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1

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

SciTech Connect (OSTI)

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.

NONE

1997-07-01T23:59:59.000Z

2

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

Broader source: Energy.gov [DOE]

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

Neutronics, steady-state, and transient analyses for the Poland MARIA reactor for irradiation testing of LEU lead test fuel assemblies from CERCA : ANL independent verification results.  

SciTech Connect (OSTI)

The MARIA reactor at the Institute of Atomic Energy (IAE) in Swierk (30 km SE of Warsaw) in the Republic of Poland is considering conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel assemblies (FA). The FA design in MARIA is rather unique; a suitable LEU FA has never been designed or tested. IAE has contracted with CERCA (the fuel supply portion of AREVA in France) to supply 2 lead test assemblies (LTA). The LTAs will be irradiated in MARIA to burnup level of at least 40% for both LTAs and to 60% for one LTA. IAE may decide to purchase additional LEU FAs for a full core conversion after the test irradiation. The Reactor Safety Committee within IAE and the National Atomic Energy Agency in Poland (PAA) must approve the LTA irradiation process. The approval will be based, in part, on IAE submitting revisions to portions of the Safety Analysis Report (SAR) which are affected by the insertion of the LTAs. (A similar process will be required for the full core conversion to LEU fuel.) The analysis required was established during working meetings between Argonne National Laboratory (ANL) and IAE staff during August 2006, subsequent email correspondence, and subsequent staff visits. The analysis needs to consider the current high-enriched uranium (HEU) core and 4 core configurations containing 1 and 2 LEU LTAs in various core positions. Calculations have been performed at ANL in support of the LTA irradiation. These calculations are summarized in this report and include criticality, burn-up, neutronics parameters, steady-state thermal hydraulics, and postulated transients. These calculations have been performed at the request of the IAE staff, who are performing similar calculations to be used in their SAR amendment submittal to the PAA. The ANL analysis has been performed independently from that being performed by IAE and should only be used as one step in the verification process.

Garner, P. L.; Hanan, N. A. (Nuclear Engineering Division)

2011-06-07T23:59:59.000Z

4

Final Assembly and Initial Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor  

SciTech Connect (OSTI)

The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) 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 new 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 irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new 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.1,2 The experiments, which will each consist of 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 final design phase for the first experiment was completed in 2005, and the fabrication and assembly of the first experiment test train (designated AGR-1) as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation were completed in 2006. The experiment was inserted in the ATR in December 2006, and will serve as a shakedown test of the multi-capsule experiment design that will be used in the subsequent irradiations as well as a test of the early variants of the fuel produced under this program. The experiment test train as well as the monitoring, control, and data collection systems are discussed.

S. B. Grover

2007-05-01T23:59:59.000Z

5

FUEL ASSEMBLY SHAKER TEST SIMULATION  

SciTech Connect (OSTI)

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.

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

2013-05-30T23:59:59.000Z

6

Low temperature irradiation tests on  

E-Print Network [OSTI]

Sample cool down by He gas loop 10K ­ 20K Fast neutron flux Measured by Ni activation in 2010 1.4xK #12;reactor Cryogenics #12;Al-Cu-Mg He gas temperature near sample 12K Resistance changesLow temperature irradiation tests on stabilizer materials using reactor neutrons at KUR Makoto

McDonald, Kirk

7

TEST RESULTS FROM GAMMA IRRADIATION OF ALUMINUM OXYHYDROXIDES  

SciTech Connect (OSTI)

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.

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

2012-02-01T23:59:59.000Z

8

MITG test assembly design and fabrication  

SciTech Connect (OSTI)

The design, analysis, and evaluation of the Modular Isotopic Thermoelectric Generator (MITG), described in an earlier paper, led to a program to build and test prototypical, modules of that generator. Each test module duplicates the thermoelectric converters, thermal insulation, housing and radiator fins of a typical generator slice, and simulates its isotope heat source module by means of an electrical heater encased in a prototypical graphite box. Once the approx. 20-watt MITG module has been developed, it can be assembled in appropriate number to form a generator design yielding the desired power output. The present paper describes the design and fabrication of the MITG test assembly, which confirmed the fabricability of the multicouples and interleaved multifoil insulation called for by the design. Test plans, procedures, instrumentation, results, and post-test analyses, as well as revised designs, fabrication procedures, and performance estimates, are described in subsequent papers in these proceedings.

Schock, A.

1983-01-01T23:59:59.000Z

9

AGR-1 Irradiation Experiment Test Plan  

SciTech Connect (OSTI)

This document presents the current state of planning for the AGR-1 irradiation experiment, the first of eight planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. 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 will be irradiated in the B-10 position of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). The test will contain six independently controlled and monitored capsules. Each capsule will contain a single type, or variant, of the AGR coated fuel. The irradiation is planned for about 700 effective full power days (approximately 2.4 calendar years) with a time-averaged, volume-average temperature of approximately 1050 °C. Average fuel burnup, for the entire test, will be greater than 17.7 % FIMA, and the fuel will experience fast neutron fluences between 2.4 and 4.5 x 1025 n/m2 (E>0.18 MeV).

John T. Maki

2009-10-01T23:59:59.000Z

10

FUEL ASSEMBLY SHAKER AND TRUCK TEST SIMULATION  

SciTech Connect (OSTI)

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.

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

2014-09-25T23:59:59.000Z

11

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

SciTech Connect (OSTI)

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.

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

2012-10-01T23:59:59.000Z

12

AGC-1 Irradiation Experiment Test Plan  

SciTech Connect (OSTI)

The Advanced Graphite Capsule (AGC) irradiation test program supports the acquisition of irradiated graphite performance data to assist in the selection of the technology to be used for the VHTR. Six irradiations are planned to investigate compressive creep in graphite subjected to a neutron field and obtain irradiated mechanical properties of vibrationally molded, extruded, and iso-molded graphites for comparison. The experiments will be conducted at three temperatures: 600, 900, and 1200°C. At each temperature, two different capsules will be irradiated to different fluence levels, the first from 0.5 to 4 dpa and the second from 4 to 7 dpa. AGC-1 is the first of the six capsules designed for ATR and will focus on the prismatic fluence range.

R. L. Bratton

2006-05-01T23:59:59.000Z

13

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

SciTech Connect (OSTI)

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.

Kevin G. Field; Richard Howard; Michael Teague

2014-06-01T23:59:59.000Z

14

TEMPERATURE DEPENDANT BEHAVIOUR OBSERVED IN THE AFIP-6 IRRADIATION TEST  

SciTech Connect (OSTI)

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 contained 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, the reduced flow resulted in a sufficiently low heat transfer coefficient that failure of the fuel plates occurred. The increased fuel temperature led to significant variations in the fission gas retention behaviour of the U-Mo fuel. These variations in performance are outlined herein.

A. B. Robinson; D. M. Wachs; P. Medvedev; S.J. Miller; F. J. Rice; M. K. Meyer; D. M. Perez

2012-03-01T23:59:59.000Z

15

LWRS ATR Irradiation Testing Readiness Status  

SciTech Connect (OSTI)

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

Kristine Barrett

2012-09-01T23:59:59.000Z

16

Instrumentation to Enhance Advanced Test Reactor Irradiations  

SciTech Connect (OSTI)

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.

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

2009-09-01T23:59:59.000Z

17

Irradiation data for the MFA-1 and MFA-2 tests in the FFTF  

SciTech Connect (OSTI)

This report provides key information on the irradiation environment of the MONJU fuel tests MFA-1 and MFA-2 in the Fast Flux Test Facility (FFTF). This information includes the fission powers, neutron fluxes, sodium temperatures and sodium flow rates in MFA-I, MFA-2 and adjacent assemblies. It also includes MFA-1 and MFA-2 compositions as a function of exposure. The work was performed at the request of Power Reactor and Nuclear Fuels Corporation (PNC) of Japan.

Nelson, J.V.

1997-04-24T23:59:59.000Z

18

Furnace assembly  

DOE Patents [OSTI]

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.

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

1985-01-01T23:59:59.000Z

19

Fail-safe storage rack for irradiated fuel rod assemblies  

DOE Patents [OSTI]

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.

Lewis, Donald R. (Pocatello, ID)

1993-01-01T23:59:59.000Z

20

ALSEP CASK ASSEMBLY GEARBOX THERMAL VACUUM TEST  

E-Print Network [OSTI]

Systems Division DATE 1. 0 Introduction As a result of Chit #S-3 which was generated at the ALSEP Cask to the gearbox assembly by means of the gearbox ball chain. This chain was placed over a sprocket which was of the same diameter of the gearbox sprocket wheel and was coupled to the magnetic feedthrough by means

Rathbun, Julie A.

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


21

Test specification for decant pump and winch assembly. Revision 2  

SciTech Connect (OSTI)

This specification provides the requirements for testing of the vertical turbine decant pump including the floating suction with load sensing winch control, instrumentation and the associated PLC/PC control system. All assembly necessary for testing including piping, temporary wiring, etc., shall be performed by the Seller. All referenced figures are at the back of this document. The testing consists of performance testing, winch testing and calibration, instrumentation verification testing and run-in testing of the pump. Testing shall be done in the presence and under the direction of the Buyer in accordance with this procedure.

Staehr, T.W.

1995-02-22T23:59:59.000Z

22

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

SciTech Connect (OSTI)

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.

Not Available

1980-05-01T23:59:59.000Z

23

Design and Status of RERTR Irradiation Tests in the Advanced Test Reactor  

SciTech Connect (OSTI)

Irradiation testing of U-Mo based fuels is the central component of the Reduced Enrichment for Research and Test Reactors (RERTR) program fuel qualification plan. Several RERTR tests have recently been completed or are planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory in Idaho Falls, ID. Four mini-plate experiments in various stages of completion are described in detail, including the irradiation test design, objectives, and irradiation conditions. Observations made during and after the in-reactor RERTR-7A experiment breach are summarized. The irradiation experiment design and planned irradiation conditions for full-size plate test are described. Progress toward element testing will be reviewed.

Daniel M. Wachs; Richard G. Ambrosek; Gray Chang; Mitchell K. Meyer

2006-10-01T23:59:59.000Z

24

Construction, assembly and tests of the ATLAS electromagnetic barrel calorimeter  

E-Print Network [OSTI]

The construction and assembly of the two half barrels of the ATLAS central electromagnetic calorimeter and their insertion into the barrel cryostat are described. The results of the qualification tests of the calorimeter before installation in the LHC ATLAS pit are given.

Aubert, B; Colas, Jacques; Delebecque, P; Di Ciaccio, L; El-Kacimi, M; Ghez, P; Girard, C; Gouanère, M; Goujdami, D; Jérémie, A; Jézéquel, S; Lafaye, R; Massol, N; Perrodo, P; Przysiezniak, H; Sauvage, G; Thion, J; Wingerter-Seez, I; Zitoun, R; Zolnierowski, Y; Alforque, R; Chen, H; Farrell, J; Gordon, H; Grandinetti, R; Hackenburg, R W; Hoffmann, A; Kierstead, J A; Köhler, J; Lanni, F; Lissauer, D; Ma, H; Makowiecki, D S; Müller, T; Norton, S; Radeka, V; Rahm, David Charles; Rehak, M; Rajagopalan, S; Rescia, S; Sexton, K; Sondericker, J; Stumer, I; Takai, H; Belymam, A; Benchekroun, D; Driouichi, C; Hoummada, A; Hakimi, M; Knee, Michael; Stroynowski, R; Wakeland, B; Datskov, V I; Drobin, V; Aleksa, Martin; Bremer, J; Carli, T; Chalifour, M; Chevalley, J L; Djama, F; Ema, L; Fabre, C; Fassnacht, P; Gianotti, F; Gonidec, A; Hansen, J B; Hervás, L; Hott, T; Lacaste, C; Marin, C P; Pailler, P; Pleskatch, A; Sauvagey, D; Vandoni, Giovanna; Vuillemin, V; Wilkens, H; Albrand, S; Belhorma, B; Collot, J; de Saintignon, P; Dzahini, D; Ferrari, A; Fulachier, J; Gallin-Martel, M L; Hostachy, J Y; Laborie, G; Ledroit-Guillon, F; Martin, P; Muraz, J F; Ohlsson-Malek, F; Saboumazrag, S; Viret, S; Othegraven, R; Zeitnitz, C; Banfi, D; Carminati, L; Cavalli, D; Citterio, M; Costa, G; Delmastro, M; Fanti, M; Mandelli, L; Mazzanti, M; Tartarelli, F; Augé, E; Baffioni, S; Bonis, J; Bonivento, W; Bourdarios, C; de La Taille, C; Fayard, L; Fournier, D; Guilhem, G; Imbert, P; Iconomidou-Fayard, L; Le Meur, G; Mencik, M; Noppe, J M; Parrour, G; Puzo, P; Rousseau, D; Schaffer, A C; Seguin-Moreau, N; Serin, L; Unal, G; Veillet, J J; Wicek, F; Zerwas, D; Astesan, F; Bertoli, W; Canton, B; Fleuret, F; Imbault, D; Lacour, D; Laforge, B; Schwemling, P; Abouelouafa, M; Ben-Mansour, A; Cherkaoui, R; El-Mouahhidi, Y; Ghazlane, H; Idrissi, A; Bazizi, K; England, D; Glebov, V; Haelen, T; Lobkowicz, F; Slattery, P F; Belorgey, J; Besson, N; Boonekamp, M; Durand, D; Ernwein, J; Mansoulié, B; Molinie, F; Meyer, J P; Perrin, P; Schwindling, J; Taguet, J P; Zaccone, Henri; Lund-Jensen, B; Rydström, S; Tayalati, Y; Botchev, B; Finocchiaro, G; Hoffman, J; McCarthy, R L; Rijssenbeek, M; Steffens, J; Zdrazil, M; Braun, H M

2006-01-01T23:59:59.000Z

25

Use of laser extensometer for mechanical test on irradiated materials  

SciTech Connect (OSTI)

Techniques have been developed by EDF`s hot laboratory in Chinon for performing mechanical tests on irradiated materials. Some of these techniques aim to facilitate strain measurements, which are particularly difficult to perform on irradiated specimens at high temperatures or on subsize specimens. Recent progress has been driven by laser technology combined with software development. The use of this technique, which allows strain measurements without contact on the specimen, is described for tensile (especially on subsize specimens), fatigue and creep tests.

Brillaud, C.; Meylogan, T.; Salathe, P. [Electricite de France, Avoine (France)

1996-12-31T23:59:59.000Z

26

Reliability Testing the Die-Attach of CPV Cell Assemblies  

SciTech Connect (OSTI)

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.

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

2011-02-01T23:59:59.000Z

27

USE OF SILICON CARBIDE MONITORS IN ATR IRRADIATION TESTING  

SciTech Connect (OSTI)

In April 2007, the Department of Energy (DOE) designated the Advanced Test Reactor (ATR) a National Scientific User Facility (NSUF) to advance US leadership in nuclear science and technology. By attracting new users from universities, laboratories, and industry, the ATR will support basic and applied nuclear research and development and help address the nation's energy security needs. In support of this new program, the Idaho National Laboratory (INL) has developed in-house capabilities to fabricate, test, and qualify new and enhanced temperature sensors for irradiation testing. Although most efforts emphasize sensors capable of providing real-time data, selected tasks have been completed to enhance sensors provided in irradiation locations where instrumentation leads cannot be included, such as drop-in capsule and Hydraulic Shuttle Irradiation System (HSIS) or 'rabbit' locations. For example, silicon carbide (SiC) monitors are now available to detect peak irradiation temperatures between 200°C and 800°C. Using a resistance measurement approach, specialized equipment installed at INL's High Temperature Test Laboratory (HTTL) and specialized procedures were developed to ensure that accurate peak irradiation temperature measurements are inferred from SiC monitors irradiated at the ATR. Comparison examinations were completed by INL to demonstrate this capability, and several programs currently rely on SiC monitors for peak temperature detection. This paper discusses the use of SiC monitors at the ATR, the process used to evaluate them at the HTTL, and presents representative measurements taken using SiC monitors.

K. L. Davis; B. Chase; T. Unruh; D. Knudson; J. L. Rempe

2012-07-01T23:59:59.000Z

28

Decant pump assembly and controls qualification testing - test report  

SciTech Connect (OSTI)

This report summarizes the results of the qualification testing of the supernate decant pump and controls system to be used for in-tank sludge washing in aging waste tank AZ-101. The test was successful and all components are qualified for installation and use in the tank.

Staehr, T.W., Westinghouse Hanford

1996-05-02T23:59:59.000Z

29

Prototype Spallation Neutron Source Rotating Target Assembly Final Test Report  

SciTech Connect (OSTI)

A full-scale prototype of an extended vertical shaft, rotating target assembly based on a conceptual target design for a 1 to 3-MW spallation facility was built and tested. Key elements of the drive/coupling assembly implemented in the prototype include high integrity dynamic face seals, commercially available bearings, realistic manufacturing tolerances, effective monitoring and controls, and fail-safe shutdown features. A representative target disk suspended on a 3.5 meter prototypical shaft was coupled with the drive to complete the mechanical tests. Successful operation for 5400 hours confirmed the overall mechanical feasibility of the extended vertical shaft rotating target concept. The prototype system showed no indications of performance deterioration and the equipment did not require maintenance or relubrication.

McManamy, Thomas J [ORNL; Graves, Van [Oak Ridge National Laboratory (ORNL); Garmendia, Amaia Zarraoa [IDOM Bilbao; Sorda, Fernando [ESS Bilbao; Etxeita, Borja [IDOM Bilbao; Rennich, Mark J [ORNL

2011-01-01T23:59:59.000Z

30

Environmental Assessment LEAD TEST ASSEMBLY IRRADIATION AND ANALYSIS  

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

Administrative Code Watts Bar Nuclear Plant State of Washington Department of Fish and Wildlife Environmental Assessment ii July 1997 U . S . Department of Energy Metric...

31

AGR-2 IRRADIATION TEST FINAL AS-RUN REPORT  

SciTech Connect (OSTI)

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.

Blaise, Collin

2014-07-01T23:59:59.000Z

32

AGR-1 Irradiation Test Final As-Run Report  

SciTech Connect (OSTI)

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.

Blaise P. Collin

2012-06-01T23:59:59.000Z

33

Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor  

SciTech Connect (OSTI)

The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate 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 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. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The design of the first experiment (designated AGR-1) was completed in 2005, and the fabrication and assembly of the test train as well as the support systems and fission product monitoring system that monitor and control the experiment during irradiation were completed in September 2006. The experiment was inserted in the ATR in December 2006, and is serving as a shakedown test of the multi-capsule experiment design that will be used in the subsequent irradiations as well as a test of the early variants of the fuel produced under this program. The experiment test train as well as the monitoring, control, and data collection systems are discussed and the status of the experiment is provided.

S. Blaine Grover; David A. Petti

2008-10-01T23:59:59.000Z

34

Design of the Next Generation Nuclear Plant Graphite Creep Experiments for Irradiation in the Advanced Test Reactor  

SciTech Connect (OSTI)

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 new 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 at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain seven separate stacks of graphite specimens. Six of the specimen stacks will have half of their graphite specimens 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 be organized into pairs with a different compressive load being applied to the top half of each pair of specimen stacks. The seventh stack will not have a compressive load on the graphite specimens during irradiation. 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 final design phase for the first experiment was completed in September 2008, and the fabrication and assembly of the experiment test train as well as installation and testing of the control and support systems that will monitor and control the experiment during irradiation are being completed in early calendar 2009. The first experiment is scheduled to be ready for insertion in the ATR by April 30, 2009. This paper will discuss the design of the experiment including the test train and the temperature and compressive load monitoring, control, and data collection systems.

S. Blaine Grover

2009-05-01T23:59:59.000Z

35

CERCA LEU fuel assemblies testing in Maria Reactor - safety analysis summary and testing program scope.  

SciTech Connect (OSTI)

The presented paper contains neutronic and thermal-hydraulic (for steady and unsteady states) calculation results prepared to support annex to Safety Analysis Report for MARIA reactor in order to obtain approval for program of testing low-enriched uranium (LEU) lead test fuel assemblies (LTFA) manufactured by CERCA. This includes presentation of the limits and operational constraints to be in effect during the fuel testing investigations. Also, the scope of testing program (which began in August 2009), including additional measurements and monitoring procedures, is described.

Pytel, K.; Mieleszczenko, W.; Lechniak, J.; Moldysz, A.; Andrzejewski, K.; Kulikowska, T.; Marcinkowska, A.; Garner, P. L.; Hanan, N. A.; Nuclear Engineering Division; Institute of Atomic Energy (Poland)

2010-03-01T23:59:59.000Z

36

Updated FY12 Ceramic Fuels Irradiation Test Plan  

SciTech Connect (OSTI)

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.

Nelson, Andrew T. [Los Alamos National Laboratory

2012-05-24T23:59:59.000Z

37

In situ investigation of formation of self-assembled nanodomain structure in lithium niobate after pulse laser irradiation  

SciTech Connect (OSTI)

The evolution of the self-assembled quasi-regular micro- and nanodomain structures after pulse infrared laser irradiation was studied by in situ optical observation. The average periods of the structures are much less than the sizes of the laser spots. The polarization reversal occurs through covering of the whole irradiated area by the nets of the spatially separated nanodomain chains and microdomain rays--''hatching effect.'' The main stages of the anisotropic nanodomain kinetics: nucleation, growth, and branching, have been singled out. The observed abnormal domain kinetics was attributed to the action of the pyroelectric field arising during cooling after laser heating.

Shur, V. Ya.; Kuznetsov, D. K.; Mingaliev, E. A.; Yakunina, E. M.; Lobov, A. I.; Ievlev, A. V. [Ferroelectric Laboratory, Institute of Physics and Applied Mathematics, Ural State University, Lenin Ave. 51, Ekaterinburg 620083 (Russian Federation)

2011-08-22T23:59:59.000Z

38

System support software for TSTA (Tritium Systems Test Assembly)  

SciTech Connect (OSTI)

The fact that Tritium Systems Test Assembly (TSTA) is an experimental facility makes it impossible and undesirable to try to forecast the exact software requirements. Thus the software had to be written in a manner that would allow modifications without compromising the safety requirements imposed by the handling of tritium. This suggested a multi-level approach to the software. In this approach (much like the ISO network model) each level is isolated from the level below and above by cleanly defined interfaces. For example, the subsystem support level interfaces with the subsystem hardware through the software support level. Routines in the software support level provide operations like ''OPEN VALVE'' and CLOSE VALVE'' to the subsystem level. This isolates the subsystem level from the actual hardware. This is advantageous because changes can occur in any level without the need for propagating the change to any other level. The TSTA control system consists of the hardware level, the data conversion level, the operator interface level, and the subsystem process level. These levels are described.

Claborn, G.W.; Mann, L.W.; Nielson, C.W.

1987-10-01T23:59:59.000Z

39

Improving the AGR Fuel Testing Power Density Profile Versus Irradiation-Time in the Advanced Test Reactor  

SciTech Connect (OSTI)

The Very High Temperature gas-cooled Reactor (VHTR), which is currently being developed, achieves simplification of safety through reliance on ceramic-coated fuel particles. Each TRISO-coated fuel particle has its own containment which serves as the principal barrier against radionuclide release under normal operating and accident conditions. These fuel particles, in the form of graphite fuel compacts, are currently undergoing a series of irradiation tests in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) to support the Advanced Gas-Cooled Reactor (AGR) fuel qualification program. A representive coated fuel particle with an 235U enrichment of 19.8 wt% was used in this analysis. The fuel burnup analysis tool used to perform the neutronics study reported herein, couples the Monte Carlo transport code MCNP, with the radioactive decay and burnup code ORIGEN2. The fuel burnup methodology known as Monte-Carlo with ORIGEN2 (MCWO) was used to evaluate the AGR experiment assembly and demonstrate compliance with ATR safety requirements. For the AGR graphite fuel compacts, the MCWO-calculated fission power density (FPD) due to neutron fission in 235U is an important design parameter. One of the more important AGR fuel testing requirements is to maintain the peak fuel compact temperature close to 1250°C throughout the proposed irradiation campaign of 550 effective full power days (EFPDs). Based on the MCWO-calculated FPD, a fixed gas gap size was designed to allow regulation of the fuel compact temperatures throughout the entire fuel irradiation campaign by filling the gap with a mixture of helium and neon gases. The chosen fixed gas gap can only regulate the peak fuel compact temperature in the desired range during the irradiation test if the ratio of the peak power density to the time-dependent low power density (P/T) at 550 EFPDs is less than 2.5. However, given the near constant neutron flux within the ATR driver core and the depletion of 235U in the graphite fuel compacts versus EFPD, the P/T ratio was calculated to be 5.3, which is unacceptable given the fuel compact temperature control requirement. To flatten the FPD profile versus EFPDs, two proposed options are – (a) add fertile (232Th) particles to the fuel compact and (b) add burnable absorber (B4C) to the graphite holder. The effectiveness of these two proposed options to flatten the FPD profile versus EFPDs were investigated and the results are compared in this study.

Gray S. Chang; David A. Petti; John T. Maki; Misti A. Lillo

2009-05-01T23:59:59.000Z

40

Irradiation testing of a niobium-molybdenum developmental thermocouple  

SciTech Connect (OSTI)

A need exists for a radiation-resistant thermocouple capable of monitoring temperatures in excess of the limits of the chromel/alumel system. Tungsten/rhenium and platinum/rhodium thermocouples have sufficient temperature capability but have proven to be unstable because of irradiation-induced decalibration. The niobium/molybdenum system is believed to hold great potential for nuclear applications at temperatures up to 2000 K. However, the fragility of pure niobium and fabrication problems with niobium/molybdenum alloys have limited development of this system. Utilizing the Fast Flux Test Facility, a developmental thermocouple with a thermoelement pair consisting of a pure molybdenum and a niobium-1%zirconium alloy wire was irradiated fro 7200 hours at a temperature of 1070 K. The thermocouple performed flawlessly for the duration of the experiment and exhibited stability comparable to a companion chromel/alumel unit. A second thermocouple, operating at 1375 K, is currently being employed to monitor a fusion materials experiment in the Fast Flux Test Facility. This experiment, also scheduled for 7200 hours, will serve to further evaluate the potential of the niobium-1%zirconium/molybdenum thermoelement system. 7 refs., 7 figs.

Knight, R.C.; Greenslade, D.L.

1991-10-01T23:59:59.000Z

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


41

Results of irradiated cladding tests and clad plate experiments  

SciTech Connect (OSTI)

Two aspects critical to the fracture behavior of three-wire stainless steel cladding were investigated by the Heavy-Section Steel Technology (HSST) Program: (1) radiation effects on cladding strength and toughness, and (2) the response of mechanically loaded, flawed structures in the presence of cladding (clad plate experiments). Postirradiation testing results show that, in the test temperature range from /minus/125 to 288/degree/C, the yield strength increased, and ductility insignificantly increased, while there was almost no change in ultimate tensile strength. All cladding exhibited ductile-to-brittle transition behavior during Charpy impact testing. Radiation damage decreased the Charpy upper-shelf energy by 15 to 20% and resulted in up to 28/degree/C shifts of the Charpy impact transition temperature. Results of irradiated 12.5-mm-thick compact specimens (0.5TCS) show consistent decreases in the ductile fracture toughness, J/sub Ic/, and the tearing modulus. Results from clad plate tests have shown that (1) a tough surface layer composed of cladding and/or heat-affected zone has arrested running flaws under conditions where unclad plates have ruptured, and (2) the residual load-bearing capacity of clad plates with large subclad flaws significantly exceeded that of an unclad plate. 13 figs., 1 tab.

Haggag, F.M.; Iskander, S.K.

1988-01-01T23:59:59.000Z

42

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

SciTech Connect (OSTI)

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)

Yang, Yong Sik; Bang, Je Geon; Kim, Sun Ki; Song, Kun Woo [LWR Fuel Development Division, Korea Atomic Energy Research Institute, 1045, Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of); Park, Su Ki [HANARO Utilization Technology Development Division, Korea Atomic Energy Research Institute, 1045, Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of); Seo, Chul Gyo [HANARO Management Division, Korea Atomic Energy Research Institute, 1045, Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

2007-07-01T23:59:59.000Z

43

Irradiation Testing of Blanket Materials at the HFR Petten with On Line Tritium Monitoring  

SciTech Connect (OSTI)

Irradiation experiments are performed in support of fusion blanket technology development. These comprise ceramic solid breeder materials, and a liquid Lithium Lead alloy, as well as blanket subassemblies and components. Experimental facilities at the HFR to study tritium release, permeation characteristics, and neutron irradiation performance, have recently been extended. This paper gives an overview on the tritium breeding materials irradiation programme and describes the facilities required for irradiation testing and on-line tritium measurement.

Magielsen, A.J.; Laan, J.G. van der; Hegeman, J.B.J.; Stijkel, M.P.; Ooijevaar, M.A.G

2005-07-15T23:59:59.000Z

44

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.

45

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, MD 1. Objective This note describes neutron irradiation tests of an S-LINK [1] source card (LSC/1024 transmitter/receiver chips operated in 20-bit, single frame mode with a 40 MHz clock. The optical transmitter

46

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

SciTech Connect (OSTI)

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.

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

2003-11-01T23:59:59.000Z

47

Neutron Irradiation Tests of Pressure Transducers in Liquid Helium  

E-Print Network [OSTI]

The superconducting magnets of the future Large Hadron Collider (LHC) at CERN will operate in pressurised superfluid helium (1 bar, 1.9 K). About 500 pressure transducers will be placed in the liquid helium bath for monitoring the filling and the pressure transients after resistive transitions. Their precision must remain better than 100 mbar at pressures below 2 bar and better than 5% for higher pressures (up to 20 bar), with temperatures ranging from 1.8 K to 300 K. All the tested transducers are based on the same principle: the fluid or gas is separated from a sealed reference vacuum by an elastic membrane; its deformation indicates the pressure. The transducers will be exposed to high neutron fluence (2 kGy, 1014 n/cm2 per year) during the 20 years of machine operation. This irradiation may induce changes both on the membranes characteristics (leakage, modification of elasticity) and on gauges which measure their deformations. To investigate these effects and select the transducer to be used in the LHC, a...

Amand, J F; Casas-Cubillos, J; Thermeau, J P

1999-01-01T23:59:59.000Z

48

Re-START: The second operational test of the String Thermionic Assembly Research Testbed  

SciTech Connect (OSTI)

The second operational test of the String Thermionic Assembly Research Testbed -- Re-START -- was carried out from June 9 to June 14, 1997. This test series was designed to help qualify and validate the designs and test methods proposed for the Integrated Solar Upper Stage (ISUS) power converters for use during critical evaluations of the complete ISUS bimodal system during the Engine Ground Demonstration (EGD). The test article consisted of eight ISUS prototype thermionic converter diodes electrically connected in series.

Wyant, F.J. [Sandia National Labs., Albuquerque, NM (United States); Luchau, D. [TEAM Specialty Services, Inc., Albuquerque, NM (United States); McCarson, T.D. [New Mexico Engineering Research Inst., Albuquerque, NM (United States)

1998-01-01T23:59:59.000Z

49

IRRADIATION TESTING OF THE RERTR FUEL MINIPLATES WITH BURNABLE ABSORBERS IN THE ADVANCED TEST REACTOR  

SciTech Connect (OSTI)

Based on the results of the reactor physics assessment, conversion of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) can be potentially accomplished in two ways, by either using U-10Mo monolithic or U-7Mo dispersion type plates in the ATR fuel element. Both designs, however, would require incorporation of the burnable absorber in several plates of the fuel element to compensate for the excess reactivity and to flatten the radial power profile. Several different types of burnable absorbers were considered initially, but only borated compounds, such as B4C, ZrB2 and Al-B alloys, were selected for testing primarily due to the length of the ATR fuel cycle and fuel manufacturing constraints. To assess and compare irradiation performance of the U-Mo fuels with different burnable absorbers we have designed and manufactured 28 RERTR miniplates (20 fueled and 8 non-fueled) containing fore-mentioned borated compounds. These miniplates will be tested in the ATR as part of the RERTR-13 experiment, which is described in this paper. Detailed plate design, compositions and irradiations conditions are discussed.

I. Glagolenko; D. Wachs; N. Woolstenhulme; G. Chang; B. Rabin; C. Clark; T. Wiencek

2010-10-01T23:59:59.000Z

50

Streamlined Approach for Environmental Restoration Work Plan for Corrective Action Unit 461: Joint Test Assembly Sites and Corrective Action Unit 495: Unconfirmed Joint Test Assembly Sites Tonopah Test Range, Nevada  

SciTech Connect (OSTI)

This Streamlined Approach for Environmental Restoration plan addresses the action necessary for the clean closure of Corrective Action Unit 461 (Test Area Joint Test Assembly Sites) and Corrective Action Unit 495 (Unconfirmed Joint Test Assembly Sites). The Corrective Action Units are located at the Tonopah Test Range in south central Nevada. Closure for these sites will be completed by excavating and evaluating the condition of each artillery round (if found); detonating the rounds (if necessary); excavating the impacted soil and debris; collecting verification samples; backfilling the excavations; disposing of the impacted soil and debris at an approved low-level waste repository at the Nevada Test Site

Jeff Smith

1998-08-01T23:59:59.000Z

51

Test specification for decant pump and winch assembly  

SciTech Connect (OSTI)

This specification provides the requirements for testing of the vertical turbine decant pump including the floating suction arm with load sensing winch control, instrumentation and the associated PLC/PC control system.

Staehr, T.W.

1994-12-07T23:59:59.000Z

52

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.

53

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

SciTech Connect (OSTI)

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

S. Blaine Grover; David A. Petti

2014-05-01T23:59:59.000Z

54

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

SciTech Connect (OSTI)

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.

Blaine Grover

2012-10-01T23:59:59.000Z

55

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

SciTech Connect (OSTI)

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.

G. Borges

2006-01-31T23:59:59.000Z

56

Irradiated Materials Testing Complex (IMTL) The Irradiated Materials Testing Laboratory provides the capability to conduct high temperature  

E-Print Network [OSTI]

provides the capability to conduct high temperature corrosion and stress corrosion cracking of neutron next to a hot cell. This configuration allows us to disconnect the autoclave from its water loop, maneuver it into the hot cell, where the neutron irradiated specimens can be safely mounted

Kamat, Vineet R.

57

ADONIS, high count-rate HP-Ge {gamma} spectrometry algorithm: Irradiated fuel assembly measurement  

SciTech Connect (OSTI)

ADONIS is a digital system for gamma-ray spectrometry, developed by CEA. This system achieves high count-rate gamma-ray spectrometry with correct dynamic dead-time correction, up to, at least, more than an incoming count rate of 3.10{sup 6} events per second. An application of such a system at AREVA NC's La Hague plant is the irradiated fuel scanning facility before reprocessing. The ADONIS system is presented, then the measurement set-up and, last, the measurement results with reference measurements. (authors)

Pin, P. [AREVA NC La Hague - Nuclear Measurement Team, 50444 Beaumont-Hague Cedex (France); Barat, E.; Dautremer, T.; Montagu, T. [CEA - Saclay, LIST, Electronics and Signal Processing Laboratory, 91191 Gif sur Yvette (France); Normand, S. [CEA - Saclay, LIST, Sensors and Electronic Architectures Laboratory, 91191 Gif sur Yvette (France)

2011-07-01T23:59:59.000Z

58

Experiment Safety Assurance Package for the 40- to 50-GWd/MT Burnup Phase of Mixed Oxide Fuel Irradiation in Small I-Hole Positions in the Advanced Test Reactor  

SciTech Connect (OSTI)

This experiment safety assurance package (ESAP) is a revision of the last MOX ESAP issued in February 2001(Khericha 2001). The purpose of this revision is to identify the changes in the loading pattern and to provide a basis to continue irradiation up to {approx}42 GWd/MT burnup (+ 2.5%) as predicted by MCNP (Monte Carlo N-Particle) transport code before the preliminary postirradiation examination (PIE) results for 40 GWd/MT burnup are available. Note that the safety analysis performed for the last ESAP is still applicable and no additional analysis is required (Khericha 2001). In July 2001, it was decided to reconfigure the test assembly using the loading pattern for Phase IV, Part 3, at the end of Phase IV, Part 1, as the loading pattern for Phase IV, Parts 2 and 3. Three capsule assemblies will be irradiated until the highest burnup capsule assembly accumulates: {approx}50 GWd/MT burnup, based on the MCNP code predictions. The last ESAP suggests that at the end of Phase IV, Part 1, we remove the two highest burnup capsule assemblies ({at} {approx}40 GWd/MT burnup) and send them to ORNL for PIE. Then, irradiate the test assembly using the loading pattern for Phase IV, Part 2, until the highest burnup capsule reaches {approx}40 GWd/MT burnup per MCNP-predicted values.

Khericha, S.T.

2002-06-30T23:59:59.000Z

59

Experiment Safety Assurance Package for the 40- to 50-GWd/MT Burnup Phase of Mixed Oxide Fuel Irradiation in Small I-Hole Positions in the Advanced Test Reactor  

SciTech Connect (OSTI)

This experiment safety assurance package (ESAP) is a revision of the last MOX ESAP issued in February 2001(Khericha 2001). The purpose of this revision is to identify the changes in the loading pattern and to provide a basis to continue irradiation up to ~42 GWd/MT burnup (+ 2.5% as predicted by MCNP (Monte Carlo N-Particle) transport code before the preliminary postirradiation examination (PIE) results for 40 GWd/MT burnup are available. Note that the safety analysis performed for the last ESAP is still applicable and no additional analysis is required (Khericha 2001). In July 2001, it was decided to reconfigure the test assembly using the loading pattern for Phase IV, Part 3, at the end of Phase IV, Part 1, as the loading pattern for Phase IV, Parts 2 and 3. Three capsule assemblies will be irradiated until the highest burnup capsule assembly accumulates: ~50 GWd/MT burnup, based on the MCNP code predictions. The last ESAP suggests that at the end of Phase IV, Part 1, we remove the two highest burnup capsule assemblies (@ ~40 GWd/MT burnup) and send them to ORNL for PIE. Then, irradiate the test assembly using the loading pattern for Phase IV, Part 2, until the highest burnup capsule reaches ~40 GWd/MT burnup per MCNP-predicted values.

Khericha, Soli T

2002-06-01T23:59:59.000Z

60

MELT WIRE SENSORS AVAILABLE TO DETERMINE PEAK TEMPERATURES IN ATR IRRADIATION TESTING  

SciTech Connect (OSTI)

In April 2007, the Department of Energy (DOE) designated the Advanced Test Reactor (ATR) a National Scientific User Facility (NSUF) to advance US leadership in nuclear science and technology. By attracting new users from universities, laboratories, and industry, the ATR will support basic and applied nuclear research and development and help address the nation's energy security needs. In support of this new program, the Idaho National Laboratory (INL) has developed in-house capabilities to fabricate, test, and qualify new and enhanced temperature sensors for irradiation testing. Although most efforts emphasize sensors capable of providing real-time data, selected tasks have been completed to enhance sensors provided in irradiation locations where instrumentation leads cannot be included, such as drop-in capsule and Hydraulic Shuttle Irradiation System (HSIS) or 'rabbit' locations. To meet the need for these locations, the INL has developed melt wire temperature sensors for use in ATR irradiation testing. Differential scanning calorimetry and environmental testing of prototypical sensors was used to develop a library of 28 melt wire materials, capable of detecting peak irradiation temperatures ranging from 85 to 1500°C. This paper will discuss the development work and present test results.

K. L. Davis; D. Knudson; J. Daw; J. Palmer; J. L. Rempe

2012-07-01T23:59:59.000Z

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


61

Test plan for the irradiation of nonmetallic materials.  

SciTech Connect (OSTI)

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.

Brush, Laurence H.; Farnum, Cathy Ottinger; Dahl, M. [ARES Corporation, Richland, WA; Joslyn, C. C. [Washington River Protection Solutions, Richland, WA; Venetz, T. J. [Washington River Protection Solutions, Richland, WA

2013-05-01T23:59:59.000Z

62

Test plan for the irradiation of nonmetallic materials.  

SciTech Connect (OSTI)

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.

Brush, Laurence H.; Farnum, Cathy Ottinger; Gelbard, Fred; Dahl, M. [ARES Corporation, Richland, WA; Joslyn, C. C. [Washington River Protection Solutions, Richland, WA; Venetz, T. J. [Washington River Protection Solutions, Richland, WA

2013-03-01T23:59:59.000Z

63

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

SciTech Connect (OSTI)

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.

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

2014-04-01T23:59:59.000Z

64

The Advanced Test Reactor Irradiation Capabilities Available as a National Scientific User Facility  

SciTech Connect (OSTI)

The Advanced Test Reactor (ATR) is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. The ATR is a very versatile facility with a wide variety of experimental test capabilities for providing the environment needed in an irradiation experiment. These capabilities include simple capsule experiments, instrumented and/or temperature-controlled experiments, and pressurized water loop experiment facilities. Monitoring systems have also been utilized to monitor different parameters such as fission gases for fuel experiments, to measure specimen performance during irradiation. ATR’s control system provides a stable axial flux profile throughout each reactor operating cycle, and allows the thermal and fast neutron fluxes to be controlled separately in different sections of the core. The ATR irradiation positions vary in diameter from 16 mm to 127 mm over an active core height of 1.2 m. This paper discusses the different irradiation capabilities with examples of different experiments and the cost/benefit issues related to each capability. The recent designation of ATR as a national scientific user facility will make the ATR much more accessible at very low to no cost for research by universities and possibly commercial entities.

S. Blaine Grover

2008-09-01T23:59:59.000Z

65

Advanced Test Reactor Capabilities and Future Irradiation Plans  

SciTech Connect (OSTI)

The Advanced Test Reactor (ATR), located at the Idaho National Laboratory (INL), is one of the most versatile operating research reactors in the Untied States. The ATR has a long history of supporting reactor fuel and material research for the US government and other test sponsors. The INL is owned by the US Department of Energy (DOE) and currently operated by Battelle Energy Alliance (BEA). The ATR is the third generation of test reactors built at the Test Reactor Area, now named the Reactor Technology Complex (RTC), whose mission is to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The current experiments in the ATR are for a variety of customers--US DOE, foreign governments and private researchers, and commercial companies that need neutrons. The ATR has several unique features that enable the reactor to perform diverse simultaneous tests for multiple test sponsors. The ATR has been operating since 1967, and is expected to continue operating for several more decades. The remainder of this paper discusses the ATR design features, testing options, previous experiment programs, future plans for the ATR capabilities and experiments, and some introduction to the INL and DOE's expectations for nuclear research in the future.

Frances M. Marshall

2006-10-01T23:59:59.000Z

66

COTS FPGA/SRAM Irradiations Using a Dedicated Testing Infrastructure for Characterization of Large Component Batches  

E-Print Network [OSTI]

This paper introduces a new testing platform for irradiation of large batches of COTS FPGA and SRAMs. The main objective is measurement of component radiation response and assessment of component-to-component variability within one batch. The first validation and test results using the testing platform are presented for 150nm TFT SRAM (Renesas) and different sizes of the 130nm ProASIC3 FPGA (Microsemi).

Slawosz, Uznanski; Johannes, Walter; Andrea, Vilar-Villanueva

2015-01-01T23:59:59.000Z

67

Design of the Advanced Gas Reactor Fuel Experiments for Irradiation in the Advanced Test Reactor  

SciTech Connect (OSTI)

The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight particle fuel tests in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL) to support development of the next generation Very High Temperature Reactor (VHTR) in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments will be irradiated in an inert sweep gas atmosphere with on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The final design phase has just been completed on the first experiment (AGR-1) in this series and the support systems and fission product monitoring system that will monitor and control the experiment during irradiation. This paper discusses the development of the experimental hardware and support system designs and the status of the experiment.

S. Blaine Grover

2005-10-01T23:59:59.000Z

68

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

SciTech Connect (OSTI)

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.

Blaise Collin

2014-08-01T23:59:59.000Z

69

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

SciTech Connect (OSTI)

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.

Collin, Blaise P.

2014-08-01T23:59:59.000Z

70

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

SciTech Connect (OSTI)

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.

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

2013-07-01T23:59:59.000Z

71

Design and Testing of a Prototype Spallation Neutron Source Rotating Target Assembly  

SciTech Connect (OSTI)

The mechanical aspects of an extended vertical shaft rotating target have been evaluated in a full-scale mockup test. A prototype assembly based on a conceptual target design for a 1 to 3-MW spallation facility was built and tested. Key elements of the drive/coupling assembly implemented in the prototype include high integrity dynamic face seals, commercially available bearings, realistic manufacturing tolerances, effective monitoring and controls, and fail-safe shutdown features. A representative target disk suspended on a 3.5 meter prototypical shaft was coupled with the drive to complete the mechanical tests. After1800 hours of operation the test program has confirmed the overall mechanical feasibility of the extended vertical shaft rotating target concept. Precision alignment of the suspended target disk; successful containment of the water and verification of operational stability over the full speed range of 30 to 60 rpm were primary indications the proposed mechanical design is valid for use in a high power target station.

Rennich, Mark J [ORNL; McManamy, Thomas J [ORNL; Graves, Van [Oak Ridge National Laboratory (ORNL); Garmendia, Amaia Zarraoa [IDOM Bilbao; Sorda, Fernando [ESS Bilbao

2010-01-01T23:59:59.000Z

72

AGR-1 Irradiation Test Final As-Run Report , Rev 2  

SciTech Connect (OSTI)

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.

Blaise P. Collin

2015-01-01T23:59:59.000Z

73

Charpy impact test results for low activation ferritic alloys irradiated to 30 dpa  

SciTech Connect (OSTI)

Miniature specimens of six low activation ferritic alloys have been impact field tested following irradiation at 370{degrees}C to 30 dpa. Comparison of the results with those of control specimens and specimens irradiated to 10 dpa indicates that degradation in the impact behavior appears to have saturated by {approx}10 dpa in at least four of these alloys. The 7.5Cr-2W alloy referred to as GA3X appears most promising for further consideration as a candidate structural material in fusion reactor applications, although the 9Cr-1V alloy may also warrant further investigation.

Schubert, L.E.; Hamilton, M.L.; Gelles, D.S. [Pacific Northwest National Laboratory, Richland, WA (United States)

1996-04-01T23:59:59.000Z

74

Environmental assessment for device assembly facility operations, Nevada Test Site, Nye County, Nevada. Final report  

SciTech Connect (OSTI)

The U.S. Department of Energy, Nevada Operations Office (DOE/NV), has prepared an environmental assessment (EA), (DOE/EA-0971), to evaluate the impacts of consolidating all nuclear explosive operations at the newly constructed Device Assembly Facility (DAF) in Area 6 of the Nevada Test Site. These operations generally include assembly, disassembly or modification, staging, transportation, testing, maintenance, repair, retrofit, and surveillance. Such operations have previously been conducted at the Nevada Test Site in older facilities located in Area 27. The DAF will provide enhanced capabilities in a state-of-the-art facility for the safe, secure, and efficient handling of high explosives in combination with special nuclear materials (plutonium and highly enriched uranium). Based on the information and analyses in the EA, DOE has determined that the proposed action would not constitute a major federal action significantly affecting the quality of the human environment within the meaning of the National Environmental Policy Act of 1969 (42 U.S.C. 4321 et seq.). Therefore, an environmental impact statement is not required, and DOE is issuing this finding of no significant impact.

NONE

1995-05-01T23:59:59.000Z

75

The data collection system for failure/maintenance at the Tritium Systems Test Assembly  

SciTech Connect (OSTI)

A data collection system for obtaining information which can be used to help determine the reliability and vailability of future fusion power plants has been installed at the Los Alamos National Laboratory's Tritium Systems Test Assembly (TSTA). Failure and maintenance data on components of TSTA's tritium systems have been collected since 1984. The focus of the data collection has been TSTA's Tritium Waste Tratment System (TWT), which has maintained high availability since it became operation in 1982. Data collection is still in progress and a total of 291 failure reports are in the data collection system at this time, 47 of which are from the TWT. 6 refs., 2 figs., 2 tabs.

Casey, M.A.; Gruetzmacher, K.M.; Bartlit, J.R.; Cadwallader, L.C.

1988-01-01T23:59:59.000Z

76

Benchmark physics tests in the metallic-fuelled assembly ZPPR-15  

SciTech Connect (OSTI)

Results of the first benchmark physics tests of a metallic-fueled, demonstration-size, liquid metal reactor are reported. A simple, two-zone, cylindrical conventional assembly was built with three distinctly different compositions to represent the stages of the Integral Fast Reactor fuel cycle. Experiments included criticality, control, power distribution, reaction rate ratios, reactivity coefficients, shielding, kinetics and spectrum. Analysis was done with 3-D nodal diffusion calculations and ENDFIB-V.2 cross sections. Predictions of the ZPPR-15 reactor physics parameters agreed sufficiently well with the measured values to justify confidence in design analyses for metallic-fueled LMRs.

McFarlane, H.F.; Brumbach, S.B.; Carpenter, S.G.; Collins, P.J.

1987-01-01T23:59:59.000Z

77

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

SciTech Connect (OSTI)

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 effects of increased exposure to irradiation, stress, and/or coolant can substantially increase susceptibility to stress-corrosion cracking of austenitic steels in high-temperature water environments. . Despite 30 years of experience, the underlying mechanisms of IASCC are unknown. Extended service conditions will increase the exposure to irradiation, stress, and corrosive environment for all core internal components. The objective of this effort within the Light Water Reactor Sustainability program is to evaluate the response and mechanisms of IASCC in austenitic stainless steels with single variable experiments. A series of high-value irradiated specimens has been acquired from the past international research programs, providing a valuable opportunity to examine the mechanisms of IASCC. This batch of irradiated specimens has been received and inventoried. In addition, visual examination and sample cleaning has been completed. Microhardness testing has been performed on these specimens. All samples show evidence of hardening, as expected, although the degree of hardening has saturated and no trend with dose is observed. Further, the change in hardening can be converted to changes in mechanical properties. The calculated yield stress is consistent with previous data from light water reactor conditions. In addition, some evidence of changes in deformation mode was identified via examination of the microhardness indents. This analysis may provide further insights into the deformation mode under larger scale tests. Finally, swelling analysis was performed using immersion density methods. Most alloys showed some evidence of swelling, consistent with the expected trends for this class of alloy. The Hf-doped alloy showed densification rather than swelling. This observation may be related to the formation of second-phases under irradiation, although further examination is required

Busby, Jeremy T [ORNL; Gussev, Maxim N [ORNL

2011-04-01T23:59:59.000Z

78

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

SciTech Connect (OSTI)

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.

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

2013-07-01T23:59:59.000Z

79

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

SciTech Connect (OSTI)

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.

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

2014-01-01T23:59:59.000Z

80

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

SciTech Connect (OSTI)

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.

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

2012-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "test assembly irradiation" from the National Library of EnergyBeta (NLEBeta).
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81

Assembly and Testing of a Radioisotope Power System for the New Horizons Spacecraft  

SciTech Connect (OSTI)

The Idaho National Laboratory (INL) recently fueled and assembled a radioisotope power system (RPS) that was used upon the New Horizons spacecraft which was launched in January 2006. New Horizons is the first mission to the last planet - the initial reconnaissance of Pluto-Charon and the Kuiper Belt, exploring the mysterious worlds at the edge of our solar system. The RPS otherwise known as a "space battery" converts thermal heat into electrical energy. The thermal heat source contains plutonium dioxide in the form of ceramic pellets encapsulated in iridium metal. The space battery was assembled in a new facility at the Idaho National Laboratory site near Idaho Falls, Idaho. The new facility has all the fueling and testing capabilities including the following: the ability to handle all the shipping containers currently certified to ship Pu-238, the ability to fuel a variety of RPS designs, the ability to perform vibrational testing to simulate transportation and launch environments, welding systems, a center of mass determination device, and various other support systems.

Kenneth E. Rosenberg; Stephen G. Johnson

2006-06-01T23:59:59.000Z

82

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

SciTech Connect (OSTI)

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.

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)

2010-10-15T23:59:59.000Z

83

Completing the Design of the Advanced Gas Reactor Fuel Development and Qualification Experiments for Irradiation in the Advanced Test Reactor  

SciTech Connect (OSTI)

The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the newly formed 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 ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new 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 AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new 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 six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

S. Blaine Grover

2006-10-01T23:59:59.000Z

84

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

SciTech Connect (OSTI)

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.

Blaine Grover

2012-10-01T23:59:59.000Z

85

Physics calculations for the RIA 1-3 irradiated rod test  

SciTech Connect (OSTI)

The RIA 1-3 test would employ a square array of four pre-irradiated BWR rods to provide information on fuel failure modes and consequences of postulated Reactivity Initiated Accidents in power reactors. Calculations were done to: (1) predict R-O power distributions in the test rods for thermal-hydraulic and fuel-failure analysis; and (2) predict the steady-state and transient ratios of test fuel energy deposition to core energy deposition (Figures of Merit). Fission distributions for the test were computed with the RAFFL Monte Carlo code using an external neutron current source from a complete-reactor radial calculation with the SCAMP S/sub n/ code. Energies per fission for the rods were computed using the SINBAD buildup and depletion code, the GAMSOR gamma ray source code, and the QAD-BSA point-kernel shielding code. The calculated rod average-to-test average energy deposition ratios are 0.99, 0.99, and 0.97 for the rods irradiated to approximately 12 CWd/tu, and 1.04 for the rod irradiated to 4.8 GWd/tu. The maximum deviation of the power density of 1/12-rod azimuthal segments from the rod average is 4%. For an estimated control rod position of 0.591 m withdrawn the predicted radial average energy deposition at the axial peak in an average test rod is 1.71 (kW/m)/MW during preconditioning, and 1.84 (kJ/kg UO/sub 2/) MW.S during the burst. 16 figures, 7 tables.

Young, T.E.

1981-06-01T23:59:59.000Z

86

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

SciTech Connect (OSTI)

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.

Andreani, C.; Pietropaolo, A.; Salsano, A. [Centro NAST, Universita degli Studi di Roma Tor Vergata (Italy); Gorini, G.; Tardocchi, M. [Dipartimento di Fisica 'G. Occhialini', Universita degli Studi di Milano-Bicocca (Italy); Paccagnella, A.; Gerardin, S. [Dipartimento di Ingegneria dell'Informazione, Universita di Padova (Italy); Frost, C. D.; Ansell, S. [ISIS Facility, Rutherford Appleton Laboratory, Chilton, Didcot, Oxfordshire OX11 0QX (United Kingdom); Platt, S. P. [School of Computing, Engineering and Physical Sciences, University of Central Lancashire, Preston, Lancs. PR1 2HE (United Kingdom)

2008-03-17T23:59:59.000Z

87

The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor  

SciTech Connect (OSTI)

The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating eight 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). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new 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 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 ten years to demonstrate and qualify new 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 the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In addition, the purpose and differences between the two experiments will be compared and the irradiation results to date on the first experiment will be presented.

S. Blaine Grover

2009-09-01T23:59:59.000Z

88

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)

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.

M. Chen; CM Regan; D. Noe

2006-01-09T23:59:59.000Z

89

Further Charpy impact test results of low activation ferritic alloys, irradiated at 430{degrees}C to 67 dpa  

SciTech Connect (OSTI)

Miniature CVN specimens of four ferritic alloys, GA3X, F82H, GA4X and HT9, have been impact tested following irradiation at 430{degrees}C to 67 dpa. Comparison of the results with those of the previously tested lower dose irradiation condition indicates that the GA3X and F82H alloys, two primary candidate low activation alloys, exhibit virtually identical behavior following irradiation at 430{degrees}C to {approximately}67 dpa and at 370{degrees}C to {approximately}15 dpa. Very little shift is observed in either DBTT or USE relative to the unirradiated condition. The shifts in DBTT and USE observed in both GA4X and HT9 were smaller after irradiation at 430{degrees}C to {approximately}67 dpa than after irradiation at 370{degrees}C to {approximately}15 dpa.

Schubert, L.E.; Hamilton, M.L.; Gelles, D.S. [Pacific Northwest National Lab., Richland, WA (United States)

1997-04-01T23:59:59.000Z

90

INITIAL IRRADIATION OF THE FIRST ADVANCED GAS REACTOR FUEL DEVELOPMENT AND QUALIFICATION EXPERIMENT IN THE ADVANCED TEST REACTOR  

SciTech Connect (OSTI)

The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate 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 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 AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

S. Blaine Grover; David A. Petti

2007-09-01T23:59:59.000Z

91

AGR-1 Irradiated Test Train Preliminary Inspection and Disassembly First Look  

SciTech Connect (OSTI)

The AGR-1 irradiation experiment ended on November 6, 2009, after 620 effective full power days in the Advanced Test Reactor, achieving a peak burnup of 19.6% FIMA. The test train was shipped to the Materials and Fuels Complex in March 2010 for post-irradiation examination. The first PIE activities included non-destructive 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 the graphite fuel holders. Dimensional measurements of the compacts, graphite holders, and steel capsules shells were performed using a custom vision measurement system (for outer diameters and lengths) and conventional bore gauges (for inner diameters). 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. Neutron radiography of the intact Capsule 2 showed a high degree of detail of interior components and confirmed the observation that there was no major damage to the capsule. Disassembly of the capsules was initiated using procedures qualified during out-of-cell mockup testing. Difficulties were encountered during capsule disassembly due to irradiation-induced changes in some of the capsule components’ properties, including embrittled niobium and molybdenum parts that were susceptible to fracture and swelling of the graphite fuel holders that affected their removal from the capsule shells. This required various improvised modifications to the disassembly procedure to avoid damage to the fuel compacts. Ultimately the capsule disassembly was successful and only one compact from Capsule 4 (out of 72 total in the test train) sustained damage during the disassembly process, along with the associated graphite holder. The compacts were generally in very good condition upon removal. Only relatively minor damage or markings were visible using high resolution photographic inspection. Compact dimensional measurements indicated diametrical shrinkage of 0.9 to 1. 4%, and length shrinkage of 0.2 to 1.1%. The shrinkage was somewhat dependent on compact location within each capsule and within the test train. Compacts exhibited a maximum diametrical shrinkage at a fast neutron fluence of approximately 3×1021 n/cm2. A multivariate statistical analysis indicates that fast neutron fluence as well as compact position in the test train influence compact shrinkage.

Paul Demkowicz; Lance Cole; Scott Ploger; Philip Winston; Binh Pham; Michael Abbott

2011-01-01T23:59:59.000Z

92

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

SciTech Connect (OSTI)

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.

S. Blaine Grover

2014-05-01T23:59:59.000Z

93

Overview of tritium processing development at the tritium systems test assembly  

SciTech Connect (OSTI)

The Tritium Systems Test Assembly (TSTA) at the Los Alamos National Laboratory has been operating with tritium since June 1984. Presently there are some 50 g of tritium in the main processing loop. This 50 g has been sufficient to do a number of experiments involving the cryogenic distillation isotope separation system and to integrate the fuel cleanup system into the main fuel processing loop. In January 1986 two major experiments were conducted. During these experiments the fuel cleanup system was integrated, through the transfer pumping system, with the isotope separation system, thus permitting testing on the integrated fuel processing loop. This integration of these systems leaves only the main vacuum system to be integrated into the TSTA fuel processing loop. In September 1986 another major tritium experiment was performed in which the integrated loop was operated, the tritium inventory increased to 50 g and additional measurements on the performance of the distillation system were taken. In the period June 1984 through September 1986 the TSTA system has processed well over 10/sup 8/ Ci of tritium. Total tritium emissions to the environment over this period have been less than 15 Ci. Personnel exposures during this period have totaled less than 100 person-mRem. To date, the development of tritium technology at TSTA has proceeded in progressive and orderly steps. In two years of operation with tritium, no major design flows have been uncovered.

Anderson, J.L.

1986-10-22T23:59:59.000Z

94

Self-aligning hydraulic piston assembly for tensile testing of ceramic  

DOE Patents [OSTI]

The present invention is directed to a self-aligning grip housing assembly that can transmit an uniaxial load to a tensil 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 irregardless whether there is initial misalignment between the specimen load column assembly and housing axis.

Liu, Kenneth C. (Oak Ridge, TN)

1987-01-01T23:59:59.000Z

95

Self-aligning hydraulic piston assembly for tensile testing of ceramic  

DOE Patents [OSTI]

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.

Liu, K.C.

1987-08-18T23:59:59.000Z

96

Gas Generation from K East Basin Sludges and Irradiated Metallic Uranium Fuel Particles Series III Testing  

SciTech Connect (OSTI)

The path forward for managing of Hanford K Basin sludge calls for it to be packaged, shipped, and stored at T Plant until final processing at a future date. An important consideration for the design and cost of retrieval, transportation, and storage systems is the potential for heat and gas generation through oxidation reactions between uranium metal and water. This report, the third in a series (Series III), describes work performed at the Pacific Northwest National Laboratory (PNNL) to assess corrosion and gas generation from irradiated metallic uranium particles (fuel particles) with and without K Basin sludge addition. The testing described in this report consisted of 12 tests. In 10 of the tests, 4.3 to 26.4 g of fuel particles of selected size distribution were placed into 60- or 800-ml reaction vessels with 0 to 100 g settled sludge. In another test, a single 3.72-g fuel fragment (i.e., 7150-mm particle) was placed in a 60 ml reaction vessel with no added sludge. The twelfth test contained only sludge. The fuel particles were prepared by crushing archived coupons (samples) from an irradiated metallic uranium fuel element. After loading the sludge materials (whether fuel particles, mixtures of fuel particles and sludge, or sludge-only) into reaction vessels, the solids were covered with an excess of K Basin water, the vessels closed and connected to a gas measurement manifold, and the vessels back-flushed with inert neon cover gas. The vessels were then heated to a constant temperature. The gas pressures and temperatures were monitored continuously from the times the vessels were purged. Gas samples were collected at various times during the tests, and the samples analyzed by mass spectrometry. Data on the reaction rates of uranium metal fuel particles with water as a function of temperature and particle size were generated. The data were compared with published studies on metallic uranium corrosion kinetics. The effects of an intimate overlying sludge layer (''blanket'') on the uranium metal corrosion rates were also evaluated.

Schmidt, Andrew J.; Delegard, Calvin H.; Bryan, Samuel A.; Elmore, Monte R.; Sell, Rachel L.; Silvers, Kurt L.; Gano, Susan R.; Thornton, Brenda M.

2003-08-01T23:59:59.000Z

97

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

SciTech Connect (OSTI)

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.

Blaine Grover

2012-10-01T23:59:59.000Z

98

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

SciTech Connect (OSTI)

In support of the Large Hadron Collider (LHC) luminosity upgrade, the US LHC Accelerator Research Program (LARP) has been developing a 1-meter long, 120 mm bore Nb{sub 3}Sn IR quadrupole magnet (HQ). With a design short sample gradient of 219 T/m at 1.9 K and a peak field approaching 15 T, one of the main challenges of this magnet is to provide appropriate mechanical support to the coils. Compared to the previous LARP Technology Quadrupole and Long Quadrupole magnets, the purpose of HQ is also to demonstrate accelerator quality features such as alignment and cooling. So far, 8 HQ coils have been fabricated and 4 of them have been assembled and tested in HQ01a. This paper presents the mechanical assembly and test results of HQ01a.

Felice, H.; Caspi, S.; Cheng, D.; Dietderich, D.; Ferracin, P.; Hafalia, R.; Joseph, J.; Lizarazo, J.; Sabbi, G. L.; Wang, X.; Anerella, M.; Ghosh, A. K.; Schmalzle, J.; Wanderer, P.; Ambrosio, G.; Bossert, R.; Zlobin, A. V.

2010-05-23T23:59:59.000Z

99

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

SciTech Connect (OSTI)

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.

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

2011-03-01T23:59:59.000Z

100

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

SciTech Connect (OSTI)

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.

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

2014-01-01T23:59:59.000Z

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


101

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

SciTech Connect (OSTI)

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.

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

1995-02-01T23:59:59.000Z

102

Board scheduling for circuit board assembly : computational testing of an integer programming approach  

E-Print Network [OSTI]

, and decreasing costs. Optimization techniques involve process planning and lead to shortened cycle times and reduced costs. No generic workload-balancing tool is available to optimize the assembly process. This study is part of a larger, ATP-sponsored project...

Zetangue, Natalie F

2013-02-22T23:59:59.000Z

103

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

SciTech Connect (OSTI)

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.

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

2014-05-01T23:59:59.000Z

104

Thermally Simulated 32kW Direct-Drive Gas-Cooled Reactor: Design, Assembly, and Test  

SciTech Connect (OSTI)

One of the power systems under consideration for nuclear electric propulsion is a direct-drive gas-cooled reactor coupled to a Brayton cycle. In this system, power is transferred from the reactor to the Brayton system via a circulated closed loop gas. To allow early utilization, system designs must be relatively simple, easy to fabricate, and easy to test using non-nuclear heaters to closely mimic heat from fission. This combination of attributes will allow pre-prototypic systems to be designed, fabricated, and tested quickly and affordably. The ability to build and test units is key to the success of a nuclear program, especially if an early flight is desired. The ability to perform very realistic non-nuclear testing increases the success probability of the system. In addition, the technologies required by a concept will substantially impact the cost, time, and resources required to develop a successful space reactor power system. This paper describes design features, assembly, and test matrix for the testing of a thermally simulated 32kW direct-drive gas-cooled reactor in the Early Flight Fission - Test Facility (EFF-TF) at Marshall Space Flight Center. The reactor design and test matrix are provided by Los Alamos National Laboratories.

Godfroy, Thomas J.; Bragg-Sitton, Shannon M. [NASA Marshall Space Flight Center, TD40, Huntsville, Alabama, 35812 (United States); University of Michgan, Dept. of Nuclear Engineering and Radiological Sciences, Ann Arbor MI 48109 (United States); Kapernick, Richard J. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

2004-02-04T23:59:59.000Z

105

Experimentally testing and assessing the predictive power of species assembly rules for tropical canopy ants  

E-Print Network [OSTI]

to be sufficient to stabilise this quantity for all assembly rules of interest (with the exception of the null rule, which is, by definition, stochastic). To avoid gradual reduction of diversity due to local extinction of species, we reintroduced extinct species... ). The impact of forest conversion to oil palm on arthropod abundance and biomass in Sabah, Malaysia. J. Trop. Ecol., 25, 23–30. Turner, E.C., Snaddon, J.L., Johnson, H.R. & Foster, W.A. (2007). The impact of bird’s nest ferns on stemflow nutrient concentration...

Fayle, Tom M.; Eggleton, Paul; Manica, Andrea; Yusah, Kalsum M.; Foster, William A.

2015-01-27T23:59:59.000Z

106

Assembly Bias & Redshift-Space Distortions: Impact on cluster dynamics tests of general relativity  

E-Print Network [OSTI]

The redshift-space distortion (RSD) of galaxies surrounding massive clusters is emerging as a promising testbed for theories of modified gravity. Conventional applications of this method rely upon the assumption that the velocity field in the cluster environment is uniquely determined by the cluster mass profile. Yet, real dark matter halos in N-body simulations are known to violate the assumption that virial mass determines the configuration space distribution, an effect known as assembly bias. In this Letter, I show that assembly bias in simulated dark matter halos also manifests in velocity space. In the 1-10 Mpc environment surrounding a cluster, high-concentration "tracer" halos exhibit a 10-20% larger pairwise-velocity dispersion profile relative to low-concentration tracer halos of the same mass. This difference is comparable to the size of the RSD signal predicted by f(R) models designed to account for the cosmic acceleration. I use the age matching technique to study how color-selection effects may i...

Hearin, Andrew P

2015-01-01T23:59:59.000Z

107

Design, Feasibility, and Testing of Instrumented Rod Bundles to Improve Heat Transfer Knowledge in PWR Fuel Assemblies  

SciTech Connect (OSTI)

Two 5 x 5 test rod bundles mimicking the PWR fuel assembly have been adapted into two suitable test loop facilities, respectively, to carry out sufficiently detailed hydraulic and thermal measurements in identical geometric configuration. The objective is to investigate heat transfer phenomena in single-phase as well as with onset of nucleate boiling (ONB). The accuracy and reproducibility of the temperature measurements using the sliding-traversing thermocouple device under typical PWR conditions has been demonstrated in the thermal test facility. In the hydraulic loop, a Laser Doppler Velocimetry (LDV) system to precisely scan the local axial velocity component in each sub-channel has been implemented. The approach is to utilize mean sub-channel axial velocity distributions and pressure drop data from the hydraulic loop and the global boundary conditions (Pressure, Temperature, flow rate) from the thermal loop to simulate sub-channels in appropriate T/H codes. This permits computation of sub-channel averaged fluid temperatures (as well as mass velocity) in various subchannels within the test bundle. Subsequently, in conjunction with the wall temperatures and applied heat flux values from the thermal loop, it is possible to develop a complete map of heat transfer coefficients along the 9 instrumented central heater rods. Locations downstream of spacer grids would be of special interest. Depending on pressure, mass velocity and heat flux conditions of a given test, the inlet temperature will be a parameter to be varied so that the ONB boundary can be observed within the bundle. Detailed designs of the test section, required loop modifications, and adaptation of specialized instrumentation and data acquisition systems have been accomplished in both test loops. Further we have established that based on such detailed rod surface temperature and sub-channel axial velocity measurements, it is possible to achieve sufficient accuracy in the temperature measurements to meet the objective of improving the heat transfer correlations applicable to PWR cores. (authors)

Bergeron, A. [CEA, Saclay (France); Chataing, T.; Garnier, J. [CEA, Genoble (France); Decossin, E.; Peturaud, P. [EDF/R and D, Chatou (France); Yagnik, S.K. [Electric Power Research Institute - EPRI (United States)

2007-07-01T23:59:59.000Z

108

1994 Baseline biological studies for the Device Assembly Facility at the Nevada Test Site  

SciTech Connect (OSTI)

This report describes environmental work performed at the Device Assembly Facility (DAF) in 1994 by the Basic Environmental Monitoring and Compliance Program (BECAMP). The DAF is located near the Mojave-Great Basin desert transition zone 27 km north of Mercury. The area immediately around the DAF building complex is a gentle slope cut by 1 to 3 m deep arroyos, and occupied by transitional vegetation. In 1994, construction activities were largely limited to work inside the perimeter fence. The DAF was still in a preoperational mode in 1994, and no nuclear materials were present. The DAF facilities were being occupied so there was water in the sewage settling pond, and the roads and lights were in use. Sampling activities in 1994 represent the first year in the proposed monitoring scheme. The proposed biological monitoring plan gives detailed experimental protocols. Plant, lizard, tortoise, small mammal, and bird surveys were performed in 1994. The authors briefly outline procedures employed in 1994. Studies performed on each taxon are reviewed separately then summarized in a concluding section.

Townsend, Y.E. [ed.; Woodward, B.D.; Hunter, R.B.; Greger, P.D.; Saethre, M.B.

1995-02-01T23:59:59.000Z

109

Assembly and bench testing of a spiral fiber tracker for the J-PARC TREK/E36 experiment  

E-Print Network [OSTI]

This study presents the recent progress made in developing a spiral fiber tracker (SFT) for use in the experiment TREK/E36 planned at the Japan Proton Accelerator Research Complex. This kaon decay experiment uses a stopped positive kaon beam to search for physics beyond the Standard Model through precision measurements of lepton universality and through searches for a heavy sterile neutrino and a dark photon. Detecting and tracking positrons and positive muons from kaon decays are of importance in achieving high-precision measurements; therefore, we designed and are developing the new tracking detector using a scintillating fiber. The SFT was completely assembled, and in a bench test, no dead channel was determined.

Makoto Tabata; Sébastien Bianchin; Michael D. Hasinoff; Robert S. Henderson; Keito Horie; Youichi Igarashi; Jun Imazato; Hiroshi Ito; Alexander Ivashkin; Hideyuki Kawai; Yury Kudenko; Oleg Mineev; Suguru Shimizu; Akihisa Toyoda; Hirohito Yamazaki

2014-11-29T23:59:59.000Z

110

Advanced industrial gas turbine technology readiness demonstration program. Phase II. Final report: compressor rig fabrication assembly and test  

SciTech Connect (OSTI)

The results of a component technology demonstration program to fabricate, assemble and test an advanced axial/centrifugal compressor are presented. This work was conducted to demonstrate the utilization of advanced aircraft gas turbine cooling and high pressure compressor technology to improve the performance and reliability of future industrial gas turbines. Specific objectives of the compressor component testing were to demonstrate 18:1 pressure ratio on a single spool at 90% polytropic efficiency with 80% fewer airfoils as compared to current industrial gas turbine compressors. The compressor design configuration utilizes low aspect ratio/highly-loaded axial compressor blading combined with a centrifugal backend stage to achieve the 18:1 design pressure ratio in only 7 stages and 281 axial compressor airfoils. Initial testing of the compressor test rig was conducted with a vaneless centrifugal stage diffuser to allow documentation of the axial compressor performance. Peak design speed axial compressor performance demonstrated was 91.8% polytropic efficiency at 6.5:1 pressure ratio. Subsequent documentation of the combined axial/centrifugal performance with a centrifugal stage pipe diffuser resulted in the demonstration of 91.5% polytropic efficiency and 14% stall margin at the 18:1 overall compressor design pressure ratio. The demonstrated performance not only exceeded the contract performance goals, but also represents the highest known demonstrated compressor performance in this pressure ratio and flow class. The performance demonstrated is particularly significant in that it was accomplished at airfoil loading levels approximately 15% higher than that of current production engine compressor designs. The test results provide conclusive verification of the advanced low aspect ratio axial compressor and centrifugal stage technologies utilized.

Schweitzer, J. K.; Smith, J. D.

1981-03-01T23:59:59.000Z

111

Charpy impact test results of four low activation ferritic alloys irradiated at 370{degrees}C to 15 DPA  

SciTech Connect (OSTI)

Miniature CVN specimens of four low activation ferritic alloys have been impact tested following irradiation at 370{degrees}C to 15 dpa. Comparison of the results with those of control specimens indicates that degradation in the impact behavior occurs in each of these four alloys. The 9Cr-2W alloy referred to as GA3X and the similar alloy F82H with 7.8Cr-2W appear most promising for further consideration as candidate structural materials in fusion energy system applications. These two alloys exhibit a small DBTT shift to higher temperatures but show increased absorbed energy on the upper shelf.

Schubert, L.E.; Hamilton, M.L.; Gelles, D.S. [Pacific Northwest National Lab., Richland, WA (United States)

1996-10-01T23:59:59.000Z

112

RIS-M-2185 CALCULATION OF HEAT RATING AND BURN-UP FOR TEST FUEL PINS  

E-Print Network [OSTI]

RISØ-M-2185 CALCULATION OF HEAT RATING AND BURN-UP FOR TEST FUEL PINS IRRADIATED IN DR3 C. Bagger of fuel pins irradiated in HP1 rigs. The calculations are carried out rather detailed, especially of the data. INIS Descriptors . BURN-UP, CALORIMETRY, COMPUTER CALCULATIONS, DR-3, FISSION, FUEL ASSEMBLIES

113

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

SciTech Connect (OSTI)

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

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

1988-01-01T23:59:59.000Z

114

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

SciTech Connect (OSTI)

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

Thomas Blue; Wolfgang Windl; Bryan Dickerson

2013-01-03T23:59:59.000Z

115

Imminent: Irradiation Testing of (Th,Pu)O{sub 2} Fuel - 13560  

SciTech Connect (OSTI)

Commercial-prototype thorium-plutonium oxide (Th-MOX) fuel pellets have been loaded into the material test reactor in Halden, Norway. The fuel is being operated at full power - with instrumentation - in simulated LWR / PHWR conditions and its behaviour is measured 'on-line' as it operates to high burn-up. This is a vital test on the commercialization pathway for this robust new thoria-based fuel. The performance data that is collected will support a fuel modeling effort to support its safety qualification. Several different samples of Th-MOX fuel will be tested, thereby collecting information on ceramic behaviours and their microstructure dependency. The fuel-cycle reasoning underpinning the test campaign is that commercial Th- MOX fuels are an achievable intermediate / near-term SNF management strategy that integrates well with a fast reactor future. (authors)

Kelly, Julian F. [Thor Energy AS, Sommerrogaten 13-15, Oslo 0255 (Norway)] [Thor Energy AS, Sommerrogaten 13-15, Oslo 0255 (Norway); Franceschini, Fausto [Westinghouse Electric Company LLC, 1000 Cranberry Woods Drive, Cranberry Township, PA 16066 (United States)] [Westinghouse Electric Company LLC, 1000 Cranberry Woods Drive, Cranberry Township, PA 16066 (United States)

2013-07-01T23:59:59.000Z

116

Microsoft Word - AGR-1_Irradiation-Test-Final-As-Run-Report_rev1...  

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

objectives of the AGR-1 experiment are to: (a) Gain experience with multi-capsule test train design, fabrication, and operation with the intent to reduce the probability of...

117

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

SciTech Connect (OSTI)

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.

Troy Unruh; Michael Reichenberger; Phillip Ugorowski

2013-09-01T23:59:59.000Z

118

High speed door assembly  

DOE Patents [OSTI]

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.

Shapiro, C.

1993-04-27T23:59:59.000Z

119

High speed door assembly  

DOE Patents [OSTI]

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.

Shapiro, Carolyn (Idaho Falls, ID)

1993-01-01T23:59:59.000Z

120

Design of irradiation rig for reactor testing of prototype bolometers for ITER  

SciTech Connect (OSTI)

We describe the design of an experimental rig, which was developed to allow reactor testing at relevant conditions, i.e. vacuum and {approx}400 deg.C temperature, of prototype resistive bolometers, which will be used in ITER to acquire information on the radiated power distribution from the main plasma and in the diverter region. The main feature of the design is that the rig has no active temperature control. (authors)

Gusarov, A.; Huysmans, S. [SCK.CEN Belgian Nucrear Research Center, 2400 Mol (Belgium); Meister, H. [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstr. 2, D-85748 Garching b. Muenchen (Germany); Hodgson, E. [Euratom/CIEMAT Fusion Association, Avenida Complutense 22, 28040 Madrid (Spain)

2011-07-01T23:59:59.000Z

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


121

Experiment Safety Assurance Package for the 40- to 52-GWd/MT Burnup Phase of Mixed Oxide Fuel Irradiation in Small I-hole Positions in the Advanced Test Reactor  

SciTech Connect (OSTI)

This experiment safety assurance package (ESAP) is a revision of the last mixed uranium and plutonium oxide (MOX) ESAP issued in June 2002). The purpose of this revision is to provide a basis to continue irradiation up to 52 GWd/MT burnup [as predicted by MCNP (Monte Carlo N-Particle) transport code The last ESAP provided basis for irradiation, at a linear heat generation rate (LHGR) no greater than 9 kW/ft, of the highest burnup capsule assembly to 50 GWd/MT. This ESAP extends the basis for irradiation, at a LHGR no greater than 5 kW/ft, of the highest burnup capsule assembly from 50 to 52 GWd/MT.

S. T. Khericha; R. C. Pedersen

2003-09-01T23:59:59.000Z

122

Experiment Safety Assurance Package for Mixed Oxide Fuel Irradiation in an Average Power Position (I-24) in the Advanced Test Reactor  

SciTech Connect (OSTI)

The Fissile Material Disposition Program Light Water Reactor Mixed Oxide Fuel Irradiation Test Project Plan details a series of test irradiations designed to investigate the use of weapons-grade plutonium in MOX fuel for light water reactors (LWR) (Cowell 1996a, Cowell 1997a, Thoms 1997a). Commercial MOX fuel has been successfully used in overseas reactors for many years; however, weapons-derived test fuel contains small amounts of gallium (about 2 parts per million). A concern exists that the gallium may migrate out of the fuel and into the clad, inducing embrittlement. For preliminary out-of-pile experiments, Wilson (1997) states that intermetallic compound formation is the principal interaction mechanism between zircaloy cladding and gallium. This interaction is very limited by the low mass of gallium, so problems are not expected with the zircaloy cladding, but an in-pile experiment is needed to confirm the out-of-pile experiments. Ryskamp (1998) provides an overview of this experiment and its documentation. The purpose of this Experiment Safety Assurance Package (ESAP) is to demonstrate the safe irradiation and handling of the mixed uranium and plutonium oxide (MOX) Fuel Average Power Test (APT) experiment as required by Advanced Test Reactor (ATR) Technical Safety Requirement (TSR) 3.9.1 (LMITCO 1998). This ESAP addresses the specific operation of the MOX Fuel APT experiment with respect to the operating envelope for irradiation established by the Upgraded Final Safety Analysis Report (UFSAR) Lockheed Martin Idaho Technologies Company (LMITCO 1997a). Experiment handling activities are discussed herein.

J. M . Ryskamp; R. C. Howard; R. C. Pedersen; S. T. Khericha

1998-10-01T23:59:59.000Z

123

Irradiation-Free, Columnar Defects Comprised of Self-Assembled Nanodots and Nanorods Resulting in Strongly Enhanced Flux-Pinning in YBa2Cu307-? Films  

SciTech Connect (OSTI)

The development of biaxially textured, second-generation, high-temperature superconducting (HTS) wires is expected to enable most large-scale applications of HTS materials, in particular electric-power applications. For many potential applications, high critical currents in applied magnetic fields are required. It is well known that columnar defects generated by irradiating high-temperature superconducting materials with heavy ions significantly enhance the in-field critical current density. Hence, for over a decade scientists world-wide have sought means to produce such columnar defects in HTS materials without the expense and complexity of ionizing radiation. Using a simple and practically scalable technique, we have succeeded in producing long, nearly continuous vortex pins along the c-axis in YBa{sub 2}Cu{sub 3}O{sub 7-{delta}} (YBCO), in the form of self-assembled stacks of BaZrO{sub 3} (BZO) nanodots and nanorods. The nanodots and nanorods have a diameter of {approx}2-3 nm and an areal density ('matching field') of 8-10 T for 2 vol.% incorporation of BaZrO{sub 3}. In addition, four misfit dislocations around each nanodot or nanorod are aligned and act as extended columnar defects. YBCO films with such defects exhibit significantly enhanced pinning with less sensitivity to magnetic fields H. In particular, at intermediate field values, the current density, J{sub c}, varies as J{sub c} {approx}H{sup -{alpha}}, with {alpha} {approx} 0.3 rather than the usual values 0.5-0.65. Similar results were also obtained for CaZrO{sub 3} (CZO) and YSZ incorporation in the form of nanodots and nanorods within YBCO, indicating the broad applicability of the developed process. The process could also be used to incorporate self-assembled nanodots and nanorods within matrices of other materials for different applications, such as magnetic materials.

Goyal, Amit [ORNL; Kang, Sukill [ORNL; Leonard, Keith J [ORNL; Martin, Patrick M [ORNL; Gapud, Albert Agcaoili [ORNL; Varela del Arco, Maria [ORNL; Paranthaman, Mariappan Parans [ORNL; Ijaduola, Anota O [ORNL; Specht, Eliot D [ORNL; Thompson, James R [ORNL; Christen, David K [ORNL; Pennycook, Stephen J [ORNL; List III, Frederick Alyious [ORNL

2005-11-01T23:59:59.000Z

124

BWR Fuel Assembly BWR Fuel Assembly PWR Fuel Assembly  

National Nuclear Security Administration (NNSA)

BWR Fuel Assembly BWR Fuel Assembly PWR Fuel Assembly PWR Fuel Assembly The PWR 17x17 assembly is approximately 160 inches long (13.3 feet), 8 inches across, and weighs 1,500 lbs....

125

Tests of the radiation hardness of VLSI Integrated Circuits and Silicon Strip Detectors for the SSC (Superconducting Super Collider) under neutron, proton, and gamma irradiation  

SciTech Connect (OSTI)

As part of a program to develop a silicon strip central tracking detector system for the Superconducting Super Collider (SSC) we are studying the effects of radiation damage in silicon detectors and their associated front-end readout electronics. We report on the results of neutron and proton irradiations at the Los Alamos National Laboratory (LANL) and {gamma}-ray irradiations at UC Santa Cruz (UCSC). Individual components on single-sided AC-coupled silicon strip detectors and on test structures were tested. Circuits fabricated in a radiation hard CMOS process and individual transistors fabricated using dielectric isolation bipolar technology were also studied. Results indicate that a silicon strip tracking detector system should have a lifetime of at least one decade at the SSC. 17 refs., 17 figs.

Ziock, H.J.; Milner, C.; Sommer, W.F. (Los Alamos National Lab., NM (USA)); Carteglia, N.; DeWitt, J.; Dorfan, D.; Hubbard, B.; Leslie, J.; O'Shaughnessy, K.F.; Pitzl, D.; Rowe, W.A.; Sadrozinski, H.F.W.; Seiden, A.; Spencer, E. (California Univ., Santa Cruz, CA (USA). Inst. for Particle Physics); Ellison, J.A. (California Univ., Riverside, CA (USA)); Ferguson, P. (Missouri Univ., Rolla, MO (USA)); Giubellino

1990-01-01T23:59:59.000Z

126

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

SciTech Connect (OSTI)

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.

G. Palmiotti

2011-12-01T23:59:59.000Z

127

High speed door assembly  

SciTech Connect (OSTI)

This invention is comprised of 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.

Shapiro, C.

1991-12-31T23:59:59.000Z

128

Activity testing of alveolar macrophages and changes in surfactant phospholipids after irradiation in bronchoalveolar lavage: Experimental and clinical data  

SciTech Connect (OSTI)

This study presents results of bronchoalveolar lavage (BAL) after irradiation to the lungs in mice as well as clinical data. The number of BAL cells, mainly macrophages, lymphocytes, and granulocytes, changed in a time-dependent manner. The phagocytic activity of the macrophages measured as the phagocytosis of microbeads and measured as the esterase activity also showed a strong time-dependent increase during the acute phase up to 21 days after irradiation. The contents of surfactant phospholipids (SF) and sphingomyelin (SPH; as a parameter for cell death) were quantified by HPLC. Both were significantly changed between day 2 and 21 after irradiation. Three BALs of a patient with idiopathic interstitial pneumonitis, who had received an allogenic bone marrow graft after total body irradiation with 10 Gy, showed similar effects in the cellular and surfactant parameters. These data indicate that there are positive interactions between the number of different BAL cells, macrophage activity, and SF and SPH content in the preclinical model of the mouse as well as in the clinical situation after lung irradiation. 30 refs., 7 figs., 3 tabs.

Steinberg, F.; Rehn, B.; Kraus, R.; Quabeck, K.; Bruch, J.; Beelen, D.W.; Schaefer, U.W.; Streffer, C. (Univ. Clinics, Essen (Germany))

1992-07-01T23:59:59.000Z

129

The second and third NGNP advanced gas reactor fuel irradiation experiments  

SciTech Connect (OSTI)

The United States Dept. of Energy's Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is currently scheduled to irradiate a total of five low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The irradiations are being accomplished to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas cooled reactors. The experiments will each consist of at least six separate capsules, and will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The effluent sweep gas will also have 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 completed a very successful irradiation in early November 2009. The second experiment (AGR-2) started irradiation in June 2010, and the third and fourth experiments have been combined into a single larger irradiation (AGR-3/4) that is currently being assembled. The design and status of the second through fourth experiments as well as the irradiation results of the second experiment to date are discussed. (authors)

Grover, S. B.; Petti, D. A. [Idaho National Laboratory, 2525 N. Fremont Ave., Idaho Falls, ID 83415 (United States)

2012-07-01T23:59:59.000Z

130

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

SciTech Connect (OSTI)

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).

J. L. Smith

2001-01-01T23:59:59.000Z

131

Test plan for N2 HEPA filters assembly shop stock used on PFP E4 exhaust system  

SciTech Connect (OSTI)

At Plutonium Finishing Plant (PFP) and Plutonium Reclamation Facility (PRF) Self-contained HEPA filters, encased in wooden frames and boxes, are installed in the E4 Exhaust Ventilation System to provide confinement of radioactive releases to the environment and confinement of radioactive contamination within designated zones inside the facility. Recently during the routine testing in-leakage was discovered downstream of the Self-contained HEPA filters boxes. This Test Plan describes the approach to conduct investigation of the root causes for the in-leakage of HEPA filters.

DICK, J.D.

1999-09-01T23:59:59.000Z

132

Effects of chronic, low-intensity gamma irradiation on the gonadotropic response of the ovaries and testes of the albino mouse  

E-Print Network [OSTI]

i senor icsl Cells, of Ter=s osrtiel fuli' illr, ent c f tho recuire e ct for tne oe ree of i ". ST::R GF SCI-~;CE :~u ust, , l?tO J sjcr Subject: I'olo y (ioolo y) FFFECTS OF CHRONIC, LOvk-INTENSITY Ghhhh& IRRaDI&TION ON THE, GONADOTROPIC... made re. . ardin =" th? effect of irradi?tion on endocrin . secret' on by the testes. abbott (195') reoorted that ardro?enic caoacity was rot reduced as juc:, ed by the wei;hts of accessory sex or-ans in the course of twerty-five weeks after...

Hunter, Jerry Don

1960-01-01T23:59:59.000Z

133

Seal assembly  

DOE Patents [OSTI]

A seal assembly that seals a gap formed by a groove comprises a seal body, a biasing element, and a connection that connects the seal body to the biasing element to form the seal assembly. The seal assembly further comprises a concave-shaped center section and convex-shaped contact portions at each end of the seal body. The biasing element is formed from an elastic material and comprises a convex-shaped center section and concave-shaped biasing zones that are opposed to the convex-shaped contact portions. The biasing element is adapted to be compressed to change a width of the seal assembly from a first width to a second width that is smaller than the first width. In the compressed state, the seal assembly can be disposed in the groove. After release of the compressing force, the seal assembly expands. The contact portions will move toward a surface of the groove and the biasing zones will move into contact with another surface of the groove. The biasing zones will bias the contact portions of the seal body against the surface of the groove.

Johnson, Roger Neal (Hagaman, NY); Longfritz, William David (Fonda, NY)

2001-01-01T23:59:59.000Z

134

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

SciTech Connect (OSTI)

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.).

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

2013-11-01T23:59:59.000Z

135

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

SciTech Connect (OSTI)

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.

Ellis, Ronald James [ORNL; Rapp, Juergen [ORNL

2014-01-01T23:59:59.000Z

136

DISSOLUTION OF IRRADIATED MURR FUEL ASSEMBLIES  

SciTech Connect (OSTI)

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.

Kyser, E.

2010-06-17T23:59:59.000Z

137

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)

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.

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

2009-01-01T23:59:59.000Z

138

Irradiation Creep in Graphite  

SciTech Connect (OSTI)

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.

Ubic, Rick; Butt, Darryl; Windes, William

2014-03-13T23:59:59.000Z

139

Latch assembly  

DOE Patents [OSTI]

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.

Frederickson, James R. (Richland, WA); Harper, William H. (Richland, WA); Perez, Raymond (Lynnwood, WA)

1986-01-01T23:59:59.000Z

140

Latch assembly  

DOE Patents [OSTI]

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.

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

1984-08-17T23:59:59.000Z

Note: This page contains sample records for the topic "test assembly irradiation" from the National Library of EnergyBeta (NLEBeta).
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141

Viral Assembly | EMSL  

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

Assembly Viral Assembly Current HIV research moves forward with help from EMSL HIV-1 CA protein assemblies are amenable to structural studies: This transmission electron...

142

Cutter assembly  

SciTech Connect (OSTI)

A drill bit with multiple fluid jet cutting nozzles designed so that the drill bit workface including the cutters is a separate piece from the drill bit body that houses the fluid jet nozzle orifice mounts. The cutter assembly protects the nozzle housing from rapid wear and it can be easily removed from the nozzle housing without disturbing or removing any of the nozzle orifice mounts.

O'Hanlon, T. A.

1985-09-10T23:59:59.000Z

143

Dump assembly  

DOE Patents [OSTI]

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.

Goldmann, Louis H. (Benton City, WA)

1986-01-01T23:59:59.000Z

144

Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment  

SciTech Connect (OSTI)

This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations.

MH Lane

2006-02-15T23:59:59.000Z

145

IRRADIATION PERFORMANCE OF U-Mo MONOLITHIC FUEL  

SciTech Connect (OSTI)

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.

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

2014-04-01T23:59:59.000Z

146

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

SciTech Connect (OSTI)

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.

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

2011-10-01T23:59:59.000Z

147

Shingle assembly  

DOE Patents [OSTI]

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.

Dinwoodie, Thomas L.

2007-02-20T23:59:59.000Z

148

Pushrod assembly  

DOE Patents [OSTI]

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.

Potter, Jerry D. (Kennewick, WA)

1987-01-01T23:59:59.000Z

149

Dump assembly  

DOE Patents [OSTI]

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.

Goldmann, L.H.

1984-12-06T23:59:59.000Z

150

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)

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.

Jeremy Gwin and Douglas Frenette

2010-04-08T23:59:59.000Z

151

The MARVEL assembly for neutron multiplication  

SciTech Connect (OSTI)

A new multiplying test assembly is under development at Idaho National Laboratory to support research, validation, evaluation, and learning. The item is comprised of three stacked, highly-enriched uranium (HEU) cylinders, each 11.4 cm in diameter and having a combined height of up to 11.7 cm. The combined mass of all three cylinders is 20.3 kg of HEU. Calculations for the bare configuration of the assembly indicate a multiplication level of >3.5 (keff=0.72). Reflected configurations of the assembly, using either polyethylene or tungsten, are possible and have the capability of raising the assembly's multiplication level to greater than 10. This paper describes simulations performed to assess the assembly's multiplication level under different conditions and describes the resources available at INL to support the use of these materials. We also describe some preliminary calculations and test activities using the assembly to study neutron multiplication.

David L. Chichester; Mathew T. Kinlaw

2013-10-01T23:59:59.000Z

152

Micromanifold assembly  

DOE Patents [OSTI]

A micromanifold for connecting external capillaries to the inlet and/or outlet ports of a microfluidic device can employ a ferrule/capillary assembly that includes: (a) a ferrule comprising an elongated member and having a bore traversing from a proximal end to a distal end of the member, wherein the bore has an inner surface and wherein the distal end of the ferrule has a tapered, threaded exterior surface, and (b) a capillary that is positioned within the bore wherein the capillary's outer surface is in direct contact with the bore's inner surface. No mating sleeve is required for the one-piece ferrule. Alternatively, the capillaries can be bonded to channels that traverse the manifold and therefore obviate the need for a ferrule.

Renzi, Ronald F. (Tracy, CA); Ferko, Scott M. (Livermore, CA)

2012-04-24T23:59:59.000Z

153

Micromanifold assembly  

SciTech Connect (OSTI)

A micromanifold for connecting external capillaries to the inlet and/or outlet ports of a microfluidic device can employ a ferrule/capillary assembly that includes: (a) a ferrule comprising an elongated member and having a bore traversing from a proximal end to a distal end of the member, wherein the bore has an inner surface and wherein the distal end of the ferrule has a tapered, threaded exterior surface, and (b) a capillary that is positioned within the bore wherein the capillary's outer surface is in direct contact with the bore's inner surface. No mating sleeve is required for the one-piece ferrule. Alternatively, the capillaries can be bonded to channels that traverse the manifold and therefore obviate the need for a ferrule.

Renzi, Ronald F. (Tracey, CA); Ferko, Scott (Livermore, CA)

2009-06-30T23:59:59.000Z

154

Flexible Assembly Solar Technology  

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

Assembly Solar Technology BrightSource DE-EE0005792 | February 15, 2013 | Toister * The proposed assembly process is based on small, cost-effective assembly cells (to be designed...

155

Self assembling magnetic tiles  

E-Print Network [OSTI]

Self assembly is an emerging technology in the field of manufacturing. Inspired by nature's ability to self assembly proteins from amino acids, this thesis attempts to demonstrate self assembly on the macro-scale. The ...

Rabl, Jessica A. (Jessica Ann)

2006-01-01T23:59:59.000Z

156

IRRADIATION EXPERIMENTS &  

E-Print Network [OSTI]

IRRADIATION EXPERIMENTS & FACILITIES AT BNL: BLIP & NSLS II Peter Wanderer Superconducting Magnet). Current user: LBNE ­ materials for Project X. · Long Baseline Neutrino Experiment ­ Abandoned gold mine

McDonald, Kirk

157

Sequence Assembly Validation by Restriction Digest Fingerprint  

E-Print Network [OSTI]

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

Rouchka, Eric

158

Inlet nozzle assembly  

DOE Patents [OSTI]

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.

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

1985-09-09T23:59:59.000Z

159

Inlet nozzle assembly  

DOE Patents [OSTI]

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.

Christiansen, David W. (Kennewick, WA); Karnesky, Richard A. (Richland, WA); Precechtel, Donald R. (Richland, WA); Smith, Bob G. (Richland, WA); Knight, Ronald C. (Richland, WA)

1987-01-01T23:59:59.000Z

160

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)

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.

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

2002-06-01T23:59:59.000Z

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


161

Installation and Final Testing of an On-Line, Multi-Spectrometer Fission Product Monitoring System (FPMS) to Support Advanced Gas Reactor (AGR) Fuel Testing and Qualification in the Advanced Test Reactor  

SciTech Connect (OSTI)

The US Department of Energy (DOE) is initiating tests of reactor fuel for use in an Advanced Gas Reactor (AGR). The AGR will use helium coolant, a low-power-density ceramic core, and coated-particle fuel. A series of eight (8) fuel irradiation tests are planned for the Idaho National Laboratory’s (INL’s) Advanced Test Reactor (ATR). One important measure of fuel performance in these tests is quantification of the fission gas releases over the nominal 2-year duration of each irradiation experiment. This test objective will be met using the AGR Fission Product Monitoring System (FPMS) which includes seven (7) on-line detection stations viewing each of the six test capsule effluent lines (plus one spare). Each station incorporates both a heavily-shielded high-purity germanium (HPGe) gamma-ray spectrometer for quantification of the isotopic releases, and a NaI(Tl) scintillation detector to monitor the total count rate and identify the timing of the releases. The AGR-1 experiment will begin irradiation after October 1, 2006. To support this experiment, the FPMS has been completely assembled, tested, and calibrated in a laboratory at the INL, and then reassembled and tested in its final location in the ATR reactor basement. This paper presents the details of the equipment performance, the control and acquisition software, the test plan for the irradiation monitoring, and the installation in the ATR basement. Preliminary on-line data may be available by the Conference date.

J. K. Hartwell; D. M. Scates; M. W. Drigert; J. B. Walter

2006-10-01T23:59:59.000Z

162

Tilt assembly for tracking solar collector assembly  

DOE Patents [OSTI]

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.

Almy, Charles; Peurach, John; Sandler, Reuben

2012-01-24T23:59:59.000Z

163

Horizontal modular dry irradiated fuel storage system  

DOE Patents [OSTI]

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).

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

1988-01-01T23:59:59.000Z

164

Experimental Investigation of Microbially Induced Corrosion of Test Samples and Effect of Self-Assembled Hydrophobic Monolayers. Exposure of Test Samples to Continuous Microbial Cultures, Chemical Analysis, and Biochemical Studies  

SciTech Connect (OSTI)

The study of biocorrosion of aluminum and beryllium samples were performed under conditions of continuous fermentation of thermophilic anaerobic microorganisms of different groups. This allowed us to examine the effect of various types of metabolic reactions of reduction-oxidation proceeding at different pH and temperatures under highly reduced conditions on aluminum and beryllium corrosion and effect of self-assembled hydrophobic monolayers.

Laurinavichius, K.S.

1998-09-30T23:59:59.000Z

165

Self assembly of complex structures.  

E-Print Network [OSTI]

??The state of the art in artificial micro self assembly concepts are reviewed. The history of assembly is presented with a comparison to macro assembly,… (more)

Nellis, Michael

2007-01-01T23:59:59.000Z

166

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)

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.

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

2009-11-04T23:59:59.000Z

167

Finding of No Significant Impact and Final Environmental Assessment for the Future Location of Heat Source/Radioisotope Power System Assembly and Testing and Operations Currently Located at the Mound Site  

SciTech Connect (OSTI)

The U.S. Department of Energy (the Department) has completed an Environmental Assessment for the Future Location of the Heat Source/Radioisotope Power System Assembly and Test. Operations Currently Located at the Mound Site. Based on the analysis in the environmental assessment, the Department has determined that the proposed action, the relocation of the Department's heat source and radioisotope power system operations, does not constitute a major Federal action significantly affecting the quality of the human environment within the meaning of the ''National Environmental Policy Act'' of 1969 (NEPA). Therefore, the preparation of an Environmental Impact Statement is not required, and the Department is issuing this Finding of No Significant Impact (FONSI).

N /A

2002-08-30T23:59:59.000Z

168

Membrane module assembly  

DOE Patents [OSTI]

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.

Kaschemekat, J.

1994-03-15T23:59:59.000Z

169

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)

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.

Not Available

1989-08-01T23:59:59.000Z

170

Application for approval for construction of the Fueled Clad Fabrication System, the Radioisotope Power Systems Facility, and the Fuel Assembly Area  

SciTech Connect (OSTI)

The following ''Application for Approval of Construction'' is being submitted by the US Department of Energy-Richland Operations Office, pursuant to 40 CFR 61.07, for three new sources of airborne 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 canceled 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 and 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. 2 refs., 16 figs., 12 tabs.

Not Available

1989-08-01T23:59:59.000Z

171

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)

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.

Not Available

1989-08-01T23:59:59.000Z

172

RERTR-6 Irradiation Summary Report  

SciTech Connect (OSTI)

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-6 was designed to evaluate several modified fuel designs that were proposed to address the possibility of breakaway swelling due to porosity within the (U. Mo) Al interaction product observed in the full-size plate tests performed in Russia and France1. The following report summarizes the life of the RERTR-6 experiment through end of irradiation, including as-run neutronic analyses, thermal analyses and hydraulic testing results.

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

2011-12-01T23:59:59.000Z

173

Safer Food with Irradiation  

E-Print Network [OSTI]

This publication answers questions about food irradiation and how it helps prevent foodborne illnesses. Included are explanations of how irradiation works and its benefits. Irradiation is a safe method of preserving food quality and ensuring its...

Thompson, Britta; Vestal, Andy; Van Laanen, Peggy

2003-01-21T23:59:59.000Z

174

Decline in Tested and Self-Reported Cognitive Functioning After Prophylactic Cranial Irradiation for Lung Cancer: Pooled Secondary Analysis of Radiation Therapy Oncology Group Randomized Trials 0212 and 0214  

SciTech Connect (OSTI)

Purpose: To assess the impact of prophylactic cranial irradiation (PCI) on self-reported cognitive functioning (SRCF), a functional scale on the European Organization for Research and Treatment of Cancer Core Quality of Life Questionnaire (EORTC QLQ-C30). Methods and Materials: Radiation Therapy Oncology Group (RTOG) protocol 0214 randomized patients with locally advanced non-small cell lung cancer to PCI or observation; RTOG 0212 randomized patients with limited-disease small cell lung cancer to high- or standard-dose PCI. In both trials, Hopkins Verbal Learning Test (HVLT)-Recall and -Delayed Recall and SRCF were assessed at baseline (after locoregional therapy but before PCI or observation) and at 6 and 12 months. Patients developing brain relapse before follow-up evaluation were excluded. Decline was defined using the reliable change index method and correlated with receipt of PCI versus observation using logistic regression modeling. Fisher's exact test correlated decline in SRCF with HVLT decline. Results: Of the eligible patients pooled from RTOG 0212 and RTOG 0214, 410 (93%) receiving PCI and 173 (96%) undergoing observation completed baseline HVLT or EORTC QLQ-C30 testing and were included in this analysis. Prophylactic cranial irradiation was associated with a higher risk of decline in SRCF at 6 months (odds ratio 3.60, 95% confidence interval 2.34-6.37, P<.0001) and 12 months (odds ratio 3.44, 95% confidence interval 1.84-6.44, P<.0001). Decline on HVLT-Recall at 6 and 12 months was also associated with PCI (P=.002 and P=.002, respectively) but was not closely correlated with decline in SRCF at the same time points (P=.05 and P=.86, respectively). Conclusions: In lung cancer patients who do not develop brain relapse, PCI is associated with decline in HVLT-tested and self-reported cognitive functioning. Decline in HVLT and decline in SRCF are not closely correlated, suggesting that they may represent distinct elements of the cognitive spectrum.

Gondi, Vinai, E-mail: vgondi@chicagocancer.org [Central Dupage Hospital Cancer Center, Warrenville, Illinois (United States) [Central Dupage Hospital Cancer Center, Warrenville, Illinois (United States); University of Wisconsin Comprehensive Cancer Center, Madison, Wisconsin (United States); Paulus, Rebecca [Radiation Therapy Oncology Group Statistical Center, Philadelphia, Pennsylvania (United States)] [Radiation Therapy Oncology Group Statistical Center, Philadelphia, Pennsylvania (United States); Bruner, Deborah W. [Nell Hodgson Woodfull School of Nursing, Emory University, Atlanta, Georgia (United States)] [Nell Hodgson Woodfull School of Nursing, Emory University, Atlanta, Georgia (United States); Meyers, Christina A. [University of Texas MD Anderson Cancer Center, Houston, Texas (United States)] [University of Texas MD Anderson Cancer Center, Houston, Texas (United States); Gore, Elizabeth M. [Medical College of Wisconsin, Milwaukee, Wisconsin (United States)] [Medical College of Wisconsin, Milwaukee, Wisconsin (United States); Wolfson, Aaron [University of Miami School of Medicine, Miami, Florida (United States)] [University of Miami School of Medicine, Miami, Florida (United States); Werner-Wasik, Maria [Thomas Jefferson University Hospital, Philadelphia, Pennsylvania (United States)] [Thomas Jefferson University Hospital, Philadelphia, Pennsylvania (United States); Sun, Alexander Y. [Princess Margaret Hospital, Toronto, ON (Canada)] [Princess Margaret Hospital, Toronto, ON (Canada); Choy, Hak [University of Texas Southwestern Moncreif Cancer Center, Fort Worth, Texas (United States)] [University of Texas Southwestern Moncreif Cancer Center, Fort Worth, Texas (United States); Movsas, Benjamin [Henry Ford Health System, Detroit, Michigan (United States)] [Henry Ford Health System, Detroit, Michigan (United States)

2013-07-15T23:59:59.000Z

175

Proton irradiation effect on SCDs  

E-Print Network [OSTI]

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.

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

2014-04-19T23:59:59.000Z

176

Self-Assembling Circuits Plasticity in Self-Assembly: Templating  

E-Print Network [OSTI]

Self-Assembling Circuits Plasticity in Self-Assembly: Templating Generates Functionally Different on a combination of self-assembly and external guidance. We demonstrate the self-assembly of mm-sized components of a template--that is, by the geometry of the volume in which the self-assembly proceeds. The components carry

Prentiss, Mara

177

Flexible Assembly Solar Technology  

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

2007-2010 BrightSource Energy, Inc. All rights reserved. 1 Flexible Assembly Solar Technology Binyamin Koretz Director, Strategic Planning & IP 2 Proprietary &...

178

Composite turbine bucket assembly  

DOE Patents [OSTI]

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.

Liotta, Gary Charles; Garcia-Crespo, Andres

2014-05-20T23:59:59.000Z

179

RERTR-13 Irradiation Summary Report  

SciTech Connect (OSTI)

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.

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

2012-09-01T23:59:59.000Z

180

Laser bottom hole assembly  

DOE Patents [OSTI]

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.

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

2014-01-14T23:59:59.000Z

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


181

Turbine disc sealing assembly  

DOE Patents [OSTI]

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.

Diakunchak, Ihor S.

2013-03-05T23:59:59.000Z

182

NDA safeguards techniques for LMFBR assemblies  

SciTech Connect (OSTI)

The significant safeguards concerns for liquid-metal fast breeder reactors (LMFBRs), and for the LMFBR fuel handling systems are the accountability, surveillance, and identification of fuel and blanket assemblies. The introduction of fuel assemblies with a high content of Pu into the receiving and shipping areas of the LMFBR fuel cycle does allow a more direct near-real-time assay profile of the disposition of Pu. Isotope correlations and neutron assay methods have been investigated and implemented for determining plutonium and burnup in fresh and spent LMFBR fuel assemblies. The methods are based on active and passive neutron coincidence counting (NCC) techniques. Preliminary studies on neutron yield rates from the spontaneous fission of plutonium and curium isotopes have indicated that the NCC system is a most effective measure in the verification of nuclear material flow in assembly form for the entire reactor fuel handling cycle, i.e., from the fresh- to the spent-fuel stage. A consequence of the high plutonium concentration level throughout the fuel irradiation period in an LMFBR, is that the spontaneous fission neutron yield from the 242-curium and 244-curium does not dominate the spontaneous fission neutron yield from the plutonium isotopes in the spent fuel stage.

Persiani, P.J.; Gundy, M.L.

1982-08-01T23:59:59.000Z

183

Constrained space camera assembly  

DOE Patents [OSTI]

A constrained space camera assembly which is intended to be lowered through a hole into a tank, a borehole or another cavity. 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.

Heckendorn, Frank M. (Aiken, SC); Anderson, Erin K. (Augusta, GA); Robinson, Casandra W. (Trenton, SC); Haynes, Harriet B. (Aiken, SC)

1999-01-01T23:59:59.000Z

184

Superconducting radiofrequency window assembly  

DOE Patents [OSTI]

The present invention is a superconducting radiofrequency window assembly for use in an electron beam accelerator. The srf window assembly has a superconducting metal-ceramic design. The srf window assembly comprises a superconducting frame, a ceramic plate having a superconducting metallized area, and a superconducting eyelet for sealing plate into frame. The plate is brazed to eyelet which is then electron beam welded to frame. 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. 11 figs.

Phillips, H.L.; Elliott, T.S.

1997-03-11T23:59:59.000Z

185

Superconductive radiofrequency window assembly  

DOE Patents [OSTI]

The present invention is a superconducting radiofrequency window assembly for use in an electron beam accelerator. The SRF window assembly has a superconducting metal-ceramic design. The SRF window assembly comprises a superconducting frame, a ceramic plate having a superconducting metallized area, and a superconducting eyelet for sealing plate into frame. The plate is brazed to eyelet which is then electron beam welded to frame. 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. 11 figs.

Phillips, H.L.; Elliott, T.S.

1998-05-19T23:59:59.000Z

186

RERTR-12 Insertion 2 Irradiation Summary Report  

SciTech Connect (OSTI)

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-12 was designed to provide comprehensive information on the performance of uranium-molybdenum (U-Mo) based monolithic fuels for research reactor applications.1 RERTR-12 insertion 2 includes the capsules irradiated during the last three irradiation cycles. These capsules include Z, Y1, Y2 and Y3 type capsules. The following report summarizes the life of the RERTR-12 insertion 2 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

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

2012-09-01T23:59:59.000Z

187

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

SciTech Connect (OSTI)

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.

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

2000-08-07T23:59:59.000Z

188

Automated Assembly Using Feature Localization  

E-Print Network [OSTI]

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 ...

Gordon, Steven Jeffrey

1986-12-01T23:59:59.000Z

189

Knowledge based process planning system for electronic assembly  

E-Print Network [OSTI]

. In both cases the connection is made permanent by soldering the leads. The two methods require different component geometries, assembly machines and soldering methods. Of these, surface mount method is a relatively new technology and is becoming... of solder paste or adhesives, cleaning processes, material handling, inspection and testing Operations [11, 12, 13]. Assembly Operations are concerned with the delivery of components to a specified location on the board. Assembly Operations can...

Sabapathy, Arvindh

2012-06-07T23:59:59.000Z

190

GTL-1 Irradiation Summary Report  

SciTech Connect (OSTI)

The primary objective of the Gas Test Loop (GTL-1) miniplate experiment is to confirm acceptable performance of high-density (i.e., 4.8 g-U/cm3) U3Si2/Al dispersion fuel plates clad in Al-6061 and irradiated under the relatively aggressive Booster Fast Flux Loop (BFFL) booster fuel conditions, namely a peak plate surface heat flux of 450 W/cm2. As secondary objectives, several design and fabrication variations were included in the test matrix that may have the potential to improve the high-heat flux, high-temperature performance of the base fuel plate design.1, 2 The following report summarizes the life of the GTL-1 experiment through end of irradiation, including as-run neutronic analysis, thermal analysis and hydraulic testing results.

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

2012-01-01T23:59:59.000Z

191

Nuclear reactor control assembly  

SciTech Connect (OSTI)

This patent describes an assembly for providing global power control in a nuclear reactor having the core split into two halves. It comprises a disk assembly formed from at least two disks each machined with an identical surface hole pattern such that rotation of one disk relative to the other causes the hole pattern to open or close, the disk assembly being positioned substantially at the longitudinal center of and coaxial with the core halves; and means for rotating at least one of the disks relative to the other.

Negron, S.B.

1991-06-11T23:59:59.000Z

192

DC source assemblies  

DOE Patents [OSTI]

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.

Campbell, Jeremy B; Newson, Steve

2013-02-26T23:59:59.000Z

193

Assembly flow simulation of a radar  

SciTech Connect (OSTI)

A discrete event simulation model has been developed to predict the assembly flow time of a new radar product. The simulation was the key tool employed to identify flow constraints. The radar, production facility, and equipment complement were designed, arranged, and selected to provide the most manufacturable assembly possible. A goal was to reduce the assembly and testing cycle time from twenty-six weeks to six weeks. A computer software simulation package (SLAM II) was utilized as the foundation a for simulating the assembly flow time. FORTRAN subroutines were incorporated into the software to deal with unique flow circumstances that were not accommodated by the software. Detailed information relating to the assembly operations was provided by a team selected from the engineering, manufacturing management, inspection, and production assembly staff. The simulation verified that it would be possible to achieve the cycle time goal of six weeks. Equipment and manpower constraints were identified during the simulation process and adjusted as required to achieve the flow with a given monthly production requirement. The simulation is being maintained as a planning tool to be used to identify constraints in the event that monthly output is increased. ``What-if`` studies have been conducted to identify the cost of reducing constraints caused by increases in output requirement.

Rutherford, W.C.; Biggs, P.M.

1993-10-01T23:59:59.000Z

194

Hg System Assembly and Testing Status  

E-Print Network [OSTI]

OF ENERGY Current Status / Next Steps · Populate secondary box ports · Complete plumbing inside secondary #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

McDonald, Kirk

195

Hg System Assembly and Testing Status  

E-Print Network [OSTI]

#12;2VRVS Meeting 8 Nov 2006 OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY Hg Drain & Spill Ports Completed Hg drain from cylinder Hg leak sump with float switch Hg spill extraction port #12;3VRVS Meeting 8 Nov 2006 OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY Local Hg Vapor Filters · Hg

McDonald, Kirk

196

Flexible Assembly Solar Technology  

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

platform that can be used for rapid assembly and installation of heliostats at a solar power tower plant. BrightSource will design and deploy a FAST prototype under this...

197

Steam separator latch assembly  

DOE Patents [OSTI]

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.

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

1994-02-01T23:59:59.000Z

198

Spring bypass assembly. [LMFBR  

SciTech Connect (OSTI)

Pipe clamp comprises two substantially semicircular rim halves biased toward each other by spring assemblies. Adjustable stop means 5 limit separation of the rim halves when the pipe expands.

Jablonski, H.; Roughgarden, J.D.

1982-06-02T23:59:59.000Z

199

Magnetic assisted statistical assembly  

E-Print Network [OSTI]

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 ...

Cheng, Diana I

2008-01-01T23:59:59.000Z

200

Steam separator latch assembly  

DOE Patents [OSTI]

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.

Challberg, Roy C. (Livermore, CA); Kobsa, Irvin R. (San Jose, CA)

1994-01-01T23:59:59.000Z

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


201

Recuperator assembly and procedures  

DOE Patents [OSTI]

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.

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

2008-08-26T23:59:59.000Z

202

Recuperator assembly and procedures  

DOE Patents [OSTI]

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.

Kang, Yungmo; McKeirnan Jr., Robert D.

2006-06-27T23:59:59.000Z

203

Derivation of criticality safety benchmarks from ZPR fast critical assemblies  

SciTech Connect (OSTI)

Scores of critical assemblies were constructed, over a period of about three decades, at the Argonne National Laboratory ZPR-3, ZPR-6, ZPR-9, and ZPPR fast critical assembly facilities. Most of the assemblies were mockups of various liquid-metal fast breeder reactor designs. These tended to be complex, containing, for example, mockups of control rods and control rod positions. Some assemblies, however, were `physics benchmarks`. These relatively `clean` assemblies had uniform compositions and simple geometry and were designed to test fast reactor physics data and methods. Assemblies in this last category are well suited to form the basis for new criticality safety benchmarks. The purpose of this paper is to present an overview of some of these benchmark candidates and to describe the strategy being used to create the benchmarks.

Schaefer, R.W.; McKnight, R.D.

1997-09-01T23:59:59.000Z

204

Nuclear fuel scoping: implementation of a four node per assembly algorithm as the neutronic module for microscope  

E-Print Network [OSTI]

, . APPENDIX B INPUT FOR TEST CASES . SAMPLE INPUT FOR ONE NODE PER ASSEMBLY. . . . SAMPLE INPUT FOR FOUR NODES PER ASSEMBLY. . . APPENDIX C OUTPUT FOR TEST CASES SAMPLE OUTPUT FOR ONE NODE PER ASSEMBLY. . . SAMPLE OUTPUT FOR FOUR NODES PER ASSEMBLY... on which to simulate and test out some of his ideas or intuitions regarding ways of enhancing fuel procurement and bumup. It allows him to easily and cheaply experiment with the decision variables associated with fuel procurement and burnup, and to see...

Shofolu, Babatunde Olayemi

2012-06-07T23:59:59.000Z

205

AGC-1 Post Irradiation Examination Status  

SciTech Connect (OSTI)

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 six AGC capsules in the experiment will be irradiated in the Advanced Test Reactor (ATR), disassembled in the Hot Fuel Examination Facility (HFEF), and examined at the INL Research Center (IRC) or Oak Ridge National Laboratory (ORNL). This is the first in a series of status reports on the progress of the AGC experiment. As the first capsule, AGC1 was irradiated from September 2009 to January 2011 to a maximum dose level of 6-7 dpa. The capsule was removed from ATR and transferred to the HFEF in April 2011 where the capsule was disassembled and test specimens extracted from the capsules. The first irradiated samples from AGC1 were shipped to the IRC in July 2011and initial post irradiation examination (PIE) activities were begun on the first 37 samples received. PIE activities continue for the remainder of the AGC1 specimen as they are received at the IRC.

David Swank

2011-09-01T23:59:59.000Z

206

Nuclear cask testing films misleading and misused  

SciTech Connect (OSTI)

In 1977 and 1978, Sandia National Laboratories, located in Albuquerque, New Mexico, and operated for the US Department of Energy (DOE), filmed a series of crash and fire tests performed on three casks designed to transport irradiated nuclear fuel assemblies. While the tests were performed to assess the applicability of scale and computer modeling techniques to actual accidents, films of them were quickly pressed into service by the DOE and nuclear utilities as ``proof`` to the public of the safety of the casks. In the public debate over the safety of irradiated nuclear fuel transportation, the films have served as the mainstay for the nuclear industry. Although the scripts of all the films were reviewed by USDOE officials before production, they contain numerous misleading concepts and images, and omit significant facts. The shorter versions eliminated qualifying statements contained in the longer version, and created false impressions. This paper discusses factors which cast doubt on the veracity of the films and the results of the tests.

Audin, L. [Audin (Lindsay), Ossining, NY (United States)

1991-10-01T23:59:59.000Z

207

Nuclear cask testing films misleading and misused  

SciTech Connect (OSTI)

In 1977 and 1978, Sandia National Laboratories, located in Albuquerque, New Mexico, and operated for the US Department of Energy (DOE), filmed a series of crash and fire tests performed on three casks designed to transport irradiated nuclear fuel assemblies. While the tests were performed to assess the applicability of scale and computer modeling techniques to actual accidents, films of them were quickly pressed into service by the DOE and nuclear utilities as proof'' to the public of the safety of the casks. In the public debate over the safety of irradiated nuclear fuel transportation, the films have served as the mainstay for the nuclear industry. Although the scripts of all the films were reviewed by USDOE officials before production, they contain numerous misleading concepts and images, and omit significant facts. The shorter versions eliminated qualifying statements contained in the longer version, and created false impressions. This paper discusses factors which cast doubt on the veracity of the films and the results of the tests.

Audin, L. (Audin (Lindsay), Ossining, NY (United States))

1991-10-01T23:59:59.000Z

208

Constrained space camera assembly  

DOE Patents [OSTI]

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.

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

1999-05-11T23:59:59.000Z

209

Superconducting radiofrequency window assembly  

DOE Patents [OSTI]

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.

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

1997-01-01T23:59:59.000Z

210

Superconductive radiofrequency window assembly  

DOE Patents [OSTI]

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.

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

1998-01-01T23:59:59.000Z

211

Power module assembly  

DOE Patents [OSTI]

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.

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

2011-11-15T23:59:59.000Z

212

Solar central receiver heliostat reflector assembly  

DOE Patents [OSTI]

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.

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

1980-01-01T23:59:59.000Z

213

Lightweight, self-ballasting photovoltaic roofing assembly  

DOE Patents [OSTI]

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.

Dinwoodie, Thomas L. (Berkeley, CA)

1998-01-01T23:59:59.000Z

214

Lightweight, self-ballasting photovoltaic roofing assembly  

DOE Patents [OSTI]

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.

Dinwoodie, T.L.

1998-05-05T23:59:59.000Z

215

Lightweight, self-ballasting photovoltaic roofing assembly  

DOE Patents [OSTI]

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.

Dinwoodie, Thomas L.

2006-02-28T23:59:59.000Z

216

Low inductance connector assembly  

DOE Patents [OSTI]

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.

Holbrook, Meghan Ann; Carlson, Douglas S

2013-07-09T23:59:59.000Z

217

Biofilm assembly | EMSL  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to someone6 M. Babzien, I.ProgramBig SolBiofilm assembly Biofilm assembly

218

E-Print Network 3.0 - assembly leakage unit Sample Search Results  

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

and Utilization 2 ERNEST ORLANDO LAWRENCE BERKELEY NATIONAL LABORATORY Summary: precision orifices was used for the leakage testing. The forces required to assemble and pull...

219

SELF-ASSEMBLY AND CONTROLLED ASSEMBLY OF NANOPARTICLES.  

E-Print Network [OSTI]

??This thesis describes an exploration of interactions between metal nanoparticles and new techniques for their assembly. In Chapter 2, the self-assembly of 300-nm diameter Au,… (more)

Dillenback, Lisa

2008-01-01T23:59:59.000Z

220

Corium protection assembly  

DOE Patents [OSTI]

A corium protection assembly includes a perforated base grid disposed below a pressure vessel containing a nuclear reactor core and spaced vertically above a containment vessel floor to define a sump therebetween. A plurality of layers of protective blocks are disposed on the grid for protecting the containment vessel floor from the corium.

Gou, Perng-Fei (Saratoga, CA); Townsend, Harold E. (Campbell, CA); Barbanti, Giancarlo (Sirtori, IT)

1994-01-01T23:59:59.000Z

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


221

Segmented stator assembly  

DOE Patents [OSTI]

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.

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

2013-04-02T23:59:59.000Z

222

Rotary shaft sealing assembly  

DOE Patents [OSTI]

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.

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

2010-09-21T23:59:59.000Z

223

Rotary shaft sealing assembly  

DOE Patents [OSTI]

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.

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

2013-08-13T23:59:59.000Z

224

Irradiation creep of vanadium-base alloys.  

SciTech Connect (OSTI)

A study of irradiation creep in vanadium-base alloys is underway with experiments in the Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR) in the US. Test specimens are thin-wall sealed tubes with internal pressure loading. The results from the initial ATR irradiation at low temperature (200-300 C) to a neutron damage level of 4.7 dpa show creep rates ranging from {approx}0 to 1.2 x 10{sup {minus}5}/dpa/MPa for a 500-kg heat of V-4Cr-4Ti alloy. These rates were generally lower than reported from a previous experiment in BR-10. Because both the attained neutron damage levels and the creep strains were low in the present study, however, these creep rates should be regarded as only preliminary. Substantially more testing is required before a data base on irradiation creep of vanadium alloys can be developed and used with confidence.

Tsai, H.; Matsui, H.; Billone, M. C.; Strain, R. V.; Smith, D. L.

1998-05-18T23:59:59.000Z

225

RERTR-7 Irradiation Summary Report  

SciTech Connect (OSTI)

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.

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

2011-12-01T23:59:59.000Z

226

REVIEW OF FAST FLUX TEST FACILITY (FFTF) FUEL EXPERIMENTS FOR STORAGE IN INTERIM STORAGE CASKS (ISC)  

SciTech Connect (OSTI)

Appendix H, Section H.3.3.10.11 of the Final Safety Analysis Report (FSAR), provides the limits to be observed for fueled components authorized for storage in the Fast Flux Test Facility (FFTF) spent fuel storage system. Currently, the authorization basis allows standard driver fuel assemblies (DFA), as described in the FSAR Chapter 17, Section 17.5.3.1, to be stored provided decay power per assembly is {le} 250 watts, post-irradiation time is four years minimum, average assembly burn-up is 150,000 MWD/MTHM maximum and the pre-irradiation enrichment is 29.3% maximum (per H.3.3.10.11). In addition, driver evaluation (DE), core characterizer assemblies (CCA), and run-to-cladding-breach (RTCB) assemblies are included based on their similarities to a standard DFA. Ident-69 pin containers with fuel pins from these DFAs can also be stored. Section H.3.3.10.11 states that fuel types outside the specification criteria above will be addressed on a case-by-case basis. There are many different types of fuel and blanket experiments that were irradiated in the FFTF which now require offload to the spent fuel storage system. Two reviews were completed for a portion of these special type fuel components to determine if placement into the Core Component Container (CCC)/Interim Storage Cask (ISC) would require any special considerations or changes to the authorization basis. Project mission priorities coupled with availability of resources and analysts prevented these evaluations from being completed as a single effort. Areas of review have included radiological accident release consequences, radiological shielding adequacy, criticality safety, thermal limits, confinement, and stress. The results of these reviews are available in WHC-SD-FF-RPT-005, Rev. 0 and 1, ''Review of FFTF Fuel Experiments for Storage at ISA'', (Reference I), which subsequently allowed a large portion of these components to be included in the authorization basis (Table H.3.3-21). The report also identified additional components and actions in Section 3.0 and Table 3 that require further evaluation. The purpose of this report is to evaluate another portion of the remaining inventory (i.e., delayed neutron signal fuel, blanket assemblies, highly enriched assemblies, newly loaded Ident-69 pin containers, and returned fuel) to ensure it can be safely off loaded to the FFTF spent fuel storage system.

CHASTAIN, S.A.

2005-10-24T23:59:59.000Z

227

Molecular Computing with DNA Self-Assembly  

E-Print Network [OSTI]

Molecular Computing with DNA Self-Assembly Urmi Majumder #12;Self-Assembly in Nature #12;Key to DNA for Molecular Computing with DNA Self-Assembly Compact: Small library of assembly primitives Complex: Capable in Tiling Assembly: vitroation tural DNA self-assembly has powerful echanisms for error correction

Reif, John H.

228

Low inductance busbar assembly  

DOE Patents [OSTI]

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.

Holbrook, Meghan Ann (Manhattan Beach, CA)

2010-09-21T23:59:59.000Z

229

Vacuum breaker valve assembly  

DOE Patents [OSTI]

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.

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

1999-04-27T23:59:59.000Z

230

Vacuum breaker valve assembly  

DOE Patents [OSTI]

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.

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

1999-04-27T23:59:59.000Z

231

Solution deposition assembly  

DOE Patents [OSTI]

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.

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

2014-01-21T23:59:59.000Z

232

Fuel nozzle assembly  

DOE Patents [OSTI]

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.

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)

2011-08-30T23:59:59.000Z

233

Turbine seal assembly  

DOE Patents [OSTI]

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.

Little, David A.

2013-04-16T23:59:59.000Z

234

Mechanical seal assembly  

DOE Patents [OSTI]

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.

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

2002-01-01T23:59:59.000Z

235

Mechanical seal assembly  

DOE Patents [OSTI]

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.

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

2001-01-01T23:59:59.000Z

236

Ingestion resistant seal assembly  

DOE Patents [OSTI]

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.

Little, David A. (Chuluota, FL)

2011-12-13T23:59:59.000Z

237

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

SciTech Connect (OSTI)

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.

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

1997-12-01T23:59:59.000Z

238

Pull rod assembly  

DOE Patents [OSTI]

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.

Cioletti, Olisse C. (Pittsburgh, PA)

1990-01-01T23:59:59.000Z

239

Development of a Safeguards Verification Method and Instrument to Detect Pin Diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies  

SciTech Connect (OSTI)

A technical safeguards challenge has remained for decades for the IAEA to identify possible diversion of nuclear fuel pins from Light Water Reactor (LWR) spent fuel assemblies. In fact, as modern nuclear power plants are pushed to higher power levels and longer fuel cycles, fuel failures (i.e., ''leakers'') as well as the corresponding fuel assembly repairs (i.e., ''reconstitutions'') are commonplace occurrences within the industry. Fuel vendors have performed hundreds of reconstitutions in the past two decades, thus, an evolved know-how and sophisticated tools exist to disassemble irradiated fuel assemblies and replace damaged pins with dummy stainless steel or other type rods. Various attempts have been made in the past two decades to develop a technology to identify a possible diversion of pin(s) and to determine whether some pins are missing or replaced with dummy or fresh fuel pins. However, to date, there are no safeguards instruments that can detect a possible pin diversion scenario to the requirements of the IAEA. The FORK detector system [1-2] can characterize spent fuel assemblies using operator declared data, but it is not sensitive enough to detect missing pins from spent fuel assemblies. Likewise, an emission computed tomography system [3] has been used to try to detect missing pins from a spent fuel assembly, which has shown some potential for identifying possible missing pins but this capability has not yet been fully demonstrated. The use of such a device in the future would not be envisaged, especially in an inexpensive, easy to handle setting for field applications. In this article, we describe a concept and ongoing research to help develop a new safeguards instrument for the detection of pin diversions in a PWR spent fuel assembly. The proposed instrument is based on one or more very thin radiation detectors that could be inserted within the guide tubes of a Pressurized Water Reactor (PWR) assembly. Ultimately, this work could lead to the development of a detector cluster and corresponding high-precision driving system to collect radiation signatures inside PWR spent fuel assemblies. The data obtained would provide the spatial distribution of the neutron and gamma flux fields within the spent fuel assembly, while the data analysis would be used to help identify missing or replaced pins. Monte Carlo simulations have been performed to help validate this concept using a realistic 17 x 17 PWR spent fuel assembly [4-5]. The initial results of this study show that neutron profile in the guide tubes, when obtained in the presence of missing pins, can be identifiably different from the profiles obtained without missing pins, Our latest simulations have focused upon a specific type of fission chamber that could be tested for this application.

Ham, Y S; Maldonado, G I; Burdo, J; He, T

2006-10-10T23:59:59.000Z

240

Very large assemblies: Optimizing for automatic generation of assembly sequences  

SciTech Connect (OSTI)

Sandia's Archimedes 3.0{copyright} Automated Assembly Analysis System has been applied successfully to several large industrial and weapon assemblies. These have included Sandia assemblies such as portions of the B61 bomb, and assemblies from external customers such as Cummins Engine Inc., Raytheon (formerly Hughes) Missile Systems and Sikorsky Aircraft. While Archimedes 3.0{copyright} represents the state-of-the-art in automated assembly planning software, applications of the software made prior to the technological advancements presented here showed several limitations of the system, and identified the need for extensive modifications to support practical analysis of assemblies with several hundred to a few thousand parts. It was believed that there was substantial potential for enhancing Archimedes 3.0{copyright} to routinely handle much larger models and/or to handle more modestly sized assemblies more efficiently. Such a mature assembly analysis capability was needed to support routine application to industrial assemblies that overstressed the system, such as full nuclear weapon assemblies or full-scale aerospace or military vehicles.

CALTON,TERRI L.

2000-02-01T23:59:59.000Z

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


241

Comminuting irradiated ferritic steel  

DOE Patents [OSTI]

Disclosed is a method of comminuting irradiated ferritic steel by placing the steel in a solution of a compound selected from the group consisting of sulfamic acid, bisulfate, and mixtures thereof. The ferritic steel is used as cladding on nuclear fuel rods or other irradiated components.

Bauer, Roger E. (Kennewick, WA); Straalsund, Jerry L. (Kennewick, WA); Chin, Bryan A. (Auburn, AL)

1985-01-01T23:59:59.000Z

242

EA-1210: Final Environmental Assessment  

Broader source: Energy.gov [DOE]

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

243

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

244

Evaluation of Neutron Irradiated Silicon Carbide and Silicon Carbide Composites  

SciTech Connect (OSTI)

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.

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

2007-03-26T23:59:59.000Z

245

Nanomechanical testing system  

DOE Patents [OSTI]

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.

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

2014-07-08T23:59:59.000Z

246

Removable feedwater sparger assembly  

DOE Patents [OSTI]

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.

Challberg, Roy C. (Livermore, CA)

1994-01-01T23:59:59.000Z

247

Precision Robotic Assembly Machine  

ScienceCinema (OSTI)

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.

None

2010-09-01T23:59:59.000Z

248

Removable feedwater sparger assembly  

DOE Patents [OSTI]

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.

Challberg, R.C.

1994-10-04T23:59:59.000Z

249

CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 8 AND 9 CRYSTAL RIVER UNIT 3  

SciTech Connect (OSTI)

The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 8 and 9 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.

Michael L. Wilson

2001-02-08T23:59:59.000Z

250

Gas separation membrane module assembly  

DOE Patents [OSTI]

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.

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

2009-03-31T23:59:59.000Z

251

KJRR-FAI Hydraulic Flow Testing Input Package  

SciTech Connect (OSTI)

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.

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

2013-12-01T23:59:59.000Z

252

Irradiation facilities at the Los Alamos Meson Physics Facility  

SciTech Connect (OSTI)

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.

Sandberg, V.

1990-01-01T23:59:59.000Z

253

Next-generation transcriptome assembly  

SciTech Connect (OSTI)

Transcriptomics studies often rely on partial reference transcriptomes that fail to capture the full catalog of transcripts and their variations. Recent advances in sequencing technologies and assembly algorithms have facilitated the reconstruction of the entire transcriptome by deep RNA sequencing (RNA-seq), even without a reference genome. However, transcriptome assembly from billions of RNA-seq reads, which are often very short, poses a significant informatics challenge. This Review summarizes the recent developments in transcriptome assembly approaches - reference-based, de novo and combined strategies-along with some perspectives on transcriptome assembly in the near future.

Martin, Jeffrey A.; Wang, Zhong

2011-09-01T23:59:59.000Z

254

Methanation assembly using multiple reactors  

DOE Patents [OSTI]

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.

Jahnke, Fred C.; Parab, Sanjay C.

2007-07-24T23:59:59.000Z

255

Airfoil nozzle and shroud assembly  

DOE Patents [OSTI]

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.

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

1997-06-03T23:59:59.000Z

256

Airfoil nozzle and shroud assembly  

DOE Patents [OSTI]

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.

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

1997-01-01T23:59:59.000Z

257

Bottom head assembly  

DOE Patents [OSTI]

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.

Fife, A.B.

1998-09-01T23:59:59.000Z

258

Microchannel heat sink assembly  

DOE Patents [OSTI]

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.

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

1992-03-24T23:59:59.000Z

259

E-Print Network 3.0 - acute gamma irradiation Sample Search Results  

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

Upgraded D Detector D Collaboration Summary: is due to the bulk silicon properties, photodiode test structures from the same wafer were irradiated Source: Fermi National...

260

Assembling thefacebook: Using heterogeneity to understand online social network assembly  

E-Print Network [OSTI]

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...

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

2015-01-01T23:59:59.000Z

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


261

Characterization of LWR spent fuel MCC-approved testing material-ATM-101  

SciTech Connect (OSTI)

The characterization data, obtained to date, for Materials Characterization Center (MCC) Approved Testing Materials (ATM)-101, spent fuel from H.B. Robinson, Unit 2, Assembly BO-5, are described. ATM-101 consists of 27 equal-length segments from nine fuel rods. Characterizations provided for ATM-101 include, (1) reactor, assembly, and fuel rod descriptions, (2) Assembly BO-5 irradiation history, (3) a description of unusual incidents that occurred to the rods, (4) fission gas release measurements, (5) results of ceramography/metallography examinations, (6) fuel burnup measurement results and correlations, (7) results of gamma scanning, (8) calculated values of the radionuclide inventory, and (9) results of a radionuclide chemical overcheck. Calculations for and measurement of radial distributions of selected radionuclides are planned. A description of pertinent results from other studies on sibling rods from Assembly BO-5 is also included. The distribution of ATM-101 to date is described along with characterization results on specially processed material. It is intended that this report be revised and updated as additional characterization data become available. 6 references, 23 figures, 19 tables.

Barner, J.O.

1984-06-01T23:59:59.000Z

262

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

SciTech Connect (OSTI)

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

Serell, D.C.; Kaplan, S.

1980-09-01T23:59:59.000Z

263

Stator and method of assembly  

DOE Patents [OSTI]

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.

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

2013-06-18T23:59:59.000Z

264

Three dimensional colorimetric assay assemblies  

DOE Patents [OSTI]

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.

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

2000-01-01T23:59:59.000Z

265

Moisture Research - Optimizing Wall Assemblies  

SciTech Connect (OSTI)

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

Arena, L.; Mantha, P.

2013-05-01T23:59:59.000Z

266

Laboratory for Characterization of Irradiated Graphite  

SciTech Connect (OSTI)

The newly completed Idaho National Laboratory (INL) Carbon Characterization Laboratory (CCL) is located in Labs C19 and C20 of the Idaho National Laboratory Research Center (IRC). The CCL was established under the Next Generation Nuclear Plant (NGNP) Project to support graphite and ceramic composite research and development activities. The research is in support of the Advanced Graphite Creep (AGC) experiment — a major material irradiation experiment within the NGNP Graphite program. The CCL is designed to characterize and test low activated irradiated materials such as high purity graphite, carbon-carbon composites, and silicon-carbide composite materials. The laboratory is fully capable of characterizing material properties for both irradiated and nonirradiated materials.

Karen A. Moore

2010-03-01T23:59:59.000Z

267

Flexible cloth seal assembly  

DOE Patents [OSTI]

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.

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

1999-06-29T23:59:59.000Z

268

Photovoltaic cell assembly  

DOE Patents [OSTI]

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.

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

1990-01-01T23:59:59.000Z

269

Crank shaft support assembly  

DOE Patents [OSTI]

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.

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)

2007-10-23T23:59:59.000Z

270

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

SciTech Connect (OSTI)

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.

Baek, Jong-Hyuk [KAERI] [KAERI; Byun, Thak Sang [ORNL] [ORNL; Maloy, S [Los Alamos National Laboratory (LANL)] [Los Alamos National Laboratory (LANL); Toloczko, M [Pacific Northwest National Laboratory (PNNL)] [Pacific Northwest National Laboratory (PNNL)

2014-01-01T23:59:59.000Z

271

Emulation of reactor irradiation damage using ion beams  

SciTech Connect (OSTI)

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.

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

2014-10-01T23:59:59.000Z

272

Advanced gray rod control assembly  

DOE Patents [OSTI]

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.

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

2013-09-17T23:59:59.000Z

273

Functionalized Methionine Polypeptides And Their Self Assembly  

E-Print Network [OSTI]

1.1: Schematic showing the self-assembly of M Ox (rac-L) yas well as their self-assembly into micelles. Both poly((Polypeptides And Their Self Assembly A thesis submitted in

Higgins, Robin

2013-01-01T23:59:59.000Z

274

Metal-directed protein self-assembly  

E-Print Network [OSTI]

Metal-Directed Protein Self- Assembly. Acc. Chem. Res. 43,Metal-directed protein self-assembly. Acc. Chem. Res. 43,Metal- mediated self-assembly of protein superstructures:

Salgado. Eric N.

2010-01-01T23:59:59.000Z

275

ASSEMBLY TRANSFER SYSTEM DESCRIPTION DOCUMENT  

SciTech Connect (OSTI)

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.

B. Gorpani

2000-06-26T23:59:59.000Z

276

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

SciTech Connect (OSTI)

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.

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

2011-09-01T23:59:59.000Z

277

Recent experience measuring breeder fresh fuel assemblies  

SciTech Connect (OSTI)

The International Atomic Energy Agency (IAEA) is required to conduct independent on-site verification of nuclear material held under safeguards agreements with member states. The nuclear material contained in liquid-metal fast breeder reactor (LMFBR) fresh fuel assemblies presents unique safeguards and measurement problems. Since LMFBR fresh fuel may contain uranium of various enrichments, plutonium, or mixtures of uranium and plutonium, a combination of nondestructive assay (NDA) methods and equipment must be used to achieve independent verification of the nuclear material contained in LMFBR fresh fuel assemblies. During 1985 and 1986, a number of measurements were carried out at the BOR-60 LMFBR facility near Dimitrovgrad, USSR to train IAEA inspectors in the use of standard NDA equipment and measurement procedures that can be employed to verify the nuclear material content of LMFBR fresh fuel. Since these measurements were conducted at an operation LMFBR facility, agency inspectors had an opportunity to receive training under actual field conditions. These activities also presented the first opportunity for the agency to test NDA measurement methods on LMFBR fresh fuel of the BOR-60 design. The measurements conducted at the BOR-60 site established that standard agency NDA equipment and procedures can be employed to independently verify the nuclear material content of LMFBR fresh fuel assemblies.

Rizhikov, V.; Fager, J.; Menlove, H.O.

1987-01-01T23:59:59.000Z

278

Automated self-assembly programming paradigm.  

E-Print Network [OSTI]

??Self-assembly is a ubiquitous process in nature in which a disordered set of components autonomously assemble into a complex and more ordered structure. Components interact… (more)

Li, Lin

2008-01-01T23:59:59.000Z

279

Self assembly of acetylcholinesterase on a goldnanoparticles...  

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

Self assembly of acetylcholinesterase on a gold nanoparticles–graphene nanosheet hybrid for organophosphate pesticide Self assembly of acetylcholinesterase on a gold...

280

Natural circulation in simulated LMFBR fuel assemblies  

SciTech Connect (OSTI)

Natural circulation experiments have been performed using simulated liquid metal fast breeder reactor fuel assemblies in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility, an engineering-scale sodium loop. Objective of these tests has been to provide experimental data under conditions that might be encountered during a partial or total loss of the shutdown heat removal system (SHRS) in a reactor. The experiments have included single- and two-phase tests under quasi-steady and transient conditions, at both nominal and non-nominal system conditions. Results from these test indicate that the potential for reactor damage during degraded SHRS operation is extremely slight, and that natural circulation can be a major contributor to safe operation of the system in both single- and two-phase flow during such operation.

Levin, A.E.; Carbajo, J.J.; Lloyd, D.B.; Montgomery, B.H.; Rose, S.D.; Wantland, J.L.

1985-01-01T23:59:59.000Z

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


281

Drive piston assembly for a valve actuator assembly  

DOE Patents [OSTI]

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.

Sun, Zongxuan (Troy, MI)

2010-02-23T23:59:59.000Z

282

Cooling assembly for fuel cells  

DOE Patents [OSTI]

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.

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

1990-01-01T23:59:59.000Z

283

Concentric tube support assembly  

SciTech Connect (OSTI)

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.

Rubio, Mark F.; Glessner, John C.

2012-09-04T23:59:59.000Z

284

Packer sealing assembly  

SciTech Connect (OSTI)

A sealing assembly for a packer includes a generally cylindrical elastomeric sealing element telescoped onto a mandrel between upper and lower expander heads. A sealing ring is disposed between each expander head and the sealing element at each end thereof for expanding radially outward toward engagement with the inside wall of a casing to keep the element from extruding between the expander heads and casing when setting the packer. A substantially non-expandable retaining ring surrounds the mandrel adjacent each of said sealing rings and an annular receptacle surrounds the mandrel and is located between each of the expander heads and the opposite ends of the sealing element. The receptacle has inner and outer malleable annular walls which are normally spaced radially outward from the mandrel and an end wall is integrally connected between these inner and outer walls so as to define an annular trough opening toward the sealing element. Extending into the trough is an annular protrusion which is integrally formed with the sealing element in each of the ends thereof so as to deform the inner walls radially inward into sealing engagement with the mandrel against elastomeric extrusion therebetween and so as to deform the outer walls radially outward into sealing engagement with the retaining ring against elastomeric extrusion therebetween when setting the packer.

Buckner, R.K.

1984-06-05T23:59:59.000Z

285

CU-LASP Test Facilities ! and Instrument Calibration Capabilities"  

E-Print Network [OSTI]

­ Star tracker ­ Solar position sensors ­ Test & calibration applications ­ End-to-end instrument;Total Solar Irradiance Radiometer Facility (TRF) · Total Solar Irradiance (TSI) instrument calibrations

Mojzsis, Stephen J.

286

Irradiation Stability of Carbon Nanotubes  

E-Print Network [OSTI]

Ion irradiation of carbon nanotubes is a tool that can be used to achieve modification of the structure. Irradiation stability of carbon nanotubes was studied by ion and electron bombardment of the samples. Different ion species at various energies...

Aitkaliyeva, Assel

2010-01-14T23:59:59.000Z

287

Characterization of spent fuel approved testing material: ATM-103  

SciTech Connect (OSTI)

The characterization data obtained to date are described for Approved Testing Material (ATM)-103, which is spent fuel from Assembly D101 of pressurized-water reactor Calvert Cliffs, No. 1. This report is one in a series being written by the Materials Characterization Center (MCC) at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US nuclear waste repository program. ATM-103 consists of 176 full-length irradiated fuel rods with rod-average burnups of about 2600 GJ/kgM (30 MWd/kgM) and less than 1% fission gas release. Characterization data include 1) as-fabricated fuel design, irradiation history, and subsequent storage and handling; 2) isotopic gamma scans; 3) fission gas analyses; 4) ceramography of the fuel and metallography of the cladding; 5) special fuels studies involving analytical transmission electron microscopy (AEM); 6) calculated nuclide inventories and radioactivities in the fuel and cladding; and 7) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel are being conducted and will be included in planned revisions of this report. 10 refs., 103 figs., 63 tabs.

Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

1988-04-01T23:59:59.000Z

288

Characterization of spent fuel approved testing material: ATM-106  

SciTech Connect (OSTI)

The characterization data obtained to date are described for Approved Testing Material (ATM)-106 spent fuel from Assembly BT03 of pressurized-water reactor Calvert Cliffs No. 1. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well- characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCWRM) program. ATM-106 consists of 20 full-length irradiated fuel rods with rod-average burnups of about 3700 GJ/kgM (43 MWd/kgM) and expected fission gas release of /approximately/10%. Characterization data include (1) as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) calculated nuclide inventories and radioactivities in the fuel and cladding; and (6) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel rod are being conducted and will be included in planned revisions of this report. 12 refs., 110 figs., 81 tabs.

Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thornhill, C.K.

1988-10-01T23:59:59.000Z

289

Characterization of spent fuel approved testing material--ATM-104  

SciTech Connect (OSTI)

The characterization data obtained to date are described for Approved Testing Material 104 (ATM-104), which is spent fuel from Assembly DO47 of the Calvert Cliffs Nuclear Power Plant (Unit 1), a pressurized-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-104 consists of 128 full-length irradiated fuel rods with rod-average burnups of about 42 MWd/kgM and expected fission gas release of about 1%. A variety of analyses were performed to investigate cladding characteristics, radionuclide inventory, and redistribution of fission products. Characterization data include (1) fabricated fuel design, irradiation history, and subsequent storage and handling history; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM) and electron probe microanalyses (EPMA); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding.

Guenther, R.J.; Blahnik, D.E.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

1991-12-01T23:59:59.000Z

290

Magnet Girder Assembly and Installation  

ScienceCinema (OSTI)

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.

None

2013-07-17T23:59:59.000Z

291

Direct hierarchical assembly of nanoparticles  

SciTech Connect (OSTI)

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.

Xu, Ting; Zhao, Yue; Thorkelsson, Kari

2014-07-22T23:59:59.000Z

292

Additive assembly of digital materials  

E-Print Network [OSTI]

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 ...

Ward, Jonathan (Jonathan Daniel)

2010-01-01T23:59:59.000Z

293

Magnet Girder Assembly and Installation  

SciTech Connect (OSTI)

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.

None

2012-12-12T23:59:59.000Z

294

Wafer scale micromachine assembly method  

DOE Patents [OSTI]

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.

Christenson, Todd R. (Albuquerque, NM)

2001-01-01T23:59:59.000Z

295

Microstructure and mechanical behavior of neutron irradiated ultrafine grained ferritic steel  

SciTech Connect (OSTI)

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.

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

2014-10-01T23:59:59.000Z

296

Hear Exchange Assembly  

DOE Patents [OSTI]

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.

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

2003-05-27T23:59:59.000Z

297

INTRODUCTION 1.1 Self Assembly  

E-Print Network [OSTI]

CHAPTER 1 INTRODUCTION 1.1 Self Assembly Nearly all complex biological systems are self-assembled to some degree: from these systems came the inspiration for directed self-assembly; that, by manipulating the same forces governing natural self-assembly, researchers would be able to decide at a basic level what

Braun, Paul

298

Molecular Coordination of Hierarchical Self-Assembly  

E-Print Network [OSTI]

Molecular Coordination of Hierarchical Self-Assembly Technical Report UT-CS-10-662 Bruce J. Mac to self-assemble into multiscale complex hierarchical systems. Keywords: algorithmic assembly, embodied, nano communication, nanofabrication, nanotechnology, Moore's Law, self-assembly, self-organization. 1

MacLennan, Bruce

299

Towards Assembly Automation at Small Size Scales  

E-Print Network [OSTI]

environmental impact Robotics Injection Molding Advanced Polymer Composites 3D Printing #12;In-Mold Assembly

Gupta, Satyandra K.

300

Irradiation hardening and loss of ductility of type 316L(N) stainless steel plate material due to neutron-irradiation  

SciTech Connect (OSTI)

Type 316 stainless steel is the primary candidate austenitic structural material for fusion first wall constructions. Here, type 316L(N) stainless steel plate material has been irradiated up to 10 dpa at temperatures of 80, 225, 325, and 425 C in the High Flux Reactor (HFR) of Petten. Tensile tests have been performed in the temperature range from RT to 575 C at a conventional strain rate of 5 {times} 10{sup {minus}4} s{sup {minus}1}. The results of the tensile tests are analyzed in terms of irradiation hardening and loss of ductility due to irradiation. Tensile properties saturate in the early stage (within 0.65 dpa) at the lowest applied irradiation temperature. It is indicated that the most severe degradation of tensile ductility occurs in the temperature range of 275 to 350 C. Comparison with literature data revealed a large scatter in irradiation hardening at irradiation temperatures above 325 C.

Horsten, M.G.; Vries, M.I. de [Netherlands Energy Research Foundation, Petten (Netherlands)

1996-12-31T23:59:59.000Z

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


301

Method for monitoring irradiated fuel using Cerenkov radiation  

DOE Patents [OSTI]

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.

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

1980-05-21T23:59:59.000Z

302

Neutron irradiation of beryllium: Recent Russian results  

SciTech Connect (OSTI)

Results on postirradiation tensile and compression testing, swelling and bubble growth during annealing for various grades of beryllium are presented. It is shown that swelling at temperatures above 550{degrees}C is sensitive to material condition and response is correlated with oxygen content. Swelling on the order of 15% can be expected at 700{degrees}C for doses on the order of 10{sup 22} n/cm{sup 2}. Bubble growth response depends on irradiation fluence.

Gelles, D.S. [Pacific Northwest Lab., Richland, VA (United States)

1992-12-31T23:59:59.000Z

303

SELF-ASSEMBLING AUTOMATA: A MODEL OF CONFORMATIONAL SELF-ASSEMBLY  

E-Print Network [OSTI]

SELF-ASSEMBLING AUTOMATA: A MODEL OF CONFORMATIONAL SELF-ASSEMBLY KAZUHIRO SAITOU Department An abstract model of self-assembling systems is presented where assembly instruc- tions are written as conformational switches ­ local rules that specify conforma- tional changes of a component. The model, the self-assembling

Saitou, Kazuhiro "Kazu"

304

Triangular and Hexagonal Tile Self-Assembly Systems Triangular and Hexagonal Tile Self-Assembly Systems  

E-Print Network [OSTI]

Triangular and Hexagonal Tile Self-Assembly Systems Triangular and Hexagonal Tile Self-Assembly theoretical aspects of the self-assembly of triangular tiles, in particular, right triangular tiles and equilateral triangular tiles, and the self-assembly of hexagonal tiles. We show that triangular tile assembly

Kari, Lila

305

Self-Assembly Using Hydrogen Bonds to Direct the Assembly of  

E-Print Network [OSTI]

Self-Assembly Using Hydrogen Bonds to Direct the Assembly of Crowded Aromatics Mark L. Bushey, Thuc · molecular recognition · nanotechnology · self-assembly 1. Introduction Self-assembly is a powerful tool such self-assembled system.[2] This relatively new class of liquid crystalline compounds, discovered in 1977

Hone, James

306

Fission product release from irradiated LWR fuel under accident conditions  

SciTech Connect (OSTI)

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.

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

1984-01-01T23:59:59.000Z

307

Continuous wave laser irradiation of explosives  

SciTech Connect (OSTI)

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.

McGrane, Shawn D.; Moore, David S.

2010-12-01T23:59:59.000Z

308

Cylinder Test Specification  

SciTech Connect (OSTI)

The purpose of the cylinder testis two-fold: (1) to characterize the metal-pushing ability of an explosive relative to that of other explosives as evaluated by the E{sub 19} cylinder energy and the G{sub 19} Gurney energy and (2) to help establish the explosive product equation-of-state (historically, the Jones-Wilkins-Lee (JWL) equation). This specification details the material requirements and procedures necessary to assemble and fire a typical Los Alamos National Laboratory (LANL) cylinder test. Strict adherence to the cylinder. material properties, machining tolerances, material heat-treatment and etching processes, and high explosive machining tolerances is essential for test-to-test consistency and to maximize radial wall expansions. Assembly and setup of the cylinder test require precise attention to detail, especially when placing intricate pin wires on the cylinder wall. The cylinder test is typically fired outdoors and at ambient temperature.

Richard Catanach; Larry Hill; Herbert Harry; Ernest Aragon; Don Murk

1999-10-01T23:59:59.000Z

309

Development of a New Multiplying Assembly for Research, Validation, Evaluation, and Learning  

SciTech Connect (OSTI)

A new multiplying test assembly is under development at Idaho National Laboratory (INL) to support research, validation, evaluation, and learning. The item is comprised of two stacked highly-enriched uranium (HEU) cylinders each 11.4 cm in diameter and having a combined height of 8.4 cm. The combined mass is 14.4 kg of HEU. Calculations for the bare configuration of the assembly indicate a multiplication level of >2.5 (keff = 0.62). Reflected configurations of the assembly, using either polyethylene or tungsten, are possible and have the capability of raising its multiplication level to approximately 8. This paper will describe the MCNP calculations performed to assess the assembly's multiplication level under different conditions and describe the resource available at INL to support visiting researchers in their use of the material. We will also describe some preliminary calculations and test activities using the assembly to study neutron multiplicity.

David L. Chichester

2012-10-01T23:59:59.000Z

310

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

E-Print Network [OSTI]

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 ...

Kopesky, Paul Wayne

311

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

SciTech Connect (OSTI)

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.

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

2013-09-01T23:59:59.000Z

312

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

E-Print Network [OSTI]

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...

Cheboli, Ramakrishna

2012-06-07T23:59:59.000Z

313

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

SciTech Connect (OSTI)

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

Isabella J van Rooyen

2012-09-01T23:59:59.000Z

314

Test Automation Test Automation  

E-Print Network [OSTI]

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

Mousavi, Mohammad

315

Insulation assembly for electric machine  

SciTech Connect (OSTI)

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.

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

2013-10-15T23:59:59.000Z

316

Dynamics of assembly production flow  

E-Print Network [OSTI]

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...

Ezaki, Takahiro; Nishinari, Katsuhiro

2015-01-01T23:59:59.000Z

317

Understanding the Irradiation Behavior of Zirconium Carbide  

SciTech Connect (OSTI)

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.

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

2013-10-11T23:59:59.000Z

318

SciTech Connect: Normal Conditions of Transport Truck Test of...  

Office of Scientific and Technical Information (OSTI)

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...

319

DNA-guided nanoparticle assemblies  

DOE Patents [OSTI]

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.

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

2013-07-16T23:59:59.000Z

320

DOE Cell Component Accelerated Stress Test Protocols for PEM...  

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

CELL COMPONENT ACCELERATED STRESS TEST PROTOCOLS FOR PEM FUEL CELLS (Electrocatalysts, Supports, Membranes, and Membrane Electrode Assemblies) March 2007 Fuel cells, especially for...

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


321

Cell Component Accelerated Stress Test Protocols for PEM Fuel...  

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

USCAR FUEL CELL TECH TEAM CELL COMPONENT ACCELERATED STRESS TEST PROTOCOLS FOR PEM FUEL CELLS (Electrocatalysts, Supports, Membranes, and Membrane Electrode Assemblies) Revised May...

322

Microsoft Word - NTS Performance Test Rpt - Final.doc  

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

The command post tabletop performance test (TTPT) scenario involved a simulated subcritical experiment assembly at the U1a facility that was deliberately set on fire by a...

323

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

SciTech Connect (OSTI)

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.

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

2002-07-01T23:59:59.000Z

324

Simulated nuclear reactor fuel assembly  

DOE Patents [OSTI]

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.

Berta, Victor T. (Idaho Falls, ID)

1993-01-01T23:59:59.000Z

325

Simulated nuclear reactor fuel assembly  

DOE Patents [OSTI]

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.

Berta, V.T.

1993-04-06T23:59:59.000Z

326

Hot hollow cathode gun assembly  

DOE Patents [OSTI]

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.

Zeren, J.D.

1983-11-22T23:59:59.000Z

327

Annual Report CMS Spring Assembly  

E-Print Network [OSTI]

Annual Report 2007-2008 CMS Spring Assembly & Length of Service Awards March 9, 2012 #12;Annual Report 2007-2008 News & Events: Alumni David Mearns (CMS MS `86) Selected as co-recipient of USF's Distinguished Alumni Award, Fall 2011 #12;Annual Report 2007-2008 News & Events: Faculty Dr. Robert Byrne

Meyers, Steven D.

328

Student Assembly Offices Student Assembly Representatives: There are four representatives per class to the Student Assembly. Their duties are to  

E-Print Network [OSTI]

Student Assembly Offices Student Assembly Representatives: There are four representatives per class to the Student Assembly. Their duties are to: Represent the student body of the Medical College of Wisconsin-section of the student body. Fairly administer and distribute all funds including those designated as Student Activity

329

Light water reactor mixed-oxide fuel irradiation experiment  

SciTech Connect (OSTI)

The United States Department of Energy Office of Fissile Materials Disposition is sponsoring and Oak Ridge National Laboratory (ORNL) is leading an irradiation experiment to test mixed uranium-plutonium oxide (MOX) fuel made from weapons-grade (WG) plutonium. In this multiyear program, sealed capsules containing MOX fuel pellets fabricated at Los Alamos National Laboratory (LANL) are being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The planned experiments will investigate the utilization of dry-processed plutonium, the effects of WG plutonium isotopics on MOX performance, and any material interactions of gallium with Zircaloy cladding.

Hodge, S.A.; Cowell, B.S. [Oak Ridge National Lab., TN (United States); Chang, G.S.; Ryskamp, J.M. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.

1998-06-01T23:59:59.000Z

330

Transplanting assembly of individual carbon nanotubes  

E-Print Network [OSTI]

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 ...

Kim, Soohyung

2009-01-01T23:59:59.000Z

331

On the mathematics of self-assembly.  

E-Print Network [OSTI]

??Self-assembly is the ubiquitous process by which simple objects come together under simple rules to form more complex objects. Self-assembly occurs in nature to produce… (more)

Reishus, Dustin

2009-01-01T23:59:59.000Z

332

On the mathematics of self-assembly.  

E-Print Network [OSTI]

?? Self-assembly is the ubiquitous process by which simple objects come together under simple rules to form more complex objects. Self-assembly occurs in nature to… (more)

Reishus, Dustin

2010-01-01T23:59:59.000Z

333

Solder self-assembly for MEMS fabrication  

E-Print Network [OSTI]

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 ...

Au, Hin Meng, 1977-

2004-01-01T23:59:59.000Z

334

Camera assembly design proposal for SRF cavity image collection  

SciTech Connect (OSTI)

This project seeks to collect images from the inside of a superconducting radio frequency (SRF) large grain niobium cavity during vertical testing. These images will provide information on multipacting and other phenomena occurring in the SRF cavity during these tests. Multipacting, a process that involves an electron buildup in the cavity and concurrent loss of RF power, is thought to be occurring near the cathode in the SRF structure. Images of electron emission in the structure will help diagnose the source of multipacting in the cavity. Multipacting sources may be eliminated with an alteration of geometric or resonant conditions in the SRF structure. Other phenomena, including unexplained light emissions previously discovered at SLAC, may be present in the cavity. In order to effectively capture images of these events during testing, a camera assembly needs to be installed to the bottom of the RF structure. The SRF assembly operates under extreme environmental conditions: it is kept in a dewar in a bath of 2K liquid helium during these tests, is pumped down to ultra-high vacuum, and is subjected to RF voltages. Because of this, the camera needs to exist as a separate assembly attached to the bottom of the cavity. The design of the camera is constrained by a number of factors that are discussed.

Tuozzolo, S.

2011-10-10T23:59:59.000Z

335

Self-assembled DNA Structures for Nanoconstruction  

E-Print Network [OSTI]

Self-assembled DNA Structures for Nanoconstruction Hao Yan, Peng Yin, Sung Ha Park, Hanying Li methods based on DNA self-assembly. Here we review our recent experimental progress to utilize novel DNA nanostructures for self-assembly as well as for templates in the fabrication of functional nano

Yin, Peng

336

Self-assembling DNA Nanostructures for Patterned Molecular Assembly Thomas H. LaBeana  

E-Print Network [OSTI]

1 Self-assembling DNA Nanostructures for Patterned Molecular Assembly Thomas H. LaBeana , Kurt V@cs.duke.edu; Tel: (919)660-65685 Abstract The Chapter describes the use of DNA for molecular-scale self-assembly with a discussion of DNA-nanostructures, starting with the self-assembly of various building-blocks known as DNA

Reif, John H.

337

Irradiation requirements of Nb3Sn based SC magnets electrical insulation  

E-Print Network [OSTI]

Irradiation requirements of Nb3Sn based SC magnets electrical insulation developed within the Eu electrical insulation candidates · EuCARD insulators certification conditions · Post irradiation tests and neutrino factories will be subjected to very high radiation doses. · The electrical insulation employed

McDonald, Kirk

338

Hybrid microcircuit board assembly with lead-free solders  

SciTech Connect (OSTI)

An assessment was made of the manufacturability of hybrid microcircuit test vehicles assembled using three Pb-free solder compositions 96.5Sn--3.5Ag (wt.%), 91.84Sn--3.33Ag--4.83Bi, and 86.85Sn--3.15Ag--5.0Bi--5.0Au. The test vehicle substrate was 96% alumina; the thick film conductor composition was 76Au--21Pt--3Pd. Excellent registration between the LCCC or chip capacitor packages and the thick film solder pads was observed. Reduced wetting of bare (Au-coated) LCCC castellations was eliminated by hot solder dipping the I/Os prior to assembly of the circuit card. The Pb-free solders were slightly more susceptible to void formation, but not to a degree that would significantly impact joint functionality. Microstructural damage, while noted in the Sn-Pb solder joints, was not observed in the Pb-free interconnects.

Vianco, P.T.; Hernandez, C.L.; Rejent, J.A.

2000-01-11T23:59:59.000Z

339

A Simplified Shuttle Irradiation Facility for ATR  

SciTech Connect (OSTI)

During the past fifteen years there has been a steady increase in the demand for radioisotopes in nuclear medicine and a corresponding decline in the number of reactors within the U.S. capable of producing them. The Advanced Test Reactor (ATR) is the largest operating test reactor in the U.S., but its isotope production capabilities have been limited by the lack of an installed isotope shuttle irradiation system. A concept for a simple “low cost” shuttle irradiation facility for ATR has been developed. Costs were reduced (in comparison to previous ATR designs) by using a shielded trough of water installed in an occupiable cubicle as a shielding and contamination control barrier for the send and receive station. This shielding concept also allows all control valves to be operated by hand and thus the need for an automatic control system was eliminated. It was determined that 4 – 5 ft of water would be adequate to shield the isotopes of interest while shuttles are transferred to a small carrier. An additional feature of the current design is a non-isolatable by-pass line, which provides a minimum coolant flow to the test region regardless of which control valves are opened or closed. This by-pass line allows the shuttle facility to be operated without bringing reactor coolant water into the cubicle except for send and receive operations. The irradiation position selected for this concept is a 1.5 inch “B” hole (B-11). This position provides neutron fluxes of approximately: 1.6 x 1014 (<0.5 eV) and 4.0 x 1013 (>0.8 MeV) n/cm2*sec.

Palmer, Alma Joseph; Laflin, S. T.

1999-09-01T23:59:59.000Z

340

A Simplified Shuttle Irradiation Facility for ATR  

SciTech Connect (OSTI)

During the past fifteen years there has been a steady increase in the demand for radioisotopes in nuclear medicine and a corresponding decline in the number of reactors within the U.S. capable of producing them. The Advanced Test Reactor (ATR) is the largest operating test reactor in the U.S., but its isotope production capabilities have been limited by the lack of an installed isotope shuttle irradiation system. A concept for a simple "low cost" shuttle irradiation facility for ATR has been developed. Cost were reduced (in comparison to previous ATR designs) by using a shielded trough of water installed in an occupiable cubicle as a shielding and contamination control barrier for the send and receive station. This shielding concept also allows all control valves to be operated by hand and thus the need for an automatic control system was eliminated. It was determined that 4-5 ft of water would be adequate to shield the isotopes of interest while shuttles are transferred to a small carrier. An additional feature of the current design is a non-isolatable by-pass line, which provides a minimum coolant flow to the test region regardless of which control valves are opened or closed. This by-pass line allows the shuttle facility to be operated without bringing reactor coolant water into the cubicle except for send and receive operations.

A. J. Palmer; S. T. Laflin

1999-08-01T23:59:59.000Z

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


341

Metal-ceramic joint assembly  

DOE Patents [OSTI]

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.

Li, Jian (New Milford, CT)

2002-01-01T23:59:59.000Z

342

Pressure-equalizing PV assembly and method  

DOE Patents [OSTI]

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.

Dinwoodie, Thomas L.

2004-10-26T23:59:59.000Z

343

Pressure equalizing photovoltaic assembly and method  

DOE Patents [OSTI]

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.

Dinwoodie, Thomas L. (Piedmont, CA)

2003-05-27T23:59:59.000Z

344

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

SciTech Connect (OSTI)

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.

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

1998-08-01T23:59:59.000Z

345

Sodium boiling dryout correlation for LMFBR fuel assemblies  

SciTech Connect (OSTI)

Under certain postulated accident conditions for a Liquid Metal Fast Breeder Reactor (LMFBR), such as the failure of the shutdown heat removal system (SHRS), sodium boiling and clad dryout might occur in the fuel assemblies. It is important to predict the time from boiling inception to dryout, since sustained clad dryout will result in core damage. In this paper a dryout correlation is presented. This correlation is based on 21 boiling tests which resulted in dryout from the THORS BUNDLE 6A, a 19-pin full-length simulated LMFBR fuel assembly and from the THORS Bundle 9, a 61-pin full-length simulated LMFBR fuel assembly. All these tests were performed as follows: for each specified bundle power, an initial steady-state high sodium flow was established, for which sodium boiling did not occur in the bundle. The temperature at the outlet of the test section was approx. 700/sup 0/C. Then, using a programmable pump control system, the flow was reduced to a low value and boiling occurred.

Carbajo, J.J.; Rose, S.D.

1984-01-01T23:59:59.000Z

346

Research of a boundary condition quantifiable correction method in the assembly homogenization  

SciTech Connect (OSTI)

The methods and codes currently used in assembly homogenization calculation mostly adopt the reflection boundary conditions. The influences of real boundary conditions on the assembly homogenized parameters were analyzed. They were summarized into four quantifiable effects, and then the mathematical expressions could be got by linearization hypothesis. Through the calculation of a test model, it had been found that the result was close to transport calculation result when considering four boundary quantifiable effects. This method would greatly improve the precision of a core design code which using the assembly homogenization methods, but without much increase of the computing time. (authors)

Peng, L. H.; Liu, Z. H.; Zhao, J. [Inst. of Nuclear and New Energy Technology, Tsinghua Univ., Beijing, 100084 (China); Li, W. H. [China Nuclear Power Technology Research Inst., Shenzhen, 518026 (China)

2012-07-01T23:59:59.000Z

347

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...

348

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:YearRound-UpHeatMulti-Dimensional ElectricalEnergy Frozen Telescope Looks to Ends

349

Proton Irradiation Damage Assessment of Carbon Reinforced Composites  

E-Print Network [OSTI]

Proton Irradiation Damage Assessment of Carbon Reinforced Composites: 2-D & 3-D Weaved Structures carbon-carbon composite ATJ Graphite 3D CC composite AGS Beam-on-Target tests show clearly that carbon composites are better absorbers of thermo- mechanical shock. This is attributed to the very low coeff

McDonald, Kirk

350

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsrucLas ConchasPassiveSubmittedStatus TomAbout »Lab (Newport NewsStyle

351

The Assembly of the Belle II TOP Counter  

E-Print Network [OSTI]

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.

Wang, Boqun

2015-01-01T23:59:59.000Z

352

The Assembly of the Belle II TOP Counter  

E-Print Network [OSTI]

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.

Boqun Wang; for the Belle II PID Group

2015-01-13T23:59:59.000Z

353

Micro-grippers for assembly of LIGA parts  

SciTech Connect (OSTI)

This paper describes ongoing testing of two microgrippers for assembly of LIGA (Lithographie Galvanoformung Abformung) parts. The goal is to place 100 micron outside diameter (OD) LIGA gears with a 50 micron inner diameter hole onto pins ranging from 35 to 49 microns. The first micro gripper is a vacuum gripper made of a 100 micron OD stainless steel tube. The second micro gripper is a set of tweezers fabricated using the LIGA process. Nickel, Permalloy, and copper materials are tested. The tweezers are actuated by a collet mechanism which is closed by a DC linear motor.

Feddema, J.; Polosky, M.; Christenson, T.; Spletzer, B.; Simon, R.

1997-12-31T23:59:59.000Z

354

Method of monolithic module assembly  

DOE Patents [OSTI]

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.

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

1999-01-01T23:59:59.000Z

355

Functional Characterization and Surface Mapping of Frataxin (FXN) Interactions with the Fe-S Cluster Assembly Complex  

E-Print Network [OSTI]

of the protein. Kinetic and analytical ultracentrifugation studies revealed a complex heterogeneous mixture of species some of which can activate the Fe-S assembly complex. A previously identified acetylation site was also tested using mutants that mimic...

Thorstad, Melissa

2013-05-23T23:59:59.000Z

356

Algorithm for a microfluidic assembly line  

E-Print Network [OSTI]

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.

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

2011-01-19T23:59:59.000Z

357

Tensile and Charpy impact properties of irradiated reduced-activation ferritic steels  

SciTech Connect (OSTI)

Tensile tests were conducted on 8 reduced-activation Cr-W steels after irradiation to 15-17 and 26-29 dpa, and Charpy impact tests were conducted on steels irradiated to 26-29 dpa. Irradiation was in Fast Flux Test Facility at 365 C on steels containing 2.25-12% Cr, varying amounts of W, V, and Ta, and 0.1%C. Previously, tensile specimens were irradiated to 6-8 dpa and Charpy specimens to 6-8, 15- 17, and 20-24 dpa. Tensile and Charpy specimens were also thermally aged to 20,000 h at 365 C. Thermal aging had little effect on tensile properties or ductile-brittle transition temperature (DBTT), but several steels showed a slight increase in upper-shelf energy (USE). After 7 dpa, strength increased (hardened) and then remained relatively unchanged through 26-29 dpa (ie, strength saturated with fluence). Post-irradiation Charpy impact tests after 26-29 dpa showed that the loss of impact toughness (increased DBTT, decreased USE) remained relatively unchanged from the values after 20-24 dpa, which had been relatively unchanged from the earlier irradiations. As before, the two 9Cr steels had the most irradiation resistance.

Klueh, R.L.; Alexander, D.J.

1996-10-01T23:59:59.000Z

358

Natural convection heat transfer within horizontal spent nuclear fuel assemblies  

SciTech Connect (OSTI)

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.

Canaan, R.E.

1995-12-01T23:59:59.000Z

359

Flexible Assembly Solar Technology | Department of Energy  

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

Assembly Solar Technology This presentation was delivered at the SunShot Concentrating Solar Power (CSP) Program Review 2013, held April 23-25, 2013 near Phoenix, Arizona....

360

Hierarchical Assembly of Inorganic Nanostructure Building Blocks...  

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

Nanostructure Building Blocks to Octahedral Superstructures – A True Template-Free Self Hierarchical Assembly of Inorganic Nanostructure Building Blocks to Octahedral...

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


361

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

E-Print Network [OSTI]

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 ...

Lichter, Jenny, 1982-

2009-01-01T23:59:59.000Z

362

Assisted assembly: how to improve a de novo genome assembly by using related species  

E-Print Network [OSTI]

We describe a new assembly algorithm, where a genome assembly with low sequence coverage, either throughout the genome or locally, due to cloning bias, is considerably improved through an assisting process via a related ...

Gnerre, Sante

363

Corner-cutting mining assembly  

DOE Patents [OSTI]

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.

Bradley, J.A.

1981-07-01T23:59:59.000Z

364

Valve stem and packing assembly  

DOE Patents [OSTI]

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.

Wordin, J.J.

1991-09-03T23:59:59.000Z

365

Valve stem and packing assembly  

DOE Patents [OSTI]

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.

Wordin, John J. (Bingham County, ID)

1991-01-01T23:59:59.000Z

366

Ultra-precision positioning assembly  

DOE Patents [OSTI]

An apparatus and method is disclosed for ultra-precision positioning. A slide base provides a foundational support. A slide plate moves with respect to the slide base along a first geometric axis. Either a ball-screw or a piezoelectric actuator working separate or in conjunction displaces the slide plate with respect to the slide base along the first geometric axis. A linking device directs a primary force vector into a center-line of the ball-screw. The linking device consists of a first link which directs a first portion of the primary force vector to an apex point, located along the center-line of the ball-screw, and a second link for directing a second portion of the primary force vector to the apex point. A set of rails, oriented substantially parallel to the center-line of the ball-screw, direct movement of the slide plate with respect to the slide base along the first geometric axis and are positioned such that the apex point falls within a geometric plane formed by the rails. The slide base, the slide plate, the ball-screw, and the linking device together form a slide assembly. Multiple slide assemblies can be distributed about a platform. In such a configuration, the platform may be raised and lowered, or tipped and tilted by jointly or independently displacing the slide plates.

Montesanti, Richard C. (San Francisco, CA); Locke, Stanley F. (Livermore, CA); Thompson, Samuel L. (Pleasanton, CA)

2002-01-01T23:59:59.000Z

367

Mixed Stream Test Rig (MISTER) Startup Report  

SciTech Connect (OSTI)

This report describes the work accomplished to date to design, procure, assemble, authorize, and startup the Mixed Stream Test Rig (MISTER) at the Idaho National Laboratory (INL). It describes the reasons for establishing this capability, physical configuration of the test equipment, operations methodology, initial success, and plans for completing the initial 1,000 hour test.

Charles Park

2011-02-01T23:59:59.000Z

368

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

SciTech Connect (OSTI)

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.

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

1982-11-01T23:59:59.000Z

369

Passive tailoring of laser-accelerated ion beam cut-off energy by using double foil assembly  

SciTech Connect (OSTI)

A double foil assembly is shown to be effective in tailoring the maximum energy produced by a laser-accelerated proton beam. The measurements compare favorably with adiabatic expansion simulations, and particle-in-cell simulations. The arrangement proposed here offers for some applications a simple and passive way to utilize simultaneously highest irradiance lasers that have best laser-to-ion conversion efficiency while avoiding the production of undesired high-energy ions.

Chen, S. N., E-mail: sophia.chen@polytechnique.edu; Brambrink, E.; Mancic, A.; Romagnani, L.; Audebert, P.; Fuchs, J., E-mail: julien.fuchs@polytechnique.fr [Laboratoire pour l'Utilisation des Lasers Intenses, UMR 7605 CNRS-CEA-École Polytechnique-Université Paris VI, Palaiseau (France); Robinson, A. P. L. [Central Laser Facility, STFC Rutherford-Appleton Laboratory, Chilton, Didcot, Oxfordshire OX11 0QX (United Kingdom)] [Central Laser Facility, STFC Rutherford-Appleton Laboratory, Chilton, Didcot, Oxfordshire OX11 0QX (United Kingdom); Antici, P. [Laboratoire pour l'Utilisation des Lasers Intenses, UMR 7605 CNRS-CEA-École Polytechnique-Université Paris VI, Palaiseau (France) [Laboratoire pour l'Utilisation des Lasers Intenses, UMR 7605 CNRS-CEA-École Polytechnique-Université Paris VI, Palaiseau (France); Dipartimento SBAI, Università di Roma « La Sapienza », Via Scarpa 14-16, 00165 Roma (Italy); INRS-Énergie et Matériaux, 1650 bd. L. Boulet, Varennes, J3X1S2 Québec (Canada); D'Humières, E. [Physics Department, MS-220, University of Nevada, Reno, Nevada 89557 (United States) [Physics Department, MS-220, University of Nevada, Reno, Nevada 89557 (United States); Centre de Physique Théorique, CNRS-Ecole Polytechnique, 91128 Palaiseau (France); University of Bordeaux—CNRS—CEA, CELIA, UMR5107, 33405 Talence (France); Gaillard, S. [Physics Department, MS-220, University of Nevada, Reno, Nevada 89557 (United States)] [Physics Department, MS-220, University of Nevada, Reno, Nevada 89557 (United States); Grismayer, T.; Mora, P. [Centre de Physique Théorique, CNRS-Ecole Polytechnique, 91128 Palaiseau (France)] [Centre de Physique Théorique, CNRS-Ecole Polytechnique, 91128 Palaiseau (France); Pépin, H. [INRS-Énergie et Matériaux, 1650 bd. L. Boulet, Varennes, J3X1S2 Québec (Canada)] [INRS-Énergie et Matériaux, 1650 bd. L. Boulet, Varennes, J3X1S2 Québec (Canada)

2014-02-15T23:59:59.000Z

370

Testing of Gas Reactor Materials and Fuel in the Advanced Test Reactor  

SciTech Connect (OSTI)

The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations.

S. Blaine Grover

2004-10-01T23:59:59.000Z

371

TESTING OF GAS REACTOR MATERIALS AND FUEL IN THE ADVANCED TEST REACTOR  

SciTech Connect (OSTI)

The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations.

Grover, S.B.

2004-10-06T23:59:59.000Z

372

Irradiation behavior of miniature experimental uranium silicide fuel plates  

SciTech Connect (OSTI)

Uranium silicides, because of their relatively high uranium density, were selected as candidate dispersion fuels for the higher fuel densities required in the Reduced Enrichment Research and Test Reactor (RERTR) Program. Irradiation experience with this type of fuel, however, was limited to relatively modest fission densities in the bulk from, on the order of 7 x 10/sup 20/ cm/sup -3/, far short of the approximately 20 x 10/sup 20/ cm/sup -3/ goal established for the RERTR program. The purpose of the irradiation experiments on silicide fuels on the ORR, therefore, was to investigate the intrinsic irradiation behavior of uranium silicide as a dispersion fuel. Of particular interest was the interaction between the silicide particles and the aluminum matrix, the swelling behavior of the silicide particles, and the maximum volume fraction of silicide particles that could be contained in the aluminum matrix.

Hofman, G.L.; Neimark, L.A.; Mattas, R.F.

1983-01-01T23:59:59.000Z

373

Pressurized water reactor fuel assembly subchannel void fraction measurement  

SciTech Connect (OSTI)

The void fraction measurement experiment of pressurized water reactor (PWR) fuel assemblies has been conducted since 1987 under the sponsorship of the Ministry of International Trade and Industry as a Japanese national project. Two types of test sections are used in this experiment. One is a 5 x 5 array rod bundle geometry, and the other is a single-channel geometry simulating one of the subchannels in the rod bundle. Wide gamma-ray beam scanners and narrow gamma-ray beam computed tomography scanners are used to measure the subchannel void fractions under various steady-state and transient conditions. The experimental data are expected to be used to develop a void fraction prediction model relevant to PWR fuel assemblies and also to verify or improve the subchannel analysis method. The first series of experiments was conducted in 1992, and a preliminary evaluation of the data has been performed. The preliminary results of these experiments are described.

Akiyama, Yoshiei [Mitsubishi Heavy Industries, Ltd., Yokohama (Japan). Nuclear Fuel and Core Engineering Dept.; Hori, Keiichi [Mitsubishi Heavy Industries, Ltd., Hyougo (Japan); Miyazaki, Keiji [Osaka Univ. (Japan). Faculty of Engineering; Mishima, Kaichiro [Kyoto Univ., Osaka (Japan). Research Reactor Inst.; Sugiyama, Shigekazu [Nuclear Power Engineering Corp., Tokyo (Japan). Nuclear Fuel Dept.

1995-12-01T23:59:59.000Z

374

Heavily Irradiated N-in-p Thin Planar Pixel Sensors with and without Active Edges  

E-Print Network [OSTI]

We present the results of the characterization of silicon pixel modules employing n-in-p planar sensors with an active thickness of 150 $\\mathrm{\\mu}$m, produced at MPP/HLL, and 100-200 $\\mathrm{\\mu}$m thin active edge sensor devices, produced at VTT in Finland. These thin sensors are designed as candidates for the ATLAS pixel detector upgrade to be operated at the HL-LHC, as they ensure radiation hardness at high fluences. They are interconnected to the ATLAS FE-I3 and FE-I4 read-out chips. Moreover, the n-in-p technology only requires a single side processing and thereby it is a cost-effective alternative to the n-in-n pixel technology presently employed in the LHC experiments. High precision beam test measurements of the hit efficiency have been performed on these devices both at the CERN SpS and at DESY, Hamburg. We studied the behavior of these sensors at different bias voltages and different beam incident angles up to the maximum one expected for the new Insertable B-Layer of ATLAS and for HL-LHC detectors. Results obtained with 150 $\\mathrm{\\mu}$m thin sensors, assembled with the new ATLAS FE-I4 chip and irradiated up to a fluence of 4$\\times$10$^{15}\\mathrm{n}_{\\mathrm{eq}}/\\mathrm{cm}^2$, show that they are excellent candidates for larger radii of the silicon pixel tracker in the upgrade of the ATLAS detector at HL-LHC. In addition, the active edge technology of the VTT devices maximizes the active area of the sensor and reduces the material budget to suit the requirements for the innermost layers. The edge pixel performance of VTT modules has been investigated at beam test experiments and the analysis after irradiation up to a fluence of 5$\\times$10$^{15}\\mathrm{n}_{\\mathrm{eq}}/\\mathrm{cm}^2$ has been performed using radioactive sources in the laboratory.

S. Terzo; L. Andricek; A. Macchiolo; H. G. Moser; R. Nisius; R. H. Richter; P. Weigell

2014-02-19T23:59:59.000Z

375

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

SciTech Connect (OSTI)

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.

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

2010-02-16T23:59:59.000Z

376

Measurements on spent-fuel assemblies at Arkansas Nuclear One using the Fork system. Final report, January 1995  

SciTech Connect (OSTI)

The Fork measurement system has been used to examine spent-fuel assemblies at the two reactors of Arkansas Nuclear One, operated by Entergy Operations, Inc. The Unit 1 reactor is a Babcock and Wilcox (B and W) design, and the Unit 2 reactor is a Combustion Engineering (CE) design. The neutron and gamma-ray emissions from individual spent-fuel assemblies were measured in the storage pools by raising each assembly pathway out of the storage rack and performing a measurement near the center of the assembly. The overall accuracy of the measurements after corrections is about 2%. Thirty-four assemblies were examined at Unit 1, and forty-one assemblies at Unit 2. The average deviation of the burnup measurements from the calibration was 3.0% at Unit 1 and 3.5% at Unit 2, indicating 2 to 3% random variation among the reactor records. There was no indication of clearly anomalous assemblies. Axial Scans of the variation in neutron and gamma ray emission were obtained by collecting data at several locations along the length of three assemblies at Unit 2. Two of these assemblies were nonstandard in that each contained a small neutron source. The sources were detected by the axial scans. The test program was a cooperative effort involving Sandia National Laboratories, Los Alamos National Laboratory, Entergy Operations, Inc., the Electric Power Research Institute, and the Office of Civilian Radioactive Waste Management of the US Department of Energy.

Ewing, R.I.; Bronowski, D.R. [Sandia National Labs., Albuquerque, NM (United States); Bosler, G.E.; Siebelist, R. [Los Alamos National Lab., NM (United States); Priore, J.; Hansford, C.H.; Sullivan, S. [Entergy Operations, Inc., Russellville, AR (United States). Arkansas Nuclear One

1997-03-01T23:59:59.000Z

377

Post-irradiation Examination and Fission Product Inventory Analysis of AGR-1 Irradiation Capsules  

SciTech Connect (OSTI)

The AGR-1 experiment was the first in a series of Advanced Gas Reactor (AGR) experiments designed to test TRISO fuel under High Temperature Gas Reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post-irradiation examination (PIE) at INL’s Materials and Fuels Complex (MFC). The inventory and distribution of fission products, especially Ag-110m, was assessed and analyzed for all the components of the AGR-1 capsules. This data should help inform the study of fission product migration in coated particle fuel. Gamma spectrometry was used to measure the activity of various different fission products in the different components of the AGR-1 test train. Each capsule contained: 12 fuel compacts, a graphite holder that kept the fuel compacts in place, graphite spacers that were above and below the graphite holders and fuel compacts, gas lines through which a helium neon gas mixture flowed in and out of each capsule, and the stainless steel shell that contained the experiment. Gamma spectrometry results and the experimental techniques used to capture these results will be presented for all the capsule components. The components were assayed to determine the total activity of different fission products present in or on them. These totals are compared to the total expected activity of a particular fission product in the capsule based on predictions from physics simulation. Based on this metric, a significant fraction of the Ag-110m was detected outside the fuel compacts, but the amount varied highly between the 6 capsules. Very small fractions of Cs-137 (<2E-5), Cs-134 (<1e-5), and Eu-154 (<4e-4) were detected outside of the fuel compacts. Additionally, the distribution of select fission products in some of the components including the fuel compacts and the graphite holders were measured and will be discussed.

J M Harp; P D Demkowicz; S A Ploger

2012-10-01T23:59:59.000Z

378

First Assemblies Using Deep Trench Termination Diodes  

E-Print Network [OSTI]

First Assemblies Using Deep Trench Termination Diodes F. Baccar, L. Théolier, S. Azzopardi, F. Le Trench Termination (DT2 ), are analyzed in a reliability purpose. For the first time, assemblies are made. As a consequence, to improve the breakdown voltage, it is necessary to create an adequate edge termination

Paris-Sud XI, Université de

379

Microfabricated field calibration assembly for analytical instruments  

DOE Patents [OSTI]

A microfabricated field calibration assembly for use in calibrating analytical instruments and sensor systems. The assembly comprises a circuit board comprising one or more resistively heatable microbridge elements, an interface device that enables addressable heating of the microbridge elements, and, in some embodiments, a means for positioning the circuit board within an inlet structure of an analytical instrument or sensor system.

Robinson, Alex L. (Albuquerque, NM); Manginell, Ronald P. (Albuquerque, NM); Moorman, Matthew W. (Albuquerque, NM); Rodacy, Philip J. (Albuquerque, NM); Simonson, Robert J. (Cedar Crest, NM)

2011-03-29T23:59:59.000Z

380

Pv-Thermal Solar Power Assembly  

DOE Patents [OSTI]

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.

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

2001-10-02T23:59:59.000Z

Note: This page contains sample records for the topic "test assembly irradiation" from the National Library of EnergyBeta (NLEBeta).
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We encourage you to perform a real-time search of NLEBeta
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381

Redevelopment and smart growth at Assembly Square  

E-Print Network [OSTI]

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 ...

Savage, Alice (Alice Augusta)

2006-01-01T23:59:59.000Z

382

Three-dimensional colorimetric assay assemblies  

DOE Patents [OSTI]

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.

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

2001-01-01T23:59:59.000Z

383

Automatically closing swing gate closure assembly  

DOE Patents [OSTI]

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

Chang, Shih-Chih (Richland, WA); Schuck, William J. (Richland, WA); Gilmore, Richard F. (Kennewick, WA)

1988-01-01T23:59:59.000Z

384

Miniature MT optical assembly (MMTOA)  

DOE Patents [OSTI]

An optical assembly (10) includes a rigid mount (12) with a recess (26) proximate a first side thereof, a substrate (14), and an optical die (16) flip-chip bonded to the substrate (14). The substrate (14) is secured to the first side of the mount and includes a plurality of die bonding elements (40), a plurality of optical apertures (32), and a plurality of external bonding elements (42). A plurality of traces (44) interconnect the die bonding elements (40) and the external bonding elements (42). The optical die (16) includes a plurality of optical elements, each element including an optical signal interface (48), the die being bonded to the plurality of die bonding elements (40) such that the optical signal interface (48) of each element is in registry with an optical aperture (32) of the substrate (14) and the die (16) is at least partially enclosed by the recess (26).

Laughlin, Daric (Overland Park, KS); Abel, Phillip (Overland Park, KS)

2008-04-01T23:59:59.000Z

385

Photonic-powered cable assembly  

DOE Patents [OSTI]

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.

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

2013-01-22T23:59:59.000Z

386

Nanocrystal assembly for tandem catalysis  

DOE Patents [OSTI]

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.

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

2014-10-14T23:59:59.000Z

387

Photonic-powered cable assembly  

DOE Patents [OSTI]

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.

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

2014-06-24T23:59:59.000Z

388

Snubber assembly for turbine blades  

DOE Patents [OSTI]

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.

Marra, John J

2013-09-03T23:59:59.000Z

389

Low inductance power electronics assembly  

DOE Patents [OSTI]

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.

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

2012-10-02T23:59:59.000Z

390

Initiation stress threshold irradiation assisted stress corrosion cracking criterion assessment for core internals in PWR environment  

SciTech Connect (OSTI)

Irradiation assisted stress corrosion cracking (IASCC) is a problem of growing importance in pressurized water reactors (PWR). An understanding of the mechanism(s) of IASCC is required in order to provide guidance for the development of mitigation strategies. One of the principal reasons why the IASCC mechanism(s) has been so difficult to understand is the inseparability of the different IASCC potential contributors evolutions due to neutron irradiation. The potential contributors to IASCC in PWR primary water are: (i) radiation induced segregation (RIS) at grain boundaries, (ii) radiation induced microstructure (formation and growth of dislocations loops, voids, bubbles, phases), (iii) localized deformation under loading, (iv) irradiation creep and transmutations. While the development of some of the contributors (RIS, microstructure) with increasing doses are at least qualitatively well understood, the role of these changes on IASCC remains unclear. Parallel to fundamental understanding developments relative to IASCC, well controlled laboratory tests on neutron irradiated stainless steels are needed to assess the main mechanisms and also to establish an engineering criterion relative to the initiation of fracture due to IASCC. First part of this study describes the methodology carried out at CEA in order to provide more experimental data from constant load tests dedicated to the study of initiation of SCC on neutron irradiated stainless steel. A description of the autoclave recirculation loop dedicated to SCC tests on neutron irradiated materials is then given. This autoclave recirculation loop has been started on July 2010 with the first SCC test on an irradiated stainless steel (grade 316) performed at CEA. The main steps of the interrupted SCC tests are then described. Second part of this paper reports the partial results of the first test performed on a highly neutron irradiated material. (authors)

Tanguy, Benoit; Stern, Anthony; Bossis, Philippe [CEA, DEN-DMN, Gif-sur-Yvette, (France); Pokor, Cedric [EDF les Renardieres, Moret-sur-Loing, (France)

2012-07-01T23:59:59.000Z

391

Modular fuel-cell stack assembly  

DOE Patents [OSTI]

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.

Patel, Pinakin (Danbury, CT)

2010-07-13T23:59:59.000Z

392

Liquid-liquid interfacial nanoparticle assemblies  

DOE Patents [OSTI]

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.

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

2008-12-30T23:59:59.000Z

393

Locking support for nuclear fuel assemblies  

DOE Patents [OSTI]

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

Ledin, Eric (San Diego, CA)

1980-01-01T23:59:59.000Z

394

In-cell reaction rate distributions and cell-average reaction rates in fast critical assemblies  

SciTech Connect (OSTI)

Measurements are described for determining average values of fission rates in /sup 235/U, /sup 238/U and /sup 239/Pu and capture rates in /sup 238/U for heterogeneous cells used to construct fast critical assemblies. The measurements are based on irradiations of foils of /sup 238/U, /sup 235/U and /sup 239/Pu with counting of fission and capture products using gamma-ray spectroscopy. Both plate and pin cells are considered. Procedures are described for inferring cell-average reaction rate values from a single foil location based on a cell using a quantity called a cell factor. Cell factors are determined from special measurements in which several foils are irradiated within a cell. Comparisons are presented between cell factors determined by measurements and by Monte Carlo calculations which lend credibility to the measurement procedures.

Brumbach, S.B.; Gasidlo, J.M.

1985-08-01T23:59:59.000Z

395

An evolutionary fuel assembly design for high power density BWRs  

E-Print Network [OSTI]

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 ...

Karahan, Aydin

2007-01-01T23:59:59.000Z

396

PCR - Ligation Assembly Standard for BioBrick Parts  

E-Print Network [OSTI]

This Request for Comments (RFC) describes a novel method for the assembly of standard BioBrick parts. This assembly method for BioBrick parts is an improvement upon the conventional methods of BioBrick part assembly. This ...

He, Tony PeiYuan

2011-12-15T23:59:59.000Z

397

Self-assembling functionalized single-walled carbon nanotubes  

E-Print Network [OSTI]

and L. Kavan (2009). "Self-Assemblies of Cationic PorphyrinsLi and Y. Ji (2005). "Self- assembly of base-functionalizedH. Bock (2008). "Directed Self-Assembly of Surfactants in

Gao, Yan

2011-01-01T23:59:59.000Z

398

Controlled Self Assembly of Conjugated Polymer Containing Block Copolymers  

E-Print Network [OSTI]

B. D. ; Segalman, R. A. , Self-assembly of rod-coil blockF. , Synthesis and Self- Assembly of Poly(diethylhexyloxy-p-I. , Three-dimensional self- assembly of rodcoil copolymer

McCulloch, Bryan

2012-01-01T23:59:59.000Z

399

Septin Self-Assembly: Plasticity and Protein Scaffolding  

E-Print Network [OSTI]

Septin Self-Assembly: Plasticity and Protein Scaffolding BySpring 2012 Septin Self-Assembly: Plasticity and ProteinIII Abstract Septin Self-Assembly: Plasticity and Protein

Garcia, III, Galo

2012-01-01T23:59:59.000Z

400

Solution Self-Assembly of Sequence Specific Biomimetic Polymers  

E-Print Network [OSTI]

J. D. , In vitro self- assembly from a simple protein ofR. N. , Hierarchical Self-Assembly of a Biomimetic DiblockC. R. , Hierarchical self-assembly of F-actin and cationic

Murnen, Hannah

2012-01-01T23:59:59.000Z

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


401

The AGR-1 Irradiation -Objectives, Success Criteria and Risk Management  

SciTech Connect (OSTI)

The AGR-1 experiment being conducted by the US Department of Energy Advanced Gas Reactor Fuel Development and Qualification Program (AGR fuel program) will irradiate TRISO-coated particle fuel in compacts under conditions representative of a Very High Temperature Reactor (VHTR) core. The anticipated fuel performance requirements of a prismatic core VHTR significantly exceed established TRISO-coated particle fuel capability in terms of burnup, temperature and fast fluence. AGR-1 is the first in a planned series of eight irradiations leading to the qualification of low enriched uranium coated particle fuel compacts for service in a VHTR, as identified in an overall Technical Program Plan produced at the beginning of the program . The AGR-1 experiment is scheduled for insertion in the Advanced Test Reactor (ATR) in the first quarter of fiscal year 2007 and to be irradiated for a period of up to approximately two and a half years. The irradiation rig, designated a "test train" is designed to provide six independently controlled (for temperature) and monitored (for fission product gas release) capsules containing fuel samples.

James Kendall

2006-06-01T23:59:59.000Z

402

Post Irradiation Capabilities at the Idaho National Laboratory  

SciTech Connect (OSTI)

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.

Schulthess, J.L.

2011-08-01T23:59:59.000Z

403

Post Irradiation Capabilities at the Idaho National Laboratory  

SciTech Connect (OSTI)

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.

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

2011-08-01T23:59:59.000Z

404

Neutron irradiation of beryllium pebbles  

SciTech Connect (OSTI)

Seven subcapsules from the FFTF/MOTA 2B irradiation experiment containing 97 or 100% dense sintered beryllium cylindrical specimens in depleted lithium have been opened and the specimens retrieved for postirradiation examination. Irradiation conditions included 370 C to 1.6 {times} 10{sup 22} n/cm{sup 2}, 425 C to 4.8 {times} 10{sup 22} n/cm{sup 2}, and 550 C to 5.0 {times} 10{sup 22} n/cm{sup 2}. TEM specimens contained in these capsules were also retrieved, but many were broken. Density measurements of the cylindrical specimens showed as much as 1.59% swelling following irradiation at 500 C in 100% dense beryllium. Beryllium at 97% density generally gave slightly lower swelling values.

Gelles, D.S.; Ermi, R.M. [Pacific Northwest National Lab., Richland, WA (United States); Tsai, H. [Argonne National Lab., IL (United States)

1998-03-01T23:59:59.000Z

405

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

SciTech Connect (OSTI)

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.

TOFFER, H.

2006-07-18T23:59:59.000Z

406

Microsoft Word - 911135_0 SSC-4a Reactor Core Test Plan_rel.doc  

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

5 Revision 0 ENGINEERING SERVICES FOR THE NEXT GENERATION NUCLEAR PLANT (NGNP) WITH HYDROGEN PRODUCTION Test Plan for the Reactor Core Assembly Prepared by General Atomics For the...

407

Irradiation effects on base metal and welds of 9Cr-1Mo (EM10) martensitic steel  

SciTech Connect (OSTI)

9Cr martensitic steels are being developed for core components (wrapper tubes) of fast breeder reactors as well as for fusion reactor structures. Here, the effects of fast neutron irradiation on the mechanical behavior of base metal and welds of 9Cr-1Mo (EM10) martensitic steel have been studied. Two types of weldments have been produced by TIG and electron beam techniques. Half of samples have been post-weld heat treated to produce a stress-relieved structure. The irradiation has been conducted in the Phenix reactor to doses of 63--65 dpa in the temperature range 450--459 C. The characterization of the welds, before and after irradiation, includes metallographic observations, hardness measurements, tensile and Charpy tests. It is shown that the mechanical properties of the welds after irradiation are in general similar to the characteristics obtained on the base metal, which is little affected by neutron irradiation.

Alamo, A.; Seran, J.L.; Rabouille, O.; Brachet, J.C.; Maillard, A.; Touron, H.; Royer, J. [CEA Saclay, Gif-sur-Yvette (France)

1996-12-31T23:59:59.000Z

408

Radioactive isotope production for medical applications using Kharkov electron driven subcritical assembly facility.  

SciTech Connect (OSTI)

Kharkov Institute of Physics and Technology (KIPT) of Ukraine has a plan to construct an accelerator driven subcritical assembly. The main functions of the subcritical assembly are the medical isotope production, neutron thereby, and the support of the Ukraine nuclear industry. Reactor physics experiments and material research will be carried out using the capabilities of this facility. The United States of America and Ukraine have started collaboration activity for developing a conceptual design for this facility with low enrichment uranium (LEU) fuel. Different conceptual designs are being developed based on the facility mission and the engineering requirements including nuclear physics, neutronics, heat transfer, thermal hydraulics, structure, and material issues. Different fuel designs with LEU and reflector materials are considered in the design process. Safety, reliability, and environmental considerations are included in the facility conceptual design. The facility is configured to accommodate future design improvements and upgrades. This report is a part of the Argonne National Laboratory Activity within this collaboration for developing and characterizing the subcritical assembly conceptual design. In this study, the medical isotope production function of the Kharkov facility is defined. First, a review was carried out to identify the medical isotopes and its medical use. Then a preliminary assessment was performed without including the self-shielding effect of the irradiated samples. Finally, more detailed investigation was carried out including the self-shielding effect, which defined the sample size and irradiation location for producing each medical isotope. In the first part, the reaction rates were calculated as the multiplication of the cross section with the unperturbed neutron flux of the facility. Over fifty isotopes were considered and all transmutation channels are used including (n,{gamma}), (n,2n), (n,p), and ({gamma},n). In the second part, the parent isotopes with high reaction rate were explicitly modeled in the calculations. For the nuclides with a very high capture microscopic cross section, such as iridium, rhenium, and samarium, their specific activities are reduced by a factor of 30 when the self-shielding effect is included. Four irradiation locations were considered in the analyses to maximize the medical isotope production rate. The results show the self-shield effect reduces the specific activity values and changes the irradiation location for obtaining the maximum possible specific activity. The axial and radial distributions of the specific activity were used to define the irradiation sample size for producing each isotope.

Talamo, A.; Gohar, Y.; Nuclear Engineering Division

2007-05-15T23:59:59.000Z

409

Device Assembly Facility (DAF) Glovebox Radioactive Waste Characterization  

SciTech Connect (OSTI)

The Device Assembly Facility (DAF) at the Nevada Test Site (NTS) provides programmatic support to the Joint Actinide Shock Physics Experimental Research (JASPER) Facility in the form of target assembly. The target assembly activities are performed in a glovebox at DAF and include Special Nuclear Material (SNM). Currently, only activities with transuranic SNM are anticipated. Preliminary discussions with facility personnel indicate that primarily two distributions of SNM will be used: Weapons Grade Plutonium (WG-Pu), and Pu-238 enhanced WG-Pu. Nominal radionuclide distributions for the two material types are included in attachment 1. Wastes generated inside glove boxes is expected to be Transuranic (TRU) Waste which will eventually be disposed of at the Waste Isolation Pilot Plant (WIPP). Wastes generated in the Radioactive Material Area (RMA), outside of the glove box is presumed to be low level waste (LLW) which is destined for disposal at the NTS. The process knowledge quantification methods identified herein may be applied to waste generated anywhere within or around the DAF and possibly JASPER as long as the fundamental waste stream boundaries are adhered to as outlined below. The method is suitable for quantification of waste which can be directly surveyed with the Blue Alpha meter or swiped. An additional quantification methodology which requires the use of a high resolution gamma spectroscopy unit is also included and relies on the predetermined radionuclide distribution and utilizes scaling to measured nuclides for quantification.

Dominick, J L

2001-12-18T23:59:59.000Z

410

assembly modelirovanie turbulentnogo: Topics by E-print Network  

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

has the same power to the square tile assembly system in computation, which is Turing universal. By providing counter-examples, we show that the triangular tile assembly...

411

assemblies proektnye parametry: Topics by E-print Network  

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

has the same power to the square tile assembly system in computation, which is Turing universal. By providing counter-examples, we show that the triangular tile assembly...

412

Biosensor Based on Self-Assembling Acetylcholinesterase on Carbon...  

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

Biosensor Based on Self-Assembling Acetylcholinesterase on Carbon Nanotubes for Flow injectionAmperometric Detection of Biosensor Based on Self-Assembling Acetylcholinesterase on...

413

Amperometric Glucose Biosensor Based on Self-Assembling Glucose...  

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

Glucose Biosensor Based on Self-Assembling Glucose Oxidase on Carbon Nanotubes. Amperometric Glucose Biosensor Based on Self-Assembling Glucose Oxidase on Carbon Nanotubes....

414

Green approach for self-assembly of platinum nanoparticles into...  

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

Green approach for self-assembly of platinum nanoparticles into nanowires in aqueous glucose solutions. Green approach for self-assembly of platinum nanoparticles into nanowires in...

415

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

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

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

416

Substrate Changes Associated with the Chemistry of Self-Assembled...  

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

Substrate Changes Associated with the Chemistry of Self-Assembled Monolayers on Silicon. Substrate Changes Associated with the Chemistry of Self-Assembled Monolayers on Silicon....

417

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...

418

Symmetry-Driven Spontaneous Self-assembly of Nanoscale Ceria...  

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

Symmetry-Driven Spontaneous Self-assembly of Nanoscale Ceria Building Blocks to Fractal Super-octahedra. Symmetry-Driven Spontaneous Self-assembly of Nanoscale Ceria Building...

419

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...

420

Size dependent specific surface area of nanoporous film assembled...  

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

Size dependent specific surface area of nanoporous film assembled by core-shell iron nanoclusters. Size dependent specific surface area of nanoporous film assembled by core-shell...

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


421

Vehicle Technologies Office Merit Review 2014: Hierarchical Assembly...  

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

Hierarchical Assembly of InorganicOrganic Hybrid Si Negative Electrodes Vehicle Technologies Office Merit Review 2014: Hierarchical Assembly of InorganicOrganic Hybrid Si...

422

alternative assembly pushes: Topics by E-print Network  

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

lead to better understanding and exploitation of self-assembled protein (more) Huard, Dustin Johnathen Edward 2012-01-01 440 Electron-beam Directed Materials Assembly MIT -...

423

A classification scheme for LWR fuel assemblies  

SciTech Connect (OSTI)

With over 100 light water nuclear reactors operating nationwide, representing designs by four primary vendors, and with reload fuel manufactured by these vendors and additional suppliers, a wide variety of fuel assembly types are in existence. At Oak Ridge National Laboratory, both the Systems Integration Program and the Characteristics Data Base project required a classification scheme for these fuels. This scheme can be applied to other areas and is expected to be of value to many Office of Civilian Radioactive Waste Management programs. To develop the classification scheme, extensive information on the fuel assemblies that have been and are being manufactured by the various nuclear fuel vendors was compiled, reviewed, and evaluated. It was determined that it is possible to characterize assemblies in a systematic manner, using a combination of physical factors. A two-stage scheme was developed consisting of 79 assembly types, which are grouped into 22 assembly classes. The assembly classes are determined by the general design of the reactor cores in which the assemblies are, or were, used. The general BWR and PWR classes are divided differently but both are based on reactor core configuration. 2 refs., 15 tabs.

Moore, R.S.; Williamson, D.A.; Notz, K.J.

1988-11-01T23:59:59.000Z

424

Flashback resistant pre-mixer assembly  

DOE Patents [OSTI]

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.

Laster, Walter R. (Oviedo, FL); Gambacorta, Domenico (Oviedo, FL)

2012-02-14T23:59:59.000Z

425

Low energy electron irradiation of an apple  

E-Print Network [OSTI]

The viability of pathogenic organisms on the surface of fresh fruits and vegetables can be significantly reduced by low energy electron beam irradiation. The most difficult technical challenge for surface irradiation of fruits and vegetable...

Brescia, Giovanni Batista

2002-01-01T23:59:59.000Z

426

Statistical criteria for characterizing irradiance time series.  

SciTech Connect (OSTI)

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.

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

2010-10-01T23:59:59.000Z

427

Effects of neutron irradiation on thermal conductivity of SiC-based composites and monolithic ceramics  

SciTech Connect (OSTI)

A variety of SiC-based composites and monolithic ceramics were characterized by measuring their thermal diffusivity in the unirradiated, thermal annealed, and irradiated conditions over the temperature range 400 to 1,000 C. The irradiation was conducted in the EBR-II to doses of 33 and 43 dpa-SiC (185 EFPD) at a nominal temperature of 1,000 C. The annealed specimens were held at 1,010 C for 165 days to approximately duplicate the thermal exposure of the irradiated specimens. Thermal diffusivity was measured using the laser flash method, and was converted to thermal conductivity using density data and calculated specific heat values. Exposure to the 165 day anneal did not appreciably degrade the conductivity of the monolithic or particulate-reinforced composites, but the conductivity of the fiber-reinforced composites was slightly degraded. The crystalline SiC-based materials tested in this study exhibited thermal conductivity degradation of irradiation, presumably caused by the presence of irradiation-induced defects. Irradiation-induced conductivity degradation was greater at lower temperatures, and was typically more pronounced for materials with higher unirradiated conductivity. Annealing the irradiated specimens for one hour at 150 C above the irradiation temperature produced an increase in thermal conductivity, which is likely the result of interstitial-vacancy pair recombination. Multiple post-irradiation anneals on CVD {beta}-SiC indicated that a portion of the irradiation-induced damage was permanent. A possible explanation for this phenomenon was the formation of stable dislocation loops at the high irradiation temperature and/or high dose that prevented subsequent interstitial/vacancy recombination.

Senor, D.J.; Youngblood, G.E. [Pacific Northwest National Lab., Richland, WA (United States); Moore, C.E. [Auburn Univ., AL (United States); Trimble, D.J. [Westinghouse Hanford Co., Richland, WA (United States); Woods, J.J. [Lockheed Martin, Schenectady, NY (United States)

1996-06-01T23:59:59.000Z

428

Effects of neutron irradiation on thermal conductivity of SiC-based composites and monolithic ceramics  

SciTech Connect (OSTI)

A variety of SiC-based composites and monolithic ceramics were characterized by measuring their thermal diffusivity in the unirradiated, thermal annealed, and irradiated conditions over the temperature range 400 to 1,000 C. The irradiation was conducted in the EBR-II to doses of 33 and 43 dpa-SiC (185 EFPD) at a nominal temperature of 1,000 C. The annealed specimens were held at 1,010 C for 165 days to approximately duplicate the thermal exposure of the irradiated specimens. Thermal diffusivity was measured using the laser flash method, and was converted to thermal conductivity using density data and calculated specific heat values. Exposure to the 165 day anneal did not appreciably degrade the conductivity of the monolithic or particulate-reinforced composites, but the conductivity of the fiber-reinforced composites was slightly degraded. The crystalline SiC-based materials tested in this study exhibited thermal conductivity degradation after irradiation, presumably caused by the presence of irradiation-induced defects. Irradiation-induced conductivity degradation was greater at lower temperatures, and was typically more pronounced for materials with higher unirradiated conductivity. Annealing the irradiated specimens for one hour at 150 C above the irradiation temperature produced an increase in thermal conductivity, which is likely the result of interstitial-vacancy pair recombination. Multiple post-irradiation anneals on CVD {beta}-SiC indicated that a portion of the irradiation-induced damage was permanent. A possible explanation for this phenomenon was the formation of stable dislocation loops at the high irradiation temperature and/or high dose that prevented subsequent interstitial/vacancy recombination.

Senor, D.J.; Youngblood, G.E. [Pacific Northwest National Lab., Richland, WA (United States); Moore, C.E. [Auburn Univ., AL (United States); Trimble, D.J. [Westinghouse Hanford Co., Richland, WA (United States); Woods, J.J. [Lockheed Martin, Schenectady, NY (United States)

1997-05-01T23:59:59.000Z

429

Gas Test Loop Functional and Technical Requirements  

SciTech Connect (OSTI)

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.

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

2004-09-01T23:59:59.000Z

430

Directed Self-Assembly of Nanodispersions  

SciTech Connect (OSTI)

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.

Furst, Eric M [University of Delaware] [University of Delaware

2013-11-15T23:59:59.000Z

431

E-Print Network 3.0 - assembly duct irradiated Sample Search...  

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

The "perfect... 71100-draft copy: do not quote 1 of 30 New Technologies for Residential HVAC Ducts Burke Treidler... February 1995 Executive Summary There are many problems with...

432

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

SciTech Connect (OSTI)

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.

Samuel J. Miller; Hakan Ozaltun

2012-11-01T23:59:59.000Z

433

3, 895959, 2006 Irradiance and  

E-Print Network [OSTI]

and corals. However, the contribution of benthic communities to the primary production of the global coastal energy source fueling marine primary prBGD 3, 895­959, 2006 Irradiance and primary production in the coastal ocean J.-P. Gattuso et al

Paris-Sud XI, Université de

434

sterilization by irradiation Arne Miller  

E-Print Network [OSTI]

-1:2006 Equipment characterization (6) Product definition (7) Process definition (8) Installation Qualification (9.1) Operational Qualification (9.2) · Performance Qualification (9.3) - later #12;3 Equipment characterization samples shall be irradiated to defined and uniform doses. #12;9 9.1 Installation qualification (A.9

435

High aspect ratio, remote controlled pumping assembly  

DOE Patents [OSTI]

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.

Brown, S.B.; Milanovich, F.P.

1995-11-14T23:59:59.000Z

436

Dual-axis resonance testing of wind turbine blades  

DOE Patents [OSTI]

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.

Hughes, Scott; Musial, Walter; White, Darris

2014-01-07T23:59:59.000Z

437

Irradiation-induced phenomena in carbon  

E-Print Network [OSTI]

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-induced phenomena in carbon nanotubes 1 1.1 Introduction

Krasheninnikov, Arkady V.

438

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

SciTech Connect (OSTI)

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.

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

2014-09-01T23:59:59.000Z

439

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

SciTech Connect (OSTI)

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

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

1996-12-31T23:59:59.000Z

440

Transition from Irradiation-Induced Amorphization to Crystallization...  

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

from Irradiation-Induced Amorphization to Crystallization in Nanocrystalline Silicon Carbide. Transition from Irradiation-Induced Amorphization to Crystallization in...

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441

Occlusion-Aware Hessians for Error Control in Irradiance Caching /  

E-Print Network [OSTI]

Control for Irradiance Caching. ” In ACM Transactions on Graphics,Control for Irradiance Caching. ” In ACM Transactions on Graphics,

Schwarzhaupt, Jorge Andres

2013-01-01T23:59:59.000Z

442

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

443

Irradiated Materials Examination and Testing Facility (IMET) | ORNL  

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

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444

Researchers Devise New Stress Test for Irradiated Materials | Department of  

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

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