises a roofing membrane (102), a plurality of photovoltaic modules (104, 106, 108) disposed as a layer on top of the roofing membrane (102), and a plurality of pre-formed spacers, pedestals or supports (112, 114, 116, 118, 120, 122) which are respectively disposed below the plurality of photovoltaic modules (104, 106, 108) and integral therewith, or fixed thereto. Spacers (112, 114, 116, 118, 120, 122) are disposed on top of roofing membrane (102). Membrane (102) is supported on conventional roof framing, and attached thereto by conventional methods. In an alternative embodiment, the roofing assembly may have insulation block (322) below the spacers (314, 314', 315, 315'). The geometry of the pre-formed spacers (112, 114, 116, 118, 120, 122, 314, 314', 315, 315') is such that wind tunnel testing has shown its maximum effectiveness in reducing net forces of wind uplift on the overall assembly. Such construction results in a simple, lightweight, self-ballasting, readily assembled roofing assembly which resists the forces of wind uplift using no roofing penetrations.

  3. Assembling and Installing LRUs for NIF

    SciTech Connect (OSTI)

    Bonanno, R E

    2003-12-31

    Within the 192 National Ignition Facility (NIF) beamlines, there are over 7000 large (40 x 40 cm) optical components, including laser glass, mirrors, lenses, and polarizers. These optics are held in large opto-mechanical assemblies called line-replaceable units (LRUs). Each LRU has strict specifications with respect to cleanliness, alignment, and wavefront so that once activated, each NIF beamline will meet its performance requirements. NIF LRUs are assembled, tested, and refurbished in on-site cleanroom facilities. The assembled LRUs weigh up to 1800 kilograms, and are about the size of a phone booth. They are transported in portable clean canisters and inserted into the NIF beampath using robotic transporters. This plug and play design allows LRUs to be easily removed from the beampath for maintenance or upgrades. Commissioning of the first NIF quad, an activity known as NIF Early Light (NEL), has validated LRU designs and architecture, as well as demonstrated that LRUs can be assembled and installed as designed. Furthermore, it has served to develop key processes and tools forming the foundation for NIF s long-term LRU production and maintenance strategy. As we look forward to building out the rest of NIF, the challenge lies in scaling up the production rate while maintaining quality, implementing process improvements, and fully leveraging the learning and experience gained from NEL. This paper provides an overview of the facilities, equipment and processes used to assemble and install LRUs in NIF.

  4. Solar central receiver heliostat reflector assembly

    DOE Patents [OSTI]

    Horton, Richard H. (Schenectady, NY); Zdeb, John J. (Clifton Park, NY)

    1980-01-01

    A heliostat reflector assembly for a solar central receiver system comprises a light-weight, readily assemblable frame which supports a sheet of stretchable reflective material and includes mechanism for selectively applying tension to and positioning the sheet to stretch it to optical flatness. The frame is mounted on and supported by a pipe pedestal assembly that, in turn, is installed in the ground. The frame 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. central receiver, which forms part of a solar energy utilization system, such as a solar energy fueled electrical power generation system. The frame may include a built-in system for testing for optical flatness of the reflector. The preferable geometric configuration of the reflector is octagonal; however, it may be other shapes, such as hexagonal, pentagonal or square. Several different embodiments of means for tensioning and positioning the reflector to achieve optical flatness are disclosed. The reflector assembly is based on the stretch frame concept which provides an extremely light-weight, simple, low-cost reflector assembly that may be driven for positioning and tracking by a light-weight, inexpensive drive system.

  5. Lightweight, self-ballasting photovoltaic roofing assembly

    DOE Patents [OSTI]

    Dinwoodie, T.L.

    1998-05-05

    A photovoltaic roofing assembly comprises a roofing membrane (102), a plurality of photovoltaic modules (104, 106, 108) disposed as a layer on top of the roofing membrane (102), and a plurality of pre-formed spacers, pedestals or supports (112, 114, 116, 118, 120, 122) which are respectively disposed below the plurality of photovoltaic modules (104, 106, 108) and integral therewith, or fixed thereto. Spacers (112, 114, 116, 118, 120, 122) are disposed on top of roofing membrane (102). Membrane (102) is supported on conventional roof framing, and attached thereto by conventional methods. In an alternative embodiment, the roofing assembly may have insulation block (322) below the spacers (314, 314', 315, 315'). The geometry of the preformed spacers (112, 114, 116, 118, 120, 122, 314, 314', 315, 315') is such that wind tunnel testing has shown its maximum effectiveness in reducing net forces of wind uplift on the overall assembly. Such construction results in a simple, lightweight, self-ballasting, readily assembled roofing assembly which resists the forces of wind uplift using no roofing penetrations.

  6. Lightweight, self-ballasting photovoltaic roofing assembly

    DOE Patents [OSTI]

    Dinwoodie, Thomas L. (Berkeley, CA)

    1998-01-01

    A photovoltaic roofing assembly comprises a roofing membrane (102), a plurality of photovoltaic modules (104, 106, 108) disposed as a layer on top of the roofing membrane (102), and a plurality of pre-formed spacers, pedestals or supports (112, 114, 116, 118, 120, 122) which are respectively disposed below the plurality of photovoltaic modules (104, 106, 108) and integral therewith, or fixed thereto. Spacers (112, 114, 116, 118, 120, 122) are disposed on top of roofing membrane (102). Membrane (102) is supported on conventional roof framing, and attached thereto by conventional methods. In an alternative embodiment, the roofing assembly may have insulation block (322) below the spacers (314, 314', 315, 315'). The geometry of the preformed spacers (112, 114, 116, 118, 120, 122, 314, 314', 315, 315') is such that wind tunnel testing has shown its maximum effectiveness in reducing net forces of wind uplift on the overall assembly. Such construction results in a simple, lightweight, self-ballasting, readily assembled roofing assembly which resists the forces of wind uplift using no roofing penetrations.

  7. Low inductance connector assembly

    DOE Patents [OSTI]

    Holbrook, Meghan Ann; Carlson, Douglas S

    2013-07-09

    A busbar connector assembly for coupling first and second terminals on a two-terminal device to first and second contacts on a power module is provided. The first terminal resides proximate the first contact and the second terminal resides proximate the second contact. The assembly comprises a first bridge having a first end configured to be electrically coupled to the first terminal, and a second end configured to be electrically coupled to the second contact, and a second bridge substantially overlapping the first bridge and having a first end electrically coupled to the first contact, and a second end electrically coupled to the second terminal.

  8. Self assembling proteins

    DOE Patents [OSTI]

    Yeates, Todd O.; Padilla, Jennifer; Colovos, Chris

    2004-06-29

    Novel fusion proteins capable of self-assembling into regular structures, as well as nucleic acids encoding the same, are provided. The subject fusion proteins comprise at least two oligomerization domains rigidly linked together, e.g. through an alpha helical linking group. Also provided are regular structures comprising a plurality of self-assembled fusion proteins of the subject invention, and methods for producing the same. The subject fusion proteins find use in the preparation of a variety of nanostructures, where such structures include: cages, shells, double-layer rings, two-dimensional layers, three-dimensional crystals, filaments, and tubes.

  9. Durable Fuel Cell Membrane Electrode Assembly (MEA)

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

    Durable Fuel Cell Membrane Electrode Assembly (MEA) Durable Fuel Cell Membrane Electrode Assembly (MEA) A revolutionary method of building a membrane electrode assembly (MEA) for...

  10. Segmented stator assembly

    DOE Patents [OSTI]

    Lokhandwalla, Murtuza; Alexander, James Pellegrino; El-Refaie, Ayman Mohamed Fawzi; Shah, Manoj Ramprasad; Quirion, Owen Scott

    2013-04-02

    An electric machine and stator assembly are provided that include a continuous stator portion having stator teeth, and a tooth tip portion including tooth tips corresponding to the stator teeth of the continuous stator portion, respectively. The tooth tip portion is mounted onto the continuous stator portion.

  11. Rotary shaft sealing assembly

    DOE Patents [OSTI]

    Dietle, Lannie L. (Houston, TX); Schroeder, John E. (Richmond, TX); Kalsi, Manmohan S. (Houston, TX); Alvarez, Patricio D. (Richmond, TX)

    2010-09-21

    A rotary shaft sealing assembly in which a first fluid is partitioned from a second fluid in a housing assembly having a rotary shaft located at least partially within. In one embodiment a lip seal is lubricated and flushed with a pressure-generating seal ring preferably having an angled diverting feature. The pressure-generating seal ring and a hydrodynamic seal may be used to define a lubricant-filled region with each of the seals having hydrodynamic inlets facing the lubricant-filled region. Another aspect of the sealing assembly is having a seal to contain pressurized lubricant while withstanding high rotary speeds. Another rotary shaft sealing assembly embodiment includes a lubricant supply providing a lubricant at an elevated pressure to a region between a lip seal and a hydrodynamic seal with a flow control regulating the flow of lubricant past the lip seal. The hydrodynamic seal may include an energizer element having a modulus of elasticity greater than the modulus of elasticity of a sealing lip of the hydrodynamic seal.

  12. Rotary shaft sealing assembly

    DOE Patents [OSTI]

    Dietle, Lannie L; Schroeder, John E; Kalsi, Manmohan S; Alvarez, Patricio D

    2013-08-13

    A rotary shaft sealing assembly in which a first fluid is partitioned from a second fluid in a housing assembly having a rotary shaft located at least partially within. In one embodiment a lip seal is lubricated and flushed with a pressure-generating seal ring preferably having an angled diverting feature. The pressure-generating seal ring and a hydrodynamic seal may be used to define a lubricant-filled region with each of the seals having hydrodynamic inlets facing the lubricant-filled region. Another aspect of the sealing assembly is having a seal to contain pressurized lubricant while withstanding high rotary speeds. Another rotary shaft sealing assembly embodiment includes a lubricant supply providing a lubricant at an elevated pressure to a region between a lip seal and a hydrodynamic seal with a flow control regulating the flow of lubricant past the lip seal. The hydrodynamic seal may include an energizer element having a modulus of elasticity greater than the modulus of elasticity of a sealing lip of the hydrodynamic seal.

  13. Evaluation of Neutron Irradiated Silicon Carbide and Silicon Carbide Composites

    SciTech Connect (OSTI)

    Newsome G, Snead L, Hinoki T, Katoh Y, Peters D

    2007-03-26

    The effects of fast neutron irradiation on SiC and SiC composites have been studied. The materials used were chemical vapor deposition (CVD) SiC and SiC/SiC composites reinforced with either Hi-Nicalon{trademark} Type-S, Hi-Nicalon{trademark} or Sylramic{trademark} fibers fabricated by chemical vapor infiltration. Statistically significant numbers of flexural samples were irradiated up to 4.6 x 10{sup 25} n/m{sup 2} (E>0.1 MeV) at 300, 500 and 800 C in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Dimensions and weights of the flexural bars were measured before and after the neutron irradiation. Mechanical properties were evaluated by four point flexural testing. Volume increase was seen for all bend bars following neutron irradiation. Magnitude of swelling depended on irradiation temperature and material, while it was nearly independent of irradiation fluence over the fluence range studied. Flexural strength of CVD SiC increased following irradiation depending on irradiation temperature. Over the temperature range studied, no significant degradation in mechanical properties was seen for composites fabricated with Hi-Nicalon{trademark} Type-S, while composites reinforced with Hi-Nicalon{trademark} or Sylramic fibers showed significant degradation. The effects of irradiation on the Weibull failure statistics are also presented suggesting a reduction in the Weibull modulus upon irradiation. The cause of this potential reduction is not known.

  14. Los Alamos National Laboratory summary plan to fabricate mixed oxide lead assemblies for the fissile material disposition program

    SciTech Connect (OSTI)

    Buksa, J.J.; Eaton, S.L.; Trellue, H.R.; Chidester, K.; Bowidowicz, M.; Morley, R.A.; Barr, M.

    1997-12-01

    This report summarizes an approach for using existing Los Alamos National Laboratory (Laboratory) mixed oxide (MOX) fuel-fabrication and plutonium processing capabilities to expedite and assure progress in the MOX/Reactor Plutonium Disposition Program. Lead Assembly MOX fabrication is required to provide prototypic fuel for testing in support of fuel qualification and licensing requirements. It is also required to provide a bridge for the full utilization of the European fabrication experience. In part, this bridge helps establish, for the first time since the early 1980s, a US experience base for meeting the safety, licensing, safeguards, security, and materials control and accountability requirements of the Department of Energy and Nuclear Regulatory Commission. In addition, a link is needed between the current research and development program and the production of disposition mission fuel. This link would also help provide a knowledge base for US regulators. Early MOX fabrication and irradiation testing in commercial nuclear reactors would provide a positive demonstration to Russia (and to potential vendors, designers, fabricators, and utilities) that the US has serious intent to proceed with plutonium disposition. This report summarizes an approach to fabricating lead assembly MOX fuel using the existing MOX fuel-fabrication infrastructure at the Laboratory.

  15. EA-1210: Finding of No Significant Impact

    Broader source: Energy.gov [DOE]

    Lead Test Assembly Irradiation and Analysis Watts Bar Nuclear Plant, Tennessee and Hanford Site, Richland, Washington

  16. Irradiation requirements of Nb3Sn based SC magnets electrical insulation

    E-Print Network [OSTI]

    McDonald, Kirk

    Irradiation requirements of Nb3Sn based SC magnets electrical insulation developed within the Eu and Power Engineering #12;Outline · Motivation of launching EuCARD irradiation task · Nb3Sn SC magnet coils electrical insulation candidates · EuCARD insulators certification conditions · Post irradiation tests

  17. Irradiation facilities at the Los Alamos Meson Physics Facility

    SciTech Connect (OSTI)

    Sandberg, V.

    1990-01-01

    The irradiation facilities for testing SSC components and detector systems are described. Very high intensity proton, neutron, and pion fluxes are available with beam kinetic energies of up to 800 MeV. 4 refs., 12 figs., 2 tabs.

  18. Design Strategies for Self-Assembly of Discrete Targets

    E-Print Network [OSTI]

    Jim Madge; Mark A. Miller

    2015-08-11

    Both biological and artificial self-assembly processes can take place by a range of different schemes, from the successive addition of identical building blocks, to hierarchical sequences of intermediates, all the way to the fully addressable limit in which each component is unique. In this paper we introduce an idealized model of cubic particles with patterned faces that allows self-assembly strategies to be compared and tested. We consider a simple octameric target, starting with the minimal requirements for successful self-assembly and comparing the benefits and limitations of more sophisticated hierarchical and addressable schemes. Simulations are performed using a hybrid dynamical Monte Carlo protocol that allows self-assembling clusters to rearrange internally while still providing Stokes-Einstein-like diffusion of aggregates of different sizes. Our simulations explicitly capture the thermodynamic, dynamic and steric challenges typically faced by self-assembly processes, including competition between multiple partially-completed structures. Self-assembly pathways are extracted from the simulation trajectories by a fully extendable scheme for identifying structural fragments, which are then assembled into history diagrams for successfully completed target structures. For the simple target, a one-component assembly scheme is most efficient and robust overall, but hierarchical and addressable strategies can have an advantage under some conditions if high yield is a priority.

  19. Vacuum breaker valve assembly

    DOE Patents [OSTI]

    Thompson, J.L.; Upton, H.A.

    1999-04-27

    Breaker valve assemblies for a simplified boiling water nuclear reactor are described. The breaker valve assembly, in one form, includes a valve body and a breaker valve. The valve body includes an interior chamber, and an inlet passage extends from the chamber and through an inlet opening to facilitate transporting particles from outside of the valve body to the interior chamber. The breaker valve is positioned in the chamber and is configured to substantially seal the inlet opening. Particularly, the breaker valve includes a disk which is sized to cover the inlet opening. The disk is movably coupled to the valve body and is configured to move substantially concentrically with respect to the valve opening between a first position, where the disk completely covers the inlet opening, and a second position, where the disk does not completely cover the inlet opening. 1 fig.

  20. Vacuum breaker valve assembly

    DOE Patents [OSTI]

    Thompson, Jeffrey L. (San Jose, CA); Upton, Hubert Allen (Morgan Hill, CA)

    1999-04-27

    Breaker valve assemblies for a simplified boiling water nuclear reactor are described. The breaker valve assembly, in one form, includes a valve body and a breaker valve. The valve body includes an interior chamber, and an inlet passage extends from the chamber and through an inlet opening to facilitate transporting particles from outside of the valve body to the interior chamber. The breaker valve is positioned in the chamber and is configured to substantially seal the inlet opening. Particularly, the breaker valve includes a disk which is sized to cover the inlet opening. The disk is movably coupled to the valve body and is configured to move substantially concentrically with respect to the valve opening between a first position, where the disk completely covers the inlet opening, and a second position, where the disk does not completely cover the inlet opening.

  1. Solution deposition assembly

    SciTech Connect (OSTI)

    Roussillon, Yann; Scholz, Jeremy H; Shelton, Addison; Green, Geoff T; Utthachoo, Piyaphant

    2014-01-21

    Methods and devices are provided for improved deposition systems. In one embodiment of the present invention, a deposition system is provided for use with a solution and a substrate. The system comprises of a solution deposition apparatus; at least one heating chamber, at least one assembly for holding a solution over the substrate; and a substrate curling apparatus for curling at least one edge of the substrate to define a zone capable of containing a volume of the solution over the substrate. In another embodiment of the present invention, a deposition system for use with a substrate, the system comprising a solution deposition apparatus; at heating chamber; and at least assembly for holding solution over the substrate to allow for a depth of at least about 0.5 microns to 10 mm.

  2. Mechanical seal assembly

    DOE Patents [OSTI]

    Kotlyar, Oleg M. (Salt Lake City, UT)

    2002-01-01

    An improved mechanical seal assembly is provided for sealing rotating shafts with respect to their shaft housings, wherein the rotating shafts are subject to substantial axial vibrations. The mechanical seal assembly generally includes a rotating sealing ring fixed to the shaft, a non-rotating sealing ring adjacent to and in close contact with the rotating sealing ring for forming an annular seal about the shaft, and a mechanical diode element that applies a biasing force to the non-rotating sealing ring by means of hemispherical joint. The alignment of the mechanical diode with respect to the sealing rings is maintained by a series of linear bearings positioned axially along a desired length of the mechanical diode. Alternative embodiments include mechanical or hydraulic amplification components for amplifying axial displacement of the non-rotating sealing ring and transfering it to the mechanical diode.

  3. Turbine seal assembly

    DOE Patents [OSTI]

    Little, David A.

    2013-04-16

    A seal assembly that limits gas leakage from a hot gas path to one or more disc cavities in a turbine engine. The seal assembly includes a seal apparatus that limits gas leakage from the hot gas path to a respective one of the disc cavities. The seal apparatus comprises a plurality of blade members rotatable with a blade structure. The blade members are associated with the blade structure and extend toward adjacent stationary components. Each blade member includes a leading edge and a trailing edge, the leading edge of each blade member being located circumferentially in front of the blade member's corresponding trailing edge in a direction of rotation of the turbine rotor. The blade members are arranged such that a space having a component in a circumferential direction is defined between adjacent circumferentially spaced blade members.

  4. Ingestion resistant seal assembly

    DOE Patents [OSTI]

    Little, David A. (Chuluota, FL)

    2011-12-13

    A seal assembly limits gas leakage from a hot gas path to one or more disc cavities in a gas turbine engine. The seal assembly includes a seal apparatus associated with a blade structure including a row of airfoils. The seal apparatus includes an annular inner shroud associated with adjacent stationary components, a wing member, and a first wing flange. The wing member extends axially from the blade structure toward the annular inner shroud. The first wing flange extends radially outwardly from the wing member toward the annular inner shroud. A plurality of regions including one or more recirculation zones are defined between the blade structure and the annular inner shroud that recirculate working gas therein back toward the hot gas path.

  5. Low inductance busbar assembly

    DOE Patents [OSTI]

    Holbrook, Meghan Ann (Manhattan Beach, CA)

    2010-09-21

    A busbar assembly for electrically coupling first and second busbars to first and second contacts, respectively, on a power module is provided. The assembly comprises a first terminal integrally formed with the first busbar, a second terminal integrally formed with the second busbar and overlapping the first terminal, a first bridge electrode having a first tab electrically coupled to the first terminal and overlapping the first and second terminals, and a second tab electrically coupled to the first contact, a second bridge electrode having a third tab electrically coupled to the second terminal, and overlapping the first and second terminals and the first tab, and a fourth tab electrically coupled to the second contact, and a fastener configured to couple the first tab to the first terminal, and the third tab to the second terminal.

  6. Subsea wellhead seal assembly

    SciTech Connect (OSTI)

    Gullion, S.D.

    1988-07-26

    An annular subsea wellhead seal assembly is described for sealing against the walls in a subsea wellhead annulus above a landing seat at the lower end of the annulus comprising: an annular body having an outwardly and downwardly flaring outer skirt and an inwardly and downwardly flaring inner skirt extending from it slower surface, a landing ring having lower landing surface for landing on a landing seat at the lower end of the subsea wellhead annulus in which the assembly is to seal, and an upper flat reaction surface which is positioned immediately under and engagable with the lower ends of the skirts; and means connecting the body and the landing ring for relative movement toward each other; downward movement of the body with respect to the landing ring spreading the skirts outward and inward, respectively, into a substantially horizontal digging engagement set position with the walls of the annulus to be sealed.

  7. Mechanical seal assembly

    DOE Patents [OSTI]

    Kotlyar, Oleg M. (Salt Lake City, UT)

    2001-01-01

    An improved mechanical seal assembly is provided for sealing rotating shafts with respect to their shaft housings, wherein the rotating shafts are subject to substantial axial vibrations. The mechanical seal assembly generally includes a rotating sealing ring fixed to the shaft, a non-rotating sealing ring adjacent to and in close contact with the rotating sealing ring for forming an annular seal about the shaft, and a mechanical diode element that applies a biasing force to the non-rotating sealing ring by means of hemispherical joint. The alignment of the mechanical diode with respect to the sealing rings is maintained by a series of linear bearings positioned axially along a desired length of the mechanical diode. Alternative embodiments include mechanical or hydraulic amplification components for amplifying axial displacement of the non-rotating sealing ring and transferring it to the mechanical diode.

  8. Infrared floodlight assembly

    DOE Patents [OSTI]

    Wierzbicki, Julian J. (Peabody, MA); Chakrabarti, Kirti B. (Danvers, MA)

    1987-09-22

    An infrared floodlight assembly (10) including a cast aluminum outer housing (11) defining a central chamber (15) therein. A floodlight (14), having a tungsten halogen lamp as the light source, is spacedly positioned within a heat conducting member (43) within chamber (15) such that the floodlight is securedly positioned in an aligned manner relative to the assembly's filter (35) and lens (12) components. The invention also includes venting means (51) to allow air passage between the interior of the member (43) and the adjacent chamber (15), as well as engagement means (85) for engaging a rear surface of the floodlight (14) to retain it firmly against an internal flange of the member (43). A reflector (61), capable of being compressed to allow insertion or removal, is located within the heat conducting member's interior between the floodlight (14) and filter (35) to reflect infrared radiation toward the filter (35) and spaced lens (12).

  9. Fuel nozzle assembly

    DOE Patents [OSTI]

    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.

  10. Pull rod assembly

    DOE Patents [OSTI]

    Cioletti, O.C.

    1988-04-21

    A pull rod assembly comprising a pull rod having three peripheral grooves, a piston device including an adaptor ring and a seal ring, said piston device being mounted on the pull rod by a split ring retainer situated in one groove and extending into an interior groove in the adaptor and a resilient split ring retained in another groove and positioned to engage the piston device and to retain the seal on its adaptor.

  11. Coaxial test fixture

    DOE Patents [OSTI]

    Praeg, Walter F. (Palos Park, IL)

    1986-01-01

    An assembly is provided for testing one or more contact material samples in a vacuum environment. The samples are positioned as an inner conductive cylinder assembly which is mounted for reciprocal vertical motion as well as deflection from a vertical axis. An outer conductive cylinder is coaxially positioned around the inner cylinder and test specimen to provide a vacuum enclosure therefor. A power source needed to drive test currents through the test specimens is connected to the bottom of each conductive cylinder, through two specially formed conductive plates. The plates are similar in form, having a plurality of equal resistance current paths connecting the power source to a central connecting ring. The connecting rings are secured to the bottom of the inner conductive assembly and the outer cylinder, respectively. A hydraulic actuator is also connected to the bottom of the inner conductor assembly to adjust the pressure applied to the test specimens during testing. The test assembly controls magnetic forces such that the current distribution through the test samples is symmetrical and that contact pressure is not reduced or otherwise disturbed.

  12. Opportunities for mixed oxide fuel testing in the advanced test reactor to support plutonium disposition

    SciTech Connect (OSTI)

    Terry, W.K.; Ryskamp, J.M.; Sterbentz, J.W.

    1995-08-01

    Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification; (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania; (3) The effects of WGPu isotopic composition; (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight; (5) The effects of americium and gallium in WGPu; (6) Fission gas release from MOX fuel pellets made from WGPu; (7) Fuel/cladding gap closure; (8) The effects of power cycling and off-normal events on fuel integrity; (9) Development of radial distributions of burnup and fission products; (10) Power spiking near the interfaces of MOX and urania fuel assemblies; and (11) Fuel performance code validation. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory possesses many advantages for performing tests to resolve most of the issues identified above. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified. The facilities at Argonne National Laboratory-West can meet all potential needs for pre- and post-irradiation examination that might arise in a MOX fuel qualification program.

  13. Emulation of reactor irradiation damage using ion beams

    SciTech Connect (OSTI)

    Was, G. S.; Jiao, Z.; Getto, E.; Sun, K.; Monterrosa, A. M.; Maloy, S. A.; Anderoglu, O.; Sencer, B. H.; Hackett, M.

    2014-06-14

    The continued operation of existing light water nuclear reactors and the development of advanced nuclear reactor depend heavily on understanding how damage by radiation to levels degrades materials that serve as the structural components in reactor cores. The first high dose ion irradiation experiments on a ferritic-martensitic steel showing that ion irradiation closely emulates the full radiation damage microstructure created in-reactor are described. Ferritic-martensitic alloy HT9 (heat 84425) in the form of a hexagonal fuel bundle duct (ACO-3) accumulated 155 dpa at an average temperature of 443°C in the Fast Flux Test Facility (FFTF). Using invariance theory as a guide, irradiation of the same heat was conducted using self-ions (Fe++) at 5 MeV at a temperature of 460°C and to a dose of 188 displacements per atom. The void swelling was nearly identical between the two irradiation and the size and density of precipitates and loops following ion irradiation are within a factor of two of those for neutron irradiation. The level of agreement across all of the principal microstructure changes between ion and reactor irradiation establishes the capability of tailoring ion irradiation to emulate the reactor-irradiated microstructure.

  14. Emulation of reactor irradiation damage using ion beams

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Was, G. S.; Jiao, Z.; Getto, E.; Sun, K.; Monterrosa, A. M.; Maloy, S. A.; Anderoglu, O.; Sencer, B. H.; Hackett, M.

    2014-06-14

    The continued operation of existing light water nuclear reactors and the development of advanced nuclear reactor depend heavily on understanding how damage by radiation to levels degrades materials that serve as the structural components in reactor cores. The first high dose ion irradiation experiments on a ferritic-martensitic steel showing that ion irradiation closely emulates the full radiation damage microstructure created in-reactor are described. Ferritic-martensitic alloy HT9 (heat 84425) in the form of a hexagonal fuel bundle duct (ACO-3) accumulated 155 dpa at an average temperature of 443°C in the Fast Flux Test Facility (FFTF). Using invariance theory as a guide,more »irradiation of the same heat was conducted using self-ions (Fe++) at 5 MeV at a temperature of 460°C and to a dose of 188 displacements per atom. The void swelling was nearly identical between the two irradiation and the size and density of precipitates and loops following ion irradiation are within a factor of two of those for neutron irradiation. The level of agreement across all of the principal microstructure changes between ion and reactor irradiation establishes the capability of tailoring ion irradiation to emulate the reactor-irradiated microstructure.« less

  15. Emulation of reactor irradiation damage using ion beams

    SciTech Connect (OSTI)

    G. S. Was; Z. Jiao; E. Beckett; A. M. Monterrosa; O. Anderoglu; B. H. Sencer; M. Hackett

    2014-10-01

    The continued operation of existing light water nuclear reactors and the development of advanced nuclear reactor depend heavily on understanding how damage by radiation to levels degrades materials that serve as the structural components in reactor cores. The first high dose ion irradiation experiments on a ferritic-martensitic steel showing that ion irradiation closely emulates the full radiation damage microstructure created in-reactor are described. Ferritic-martensitic alloy HT9 (heat 84425) in the form of a hexagonal fuel bundle duct (ACO-3) accumulated 155 dpa at an average temperature of 443°C in the Fast Flux Test Facility (FFTF). Using invariance theory as a guide, irradiation of the same heat was conducted using self-ions (Fe++) at 5 MeV at a temperature of 460°C and to a dose of 188 displacements per atom. The void swelling was nearly identical between the two irradiations and the size and density of precipitates and loops following ion irradiation are within a factor of two of those for neutron irradiation. The level of agreement across all of the principal microstructure changes between ion and reactor irradiations establishes the capability of tailoring ion irradiations to emulate the reactor-irradiated microstructure.

  16. Temperature dependence of fracture toughness in HT9 steel neutron-irradiated up to 145 dpa

    SciTech Connect (OSTI)

    Baek, Jong-Hyuk; Byun, Thak Sang; Maloy, S; Toloczko, M

    2014-01-01

    The temperature dependence of fracture toughness in HT9 steel irradiated to high doses was investigated using miniature three-point bend (TPB) fracture specimens. These specimens were from the ACO-3 fuel duct wall of the Fast Flux Test Facility (FFTF), in which irradiation doses were in the range of 3.2 144.8 dpa and irradiation temperatures in the range of 380.4 502.6 oC. A miniature specimen reuse technique has been established for this investigation: the specimens used were the tested halves of miniature Charpy impact specimens (~13 3 4 mm) with diamond-saw cut in the middle. The fatigue precracking for specimens and fracture resistance (J-R) tests were carried out in a MTS servo-hydraulic testing machine with a vacuum furnace following the standard procedure described in the ASTM Standard E 1820-09. For each of five irradiated and one archive conditions, 7 to 9 J-R tests were performed at selected temperatures ranging from 22 C to 600 C. The fracture toughness of the irradiated HT9 steel was strongly dependent on irradiation temperatures rather than irradiation dose. When the irradiation temperature was below about 430 C, the fracture toughness of irradiated HT9 increased with test temperature, reached an upper shelf of 180 200 MPa m at 350 450 C and then decreased with test temperature. When the irradiation temperature 430 C, the fracture toughness was nearly unchanged until about 450 C and decreased with test temperature in higher temperature range. Similar test temperature dependence was observed for the archive material although the highest toughness values are lower after irradiation. Ductile stable crack growth occurred except for a few cases where both the irradiation temperature and test temperature are relatively low.

  17. CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 1, 2, AND 3 OF CRYSTAL RIVER UNIT 3

    SciTech Connect (OSTI)

    Kenneth D. Wright

    1997-07-29

    The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 1, 2, and 3 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies.

  18. CRC DEPLETION CALCULATIONS FOR THE RODDED ASSEMBLIES IN BATCHES 1, 2, 3, AND 1X OF CRYSTAL RIVER UNIT 3

    SciTech Connect (OSTI)

    Kenneth D. Wright

    1997-09-03

    The purpose of this design analysis is to document the SAS2H depletion calculations of certain rodded fuel assemblies from batches 1, 2, 3, and 1X of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A rodded assembly is one that contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) for some period of time during its irradiation history. The objective of this analysis is to provide SAS2H calculated isotopic compositions of depleted fuel and depleted burnable poison for each fuel assembly to be used in subsequent CRC reactivity calculations containing the fuel assemblies.

  19. CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 4 AND 5 OF CRYSTAL RIVER UNIT 3

    SciTech Connect (OSTI)

    Kenneth D. Wright

    1997-07-30

    The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 4 and 5 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies.

  20. Irradiation-induced nano-voids in strained tin precipitates in silicon

    SciTech Connect (OSTI)

    Gaiduk, P. I., E-mail: gaiduk@phys.au.dk [Department of Physics and Astronomy/iNANO, Aarhus University, Gustav Wieds Vej 14, DK-8000 Aarhus C (Denmark); Department of Physical Electronics and Nanotechnology, Belarusian State University, prosp. Nezavisimosti, 4, 220030 Minsk (Belarus); Lundsgaard Hansen, J., E-mail: johnlh@phys.au.dk; Nylandsted Larsen, A., E-mail: anl@phys.au.dk [Department of Physics and Astronomy/iNANO, Aarhus University, Gustav Wieds Vej 14, DK-8000 Aarhus C (Denmark)

    2014-04-14

    We report on self-assembling of spherically shaped voids in nanometer size strained Sn precipitates after irradiation with He{sup +} ions in different conditions. It is found that high-temperature irradiation induces vacancies which are collected by compressively strained Sn precipitates enhancing of out-diffusion of Sn atoms from the precipitates. Nano-voids formation takes place simultaneously with a ?- to ?-phase transformation in the Sn precipitates. Post-irradiation thermal treatment leads to the removal of voids and a backward transformation of the Sn phase to ?-phase. Strain-enhanced separation of point defects along with vacancy assisted Sn out-diffusion and precipitate dissolution are discussed.

  1. KJRR-FAI Hydraulic Flow Testing Input Package

    SciTech Connect (OSTI)

    N.E. Woolstenhulme; R.B. Nielson; D.B. Chapman

    2013-12-01

    The INL, in cooperation with the KAERI via Cooperative Research And Development Agreement (CRADA), undertook an effort in the latter half of calendar year 2013 to produce a conceptual design for the KJRR-FAI campaign. The outcomes of this effort are documented in further detail elsewhere [5]. The KJRR-FAI was designed to be cooled by the ATR’s Primary Coolant System (PCS) with no provision for in-pile measurement or control of the hydraulic conditions in the irradiation assembly. The irradiation assembly was designed to achieve the target hydraulic conditions via engineered hydraulic losses in a throttling orifice at the outlet of the irradiation vehicle.

  2. Irradiation study of Ti6Al4V and Ti6Al4V1B for FRIB beam dump: Aida Amroussia, PhD Student

    E-Print Network [OSTI]

    McDonald, Kirk

    Irradiation study of Ti6Al4V and Ti6Al4V1B for FRIB beam dump: Aida Amroussia, PhD Student Chemical Engineering and Materials Science Michigan State University May 19, 2014 1 #12;Outline · Irradiation-EBSD · Nano-indentation tests · Vickers Hardness tests · Insitu-tensile tests 2 #12;Irradiation experiments

  3. Normal Conditions of Transport Truck Test of a Surrogate Fuel...

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

    Normal Conditions of Transport Truck Test of a Surrogate Fuel Assembly. Citation Details In-Document Search Title: Normal Conditions of Transport Truck Test of a Surrogate Fuel...

  4. Estimating solar irradiance using a geostationary satellite

    E-Print Network [OSTI]

    Urquhart, Bryan Glenn

    2011-01-01

    Heliosat procedure for irradiance Estimates from Satellitethe pro duction of solar irradiance and daylight illuminancesatellite based solar irradiance modelling - The SOLIS clear

  5. IRRADIATION OF 3- SUBSTITUTED-2-PHENYLOXAZIRIDINES

    E-Print Network [OSTI]

    Splitter, Janet S.; Calvin, Melvin.

    2008-01-01

    1 . These oxaziridines from irradiation of the corresponding1). A complication in the irradiation of the ozaziridines isof imine formed in the irradiation of the oxaziridines. (

  6. Nanomechanical testing system

    DOE Patents [OSTI]

    Vodnick, David James; Dwivedi, Arpit; Keranen, Lucas Paul; Okerlund, Michael David; Schmitz, Roger William; Warren, Oden Lee; Young, Christopher David

    2014-07-08

    An automated testing system includes systems and methods to facilitate inline production testing of samples at a micro (multiple microns) or less scale with a mechanical testing instrument. In an example, the system includes a probe changing assembly for coupling and decoupling a probe of the instrument. The probe changing assembly includes a probe change unit configured to grasp one of a plurality of probes in a probe magazine and couple one of the probes with an instrument probe receptacle. An actuator is coupled with the probe change unit, and the actuator is configured to move and align the probe change unit with the probe magazine and the instrument probe receptacle. In another example, the automated testing system includes a multiple degree of freedom stage for aligning a sample testing location with the instrument. The stage includes a sample stage and a stage actuator assembly including translational and rotational actuators.

  7. Precision Robotic Assembly Machine

    ScienceCinema (OSTI)

    None

    2010-09-01

    The world's largest laser system is the National Ignition Facility (NIF), located at Lawrence Livermore National Laboratory. NIF's 192 laser beams are amplified to extremely high energy, and then focused onto a tiny target about the size of a BB, containing frozen hydrogen gas. The target must be perfectly machined to incredibly demanding specifications. The Laboratory's scientists and engineers have developed a device called the "Precision Robotic Assembly Machine" for this purpose. Its unique design won a prestigious R&D-100 award from R&D Magazine.

  8. Removable feedwater sparger assembly

    DOE Patents [OSTI]

    Challberg, R.C.

    1994-10-04

    A removable feedwater sparger assembly includes a sparger having an inlet pipe disposed in flow communication with the outlet end of a supply pipe. A tubular coupling includes an annular band fixedly joined to the sparger inlet pipe and a plurality of fingers extending from the band which are removably joined to a retention flange extending from the supply pipe for maintaining the sparger inlet pipe in flow communication with the supply pipe. The fingers are elastically deflectable for allowing engagement of the sparger inlet pipe with the supply pipe and for disengagement therewith. 8 figs.

  9. Irradiation Planning for Fully-Ceramic Micro-encsapsulated fuel in ATR at LWR-relevant conditions: year-end report on FY-2011

    SciTech Connect (OSTI)

    Abderrafi M. Ougouag; R. Sonat Sen; Michael A. Pope; Brian Boer

    2011-09-01

    This report presents the estimation of required ATR irradiation levels for the DB-FCM fuel design (fueled with Pu and MAs). The fuel and assembly designs are those considered in a companion report [R. S. Sen et al., FCR&D-2011- 00037 or INL/EXT-11-23269]. These results, pertaining to the DB-FCM fuel, are definitive in as much as the design of said fuel is definitive. In addition to the work performed, as required, for DB-FCM fuel, work has started in a preliminary fashion on single-cell UO2 and UN fuels. These latter activities go beyond the original charter of this project and although the corresponding work is incomplete, significant progress has been achieved. However, in this context, all that has been achieved is only preliminary because the corresponding fuel designs are neither finalized nor optimized. In particular, the UO2 case is unlikely to result in a viable fuel design if limited to enrichment at or under 20 weight % in U-235. The UN fuel allows reasonable length cycles and is likely to make an optimal design possible. Despite being limited to preliminary designs and offering only preliminary conclusions, the irradiation planning tasks for UO2 and UN fuels that are summarized in this report are useful to the overall goal of devising and deploying FCM-LWR fuel since the methods acquired and tested in this project and the overall procedure for planning will be available for planning tests for the finalized fuel design. Indeed, once the fuel design is finalized and the expected burnup level is determined, the methodology that has been assembled will allow the prompt finalization of the neutronic planning of the irradiation experiment and would provide guidance on the expected experimental performance of the fuel. Deviations from the expected behavior will then have to be analyzed and the outcome of the analysis may be corrections or modifications for the assessment models as well as, possibly, fuel design modifications, and perhaps even variation of experimental control for future experimental phases. Besides the prediction of irradiation times, preliminary work was carried out on other aspects of irradiation planning. In particular, a method for evaluating the interplay of depletion, material performance modeling and irradiation is identified by reference to a companion report. Another area that was addressed in a preliminary fashion is the identification and selection of a strategy for the physical and mechanical design of the irradiation experiments. The principal conclusion is that the similarity between the FCM fuel and the fuel compacts of the Next Generation Nuclear Plant prismatic design are strong enough to warrant using irradiation hardware designs and instrumentation adapted from the AGR irradiation tests. Modifications, if found necessary, will probably be few and small, except as pertains to the water environment and its implications on the use of SiC cladding or SiC matrix with no additional cladding.

  10. Gas separation membrane module assembly

    DOE Patents [OSTI]

    Wynn, Nicholas P (Palo Alto, CA); Fulton, Donald A. (Fairfield, CA)

    2009-03-31

    A gas-separation membrane module assembly and a gas-separation process using the assembly. The assembly includes a set of tubes, each containing gas-separation membranes, arranged within a housing. The housing contains a tube sheet that divides the space within the housing into two gas-tight spaces. A permeate collection system within the housing gathers permeate gas from the tubes for discharge from the housing.

  11. Airfoil nozzle and shroud assembly

    DOE Patents [OSTI]

    Shaffer, James E. (Maitland, FL); Norton, Paul F. (San Diego, CA)

    1997-01-01

    An airfoil and nozzle assembly including an outer shroud having a plurality of vane members attached to an inner surface and having a cantilevered end. The assembly further includes a inner shroud being formed by a plurality of segments. Each of the segments having a first end and a second end and having a recess positioned in each of the ends. The cantilevered end of the vane member being positioned in the recess. The airfoil and nozzle assembly being made from a material having a lower rate of thermal expansion than that of the components to which the airfoil and nozzle assembly is attached.

  12. Airfoil nozzle and shroud assembly

    DOE Patents [OSTI]

    Shaffer, J.E.; Norton, P.F.

    1997-06-03

    An airfoil and nozzle assembly are disclosed including an outer shroud having a plurality of vane members attached to an inner surface and having a cantilevered end. The assembly further includes a inner shroud being formed by a plurality of segments. Each of the segments having a first end and a second end and having a recess positioned in each of the ends. The cantilevered end of the vane member being positioned in the recess. The airfoil and nozzle assembly being made from a material having a lower rate of thermal expansion than that of the components to which the airfoil and nozzle assembly is attached. 5 figs.

  13. Multi-position photovoltaic assembly

    DOE Patents [OSTI]

    Dinwoodie, Thomas L. (Piedmont, CA)

    2003-03-18

    The invention is directed to a PV assembly, for use on a support surface, comprising a base, a PV module, a multi-position module support assembly, securing the module to the base at shipping and inclined-use angles, a deflector, a multi-position deflector support securing the deflector to the base at deflector shipping and deflector inclined-use angles, the module and deflector having opposed edges defining a gap therebetween. The invention permits transport of the PV assemblies in a relatively compact form, thus lowering shipping costs, while facilitating installation of the PV assemblies with the PV module at the proper inclination.

  14. Moisture Research - Optimizing Wall Assemblies

    SciTech Connect (OSTI)

    Arena, Lois; Mantha, Pallavi

    2013-05-01

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

  15. Methanation assembly using multiple reactors

    DOE Patents [OSTI]

    Jahnke, Fred C.; Parab, Sanjay C.

    2007-07-24

    A methanation assembly for use with a water supply and a gas supply containing gas to be methanated in which a reactor assembly has a plurality of methanation reactors each for methanating gas input to the assembly and a gas delivery and cooling assembly adapted to deliver gas from the gas supply to each of said methanation reactors and to combine water from the water supply with the output of each methanation reactor being conveyed to a next methanation reactor and carry the mixture to such next methanation reactor.

  16. Microchannel heat sink assembly

    DOE Patents [OSTI]

    Bonde, W.L.; Contolini, R.J.

    1992-03-24

    The present invention provides a microchannel heat sink with a thermal range from cryogenic temperatures to several hundred degrees centigrade. The heat sink can be used with a variety of fluids, such as cryogenic or corrosive fluids, and can be operated at a high pressure. The heat sink comprises a microchannel layer preferably formed of silicon, and a manifold layer preferably formed of glass. The manifold layer comprises an inlet groove and outlet groove which define an inlet manifold and an outlet manifold. The inlet manifold delivers coolant to the inlet section of the microchannels, and the outlet manifold receives coolant from the outlet section of the microchannels. In one embodiment, the manifold layer comprises an inlet hole extending through the manifold layer to the inlet manifold, and an outlet hole extending through the manifold layer to the outlet manifold. Coolant is supplied to the heat sink through a conduit assembly connected to the heat sink. A resilient seal, such as a gasket or an O-ring, is disposed between the conduit and the hole in the heat sink in order to provide a watertight seal. In other embodiments, the conduit assembly may comprise a metal tube which is connected to the heat sink by a soft solder. In still other embodiments, the heat sink may comprise inlet and outlet nipples. The present invention has application in supercomputers, integrated circuits and other electronic devices, and is suitable for cooling materials to superconducting temperatures. 13 figs.

  17. Bottom head assembly

    DOE Patents [OSTI]

    Fife, A.B.

    1998-09-01

    A bottom head dome assembly is described which includes, in one embodiment, a bottom head dome and a liner configured to be positioned proximate the bottom head dome. The bottom head dome has a plurality of openings extending there through. The liner also has a plurality of openings extending there through, and each liner opening aligns with a respective bottom head dome opening. A seal is formed, such as by welding, between the liner and the bottom head dome to resist entry of water between the liner and the bottom head dome at the edge of the liner. In the one embodiment, a plurality of stub tubes are secured to the liner. Each stub tube has a bore extending there through, and each stub tube bore is coaxially aligned with a respective liner opening. A seat portion is formed by each liner opening for receiving a portion of the respective stub tube. The assembly also includes a plurality of support shims positioned between the bottom head dome and the liner for supporting the liner. In one embodiment, each support shim includes a support stub having a bore there through, and each support stub bore aligns with a respective bottom head dome opening. 2 figs.

  18. Microchannel heat sink assembly

    DOE Patents [OSTI]

    Bonde, Wayne L. (Livermore, CA); Contolini, Robert J. (Pleasanton, CA)

    1992-01-01

    The present invention provides a microchannel heat sink with a thermal range from cryogenic temperatures to several hundred degrees centigrade. The heat sink can be used with a variety of fluids, such as cryogenic or corrosive fluids, and can be operated at a high pressure. The heat sink comprises a microchannel layer preferably formed of silicon, and a manifold layer preferably formed of glass. The manifold layer comprises an inlet groove and outlet groove which define an inlet manifold and an outlet manifold. The inlet manifold delivers coolant to the inlet section of the microchannels, and the outlet manifold receives coolant from the outlet section of the microchannels. In one embodiment, the manifold layer comprises an inlet hole extending through the manifold layer to the inlet manifold, and an outlet hole extending through the manifold layer to the outlet manifold. Coolant is supplied to the heat sink through a conduit assembly connected to the heat sink. A resilient seal, such as a gasket or an O-ring, is disposed between the conduit and the hole in the heat sink in order to provide a watetight seal. In other embodiments, the conduit assembly may comprise a metal tube which is connected to the heat sink by a soft solder. In still other embodiments, the heat sink may comprise inlet and outlet nipples. The present invention has application in supercomputers, integrated circuits and other electronic devices, and is suitable for cooling materials to superconducting temperatures.

  19. Assembling thefacebook: Using heterogeneity to understand online social network assembly

    E-Print Network [OSTI]

    Jacobs, Abigail Z; Ugander, Johan; Clauset, Aaron

    2015-01-01

    Online social networks represent a popular and highly diverse class of social media systems. Despite this variety, each of these systems undergoes a general process of online social network assembly, which represents the complicated and heterogeneous changes that transform newly born systems into mature platforms. However, little is known about this process. For example, how much of a network's assembly is driven by simple growth? How does a network's structure change as it matures? How does network structure vary with adoption rates and user heterogeneity, and do these properties play different roles at different points in the assembly? We investigate these and other questions using a unique dataset of online connections among the roughly one million users at the first 100 colleges admitted to Facebook, captured just 20 months after its launch. We first show that different vintages and adoption rates across this population of networks reveal temporal dynamics of the assembly process, and that assembly is onl...

  20. Irradiation Stability of Carbon Nanotubes 

    E-Print Network [OSTI]

    Aitkaliyeva, Assel

    2010-01-14

    was revealed by Raman spectroscopy study of irradiated carbon buckypapers. The deviations from linear relationship were observed for the samples irradiated to very high fluence values. Annealing of irradiated samples was able to reduce the value of ID/IG ratio...

  1. Three dimensional colorimetric assay assemblies

    DOE Patents [OSTI]

    Charych, Deborah (Albany, CA); Reichart, Anke (Albany, CA)

    2000-01-01

    A direct assay is described using novel three-dimensional polymeric assemblies which change from a blue to red color when exposed to an analyte, in one case a flu virus. The assemblies are typically in the form of liposomes which can be maintained in a suspension, and show great intensity in their color changes. Their method of production is also described.

  2. Moisture Research - Optimizing Wall Assemblies

    SciTech Connect (OSTI)

    Arena, L.; Mantha, P.

    2013-05-01

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

  3. Stator and method of assembly

    DOE Patents [OSTI]

    Alexander, James Pellegrino; El-Refaie, Ayman Mohamed Fawzi; Shen, Xiaochun

    2013-06-18

    The present application provides a stator. The stator may include a number of poles and a stator tip and cooling assembly. The stator tip and cooling assembly may include a number of stator tips with a number of cooling tubes adjacent thereto such that the stator tips align with the poles and the cooling tubes cool the poles.

  4. UNC Faculty Assembly Reports to the Faculty Senate Cynthia Deale, Faculty Assembly Delegate

    E-Print Network [OSTI]

    Gopalakrishnan, K.

    UNC Faculty Assembly Reports to the Faculty Senate Cynthia Deale, Faculty Assembly Delegate Report Assembly Delegates Report on February 20, 2015, UNC Faculty Assembly Meeting. James Holloway, Faculty Assembly Delegate Report on January 9, 2015, UNC Faculty Assembly Meeting. Robert Kulesher, Faculty

  5. Microstructure and mechanical behavior of neutron irradiated ultrafine grained ferritic steel

    SciTech Connect (OSTI)

    Ahmad Alsabbagh; Apu Sarkar; Brandon Miller; Jatuporn Burns; Leah Squires; Douglas Porter; James I. Cole; K. L. Murty

    2014-10-01

    Neutron irradiation effects on ultra-fine grain (UFG) low carbon steel prepared by equal channel angular pressing (ECAP) has been examined. Counterpart samples with conventional grain (CG) sizes have been irradiated alongside with the UFG ones for comparison. Samples were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to 1.24 dpa. Atom probe tomography revealed manganese, silicon-enriched clusters in both ECAP and CG steel after neutron irradiation. X-ray quantitative analysis showed that dislocation density in CG increased after irradiation. However, no significant change was observed in UFG steel revealing better radiation tolerance.

  6. Photovoltaic cell assembly

    DOE Patents [OSTI]

    Beavis, Leonard C. (Albuquerque, NM); Panitz, Janda K. G. (Edgewood, NM); Sharp, Donald J. (Albuquerque, NM)

    1990-01-01

    A photovoltaic assembly for converting high intensity solar radiation into lectrical energy in which a solar cell is separated from a heat sink by a thin layer of a composite material which has excellent dielectric properties and good thermal conductivity. This composite material is a thin film of porous Al.sub.2 O.sub.3 in which the pores have been substantially filled with an electrophoretically-deposited layer of a styrene-acrylate resin. This composite provides electrical breakdown strengths greater than that of a layer consisting essentially of Al.sub.2 O.sub.3 and has a higher thermal conductivity than a layer of styrene-acrylate alone.

  7. Microfluidic self assembly

    E-Print Network [OSTI]

    Bingqing Shen; Joshua Ricouvier; Mathilde Reyssat; Florent Malloggi; Patrick Tabeling

    2014-09-14

    Recent progress in colloidal science has led to elaborate self-assembled structures whose complexity raises hopes for elaborating new materials. However, the throughputs are extremely low and consequently, the chance to produce materials of industrial interest, for instance, groundbreaking optical devices, harnessing complete three-dimensional band gaps, is markedly low. We discovered a novel hydrodynamic effect that may unlock this bottleneck. It is based on the dipolar flow interactions that build up when droplets are slowed down by the microchannel walls along which they are transported. Coupled with depletion forces, we succeeded to form, via a continuous flow process, at unprecedented speeds and under exquisite control, a rich ensemble of monodisperse planar and tridimensional clusters, such as chains, triangles, diamonds, tetahedrons, heterotrimers, possessing geometrical, chemical, and/or magnetic anisotropies enabling directional bonding. Continuous productions of millions of building blocks per second for elaborating new functional materials can be envisioned.

  8. Flexible cloth seal assembly

    DOE Patents [OSTI]

    Bagepalli, B.S.; Taura, J.C.; Aksit, M.F.; Demiroglu, M.; Predmore, D.R.

    1999-06-29

    A seal assembly is described having a flexible cloth seal which includes a shim assemblage surrounded by a cloth assemblage. A first tubular end portion, such as a gas turbine combustor, includes a longitudinal axis and has smooth and spaced-apart first and second surface portions defining a notch there between which is wider at its top than at its bottom and which extends outward from the axis. The second surface portion is outside curved, and a first edge of the cloth seal is positioned in the bottom of the notch. A second tubular end portion, such as a first stage nozzle, is located near, spaced apart from, and coaxially aligned with, the first tubular end portion. The second tubular end portion has a smooth third surface portion which surrounds at least a portion of the first tubular end portion and which is contacted by the cloth seal. 7 figs.

  9. Crank shaft support assembly

    DOE Patents [OSTI]

    Natkin, Robert J. (Canton, MI); Oltmans, Bret (Stacy, MN); Allison, John E. (Ann Arbor, MI); Heater, Thomas J. (Milford, MI); Hines, Joy Adair (Plymouth, MI); Tappen, Grant K. (Washington, MI); Peiskammer, Dietmar (Rochester, MI)

    2007-10-23

    A crank shaft support assembly for increasing stiffness and reducing thermal mismatch distortion in a crank shaft bore of an engine comprising different materials. A cylinder block comprises a first material and at least two crank journal inserts are insert-molded into respective crank journal regions of the cylinder block and comprise a second material having greater stiffness and a lower thermal coefficient of expansion that the first material. At least two bearing caps are bolted to the respective crank journal inserts and define, along with the crank journal inserts, at least two crank shaft support rings defining a crank shaft bore coaxially aligned with a crank shaft axis. The bearing caps comprise a material having higher stiffness and a lower thermal coefficient of expansion than the first material and are supported on the respective crank journal inserts independently of any direct connection to the cylinder block.

  10. Advanced gray rod control assembly

    DOE Patents [OSTI]

    Drudy, Keith J; Carlson, William R; Conner, Michael E; Goldenfield, Mark; Hone, Michael J; Long, Jr., Carroll J; Parkinson, Jerod; Pomirleanu, Radu O

    2013-09-17

    An advanced gray rod control assembly (GRCA) for a nuclear reactor. The GRCA provides controlled insertion of gray rod assemblies into the reactor, thereby controlling the rate of power produced by the reactor and providing reactivity control at full power. Each gray rod assembly includes an elongated tubular member, a primary neutron-absorber disposed within the tubular member said neutron-absorber comprising an absorber material, preferably tungsten, having a 2200 m/s neutron absorption microscopic capture cross-section of from 10 to 30 barns. An internal support tube can be positioned between the primary absorber and the tubular member as a secondary absorber to enhance neutron absorption, absorber depletion, assembly weight, and assembly heat transfer characteristics.

  11. Shape Allophiles Improve Entropic Assembly

    E-Print Network [OSTI]

    Eric S. Harper; Ryan L. Marson; Joshua A. Anderson; Greg van Anders; Sharon C. Glotzer

    2015-06-25

    We investigate a class of "shape allophiles" that fit together like puzzle pieces as a method to access and stabilize desired structures by controlling directional entropic forces. Squares are cut into rectangular halves, which are shaped in an allophilic manner with the goal of re-assembling the squares while self-assembling the square lattice. We examine the assembly characteristics of this system via the potential of mean force and torque, and the fraction of particles that entropically bind. We generalize our findings and apply them to self-assemble triangles into a square lattice via allophilic shaping. Through these studies we show how shape allophiles can be useful in assembling and stabilizing desired phases with appropriate allophilic design.

  12. Molecular Self-Assembly

    SciTech Connect (OSTI)

    CURRO, JOHN G.; MCCOY, JOHN DWANE; FRISCHKNECHT, AMALIE L.; YU, KUI

    2001-11-01

    This report is divided into two parts: a study of the glass transition in confined geometries, and formation mechanisms of block copolymer mesophases by solvent evaporation-induced self-assembly. The effect of geometrical confinement on the glass transition of polymers is a very important consideration for applications of polymers in nanotechnology applications. We hypothesize that the shift of the glass transition temperature of polymers in confined geometries can be attributed to the inhomogeneous density profile of the liquid. Accordingly, we assume that the glass temperature in the inhomogeneous state can be approximated by the Tg of a corresponding homogeneous, bulk polymer, but at a density equal to the average density of the inhomogeneous system. Simple models based on this hypothesis give results that are in remarkable agreement with experimental measurements of the glass transition of confined liquids. Evaporation-induced self-assembly (EISA) of block copolymers is a versatile process for producing novel, nanostructured materials and is the focus of much of the experimental work at Sandia in the Brinker group. In the EISA process, as the solvent preferentially evaporates from a cast film, two possible scenarios can occur: microphase separation or micellization of the block copolymers in solution. In the present investigation, we established the conditions that dictate which scenario takes place. Our approach makes use of scaling arguments to determine whether the overlap concentration c* occurs before or after the critical micelle concentration (CMC). These theoretical arguments are used to interpret recent experimental results of Yu and collaborators on EISA experiments on Silica/PS-PEO systems.

  13. Method for monitoring irradiated fuel using Cerenkov radiation

    DOE Patents [OSTI]

    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.

  14. ASSEMBLY TRANSFER SYSTEM DESCRIPTION DOCUMENT

    SciTech Connect (OSTI)

    B. Gorpani

    2000-06-26

    The Assembly Transfer System (ATS) receives, cools, and opens rail and truck transportation casks from the Carrier/Cask Handling System (CCHS). The system unloads transportation casks consisting of bare Spent Nuclear Fuel (SNF) assemblies, single element canisters, and Dual Purpose Canisters (DPCs). For casks containing DPCs, the system opens the DPCs and unloads the SNF. The system stages the assemblies, transfer assemblies to and from fuel-blending inventory pools, loads them into Disposal Containers (DCs), temporarily seals and inerts the DC, decontaminates the DC and transfers it to the Disposal Container Handling System. The system also prepares empty casks and DPCs for off-site shipment. Two identical Assembly Transfer System lines are provided in the Waste Handling Building (WHB). Each line operates independently to handle the waste transfer throughput and to support maintenance operations. Each system line primarily consists of wet and dry handling areas. The wet handling area includes a cask transport system, cask and DPC preparation system, and a wet assembly handling system. The basket transport system forms the transition between the wet and dry handling areas. The dry handling area includes the dry assembly handling system, assembly drying system, DC preparation system, and DC transport system. Both the wet and dry handling areas are controlled by the control and tracking system. The system operating sequence begins with moving transportation casks to the cask preparation area. The cask preparation operations consist of cask cavity gas sampling, cask venting, cask cool-down, outer lid removal, and inner shield plug lifting fixture attachment. Casks containing bare SNF (no DPC) are filled with water and placed in the cask unloading pool. The inner shield plugs are removed underwater. For casks containing a DPC, the cask lid(s) is removed, and the DPC is penetrated, sampled, vented, and cooled. A DPC lifting fixture is attached and the cask is placed into the cask unloading pool. In the cask unloading pool the DPC is removed from the cask and placed in an overpack and the DPC lid is severed and removed. Assemblies are removed from either an open cask or DPC and loaded into assembly baskets positioned in the basket staging rack in the assembly unloading pool. A method called ''blending'' is utilized to load DCs with a heat output of less than 11.8 kW. This involves combining hotter and cooler assemblies from different baskets. Blending requires storing some of the hotter fuel assemblies in fuel-blending inventory pools until cooler assemblies are available. The assembly baskets are then transferred from the basket staging rack to the assembly handling cell and loaded into the assembly drying vessels. After drying, the assemblies are removed from the assembly drying vessels and loaded into a DC positioned below the DC load port. After installation of a DC inner lid and temporary sealing device, the DC is transferred to the DC decontamination cell where the top area of the DC, the DC lifting collar, and the DC inner lid and temporary sealing device are decontaminated, and the DC is evacuated and backfilled with inert gas to prevent prolonged clad exposure to air. The DC is then transferred to the Disposal Container Handling System for lid welding. In another cask preparation and decontamination area, lids are replaced on the empty transportation casks and DPC overpacks, the casks and DPC overpacks are decontaminated, inspected, and transferred to the Carrier/Cask Handling System for shipment off-site. All system equipment is designed to facilitate manual or remote operation, decontamination, and maintenance. The system interfaces with the Carrier/Cask Handling System for incoming and outgoing transportation casks and DPCs. The system also interfaces with the Disposal Container Handling System, which prepares the DC for loading and subsequently seals the loaded DC. The system support interfaces are the Waste Handling Building System and other internal WHB support systems.

  15. Fission product release from irradiated LWR fuel under accident conditions

    SciTech Connect (OSTI)

    Strain, R.V.; Sanecki, J.E.; Osborne, M.F.

    1984-01-01

    Fission product release from irradiated LWR fuel is being studied by heating fuel rod segments in flowing steam and an inert carrier gas to simulate accident conditions. Fuels with a range of irradiation histories are being subjected to several steam flow rates over a wide range of temperatures. Fission product release during each test is measured by gamma spectroscopy and by detailed examination of the collection apparatus after the test has been completed. These release results are complemented by a detailed posttest examination of samples of the fuel rod segment. Results of release measurements and fuel rod characterizations for tests at 1400 through 2000/sup 0/C are presented in this paper.

  16. Hybrid reduced order modeling for assembly calculations

    SciTech Connect (OSTI)

    Bang, Y.; Abdel-Khalik, H. S.; Jessee, M. A.; Mertyurek, U.

    2013-07-01

    While the accuracy of assembly calculations has considerably improved due to the increase in computer power enabling more refined description of the phase space and use of more sophisticated numerical algorithms, the computational cost continues to increase which limits the full utilization of their effectiveness for routine engineering analysis. Reduced order modeling is a mathematical vehicle that scales down the dimensionality of large-scale numerical problems to enable their repeated executions on small computing environment, often available to end users. This is done by capturing the most dominant underlying relationships between the model's inputs and outputs. Previous works demonstrated the use of the reduced order modeling for a single physics code, such as a radiation transport calculation. This manuscript extends those works to coupled code systems as currently employed in assembly calculations. Numerical tests are conducted using realistic SCALE assembly models with resonance self-shielding, neutron transport, and nuclides transmutation/depletion models representing the components of the coupled code system. (authors)

  17. Assembly and electrical transport characterization of nanostructures...

    Office of Scientific and Technical Information (OSTI)

    Conference: Assembly and electrical transport characterization of nanostructures. Citation Details In-Document Search Title: Assembly and electrical transport characterization of...

  18. Continuous wave laser irradiation of explosives

    SciTech Connect (OSTI)

    McGrane, Shawn D.; Moore, David S.

    2010-12-01

    Quantitative measurements of the levels of continuous wave (CW) laser light that can be safely applied to bare explosives during contact operations were obtained at 532 nm, 785 nm, and 1550 nm wavelengths. A thermal camera was used to record the temperature of explosive pressed pellets and single crystals while they were irradiated using a measured laser power and laser spot size. A visible light image of the sample surface was obtained before and after the laser irradiation. Laser irradiation thresholds were obtained for the onset of any visible change to the explosive sample and for the onset of any visible chemical reaction. Deflagration to detonation transitions were not observed using any of these CW laser wavelengths on single crystals or pressed pellets in the unconfined geometry tested. Except for the photochemistry of DAAF, TATB and PBX 9502, all reactions appeared to be thermal using a 532 nm wavelength laser. For a 1550 nm wavelength laser, no photochemistry was evident, but the laser power thresholds for thermal damage in some of the materials were significantly lower than for the 532 nm laser wavelength. No reactions were observed in any of the studied explosives using the available 300 mW laser at 785 nm wavelength. Tables of laser irradiance damage and reaction thresholds are presented for pressed pellets of PBX9501, PBX9502, Composition B, HMX, TATB, RDX, DAAF, PETN, and TNT and single crystals of RDX, HMX, and PETN for each of the laser wavelengths.

  19. Bearing assemblies, apparatuses, and motor assemblies using the same

    DOE Patents [OSTI]

    Sexton, Timothy N.; Cooley, Craig H.; Knuteson, Cody W.

    2015-12-29

    Various embodiments of the invention relate to bearing assemblies, apparatuses and motor assemblies that include geometric features configured to impart a selected amount of heat transfer and/or hydrodynamic film formation. In an embodiment, a bearing assembly may include a plurality of superhard bearing pads distributed circumferentially about an axis. At least some of the plurality of superhard bearing pads may include a plurality of sub-superhard bearing elements defining a bearing surface. At least some of the plurality of sub-superhard bearing elements may be spaced from one another by one or more voids to impart a selected amount of heat transfer and hydrodynamic film formation thereon during operation. The bearing assembly may also include a support ring that carries the plurality of superhard bearing pads. In addition, at least a portion of the sub-superhard bearing elements may extend beyond the support ring.

  20. Drive piston assembly for a valve actuator assembly

    DOE Patents [OSTI]

    Sun, Zongxuan (Troy, MI)

    2010-02-23

    A drive piston assembly is provided that is operable to selectively open a poppet valve. The drive piston assembly includes a cartridge defining a generally stepped bore. A drive piston is movable within the generally stepped bore and a boost sleeve is coaxially disposed with respect to the drive piston. A main fluid chamber is at least partially defined by the generally stepped bore, drive piston, and boost sleeve. First and second feedback chambers are at least partially defined by the drive piston and each are disposed at opposite ends of the drive piston. At least one of the drive piston and the boost sleeve is sufficiently configured to move within the generally stepped bore in response to fluid pressure within the main fluid chamber to selectively open the poppet valve. A valve actuator assembly and engine are also provided incorporating the disclosed drive piston assembly.

  1. Cooling assembly for fuel cells

    DOE Patents [OSTI]

    Kaufman, Arthur (West Orange, NJ); Werth, John (Princeton, NJ)

    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.

  2. Characterization of LWRS Hybrid SiC-CMC-Zircaloy-4 Fuel Cladding after Gamma Irradiation

    SciTech Connect (OSTI)

    Isabella J van Rooyen

    2012-09-01

    The purpose of the gamma irradiation tests conducted at the Idaho National Laboratory (INL) was to obtain a better understanding of chemical interactions and potential changes in microstructural properties of a mock-up hybrid nuclear fuel cladding rodlet design (unfueled) in a simulated PWR water environment under irradiation conditions. The hybrid fuel rodlet design is being investigated under the Light Water Reactor Sustainability (LWRS) program for further development and testing of one of the possible advanced LWR nuclear fuel cladding designs. The gamma irradiation tests were performed in preparation for neutron irradiation tests planned for a silicon carbide (SiC) ceramic matrix composite (CMC) zircaloy-4 (Zr-4) hybrid fuel rodlet that may be tested in the INL Advanced Test Reactor (ATR) if the design is selected for further development and testing

  3. Evaluation of Concepts for Mulitiple Application Thermal Reactor for Irradiation eXperiments (MATRIX)

    SciTech Connect (OSTI)

    Michael A. Pope; Hans D. Gougar; John M. Ryskamp

    2013-09-01

    The Advanced Test Reactor (ATR) is a high power density test reactor specializing in fuel and materials irradiation. For more than 45 years, the ATR has provided irradiations of materials and fuels testing along with radioisotope production. Originally operated primarily in support of the Offcie of Naval Reactors (NR), the mission has gradually expanded to cater to other customers, such as the DOE Office of Nuclear Energy (NE), private industry, and universities. Unforeseen circumstances may lead to the decommissioning of ATR, thus leaving the U.S. Government without a large-scale materials irradiation capability to meet the needs of its nuclear energy and naval reactor missions. In anticipation of this possibility, work was performed under the Laboratory Directed Research and Development (LDRD) program to investigate test reactor concepts that could satisfy the current missions of the ATR along with an expanded set of secondary missions. This work can be viewed as an update to a project from the 1990’s called the Broad Application Test Reactor (BATR). In FY 2012, a survey of anticipated customer needs was performed, followed by analysis of the original BATR concepts with fuel changed to low-enriched uranium. Departing from these original BATR designs, four concepts were identified for further analysis in FY2013. The project informally adopted the acronym MATRIX (Multiple-Application Thermal Reactor for Irradiation eXperiments). This report discusses analysis of the four MATRIX concepts along with a number of variations on these main concepts. Designs were evaluated based on their satisfaction of anticipated customer requirements and the “Cylindrical” variant was selected for further analysis of options. This downselection should be considered preliminary and the backup alternatives should include the other three main designs. The baseline Cylindrical MATRIX design is expected to be capable of higher burnup than the ATR (or longer cycle length given a particular batch scheme). The volume of test space in IPTs is larger in MATRIX than in ATR with comparable magnitude of neutron flux. In addition to the IPTs, the Cylindrical MATRIX concept features test spaces at the centers of fuel assemblies where very high fast flux can be achieved. This magnitude of fast flux is similar to that achieved in the ATR A-positions, however, the available volume having these conditions is greater in the MATRIX design than in the ATR. From the analyses performed in this work, it appears that the Cylindrical MATRIX design can be designed to meet the anticipated needs of the ATR replacement reactor. However, this statement must be qualified by acknowledging that this design is quite immature, and therefore any requirements currently met must be re-evaluated as the design matures. Also, some of the requirements were not strictly met, but are believed to be achievable once features to be added later are designed.

  4. Concentric tube support assembly

    DOE Patents [OSTI]

    Rubio, Mark F.; Glessner, John C.

    2012-09-04

    An assembly (45) includes a plurality of separate pie-shaped segments (72) forming a disk (70) around a central region (48) for retaining a plurality of tubes (46) in a concentrically spaced apart configuration. Each segment includes a support member (94) radially extending along an upstream face (96) of the segment and a plurality of annularly curved support arms (98) transversely attached to the support member and radially spaced apart from one another away from the central region for receiving respective upstream end portions of the tubes in arc-shaped spaces (100) between the arms. Each segment also includes a radial passageway (102) formed in the support member for receiving a fluid segment portion (106) and a plurality of annular passageways (104) formed in the support arms for receiving respective arm portions (108) of the fluid segment portion from the radial passageway and for conducting the respective arm portions into corresponding annular spaces (47) formed between the tubes retained by the disk.

  5. Understanding the Irradiation Behavior of Zirconium Carbide

    SciTech Connect (OSTI)

    Motta, Arthur; Sridharan, Kumar; Morgan, Dane; Szlufarska, Izabela

    2013-10-11

    Zirconium carbide (ZrC) is being considered for utilization in high-temperature gas-cooled reactor fuels in deep-burn TRISO fuel. Zirconium carbide possesses a cubic B1-type crystal structure with a high melting point, exceptional hardness, and good thermal and electrical conductivities. The use of ZrC as part of the TRISO fuel requires a thorough understanding of its irradiation response. However, the radiation effects on ZrC are still poorly understood. The majority of the existing research is focused on the radiation damage phenomena at higher temperatures (>450{degree}C) where many fundamental aspects of defect production and kinetics cannot be easily distinguished. Little is known about basic defect formation, clustering, and evolution of ZrC under irradiation, although some atomistic simulation and phenomenological studies have been performed. Such detailed information is needed to construct a model describing the microstructural evolution in fast-neutron irradiated materials that will be of great technological importance for the development of ZrC- based fuel. The goal of the proposed project is to gain fundamental understanding of the radiation-induced defect formation in zirconium carbide and irradiation response (ZrC) by using a combination of state-of-the-art experimental methods and atomistic modeling. This project will combine (1) in situ ion irradiation at a specialized facility at a national laboratory, (2) controlled temperature proton irradiation on bulk samples, and (3) atomistic modeling to gain a fundamental understanding of defect formation in ZrC. The proposed project will cover the irradiation temperatures from cryogenic temperature to as high as 800{degree}C, and dose ranges from 0.1 to 100 dpa. The examination of this wide range of temperatures and doses allows us to obtain an experimental data set that can be effectively used to exercise and benchmark the computer calculations of defect properties. Combining the examination of radiation-induced microstructures mapped spatially and temporally, microstructural evolution during post-irradiation annealing, and atomistic modeling of defect formation and transport energetics will provide new, critical understanding about property changes in ZrC. The behavior of materials under irradiation is determined by the balance between damage production, defect clustering, and lattice response. In order to predict those effects at high temperatures so targeted testing can be expanded and extrapolated beyond the known database, it is necessary to determine the defect energetics and mobilities as these control damage accumulation and annealing. In particular, low-temperature irradiations are invaluable for determining the regions of defect mobility. Computer simulation techniques are particularly useful for identifying basic defect properties, especially if closely coupled with a well-constructed and complete experimental database. The close coupling of calculation and experiment in this project will provide mutual benchmarking and allow us to glean a deeper understanding of the irradiation response of ZrC, which can then be applied to the prediction of its behavior in reactor conditions.

  6. Additive assembly of digital materials

    E-Print Network [OSTI]

    Ward, Jonathan (Jonathan Daniel)

    2010-01-01

    This thesis develops the use of additive assembly of press-fit digital materials as a new rapid-prototyping process. Digital materials consist of a finite set of parts that have discrete connections and occupy discrete ...

  7. Magnet Girder Assembly and Installation

    ScienceCinema (OSTI)

    None

    2013-07-17

    It takes teamwork to assemble and install magnet girders for the storage ring of the National Synchrotron Light Source II. NSLS-II is now under construction at Brookhaven Lab.

  8. Direct hierarchical assembly of nanoparticles

    SciTech Connect (OSTI)

    Xu, Ting; Zhao, Yue; Thorkelsson, Kari

    2014-07-22

    The present invention provides hierarchical assemblies of a block copolymer, a bifunctional linking compound and a nanoparticle. The block copolymers form one micro-domain and the nanoparticles another micro-domain.

  9. AGC-1 Pre-Irradiation Data Report Status

    SciTech Connect (OSTI)

    William Windes

    2011-08-01

    The Next Generation Nuclear Plant (NGNP) Graphite R&D program is currently measuring irradiated material property changes in several grades of nuclear graphite for predicting their behavior and operating performance within the core of new Very High Temperature Reactor (VHTR) designs. The Advanced Graphite Creep (AGC) experiment consisting of six irradiation capsules will generate this irradiated graphite performance data for NGNP reactor operating conditions. All samples in the experiment will be fully characterized before irradiation, irradiated in the Advanced Test Reactor (ATR), and then re-examined to determine the irradiation induced changes to key materials properties in the different graphite grades. The information generated during the AGC experiment will be utilized for NRC licensing of NGNP reactor designs, shared with international collaborators in the Generation IV Information Forum (GIF), and eventually utilized in ASME design code for graphite nuclear applications. This status report will describe the process the NGNP Graphite R&D program has developed to record the AGC1 pre-irradiation examination data.

  10. Los Alamos Critical Assemblies Facility

    SciTech Connect (OSTI)

    Malenfant, R.E.

    1981-06-01

    The Critical Assemblies Facility of the Los Alamos National Laboratory has been in existence for thirty-five years. In that period, many thousands of measurements have been made on assemblies of /sup 235/U, /sup 233/U, and /sup 239/Pu in various configurations, including the nitrate, sulfate, fluoride, carbide, and oxide chemical compositions and the solid, liquid, and gaseous states. The present complex of eleven operating machines is described, and typical applications are presented.

  11. Wafer scale micromachine assembly method

    DOE Patents [OSTI]

    Christenson, Todd R. (Albuquerque, NM)

    2001-01-01

    A method for fusing together, using diffusion bonding, micromachine subassemblies which are separately fabricated is described. A first and second micromachine subassembly are fabricated on a first and second substrate, respectively. The substrates are positioned so that the upper surfaces of the two micromachine subassemblies face each other and are aligned so that the desired assembly results from their fusion. The upper surfaces are then brought into contact, and the assembly is subjected to conditions suited to the desired diffusion bonding.

  12. Microelectromechanical system assembled ion optics: An advance to miniaturization and assembly of electron and ion optics

    SciTech Connect (OSTI)

    Fox, J.; Verbeck, G. [University of North Texas, Denton, Texas 76203 (United States); Saini, R.; Tsui, K. [Zyvex, Richardson, Texas 75081 (United States)

    2009-09-15

    Deep-reactive ion etching of n-doped silicon-on-insulator is utilized to make ion optical components to aid in the miniaturization of mass analyzers. The microelectromechanical system components are bound to aluminum nitride substrates and employed three-dimensional assembly. The assembly methods are tested for breakdown (V{sub b}), durability, and alignment. Demonstration of ion manipulation is shown with a 1 mm Bradbury-Nielsen gate, 500 {mu}m Einzel lens, 500 {mu}m coaxial ring ion trap, and reflectron optics. Data are presented showing the resolution, attenuation, and performance of each of these devices. We demonstrate advantages and disadvantages of this technology and its applications to mass analysis.

  13. Heat exchange assembly

    DOE Patents [OSTI]

    Lowenstein, Andrew; Sibilia, Marc; Miller, Jeffrey; Tonon, Thomas S.

    2004-06-08

    A heat exchange assembly comprises a plurality of plates disposed in a spaced-apart arrangement, each of the plurality of plates includes a plurality of passages extending internally from a first end to a second end for directing flow of a heat transfer fluid in a first plane, a plurality of first end-piece members equaling the number of plates and a plurality of second end-piece members also equaling the number of plates, each of the first and second end-piece members including a recessed region adapted to fluidly connect and couple with the first and second ends of the plate, respectively, and further adapted to be affixed to respective adjacent first and second end-piece members in a stacked formation, and each of the first and second end-piece members further including at least one cavity for enabling entry of the heat transfer fluid into the plate, exit of the heat transfer fluid from the plate, or 180.degree. turning of the fluid within the plate to create a serpentine-like fluid flow path between points of entry and exit of the fluid, and at least two fluid conduits extending through the stacked plurality of first and second end-piece members for providing first fluid connections between the parallel fluid entry points of adjacent plates and a fluid supply inlet, and second fluid connections between the parallel fluid exit points of adjacent plates and a fluid discharge outlet so that the heat transfer fluid travels in parallel paths through each respective plate.

  14. Hear Exchange Assembly

    DOE Patents [OSTI]

    Lowenstein, Andrew (Princeton, NJ); Sibilia, Marc (Princeton, NJ); Miller, Jeffrey (Rocky Hill, NJ); Tonon, Thomas S. (Princeton, NJ)

    2003-05-27

    A heat exchange assembly comprises a plurality of plates disposed in a spaced-apart arrangement, each of the plurality of plates includes a plurality of passages extending internally from a first end to a second end for directing flow of a heat transfer fluid in a first plane, a plurality of first end-piece members equaling the number of plates and a plurality of second end-piece members also equaling the number of plates, each of the first and second end-piece members including a recessed region adapted to fluidly connect and couple with the first and second ends of the plate, respectively, and further adapted to be affixed to respective adjacent first and second end-piece members in a stacked formation, and each of the first and second end-piece members further including at least one cavity for enabling entry of the heat transfer fluid into the plate, exit of the heat transfer fluid from the plate, or 180.degree. turning of the fluid within the plate to create a serpentine-like fluid flow path between points of entry and exit of the fluid, and at least two fluid conduits extending through the stacked plurality of first and second end-piece members for providing first fluid connections between the parallel fluid entry points of adjacent plates and a fluid supply inlet, and second fluid connections between the parallel fluid exit points of adjacent plates and a fluid discharge outlet so that the heat transfer fluid travels in parallel paths through each respective plate.

  15. Cutaneous fusariosis developing in a post-irradiation site

    E-Print Network [OSTI]

    2011-01-01

    developing in a post-irradiation site WeiWei Dai DO 1 ,irradiated site. Irradiation led to atrophy, contraction,radiotherapy. Following irradiation, which had eradicated

  16. Carbon Characterization Laboratory Readiness to Receive Irradiated Graphite Samples

    SciTech Connect (OSTI)

    Karen A. Moore

    2011-05-01

    The Carbon Characterization Laboratory (CCL) is located in Labs C19 and C20 of the Idaho National Laboratory Research Center. The CCL was established under the Next Generation Nuclear Plant Project to support graphite and ceramic composite research and development activities. The research conducted in this laboratory will support the Advanced Graphite Creep experiments—a major series of material irradiation experiments within the Next Generation Nuclear Plant Graphite program. The CCL is designed to characterize and test low activated irradiated materials such as high purity graphite, carbon-carbon composites, silicon-carbide composite, and ceramic materials. The laboratory is fully capable of characterizing material properties for both irradiated and nonirradiated materials. Major infrastructural modifications were undertaken to support this new radiological facility at Idaho National Laboratory. Facility modifications are complete, equipment has been installed, radiological controls and operating procedures have been established and work management documents have been created to place the CCL in readiness to receive irradiated graphite samples.

  17. Irradiated Beryllium Disposal Workshop, Idaho Falls, ID, May 29-30, 2002

    SciTech Connect (OSTI)

    Longhurst, Glen Reed; Anderson, Gail; Mullen, Carlan K; West, William Howard

    2002-07-01

    In 2001, while performing routine radioactive decay heat rate calculations for beryllium reflector blocks for the Advanced Test Reactor (ATR), it became evident that there may be sufficient concentrations of transuranic isotopes to require classification of this irradiated beryllium as transuranic waste. Measurements on samples from ATR reflector blocks and further calculations confirmed that for reflector blocks and outer shim control cylinders now in the ATR canal, transuranic activities are about five times the threshold for classification. That situation implies that there is no apparent disposal pathway for this material. The problem is not unique to the ATR. The High Flux Isotope Reactor at Oak Ridge National Laboratory, the Missouri University Research Reactor at Columbia, Missouri and other reactors abroad must also deal with this issue. A workshop was held in Idaho Falls Idaho on May 29-30, 2002 to acquaint stakeholders with these findings and consider a path forward in resolving the issues attendant to disposition of irradiated material. Among the findings from this workshop were (1) there is a real potential for the US to be dependent on foreign sources for metallic beryllium within about a decade; (2) there is a need for a national policy on beryllium utilization and disposition and for a beryllium coordinating committee to be assembled to provide guidance on that policy; (3) it appears it will be difficult to dispose of this material at the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico due to issues of Defense classification, facility radioactivity inventory limits, and transportation to WIPP; (4) there is a need for a funded DOE program to seek resolution of these issues including research on processing techniques that may make this waste acceptable in an existing disposal pathway or allow for its recycle.

  18. Self-assembling peptide hydrogels modulate in vitro chondrogenesis of bovine bone marrow stromal cells

    E-Print Network [OSTI]

    Kopesky, Paul Wayne

    Our objective was to test the hypothesis that self-assembling peptide hydrogel scaffolds provide cues that enhance the chondrogenic differentiation of bone marrow stromal cells (BMSCs). BMSCs were encapsulated within two ...

  19. Irradiance or luminosity changes?

    E-Print Network [OSTI]

    Sabatino Sofia; Linghuai H. Li

    2000-10-20

    Whereas a variation of the solar luminosity, L, will inevitably cause a similar change of the total solar irradiance, S, the opposite is not true. In fact, the bulk of the days to months variations of S can be explained entirely in terms of the passage of active regions across the solar disk. In this case, L remains essentially unchanged. For the total irradiance variation observed over the solar cycle, the issue is more uncertain. One view explains this modulation primarily as a combined action of active regions and magnetic network. These components would be superposed to an otherwise unchanging photosphere. the other view suggests that the activity cycle modulation of S is primarily produced by a variation of L (both in terms of R and T_{eff}) caused by structural reajustments of the interior of the Sun induced by a changing magnetic field. We will present evidence in support of this second interpretation, and a model for it. We will also present the S variations over the last 5 centuries implied by our model.

  20. Irradiated Microsphere Gamma Analyzer for Examination of Particle Fuel

    SciTech Connect (OSTI)

    Paul A. Demkowicz; Various

    2014-06-01

    Fabrication of the first series of fuel compacts for the current US tristructural isotropic (TRISO) coated particle fuel development and qualification effort was completed at Oak Ridge National Laboratory (ORNL) in 2006. In November of 2009, after almost 3 years and 620 effective full power days of irradiation in the Advanced Test Reactor at Idaho National Laboratory (INL), the first Advanced Gas Reactor irradiation test (AGR-1) was concluded. Compacts were irradiated at a calculated timeaveraged, volume-averaged temperature of 955–1136°C to a burnup ranging from 11.2–19.5% fissions per initial metal atom and a total fast fluence of 2.2–4.3·1025 n/m2 [1]. No indication of fission product release from TRISO coating failure was observed during the irradiation test, based on real-time monitoring of gaseous fission products. Post-irradiation examination (PIE) and hightemperature safety testing of the compacts has been in progress at both ORNL and INL since 2010, and have revealed small releases of a limited subset of fission products (such as silver, cesium, and europium). Past experience has shown that some elements can be released from TRISO particles when a defect forms in the SiC layer, even when one or more pyrocarbon layers remain intact and retain the gaseous fission products. Some volatile elements can also be released by diffusion through an intact SiC layer during safety testing if temperatures are high enough and the duration is long enough. In order to understand and quantify the release of certain radioactive fission products, it is sometimes necessary to individually examine each of the more than 4000 coated particles in a given compact. The Advanced Irradiated Microsphere Gamma Analyzer (Advanced- IMGA) was designed to perform this task in a remote hot cell environment. This paper describes the Advanced- IMGA equipment and examination process and gives results for a typical full compact evaluation.

  1. A framework for assembly sequence planning for computer aided design of mechanical assemblies 

    E-Print Network [OSTI]

    Cheboli, Ramakrishna

    2000-01-01

    This thesis presents a framework for interactive assembly sequence panning for mechanical assemblies. Realizing the utility of such a tool that will enable the product engineer to evaluate the assemblability of his designs and generate suitable...

  2. SPERMATOGONIAL STEM CELL RENEWAL FOLLOWING IRRADIATION

    E-Print Network [OSTI]

    Fabrikant, J.I.

    2010-01-01

    continuou-: lnw-l. -vel -,-irradiation for 15 wk, the curvesUnder Continuous Irradiation Cell Cycle Parameter (hr) >J. I. , Pathology of Irradiation, WiTiams and hiHins,

  3. MFR PAPER 1135 Irradiation of

    E-Print Network [OSTI]

    the feasibility of combining low-dose ionizing radiation of shrimp with tur- meric and salt as a preservativeMFR PAPER 1135 Irradiation of shrimp extends their storage life. Low Level Ionizing Radiation are eliminated effectively and selectively by low-level doses of irradiation, thereby resulting in extension

  4. ZEST flight test experiments, Kauai Test Facility, Hawaii. Test report

    SciTech Connect (OSTI)

    Cenkci, M.J.

    1991-07-01

    The Strategic Defense Initiative Organization (SDIO) is proposing to execute two ZEST flight experiments to obtain information related to the following objectives: validation of payload modeling; characterization of a high energy release cloud; and documentation of scientific phenomena that may occur as a result of releasing a high energy cloud. The proposed action is to design, develop, launch, and detonate two payloads carrying high energy explosives. Activities required to support this proposal include: (1) execution of component assembly tests at Space Data Division (SDD) in Chandler, Arizona and Los Alamos National Laboratory (LANL) in Los Alamos, New Mexico, and (2) execution of pre-flight flight test activities at Kauai Test Facility.

  5. Fixture for aligning motor assembly

    DOE Patents [OSTI]

    Shervington, Roger M.; Vaghani, Vallabh V.; Vanek, Laurence D.; Christensen, Scott A.

    2009-12-08

    An alignment fixture includes a rotor fixture, a stator fixture and a sensor system which measures a rotational displacement therebetween. The fixture precisely measures rotation of a generator stator assembly away from a NULL position referenced by a unique reference spline on the rotor shaft. By providing an adjustable location of the stator assembly within the housing, the magnetic axes within each generator shall be aligned to a predetermined and controlled tolerance between the generator interface mounting pin and the reference spline on the rotor shaft. Once magnetically aligned, each generator is essentially a line replaceable unit which may be readily mounted to any input of a multi-generator gearbox assembly with the assurance that the magnetic alignment will be within a predetermined tolerance.

  6. Dynamics of assembly production flow

    E-Print Network [OSTI]

    Ezaki, Takahiro; Nishinari, Katsuhiro

    2015-01-01

    Despite recent developments in management theory, maintaining a manufacturing schedule remains difficult because of production delays and fluctuations in demand and supply of materials. The response of manufacturing systems to such disruptions to dynamic behavior has been rarely studied. To capture these responses, we investigate a process that models the assembly of parts into end products. The complete assembly process is represented by a directed tree, where the smallest parts are injected at leaves and the end products are removed at the root. A discrete assembly process, represented by a node on the network, integrates parts, which are then sent to the next downstream node as a single part. The model exhibits some intriguing phenomena, including overstock cascade, phase transition in terms of demand and supply fluctuations, nonmonotonic distribution of stockout in the network, and the formation of a stockout path and stockout chains. Surprisingly, these rich phenomena result from only the nature of distr...

  7. Insulation assembly for electric machine

    DOE Patents [OSTI]

    Rhoads, Frederick W.; Titmuss, David F.; Parish, Harold; Campbell, John D.

    2013-10-15

    An insulation assembly is provided that includes a generally annularly-shaped main body and at least two spaced-apart fingers extending radially inwards from the main body. The spaced-apart fingers define a gap between the fingers. A slot liner may be inserted within the gap. The main body may include a plurality of circumferentially distributed segments. Each one of the plurality of segments may be operatively connected to another of the plurality of segments to form the continuous main body. The slot liner may be formed as a single extruded piece defining a plurality of cavities. A plurality of conductors (extendable from the stator assembly) may be axially inserted within a respective one of the plurality of cavities. The insulation assembly electrically isolates the conductors in the electric motor from the stator stack and from other conductors.

  8. Abrasive swivel assembly and method

    DOE Patents [OSTI]

    Hashish, Mohamed (Kent, WA); Marvin, Mark (Tacoma, WA)

    1990-01-01

    An abrasive swivel assembly for providing a rotating, particle-laden fluid stream and, ultimately, a rotating particle-laden fluid jet is disclosed herein. This assembly includes a tubular arrangement for providing a particle-free stream of fluid, a swivel assembly for rotating a section of the tubular arrangement, and a tubular end section for introducing solid particles into the particle-free fluid stream at a point along the rotating tubular section, whereby to produce a particle-laden fluid stream. This last-mentioned stream can then be used in combination with a cooperating nozzle arrangement for providing a rotating particle-laden fluid jet. In an actual working embodiment, the fluid stream is of sufficiently high pressure so that the abrasive jet can be used as a cutting jet.

  9. Defining the early steps in nuclear pore assembly : chromatin-associated ELYS initiates pore assembly

    E-Print Network [OSTI]

    Rasala, Beth A.

    2008-01-01

    of membrane fusion……………… 121 Figure 3.2: Cold temperaturefusion inhibitor LPC prevents FG Nups assembly into the coldfusion inhibitor LPC prevents FG Nups assembly into the cold

  10. Final Report on MEGAPIE Target Irradiation and Post-Irradiation Examination

    SciTech Connect (OSTI)

    Yong, Dai

    2015-06-30

    Megawatt pilot experiment (MEGAPIE) was successfully performed in 2006. One of the important goals of MEGAPIE is to understand the behaviour of structural materials of the target components exposed to high fluxes of high-energy protons and spallation neutrons in flowing LBE (liquid lead-bismuth eutectic) environment by conducting post-irradiation examination (PIE). The PIE includes four major parts: non-destructive test, radiochemical analysis of production and distribution of radionuclides produced by spallation reaction in LBE, analysis of LBE corrosion effects on structural materials, T91 and SS 316L steels, and mechanical testing of the T91 and SS 316L steels irradiated in the lower part of the target. The non-destructive test (NDT) including visual inspection and ultrasonic measurement was performed in the proton beam window area of the T91 calotte of the LBE container, the most intensively irradiated part of the MEGAPIE target. The visual inspection showed no visible failure and the ultrasonic measurement demonstrated no detectable change in thickness in the beam window area. Gamma mapping was also performed in the proton beam window area of the AlMg3 safety-container. The gamma mapping results were used to evaluate the accumulated proton fluence distribution profile, the input data for determining irradiation parameters. Radiochemical analysis of radionuclides produced by spallation reaction in LBE is to improve the understanding of the production and distribution of radionuclides in the target. The results demonstrate that the radionuclides of noble metals, 207Bi, 194Hg/Au are rather homogeneously distributed within the target, while radionuclides of electropositive elements are found to be deposited on the steel-LBE interface. The corrosion effect of LBE on the structural components under intensive irradiation was investigated by metallography. The results show that no evident corrosion damages. However, unexpected deep cracks were found in the EBW (electron beam weld) of the LBE container in the intensive irradiation zone of the target, which should be formed during irradiation. In the SS 316L steel of the flow guide tube, inclusions or precipitates enriched with O, Si, S, Ca, Ti and Mn were observed. Many of them are very long, up to a few mm, and located on grain boundaries along the extrusion direction of the tube. The degradation of the mechanical properties of the T91 and SS 316L steels has been investigated by conducting tensile tests on the specimens extracted from the T91 and SS 316L components in the intensive irradiation region. The results obtained from the proton beam window of the T91 calotte exhibit a good ductility of T91 steel after irradiation at 6-7 dpa (displacement per atom) in contact with flowing LBE.

  11. Normal Conditions of Transport Truck Test of a Surrogate Fuel...

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

    Normal Conditions of Transport Truck Test of a Surrogate Fuel Assembly. McConnell, Paul E.; Wauneka, Robert; Saltzstein, Sylvia J.; Sorenson, Ken B. Abstract not provided. Sandia...

  12. Advanced fuel assembly characterization capabilities based on gamma tomography at the Halden boiling water reactor

    SciTech Connect (OSTI)

    Holcombe, S.; Eitrheim, K.; Svaerd, S. J.; Hallstadius, L.; Willman, C.

    2012-07-01

    Characterization of individual fuel rods using gamma spectroscopy is a standard part of the Post Irradiation Examinations performed on experimental fuel at the Halden Boiling Water Reactor. However, due to handling and radiological safety concerns, these measurements are presently carried out only at the end of life of the fuel, and not earlier than several days or weeks after its removal from the reactor core. In order to enhance the fuel characterization capabilities at the Halden facilities, a gamma tomography measurement system is now being constructed, capable of characterizing fuel assemblies on a rod-by-rod basis in a more timely and efficient manner. Gamma tomography for measuring nuclear fuel is based on gamma spectroscopy measurements and tomographic reconstruction techniques. The technique, previously demonstrated on irradiated commercial fuel assemblies, is capable of determining rod-by-rod information without the need to dismantle the fuel. The new gamma tomography system will be stationed close to the Halden reactor in order to limit the need for fuel transport, and it will significantly reduce the time required to perform fuel characterization measurements. Furthermore, it will allow rod-by-rod fuel characterization to occur between irradiation cycles, thus allowing for measurement of experimental fuel repeatedly during its irradiation lifetime. The development of the gamma tomography measurement system is a joint project between the Inst. for Energy Technology - OECD Halden Reactor Project, Westinghouse (Sweden), and Uppsala Univ.. (authors)

  13. DNA-guided nanoparticle assemblies

    DOE Patents [OSTI]

    Gang, Oleg; Nykypanchuk, Dmytro; Maye, Mathew; van der Lelie, Daniel

    2013-07-16

    In some embodiments, DNA-capped nanoparticles are used to define a degree of crystalline order in assemblies thereof. In some embodiments, thermodynamically reversible and stable body-centered cubic (bcc) structures, with particles occupying <.about.10% of the unit cell, are formed. Designs and pathways amenable to the crystallization of particle assemblies are identified. In some embodiments, a plasmonic crystal is provided. In some aspects, a method for controlling the properties of particle assemblages is provided. In some embodiments a catalyst is formed from nanoparticles linked by nucleic acid sequences and forming an open crystal structure with catalytically active agents attached to the crystal on its surface or in interstices.

  14. Heavy-ion irradiation induced diamond formation in carbonaceous materials.

    SciTech Connect (OSTI)

    Daulton, T. L.

    1999-01-08

    The basic mechanisms of metastable phase formation produced under highly non-equilibrium thermodynamic conditions within high-energy particle tracks are investigated. In particular, the possible formation of diamond by heavy-ion irradiation of graphite at ambient temperature is examined. This work was motivated, in part, by earlier studies which discovered nanometer-grain polycrystalline diamond aggregates of submicron-size in uranium-rich carbonaceous mineral assemblages of Precambrian age. It was proposed that the radioactive decay of uranium formed diamond in the fission particle tracks produced in the carbonaceous minerals. To test the hypothesis that nanodiamonds can form by ion irradiation, fine-grain polycrystalline graphite sheets were irradiated with 400 MeV Kr ions. The ion irradiated graphite (and unirradiated graphite control) were then subjected to acid dissolution treatments to remove the graphite and isolate any diamonds that were produced. The acid residues were then characterized by analytical and high-resolution transmission electron microscopy. The acid residues of the ion-irradiated graphite were found to contain ppm concentrations of nanodiamonds, suggesting that ion irradiation of bulk graphite at ambient temperature can produce diamond.

  15. Test Automation Test Automation

    E-Print Network [OSTI]

    Mousavi, Mohammad

    Test Automation Test Automation Mohammad Mousavi Eindhoven University of Technology, The Netherlands Software Testing 2013 Mousavi: Test Automation #12;Test Automation Outline Test Automation Mousavi: Test Automation #12;Test Automation Why? Challenges of Manual Testing Test-case design: Choosing inputs

  16. Irradiation of commercial, high-Tc superconducting tape for potential fusion applications: electromagnetic transport properties

    SciTech Connect (OSTI)

    Aytug, Tolga [ORNL; Gapud, Albert A. [University of South Alabama, Mobile; List III, Frederick Alyious [ORNL; Leonard, Keith J [ORNL; Rupich, Marty [American Superconductor Corporation, Westborough, MA; Zhang, Yanwen [ORNL; Greenwood, N T [University of South Alabama, Mobile; Alexander, J A [University of South Alabama, Mobile; Khan, A [University of South Alabama, Mobile

    2015-01-01

    Effects of low dose irradiation on the electrical transport current properties of commercially available high-temperature superconducting, coated-conductor tapes were investigated, in view of potential applications in the irradiative environment of fusion reactors. Three different tapes, each with unique as-grown flux-pinning structures, were irradiated with Au and Ni ions at energies that provide a range of damage effects, with accumulated damage levels near that expected for conductors in a fusion reactor environment. Measurements using transport current determined the pre- and post-irradiation resistivity, critical current density, and pinning force density, yielding critical temperatures, irreversibility lines, and inferred vortex creep rates. Results show that at the irradiation damage levels tested, any detriment to as-grown pre-irradiation properties is modest; indeed in one case already-superior pinning forces are enhanced, leading to higher critical currents.

  17. MATERIAL IRRADIATION STUDIES HIGH-INTENSITY PROTON BEAM TARGETS

    E-Print Network [OSTI]

    McDonald, Kirk

    loss Strength loss/gain Fracture toughness Physical property changes CTE Diffusivity Heat capacityBeMet Nickel-Plated Aluminum Material Matrix TESTS (on-going + upcoming) Mechanical property changes Ductility (C) Strain(%) Gum (unirradiated) Gum (irradiated) GUM Metal Strengthens but clearly looses the "super-ductility

  18. Monolithic fiber optic sensor assembly

    DOE Patents [OSTI]

    Sanders, Scott

    2015-02-10

    A remote sensor element for spectrographic measurements employs a monolithic assembly of one or two fiber optics to two optical elements separated by a supporting structure to allow the flow of gases or particulates therebetween. In a preferred embodiment, the sensor element components are fused ceramic to resist high temperatures and failure from large temperature changes.

  19. Simulated nuclear reactor fuel assembly

    DOE Patents [OSTI]

    Berta, Victor T. (Idaho Falls, ID)

    1993-01-01

    An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

  20. Vacuum vapor deposition gun assembly

    DOE Patents [OSTI]

    Zeren, Joseph D. (Boulder, CO)

    1985-01-01

    A vapor deposition gun assembly includes a hollow body having a cylindrical outer surface and an end plate for holding an adjustable heat sink, a hot hollow cathode gun, two magnets for steering the plasma from the gun into a crucible on the heat sink, and a shutter for selectively covering and uncovering the crucible.

  1. Hot hollow cathode gun assembly

    DOE Patents [OSTI]

    Zeren, J.D.

    1983-11-22

    A hot hollow cathode deposition gun assembly includes a hollow body having a cylindrical outer surface and an end plate for holding an adjustable heat sink, the hot hollow cathode gun, two magnets for steering the plasma from the gun into a crucible on the heat sink, and a shutter for selectively covering and uncovering the crucible.

  2. Inorganic Materials and Assembly Techniques

    E-Print Network [OSTI]

    Rogers, John A.

    a transformational set of capabilities for high performance flexible/stretchable electronics. KEYWORDS | Bio-integrated electronics; flexible electronics; inorganic materials; integrated systems; stretchable electron- icsINVITED P A P E R Inorganic Materials and Assembly Techniques for Flexible and Stretchable

  3. Corner-cutting mining assembly

    DOE Patents [OSTI]

    Bradley, John A. (San Antonio, TX)

    1983-01-01

    A mining assembly includes a primary rotary cutter mounted on one end of a support shaft and four secondary rotary cutters carried on the same support shaft and positioned behind the primary cutters for cutting corners in the hole cut by the latter.

  4. Simulated nuclear reactor fuel assembly

    DOE Patents [OSTI]

    Berta, V.T.

    1993-04-06

    An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

  5. Wesleyan Student Assembly WSA Meeting

    E-Print Network [OSTI]

    Royer, Dana

    be like cutting studios - Motion to previous question passed Not Acceptable Label accepted by a 26-3-3 #12;Wesleyan Student Assembly - Phil: Joe Bruno said we will review it later Acceptable Label for Reorganizing Academic Departments 26-0-3 - David; did you consider selling off land? - Mike: Yes but not a good idea

  6. Assembly of quasicrystalline photonic heterostructures

    DOE Patents [OSTI]

    Grier, David G.; Roichman, Yael; Man, Weining; Chaikin, Paul Michael; Steinhardt, Paul Joseph

    2013-03-12

    A method and system for assembling a quasicrystalline heterostructure. A plurality of particles is provided with desirable predetermined character. The particles are suspended in a medium, and holographic optical traps are used to position the particles in a way to achieve an arrangement which provides a desired property.

  7. Assembly of quasicrystalline photonic heterostructures

    DOE Patents [OSTI]

    Grier, David G. (New York, NY); Roichman, Yael (New York, NY); Man, Weining (Princeton, NJ); Chaikin, Paul Michael (Pennington, NJ); Steinhardt, Paul Joseph (Princeton, NJ)

    2011-07-19

    A method and system for assembling a quasicrystalline heterostructure. A plurality of particles is provided with desirable predetermined character. The particles are suspended in a medium, and holographic optical traps are used to position the particles in a way to achieve an arrangement which provides a desired property.

  8. Solder self-assembly for MEMS fabrication

    E-Print Network [OSTI]

    Au, Hin Meng, 1977-

    2004-01-01

    This thesis examines and demonstrates self-assembly of MEMS components on the 25 micron scale onto substrates using the capillary force of solder. This is an order of magnitude smaller than current solder self-assembly in ...

  9. Multivariate Statistical Analysis of Assembly Tolerance Specifications

    E-Print Network [OSTI]

    i Multivariate Statistical Analysis of Assembly Tolerance Specifications A Thesis Presented ..........................................................................................................12 2.6.1 Multivariate Normal Distributions.8.3 Combined Assembly Specification Quality...............................................22 2.9 Multivariate

  10. Metal-templated assembly of protein cages

    E-Print Network [OSTI]

    Huard, Dustin Johnathen Edward

    2012-01-01

    Chapter 2. Generation of Metal-Responsive HuHF Buildingprotein interactions through metal coordination: Assembly ofSalgado, E.N. , et al. , Metal-mediated self-assembly of

  11. Transplanting assembly of individual carbon nanotubes

    E-Print Network [OSTI]

    Kim, Soohyung

    2009-01-01

    Handling and assembling individual nanostructures to bigger scale systems such as MEMS have been the biggest challenge. A deterministic assembly of individual carbon nanotubes by transplanting them to MEMS structures is ...

  12. Non-latching relay switch assembly

    DOE Patents [OSTI]

    Duimstra, Frederick A. (Anaheim Hills, CA)

    1991-01-01

    A non-latching relay switch assembly which includes a coil section and a switch or contact section. The coil section includes a permanent magnet and an electromagnet. The respective sections are arranged in separate locations or cavities in the assembly. The switch has a "normal" position and is selectively switched by an overriding electromagnetic assembly. The switch returns to the "normal" position when the overriding electromagnetic assembly is inactive.

  13. Fringing Field Directed Assembly of Nanomaterials

    E-Print Network [OSTI]

    Jacobs, Heiko O.

    using electrostatics14-18 and electrospinning,12 are exclusive to the assembly from the liquid phase

  14. Irradiation-induced phenomena in carbon

    E-Print Network [OSTI]

    Krasheninnikov, Arkady V.

    Chapter 1 Irradiation-induced phenomena in carbon nanotubes To appear in "Chemistry of Carbon@acclab.helsinki.fi 1 #12;2CHAPTER 1. IRRADIATION-INDUCED PHENOMENA IN CARBON NANOTUBES #12;Contents 1 Irradiation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.2 Interaction of energetic particles with solid targets . . . . . . . . 6 1.2.1 Ion irradiation

  15. EFFECTS OF GAMMA IRRADIATION ON EPDM ELASTOMERS

    SciTech Connect (OSTI)

    Clark, E.

    2011-09-22

    Two formulations of EPDM elastomer, one substituting a UV stabilizer for the normal antioxidant in this polymer, and the other the normal formulation, were synthesized and samples of each were exposed to gamma irradiation in initially pure deuterium gas to compare their radiation stability. Stainless steel containers having rupture disks were designed for this task. After 130 MRad dose of cobalt-60 radiation in the SRNL Gamma Irradiation Facility, a significant amount of gas was created by radiolysis; however the composition indicated by mass spectroscopy indicated an unexpected increase in the total amount deuterium in both formulations. The irradiated samples retained their ductility in a bend test. No change of sample weight, dimensions, or density was observed. No change of the glass transition temperature as measured by dynamic mechanical analysis was observed, and most of the other dynamic mechanical properties remained unchanged. There appeared to be an increase in the storage modulus of the irradiated samples containing the UV stabilizer above the glass transition, which may indicate hardening of the material by radiation damage. Polymeric materials become damaged by exposure over time to ionizing radiation. Despite the limited lifetime, polymers have unique engineering material properties and polymers continue to be used in tritium handling systems. In tritium handling systems, polymers are employed mainly in joining applications such as valve sealing surfaces (eg. Stem tips, valve packing, and O-rings). Because of the continued need to employ polymers in tritium systems, over the past several years, programs at the Savannah River National Laboratory have been studying the effect of tritium on various polymers of interest. In these studies, samples of materials of interest to the SRS Tritium Facilities (ultra-high molecular weight polyethylene (UHMW-PE), polytetrafluoroethylene (PTFE, Teflon{reg_sign}), Vespel{reg_sign} polyimide, and the elastomer ethylene propylene diene monomer (EPDM)) have been exposed in closed containers to tritium gas initially at 1 atmosphere pressure. These studies have demonstrated the degradation of properties when exposed to tritium gas. Also, the radiolytic production of significant amounts of hydrogen has been observed for UHMW-PE and EPDM. The study documented in this report exposes two similar formulations of EPDM elastomer to gamma irradiation in a closed container backfilled with deuterium. Deuterium is chemically identical to protium and tritium, but allows the identification of protium that is radiolytically produced from the samples. The goal of this program is to compare and contrast the response of EPDM exposure to two different types of ionizing radiation in a similar chemical environment.

  16. Solar irradiance forecasting at multiple time horizons and novel methods to evaluate uncertainty

    E-Print Network [OSTI]

    Marquez, Ricardo

    2012-01-01

    Solar irradiance data . . . . . . . . . . . . .Irradiance . . . . . . . . . . . . . . . . . . . . . . . . . . .4 Forecasting Solar Irradiance With GOES-West Satellite

  17. LANL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    SciTech Connect (OSTI)

    Fisher, S.E.; Holdaway, R.; Ludwig, S.B. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. LANL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO{sub 2} powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within both Category 1 and 2 areas. Technical Area (TA) 55/Plutonium Facility 4 will be used to store the bulk PuO{sub 2} powder, fabricate MOX fuel pellets, assemble rods, and store fuel bundles. Bundles will be assembled at a separate facility, several of which have been identified as suitable for that activity. The Chemistry and Metallurgy Research Building (at TA-3) will be used for analytical chemistry support. Waste operations will be conducted in TA-50 and TA-54. Only very minor modifications will be needed to accommodate the LA program. These modifications consist mostly of minor equipment upgrades. A commercial reactor operator has not been identified for the LA irradiation. Postirradiation examination (PIE) of the irradiated fuel will take place at either Oak Ridge National Laboratory or ANL-W. The only modifications required at either PIE site would be to accommodate full-length irradiated fuel rods. Results from this program are critical to the overall plutonium distribution schedule.

  18. Genome Sequence Databases (Overview): Sequencing and Assembly

    SciTech Connect (OSTI)

    Lapidus, Alla L.

    2009-01-01

    From the date its role in heredity was discovered, DNA has been generating interest among scientists from different fields of knowledge: physicists have studied the three dimensional structure of the DNA molecule, biologists tried to decode the secrets of life hidden within these long molecules, and technologists invent and improve methods of DNA analysis. The analysis of the nucleotide sequence of DNA occupies a special place among the methods developed. Thanks to the variety of sequencing technologies available, the process of decoding the sequence of genomic DNA (or whole genome sequencing) has become robust and inexpensive. Meanwhile the assembly of whole genome sequences remains a challenging task. In addition to the need to assemble millions of DNA fragments of different length (from 35 bp (Solexa) to 800 bp (Sanger)), great interest in analysis of microbial communities (metagenomes) of different complexities raises new problems and pushes some new requirements for sequence assembly tools to the forefront. The genome assembly process can be divided into two steps: draft assembly and assembly improvement (finishing). Despite the fact that automatically performed assembly (or draft assembly) is capable of covering up to 98% of the genome, in most cases, it still contains incorrectly assembled reads. The error rate of the consensus sequence produced at this stage is about 1/2000 bp. A finished genome represents the genome assembly of much higher accuracy (with no gaps or incorrectly assembled areas) and quality ({approx}1 error/10,000 bp), validated through a number of computer and laboratory experiments.

  19. Nuclear reactor composite fuel assembly

    DOE Patents [OSTI]

    Burgess, Donn M. (Richland, WA); Marr, Duane R. (West Richland, WA); Cappiello, Michael W. (Richland, WA); Omberg, Ronald P. (Richland, WA)

    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.

  20. Regenerator cross arm seal assembly

    DOE Patents [OSTI]

    Jackman, Anthony V. (Indianapolis, IN)

    1988-01-01

    A seal assembly for disposition between a cross arm on a gas turbine engine block and a regenerator disc, the seal assembly including a platform coextensive with the cross arm, a seal and wear layer sealingly and slidingly engaging the regenerator disc, a porous and compliant support layer between the platform and the seal and wear layer porous enough to permit flow of cooling air therethrough and compliant to accommodate relative thermal growth and distortion, a dike between the seal and wear layer and the platform for preventing cross flow through the support layer between engine exhaust and pressurized air passages, and air diversion passages for directing unregenerated pressurized air through the support layer to cool the seal and wear layer and then back into the flow of regenerated pressurized air.

  1. Abrasive swivel assembly and method

    DOE Patents [OSTI]

    Hashish, Mohamed (Kent, WA); Marvin, Mark (Tacoma, WA)

    1989-01-01

    An abrasive swivel assembly for providing a rotating, particle-laden fluid stream and, ultimately, a rotating particle-laden fluid jet is disclosed herein. This assembly includes a tubular arrangement for providing a particle-free stream of fluid, means for rotating a section of the tubular arrangement, and means for introducing solid particles into the particle-free fluid stream at a point along the rotating tubular section, whereby to produce a particle-laden fluid stream. This last-mentioned stream can then be used in combination with a cooperating nozzle arrangement for providing a rotating particle-laden fluid jet. In an actual working embodiment, the fluid stream is of sufficiently high pressure so that the abrasive jet can be used as a cutting jet.

  2. Dynamic pathways for viral capsid assembly

    SciTech Connect (OSTI)

    Hagan, Michael F.; Chandler, David

    2006-02-09

    We develop a class of models with which we simulate the assembly of particles into T1 capsid-like objects using Newtonian dynamics. By simulating assembly for many different values of system parameters, we vary the forces that drive assembly. For some ranges of parameters, assembly is facile, while for others, assembly is dynamically frustrated by kinetic traps corresponding to malformed or incompletely formed capsids. Our simulations sample many independent trajectories at various capsomer concentrations, allowing for statistically meaningful conclusions. Depending on subunit (i.e., capsomer) geometries, successful assembly proceeds by several mechanisms involving binding of intermediates of various sizes. We discuss the relationship between these mechanisms and experimental evaluations of capsid assembly processes.

  3. Pressure equalizing photovoltaic assembly and method

    DOE Patents [OSTI]

    Dinwoodie, Thomas L. (Piedmont, CA)

    2003-05-27

    Each PV assembly of an array of PV assemblies comprises a base, a PV module and a support assembly securing the PV module to a position overlying the upper surface of the base. Vents are formed through the base. A pressure equalization path extends from the outer surface of the PV module, past the peripheral edge of the PV module, to and through at least one of the vents, and to the lower surface of the base to help reduce wind uplift forces on the PV assembly. The PV assemblies may be interengaged, such as by interengaging the bases of adjacent PV assemblies. The base may include a main portion and a cover and the bases of adjacent PV assemblies may be interengaged by securing the covers of adjacent bases together.

  4. Pressure-equalizing PV assembly and method

    DOE Patents [OSTI]

    Dinwoodie, Thomas L.

    2004-10-26

    Each PV assembly of an array of PV assemblies comprises a base, a PV module and a support assembly securing the PV module to a position overlying the upper surface of the base. Vents are formed through the base. A pressure equalization path extends from the outer surface of the PV module, past the PV module, to and through at least one of the vents, and to the lower surface of the base to help reduce wind uplift forces on the PV assembly. The PV assemblies may be interengaged, such as by interengaging the bases of adjacent PV assemblies. The base may include a main portion and a cover and the bases of adjacent PV assemblies may be interengaged by securing the covers of adjacent bases together.

  5. Metal-ceramic joint assembly

    DOE Patents [OSTI]

    Li, Jian (New Milford, CT)

    2002-01-01

    A metal-ceramic joint assembly in which a brazing alloy is situated between metallic and ceramic members. The metallic member is either an aluminum-containing stainless steel, a high chromium-content ferritic stainless steel or an iron nickel alloy with a corrosion protection coating. The brazing alloy, in turn, is either an Au-based or Ni-based alloy with a brazing temperature in the range of 9500 to 1200.degree. C.

  6. EFFECTS OF GAMMA IRRADIATION ON EPDM ELASTOMERS (REVISION 1)

    SciTech Connect (OSTI)

    Clark, E.

    2013-09-13

    Two formulations of EPDM elastomer, one substituting a UV stabilizer for the normal antioxidant in this polymer, and the other the normal formulation, were synthesized and samples of each were exposed to gamma irradiation in initially pure deuterium gas to compare their radiation stability. Stainless steel containers having rupture disks were designed for this task. After 130 MRad dose of cobalt-60 radiation in the SRNL Gamma Irradiation Facility, a significant amount of gas was created by radiolysis; however the composition indicated by mass spectroscopy indicated an unexpected increase in the total amount deuterium in both formulations. The irradiated samples retained their ductility in a bend test. No change of sample weight, dimensions, or density was observed. No change of the glass transition temperature as measured by dynamic mechanical analysis was observed, and most of the other dynamic mechanical properties remained unchanged. There appeared to be an increase in the storage modulus of the irradiated samples containing the UV stabilizer above the glass transition, which may indicate hardening of the material by radiation damage. Revision 1 adds a comparison with results of a study of tritium exposed EPDM. The amount of gas produced by the gamma irradiation was found to be equivalent to about 280 days exposure to initially pure tritium gas at one atmosphere. The glass transition temperature of the tritium exposed EPDM rose about 10 ?C. over 280 days, while no glass transition temperature change was observed for gamma irradiated EPDM. This means that gamma irradiation in deuterium cannot be used as a surrogate for tritium exposure.

  7. Rotor assembly and assay method

    DOE Patents [OSTI]

    Burtis, Carl A. (Oak Ridge, TN); Johnson, Wayne F. (Loudon, TN); Walker, William A. (Knoxville, TN)

    1993-01-01

    A rotor assembly for carrying out an assay includes a rotor body which is rotatable about an axis of rotation, and has a central chamber and first, second, third, fourth, fifth, and sixth chambers which are in communication with and radiate from the central chamber. The rotor assembly further includes a shuttle which is movable through the central chamber and insertable into any of the chambers, the shuttle including a reaction cup carrying an immobilized antigen or an antibody for transport among the chambers. A method for carrying out an assay using the rotor assembly includes moving the reaction cup among the six chambers by passing the cup through the central chamber between centrifugation steps in order to perform the steps of: separating plasma from blood cells, binding plasma antibody or antigen, washing, drying, binding enzyme conjugate, reacting with enzyme substrate and optically comparing the resulting reaction product with unreacted enzyme substrate solution. The movement of the reaction cup can be provided by attaching a magnet to the reaction cup and supplying a moving magnetic field to the rotor.

  8. Rotor assembly and assay method

    DOE Patents [OSTI]

    Burtis, C.A.; Johnson, W.F.; Walker, W.A.

    1993-09-07

    A rotor assembly for carrying out an assay includes a rotor body which is rotatable about an axis of rotation, and has a central chamber and first, second, third, fourth, fifth, and sixth chambers which are in communication with and radiate from the central chamber. The rotor assembly further includes a shuttle which is movable through the central chamber and insertable into any of the chambers, the shuttle including a reaction cup carrying an immobilized antigen or an antibody for transport among the chambers. A method for carrying out an assay using the rotor assembly includes moving the reaction cup among the six chambers by passing the cup through the central chamber between centrifugation steps in order to perform the steps of: separating plasma from blood cells, binding plasma antibody or antigen, washing, drying, binding enzyme conjugate, reacting with enzyme substrate and optically comparing the resulting reaction product with unreacted enzyme substrate solution. The movement of the reaction cup can be provided by attaching a magnet to the reaction cup and supplying a moving magnetic field to the rotor. 34 figures.

  9. Cracking behavior and microstructure of austenitic stainless steels and alloy 690 irradiated in BOR-60 reactor, phase I.

    SciTech Connect (OSTI)

    Chen, Y.; Chopra, O. K.; Soppet, W. K.; Shack, W. J.; Yang, Y.; Allen, T. R.; Univ. of Wisconsin at Madison

    2010-02-16

    Cracking behavior of stainless steels specimens irradiated in the BOR-60 at about 320 C is studied. The primary objective of this research is to improve the mechanistic understanding of irradiation-assisted stress corrosion cracking (IASCC) of core internal components under conditions relevant to pressurized water reactors. The current report covers several baseline tests in air, a comparison study in high-dissolved-oxygen environment, and TEM characterization of irradiation defect structure. Slow strain rate tensile (SSRT) tests were conducted in air and in high-dissolved-oxygen (DO) water with selected 5- and 10-dpa specimens. The results in high-DO water were compared with those from earlier tests with identical materials irradiated in the Halden reactor to a similar dose. The SSRT tests produced similar results among different materials irradiated in the Halden and BOR-60 reactors. However, the post-irradiation strength for the BOR-60 specimens was consistently lower than that of the corresponding Halden specimens. The elongation of the BOR-60 specimens was also greater than that of their Halden specimens. Intergranular cracking in high-DO water was consistent for most of the tested materials in the Halden and BOR-60 irradiations. Nonetheless, the BOR-60 irradiation was somewhat less effective in stimulating IG fracture among the tested materials. Microstructural characterization was also carried out using transmission electron microscopy on selected BOR-60 specimens irradiated to {approx}25 dpa. No voids were observed in irradiated austenitic stainless steels and cast stainless steels, while a few voids were found in base and grain-boundary-engineered Alloy 690. All the irradiated microstructures were dominated by a high density of Frank loops, which varied in mean size and density for different alloys.

  10. Electrical Quality Assurance of the Superconducting Circuits during LHC Machine Assembly

    E-Print Network [OSTI]

    Bozzini, D; Desebe, O; Mess, K H; Russenschuck, Stephan; Bednarek, M; Dworak, D; Górnicki, E; Jurkiewicz, P; Kapusta, P; Kotarba, A; Ludwin, J; Olek, S; Talach, M; Zieblinski, M; Klisch, M; Prochal, B

    2008-01-01

    Based on the LHC powering reference database, all-together 1750 superconducting circuits were connected in the various cryogenic transfer lines of the LHC machine. Testing the continuity, magnet polarity, and the quality of the electrical insulation were the main tasks of the Electrical Quality Assurance (ELQA) activities during the LHC machine assembly. With the assembly of the LHC now complete, the paper reviews the work flow, resources, and the qualification results including the different types of electrical non-conformities.

  11. Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments

    SciTech Connect (OSTI)

    Chen, Y.; Chopra, O. K.; Gruber, Eugene E.; Shack, William J.

    2010-06-01

    The internal components of light water reactors are exposed to high-energy neutron irradiation and high-temperature reactor coolant. The exposure to neutron irradiation increases the susceptibility of austenitic stainless steels (SSs) to stress corrosion cracking (SCC) because of the elevated corrosion potential of the reactor coolant and the introduction of new embrittlement mechanisms through radiation damage. Various nonsensitized SSs and nickel alloys have been found to be prone to intergranular cracking after extended neutron exposure. Such cracks have been seen in a number of internal components in boiling water reactors (BWRs). The elevated susceptibility to SCC in irradiated materials, commonly referred to as irradiation-assisted stress corrosion cracking (IASCC), is a complex phenomenon that involves simultaneous actions of irradiation, stress, and corrosion. In recent years, as nuclear power plants have aged and irradiation dose increased, IASCC has become an increasingly important issue. Post-irradiation crack growth rate and fracture toughness tests have been performed to provide data and technical support for the NRC to address various issues related to aging degradation of reactor-core internal structures and components. This report summarizes the results of the last group of tests on compact tension specimens from the Halden-II irradiation. The IASCC susceptibility of austenitic SSs and heat-affected-zone (HAZ) materials sectioned from submerged arc and shielded metal arc welds was evaluated by conducting crack growth rate and fracture toughness tests in a simulated BWR environment. The fracture and cracking behavior of HAZ materials, thermally sensitized SSs and grain-boundary engineered SSs was investigated at several doses (?3 dpa). These latest results were combined with previous results from Halden-I and II irradiations to analyze the effects of neutron dose, water chemistry, alloy compositions, and welding and processing conditions on IASCC. The effect of neutron irradiation on the fracture toughness of austenitic SSs was also evaluated at dose levels relevant to BWR internals.

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

    DOE Patents [OSTI]

    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.

  13. Irradiation dose and temperature dependence of fracture toughness in high dose HT9 steel from the fuel duct of FFTF

    SciTech Connect (OSTI)

    Byun, Thak Sang; Toloczko, Mychailo B.; Saleh, Tarik A.; Maloy, Stuart A.

    2013-01-14

    To expand the knowledge base for fast reactor core materials, fracture toughness has been evaluated for high dose HT9 steel using miniature disk compact tension (DCT) specimens. The HT9 steel DCT specimens were machined from the ACO-3 fuel duct of the Fast Flux Test Facility (FFTF), which achieved high doses in the range of 3–148 dpa at 378–504 C. The static fracture resistance (J-R) tests have been performed in a servohydraulic testing machine in vacuum at selected temperatures including room temperature, 200 C, and each irradiation temperature. Brittle fracture with a low toughness less than 50 MPa pm occurred in room temperature tests when irradiation temperature was below 400 C, while ductile fracture with stable crack growth was observed when irradiation temperature was higher. No fracture toughness less than 100 MPa pm was measured when the irradiation temperature was above 430 C. It was shown that the influence of irradiation temperature was dominant in fracture toughness while the irradiation dose has only limited influence over the wide dose range 3–148 dpa. A slow decrease of fracture toughness with test temperature above room temperature was observed for the nonirradiated and high temperature (>430 *C) irradiation cases, which indicates that the ductile–brittle transition temperatures (DBTTs) in those conditions are lower than room temperature. A comparison with the collection of existing data confirmed the dominance of irradiation temperature in the fracture toughness of HT9 steels.

  14. Occlusion-Aware Hessians for Error Control in Irradiance Caching /

    E-Print Network [OSTI]

    Schwarzhaupt, Jorge Andres

    2013-01-01

    5.2 Derivation of the Irradiance Hessian . . 5.3of the translational irradiance gradient. . . . . .1.1.1 Occlusion-Aware Irradiance Hessian . . . . . . . .

  15. High-frequency irradiance fluctuations and geographic smoothing

    E-Print Network [OSTI]

    Lave, Matthew; Kleissl, Jan; Arias-Castro, Ery

    2012-01-01

    of the variability of irradiance and PV power time series1997. Study of areal solar irradiance for analyzing areally-High-frequency irradiance fluctuations and geographic

  16. Tritium Related Material Research -Irradiation Effect on Isotropic...

    Office of Environmental Management (EM)

    Related Material Research -Irradiation Effect on Isotropic Graphite Utilizing Heavy Ion-Irradiation- Tritium Related Material Research -Irradiation Effect on Isotropic Graphite...

  17. Nonlinear broadband photoluminescence of graphene induced by femtosecond laser irradiation

    E-Print Network [OSTI]

    Liu, Wei-Tao

    2011-01-01

    by femtosecond laser irradiation Wei-Tao Liu, 1 S. W. Wu, 2USA Upon femtosecond laser irradiation, a bright, broadbandthe experimental etup. laser irradiation, strong PL signals

  18. Fuel Assembly Shaker Test for Determining Loads on a PWR Assembly...

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

    current approach of long-term storage at its nuclear power plants and independent spent fuel storage installation, and deferred transportation of used nuclear fuel (UNF), along...

  19. Fuel Assembly Shaker Test for Determining Loads on a PWR Assembly under

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:FinancingPetroleum12, 2015ExecutiveFluorescentDan O"HaganTalley,Surrogate Normal Conditions of

  20. Small Specimen Mechanical Property Testing at PNNL

    E-Print Network [OSTI]

    McDonald, Kirk

    to measure crack length in-situ. Fully automated precracking system. 7 Brittle fracture of ferritic/martensitic steel fracture toughness specimen #12;Fracture Toughness Testing of Unirradiated and Irradiated

  1. FETS*-HIPSTER** * Front End Test Stand

    E-Print Network [OSTI]

    McDonald, Kirk

    A · Candidate materials for irradiation testing: Be, C, Ti, Steels, W #12;FETS Future and long term plans of completing the current phase. FETS continues discussions with the neutron, medical and fusion materials variable, dual beam in late 2015

  2. The Assembly of the Belle II TOP Counter

    E-Print Network [OSTI]

    Boqun Wang; for the Belle II PID Group

    2015-01-13

    A new type of ring-imaging Cherenkov counter, called TOP counter, has been developed for particle identification at the Belle II experiment to run at the SuperKEKB accelerator in KEK, Japan. The detector consists of 16 identical modules arranged azimuthally around the beam line. The assembly procedure for a TOP module is described. This procedure includes acceptance testing of the quartz mirror, prism, and quartz bar radiators. The acceptance tests include a chip search and measurements of bulk transmittance and total internal reflectance. The process for aligning and gluing the optical components together is described.

  3. Post Irradiation Capabilities at the Idaho National Laboratory

    SciTech Connect (OSTI)

    Schulthess, J.L.; Robert D. Mariani; Rory Kennedy; Doug Toomer

    2011-08-01

    The U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) oversees the research, development, and demonstration activities that ensure nuclear energy remains a viable energy option for the United States. Fuel and material development through fabrication, irradiation, and characterization play a significant role in accomplishing the research needed to support nuclear energy. All fuel and material development requires the understanding of irradiation effects on the fuel performance and relies on irradiation experiments ranging from tests aimed at targeted scientific questions to integral effects under representative and prototypic conditions. The DOE recently emphasized a solution-driven, goal-oriented, science-based approach to nuclear energy development. Nuclear power systems and materials were initially developed during the latter half of the 20th century and greatly facilitated by the United States’ ability and willingness to conduct large-scale experiments. Fifty-two research and test reactors with associated facilities for performing fabrication and pre and post irradiation examinations were constructed at what is now Idaho National Laboratory (INL), another 14 at Oak Ridge National Laboratory (ORNL), and a few more at other national laboratory sites. Building on the scientific advances of the last several decades, our understanding of fundamental nuclear science, improvements in computational platforms, and other tools now enable technological advancements with less reliance on large-scale experimentation.

  4. Post Irradiation Capabilities at the Idaho National Laboratory

    SciTech Connect (OSTI)

    Schulthess, J.L.

    2011-08-01

    The U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) oversees the research, development, and demonstration activities that ensure nuclear energy remains a viable energy option for the United States. Fuel and material development through fabrication, irradiation, and characterization play a significant role in accomplishing the research needed to support nuclear energy. All fuel and material development requires the understanding of irradiation effects on the fuel performance and relies on irradiation experiments ranging from tests aimed at targeted scientific questions to integral effects under representative and prototypic conditions. The DOE recently emphasized a solution-driven, goal-oriented, science-based approach to nuclear energy development. Nuclear power systems and materials were initially developed during the latter half of the 20th century and greatly facilitated by the United States ability and willingness to conduct large-scale experiments. Fifty-two research and test reactors with associated facilities for performing fabrication and pre and post irradiation examinations were constructed at what is now Idaho National Laboratory (INL), another 14 at Oak Ridge National Laboratory (ORNL), and a few more at other national laboratory sites. Building on the scientific advances of the last several decades, our understanding of fundamental nuclear science, improvements in computational platforms, and other tools now enable technological advancements with less reliance on large-scale experimentation.

  5. Irradiation-assisted stress corrosion cracking of HTH Alloy X-750 and Alloy 625

    SciTech Connect (OSTI)

    Bajaj, R.; Mills, W.J.; Lebo, M.R.; Hyatt, B.Z.; Burke, M.G.

    1995-07-01

    In-reactor testing of bolt-loaded compact tension specimens was performed in 360 C water. New data confirms previous results that high irradiation levels reduce SCC resistance in Alloy X-750. Low boron heats show improved IASCC (irradiation-assisted stress corrosion cracking). Alloy 625 is resistant to IASCC. Microstructural, microchemical, and deformation studies were carried out. Irradiation of X-750 caused significant strengthening and ductility loss associated with formation of cavities and dislocation loops. High irradiation did not cause segregation in X-750. Irradiation of 625 resulted in formation of small dislocation loops and a fine body-centered-orthorhombic phase. The strengthening due to loops and precipitates was apparently offset in 625 by partial dissolution of {gamma} precipitates. Transmutation of boron to helium at grain boundaries, coupled with matrix strengthening, is believed to be responsible for IASCC in X-750, and the absence of these two effects results in superior IASCC resistance in 625.

  6. Design, Fabrication, Assembly and Initial Testing of a SMART Rotor

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration would like submit theCovalent BondingMeeting |Design Competitions Designthe 29 th ASME Wind

  7. Method of monolithic module assembly

    DOE Patents [OSTI]

    Gee, James M. (Albuquerque, NM); Garrett, Stephen E. (Albuquerque, NM); Morgan, William P. (Albuquerque, NM); Worobey, Walter (Albuquerque, NM)

    1999-01-01

    Methods for "monolithic module assembly" which translate many of the advantages of monolithic module construction of thin-film PV modules to wafered c-Si PV modules. Methods employ using back-contact solar cells positioned atop electrically conductive circuit elements affixed to a planar support so that a circuit capable of generating electric power is created. The modules are encapsulated using encapsulant materials such as EVA which are commonly used in photovoltaic module manufacture. The methods of the invention allow multiple cells to be electrically connected in a single encapsulation step rather than by sequential soldering which characterizes the currently used commercial practices.

  8. Nanoengineered membrane electrode assembly interface

    DOE Patents [OSTI]

    Song, Yujiang; Shelnutt, John A

    2013-08-06

    A membrane electrode structure suitable for use in a membrane electrode assembly (MEA) that comprises membrane-affixed metal nanoparticles whose formation is controlled by a photochemical process that controls deposition of the metal nanoparticles using a photocatalyst integrated with a polymer electrolyte membrane, such as an ionomer membrane. Impregnation of the polymer membrane with the photocatalyst prior to metal deposition greatly reduces the required amount of metal precursor in the deposition reaction solution by restricting metal reduction substantially to the formation of metal nanoparticles affixed on or near the surface of the polymer membrane with minimal formation of metallic particles not directly associated with the membrane.

  9. Algorithm for a microfluidic assembly line

    E-Print Network [OSTI]

    Tobias M. Schneider; Shreyas Mandre; Michael P. Brenner

    2011-01-19

    Microfluidic technology has revolutionized the control of flows at small scales giving rise to new possibilities for assembling complex structures on the microscale. We analyze different possible algorithms for assembling arbitrary structures, and demonstrate that a sequential assembly algorithm can manufacture arbitrary 3D structures from identical constituents. We illustrate the algorithm by showing that a modified Hele-Shaw cell with 7 controlled flowrates can be designed to construct the entire English alphabet from particles that irreversibly stick to each other.

  10. Natural convection heat transfer within horizontal spent nuclear fuel assemblies

    SciTech Connect (OSTI)

    Canaan, R.E.

    1995-12-01

    Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array.

  11. Heavy duty insulator assemblies for 500-kV bulk power transmission line with large diameter octagonalbundled conductor

    SciTech Connect (OSTI)

    Tsujimoto, K.; Hayase, I.; Hirai, J.; Inove, M.; Naito, K.; Yukino, T.

    1982-11-01

    This paper describes the design procedure and the results of field tests on mechanical performances of insulator assemblies newly developed to support octagonal-bundled conductors for 500-kV bulk power transmission. Taking account of conductor-motion-induced peak tensile load, fatigue, torsional torque and others, a successful design has been achieved in two prototype assemblies for such heavy mechanical duties as encountered during conductor galloping or swing. This has been proved throughout three years of the field tests.

  12. SUB-LEU-METAL-THERM-001 SUBCRITICAL MEASUREMENTS OF LOW ENRICHED TUBULAR URANIUM METAL FUEL ELEMENTS BEFORE & AFTER IRRADIATION

    SciTech Connect (OSTI)

    TOFFER, H.

    2006-07-18

    With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Fuel that had experienced a neutron environment in a reactor is known as spent, exposed, or irradiated fuel. In contrast fuel that has not yet been placed in a reactor is known as green, unexposed, or unirradiated fuel. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled (References 1 and 2) and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements (Reference 3). The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprised of two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with unirradiated fuel and one with irradiated fuel. Both the unirradiated and irradiated fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k{sub eff} = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, three (3) green fuel and four (4) spent fuel loading configurations were considered to serve as benchmark models. However, shortcomings in experimental data, such as the uncertainty in fuel exposure impact on reactivity and the pulse neutron data evaluation methodology, failed to meet the high standards for a benchmark problem. Nevertheless, the data provided by these subcritical measurements supply useful information to analysts evaluating spent fuel subcriticality. The original purpose of the subcritical measurements was to validate computer model predictions that spent N Reactor fuel of a particular, typical exposure (2740 MWd/t) had a critical mass equal to twice that of unexposed fuel of the same type. The motivation for performing this work was driven by the need to increase spent fuel storage limits. These subcritical measurements confirmed the computer model predictions.

  13. Assembly and post-assembly manipulation of polyelectrolyte multilayers for control of bacterial attachment and viability

    E-Print Network [OSTI]

    Lichter, Jenny, 1982-

    2009-01-01

    The overall goal of this thesis was to exploit the versatility of the polyelectrolyte multilayer (PEM) platform to consider bacteria-substrata interactions by varying multilayer assembly and post-assembly conditions. We ...

  14. Metal-templated assembly of protein cages

    E-Print Network [OSTI]

    Huard, Dustin Johnathen Edward

    2012-01-01

    self- assembly. Far-UV CD measurements show that the MIC1spectrometry to UV-VIS absorbance measurements to ensure

  15. Method of cleaning a spent fuel assembly

    SciTech Connect (OSTI)

    Chung, D.K.; Jones, C.E. Jr.

    1989-05-09

    A method is described of cleaning a fuel assembly including surfaces thereof prior to decladding, each assembly surface contaminated with a radioactive alkali metal and comprising a plurality of pressurized metallic fuel pins containing a spent fissible material, the method comprising the sequential steps of: (a) placing the fuel assembly in a sealed chamber; (b) passing a heated, inert gas through the chamber to heat the fuel assembly to a temperature sufficient to cause volatilization of the alkali metal but insufficient to rupture the pressurized metal pins; (c) evacuating the chamber to a pressure of less than 0.5 mm of Hg to further enhance volatilization and removal of the alkali metal and maintaining the chamber at that pressure until the decay heat of the fissile materials causes the temperature of the fuel assembly to increase to a level which would be detrimental to the integrity of the metal pins; (d) cooling the fuel assembly by passing a cool, inert gas through the chamber to reduce the temperature of the fuel assembly to a desired level; (e) repeating the evacuation and cooling steps as required to insure removal of substantially all of the radioactive alkali metal from the assembly surface; and (f) recovering the cleaned fuel assembly from the chamber.

  16. Optically Directed Assembly of Continuous Mesoscale Filaments...

    Office of Scientific and Technical Information (OSTI)

    Optically Directed Assembly of Continuous Mesoscale Filaments Bahns, J. T.; Sankaranarayanan, S. K. R. S.; Gray, S. K.; Chen, L. Not Available American Physical Society None USDOE...

  17. Statistical criteria for characterizing irradiance time series.

    SciTech Connect (OSTI)

    Stein, Joshua S.; Ellis, Abraham; Hansen, Clifford W.

    2010-10-01

    We propose and examine several statistical criteria for characterizing time series of solar irradiance. Time series of irradiance are used in analyses that seek to quantify the performance of photovoltaic (PV) power systems over time. Time series of irradiance are either measured or are simulated using models. Simulations of irradiance are often calibrated to or generated from statistics for observed irradiance and simulations are validated by comparing the simulation output to the observed irradiance. Criteria used in this comparison should derive from the context of the analyses in which the simulated irradiance is to be used. We examine three statistics that characterize time series and their use as criteria for comparing time series. We demonstrate these statistics using observed irradiance data recorded in August 2007 in Las Vegas, Nevada, and in June 2009 in Albuquerque, New Mexico.

  18. Effects of Irradiation on the Microstructure of U-7Mo Dispersion Fuel with Al-2Si Matrix

    SciTech Connect (OSTI)

    Dennis D. Keiser, Jr.; Jan-Fong Jue; Adam B. Robinson; Pavel Medvedev; Jian Gan; Brandon D. Miller; Daniel M. Wachs; Glenn A. Moore; Curtis R. Clark; Mitchell K. Meyer; M. Ross Finlay

    2012-06-01

    The Reduced Enrichment for Research and Test Reactor program is developing low-enriched uranium U-Mo dispersion fuels for application in research and test reactors around the world. As part of this development, fuel plates have been irradiated in the Advanced Test Reactor and then characterized using optical metallography (OM) and scanning electron microscopy (SEM) to determine the as-irradiated microstructure. To demonstrate the irradiation performance of U-7Mo dispersion fuel plates with 2 wt% Si added to the matrix, fuel plates were tested to medium burnups at intermediate fission rates as part of the RERTR-6 experiment. Further testing was performed to higher fission rates as part of the RERTR-7A experiment, and very aggressive testing (high temperature, high fission density, high fission rate) was performed in the RERTR-9A, RERTR-9B and AFIP-1 experiments. As-irradiated microstructures were compared to those observed after fabrication to determine the effects of irradiation on the microstructure. Based on comparison of the microstructural characterization results for each irradiated sample, some general conclusions can be drawn about how the microstructure evolves during irradiation: there is growth of the fuel/matrix interaction layer (FMI), which was present in the samples to some degree after fabrication, during irradiation; Si diffuses from the FMI layer to deeper depths in the U-7Mo particles as the irradiation conditions are made more aggressive; lowering of the Si content in the FMI layer results in an increase in the size of the fission gas bubbles; as the FMI layer grows during irradiation more Si diffuses from the matrix to the FMI layer/matrix interface, and interlinking of fission gas bubbles in the fuel plate microstructure that may indicate breakaway swelling is not observed.

  19. Burnup Predictions for Metal Fuel Tests in the Fast Flux Test Facility

    SciTech Connect (OSTI)

    Wootan, David W.; Nelson, Joseph V.

    2012-06-01

    The Fast Flux Test Facility (FFTF) is the most recent Liquid Metal Reactor (LMR) to be designed, constructed, and operated by the U.S. Department of Energy (DOE). The FFTF operated successfully from initial startup in 1980 through the end of the last operating cycle in March, 1992. A variety of fuel tests were irradiated in FFTF to provide performance data over a range of conditions. The MFF-3 and MFF-5 tests were U10Zr metal fuel tests with HT9 cladding. The MFF-3 and MFF-5 tests were both aggressive irradiation tests of U10Zr metal fuel pins with HT9 cladding that were prototypic of full scale LMR designs. MFF-3 was irradiated for 726 Effective Full Power Days (EFPD), starting from Cycle 10C1 (from November 1988 through March 1992), and MFF-5 was irradiated for 503 EFPD starting from Cycle 11B1 (from January 1990 through March 1992). A group of fuel pins from these two tests are undergoing post irradiation examination at the Idaho National Laboratory (INL) for the Fuel Cycle Research and Development Program (FCRD). The generation of a data package of key information on the irradiation environment and current pin detailed compositions for these tests is described. This information will be used in interpreting the results of these examinations.

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

    SciTech Connect (OSTI)

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

    1984-01-01

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

  1. Valve stem and packing assembly

    DOE Patents [OSTI]

    Wordin, J.J.

    1991-09-03

    A valve stem and packing assembly is provided in which a rotatable valve stem includes a first tractrix surface for sliding contact with a stem packing and also includes a second tractrix surface for sliding contact with a bonnet. Force is applied by means of a spring, gland flange, and gland on the stem packing so the stem packing seals to the valve stem and bonnet. This configuration serves to create and maintain a reliable seal between the stem packing and the valve stem. The bonnet includes a second complementary tractrix surface for contacting the second sliding tractrix surface, the combination serving as a journal bearing for the entire valve stem and packing assembly. The journal bearing so configured is known as a Schiele's pivot. The Schiele's pivot also serves to maintain proper alignment of the valve stem with respect to the bonnet. Vertical wear between the surfaces of the Schiele's pivot is uniform at all points of contact between the second sliding tractrix surface and the second complementary tractrix surface of a bonnet. The valve stem is connected to a valve plug by means of a slip joint. The valve is opened and closed by rotating the valve stem. The slip joint compensates for wear on the Schiele's pivot and on the valve plug. A ledge is provided on the valve bonnet for the retaining nut to bear against. The ledge prevents over tightening of the retaining nut and the resulting excessive friction between stem and stem packing. 2 figures.

  2. Valve stem and packing assembly

    DOE Patents [OSTI]

    Wordin, John J. (Bingham County, ID)

    1991-01-01

    A valve stem and packing assembly is provided in which a rotatable valve stem includes a first tractrix surface for sliding contact with a stem packing and also includes a second tractrix surface for sliding contact with a bonnet. Force is applied by means of a spring, gland flange, and gland on the stem packing so the stem packing seals to the valve stem and bonnet. This configuration serves to create and maintain a reliable seal between the stem packing and the valve stem. The bonnet includes a second complementary tractrix surface for contacting the second sliding tractrix surface, the combination serving as a journal bearing for the entire valve stem and packing assembly. The journal bearing so configured is known as a Schiele's pivot. The Schiele's pivot also serves to maintain proper alignment of the valve stem with respect to the bonnet. Vertical wear between the surfaces of the Schiele's pivot is uniform at all points of contact between the second sliding tractrix surface and the second complementary tractrix surface of a bonnet. The valve stem is connected to a valve plug by means of a slip joint. The valve is opened and closed by rotating the valve stem. The slip joint compensates for wear on the Schiele's pivot and on the valve plug. A ledge is provided on the valve bonnet for the retaining nut to bear against. The ledge prevents overtightening of the retaining nut and the resulting excessive friction between stem and stem packing.

  3. Corner-cutting mining assembly

    DOE Patents [OSTI]

    Bradley, J.A.

    1981-07-01

    This invention resulted from a contract with the United States Department of Energy and relates to a mining tool. More particularly, the invention relates to an assembly capable of drilling a hole having a square cross-sectional shape with radiused corners. In mining operations in which conventional auger-type drills are used to form a series of parallel, cylindrical holes in a coal seam, a large amount of coal remains in place in the seam because the shape of the holes leaves thick webs between the holes. A higher percentage of coal can be mined from a seam by a means capable of drilling holes having a substantially square cross section. It is an object of this invention to provide an improved mining apparatus by means of which the amount of coal recovered from a seam deposit can be increased. Another object of the invention is to provide a drilling assembly which cuts corners in a hole having a circular cross section. These objects and other advantages are attained by a preferred embodiment of the invention.

  4. Nanoparticle Assemblies at Fluid Interfaces

    SciTech Connect (OSTI)

    Russell, Thomas P.

    2015-03-10

    A systematic study of the structure and dynamics of nanoparticles (NP) and NP-surfactants was performed. The ligands attached to both the NPs and NP-surfactants dictate the manner in which the nanoscopic materials assemble at fluid interfaces. Studies have shown that a single layer of the nanoscpic materials form at the interface to reduce the interactions between the two immiscible fluids. The shape of the NP is, also, important, where for spherical particles, a disordered, liquid-like monolayer forms, and, for nanorods, ordered domains at the interface is found and, if the monolayers are compressed, the orientation of the nanorods with respect to the interface can change. By associating end-functionalized polymers to the NPs assembled at the interface, NP-surfactants are formed that increase the energetic gain in segregating each NP at the interface which allows the NP-surfactants to jam at the interface when compressed. This has opened the possibility of structuring the two liquids by freezing in shape changes of the liquids.

  5. Breakdown properties of irradiated MOS capacitors

    SciTech Connect (OSTI)

    Paccagnella, A.; Candelori, A.; Milani, A.; Formigoni, E.; Ghidini, G.; Drera, D.; Pellizzer, F.; Fuochi, P.G.; Lavale, M.

    1996-12-01

    The authors have studied the effects of ionizing and non-ionizing radiation on the breakdown properties of different types of MOS capacitors, with thick (200 nm) and thin (down to 8 nm) oxides. In general, no large variations of the average breakdown field, time-to-breakdown at constant voltage, or charge-to-breakdown at constant voltage, or charge-to-breakdown values have been observed after high dose irradiation (20 Mrad(Si) 9 MeV electrons on thin and thick oxides, 17(Si) Mrad Co{sup 60} gamma and 10{sup 14} neutrons/cm{sup 2} only on thick oxides). However, some modifications of the cumulative failure distributions have been observed in few of the oxides tested.

  6. EVALUATION OF U10MO FUEL PLATE IRRADIATION BEHAVIOR VIA NUMERICAL AND EXPERIMENTAL BENCHMARKING

    SciTech Connect (OSTI)

    Samuel J. Miller; Hakan Ozaltun

    2012-11-01

    This article analyzes dimensional changes due to irradiation of monolithic plate-type nuclear fuel and compares results with finite element analysis of the plates during fabrication and irradiation. Monolithic fuel plates tested in the Advanced Test Reactor (ATR) at Idaho National Lab (INL) are being used to benchmark proposed fuel performance for several high power research reactors. Post-irradiation metallographic images of plates sectioned at the midpoint were analyzed to determine dimensional changes of the fuel and the cladding response. A constitutive model of the fabrication process and irradiation behavior of the tested plates was developed using the general purpose commercial finite element analysis package, Abaqus. Using calculated burn-up profiles of irradiated plates to model the power distribution and including irradiation behaviors such as swelling and irradiation enhanced creep, model simulations allow analysis of plate parameters that are either impossible or infeasible in an experimental setting. The development and progression of fabrication induced stress concentrations at the plate edges was of primary interest, as these locations have a unique stress profile during irradiation. Additionally, comparison between 2D and 3D models was performed to optimize analysis methodology. In particular, the ability of 2D and 3D models account for out of plane stresses which result in 3-dimensional creep behavior that is a product of these components. Results show that assumptions made in 2D models for the out-of-plane stresses and strains cannot capture the 3-dimensional physics accurately and thus 2D approximations are not computationally accurate. Stress-strain fields are dependent on plate geometry and irradiation conditions, thus, if stress based criteria is used to predict plate behavior (as opposed to material impurities, fine micro-structural defects, or sharp power gradients), unique 3D finite element formulation for each plate is required.

  7. Part X. The Faculty Assembly of the University of North Carolina I. Faculty Assembly Organization

    E-Print Network [OSTI]

    Gopalakrishnan, K.

    Part X. The Faculty Assembly of the University of North Carolina I. Faculty Assembly Organization The University of North Carolina Faculty Assembly was formed in 1972 when all of the State's public senior institutions were brought into the University of North Carolina. The Charter of the University of North

  8. Optimizing Robot Motion Strategies for Assembly with Stochastic Models of`the Assembly Process

    E-Print Network [OSTI]

    Hutchinson, Seth

    at this level) address the flow of parts through the assembly system. Production control deter- mines what at Urbana-Champaign 405 N. Mathews Avenue, Urbana, IL 61801 Abstract Gross-motion planning for assembly-motion planning. In this pa- per we formulate the problem of gross-motion planning for assembly in a manner

  9. UK Irradiated Materials Archive Steve Roberts

    E-Print Network [OSTI]

    UK Irradiated Materials Archive Steve Roberts University of Oxford MRF meeting ­ CCFE - 7th July 2015 #12;UK Irradiated Materials Archive Background · Establishment of National Nuclear User Facility exposure · Materials well suited for use in future university-based research programmes #12;UK Irradiated

  10. SIPS: Solar Irradiance Prediction System Stefan Achleitner

    E-Print Network [OSTI]

    Cerpa, Alberto E.

    SIPS: Solar Irradiance Prediction System Stefan Achleitner Computer Science and Engineering the variability and dynamics are the largest. We propose SIPS, Solar Irradiance Prediction System, a novel sensing infrastructure using wireless sensor networks (WSNs) to enable sensing of solar irradiance for solar power

  11. Vol. [VOL], No. [ISS]: 111 Irradiance Rigs

    E-Print Network [OSTI]

    Toronto, University of

    Vol. [VOL], No. [ISS]: 1­11 Irradiance Rigs Hong Yuan UMass Amherst, Disney Interactive Studios that conveys their placement in the environment. Irradiance Volumes [Greger et al. 98] were designed specifically for this problem. The spatially varying irradiance in a volume is explicitly computed

  12. STATUS OF HIGH FLUX ISOTOPE REACTOR IRRADIATION OF SILICON CARBIDE/SILICON CARBIDE JOINTS

    SciTech Connect (OSTI)

    Katoh, Yutai; Koyanagi, Takaaki; Kiggans, Jim; Cetiner, Nesrin; McDuffee, Joel

    2014-09-01

    Development of silicon carbide (SiC) joints that retain adequate structural and functional properties in the anticipated service conditions is a critical milestone toward establishment of advanced SiC composite technology for the accident-tolerant light water reactor (LWR) fuels and core structures. Neutron irradiation is among the most critical factors that define the harsh service condition of LWR fuel during the normal operation. The overarching goal of the present joining and irradiation studies is to establish technologies for joining SiC-based materials for use as the LWR fuel cladding. The purpose of this work is to fabricate SiC joint specimens, characterize those joints in an unirradiated condition, and prepare rabbit capsules for neutron irradiation study on the fabricated specimens in the High Flux Isotope Reactor (HFIR). Torsional shear test specimens of chemically vapor-deposited SiC were prepared by seven different joining methods either at Oak Ridge National Laboratory or by industrial partners. The joint test specimens were characterized for shear strength and microstructures in an unirradiated condition. Rabbit irradiation capsules were designed and fabricated for neutron irradiation of these joint specimens at an LWR-relevant temperature. These rabbit capsules, already started irradiation in HFIR, are scheduled to complete irradiation to an LWR-relevant dose level in early 2015.

  13. Microquasar Jet Irradiation of the Proto-Solar Nebula?

    E-Print Network [OSTI]

    Yousaf Butt; Nikos Prantzos

    2006-04-19

    We explore the possibility that a now-extinct microquasar may have irradiated the proto-solar neighborhood, causing the 'anomalously' high local 11B/10B isotopic ratio. Given the population and typical lifetimes of radio-emitting X-ray binaries, we find the probability of such an event having occurred is not unreasonable. We comment on some tests of the scenario that could be carried out by observing the elemental abundances in the vicinity of microquasars, in particular SS433.

  14. Assembly mechanism of recombinant spider silk proteins

    E-Print Network [OSTI]

    engineered and recombinantly produced spider dragline silk proteins eADF3 (engineered Araneus diadematus and identification of the essential parameters of dragline silk assembly. Changes in ionic conditions and pH result of these results, we propose a model for dragline silk aggregation and early steps of fiber assembly

  15. Redevelopment and smart growth at Assembly Square

    E-Print Network [OSTI]

    Savage, Alice (Alice Augusta)

    2006-01-01

    The story of Assembly Square is not yet finished. To tell the complete story of Assembly Square would require much more time to write than I had, and more time to read than the reader would likely care to devote. An earlier ...

  16. Pv-Thermal Solar Power Assembly

    DOE Patents [OSTI]

    Ansley, Jeffrey H. (El Cerrito, CA); Botkin, Jonathan D. (El Cerrito, CA); Dinwoodie, Thomas L. (Piedmont, CA)

    2001-10-02

    A flexible solar power assembly includes a flexible photovoltaic device attached to a flexible thermal solar collector. The solar power assembly can be rolled up for transport and then unrolled for installation on a surface, such as the roof or side wall of a building or other structure, by use of adhesive and/or other types of fasteners.

  17. Three-dimensional colorimetric assay assemblies

    DOE Patents [OSTI]

    Charych, Deborah (Albany, CA); Reichert, Anke (Albany, CA)

    2001-01-01

    A direct assay is described using novel three-dimensional polymeric assemblies which change from a blue to red color when exposed to an analyte, in one case a flue virus. The assemblies are typically in the form of liposomes which can be maintained in a suspension, and show great intensity in their color changes. Their method of production is also described.

  18. Multi-component assembly casting

    DOE Patents [OSTI]

    James, Allister W.

    2015-10-13

    Multi-component vane segment and method for forming the same. Assembly includes: positioning a pre-formed airfoil component (12) and a preformed shroud heat resistant material (18) in a mold, wherein the airfoil component (12) and the shroud heat resistant material (18) each comprises an interlocking feature (24); preheating the mold; introducing molten structural material (46) into the mold; and solidifying the molten structural material such that it interlocks the pre-formed airfoil component (12) with respect to the preformed shroud heat resistant material (18) and is effective to provide structural support for the shroud heat resistant material (18). Surfaces between the airfoil component (12) and the structural material (46), between the airfoil component (12) and the shroud heat resistant material (18), and between the shroud heat resistant material (18) and the structural material (46) are free of metallurgical bonds.

  19. Photonic-powered cable assembly

    DOE Patents [OSTI]

    Sanderson, Stephen N.; Appel, Titus James; Wrye, IV, Walter C.

    2013-01-22

    A photonic-cable assembly includes a power source cable connector ("PSCC") coupled to a power receive cable connector ("PRCC") via a fiber cable. The PSCC electrically connects to a first electronic device and houses a photonic power source and an optical data transmitter. The fiber cable includes an optical transmit data path coupled to the optical data transmitter, an optical power path coupled to the photonic power source, and an optical feedback path coupled to provide feedback control to the photonic power source. The PRCC electrically connects to a second electronic device and houses an optical data receiver coupled to the optical transmit data path, a feedback controller coupled to the optical feedback path to control the photonic power source, and a photonic power converter coupled to the optical power path to convert photonic energy received over the optical power path to electrical energy to power components of the PRCC.

  20. Photonic-powered cable assembly

    DOE Patents [OSTI]

    Sanderson, Stephen N; Appel, Titus James; Wrye, IV, Walter C

    2014-06-24

    A photonic-cable assembly includes a power source cable connector ("PSCC") coupled to a power receive cable connector ("PRCC") via a fiber cable. The PSCC electrically connects to a first electronic device and houses a photonic power source and an optical data transmitter. The fiber cable includes an optical transmit data path coupled to the optical data transmitter, an optical power path coupled to the photonic power source, and an optical feedback path coupled to provide feedback control to the photonic power source. The PRCC electrically connects to a second electronic device and houses an optical data receiver coupled to the optical transmit data path, a feedback controller coupled to the optical feedback path to control the photonic power source, and a photonic power converter coupled to the optical power path to convert photonic energy received over the optical power path to electrical energy to power components of the PRCC.

  1. Snubber assembly for turbine blades

    DOE Patents [OSTI]

    Marra, John J

    2013-09-03

    A snubber associated with a rotatable turbine blade in a turbine engine, the turbine blade including a pressure sidewall and a suction sidewall opposed from the pressure wall. The snubber assembly includes a first snubber structure associated with the pressure sidewall of the turbine blade, a second snubber structure associated with the suction sidewall of the turbine blade, and a support structure. The support structure extends through the blade and is rigidly coupled at a first end portion thereof to the first snubber structure and at a second end portion thereof to the second snubber structure. Centrifugal loads exerted by the first and second snubber structures caused by rotation thereof during operation of the engine are at least partially transferred to the support structure, such that centrifugal loads exerted on the pressure and suctions sidewalls of the turbine blade by the first and second snubber structures are reduced.

  2. Dual contact pogo pin assembly

    SciTech Connect (OSTI)

    Hatch, Stephen McGarry

    2015-01-20

    A contact assembly includes a base and a pair of electrical contacts supported by the base. A first end of the first electrical contact corresponds to a first end of the base and is configured to engage a first external conductive circuit element. A first end of the second electrical contact also corresponds to the first end of the base and is configured to engage a second external conductive circuit element. The first contact and the second contact are electrically isolated from one another and configured to compress when engaging an external connector element. The base includes an aperture positioned on a second end of the base outboard of a second end of the first and second electrical contacts. The aperture presents a narrowing shape with a wide mouth distal the electrical contacts and a narrow internal through-hole proximate the electrical contacts.

  3. Nanocrystal assembly for tandem catalysis

    DOE Patents [OSTI]

    Yang, Peidong; Somorjai, Gabor; Yamada, Yusuke; Tsung, Chia-Kuang; Huang, Wenyu

    2014-10-14

    The present invention provides a nanocrystal tandem catalyst comprising at least two metal-metal oxide interfaces for the catalysis of sequential reactions. One embodiment utilizes a nanocrystal bilayer structure formed by assembling sub-10 nm platinum and cerium oxide nanocube monolayers on a silica substrate. The two distinct metal-metal oxide interfaces, CeO.sub.2--Pt and Pt--SiO.sub.2, can be used to catalyze two distinct sequential reactions. The CeO.sub.2--Pt interface catalyzed methanol decomposition to produce CO and H.sub.2, which were then subsequently used for ethylene hydroformylation catalyzed by the nearby Pt--SiO.sub.2 interface. Consequently, propanal was selectively produced on this nanocrystal bilayer tandem catalyst.

  4. Low inductance power electronics assembly

    DOE Patents [OSTI]

    Herron, Nicholas Hayden; Mann, Brooks S.; Korich, Mark D.; Chou, Cindy; Tang, David; Carlson, Douglas S.; Barry, Alan L.

    2012-10-02

    A power electronics assembly is provided. A first support member includes a first plurality of conductors. A first plurality of power switching devices are coupled to the first support member. A first capacitor is coupled to the first support member. A second support member includes a second plurality of conductors. A second plurality of power switching devices are coupled to the second support member. A second capacitor is coupled to the second support member. The first and second pluralities of conductors, the first and second pluralities of power switching devices, and the first and second capacitors are electrically connected such that the first plurality of power switching devices is connected in parallel with the first capacitor and the second capacitor and the second plurality of power switching devices is connected in parallel with the second capacitor and the first capacitor.

  5. Diverter assembly for radioactive material

    DOE Patents [OSTI]

    Andrews, K.M.; Starenchak, R.W.

    1988-04-11

    A diverter assembly for diverting a pneumatically conveyed holder for a radioactive material between a central conveying tube and one of a plurality of radially offset conveying tubes includes an airtight container. A diverter tube having an offset end is suitably mounted in the container for rotation. A rotary seal seals one end of the diverter tube during and after rotation of the diverter tube while a spring biased seal seals the other end of the diverter tube which moves between various offset conveying tubes. An indexing device rotatably indexes the diverter tube and this indexing device is driven by a suitable drive. The indexing mechanism is preferably a geneva-type mechanism to provide a locking of the diverter tube in place. 3 figs.

  6. Liquid-liquid interfacial nanoparticle assemblies

    DOE Patents [OSTI]

    Emrick, Todd S. (South Deerfield, MA); Russell, Thomas P. (Amherst, MA); Dinsmore, Anthony (Amherst, MA); Skaff, Habib (Amherst, MA); Lin, Yao (Amherst, MA)

    2008-12-30

    Self-assembly of nanoparticles at the interface between two fluids, and methods to control such self-assembly process, e.g., the surface density of particles assembling at the interface; to utilize the assembled nanoparticles and their ligands in fabrication of capsules, where the elastic properties of the capsules can be varied from soft to tough; to develop capsules with well-defined porosities for ultimate use as delivery systems; and to develop chemistries whereby multiple ligands or ligands with multiple functionalities can be attached to the nanoparticles to promote the interfacial segregation and assembly of the nanoparticles. Certain embodiments use cadmium selenide (CdSe) nanoparticles, since the photoluminescence of the particles provides a convenient means by which the spatial location and organization of the particles can be probed. However, the systems and methodologies presented here are general and can, with suitable modification of the chemistries, be adapted to any type of nanoparticle.

  7. Forwardly movable assembly for a firearm

    DOE Patents [OSTI]

    Crandall, David L. (Idaho Falls, ID); Watson, Richard W. (Blackfoot, ID)

    2007-06-05

    A forwardly movable assembly for a firearm, the forwardly movable assembly adapted to be disposed in operative relationship relative to the other operative parts of a firearm, the firearm having in operative relationship each with one or more of the others: a barrel, a receiver, and at least one firing mechanism; the forwardly movable assembly comprising: the barrel and the receiver operatively connected with each other; a movable hand support structure to which at least one of the barrel and the receiver is connected, the barrel being movable therewith, the movable hand support structure being adapted to be gripped by an operator of the firearm; the forwardly movable assembly being adapted to be moved forward by an operator upon gripping the movable hand support structure and manually maneuvering the hand support structure forwardly; and, as the forwardly movable assembly is moved forwardly, the firing mechanism is completely disengaged therefrom and held substantially stationary relative thereto.

  8. Forwardly-placed firearm fire control assembly

    DOE Patents [OSTI]

    Frickey, Steven J. (Rigby, ID)

    2001-12-22

    A firearm fire control assembly for disposition in a forwardly placed support-hand operative relationship within a firearm having a combination of a firing pin and a firearm hammer adapted to engage and fire a cartridge, a sear assembly to alternately engage and disengage the combination of the firearm hammer and firing pin, and a trigger assembly including a movable trigger mechanism that is operable to engage the sear assembly to cause the firearm hammer firing pin combination to fire the firearm, a fire control assembly including a fire control depression member and a fire control rod operably connected to the depression member, and being positioned in a forward disposition disposed within a forestock of the firearm, and the depression member adapted to be operably engaged and depressed by the user's conventional forwardly placed support hand to maneuver the fire control rod to provide firing control of the firing of the firearm.

  9. Modular fuel-cell stack assembly

    DOE Patents [OSTI]

    Patel, Pinakin (Danbury, CT)

    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.

  10. High temperature control rod assembly

    DOE Patents [OSTI]

    Vollman, Russell E. (Solana Beach, CA)

    1991-01-01

    A high temperature nuclear control rod assembly comprises a plurality of substantially cylindrical segments flexibly joined together in succession by ball joints. The segments are made of a high temperature graphite or carbon-carbon composite. The segment includes a hollow cylindrical sleeve which has an opening for receiving neutron-absorbing material in the form of pellets or compacted rings. The sleeve has a threaded sleeve bore and outer threaded surface. A cylindrical support post has a threaded shaft at one end which is threadably engaged with the sleeve bore to rigidly couple the support post to the sleeve. The other end of the post is formed with a ball portion. A hollow cylindrical collar has an inner threaded surface engageable with the outer threaded surface of the sleeve to rigidly couple the collar to the sleeve. the collar also has a socket portion which cooperates with the ball portion to flexibly connect segments together to form a ball and socket-type joint. In another embodiment, the segment comprises a support member which has a threaded shaft portion and a ball surface portion. The threaded shaft portion is engageable with an inner threaded surface of a ring for rigidly coupling the support member to the ring. The ring in turn has an outer surface at one end which is threadably engageably with a hollow cylindrical sleeve. The other end of the sleeve is formed with a socket portion for engagement with a ball portion of the support member. In yet another embodiment, a secondary rod is slidably inserted in a hollow channel through the center of the segment to provide additional strength. A method for controlling a nuclear reactor utilizing the control rod assembly is also included.

  11. An evolutionary fuel assembly design for high power density BWRs

    E-Print Network [OSTI]

    Karahan, Aydin

    2007-01-01

    An evolutionary BWR fuel assembly design was studied as a means to increase the power density of current and future BWR cores. The new assembly concept is based on replacing four traditional assemblies and large water gap ...

  12. On-Orbit Assembly of Flexible Space Structures with SWARM

    E-Print Network [OSTI]

    Mohan, Swati

    On-orbit assembly is an enabling technology for many space applications. However, current methods of human assisted assembly are high in cost and risk to the crew, motivating a desire to automate the on-orbit assembly ...

  13. Easy orientation of diblock copolymers on self-assembled monolayers using UV irradiation

    E-Print Network [OSTI]

    Pang-Hung Liu; Patrick Guenoun; Jean Daillant

    2009-02-15

    A simple method based on UV/ozone treatment is proposed to control the surface energy of dense grafted silane layers for orientating block copolymer mesophases. Our method allows one to tune the surface energy down to a fraction of a mN/m. We show that related to the surface, perpendicular orientation of a lamellar phase of a PS-PMMA diblock copolymer (neutral surface) is obtained for a critical surface energy of 23.9-25.7 mN/m. Perpendicular cylinders are obtained for 24.6 mN/m and parallel cylinders for 26.8 mN/m.

  14. Gas Test Loop Functional and Technical Requirements

    SciTech Connect (OSTI)

    Glen R. Longhurst; Soli T. Khericha; James L. Jones

    2004-09-01

    This document defines the technical and functional requirements for a gas test loop (GTL) to be constructed for the purpose of providing a high intensity fast-flux irradiation environment for developers of advanced concept nuclear reactors. This capability is needed to meet fuels and materials testing requirements of the designers of Generation IV (GEN IV) reactors and other programs within the purview of the Advanced Fuel Cycle Initiative (AFCI). Space nuclear power development programs may also benefit by the services the GTL will offer. The overall GTL technical objective is to provide developers with the means for investigating and qualifying fuels and materials needed for advanced reactor concepts. The testing environment includes a fast-flux neutron spectrum of sufficient intensity to perform accelerated irradiation testing. Appropriate irradiation temperature, gaseous environment, test volume, diagnostics, and access and handling features are also needed. This document serves to identify those requirements as well as generic requirements applicable to any system of this kind.

  15. Beam Test of a Large Area nonn Silicon Strip Detector with Fast Binary Readout Electronics

    E-Print Network [OSTI]

    Beam Test of a Large Area n­on­n Silicon Strip Detector with Fast Binary Readout Electronics Y test was carried out for the non­irradiated and the irradiated detector modules. Efficiency, noise occupancy and performance in the edge regions were analyzed using the beam test data. High efficiency

  16. Self-Assembly of Polymer Nano-Elements on Sapphire

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

    Self-Assembly of Polymer Nano-Elements on Sapphire Print Self-assembly of polymers promises to vastly improve the properties and manufacturing processes of nanostructured...

  17. Bio-inspired nanocomposite assemblies as smart skin components...

    Office of Scientific and Technical Information (OSTI)

    Bio-inspired nanocomposite assemblies as smart skin components. Citation Details In-Document Search Title: Bio-inspired nanocomposite assemblies as smart skin components. There is...

  18. The Biochemistry, Ultrastructure, and Subunit Assembly Mechanism of AMPA Receptors

    E-Print Network [OSTI]

    Nakagawa, Terunaga

    2010-01-01

    to understanding the precise mechanisms of AMPA-R-mediatedSubunit assembly mechanism . Subunit assembly . Subunit .selectivity and mechanisms at recombinant heteromeric

  19. Self-Assembled, Nanostructured Carbon for Energy Storage and...

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

    Self-Assembled, Nanostructured Carbon for Energy Storage and Water Treatment Self-Assembled, Nanostructured Carbon for Energy Storage and Water Treatment nanostructuredcarbon.pdf...

  20. Chrysler: Save Energy Now Assessment Enables a Vehicle Assembly...

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

    Energy Now Assessment Enables a Vehicle Assembly Complex to Achieve Significant Natural Gas Savings Chrysler: Save Energy Now Assessment Enables a Vehicle Assembly Complex to...

  1. Analysis of the Durability of PEM FC Membrane Electrode Assemblies...

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

    the Durability of PEM FC Membrane Electrode Assemblies in Automotive Applications Analysis of the Durability of PEM FC Membrane Electrode Assemblies in Automotive Applications...

  2. Results of charpy V-notch impact testing of structural steel...

    Office of Scientific and Technical Information (OSTI)

    MATERIALS SCIENCE; FERRITIC STEELS; PHYSICAL RADIATION EFFECTS; TENSILE PROPERTIES; IRRADIATION; CHARPY TEST A capsule containing Charpy V-notch (CVN) and mini-tensile specimens...

  3. Flashback resistant pre-mixer assembly

    DOE Patents [OSTI]

    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.

  4. Dose dependence of mechanical properties in tantalum and tantalum alloys after low temperature irradiation

    SciTech Connect (OSTI)

    Byun, Thak Sang

    2008-01-01

    The dose dependence of mechanical properties was investigated for tantalum and tantalum alloys after low temperature irradiation. Miniature tensile specimens of three pure tantalum metals, ISIS Ta, Aesar Ta1, Aesar Ta2, and one tantalum alloy, Ta-1W, were irradiated by neutrons in the High Flux Isotope Reactor (HFIR) at ORNL to doses ranging from 0.00004 to 0.14 displacements per atom (dpa) in the temperature range 60 C 100 oC. Also, two tantalum-tungsten alloys, Ta-1W and Ta-10W, were irradiated by protons and spallation neutrons in the LANSCE facility at LANL to doses ranging from 0.7 to 7.5 dpa and from 0.7 to 25.2 dpa, respectively, in the temperature range 50 C 160 oC. Tensile tests were performed at room temperature and at 250oC at nominal strain rates of about 10-3 s-1. All neutron-irradiated materials underwent progressive irradiation hardening and loss of ductility with increasing dose. The ISIS Ta experienced embrittlement at 0.14 dpa, while the other metals retained significant necking ductility. Such a premature embrittlement in ISIS Ta is believed to be because of high initial oxygen concentrations picked up during a pre-irradiation anneal. The Ta-1W and Ta-10W specimens irradiated in spallation condition experienced prompt necking at yield since irradiation doses for those specimens were high ( 0.7 dpa). At the highest dose, 25.2 dpa, the Ta-10W alloy specimen broke with little necking strain. Among the test materials, the Ta-1W alloy displayed the best combination of strength and ductility. The plastic instability stress and true fracture stress were nearly independent of dose. Increasing test temperature decreased strength and delayed the onset of necking at yield.

  5. Researchers Devise New Stress Test for Irradiated Materials | Department of

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankADVANCED MANUFACTURINGEnergy BillsNo. 195 - Oct.7, 2015Verizon andSectionBioenergy

  6. Voltages across assembly joints due to direct-strike lightning currents

    SciTech Connect (OSTI)

    Dinallo, M.S. [Quatro Corp., Albuquerque, NM (United States); Fisher, R.J. [Sandia National Labs., Albuquerque, NM (United States)

    1994-08-01

    An extensive set of direct-strike lightning tests has been carried out on a set of facsimile assembly joints of the kinds employed in the design of nuclear weapon cases. Taken as a whole, the test hardware included all the conceptual design elements that are embodied, either singly or in combination, in any specific assembly joint incorporated into any stockpiled weapon. During the present testing, the effects of all key design parameters on the voltages developed across the interior of the joints were investigated under a range of lightning stroke current parameter values. Design parameter variations included the types and number of joint fasteners, mechanical preload, surface finish tolerance and coatings, and the material from which the joint assembly was fabricated. Variations of the simulated lightning stroke current included amplitude (30-, 100-, and 200-kA peak), rise time, and decay time. The maximum voltage observed on any of the test joints that incorporated proper metal-to-metal surface contact was 65 V. Typical response values were more on the order of 20 V. In order to assess the effect of the presence of a dielectric coating (either intentional or as a result of corrosion) between the mating surfaces of a joint, a special configuration was tested in which the mating parts of the test assembly were coated with a 1-mil-thick dielectric anodizing layer. First strokes to these test assemblies resulted in very narrow voltage spikes of amplitudes up to 900 V. The durations of these spikes were less than 0.1 {mu}s. However, beyond these initial spikes, the voltages typically had amplitudes of up to 400 V for durations of 3 to 5 {mu}s.

  7. Directed Self-Assembly of Nanodispersions

    SciTech Connect (OSTI)

    Furst, Eric M

    2013-11-15

    Directed self-assembly promises to be the technologically and economically optimal approach to industrial-scale nanotechnology, and will enable the realization of inexpensive, reproducible and active nanostructured materials with tailored photonic, transport and mechanical properties. These new nanomaterials will play a critical role in meeting the 21st century grand challenges of the US, including energy diversity and sustainability, national security and economic competitiveness. The goal of this work was to develop and fundamentally validate methods of directed selfassembly of nanomaterials and nanodispersion processing. The specific aims were: 1. Nanocolloid self-assembly and interactions in AC electric fields. In an effort to reduce the particle sizes used in AC electric field self-assembly to lengthscales, we propose detailed characterizations of field-driven structures and studies of the fundamental underlying particle interactions. We will utilize microscopy and light scattering to assess order-disorder transitions and self-assembled structures under a variety of field and physicochemical conditions. Optical trapping will be used to measure particle interactions. These experiments will be synergetic with calculations of the particle polarizability, enabling us to both validate interactions and predict the order-disorder transition for nanocolloids. 2. Assembly of anisotropic nanocolloids. Particle shape has profound effects on structure and flow behavior of dispersions, and greatly complicates their processing and self-assembly. The methods developed to study the self-assembled structures and underlying particle interactions for dispersions of isotropic nanocolloids will be extended to systems composed of anisotropic particles. This report reviews several key advances that have been made during this project, including, (1) advances in the measurement of particle polarization mechanisms underlying field-directed self-assembly, and (2) progress in the directed self-assembly of anisotropic nanoparticles and their unique physical properties.

  8. Dual-axis resonance testing of wind turbine blades

    DOE Patents [OSTI]

    Hughes, Scott; Musial, Walter; White, Darris

    2014-01-07

    An apparatus (100) for fatigue testing test articles (104) including wind turbine blades. The apparatus (100) includes a test stand (110) that rigidly supports an end (106) of the test article (104). An actuator assembly (120) is attached to the test article (104) and is adapted for substantially concurrently imparting first and second forcing functions in first and second directions on the test article (104), with the first and second directions being perpendicular to a longitudinal axis. A controller (130) transmits first and second sets of displacement signals (160, 164) to the actuator assembly (120) at two resonant frequencies of the test system (104). The displacement signals (160, 164) initiate the actuator assembly (120) to impart the forcing loads to concurrently oscillate the test article (104) in the first and second directions. With turbine blades, the blades (104) are resonant tested concurrently for fatigue in the flapwise and edgewise directions.

  9. Self-assembled lipid bilayer materials

    DOE Patents [OSTI]

    Sasaki, Darryl Y.; Waggoner, Tina A.; Last, Julie A.

    2005-11-08

    The present invention is a self-assembling material comprised of stacks of lipid bilayers formed in a columnar structure, where the assembly process is mediated and regulated by chemical recognition events. The material, through the chemical recognition interactions, has a self-regulating system that corrects the radial size of the assembly creating a uniform diameter throughout most of the structure. The materials form and are stable in aqueous solution. These materials are useful as structural elements for the architecture of materials and components in nanotechnology, efficient light harvesting systems for optical sensing, chemical processing centers, and drug delivery vehicles.

  10. High aspect ratio, remote controlled pumping assembly

    DOE Patents [OSTI]

    Brown, S.B.; Milanovich, F.P.

    1995-11-14

    A miniature dual syringe-type pump assembly is described which has a high aspect ratio and which is remotely controlled, for use such as in a small diameter penetrometer cone or well packer used in water contamination applications. The pump assembly may be used to supply and remove a reagent to a water contamination sensor, for example, and includes a motor, gearhead and motor encoder assembly for turning a drive screw for an actuator which provides pushing on one syringe and pulling on the other syringe for injecting new reagent and withdrawing used reagent from an associated sensor. 4 figs.

  11. High aspect ratio, remote controlled pumping assembly

    DOE Patents [OSTI]

    Brown, Steve B. (Livermore, CA); Milanovich, Fred P. (Lafayette, CA)

    1995-01-01

    A miniature dual syringe-type pump assembly which has a high aspect ratio and which is remotely controlled, for use such as in a small diameter penetrometer cone or well packer used in water contamination applications. The pump assembly may be used to supply and remove a reagent to a water contamination sensor, for example, and includes a motor, gearhead and motor encoder assembly for turning a drive screw for an actuator which provides pushing on one syringe and pulling on the other syringe for injecting new reagent and withdrawing used reagent from an associated sensor.

  12. Design of quantum dot lattices in amorphous matrices by ion beam irradiation

    SciTech Connect (OSTI)

    Buljan, M.; Bogdanovic-Radovic, I.; Karlusic, M.; Desnica, U. V.; Radic, N.; Jaksic, M.; Salamon, K.; Drazic, G.; Bernstorff, S.; Holy, V.

    2011-10-15

    We report on the highly controllable self-assembly of semiconductor quantum dots and metallic nanoparticles in a solid amorphous matrix, induced by ion beam irradiation of an amorphous multilayer. We demonstrate experimentally and theoretically a possibility to tune the basic structural properties of the quantum dots in a wide range. Furthermore, the sizes, distances, and arrangement type of the quantum dots follow simple equations dependent on the irradiation and the multilayer properties. We present a Monte Carlo model for the simulation and prediction of the structural properties of the materials formed by this method. The presented results enable engineering and simple production of functional materials or simple devices interesting for applications in nanotechnology.

  13. Principles and practices of irradiation creep experiment using pressurized mini-bellows

    SciTech Connect (OSTI)

    Byun, Thak Sang [ORNL; Li, Meimei [Argonne National Laboratory (ANL); Snead, Lance Lewis [ORNL; Katoh, Yutai [ORNL; Burchell, Timothy D [ORNL; McDuffee, Joel Lee [ORNL

    2013-01-01

    This article is to describe the key design principles and application practices of the newly developed in-reactor irradiation creep testing technology using pressurized mini-bellows. Miniature creep test frames were designed to fit into the high flux isotope reactor (HFIR) rabbit capsule whose internal diameter is slightly less than 10 mm. The most important consideration for this in-reactor creep testing technology was the ability of the small pressurized metallic bellows to survive irradiation at elevated temperatures while maintaining applied load to the specimen. Conceptual designs have been developed for inducing tension and compression stresses in specimens. Both the theoretical model and the in-furnace test confirmed that a gas-pressurized bellows can produce high enough stress to induce irradiation creep in subsize specimens. Discussion focuses on the possible stress range in specimens induced by the miniature gas-pressurized bellows and the limitations imposed by the size and structure of thin-walled bellows. A brief introduction to the in-reactor creep experiment for graphite is provided to connect to the companion paper describing the application practices and irradiation creep data. An experimental and calculation procedure to obtain in-situ applied stress values from post irradiation in-furnace force measurements is also presented.

  14. Advanced Numerical Weather Prediction Techniques for Solar Irradiance Forecasting : : Statistical, Data-Assimilation, and Ensemble Forecasting

    E-Print Network [OSTI]

    Mathiesen, Patrick James

    2013-01-01

    of numerical weather prediction solar irradiance forecastsof numerical weather prediction solar irradiance forecastsnumerical weather prediction model for solar irradiance

  15. Review of the MERIT Experiment Magnet Status and Testing Plans

    E-Print Network [OSTI]

    McDonald, Kirk

    Review of the MERIT Experiment Magnet Status and Testing Plans Monday December 10, 2005 Peter H;CVIP Vessel Manufacturing Status Final alignment of terminations was made. Gland nut components assembled well ad will be readily changed-out if needed. Instrumentation was assembled, -Not soldered

  16. Irradiation preservation of seafood: Literature review

    SciTech Connect (OSTI)

    Molton, P.M.

    1987-10-01

    The application of gamma-irradiation for extending the shelf life of seafood has been of interest for many years. This report reviews a number of studies on seafood irradiation conducted over the past several years. Topics covered include seafood irradiation techniques and dosages, species applicability and differences, the effects of packaging on seafood preservation, and changes in organoleptic acceptability as a result of irradiation. Particular attention is given to radiation effects (likely and unlikely) of concern to the public. These include the potential for generation of toxic chemical products, botulinum toxin production, and other health concerns. No scientifically defensible evidence of any kind was found for any harmful effect of irradiation of seafoods at the doses being considered (less than 300 krad), and all indications are that irradiation is an acceptable and needed additional tool for seafood preservation. 49 refs., 14 figs., 14 tabs.

  17. INFLUENCE OF SPECIMEN SIZE/TYPE ON THE FRACTURE TOUGHNESS OF FIVE IRRADIATED RPV MATERIALS

    SciTech Connect (OSTI)

    Sokolov, Mikhail A; Lucon, Enrico

    2015-01-01

    The Heavy-Section Steel Irradiation (HSSI) Program had previously irradiated five reactor pressure vessel (RPV) steels/welds at fast neutron fluxes of about 4 to 8 1011 n/cm2/s (>1 MeV) to fluences from 0.5 to 3.4 1019 n/cm2 and at 288 C. The unirradiated fracture toughness tests were performed by Oak Ridge National Laboratory with 12.7-mm and 25.4-mm thick (0.5T and 1T) compact specimens, while the HSSI Program provided tensile and 5 10-mm three-point bend specimens to SCK CEN for irradiation in the in-pile section of the Belgian Reactor BR2 at fluxes >1013 n/cm2/s and subsequent testing by SCK CEN. The BR2 irradiations were conducted at about 2 and 4 1013 n/cm2/s with irradiation temperature between 295 C and 300 C (water temperature), and to fluences between 6 and10 1019 n/cm2. The irradiation-induced shifts of the Master Curve reference temperatures, T0, for most of the materials deviated from the embrittlement correlations much more than expected, motivating the testing of 5 10-mm three-point bend specimens of all five materials in the unirradiated condition to eliminate specimen size and geometry as a variable. Tests of the unirradiated small bend specimens resulted in Master Curve reference temperatures, T0, 25 C to 53 C lower than those from the larger compact specimens, meaning that the irradiation-induced reference temperature shifts, T0, were larger than the initial measurements, resulting in much improved agreement between the measured and predicted fracture toughness shifts.

  18. DNA Assembly Line for Nano-Construction

    ScienceCinema (OSTI)

    Oleg Gang

    2010-01-08

    Building on the idea of using DNA to link up nanoparticles scientists at Brookhaven National Lab have designed a molecular assembly line for high-precision nano-construction. Nanofabrication is essential for exploiting the unique properties of nanoparticl

  19. Method of forming and assembly of parts

    DOE Patents [OSTI]

    Ripley, Edward B. (Knoxville, TN)

    2010-12-28

    A method of assembling two or more parts together that may be metal, ceramic, metal and ceramic parts, or parts that have different CTE. Individual parts are formed and sintered from particles that leave a network of interconnecting porosity in each sintered part. The separate parts are assembled together and then a fill material is infiltrated into the assembled, sintered parts using a method such as capillary action, gravity, and/or pressure. The assembly is then cured to yield a bonded and fully or near-fully dense part that has the desired physical and mechanical properties for the part's intended purpose. Structural strength may be added to the parts by the inclusion of fibrous materials.

  20. Rack assembly for mounting solar modules

    DOE Patents [OSTI]

    Plaisted, Joshua Reed (Oakland, CA); West, Brian (San Francisco, CA)

    2010-12-28

    A rack assembly is provided for mounting solar modules over an underlying body. The rack assembly may include a plurality of rail structures that are arrangeable over the underlying body to form an overall perimeter for the rack assembly. One or more retention structures may be provided with the plurality of rail structures, where each retention structure is configured to support one or more solar modules at a given height above the underlying body. At least some of the plurality of rail structures are adapted to enable individual rail structures o be sealed over the underlying body so as to constrain air flow underneath the solar modules. Additionally, at least one of (i) one or more of the rail structures, or (ii) the one or more retention structures are adjustable so as to adapt the rack assembly to accommodate solar modules of varying forms or dimensions.

  1. BWR Assembly Optimization for Minor Actinide Recycling

    SciTech Connect (OSTI)

    G. Ivan Maldonado; John M. Christenson; J.P. Renier; T.F. Marcille; J. Casal

    2010-03-22

    The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs).

  2. Helicopter final assembly critical path analysis

    E-Print Network [OSTI]

    Daigh, Sara L. (Sarah Louise), 1981-

    2012-01-01

    Helicopter final assembly involves the installation of hundreds of components into the aircraft and takes thousands of man-hours. Meeting production targets such as total build days and total aircraft man-hours can be ...

  3. Method of making a unitized electrode assembly

    DOE Patents [OSTI]

    Niksa, M.J.; Pohto, G.R.; Lakatos, L.K.; Wheeler, D.J.; Solomon, F.; Niksa, A.J.; Schue, T.J.; Genodman, Y.; Turk, T.R.; Hagel, D.P.

    1988-12-06

    A battery assembly of the consumable metal anode type has now been constructed for ready assembly as well as disassembly. In a non-conductive and at least substantially inert cell body, space is provided for receiving an open-structured, non-consumable anode cage. The cage has an open top for facilitating insertion of an anode. A modular cathode is used, comprising a peripheral current conductor frame clamped about a grid reinforced air cathode in sheet form. The air cathode may be double gridded. The cathode frame can be sealed, during assembly, with electrolyte-resistant-sealant as well as with adhesive. The resulting cathode module can be assembled outside the cell body and readily inserted therein, or can later be easily removed therefrom. 6 figs.

  4. Reactivity control assembly for nuclear reactor

    DOE Patents [OSTI]

    Bollinger, Lawrence R. (Schenectady, NY)

    1984-01-01

    Reactivity control assembly for nuclear reactor comprises supports stacked above reactor core for holding control rods. Couplers associated with the supports and a vertically movable drive shaft have lugs at their lower ends for engagement with the supports.

  5. Nozzle and shroud assembly mounting structure

    DOE Patents [OSTI]

    Faulder, Leslie J. (San Diego, CA); Frey, deceased, Gary A. (late of Seattle, WA); Nielsen, Engward W. (El Cajon, CA); Ridler, Kenneth J. (San Diego, CA)

    1997-01-01

    The present nozzle and shroud assembly mounting structure configuration increases component life and reduces maintenance by reducing internal stress between the mounting structure having a preestablished rate of thermal expansion and the nozzle and shroud assembly having a preestablished rate of thermal expansion being less than that of the mounting structure. The mounting structure includes an outer sealing portion forming a cradling member in which an annular ring member is slidably positioned. The mounting structure further includes an inner mounting portion to which a hooked end of the nozzle and shroud assembly is attached. As the inner mounting portion expands and contracts, the nozzle and shroud assembly slidably moves within the outer sealing portion.

  6. Nozzle and shroud assembly mounting structure

    DOE Patents [OSTI]

    Faulder, L.J.; Frey, G.A.; Nielsen, E.W.; Ridler, K.J.

    1997-08-05

    The present nozzle and shroud assembly mounting structure configuration increases component life and reduces maintenance by reducing internal stress between the mounting structure having a preestablished rate of thermal expansion and the nozzle and shroud assembly having a preestablished rate of thermal expansion being less than that of the mounting structure. The mounting structure includes an outer sealing portion forming a cradling member in which an annular ring member is slidably positioned. The mounting structure further includes an inner mounting portion to which a hooked end of the nozzle and shroud assembly is attached. As the inner mounting portion expands and contracts, the nozzle and shroud assembly slidably moves within the outer sealing portion. 3 figs.

  7. Productivity Improvement of a Manual Assembly Line 

    E-Print Network [OSTI]

    Yerasi, Pranavi

    2012-10-19

    The current project addresses the productivity improvement of a manual assembly line by making use of operations analysis in the framework of Lean production. A methodology is proposed that helps to improve the productivity of any production process...

  8. Rack assembly for mounting solar modules

    DOE Patents [OSTI]

    Plaisted, Joshua Reed; West, Brian

    2012-09-04

    A rack assembly is provided for mounting solar modules over an underlying body. The rack assembly may include a plurality of rail structures that are arrangeable over the underlying body to form an overall perimeter for the rack assembly. One or more retention structures may be provided with the plurality of rail structures, where each retention structure is configured to support one or more solar modules at a given height above the underlying body. At least some of the plurality of rail structures are adapted to enable individual rail structures to be sealed over the underlying body so as to constrain air flow underneath the solar modules. Additionally, at least one of (i) one or more of the rail structures, or (ii) the one or more retention structures are adjustable so as to adapt the rack assembly to accommodate solar modules of varying forms or dimensions.

  9. Defects in Self Assembled Colloidal Crystals

    E-Print Network [OSTI]

    Koh, Yaw Koon

    Colloidal self assembly is an efficient method for making 3-D ordered nanostructures suitable for materials such as photonic crystals and macroscopic solids for catalysis and sensor applications. Colloidal crystals grown ...

  10. Carbon nanotubes : a study on assembly methods

    E-Print Network [OSTI]

    Quiñones, Lisandro E. (Quiñones Ortiz)

    2008-01-01

    The urgent stipulation is to manufacture CNTs of desired properties and dimensions. The heart of this yearning lies in understanding the growth and assembly methods of CNTs, which are not yet clear. In this study, hence, ...

  11. Project Profile: Flexible Assembly Solar Technology | Department...

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

    BrightSource Energy logo BrightSource Energy, under the 2012 SunShot Concentrating Solar Power (CSP) R&D FOA, is designing and deploying an automated collector-assembly...

  12. Thermal Analysis of a TREAT Fuel Assembly

    SciTech Connect (OSTI)

    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.

  13. PV/thermal solar power assembly

    DOE Patents [OSTI]

    Ansley, Jeffrey H.; Botkin, Jonathan D.; Dinwoodie, Thomas L.

    2004-01-13

    A flexible solar power assembly (2) includes a flexible photovoltaic device (16) attached to a flexible thermal solar collector (4). The solar power assembly can be rolled up for transport and then unrolled for installation on a surface, such as the roof (20, 25) or side wall of a building or other structure, by use of adhesive and/or other types of fasteners (23).

  14. Fuel cell with electrolyte matrix assembly

    DOE Patents [OSTI]

    Kaufman, Arthur (West Orange, NJ); Pudick, Sheldon (Sayreville, NJ); Wang, Chiu L. (Edison, NJ)

    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.

  15. Characterization of the neutron irradiation system for use in the Low-Dose-Rate Irradiation Facility at Sandia National Laboratories.

    SciTech Connect (OSTI)

    Franco, Manuel,

    2014-08-01

    The objective of this work was to characterize the neutron irradiation system consisting of americium-241 beryllium (241AmBe) neutron sources placed in a polyethylene shielding for use at Sandia National Laboratories (SNL) Low Dose Rate Irradiation Facility (LDRIF). With a total activity of 0.3 TBq (9 Ci), the source consisted of three recycled 241AmBe sources of different activities that had been combined into a single source. The source in its polyethylene shielding will be used in neutron irradiation testing of components. The characterization of the source-shielding system was necessary to evaluate the radiation environment for future experiments. Characterization of the source was also necessary because the documentation for the three component sources and their relative alignment within the Special Form Capsule (SFC) was inadequate. The system consisting of the source and shielding was modeled using Monte Carlo N-Particle transport code (MCNP). The model was validated by benchmarking it against measurements using multiple techniques. To characterize the radiation fields over the full spatial geometry of the irradiation system, it was necessary to use a number of instruments of varying sensitivities. First, the computed photon radiography assisted in determining orientation of the component sources. With the capsule properly oriented inside the shielding, the neutron spectra were measured using a variety of techniques. A N-probe Microspec and a neutron Bubble Dosimeter Spectrometer (BDS) set were used to characterize the neutron spectra/field in several locations. In the third technique, neutron foil activation was used to ascertain the neutron spectra. A high purity germanium (HPGe) detector was used to characterize the photon spectrum. The experimentally measured spectra and the MCNP results compared well. Once the MCNP model was validated to an adequate level of confidence, parametric analyses was performed on the model to optimize for potential experimental configurations and neutron spectra for component irradiation. The final product of this work is a MCNP model validated by measurements, an overall understanding of neutron irradiation system including photon/neutron transport and effective dose rates throughout the system, and possible experimental configurations for future irradiation of components.

  16. National Wind Tecnology Center Provides Dual Axis Resonant Blade Testing

    ScienceCinema (OSTI)

    Felker, Fort

    2014-06-10

    NREL's Structural Testing Laboratory at the National Wind Technology Center (NWTC) provides experimental laboratories, computer facilities for analytical work, space for assembling components and turbines for atmospheric testing as well as office space for industry researchers. Fort Felker, center director at the NWTC, discusses NREL's state-of-the-art structural testing capabilities and shows a flapwise and edgewise blade test in progress.

  17. National Wind Tecnology Center Provides Dual Axis Resonant Blade Testing

    SciTech Connect (OSTI)

    Felker, Fort

    2013-11-13

    NREL's Structural Testing Laboratory at the National Wind Technology Center (NWTC) provides experimental laboratories, computer facilities for analytical work, space for assembling components and turbines for atmospheric testing as well as office space for industry researchers. Fort Felker, center director at the NWTC, discusses NREL's state-of-the-art structural testing capabilities and shows a flapwise and edgewise blade test in progress.

  18. CU-LASP Test Facilities ! and Instrument Calibration Capabilities"

    E-Print Network [OSTI]

    Mojzsis, Stephen J.

    ­ Computer-controlled monochromator for wavelength selection ­ Deuterium lamp is primary light source ­ 2°C #12;Calibration and Test Equipment 3 (CTE 3) · Characterization of detectors and detector modules;Total Solar Irradiance Radiometer Facility (TRF) · Total Solar Irradiance (TSI) instrument calibrations

  19. Enterprise Assessments, Oak Ridge National Laboratory Irradiated...

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

    of the Safety-Significant Ventilation Systems at the Irradiated Fuels Examination Laboratory Operated by UT-Battelle for the Oak Ridge National Laboratory Office of Science The...

  20. Modeling Thermal Fatigue in CPV Cell Assemblies (Presentation)

    SciTech Connect (OSTI)

    Bosco, N.; Panchagade, D.; Kurtz, S.

    2011-02-01

    This presentation outlines the modeling of thermal fatigue in concentrating photovoltaic (CPV) assemblies.

  1. Housing assembly for electric vehicle transaxle

    DOE Patents [OSTI]

    Kalns, Ilmars (Northville, MI)

    1981-01-01

    Disclosed is a drive assembly (10) for an electrically powered vehicle (12). The assembly includes a transaxle (16) having a two-speed transmission (40) and a drive axle differential (46) disposed in a unitary housing assembly (38), an oil-cooled prime mover or electric motor (14) for driving the transmission input shaft (42), an adapter assembly (24) for supporting the prime mover on the transaxle housing assembly, and a hydraulic system (172) providing pressurized oil flow for cooling and lubricating the electric motor and transaxle and for operating a clutch (84) and a brake (86) in the transmission to shift between the two-speed ratios of the transmission. The adapter assembly allows the prime mover to be supported in several positions on the transaxle housing. The brake is spring-applied and locks the transmission in its low-speed ratio should the hydraulic system fail. The hydraulic system pump is driven by an electric motor (212) independent of the prime mover and transaxle.

  2. System design description for GCFR-core flow test loop

    SciTech Connect (OSTI)

    Huntley, W.R.; Grindell, A.G.

    1980-12-01

    The Core Flow Test Loop is a high-pressure, high-temperature, out-of-reactor helium circulation system that is being constructed to permit detailed study of the thermomechanical and thermal performance at prototypic steady-state and transient operating conditions of simulated segments of core assemblies for a GCFR Demonstration Plant, as designed by General Atomic Company. It will also permit the expermental verification of predictive analytical models of the GCFR core assemblies needed to reduce operational and safety uncertainties of the GCFR. Full-sized blanket assemblies and segments of fuel rod and control rod fuel assemblies will be simulated with test bundles of electrically powered fuel rod or blanket rod simulators. The loop will provide the steady-state and margin test requirements of bundle power and heat removal, and of helium coolant flow rate, pressure, and temperature for test bundles having up to 91 rods; these requirements set the maximum power, coolant helium flow, and thermal requirements for the loop. However, the size of the test vessel that contains the test bundles will be determined by the bundles that simulate a full-sized GCFR blanket assembly. The loop will also provide for power and coolant transients to simulate transient operation of GCFR core assemblies, including the capability for rapid helium depressurization to simulate the depressurization class of GCFR accidents. In addition, the loop can be used as an out-of-reactor test bed for characterizing in-reactor test bundle configurations.

  3. Mechanism of Irradiation Assisted Cracking of Core Components in Light Water Reactors

    SciTech Connect (OSTI)

    Gary S. Was; Michael Atzmon; Lumin Wang

    2003-04-28

    The overall goal of the project is to determine the mechanism of irradiation assisted stress corrosion cracking (IASCC). IASCC has been linked to hardening, microstructural and microchemical changes during irradiation. Unfortunately, all of these changes occur simultaneously and at similar rates during irradiation, making attribution of IASCC to any one of these features nearly impossible to determine. The strategy set forth in this project is to develop means to separate microstructural from microchemical changes to evaluate each separately for their effect on IASCC. In the first part, post irradiation annealing (PIA) treatments are used to anneal the irradiated microstructure, leaving only radiation induced segregation (RIS) for evaluation for its contribution to IASCC. The second part of the strategy is to use low temperature irradiation to produce a radiation damage dislocation loop microstructure without radiation induced segregation in order to evaluate the effect of the dislocation microstructure alone. A radiation annealing model was developed based on the elimination of dislocation loops by vacancy absorption. Results showed that there were indeed, time-temperature annealing combinations that leave the radiation induced segregation profile largely unaltered while the dislocation microstructure is significantly reduced. Proton irradiation of 304 stainless steel irradiated with 3.2 MeV protons to 1.0 or 2.5 dpa resulted in grain boundary depletion of chromium and enrichment of nickel and a radiation damaged microstructure. Post irradiation annealing at temperatures of 500 ? 600°C for times of up to 45 min. removed the dislocation microstructure to a greater degree with increasing temperatures, or times at temperature, while leaving the radiation induced segregation profile relatively unaltered. Constant extension rate tensile (CERT) experiments in 288°C water containing 2 ppm O2 and with a conductivity of 0.2 mS/cm and at a strain rate of 3 x 10-7 s-1 showed that the IASCC susceptibility, as measured by the crack length per unit strain, decreased with very short anneals and was almost completely removed by an anneal at 500°C for 45 min. This annealing treatment removed about 15% of the dislocation microstructure and the irradiation hardening, but did not affect the grain boundary chromium depletion or nickel segregation, nor did it affect the grain boundary content of other minor impurities. These results indicate that RIS is not the sole controlling feature of IASCC in irradiated stainless steels in normal water chemistry. The isolation of the irradiated microstructure was approached using low temperature irradiation or combinations of low and high temperature irradiations to achieve a stable, irradiated microstructure without RIS. Experiments were successful in achieving a high degree of irradiation hardening without any evidence of RIS of either major or minor elements. The low temperature irradiations to doses up to 0.3 dpa at T<75°C were also very successful in producing hardening to levels considerably above that for irradiations conducted under nominal conditions of 1 dpa at 360°C. However, the microstructure consisted of an extremely fine dispersion of defect clusters of sizes that are not resolvable by either transmission electron microscopy (TEM) or small angle x-ray scattering (SAXS). The microstructure was not stable at the 288°C IASCC test temperature and resulted in rapid reduction of hardening and presumably, annealing of the defect clusters at this temperature as well. Nevertheless, the annealing studies showed that treatments that resulted in significant decreases in the hardening produced small changes in the dislocation microstructure that were confined to the elimination of the finest of loops (~1 nm). These results substantiate the importance of the very fine defect microstructure in the IASCC process. The results of this program provide the first definitive evidence that RIS is not the sole controlling factor in the irradiation assisted stress corrosion cracking of austenitic stain

  4. Post-irradiation Examination Plan for ORNL and University of California Santa Barbara Assessment of UCSB ATR-2 Irradiation Experiment

    SciTech Connect (OSTI)

    Nanstad, R. K. [Materials Science and Technology Division, Oak Ridge National Laboratory; Yamamoto, T. [University of California Santa Barbara; Sokolov, M. A. [Materials Science and Technology Division, Oak Ridge National Laboratory

    2014-01-25

    New and existing databases will be combined to support development of physically based models of transition temperature shifts (TTS) for high fluence-low flux (? < 10{sup 11}n/cm{sup 2}-s) conditions, beyond the existing surveillance database, to neutron fluences of at least 1×10{sup 20} n/cm{sup 2} (>1 MeV). All references to neutron flux and fluence in this report are for fast neutrons (>1 MeV). The reactor pressure vessel (RPV) task of the Light Water Reactor Sustainability (LWRS) Program is working with various organizations to obtain archival surveillance materials from commercial nuclear power plants to allow for comparisons of the irradiation-induced microstructural features from reactor surveillance materials with those from similar materials irradiated under high flux conditions in test reactors

  5. AGC-2 Irradiation Data Qualification Final Report

    SciTech Connect (OSTI)

    Laurence C. Hull

    2012-07-01

    The Graphite Technology Development Program will run a series of six experiments to quantify the effects of irradiation on nuclear grade graphite. The second Advanced Graphite Creep (AGC) experiment (AGC-2) began with Advanced Test Reactor (ATR) Cycle 149A on April 12, 2011, and ended with ATR Cycle 151B on May 5, 2012. The purpose of this report is to qualify AGC-2 irradiation monitoring data following INL Management and Control Procedure 2691, Data Qualification. Data that are Qualified meet the requirements for data collection and use as described in the experiment planning and quality assurance documents. Data that do not meet the requirements are Failed. Some data may not quite meet the requirements, but may still provide some useable information. These data are labeled as Trend. No Trend data were identified for the AGC-2 experiment. All thermocouples functioned throughout the AGC-2 experiment. There was one instance where spurious signals or instrument power interruption resulted in a recorded temperature value being well outside physical reality. This value was identified and labeled as Failed data. All other temperature data are Qualified. All helium and argon gas flow data are within expected ranges. Total gas flow was approximately 50 sccm through the capsule. Helium gas flow was briefly increased to 100 sccm during reactor shutdown. All gas flow data are Qualified. At the start of the experiment, moisture in the outflow gas line increased to 200 ppmv then declined to less than 10 ppmv over a period of 5 days. This increase in moisture coincides with the initial heating of the experiment and drying of the system. Moisture slightly exceeded 10 ppmv three other times during the experiment. While these moisture values exceed the 10 ppmv threshold value, the reported measurements are considered accurate and to reflect moisture conditions in the capsule. All moisture data are Qualified. Graphite creep specimens are subjected to one of three loads, 393 lbf, 491 lbf, or 589 lbf. Loads were consistently within 5% of the specified values throughout the experiment. Stack displacement increased consistently throughout the experiment with total displacement ranging from 1 to 1.5 inches. No anomalous values were identified. During reactor outages, a set of pneumatic rams are used to raise the stacks of graphite creep specimens to ensure the specimens have not become stuck within the test train. This stack raising was performed after all cycles when the capsule was in the reactor. All stacks were raised successfully after each cycle. The load and displacement data are Qualified

  6. Correlation of Clinical and Dosimetric Factors With Adverse Pulmonary Outcomes in Children After Lung Irradiation

    SciTech Connect (OSTI)

    Venkatramani, Rajkumar; Kamath, Sunil; Wong, Kenneth; Malvar, Jemily; Sposto, Richard; Goodarzian, Fariba; Freyer, David R.; Keens, Thomas G.; and others

    2013-08-01

    Purpose: To identify the incidence and the risk factors for pulmonary toxicity in children treated for cancer with contemporary lung irradiation. Methods and Materials: We analyzed clinical features, radiographic findings, pulmonary function tests, and dosimetric parameters of children receiving irradiation to the lung fields over a 10-year period. Results: We identified 109 patients (75 male patients). The median age at irradiation was 13.8 years (range, 0.04-20.9 years). The median follow-up period was 3.4 years. The median prescribed radiation dose was 21 Gy (range, 0.4-64.8 Gy). Pulmonary toxic chemotherapy included bleomycin in 58.7% of patients and cyclophosphamide in 83.5%. The following pulmonary outcomes were identified and the 5-year cumulative incidence after irradiation was determined: pneumonitis, 6%; chronic cough, 10%; pneumonia, 35%; dyspnea, 11%; supplemental oxygen requirement, 2%; radiographic interstitial lung disease, 40%; and chest wall deformity, 12%. One patient died of progressive respiratory failure. Post-irradiation pulmonary function tests available from 44 patients showed evidence of obstructive lung disease (25%), restrictive disease (11%), hyperinflation (32%), and abnormal diffusion capacity (12%). Thoracic surgery, bleomycin, age, mean lung irradiation dose (MLD), maximum lung dose, prescribed dose, and dosimetric parameters between V{sub 22} (volume of lung exposed to a radiation dose ?22 Gy) and V{sub 30} (volume of lung exposed to a radiation dose ?30 Gy) were significant for the development of adverse pulmonary outcomes on univariate analysis. MLD, maximum lung dose, and V{sub dose} (percentage of volume of lung receiving the threshold dose or greater) were highly correlated. On multivariate analysis, MLD was the sole significant predictor of adverse pulmonary outcome (P=.01). Conclusions: Significant pulmonary dysfunction occurs in children receiving lung irradiation by contemporary techniques. MLD rather than prescribed dose should be used to perform risk stratification of patients receiving lung irradiation.

  7. Biomolecular Assembly of Gold Nanocrystals

    SciTech Connect (OSTI)

    Micheel, Christine Marya

    2005-05-20

    Over the past ten years, methods have been developed to construct discrete nanostructures using nanocrystals and biomolecules. While these frequently consist of gold nanocrystals and DNA, semiconductor nanocrystals as well as antibodies and enzymes have also been used. One example of discrete nanostructures is dimers of gold nanocrystals linked together with complementary DNA. This type of nanostructure is also known as a nanocrystal molecule. Discrete nanostructures of this kind have a number of potential applications, from highly parallel self-assembly of electronics components and rapid read-out of DNA computations to biological imaging and a variety of bioassays. My research focused in three main areas. The first area, the refinement of electrophoresis as a purification and characterization method, included application of agarose gel electrophoresis to the purification of discrete gold nanocrystal/DNA conjugates and nanocrystal molecules, as well as development of a more detailed understanding of the hydrodynamic behavior of these materials in gels. The second area, the development of methods for quantitative analysis of transmission electron microscope data, used computer programs written to find pair correlations as well as higher order correlations. With these programs, it is possible to reliably locate and measure nanocrystal molecules in TEM images. The final area of research explored the use of DNA ligase in the formation of nanocrystal molecules. Synthesis of dimers of gold particles linked with a single strand of DNA possible through the use of DNA ligase opens the possibility for amplification of nanostructures in a manner similar to polymerase chain reaction. These three areas are discussed in the context of the work in the Alivisatos group, as well as the field as a whole.

  8. The Materials Test Station Eric Pitcher

    E-Print Network [OSTI]

    McDonald, Kirk

    are transferred to shipping casks in service cell #12;AHIPA Workshop, Fermilab, October 19, 2009 test fuel spectrum fuel and materials irradiation testing facility · MTS will be driven by a 1-MW proton beam delivered by the LANSCE accelerator · Spallation reactions produce 1017 neutrons per second fuel module

  9. Method of making hermetic seals for hermetic terminal assemblies

    DOE Patents [OSTI]

    Hsu, John S.; Marlino, Laura D.; Ayers, Curtis W.

    2010-04-13

    This invention teaches methods of making a hermetic terminal assembly comprising the steps of: inserting temporary stops, shims and jigs on the bottom face of a terminal assembly thereby blocking assembly core open passageways; mounting the terminal assembly inside a vacuum chamber using a temporary assembly perimeter seal and flange or threaded assembly interfaces; mixing a seal admixture and hardener in a mixer conveyor to form a polymer seal material; conveying the polymer seal material into a polymer reservoir; feeding the polymer seal material from the reservoir through a polymer outlet valve and at least one polymer outlet tube into the terminal assembly core thereby filling interstitial spaces in the core adjacent to service conduits, temporary stop, and the terminal assembly casing; drying the polymer seal material at room temperature thereby hermetically sealing the core of the terminal assembly; removing the terminal assembly from the vacuum chamber, and; removing the temporary stops, shims.

  10. MOX Lead Assembly Fabrication at the Savannah River Site

    SciTech Connect (OSTI)

    Geddes, R.L.; Spiker, D.L.; Poon, A.P.

    1997-12-01

    The U. S. Department of Energy (DOE) announced its intent to prepare an Environmental Impact Statement (EIS) under the National Environmental Policy Act (NEPA) on the disposition of the nations weapon-usable surplus plutonium.This EIS is tiered from the Storage and Disposition of Weapons-Usable Fissile Material Programmatic Environmental Impact Statement issued in December 1996,and the associated Record of Decision issued on January, 1997. The EIS will examine reasonable alternatives and potential environmental impacts for the proposed siting, construction, and operation of three types of facilities for plutonium disposition. The three types of facilities are: a pit disassembly and conversion facility, a facility to immobilize surplus plutonium in a glass or ceramic form for disposition, and a facility to fabricate plutonium oxide into mixed oxide (MOX) fuel.As an integral part of the surplus plutonium program, Oak Ridge National Laboratory (ORNL) was tasked by the DOE Office of Fissile Material Disposition(MD) as the technical lead to organize and evaluate existing facilities in the DOE complex which may meet MD`s need for a domestic MOX fuel fabrication demonstration facility. The Lead Assembly (LA) facility is to produce 1 MT of usable test fuel per year for three years. The Savannah River Site (SRS) as the only operating plutonium processing site in the DOE complex, proposes two options to carry out the fabrication of MOX fuel lead test assemblies: an all Category I facility option and a combined Category I and non-Category I facilities option.

  11. An improved characterization method for international accountancy measurements of fresh and irradiated mixed oxide (MOX) fuel: helping achieve continual monitoring and safeguards through the fuel cycle

    SciTech Connect (OSTI)

    Evans, Louise G; Croft, Stephen; Swinhoe, Martyn T; Tobin, S. J.; Boyer, B. D.; Menlove, H. O.; Schear, M. A.; Worrall, Andrew

    2010-11-24

    Nuclear fuel accountancy measurements are conducted at several points through the nuclear fuel cycle to ensure continuity of knowledge (CofK) of special nuclear material (SNM). Non-destructive assay (NDA) measurements are performed on fresh fuel (prior to irradiation in a reactor) and spent nuclear fuel (SNF) post-irradiation. We have developed a fuel assembly characterization system, based on the novel concept of 'neutron fingerprinting' with multiplicity signatures to ensure detailed CofK of nuclear fuel through the entire fuel cycle. The neutron fingerprint in this case is determined by the measurement of the various correlated neutron signatures, specific to fuel isotopic composition, and therefore offers greater sensitivity to variations in fissile content among fuel assemblies than other techniques such as gross neutron counting. This neutron fingerprint could be measured at the point of fuel dispatch (e.g. from a fuel fabrication plant prior to irradiation, or from a reactor site post-irradiation), monitored during transportation of the fuel assembly, and measured at a subsequent receiving site (e.g. at the reactor site prior to irradiation, or reprocessing facility post-irradiation); this would confirm that no unexpected changes to the fuel composition or amount have taken place during transportation and/or reactor operations. Changes may indicate an attempt to divert material for example. Here, we present the current state of the practice of fuel measurements for both fresh mixed oxide (MOX) fuel and SNF (both MOX and uranium dioxide). This is presented in the framework of international safeguards perspectives from the US and UK. We also postulate as to how the neutron fingerprinting concept could lead to improved fuel characterization (both fresh MOX and SNF) resulting in: (a) assured CofK of fuel across the nuclear fuel cycle, (b) improved detection of SNM diversion, and (c) greater confidence in safeguards of SNF transportation.

  12. An improved characterization method for international accountancy measurements of fresh and irradiated mixed oxide (MOX) fuel: helping achieve continual monitoring and safeguards through the fuel cycle

    SciTech Connect (OSTI)

    Evans, Louise G; Croft, Stephen; Swinhoe, Martyn T; Tobin, S. J.; Menlove, H. O.; Schear, M. A.; Worrall, Andrew

    2011-01-13

    Nuclear fuel accountancy measurements are conducted at several points through the nuclear fuel cycle to ensure continuity of knowledge (CofK) of special nuclear material (SNM). Non-destructive assay (NDA) measurements are performed on fresh fuel (prior to irradiation in a reactor) and spent nuclear fuel (SNF) post-irradiation. We have developed a fuel assembly characterization system, based on the novel concept of 'neutron fingerprinting' with multiplicity signatures to ensure detailed CofK of nuclear fuel through the entire fuel cycle. The neutron fingerprint in this case is determined by the measurement of the various correlated neutron signatures, specific to fuel isotopic composition, and therefore offers greater sensitivity to variations in fissile content among fuel assemblies than other techniques such as gross neutron counting. This neutron fingerprint could be measured at the point of fuel dispatch (e.g. from a fuel fabrication plant prior to irradiation, or from a reactor site post-irradiation), monitored during transportation of the fuel assembly, and measured at a subsequent receiving site (e.g. at the reactor site prior to irradiation, or reprocessing facility post-irradiation); this would confirm that no unexpected changes to the fuel composition or amount have taken place during transportation and/ or reactor operations. Changes may indicate an attempt to divert material for example. Here, we present the current state of the practice of fuel measurements for both fresh mixed oxide (MOX) fuel and SNF (both MOX and uranium dioxide). This is presented in the framework of international safeguards perspectives from the US and UK. We also postulate as to how the neutron fingerprinting concept could lead to improved fuel characterization (both fresh MOX and SNF) resulting in: (a) assured CofK of fuel across the nuclear fuel cycle, (b) improved detection of SNM diversion, and (c) greater confidence in safeguards of SNF transportation.

  13. Microstructure and Mechanical Properties of n-irradiated Fe-Cr Model Alloys

    SciTech Connect (OSTI)

    Matijasevic, Milena; Al Mazouzi, Abderrahim

    2008-07-01

    High chromium ( 9-12 wt %) ferritic/martensitic steels are candidate structural materials for future fusion reactors and other advanced systems such as accelerator driven systems (ADS). Their use for these applications requires a careful assessment of their mechanical stability under high energy neutron irradiation and in aggressive environments. In particular, the Cr concentration has been shown to be a key parameter to be optimized in order to guarantee the best corrosion and swelling resistance, together with the least embrittlement. In this work, the characterization of the neutron irradiated Fe-Cr model alloys with different Cr % with respect to microstructure and mechanical tests will be presented. The behavior of Fe-Cr alloys have been studied using tensile tests at different temperature range ( from -160 deg. C to 300 deg. C). Irradiation-induced microstructure changes have been studied by TEM for two different irradiation doses at 300 deg. C. The density and the size distribution of the defects induced have been determined. The tensile test results indicate that Cr content affects the hardening behavior of Fe-Cr binary alloys. Hardening mechanisms are discussed in terms of Orowan type of approach by correlating TEM data to the measured irradiation hardening. (authors)

  14. Microhole Coiled Tubing Bottom Hole Assemblies

    SciTech Connect (OSTI)

    Don Macune

    2008-06-30

    The original objective of the project, to deliver an integrated 3 1/8-inch diameter Measurement While Drilling (MWD) and Logging While Drilling (LWD) system for drilling small boreholes using coiled tubing drilling, has been achieved. Two prototype systems have been assembled and tested in the lab. One of the systems has been successfully tested downhole in a conventional rotary drilling environment. Development of the 3 1/8-inch system has also lead to development and commercialization of a slightly larger 3.5-inch diameter system. We are presently filling customer orders for the 3.5-inch system while continuing with commercialization of the 3 1/8-inch system. The equipment developed by this project will be offered for sale to multiple service providers around the world, enabling the more rapid expansion of both coiled tubing drilling and conventional small diameter drilling. The project was based on the reuse of existing technology whenever possible in order to minimize development costs, time, and risks. The project was begun initially by Ultima Labs, at the time a small company ({approx}12 employees) which had successfully developed a number of products for larger oil well service companies. In September, 2006, approximately 20 months after inception of the project, Ultima Labs was acquired by Sondex plc, a worldwide manufacturer of downhole instrumentation for cased hole and drilling applications. The acquisition provided access to proven technology for mud pulse telemetry, downhole directional and natural gamma ray measurements, and surface data acquisition and processing, as well as a global sales and support network. The acquisition accelerated commercialization through existing Sondex customers. Customer demand resulted in changes to the product specification to support hotter (150 C) and deeper drilling (20,000 psi pressure) than originally proposed. The Sondex acquisition resulted in some project delays as the resistivity collar was interfaced to a different MWD system and also as the mechanical design was revised for the new pressure requirements. However, the Sondex acquisition has resulted in a more robust system, secure funding for completion of the project, and more rapid commercialization.

  15. Irradiation response in weldment and HIP joint of reduced activation ferritic/martensitic steel, F82H

    SciTech Connect (OSTI)

    Hirose, Takanori [Japan Atomic Energy Agency (JAEA); Sokolov, Mikhail A [ORNL; Ando, M. [Japan Atomic Energy Agency (JAEA); Tanigawa, H. [Japan Atomic Energy Agency (JAEA); Shiba, K. [Japan Atomic Energy Agency (JAEA); Stoller, Roger E [ORNL; Odette, G.R. [University of California, Santa Barbara

    2013-11-01

    This work investigates irradiation response in the joints of F82H employed for a fusion breeding blanket. The joints, which were prepared using welding and diffusion welding, were irradiated up to 6 dpa in the High Flux Isotope Reactor at the Oak Ridge National Laboratory. Post-irradiation tests revealed hardening in weldment (WM) and base metal (BM) greater than 300 MPa. However, the heat affected zones (HAZ) exhibit about half that of WM and BM. Therefore, neutron irradiation decreased the strength of the HAZ, leaving it in danger of local deformation in this region. Further the hardening in WM made with an electron beam was larger than that in WM made with tungsten inert gas welding. However the mechanical properties of the diffusion-welded joint were very similar to those of BM even after the irradiation.

  16. THERMAL PREDICTIONS OF NEW COMPOSITE MATERIAL DURING INPILE TESTING

    SciTech Connect (OSTI)

    Donna Post Guillen; W. David Swank; Heng Ban; Kurt Harris; Adam Zabriskie

    2011-09-01

    An inpile experiment is currently underway wherein specimens comprised of a newly developed material are being irradiated at Idaho National Laboratory's Advanced Test Reactor (ATR) in conjunction with Utah State University under the auspices of the ATR National Scientific User Facility. This paper provides the thermophysical properties of this new material measured prior to irradiation. After the irradiation campaign is complete, the thermophysical properties of the specimens will be measured and compared to the preirradiation values. A finite-element model was constructed to predict bounding specimen temperatures during irradiation. Results from the thermal hydraulic modeling, including the steady-state temperatures of the specimens within sealed capsules, are presented. After the irradiation campaign is completed, best-estimate thermal predictions will be performed for the individual specimens using the actual as-run irradiation power levels.

  17. Improving Thermal Model Prediction Through Statistical Analysis of Irradiation and Post-Irradiation Data from AGR Experiments

    SciTech Connect (OSTI)

    Dr. Binh T. Pham; Grant L. Hawkes; Jeffrey J. Einerson

    2012-10-01

    As part of the Research and Development program for Next Generation High Temperature Reactors (HTR), a series of irradiation tests, designated as Advanced Gas-cooled Reactor (AGR), have been defined to support development and qualification of fuel design, fabrication process, and fuel performance under normal operation and accident conditions. The AGR tests employ fuel compacts placed in a graphite cylinder shrouded by a steel capsule and instrumented with thermocouples (TC) embedded in graphite blocks enabling temperature control. The data representing the crucial test fuel conditions (e.g., temperature, neutron fast fluence, and burnup) while impossible to obtain from direct measurements are calculated by physics and thermal models. The irradiation and post-irradiation examination (PIE) experimental data are used in model calibration effort to reduce the inherent uncertainty of simulation results. This paper is focused on fuel temperature predicted by the ABAQUS code’s finite element-based thermal models. The work follows up on a previous study, in which several statistical analysis methods were adapted, implemented in the NGNP Data Management and Analysis System (NDMAS), and applied for improving qualification of AGR-1 thermocouple data. The present work exercises the idea that the abnormal trends of measured data observed from statistical analysis may be caused by either measuring instrument deterioration or physical mechanisms in capsules that may have shifted the system thermal response. As an example, the uneven reduction of the control gas gap in Capsule 5 revealed by the capsule metrology measurements in PIE helps justify the reduction in TC readings instead of TC drift. This in turn prompts modification of thermal model to better fit with experimental data, thus help increase confidence, and in other word reduce model uncertainties in thermal simulation results of the AGR-1 test.

  18. Templated Self Assemble of Nano-Structures

    SciTech Connect (OSTI)

    Suo, Zhigang [Harvard University

    2013-04-29

    This project will identify and model mechanisms that template the self-assembly of nanostructures. We focus on a class of systems involving a two-phase monolayer of molecules adsorbed on a solid surface. At a suitably elevated temperature, the molecules diffuse on the surface to reduce the combined free energy of mixing, phase boundary, elastic field, and electrostatic field. With no template, the phases may form a pattern of stripes or disks. The feature size is on the order of 1-100 nm, selected to compromise the phase boundary energy and the long-range elastic or electrostatic interaction. Both experimental observations and our theoretical simulations have shown that the pattern resembles a periodic lattice, but has abundant imperfections. To form a perfect periodic pattern, or a designed aperiodic pattern, one must introduce a template to guide the assembly. For example, a coarse-scale pattern, lithographically defined on the substrate, will guide the assembly of the nanoscale pattern. As another example, if the molecules on the substrate surface carry strong electric dipoles, a charged object, placed in the space above the monolayer, will guide the assembly of the molecular dipoles. In particular, the charged object can be a mask with a designed nanoscale topographic pattern. A serial process (e.g., e-beam lithography) is necessary to make the mask, but the pattern transfer to the molecules on the substrate is a parallel process. The technique is potentially a high throughput, low cost process to pattern a monolayer. The monolayer pattern itself may serve as a template to fabricate a functional structure. This project will model fundamental aspects of these processes, including thermodynamics and kinetics of self-assembly, templated self-assembly, and self-assembly on unconventional substrates. It is envisioned that the theory will not only explain the available experimental observations, but also motivate new experiments.

  19. Complex Assembly Variant Design in Agile Manufacturing. Part I: System Architecture and Assembly Modeling Methodology

    E-Print Network [OSTI]

    Nagi, Rakesh

    ;1 Introduction Having experienced Mass Production, Lean Manufacturing, and Follow the Leader, American manufacComplex Assembly Variant Design in Agile Manufacturing. Part I: System Architecture and Assembly 14260, USA March 31, 2004 Abstract In the distributed and horizontally integrated manufacturing

  20. Optimizing Robot Motion Strategies for Assembly with Stochastic Models of the Assembly Process

    E-Print Network [OSTI]

    LaValle, Steven M.

    , ranging from long­term production control, which deals with entire factories and time scales on the order. Production control determines what the assembly plant should be producing from, for example, month to month­motion planning for assembly is commonly considered as a distinct, isolated step between task sequencing

  1. Simplified flangeless unisex waveguide coupler assembly

    DOE Patents [OSTI]

    Michelangelo, Dimartino (La Mesa, CA); Moeller, Charles P. (Del Mar, CA)

    1993-01-01

    A unisex coupler assembly is disclosed capable of providing a leak tight coupling for waveguides with axial alignment of the waveguides and rotational capability. The sealing means of the coupler assembly are not exposed to RF energy, and the coupler assembly does not require the provision of external flanges on the waveguides. In a preferred embodiment, O ring seals are not used and the coupler assembly is, therefore, bakeable at a temperature up to about 150.degree. C. The coupler assembly comprises a split collar which clamps around the waveguides and a second collar which fastens to the split collar. The split collar contains an inner annular groove. Each of the waveguides is provided with an external annular groove which receives a retaining ring. The split collar is clamped around one of the waveguides with the inner annular groove of the split collar engaging the retaining ring carried in the external annular groove in the waveguide. The second collar is then slipped over the second waveguide behind the annular groove and retaining ring therein and the second collar is coaxially secured by fastening means to the split collar to draw the respective waveguides together by coaxial force exerted by the second collar against the retaining ring on the second waveguide. A sealing ring is placed against an external sealing surface at a reduced external diameter end formed on one waveguide to sealingly engage a corresponding sealing surface on the other waveguide as the waveguides are urged toward each other.

  2. Microstructural Characterization of Irradiated U-7Mo/Al-5Si Dispersion to High Fission Density

    SciTech Connect (OSTI)

    J. Gan; B. D. Miller; D. D. Keiser, Jr.; A. B. Robinson; J. W. Madden; P. G. Medvedev; D. M. Wachs

    2014-11-01

    The fuel development program for research and test reactors calls for improved knowledge on the effect of microstructure on fuel performance in reactors. This work summarizes the recent TEM microstructural characterization of an irradiated U-7Mo/Al-5Si dispersion fuel plate (R3R050) irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory to 5.2×1021 fissions/cm3. While a large fraction of the fuel grains is decorated with large bubbles, there is no evidence showing interlinking of these large bubbles at the specified fission density. The attachment of solid fission product precipitates to the bubbles is likely the result of fission product diffusion into these bubbles. The process of fission gas bubble superlattice collapse appears through bubble coalescence. The results are compared with the previous TEM work of the dispersion fuels irradiated to lower fission density from the same fuel plate.

  3. Surry Unit 2 end of cycle 4 onsite fuel examination of 17 x 17 demonstration assemblies after three cycles of exposure

    SciTech Connect (OSTI)

    Balfour, M.G.; Schmidt, G.R.; Muenks, M.F.; Isaac, P.G.; Eng, G.H.

    1981-02-01

    This report presents an evaluation of the results obtained from various onsite nondestructive examinations performed during August 1979, on Surry Unit 2 fuel. The fuel consisted of 17 x 17 demonstration fuel assemblies placed in the Surry Unit 2 reactor for reactor cycles 2, 3, and 4, along with 15 x 15 fuel assemblies in the remainder of the core. Two three-cycle 17 x 17 demonstration fuel assemblies, four two-cycle 17 x 17 removable fuel rods, and eight three-cycle 17 x 17 removable fuel rods were examined. The nondestructive examinations included television visual examinations and grid cell friction force measurments performed on the demonstration assemblies and breakaway/withdrawal force measurements, television visual examinations, and profilometry measurements performed on the removable fuel rods. Summaries of the design features of the 17 x 17 demonstration assemblies, the characterization program conducted during fabrication, and previous onsite nondestructive examinations performed on them are contained within the report. A brief description of the irradiation history of the 17 x 17 demonstration assemblies and removable rods is also presented.

  4. Methods for Post Irradiation Examination of Tritium Producing...

    Office of Environmental Management (EM)

    Methods for Post Irradiation Examination of Tritium Producing Burnable Absorber Rods Methods for Post Irradiation Examination of Tritium Producing Burnable Absorber Rods...

  5. Post Irradiation Examination of Stainless Steel Cladding from...

    Office of Environmental Management (EM)

    Post Irradiation Examination of Stainless Steel Cladding from In-Reactor Permeation Experiment Post Irradiation Examination of Stainless Steel Cladding from In-Reactor Permeation...

  6. AGR-1 Safety Test Predictions using the PARFUME code

    SciTech Connect (OSTI)

    Blaise Collin

    2012-05-01

    The PARFUME modeling code was used to predict failure probability of TRISO-coated fuel particles and diffusion of fission products through these particles during safety tests following the first irradiation test of the Advanced Gas Reactor program (AGR-1). These calculations support the AGR-1 Safety Testing Experiment, which is part of the PIE effort on AGR-1. Modeling of the AGR-1 Safety Test Predictions includes a 620-day irradiation followed by a 300-hour heat-up phase of selected AGR-1 compacts. Results include fuel failure probability, palladium penetration, and fractional release of fission products. Results show that no particle failure is predicted during irradiation or heat-up, and that fractional release of fission products is limited during irradiation but that it significantly increases during heat-up.

  7. Design and Testing of Prototypic Elements Containing Monolithic Fuel

    SciTech Connect (OSTI)

    N.E. Woolstenhulme; M.K. Meyer; D.M. Wachs

    2011-10-01

    The US fuel development team has performed numerous irradiation tests on small to medium sized specimens containing low enriched uranium fuel designs. The team is now focused on qualification and demonstration of the uranium-molybdenum Base Monolithic Design and has entered the next generation of testing with the design and irradiation of prototypic elements which contain this fuel. The designs of fuel elements containing monolithic fuel, such as AFIP-7 (which is currently under irradiation) and RERTR-FE (which is currently under fabrication), are appropriate progressions relative to the technology life cycle. The culmination of this testing program will occur with the design, fabrication, and irradiation of demonstration products to include the base fuel demonstration and design demonstration experiments. Future plans show that design, fabrication, and testing activities will apply the rigor needed for a demonstration campaign.

  8. Radiation Tolerance of Neutron-Irradiated Model Fe-Cr-Al Alloys

    SciTech Connect (OSTI)

    Field, Kevin G; Hu, Xunxiang; Littrell, Ken; Yamamoto, Yukinori; Snead, Lance Lewis

    2015-01-01

    The Fe Cr Al alloy system has the potential to form an important class of enhanced accident-tolerant cladding materials in the nuclear power industry owing to the alloy system's higher oxidation resistance in high-temperature steam environments compared with traditional zirconium-based alloys. However, radiation tolerance of Fe Cr Al alloys has not been fully established. In this study, a series of Fe Cr Al alloys with 10 18 wt % Cr and 2.9 4.9 wt % Al were neutron irradiated at 382 C to 1.8 dpa to investigate the irradiation-induced microstructural and mechanical property evolution as a function of alloy composition. Dislocation loops with Burgers vector of a/2 111 and a 100 were detected and quantified. Results indicate precipitation of Cr-rich is primarily dependent on the bulk chromium composition. Mechanical testing of sub-size-irradiated tensile specimens indicates the hardening response seen after irradiation is dependent on the bulk chromium composition. A structure property relationship was developed; it indicated that the change in yield strength after irradiation is caused by the formation of these radiation-induced defects and is dominated by the large number density of Cr-rich precipitates at sufficiently high chromium contents after irradiation.

  9. Gamma irradiation effect on the chemical composition and the antioxidant activity of Ipomoea batatas L

    SciTech Connect (OSTI)

    Tahir, D., E-mail: dtahir@fmipa.unhas.ac.id; Halide, H., E-mail: dtahir@fmipa.unhas.ac.id; Kurniawan, D. [Department of Physics, Hasanuddin University, Makassar 90245 (Indonesia); Wahab, A. W. [Department of Chemistry, Hasanuddin University, Makassar 90245 (Indonesia)

    2014-09-25

    The chemical composition and antioxidant activity of Ipomoea batatas L. (sweet potato) were studied by x-ray fluorescence (XRF) spectroscopy, Fourier transform infrared spectroscopy, and by the 1,1-diphenyl-2-picrylhydrazyl (DPPH) radical scavenging capacity. The irradiation treatment was performed by using Cs-137 as a gamma sources in experimental equipment. Treatment by irradiation emerges as a possible conservation technique that has been tested successfully in several food products. The amount of chemical composition was changed and resulting new chemical for absorbed dose 40 mSv. Interestingly, it was found that gamma irradiation significantly increased the antioxidant activity, as measured by DPPH radical scavenging capacity. The antioxidant activity of Ipomoea batatas L. extract was dramatically increased in the non-irradiated sample to the sample irradiated at 40 mSv. These results indicate that gamma irradiation of Ipomoea batatas L. extract can enhance its antioxidant activity through the formation of a new chemical compound. Based on these results, increased antioxidant activity of Ipomoea batatas L. extracts by gamma rays can be applied to various industries, especially cosmetics, foodstuffs, and pharmaceuticals.

  10. Low thermal resistance power module assembly

    DOE Patents [OSTI]

    Hassani, Vahab (Denver, CO); Vlahinos, Andreas (Castle Rock, CO); Bharathan, Desikan (Arvada, CO)

    2007-03-13

    A power module assembly with low thermal resistance and enhanced heat dissipation to a cooling medium. The assembly includes a heat sink or spreader plate with passageways or openings for coolant that extend through the plate from a lower surface to an upper surface. A circuit substrate is provided and positioned on the spreader plate to cover the coolant passageways. The circuit substrate includes a bonding layer configured to extend about the periphery of each of the coolant passageways and is made up of a substantially nonporous material. The bonding layer may be solder material which bonds to the upper surface of the plate to provide a continuous seal around the upper edge of each opening in the plate. The assembly includes power modules mounted on the circuit substrate on a surface opposite the bonding layer. The power modules are positioned over or proximal to the coolant passageways.

  11. Photovoltaic assemblies and methods for transporting

    DOE Patents [OSTI]

    Almy, Charles; Campbell, Matt; Sandler, Reuben; Wares, Brian; Wayman, Elizabeth

    2014-08-05

    A PV assembly including framework, PV laminate(s), and a stiffening device. The framework includes a perimeter frame at least 10 feet in length and at least 5 feet in width. The PV laminate(s) are assembled to the perimeter frame to define a receiving zone having a depth of not more than 8 inches. The stiffening device is associated with the framework and is configured to provide a first state and a second state. In the first state, an entirety of the stiffening device is maintained within the receiving zone. In the second state, at least a portion of the stiffening device projects from the receiving zone. The stiffening device enhances a stiffness of the PV assembly in a plane of the perimeter frame, and can include rods defining truss structures.

  12. Head assembly for multiposition borehole extensometer

    DOE Patents [OSTI]

    Frank, Donald N. (Livermore, CA)

    1983-01-01

    A head assembly for a borehole extensometer and an improved extensometer for measuring subsurface subsidence. A plurality of inflatable anchors provide discrete measurement points. A metering rod is fixed to each of the anchors which are displaced when subsidence occurs, thereby translating the attached rod. The head assembly includes a sprocket wheel rotatably mounted on a standpipe and engaged by a chain which is connected at one end to the metering rod and at the other end to a counterweight. A second sprocket wheel connected to the standpipe also engages the chain and drives a connected potentiometer. The head assembly converts the linear displacement of the metering rod to the rotary motion of the second sprocket wheel, which is measured by the potentiometer, producing a continuous electrical output.

  13. Photovoltaic assemblies and methods for transporting

    DOE Patents [OSTI]

    Almy, Charles; Campbell, Matt; Sandler, Reuben; Wares, Brian; Wayman, Elizabeth

    2013-09-17

    A PV assembly including framework, PV laminate(s), and a stiffening device. The framework includes a perimeter frame at least 10 feet in length and at least 5 feet in width. The PV laminate(s) are assembled to the perimeter frame to define a receiving zone having a depth of not more than 8 inches. The stiffening device is associated with the framework and is configured to provide a first state and a second state. In the first state, an entirety of the stiffening device is maintained within the receiving zone. In the second state, at least a portion of the stiffening device projects from the receiving zone. The stiffening device enhances a stiffness of the PV assembly in a plane of the perimeter frame, and can include rods defining truss structures.

  14. Microcomponent assembly for efficient contacting of fluid

    DOE Patents [OSTI]

    Drost, Monte K. (Richland, WA); Wegeng, Robert S. (Richland, WA); Friedrich, Michele (Prosser, WA); Hanna, William T. (Gahanna, OH); Call, Charles J. (Pasco, WA); Kurath, Dean E. (Richland, WA)

    2000-01-01

    The present invention is a fundamental method and apparatus of a microcomponent assembly that overcomes the inherent limitations of state of the art chemical separations. The fundamental element enabling miniaturization is the porous contactor contained within a microcomponent assembly for mass transfer of a working compound from a first medium to a second medium. The porous contactor has a thickness, and a plurality of pores extending through the thickness. The pores are of a geometry cooperating with a boundary tension of one or the other or both of the media thereby preventing migration of one, other or both through the microporous contactor while permitting passage of the working compound. In the microcomponent assembly, the porous contactor is placed between a first laminate such that a first space or first microplenum is formed between the microporous contactor and the first laminate. Additionally, a cover sheet provides a second space or second plenum between the porous contactor and the cover sheet.

  15. Self-assembly of nanocomposite materials

    DOE Patents [OSTI]

    Brinker, C. Jeffrey (Albuquerque, NM); Sellinger, Alan (Palo Alto, CA); Lu, Yunfeng (New Orleans, LA)

    2001-01-01

    A method of making a nanocomposite self-assembly is provided where at least one hydrophilic compound, at least one hydrophobic compound, and at least one amphiphilic surfactant are mixed in an aqueous solvent with the solvent subsequently evaporated to form a self-assembled liquid crystalline mesophase material. Upon polymerization of the hydrophilic and hydrophobic compounds, a robust nanocomposite self-assembled material is formed. Importantly, in the reaction mixture, the amphiphilic surfactant has an initial concentration below the critical micelle concentration to allow formation of the liquid-phase micellar mesophase material. A variety of nanocomposite structures can be formed, depending upon the solvent evaporazation process, including layered mesophases, tubular mesophases, and a hierarchical composite coating composed of an isotropic worm-like micellar overlayer bonded to an oriented, nanolaminated underlayer.

  16. Advanced membrane electrode assemblies for fuel cells

    DOE Patents [OSTI]

    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.

  17. Advanced membrane electrode assemblies for fuel cells

    DOE Patents [OSTI]

    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.

  18. PV module mounting method and mounting assembly

    DOE Patents [OSTI]

    Lenox, Carl J.S.; Johnson, Kurt M.

    2013-04-23

    A method for mounting PV modules to a deck includes selecting PV module layout pattern so that adjacent PV module edges are spaced apart. PV mounting and support assemblies are secured to the deck according to the layout pattern using fasteners extending into the deck. The PV modules are placed on the PV mounting and support assemblies. Retaining elements are located over and secured against the upper peripheral edge surfaces of the PV modules so to secure them to the deck with the peripheral edges of the PV modules spaced apart from the deck. In some examples a PV module mounting assembly, for use on a shingled deck, comprises flashing, a base mountable on the flashing, a deck-penetrating fastener engageable with the base and securable to the deck so to secure the flashing and the base to the shingled deck, and PV module mounting hardware securable to the base.

  19. Shipping device for heater unit assembly

    DOE Patents [OSTI]

    Blaushild, Ronald M. (Export, PA); Abbott, Stephan L. (Monroeville, PA); Miller, Phillip E. (Greensburg, PA); Shaffer, Robert (Swissvale, PA)

    1991-01-01

    A shipping device for a heater unit assembly (23), the heater unit assembly (23) including a cylindrical wall (25) and a top plate (31) secured to the cylindrical wall (25) and having a flange portion which projects radially beyond the outer surface of the cylindrical wall (25), and the shipping device including: a cylindrical container (3) having a closed bottom (13); a support member (47) secured to the container (3) and having an inwardly directed flange for supporting the flange portion of the top plate (31); a supplemental supporting system (1) for positioning the heater unit assembly (23) in the container (3) at a spaced relation from the inner surface and bottom wall (13) of the container (3); a cover (15) for closing the top of the container (3); and a container supporting structure (5,7,8) supporting the container (3) in a manner to permit the container (3) to be moved, relative to the supporting structure (5,7,8 ), between a vertical position for loading and unloading the assembly (23) and a horizontal position for transport of the assembly (23). A seal (57) is interposed between the container (3) and the cover (15) for sealing the interior of the container (3) from the environment. An abutment member (41) is mounted on the container supporting structure (5,7,8) for supporting the container bottom (13), when the container (3) is in the vertical position, to prevent the container (3) from moving past the vertical position in the direction away from the horizontal position, and a retainer member (55) is secured within the cover (15) for retaining the assembly top plate (31) in contact with the support member (47) when the cover (15) closes the top of the container (3).

  20. Mutations that affect structure and assembly of light-harvesting proteins in the cyanobacterium Synechocystis sp. strain 6701

    SciTech Connect (OSTI)

    Anderson, L.K.; Rayner, M.C.; Eiserling, F.A.

    1987-01-01

    The unicellular cyanobacterium Synechocystis sp. strain 6701 was mutagenized with UV irradiation and screened for pigment changes that indicated genetic lesions involving the light-harvesting proteins of the phycobilisome. A previous examination of the pigment mutant UV16 showed an assembly defect in the phycocyanin component of the phycobilisome. Mutagenesis of UV16 produced an additional double mutant, UV16-40, with decreased phycoerythrin content. Phycocyanin and phycoerythrin were isolated from UV16-40 and compared with normal biliproteins. The results suggested that the UV16 mutation affected the alpha subunit of phycocyanin, while the phycoerythrin beta subunit from UV16-40 had lost one of its three chromophores. Characterization of the unassembled phycobilisome components in these mutants suggests that these strains will be useful for probing in vivo the regulated expression and assembly of phycobilisomes.