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1

Lead Test Assembly Irradiation and Analysis Watts Bar Nuclear...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

2

Lead Test Assembly Irradiation and Analysis Watts Bar Nuclear Plant,  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Lead Test Assembly Irradiation and Analysis Watts Bar Nuclear Lead Test Assembly Irradiation and Analysis Watts Bar Nuclear Plant, Tennessee and Hanford Site, Richland, Washington Lead Test Assembly Irradiation and Analysis Watts Bar Nuclear Plant, Tennessee and Hanford Site, Richland, Washington SUMMARY 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 light water reactor to produce tritium. PUBLIC COMMENT OPPORTUNITIES None available at this time. DOCUMENTS AVAILABLE FOR DOWNLOAD July 22, 1997 EA-1210: Finding of No Significant Impact Lead Test Assembly Irradiation and Analysis Watts Bar Nuclear Plant, Tennessee and Hanford Site, Richland, Washington July 22, 1997 EA-1210: Final Environmental Assessment

3

Environmental Assessment LEAD TEST ASSEMBLY IRRADIATION AND ANALYSIS  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

10 10 Environmental Assessment LEAD TEST ASSEMBLY IRRADIATION AND ANALYSIS WATTS BAR NUCLEAR PLANT, TENNESSEE AND HANFORD SITE, RICHLAND, WASHINGTON U. S. DEPARTMENT OF ENERGY RICHLAND OPERATIONS OFFICE COOPERATING AGENCY: TENNESSEE VALLEY AUTHORITY July 1997 ~~~~ Portions o f this dorunrat may be iIlegiile in electronic image products. Images are produced from the best available original doaxnenL DOE/EA-12 10 Environmental Assessment LEAD TEST ASSEMBLY IRRADIATION AND ANALYSIS WATTS BAR NUCLEAR PLANT, TENNESSEE AND HANFORD SITE, RICHLAND, WASHINGTON U. S. DEPARTMENT OF ENERGY RICHLAND OPERATIONS OFFICE COOPERATING AGENCY: TENNESSEE VALLEY AUTHORITY July 1997 U.S. Department of Energy ALARA ANL-W BWR CFR CEDE CEQ Ci CLWR DOE DOT EA EDE EFPD EIS FFTF

4

FFTF utilization for irradiation testing  

SciTech Connect

FFTF utilization for irradiation testing is beginning. Two Fuels Open Test Assemblies and one Vibration Open Test Assembly, both containing in-core contact instrumentation, are installed in the reactor. These assemblies will be used to confirm plant design performance predictions. Some 100 additional experiments are currently planned to follow these three. This will result in an average core loading of about 50 test assemblies throughout the early FFTF operating cycles.

Corrigan, D.C.; Julyk, L.J.; Hoth, C.W.; McGuire, J.C.; Sloan, W.R.

1980-01-01T23:59:59.000Z

5

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.  

Science Conference Proceedings (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

6

FUEL ASSEMBLY SHAKER TEST SIMULATION  

SciTech Connect

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

7

Subsea production test valve assembly  

SciTech Connect

In the subsea test assembly securable within a blowout preventer stack above a subterranean well and positionable between upper and lower portions of a tubular conduit in fluid communication with a production zone within the well, the test assembly is described including an upper subassembly carriable with the upper conduit portion, a lower subassembly carriable with the lower conduit portion, and valve means in the lower subassembly manipulatable between opened and closed positions to control fluid flow within the conduit. The improvement comprises: the upper subassembly including an upper housing and first rigid dog means fixedly secured to the upper housing; the lower subassembly including a lower housing and second rigid dog means fixedly secured to the lower housing; the first rigid dog means positionable between a latch position for latching the upper and lower subassemblies and an unlatch position for unlatching the upper and lower subassemblies upon rotational movement of the first dog means with respect to the second dog means; and lock means axially movable relative to the first and second dog means from a lock position for limiting rotational movement of the first dog means with respect to the second dog means to an unlock position for allowing the first dog means to rotate relative to the second dog means and unlatch the upper subassembly from the lower subassembly.

Yates, P.D.

1988-03-22T23:59:59.000Z

8

Irradiation test program for FFTF  

SciTech Connect

Four unique deisgn features are described which make the Fast Flux Test Facility eminently suitable for irradiation test programs. These features are a fast flux level of 7 x 10/sup 15/ neutrons/cm/sup 2//sec, a 36-inch reference (breeder reactor) core height, test volumes suitable for testing of statistical quantities of materials, and the capability for direct (contact) or indirect (proximity) instrumentation of active core experiments.

Corrigan, D.C.; Last, G.A.

1978-06-18T23:59:59.000Z

9

Portable instrument for inspecting irradiated nuclear fuel assemblies  

DOE Patents (OSTI)

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

Nicholson, Nicholas (Los Alamos, NM); Dowdy, Edward J. (Los Alamos, NM); Holt, David M. (Los Alamos, NM); Stump, Jr., Charles J. (Santa Fe, NM)

1985-01-01T23:59:59.000Z

10

Thermal analysis of the FSP-1 fuel pin irradiation test  

SciTech Connect

Thermal analysis of a pin from the FSP-1 fuels irradiation test has been completed. The purpose of the analysis was to provide predictions of fuel pin temperatures, determine the flow regime within the lithium annulus of the test assembly, and provide a standardized model for a consistent basis of comparison between pins within the test assembly. The calculations have predicted that the pin is operating at slightly above the test design temperatures and that the flow regime within the lithium annulus is a laminar buoyancy driven flow. 7 refs., 5 figs.

Lyon, W.F. III.

1990-07-25T23:59:59.000Z

11

Measured and calculated isotopes for a gadolinia lead test assembly  

Science Conference Proceedings (OSTI)

The US Department of Energy, Duke Power Company, and the B and W Fuel Company participated in an extended burnup project to develop, irradiate, and examine an advanced fuel assembly design for pressurized water reactors. The assembly uses a urania-gadolinia (UO[sub 2]-Gd[sub 2]O[sub 3]) burnable absorber fuel mixture along with other fuel performance and design features that enhance uranium utilization. Previous milestones in the gadolinia development of the extended burnup project include development and verification of a neutronics model, measurement of materials properties of gadolinia fuel, and a successful gadolinia lead test assembly (LTA) program. One LTA was discharged as planned after one cycle, four LTAs continued for two more cycles, and one LTA of these four underwent a fourth cycle and reached 58,310 MWd/ton U assembly-average burnup, a world record at the time. Hot-cell destructive examination of gadolinia and non-gadolinia fuel rods from the single-cycle LTA (406.2 effective full-power days irradiation) has been completed. The comparison of measured and calculated isotopics for this LTA is the subject of this paper. A comparison of measured and calculated power distributions is also given, because accurate prediction of core performance during power production is ultimately the most important test of a calculational model.

Hove, C.M.

1990-01-01T23:59:59.000Z

12

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Fuel Assembly Shaker Test for Determining Loads on a PWR Assembly under Surrogate Normal Conditions of Truck Transport R0.1 Fuel Assembly Shaker Test for Determining Loads on a PWR...

13

MITG test assembly design and fabrication  

DOE Green Energy (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

14

Microstructural Characterization of Test Reactor Irradiated RPV ...  

Science Conference Proceedings (OSTI)

Presentation Title, Microstructural Characterization of Test Reactor Irradiated RPV ... Evolution in High Purity Reference V-4Cr-4Ti Alloy for Fusion Reactor.

15

TEST RESULTS FROM GAMMA IRRADIATION OF ALUMINUM OXYHYDROXIDES  

DOE Green Energy (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

16

Assembly for testing weldability of sheet metal  

DOE Patents (OSTI)

A test assembly for determining the weldability of sheet metal includes (1) a base having a flat side surface with an annular groove in the side surface, a counterbore being formed in the outer wall of the groove and the surface portion of the base circumscribed by the inner wall of the groove being substantially coplanar with the bottom of the counterbore, (2) a test disk of sheet metal the periphery of which is positioned in the counterbore and the outer surface of which is coplanar with one side of the base, and (3) a clamp ring overlying the side surface of the base and the edge portion of the test disk and a plurality of clamp screws which extend through the clamp ring for holding the periphery of the test disk against the bottom of the counterbore.

David, Stan A. (Knoxville, TN); Woodhouse, John J. (Crossville, TN)

1985-01-01T23:59:59.000Z

17

Evaluation of the advanced mixed oxide fuel test FO-2 irradiated in Fast Flux Test Facility  

SciTech Connect

The advanced mixed-oxide (UO/sub 2/-PuO/sub 2/) test assembly, FO-2, irradiated in the Fast Flux Test Facility (FFTF), is undergoing postirradiation examination (PIE). This is one of the first FFTF tests examined that used the advanced ferrite-martensite alloy, HT9, which is highly resistant to irradiation swelling. The FO-2 includes the first annular fueled pins irradiated in FFTF to undergo destructive examination. The FO-2 is a lead assembly for the ongoing FFTF Core Demonstration Experiment (CDE) (Leggett and Omberg 1987) and was designed to evaluate the effects of fuel design variables, such as pellet density, smeared density, and fuel form (annular or solid fuel), on advanced pin performance. The assembly contains a total of 169 fuel pins of twelve different types. The test was irradiated for 312 equivalent full power days (EFPD) in FFTF. It had a peak pin power of 13.7 kW/ft and reached a peak burnup of 65.2 MWd/kgM with a peak fast fluence of 9.9 /times/ 10/sup 22/ n/cm/sup 2/ (E > 0.1 MeV). This document discusses the test and its results. 6 refs., 19 figs., 4 tabs.

Gilpin, L.L.; Baker, R.B.; Chastain, S.A.

1989-05-01T23:59:59.000Z

18

AGR-1 Irradiation Experiment Test Plan  

DOE Green Energy (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

19

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

Science Conference Proceedings (OSTI)

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

NONE

1997-03-01T23:59:59.000Z

20

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

Science Conference Proceedings (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

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

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Assembly Shaker Test for Determining Loads on a PWR Assembly Assembly Shaker Test for Determining Loads on a PWR Assembly under Surrogate Normal Conditions of Truck Transport R0.1 Fuel Assembly Shaker Test for Determining Loads on a PWR Assembly under Surrogate Normal Conditions of Truck Transport R0.1 The United States 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 with the trend of nuclear power plants using reactor fuel for a longer time, creates questions concerning the ability of this aged, high-burnup fuel to withstand stresses and strains seen during normal conditions of transport from its current location to a future consolidated storage facility or permanent repository. UNFD R&D conducted testing employing surrogate instrumented

22

Evaluation of the advanced mixed-oxide fuel test FO-2 irradiated in the FFTF (Fast Flux Test Facility)  

SciTech Connect

The advanced mixed-oxide (UO{sub 2}-PuO{sub 2}) test assembly, FO-2, irradiated in the Fast Flux Test Facility (FFTF) is undergoing postirradiation examination. This is one of the first FFTF tests examined that used the advanced ferrite-martensite alloy, HT9, which is highly resistant to irradiation swelling. The FO-2 includes the first annular fueled pins irradiated in FFTF to undergo destructive examination. The FO-2 is a lead assembly for the ongoing FFTF Core Demonstration Experiment (CDE) and was designed to evaluate the effects of fuel design variables, such as pellet density, smeared density, and fuel form (annular or solid fuel), on advanced pin performance. The assembly contains a total of 169 fuel pins of 12 different types. Two L (annular) fuel pins, GF02L04 (FFTF and transient tested) and GF02L09 (FFTF only), were destructively examined. Evaluation of the FO-2 fuel pins and assembly shows the excellent and predictable performance of the mixed-oxide fuels with HT9 structural material. This, combined with the robust behavior of the pins in transient tests, and the continued excellent performance of the CDE indicate this is a superior fuel system for liquid-metal reactors. It offers greatly reduced deformation during irradiation, while maintaining good operating characteristics.

Burley Gilpin, L.L.; Chastain, S.A.; Baker, R.B.

1989-01-01T23:59:59.000Z

23

Test plan for the Parallex CANDU-MOX irradiation  

Science Conference Proceedings (OSTI)

One of several options being considered by the United States and the Russian Federation for the disposition of excess plutonium from dismantled weapons is to convert it to mixed-oxide (MOX) fuel for use in Canadian uranium-deuterium (CANDU) reactors. This report describes an irradiation test demonstrating the feasibility of this concept with laboratory quantities of MOX fuel placed in the pressurized loops of the National Research Universal test reactor at the Atomic Energy of Canada, Ltd., Chalk River Laboratories. The objective of the Parallex (for parallel experiment) test is to simultaneously test laboratory-produced quantities of US and R.F. MOX fuel in a test reactor under heat generation rates representing those expected in the CANDU reactors. The MOX fuel will be produced with plutonium from disassembled weapons at the Los Alamos National Laboratory in the United States and at the Bochvar Institute in the Russian Federation. Thus, the test will serve to demonstrate the accomplishment of many parts of the disposition mission: disassembly of weapons, conversion of the plutonium to oxide, fabrication of MOX fuel, assembly of fuel elements and bundles, shipment to a reactor, irradiation, and finally, storage of the spent fuel elements awaiting eventual disposition in a geologic repository in Canada.

Copeland, G.L.

1997-06-01T23:59:59.000Z

24

TEMPERATURE DEPENDANT BEHAVIOUR OBSERVED IN THE AFIP-6 IRRADIATION TEST  

Science Conference Proceedings (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

25

Irradiation Environment of the Materials Test Station  

SciTech Connect

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

Pitcher, Eric John [Los Alamos National Laboratory

2012-06-21T23:59:59.000Z

26

Education: Digital Resource Center - VIDEOS: Assembly and Test  

Science Conference Proceedings (OSTI)

Jul 20, 2007 ... Citation: "Assembly and Test." How Microprocessors Work. February 2007. How Stuff Works, Inc. Watch Video Edited: 7/2/2008 at 11:22 AM by ...

27

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, D.R.

1993-03-23T23:59:59.000Z

28

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

29

Assembly and installation of the large coil test facility test stand  

SciTech Connect

The Large Coil Test Facility (LCTF) was built to test six tokamak-type superconducting coils, with three to be designed and built by US industrial teams and three provided by Japan, Switzerland, and Euratom under an international agreement. The facility is designed to test these coils in an environment which simulates that of a tokamak. The heart of this facility is the test stand, which is made up of four major assemblies: the Gravity Base Assembly, the Bucking Post Assembly, the Torque Ring Assembly, and the Pulse Coil Assembly. This paper provides a detailed review of the assembly and installation of the test stand components and the handling and installation of the first coil into the test stand.

Queen, C.C. Jr.

1983-01-01T23:59:59.000Z

30

LWRS ATR Irradiation Testing Readiness Status  

SciTech Connect

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

31

Instrumentation to Enhance Advanced Test Reactor Irradiations  

SciTech Connect

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

32

Multiport riser and flange assemblies acceptance test report  

DOE Green Energy (OSTI)

This document presents the results of the acceptance test for the multiport riser (MPR) and multiport flange (MPF) assemblies. The accepted MPR and MPF assemblies will be used in support of the hydrogen mitigation project for double-shell waste tank 241-SY-101 and other related projects. The testing described in this document verifies that the mechanical and interface features are operating as designed and that the unit is ready for field service. The objectives of the acceptance testing were as follows: Basic equipment functions and mechanical interfaces were verified; Installation and removal of equipment were demonstrated to the degree possible; Operation of the decon spray system and all valving was confirmed; and the accumulated leak rate of the MPR and MPF assemblies was determined.

Precechtel, D.R.; Schroeder, B.K.

1994-09-12T23:59:59.000Z

33

Irradiation and Testing of Fuels and Cladding Materials  

Science Conference Proceedings (OSTI)

Mar 14, 2012 ... Mechanical Performance of Materials for Current and Advanced Nuclear Reactors: Irradiation and Testing of Fuels and Cladding Materials

34

Thermal analysis of the FSP-1RR irradiation test  

SciTech Connect

The thermal analysis of four unirradiated fuel pins to be tested in the FSP-1RR fuels irradiation experiment was completed. This test is a follow-on experiment in the series of fuel pin irradiation tests conducted by the SP-100 Program in the Fast Flux Test Facility. One of the pins contains several meltwire temperature monitors within the fuel and the Li annulus. A post-irradiation examination will verify the accuracy of the pre-irradiation thermal analysis. The purpose of the pre-irradiation analysis was to determine the appropriate insulating gap gas compositions required to provide the design goal cladding operating temperatures and to ensure that the meltwire temperature ranges in the temperature monitored pin bracket peak irradiation temperatures. This paper discusses the methodology and summarizes the results of the analysis.

Webb, R.H.; Lyon, W.F. III

1992-10-14T23:59:59.000Z

35

Researchers Devise New Stress Test for Irradiated Materials | Department of  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Researchers Devise New Stress Test for Irradiated Materials Researchers Devise New Stress Test for Irradiated Materials Researchers Devise New Stress Test for Irradiated Materials July 20, 2011 - 3:58pm Addthis Scientists conducted compression tests of copper specimens irradiated with high-energy protons, designed to model how damage from radiation affects the mechanical properties of copper. By using a specialized in situ mechanical testing device in a transmission electron microscope at the National Center for Electron Microscopy, the team could examine — with nanoscale resolution — the localized nature of this deformation. | Courtesy of Lawrence Berkeley National Laboratory Scientists conducted compression tests of copper specimens irradiated with high-energy protons, designed to model how damage from radiation affects

36

Researchers Devise New Stress Test for Irradiated Materials | Department of  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Researchers Devise New Stress Test for Irradiated Materials Researchers Devise New Stress Test for Irradiated Materials Researchers Devise New Stress Test for Irradiated Materials July 20, 2011 - 3:58pm Addthis Scientists conducted compression tests of copper specimens irradiated with high-energy protons, designed to model how damage from radiation affects the mechanical properties of copper. By using a specialized in situ mechanical testing device in a transmission electron microscope at the National Center for Electron Microscopy, the team could examine — with nanoscale resolution — the localized nature of this deformation. | Courtesy of Lawrence Berkeley National Laboratory Scientists conducted compression tests of copper specimens irradiated with high-energy protons, designed to model how damage from radiation affects

37

Reliability Testing the Die-Attach of CPV Cell Assemblies  

DOE Green Energy (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

38

Assessment of Initial Test Conditions for Experiments to Assess Irradiation  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Assessment of Initial Test Conditions for Experiments to Assess Assessment of Initial Test Conditions for Experiments to Assess Irradiation Assisted Stress Corrosion Cracking Mechanisms Assessment of Initial Test Conditions for Experiments to Assess Irradiation Assisted Stress Corrosion Cracking Mechanisms 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 Irradiation Assisted Stress Corrosion Cracking (IASCC) are unknown. Extended service conditions will increase the exposure

39

Hydrogen storage-bed design for tritium systems test assembly  

DOE Green Energy (OSTI)

The Los Alamos National Laboratory has completed the design of a hydrogen storage bed for the Tritium Systems Test Assembly (TSTA). Our objective is to store hydrogen isotopes as uranium hydrides and recover them by dehydriding. The specific use of the storage bed is to store DT gas as U(D,T)/sub 3/ when it is required for the TSTA. The hydrogen storage bed consists of a primary container in which uranium powder is stored and a secondary container for a second level of safety in gas confinement. The primary container, inlet and outlet gas lines, cartridge heaters, and instrumentation are assembled in the secondary container. The design of the hydrogen storage bed is presented, along with the modeling and analysis of the bed behavior during hydriding-dehydriding cycles.

Cullingford, H.S.; Wheeler, M.G.; McMullen, J.W.

1981-01-01T23:59:59.000Z

40

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

Science Conference Proceedings (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

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

Acceptance test report for the AN valve pit leak detection and low point drain assembly mock up test procedure  

SciTech Connect

This document describes The Performance Mock-up Test Procedure for the Valve Pit Leak Detection and Low Point Drain Assembly Performance Mock-Up Test Procedure.

EWER, K.L.

1999-07-20T23:59:59.000Z

42

Irradiated Materials Examination and Testing Facility (IMET) | ORNL  

NLE Websites -- All DOE Office Websites (Extended Search)

Irradiated Materials Examination and Testing Facility Irradiated Materials Examination and Testing Facility May 30, 2013 The Irradiated Material Examination and Testing (IMET) Facility was designed and built as a hot cell facility. It is a two-story block and brick structure with a two-story high bay that houses six heavily shielded cells and an array of sixty shielded storage wells. It includes the Specimen Prep Lab (SPL) with its associated laboratory hood and glove boxes, an Operating Area, where the control and monitoring instruments supporting the in-cell test equipment are staged, a utility corridor, a hot equipment storage area, a tank vault room, office space, a trucking area with access to the high bay, and an outside steel building for storage. The tests and examinations are conducted in six examination "hot" cells

43

Micro-bulge testing applied to neutron irradiated materials  

SciTech Connect

Micro-bulge testing was conducted on several Fe--Ni--Cr alloys irradiated as 0.3 mm thick disks to 10 dpa at 603 and 773 K in the Oak Ridge Research Reactor. Miniature tensile tests were performed on specimens of the same alloys irradiated concurrently. Good correlation between the tensile yield strength and the bulge yield load was observed in unirradiated specimens, however, the correlation was not simple for irradiated specimens. Good correlation was also observed between the ultimate tensile strength and the maximum bulge load. While irradiation produced a significant reduction in total elongation in the tensile test, irradiation caused only a small decrease in the deflection corresponding to the maximum bulge load compared to that observed on thinner disks used in earlier experiments. The results suggest that the thinner disk is better suited for ductility evaluations than the thicker disk. The area bounded by the load-deflection traces of the bulge tests shows a systematic variation with both alloy composition and irradiation condition which is not observed in the tensile data. It is anticipated that this parameter may prove useful in the evaluation of material toughness.

Okada, A. (Hokkaido Univ., Sapporo (Japan)); Hamilton, M.L.; Garner, F.A. (Pacific Northwest Lab., Richland, WA (USA))

1990-06-01T23:59:59.000Z

44

USE OF SILICON CARBIDE MONITORS IN ATR IRRADIATION TESTING  

Science Conference Proceedings (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

45

Irradiation qualification testing of SNAP-10A components  

SciTech Connect

Selected SNAP 10A components were irradiated to about 10{sup14} nvt and 5{times} 10{sup 7} r at an average temperature of 136{degrees}F in a nominal vacuum of 2 {times} 10{sup {minus}5} torr. The components were operated periodically and the electrical characteristics recorded. Pre-irradiationand post-irradiation tests were conducted. Catastropic degradation occurred only in the low-level neutron detection system and about 1.5 {times} 10{sup 13} nvt and in the high-level neutron power supply at about 6{times} 10{sup 12} nvt. Marginal degradation occurred in the fusistors and in the silicone rubber insert material in connectors. The relays, low-voltage trip devices, expansion compensator position demodulator, resistance thermometer sensor and bridge, and the gamma detection system opearted within their respective specifications during and after irradiation. The insulation resistance of all components was adeqauate during and after irradiation.

Chesavage, A.J.

1964-02-04T23:59:59.000Z

46

ORR irradiation experiment OF-1: accelerated testing of HTGR fuel  

SciTech Connect

The OF-1 capsule, the first in a series of High-Temperature Gas-Cooled Reactor fuel irradiations in the Oak Ridge Research Reactor, was irradiated for more than 9300 hr at full reactor power (30 MW). Peak fluences of 1.08 x 10/sup 22/ neutrons/cm/sup 2/ (> 0.18 MeV) were achieved. General Atomic Company's magazine P13Q occupied the upper two-thirds of the test space and the ORNL magazine OF-1 the lower one-third. The ORNL portion tested various HTGR recycle particles and fuel bonding matrices at accelerated flux levels under reference HTGR irradiation conditions of temperature, temperature gradient, and fast fluence exposure (> 0.18 MeV).

Tiegs, T.N.; Long, E.L. Jr.; Kania, M.J.; Thoms, K.R.; Allen, E.J.

1977-08-01T23:59:59.000Z

47

AGR-1 Irradiation Test Final As-Run Report  

SciTech Connect

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

48

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

SciTech Connect

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

49

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

SciTech Connect

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

50

Assembly and method for testing the integrity of stuffing tubes  

DOE Patents (OSTI)

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

Morrison, E.F.

1996-12-31T23:59:59.000Z

51

Updated FY12 Ceramic Fuels Irradiation Test Plan  

SciTech Connect

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

52

Gas-cooled fast breeder reactor steady-state irradiation testing program  

Science Conference Proceedings (OSTI)

The requirements for the gas-cooled fast breeder reactor irradiation program are specified, and an irradiation program plan which satisfies these requirements is presented. The irradiation program plan consists of three parts and includes a schedule and a preliminary cost estimate: (1) a steady-state irradiation program, (2) irradiations in support of the design basis transient test program, and (3) irradiations in support of the GRIST-2 safety test program. Data from the liquid metal fast breeder reactor program are considered, and available irradiation facilities are examined.

Acharya, R.T.; Campana, R.J.; Langer, S.

1980-08-01T23:59:59.000Z

53

Acceptance Test Plan for Fourth-Generation Hanford Corrosion Probe Tree Assembly  

SciTech Connect

This Acceptance Test Procedure (ATP) will document the satisfactory operation of the corrosion probe tree assembly. This ATP will be performed by the manufacturer prior to delivery to the site. The objective of this procedure is to demonstrate and document the acceptance of the corrosion probe tree assembly. The test will consist of a pressure test to verify leak tightness of the probe tree body, a continuity test of the probe tree wiring, a test of the high level detector wiring, and a test of the operation of the Type K thermocouples.

NORMAN, E.C.

2000-10-17T23:59:59.000Z

54

Report of Survey of the Los Alamos Tritium Systems Test Assembly Facility |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

the Los Alamos Tritium Systems Test Assembly the Los Alamos Tritium Systems Test Assembly Facility Report of Survey of the Los Alamos Tritium Systems Test Assembly Facility 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. The primary purpose of the survey is to identify facility conditions and issues that need to be addressed to transfer responsibility for the facility from the Office of Science (SC) to the Office of Environmental Management (EM). The second purpose is to provide EM with insight regarding the facility's risks and liabilities, which may influence the management of eventual downstream life-cycle activities. The survey and this report are part of a process for implementing the

55

Effects of laser irradiation on the self-assembly of MnAs nanoparticles in a GaAs matrix  

SciTech Connect

We investigate the effects of laser irradiation on the self-assembly of MnAs nanoparticles during solid-phase decomposition in a GaAs matrix. It is found that laser irradiation suppresses the growth of MnAs nanoparticles from small to large size, and that the median diameter D{sub 1} in the size distribution of small MnAs nanoparticles depends on the incident photon energy E following D{sub 1} {approx} E{sup -1/5}. We explain this behavior by the desorption of Mn atoms on the MnAs nanoparticle surface due to resonant optical absorption, in which incident photons excite intersubband electronic transitions between the quantized energy levels in the MnAs nanoparticles.

Hai, Pham Nam; Nomura, Wataru [Department of Electrical Engineering and Information Systems, University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8656 (Japan); Yatsui, Takashi; Ohtsu, Motoichi; Tanaka, Masaaki [Department of Electrical Engineering and Information Systems, University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8656 (Japan); Nanophotonics Research Center, University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8656 (Japan)

2012-11-05T23:59:59.000Z

56

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

DOE Green Energy (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

57

Final report on graphite irradiation test OG-2  

SciTech Connect

Results are presented of dimensional, thermal expansivity, thermal conductivity, Young's modulus, and tensile strength measurements on specimens of nuclear graphites irradiated in capsule OG-2. About half the irradiation space was allocated to H-451 near-isotropic petroleum-coke-based graphite or its subsized prototype grade H-429. Most of these specimens had been previously irradiated. Virgin specimens of another near-isotropic graphite, grade TS-1240, were irradiated. Some previously irradiated specimens of needle-coke-based H-327 graphite and pitch-coke-based P$sub 3$JHAN were also included.

Price, R.J.; Beavan, L.A.

1975-12-15T23:59:59.000Z

58

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

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

59

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

DOE Patents (OSTI)

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

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

1982-05-13T23:59:59.000Z

60

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

SciTech Connect

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

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

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

DOE Patents (OSTI)

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

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

1984-01-01T23:59:59.000Z

62

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

DOE Patents (OSTI)

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

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

1985-01-01T23:59:59.000Z

63

REPORT OF SURVEY OF THE LOS ALAMOS TRITIUM SYSTEMS TEST ASSEMBLY FACILITY  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

THE LOS ALAMOS TRITIUM THE LOS ALAMOS TRITIUM SYSTEMS TEST ASSEMBLY FACILITY U.S. Department of Energy Office of Environmental Management & Office of Science Report of Survey of the Los Alamos Tritium Systems Test Assembly Facility Rev. E (Final) October 3, 2000 Contents 1. Introduction 1.1 Purpose 1.2 Facility Description 1.3 Organization Representatives 1.4 Survey Participants 2. Summary, Conclusions & Recommendations 2.1 Comparison With LCAM Requirements 2.2 Transfer Considerations 2.3 Post-Transfer EM Path Forward & Management Risk 2.4 Post-Transfer S&M Reduction via Administrative Contamination Limit Revision 2.5 Stable Metal Tritides Consideration During D&D 3. Survey Results

64

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

DOE Green Energy (OSTI)

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

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

1987-04-01T23:59:59.000Z

65

Forced-convection boiling tests performed in parallel simulated LMR fuel assemblies  

SciTech Connect

Forced-convection tests have been carried out using parallel simulated Liquid Metal Reactor fuel assemblies in an engineering-scale sodium loop, the Thermal-Hydraulic Out-of-Reactor Safety facility. The tests, performed under single- and two-phase conditions, have shown that for low forced-convection flow there is significant flow augmentation by thermal convection, an important phenomenon under degraded shutdown heat removal conditions in an LMR. The power and flows required for boiling and dryout to occur are much higher than decay heat levels. The experimental evidence supports analytical results that heat removal from an LMR is possible with a degraded shutdown heat removal system.

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

1985-04-21T23:59:59.000Z

66

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

SciTech Connect

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

67

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

Science Conference Proceedings (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

68

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

Science Conference Proceedings (OSTI)

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

D. L. Knudson; J. L. Rempe

2012-02-01T23:59:59.000Z

69

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

Science Conference Proceedings (OSTI)

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

D. L. Knudson; J. L. Rempe

2012-02-01T23:59:59.000Z

70

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

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

71

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

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

72

Test plan for the irradiation of nonmetallic materials.  

SciTech Connect

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

73

Test plan for the irradiation of nonmetallic materials.  

SciTech Connect

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

74

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

SciTech Connect

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

75

Development of the JAERI (Japan Atomic Energy Research Institute) fuel cleanup system for tests at the Tritium Systems Test Assembly  

Science Conference Proceedings (OSTI)

Tritium Process Laboratory (TPL) at the Japan Atomic Energy Research Institute (JAERI) has developed the Fuel Cleanup System (FCU) which accepts simulated fusion reactor exhaust and produces pure hydrogen isotopes and tritium-free waste. The major components are: a palladium diffuser, a catalytic reactor, cold traps, a ceramic electrolysis cell, and zirconium-cobalt beds. In 1988, an integrated loop of the FCU process was installed in the TPL and a number of hot'' runs were performed to study the system characteristics and improve system performance. Under the US-Japan collaboration program, the JAERI Fuel Cleanup System'' (JFCU) was designed and fabricated by JAERI/TPL for testing at the Tritium Systems Test Assembly (TSTA) in Los Alamos National Laboratory as a major subsystem of the simulated fusion fuel cycle. The JFCU was installed in the TSTA in early 1990.

Konishi, S.; Inoue, M.; Hayashi, T.; Okuno, K.; Naruse, Y. (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)); Barnes, J.W.; Anderson, J.L. (Los Alamos National Lab., NM (USA))

1990-01-01T23:59:59.000Z

76

IAEA Activities on Modelling and Accelerated Irradiation Testing of ...  

Science Conference Proceedings (OSTI)

... Materials Evaluation for Nuclear Application Utilizing Test Reactors, Ion Beam ... and a CRP on accelerator simulation and theoretical modelling of radiation ...

77

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

SciTech Connect

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

78

Estimation of Critical Flow Velocity for Collapse of Gas Test Loop Booster Fuel Assembly  

Science Conference Proceedings (OSTI)

This paper presents calculations performed to determine the critical flow velocity for plate collapse due to static instability for the Gas Test Loop booster fuel assembly. Long, slender plates arranged in a parallel configuration can experience static divergence and collapse at sufficiently high coolant flow rates. Such collapse was exhibited by the Oak Ridge High Flux Reactor in the 1940s and the Engineering Test Reactor at the Idaho National Laboratory in the 1950s. Theoretical formulas outlined by Miller, based upon wide-beam theory and Bernoulli’s equation, were used for the analysis. Calculations based upon Miller’s theory show that the actual coolant flow velocity is only 6% of the predicted critical flow velocity. Since there is a considerable margin between the theoretically predicted plate collapse velocity and the design velocity, the phenomena of plate collapse due to static instability is unlikely.

Guillen; Mark J. Russell

2006-07-01T23:59:59.000Z

79

Short Term Irradiation Test of Fuel Containing Minor Actinides Using the Experimental Fast Reactor Joyo  

Science Conference Proceedings (OSTI)

A mixed oxide containing minor actinides (MA-MOX) fuel irradiation program is being conducted using the experimental fast rector Joyo of the Japan Atomic Energy Agency to research early thermal behavior of MA-MOX fuel. Two irradiation experiments were conducted as part of the short-term phase of this program in May and August 2006. Six prepared fuel pins included MOX fuel containing 3% or 5% americium (Am-MOX), and MOX fuel containing 2% americium and 2% neptunium (Np/Am-MOX). The first test was conducted with high linear heat rates of approximately 430 W/cm maintained during only 10 minutes. After 10 minutes irradiation test, the test subassembly was transferred to the hot cell facility and an Am-MOX pin and a Np/Am-MOX pin were replaced with dummy pins with neutron dosimeters. The test subassembly loaded with the remaining four fuel pins was re-irradiated in Joyo for 24-hours in August 2006 at nearly the same linear power to obtain re-distribution data on MA-MOX fuel. The linear heat rate for each MA-MOX test fuel pin was calculated using the Monte Carlo calculation code MCNP. The calculated fission rates were compared with the measured data based on the Nd-148 method. The maximum linear heat rate was approximately 444{+-}19 W/cm at the actual reactor power of 119.6 MWt. Post irradiation examination of these pins to confirm the absence of fuel melting and the local concentration under irradiation of NpO{sub 2-x} or AmO{sub 2-x}, in the (U,Pu)0{sub 2-x}, fuel are underway. The test results are expected to reduce uncertainties on the margin in the thermal design for MA-MOX fuel. (authors)

Sekine, Takashi; Soga, Tomonori; Koyama, Shin-ichi; Aoyama, Takafumi [Oarai Research and Development Center, Japan Atomic Energy Agency. 4002 Narita, Oarai, Ibaraki 311-1393 (Japan); Wootan, David [Pacific Northwest National Laboratoy, M/S K8-34, P.O. Box 999 Richland, WA 99352 (United States)

2007-07-01T23:59:59.000Z

80

Development of Correlations for Pressure Loss/Drop Coefficients Obtained From Flow Testing of Fuel Assemblies in Framatome ANP's PHTF  

SciTech Connect

Thermal-hydraulic analyses of pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies are generally performed for either assembly thermal-hydraulic design, thermal-hydraulic compatibility evaluation, or cycle licensing thermal-hydraulic characterization. A key issue in all cases is the hydraulic resistance characterization of the assembly in which the assembly, its components and support plates, etc., are represented by their respective pressure loss and pressure drop coefficients. These hydraulic coefficients can be determined by single-phase flow testing in an experimental facility such as the Framatome ANP Portable Hydraulic Test Facility (PHTF) located at Richland Test Facilities (RTF) in Richland, WA. The goal of this paper is to present a uniform and consistent methodology for the development of coefficient correlations from data obtained from single phase pressure drop testing of PWR and BWR fuel assemblies and their components performed in the PHTF. This methodology reflects the years of accumulated experience from an existing facility with an ongoing test program. (authors)

Madni, Imtiaz K.; Stephens, Lance G.; Turner, Dave M. [Framatome ANP Inc. (United States)

2002-07-01T23:59:59.000Z

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

Safety Assurance for Irradiating Experiments in the Advanced Test Reactor  

SciTech Connect

The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment.

T. A. Tomberlin; S. B. Grover

2004-11-01T23:59:59.000Z

82

Beam test results of the irradiated Silicon Drift Detector for ALICE  

E-Print Network (OSTI)

The Silicon Drift Detectors will equip two of the six cylindrical layers of high precision position sensitive detectors in the ITS of the ALICE experiment at LHC. In this paper we report the beam test results of a SDD irradiated with 1 GeV electrons. The aim of this test was to verify the radiation tolerance of the device under an electron fluence equivalent to twice particle fluence expected during 10 years of ALICE operation.

Kushpil, S; Giubellino, P; Idzik, M; Kolozhvari, A; Kushpil, V; Martínez, M I; Mazza, G; Mazzoni, A; Meddi, F; Nouais, D; Petracek, V; Piemonte, C; Rashevsky, A; Riccati, L; Rivetti, A; Tosello, F; Vacchi, A; Wheadon, R; PH-TH

2006-01-01T23:59:59.000Z

83

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

SciTech Connect

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

84

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

85

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

86

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

SciTech Connect

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

87

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

Science Conference Proceedings (OSTI)

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

Miller, William A [ORNL

2005-11-01T23:59:59.000Z

88

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

SciTech Connect

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

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

1990-12-01T23:59:59.000Z

89

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

Science Conference Proceedings (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

90

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

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

91

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

SciTech Connect

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

92

Technical specification: Mixed-oxide pellets for the light-water reactor irradiation demonstration test  

Science Conference Proceedings (OSTI)

This technical specification is a Level 2 Document as defined in the Fissile Materials Disposition Program Light-Water Reactor Mixed-oxide Fuel Irradiation Test Project Plan. It is patterned after the pellet specification that was prepared by Atomic Energy of Canada, Limited, for use by Los Alamos National Laboratory in fabrication of the test fuel for the Parallex Project, adjusted as necessary to reflect the differences between the Canadian uranium-deuterium reactor and light-water reactor fuels. This specification and the associated engineering drawing are to be utilized only for preparation of test fuel as outlined in the accompanying Request for Quotation and for additional testing as directed by Oak Ridge National Laboratory or the Department of Energy.

Cowell, B.S.

1997-06-01T23:59:59.000Z

93

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

SciTech Connect

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

94

Irradiated fuel monitoring by Cerenkov glow intensity measurements  

SciTech Connect

Attribute measurement techniques for confirmation of declared irradiated fuel inventories at nuclear installations under safeguards surveillance are being investigated. High-gain measurements of the intensity of the Cerenkov glow from exposed assemblies in water-filled storage ponds are promising for this purpose. Such measurements have been made of Materials Testing Reactor plate-type fuel assemblies and Pressurized Water Reactor pin-type fuel assemblies. The measured intensities depend on cooling times as calculations predict.

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

1979-09-01T23:59:59.000Z

95

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

SciTech Connect

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

96

A CONCEPTUAL DESIGN OF A SHIELD TESTING AND MATERIALS IRRADIATION FACILITY  

SciTech Connect

A conceptual design is presented for a test reactor facility to be used for shielding experiments and component irradintions necessary for airframe development for the nuclear airplane program. To meet both requirements a modified swimming-pool reactor is used, with a dry irradintion cell of 320 cu ft of useful volume provided for component testing, while shielding experiments are performed in the pool in the usual manner. A BSR-type core is operated at 1 MW to provide a fest neutron flux in the irradiation cell of 10/sup 12/n/cm/sup 2/ sec at the core face and 10/sup 11/at a distance of 4 feet. The irradiation-cell facility is designed to avoid the need of remote operations in making up service connections to the experimental piece. The reactor is contained in a cylindrical building designed for 6 psi internal pressure to meet the conditions of the maximum credible accident. The estimated cost of the facility, including the reactor and the fabrication cost for an initial fuel charge, is 874,000. (auth)

Frankfort, J.H.

1956-11-20T23:59:59.000Z

97

3D-FBK Pixel Sensors: Recent Beam Tests Results with Irradiated Devices  

SciTech Connect

The Pixel Detector is the innermost part of the ATLAS experiment tracking device at the Large Hadron Collider, and plays a key role in the reconstruction of the primary vertices from the collisions and secondary vertices produced by short-lived particles. To cope with the high level of radiation produced during the collider operation, it is planned to add to the present three layers of silicon pixel sensors which constitute the Pixel Detector, an additional layer (Insertable B-Layer, or IBL) of sensors. 3D silicon sensors are one of the technologies which are under study for the IBL. 3D silicon technology is an innovative combination of very-large-scale integration and Micro-Electro-Mechanical-Systems where electrodes are fabricated inside the silicon bulk instead of being implanted on the wafer surfaces. 3D sensors, with electrodes fully or partially penetrating the silicon substrate, are currently fabricated at different processing facilities in Europe and USA. This paper reports on the 2010 June beam test results for irradiated 3D devices produced at FBK (Trento, Italy). The performance of these devices, all bump-bonded with the ATLAS pixel FE-I3 read-out chip, is compared to that observed before irradiation in a previous beam test.

Micelli, A.; /INFN, Trieste /Udine U.; Helle, K.; /Bergen U.; Sandaker, H.; /Bergen U.; Stugu, B.; /Bergen U.; Barbero, M.; /Bonn U.; Hugging, F.; /Bonn U.; Karagounis, M.; /Bonn U.; Kostyukhin, V.; /Bonn U.; Kruger, H.; /Bonn U.; Tsung, J.W.; /Bonn U.; Wermes, N.; /Bonn U.; Capua, M.; /Calabria U.; Fazio, S.; /Calabria U.; Mastroberardino, A.; /Calabria U.; Susinno, G.; /Calabria U.; Gallrapp, C.; /CERN; Di Girolamo, B.; /CERN; Dobos, D.; /CERN; La Rosa, A.; /CERN; Pernegger, H.; /CERN; Roe, S.; /CERN /Prague, Tech. U. /Prague, Tech. U. /Freiburg U. /Freiburg U. /Freiburg U. /INFN, Genoa /Genoa U. /INFN, Genoa /Genoa U. /INFN, Genoa /Genoa U. /INFN, Genoa /Genoa U. /INFN, Genoa /Genoa U. /Glasgow U. /Glasgow U. /Glasgow U. /Hawaii U. /Barcelona, IFAE /Barcelona, IFAE /LBL, Berkeley /Barcelona, IFAE /LBL, Berkeley /LBL, Berkeley /Manchester U. /Manchester U. /Manchester U. /Manchester U. /Manchester U. /Manchester U. /Manchester U. /Manchester U. /Manchester U. /New Mexico U. /New Mexico U. /Oslo U. /Oslo U. /Oslo U. /Oslo U. /Oslo U. /SLAC /SLAC /SLAC /SLAC /SLAC /SLAC /SLAC /SLAC /SLAC /SUNY, Stony Brook /SUNY, Stony Brook /SUNY, Stony Brook /INFN, Trento /Trento U. /INFN, Trento /Trento U. /INFN, Trento /Trento U. /INFN, Trieste /Udine U. /INFN, Trieste /Udine U. /INFN, Trieste /Udine U. /INFN, Trieste /Udine U. /INFN, Trieste /Udine U. /INFN, Trieste /Udine U. /Barcelona, Inst. Microelectron. /Barcelona, Inst. Microelectron. /Barcelona, Inst. Microelectron. /Fond. Bruno Kessler, Trento /Fond. Bruno Kessler, Trento /Fond. Bruno Kessler, Trento /Fond. Bruno Kessler, Trento /Fond. Bruno Kessler, Trento /SINTEF, Oslo /SINTEF, Oslo /SINTEF, Oslo /SINTEF, Oslo /VTT Electronics, Espoo /VTT Electronics, Espoo

2012-04-30T23:59:59.000Z

98

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

DOE Green Energy (OSTI)

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

NONE

1995-07-01T23:59:59.000Z

99

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

SciTech Connect

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

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

1994-03-01T23:59:59.000Z

100

Development of a Fissile Materials Irradiation Capability for Advanced Fuel Testing at the MIT Research Reactor  

SciTech Connect

A fissile materials irradiation capability has been developed at the Massachusetts Institute of Technology (MIT) Research Reactor (MITR) to support nuclear engineering studies in the area of advanced fuels. The focus of the expected research is to investigate the basic properties of advanced nuclear fuels using small aggregates of fissile material. As such, this program is intended to complement the ongoing fuel evaluation programs at test reactors. Candidates for study at the MITR include vibration-packed annular fuel for light water reactors and microparticle fuels for high-temperature gas reactors. Technical considerations that pertain to the design of the MITR facility are enumerated including those specified by 10 CFR 50 concerning the definition of a research reactor and those contained in a separate license amendment that was issued by the U.S. Nuclear Regulatory Commission to MIT for these types of experiments. The former includes limits on the cross-sectional area of the experiment, the physical form of the irradiated material, and the removal of heat. The latter addresses experiment reactivity worth, thermal-hydraulic considerations, avoidance of fission product release, and experiment specific temperature scrams.

Hu Linwen; Bernard, John A.; Hejzlar, Pavel; Kohse, Gordon [Massachusetts Institute of Technology (United States)

2005-05-15T23:59:59.000Z

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101

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

SciTech Connect

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

102

Results and Analyses of Irradiation/Anneal Experiments Conducted on Yankee Rowe Reactor Pressure Vessel Surrogate Materials: Yankee Atomic Electric Company Test Reactor Program  

Science Conference Proceedings (OSTI)

Many variables influence the response of reactor vessel steels to neutron irradiation. This study looks at the influence of irradiation temperature, steel heat treatment and microstructure, and nickel and phosphorus content on the irradiation response of high-copper reactor vessel steel. Also addressed are several studies evaluating the potential of thermal annealing to restore the mechanical properties of the steels tested.

1996-03-22T23:59:59.000Z

103

High-Voltage Direct Current Corona Testing of Transmission Line Hardware and Insulator Assemblies: Development of Test Methodology  

Science Conference Proceedings (OSTI)

When specifying hardware for new high-voltage direct current (HVDC) lines or replacement hardware for existing HVDC lines utilities generally require that the hardware meet specific corona performance requirements.While standards and test methods exist for testing hardware used on HVAC systems, no such material is available for HVDC systems.HVAC tests are sometimes conducted on the hardware and the results obtained are then related to HVDC by utilizing the peak HVAC line to ...

2013-12-20T23:59:59.000Z

104

Lead Fuel Assembly Programs Analysis: Utility Perspectives  

Science Conference Proceedings (OSTI)

Licensees, in association with nuclear fuel vendors, conduct lead fuel assembly (LFA) programs to test new design features prior to batch implementation. A limited number of LFAs are irradiated to obtain data and to confirm successful operation in the host reactor environment. The new LFA design features range from minor changes of dimensions and/or materials to an entirely new design from an alternate fuel vendor. LFA program elements can consist of design activities, methods development, analysis, ...

2013-10-17T23:59:59.000Z

105

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

Science Conference Proceedings (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

106

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

SciTech Connect

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

107

Numerical Analysis of the Resistance to Shear Test of Clinched Assemblies of Thin Metal Sheets  

Science Conference Proceedings (OSTI)

The work presented in this paper is part of a more extensive study the aim of which is to build a complete simulation of the clinching process and subsequent resistance tests. This paper focuses on finite element analyses - that are performed with the ABAQUS code - of the resistance of clinched points to shear test. These analyses are run up to propagation of metal sheet fracture. A simplified procedure is proposed to identify the fracture initiation and propagation models that are used to simulate this failure process. This identification process is based on Lemaitre's ductile damage model. The numerical simulations of a shear test have been compared to experimental results.

Jomaa, Moez; Billardon, Rene [LMT-Cachan - ENS de Cachan / CNRS (UMR 8535) / Universite Paris 6, 61, avenue du President Wilson F-94235 Cachan Cedex (France)

2007-05-17T23:59:59.000Z

108

Preliminary results from Charpy impact testing of irradiated JPDR weld metal and commissioning of a facility for machining of irradiated materials  

Science Conference Proceedings (OSTI)

Forty two full-size Charpy specimens were machined from eight trepans that originated from the Japan Power Demonstration Reactor (JPDR). They were also successfully tested and the preliminary results are presented in this report. The trends appear to be reasonable with respect to the location of the specimens with regards to whether they originated from the beltline or the core regions of the vessel, and also whether they were from the inside or outside regions of the vessel wall. A short synopsis regarding commissioning of the facility to machine irradiated materials is also provided.

Iskander, S.K.; Hutton, J.T.; Creech, L.E.; Nanstad, R.K.; Manneschmidt, E.T.; Rosseel, T.M.; Bishop, P.S.

1999-09-01T23:59:59.000Z

109

Assembly and Field Testing of a Ground-Based Presence-of-Cloud Detector  

Science Conference Proceedings (OSTI)

A presence-of-cloud (POC) detector has been developed for use in remote locations. The principal components of the POC detector are a moisture-sensitive resistance grid, a heater, a fan, and housing with rain shielding. Field testing at a ...

D. O. Krovetz; M. A. Reiter; J. T. Sigmon; F. S. Gilliam

1988-08-01T23:59:59.000Z

110

Fast and Thermal Data Testing of LEU, IEU, and HEU Critical Assemblies  

SciTech Connect

The purpose of this paper is to report on data testing of the ENDF/B-VI, release 5, evaluation for LEU, IEU, and HEU benchmarks. In terms of the energy spectrum, there are 10 fast, 3 intermediate, and 21 thermal cases. The characteristics of each benchmark are discussed briefly. The SCALE system (either XSDRN or KENOV.a) with the VITAMIN-B6 (199-group) cross section library were utilized. Hydrogen and U235 from the ENDF/B-VI, release 5, were used in the calculations.

Leal, L.C.; Wright, R.Q.

1999-09-20T23:59:59.000Z

111

Tritium experiments on components for fusion fuel processing at the Tritium Systems Test Assembly  

SciTech Connect

Under a collaborative agreement between US and Japan, two tritium processing components, a palladium diffuser and a ceramic electrolysis cell have been tested with tritium for application to a Fuel Cleanup System (FCU) for plasma exhaust processing at the Los Alamos National Laboratory. The fundamental characteristics, compatibility with tritium, impurities effects with tritium, and long-term behavior of the components, were studied over a three year period. Based on these studies, an integrated process loop, JAERI Fuel Cleanup System'' equipped with above components was installed at the TSTA for full scale demonstration of the plasma exhaust reprocessing.

Konishi, S.; Yoshida, H.; Naruse, Y. (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)); Carlson, R.V.; Binning, K.E.; Bartlit, J.R.; Anderson, J.L. (Los Alamos National Lab., NM (USA))

1990-01-01T23:59:59.000Z

112

Testing of coatings for the nuclear industry  

SciTech Connect

Coatings for commercial nuclear power plants need to withstand humidity, radiation exposure, and LOC accident conditions; they also must be decontaminable. Tests for decontaminability, radiation stability, and design-basis-accident (DBA) resistance are described. An irradiation test facility using spent fuel assemblies and a spray loop for simulating a DBA are described. A sample test report sheet is presented. (DLC)

Goldberg, G.

1975-01-01T23:59:59.000Z

113

TOB Module Assembly  

NLE Websites -- All DOE Office Websites (Extended Search)

SiTracker Home Page Participating Institutions and Principal Contacts Useful Links Notes Images TOB Module Assembly and Testing Project TOB Integration Data Tracker Offline DQM LHC Fluence Calculator Total US Modules Tested Graph Total US Modules Tested Graph Total US Modules Tested Total US Modules Tested US Modules Tested Graph US Modules Tested Graph US Modules Tested US Modules Tested Rod Assembly TOB Modules on a Rod TOB Rod Insertion Installation of a TOB Rod Completed TOB Completed Tracker Outer Barrel TOB Module Assembly and Testing Project All 5208 modules of the CMS Tracker Outer Barrel were assembled and tested at two production sites in the US: the Fermi National Accelerator Laboratory and the University of California at Santa Barbara. The modules were delivered to CERN in the form of rods, with the last shipment taking

114

In-Pile Tests for IASCC Growth Behavior of Irradiated 316L Stainless ...  

Science Conference Proceedings (OSTI)

Crack Growth Rates of Irradiated Commercial Stainless Steels in BWR and PWR ... Detailed Root Cause Analysis of SG Tube ODSCC Indications within the Tube Sheets of NPP Biblis Unit A .... Radiation Damage in Fe-C-Met Model Alloys ... Stress Corrosion Cracking Behavior near the Fusion Boundary of Dissimilar Weld

115

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

E-Print Network (OSTI)

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

Martin Nagel

2013-09-03T23:59:59.000Z

116

Experimental tests of irradiation-anneal-reirradiation effects on mechanical properties of RPV plate and weld materials  

Science Conference Proceedings (OSTI)

The Charpy-V (C{sub V}) notch ductility and tension test properties of three reactor pressure vessel (RPV) steel materials were determined for the 288{degree}C (550{degree}F) irradiated (I), 288{degree}C (550{degree}F) irradiated + 454{degree}C (850{degree}F)-168 h postirradiation annealed (IA), and 288{degree}C (550{degree}F) reirradiated (IAR) conditions. Total fluences of the I condition and the IAR condition were, respectively, 3.33 {times} 10{sup 19} n/cm{sup 2} and 4.18 {times} 10{sup 19} n/cm{sup 2}, E > 1 MeV. The irradiation portion of the IAR condition represents an incremental fluence increase of 1. 05 {times} 10{sup 19} n/cm{sup 2}, E > 1 MeV, over the I-condition fluence. The materials (specimens) were supplied by the Yankee Atomic Electric Company and represented high and low nickel content plates and a high nickel, high copper content weld deposit prototypical of the Yankee-Rowe reactor vessel. The promise of the IAR method for extending the fluence tolerance of radiation-sensitive steels and welds is clearly shown by the results. The annealing treatment produced full C{sub V} upper shelf recovery and full or nearly full recovery in the C{sub V} 41 J (30 ft-lb) transition temperature. The C{sub V} transition temperature increases produced by the reirradiation exposure were 22% to 43% of the increase produced by the first cycle irradiation exposure. A somewhat greater radiation embrittlement sensitivity and a somewhat greater reirradiation embrittlement sensitivity was exhibited by the low nickel content plate than the high nickel content plate. Its high phosphorus content is believed to be responsible. The IAR-condition properties of the surface vs. interior regions of the low nickel content plate are also compared.

Hawthorne, J.R. [Materials Engineering Associates, Inc., Lanham, MD (United States)

1996-01-01T23:59:59.000Z

117

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

Science Conference Proceedings (OSTI)

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

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

2012-01-01T23:59:59.000Z

118

Comparison of fracture behavior for low-swelling ferritic and austenitic alloys irradiated in the Fast Flux Test Facility (FFTF) to 180 DPA  

Science Conference Proceedings (OSTI)

Fracture toughness testing was conducted to investigate the radiation embrittlement of high-nickel superalloys, modified austenitic steels and ferritic steels. These materials have been experimentally proven to possess excellent resistance to void swelling after high neutron exposures. In addition to swelling resistance, post-irradiation fracture resistance is another important criterion for reactor material selection. By means of fracture mechanics techniques the fracture behavior of those highly irradiated alloys was characterized in terms of irradiation and test conditions. Precipitation-strengthened alloys failed by channel fracture with very low postirradiation ductility. The fracture toughness of titanium-modified austenitic stainless steel D9 deteriorates with increasing fluence to about 100 displacement per atom (dpa), the fluence level at which brittle fracture appears to occur. Ferritic steels such as HT9 are the most promising candidate materials for fast and fusion reactor applications. The upper-shelf fracture toughness of alloy HT9 remained adequate after irradiation to 180 dpa although its ductile- brittle transition temperature (DBTT) shift by low temperature irradiation rendered the material susceptible to brittle fracture at room temperature. Understanding the fracture characteristics under various irradiation and test conditions helps reduce the potential for brittle fracture by permitting appropriate measure to be taken.

Huang, F.H.

1992-02-01T23:59:59.000Z

119

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

Science Conference Proceedings (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

120

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

Science Conference Proceedings (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

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

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

High speed door assembly  

DOE Patents (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

123

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

E-Print Network (OSTI)

) The RGHB performs advanced thermal testing of full-size wall/fenestration systems. It accommodates systems content in materials, vapor pressure, temperature, heat flux, humidity, and condensation. 7. MAXLAB MAXLAB. It is adequate for testing in most residential and light commercial buildings. 12. Duct Blaster A Duct Blaster

Oak Ridge National Laboratory

124

Summary report on the HFED (High-Uranium-Loaded Fuel Element Development) miniplate irradiations for the RERTR (Reduced Enrichment Research and Test Reactor) Program  

SciTech Connect

An experiment to evaluate the irradiation characteristics of various candidate low-enriched, high-uranium content fuels for research and test reactors was performed for the US Department of Energy Reduced Enrichment Research and Test Reactor Program. The experiment included the irradiation of 244 miniature fuel plates (miniplates) in a core position in the Oak Ridge Research Reactor. The miniplates were aluminum-based, dispersion-type plates 114.3 mm long by 50.8 mm wide with overall plate thicknesses of 1.27 or 1.52 mm. Fuel core dimensions varied according to the overall plate thicknesses with a minimum clad thickness of 0.20 mm. Tested fuels included UAl/sub x/, UAl/sub 2/, U/sub 3/O/sub 8/, U/sub 3/SiAl, U/sub 3/Si, U/sub 3/Si/sub 1.5/, U/sub 3/Si/sub 2/, U/sub 3/SiCu, USi, U/sub 6/Fe, and U/sub 6/Mn/sub 1.3/ materials. Although most miniplates were made with low-enriched uranium (19.9%), some with medium-enriched uranium (40 to 45%), a few with high-enriched uranium (93%), and a few with depleted uranium (0.2 to 0.4%) were tested for comparison. These fuel materials were irradiated to burnups ranging from /approximately/27 to 98 at. % /sup 235/U depletion. Operation of the experiment, measurement of miniplate thickness as the irradiation progressed, ultimate shipment of the irradiated miniplates to various hot cells, and preliminary results are reported here. 18 refs., 12 figs., 7 tabs.

Senn, R.L.

1989-04-01T23:59:59.000Z

125

Miniaturized disk-bend testing, nano-indentation and the microstructure of ion-irradiated titanium aluminides  

SciTech Connect

Effect of ion irradiation on microstructures and mechanical properties of Ti-52Al (TiAl) and Ti-26Al (Ti{sub 3}Al) were investigated. The alloys were irradiated with 2 MeV protons and Ar{sup +} ions at low temperatures ({minus}175 to {minus}135C) to max fluences of 4.5 {times} 10{sup 15} Ar{sup +}/cm{sup 2} and 2 {times} 10{sup 17} H{sup +}/cm{sup 2}. Yield strengths of unirradiated TiAl and the fracture strength of unirradiated Ti{sub 3}Al were 367 {plus_minus} 33 MPa and 536 {plus_minus} 30 MPa, respectively, in excellent agreement with published data. Yield strength of TiAl and fracture strength of Ti{sub 3}Al increased as a result of irradiation. Strengths of both alloys were lowest for the samples irradiated to the highest Ar{sup +}-ion dose, but otherwise there was no correlation of strength with dose. The nanohardness of irradiated specimens generally increased with dose, but influence of dose on Young`s modulus was erratic. Plate-shaped defects vacancy in character, and helical dislocations were observed in irradiated TiAl by TEM. The Ar{sup +}-ion irradiation-induced microstructure of Ti{sub 3}Al contained defects producing mottled contrast at 1 dpa and black-spot contrast at 5 dpa. Irradiation-induced loss of long-range order was also observed in both alloys. Influence of these microstructural variables on mechanical behavior is discussed.

Petouhoff, N.L.; Ardell, A.J. [California Univ., Los Angeles, CA (United States). Dept. of Materials Science and Engineering; Oliver, W.C.; Lucas, B.N. [Oak Ridge National Lab., TN (United States)

1993-09-01T23:59:59.000Z

126

Tensile Hoop Behavior of Irradiated Zircaloy-4 Nuclear Fuel Cladding  

SciTech Connect

A method for evaluating the room temperature ductility behavior of irradiated Zircaloy-4 nuclear fuel cladding has been developed and applied to evaluate tensile hoop strength of material irradiated to different levels. The test utilizes a polyurethane plug fitted within a tubular cladding specimen. A cylindrical punch is used to compress the plug axially, which generates a radial displacement that acts upon the inner diameter of the specimen. Position sensors track the radial displacement of the specimen outer diameter as the compression proceeds. These measurements coupled with ram force data provide a load-displacement characterization of the cladding response to internal pressurization. The development of this simple, cost-effective, highly reproducible test for evaluating tensile hoop strain as a function of internal pressure for irradiated specimens represents a significant advance in the mechanical characterization of irradiated cladding. In this project, nuclear fuel rod assemblies using Zircaloy-4 cladding and two types of mixed uranium-plutonium oxide (MOX) fuel pellets were irradiated to varying levels of burnup. Fuel pellets were manufactured with and without thermally induced gallium removal (TIGR) processing. Fuel pellets manufactured by both methods were contained in fuel rod assemblies and irradiated to burnup levels of 9, 21, 30, 40, and 50 GWd/MT. These levels of fuel burnup correspond to fast (E > 1 MeV) fluences of 0.27, 0.68, 0.98, 1.4 and 1.7 1021 neutrons/cm2, respectively. Following irradiation, fuel rod assemblies were disassembled; fuel pellets were removed from the cladding; and the inner diameter of cladding was cleaned to remove residue materials. Tensile hoop strength of this cladding material was tested using the newly developed method. Unirradiated Zircaloy-4 cladding was also tested. With the goal of determining the effect of the two fuel types and different neutron fluences on clad ductility, tensile hoop strength tests were performed on cladding for these varying conditions. Experimental data revealed negligible performance differences for cladding containing TIGR vs non-TIGR processed fuel pellets. Irradiation hardening was observed in tensile hoop data as the strength of the cladding increased with increasing neutron dose and appeared to saturate for a fast fluence of 1.7 1021 neutrons/cm2.

Jaramillo, Roger A [ORNL; Hendrich, WILLIAM R [ORNL; Packan, Nicolas H [ORNL

2007-03-01T23:59:59.000Z

127

Intergranular Cracking of an Irradiated Ti-Stabilized Austenitic ... - TMS  

Science Conference Proceedings (OSTI)

Aug 1, 1999 ... Environmental Degradation of Materials in Nuclear Power ... Failure of an irradiated fuel assembly spacer grid sleeve was observed after three ...

128

Response comparison of a single-diode electronic dosimeter, a three-diode electronic dosimeter, and a conventional four-filter TLD assembly in several irradiation environments  

E-Print Network (OSTI)

This study was performed in order to determine and compare the response of several different dosimetry media in various exposure categofies. In order to justify the use of an electronic dosimeter as an adequate badge of record, an electronic dosimeter must be on par with accepted external dosimetry standards. Additionally, recent studies determined that a multiple diode electronic dosimeter may be considered the best candidate for use as a badge of record. The response of two commercially available electronic dosimeters, a single diode and a three-diode configuration, and a four-filter TLD packet are compared in this investigation. The exposure categofies include dose output linearity, dose rate linearity, angular dependence, incident photon energy dependence, and noble gas fission product exposure testing. These exposure categories are meant to simulate most conditions encountered in an operational setting. The responses of each dosimeter are compared to applicable industry standards. The major results include electronic dosimeter underresponse at 112 rem min-', single-diode electronic dosimeter underresponse for 100-200 keV photons, excellent deep dose agreement between the three dosimeters in the noble gas environment, and shallow dose disagreement between two TLD algorithm and the three-diode electronic dosimeter.

Charlton, Michael Aaron

1995-01-01T23:59:59.000Z

129

Gamma Irradiation | ORNL Neutron Sciences  

NLE Websites -- All DOE Office Websites (Extended Search)

Gamma Irradiation Gamma Irradiation Gamma Irradiation Facility Gamma irradiation chamber Gamma irradiation chamber. The HFIR Gamma Irradiation Facility is an experimental facility designed to irradiate materials with gamma radiation from spent fuel elements. The facility chamber is stainless steel and is made of 0.065-thick tubing to maximize the internal dimensions of the chamber. This allows for the largest samples possible that can still fit inside the cadmium post of the spent fuel loading station positions. The interior chamber is approximately 3.75 inches inside diameter and accommodates samples up to 25 inches long. There are two configurations for the chamber assembly, with the only difference being the plugs. The uninstrumented configuration has a top plug that is used for installation of the samples, to support the inert gas

130

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

131

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

132

DISSOLUTION OF IRRADIATED MURR FUEL ASSEMBLIES  

DOE Green Energy (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

133

Specific Heat Measurements and Post-Test Characterization of Irradiated and Unirradiated Urania and Gadolinia Doped Fuel  

Science Conference Proceedings (OSTI)

In pursuit of higher burnups at nuclear plants, fuel designers have introduced the use of 'advanced' fuel types, including doped fuels. Completing a systematic program to acquire data on the basic properties of these fuels, this project measured the specific heat and density of high burn-up UO2 and (U, Gd)O2 using irradiated materials of the same origin as those on which thermal diffusivity measurements had previously been made and thermal recovery phenomena investigated.

2000-12-31T23:59:59.000Z

134

The second and third NGNP advanced gas reactor fuel irradiation experiments  

SciTech Connect

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

135

Seal assembly  

DOE Patents (OSTI)

A seal assembly comprises a tube rotatable about its longitudinal axis and having two longitudinally spaced flanges projecting radially outwardly from the outer surface thereof. Slidably positioned against one of the flanges is a seal ring, and disposed between this seal ring and the other flange are two rings that are forced apart by springs, one of the latter rings being attached to a flexible wall.

Morgan, J.G.; Rennich, M.J.; Whatley, M.E.

1981-03-13T23:59:59.000Z

136

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

137

Solids irradiator  

DOE Patents (OSTI)

A novel facility for irradiation of solids embodying pathogens wherein solids are conveyed through an irradiation chamber in individual containers of an endless conveyor.

Morris, Marvin E. (Albuquerque, NM); Pierce, Jim D. (Albuquerque, NM); Whitfield, Willis J. (Albuquerque, NM)

1979-01-01T23:59:59.000Z

138

Bremsstrahlung {gamma}-ray generation by electrons from gas jets irradiated by laser pulses for radiographic testing  

Science Conference Proceedings (OSTI)

Electron generation from a gas jet irradiated by low energy femtosecond laser pulses is studied experimentally as a promising source of radiation for radioisotope-free {gamma}-ray imaging systems. The calculated yield of {gamma}-rays in the 0.5-2 MeV range, produced by low-average-power lasers and gas targets, exceeds the yields from solid tape targets up to 60 times. In addition, an effect of quasi-mono energetic electrons on {gamma}-ray imaging is also discussed.

Oishi, Yuji; Nayuki, Takuya; Zhidkov, Alexei; Fujii, Takashi; Nemoto, Koshichi [Central Research Institute of Electric Power Industry, Yokosuka, Kanagawa 240-0196 (Japan); Central Research Institute of Electric Power Industry, Yokosuka, Kanagawa 240-0196, Japan and Photon Pioneers Center in Osaka University, Yamadaoka 2-1, Suita, Osaka 565-0871 (Japan); Central Research Institute of Electric Power Industry, Yokosuka, Kanagawa 240-0196 (Japan)

2012-07-11T23:59:59.000Z

139

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

Potter, J.D.

1984-03-30T23:59:59.000Z

140

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

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.


141

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

142

Assembly of the PLT device  

SciTech Connect

The assembly of the PLT device began in June 1974 with a preassembly of the mechanical structure at a remote site. The preassembly sequence incorporated final fabrication procedures with an initial staging operation. This successful staging/fabrication procedure proved to be an invaluable asset when the final assembly was started in August 1974. The assembly continued with the initial reassembly of the previously tested structural components at the final machine site. Construction was interrupted at several points to allow for toroidal field coil, vacuum vessel, and poloidal coil installation. Two phases of toroidal field coil power tests were included in the assembly sequence prior to, and just after the vacuum vessel insertion. (auth)

Marino, R.

1975-11-01T23:59:59.000Z

143

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

Science Conference Proceedings (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

144

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

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

145

Reactor Core Assembly - HFIR Technical Parameters | ORNL Neutron Sciences  

NLE Websites -- All DOE Office Websites (Extended Search)

Home › Facilities › HFIR › Reactor Core Assembly Home › Facilities › HFIR › Reactor Core Assembly Reactor Core Assembly The reactor core assembly is contained in an 8-ft (2.44-m)-diameter pressure vessel located in a pool of water. The top of the pressure vessel is 17 ft (5.18 m) below the pool surface, and the reactor horizontal mid-plane is 27.5 ft (8.38 m) below the pool surface. The control plate drive mechanisms are located in a subpile room beneath the pressure vessel. These features provide the necessary shielding for working above the reactor core and greatly facilitate access to the pressure vessel, core, and reflector regions. In-core irradiation and experiment locations (cross section at horizontal midplane) Reactor core assembly Reactor core assembly: (1) in-core irradiation and experiment locations,

146

Thermocouple assembly  

DOE Patents (OSTI)

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

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

2002-01-01T23:59:59.000Z

147

RETORT ASSEMBLY  

DOE Patents (OSTI)

An improved retort assembly useful in the thermal reduction of volatilizable metals such as magnesium and calcium is described. In this process a high vacuum is maintained in the retort, however the retort must be heated to very high temperatures while at the same time the unloading end must bo cooled to condense the metal vapors, therefore the retention of the vacuum is frequently difficult due to the thermal stresses involved. This apparatus provides an extended condenser sleeve enclosed by the retort cover which forms the vacuum seal. Therefore, the seal is cooled by the fluid in the condenser sleeve and the extreme thermal stresses found in previous designs together with the deterioration of the sealing gasket caused by the high temperatures are avoided.

Loomis, C.C.; Ash, W.J.

1957-11-26T23:59:59.000Z

148

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

149

Public Assembly Buildings  

U.S. Energy Information Administration (EIA) Indexed Site

Assembly Assembly Characteristics by Activity... Public Assembly Public assembly buildings are those in which people gather for social or recreational activities, whether in private or non-private meeting halls. Basic Characteristics [ See also: Equipment | Activity Subcategories | Energy Use ] Public Assembly Buildings... Most public assembly buildings were not large convention centers or entertainment arenas; about two-fifths fell into the smallest size category. About one-fifth of public assembly buildings were government-owned, mostly by local governments; examples of these types of public assembly buildings are libraries and community recreational facilities. Tables: Buildings and Size Data by Basic Characteristics Establishment, Employment, and Age Data by Characteristics

150

IRRADIATION MEASUREMENTS ON THE 0.25 micro m CMOS PIXEL READOUT TEST CHIP BY A 14 MEV NEUTRON FACILITY  

E-Print Network (OSTI)

ALICE-ITS-2000-24   Abstract   A test facility station with 14 MeV neutrons was arranged at the FNG-ENEA Laboratory in Frascati (Italy) for the characterization with respect to radiation tolerance of the prototype pixel readout chips in 0.25 m m IBM technology done in edgeless design. This facility could allow to test both the readout chips and the pilot chips for the pixel readout system. In fact, both ASICs will have to survive at the same radiation level foreseen for the innermost layer (r = 4 cm) of the Inner Tracker System (ITS) in the LHC-ALICE experiment. Two test chips were exposed to an overall flux of 1.3 x 1012 14 MeV neutrons/cm2, which is larger than the expected neutron flux in ALICE during 10 years data taking. No variation in the parameters defining the chip functionality (analog and digital currents, linearity, shapes of the signal, efficiency) was observed.

Barbera, R; CERN. Geneva; Palmeri, A; Pappalardo, G S; Riggi, F; Di Liberto, S; Meddi, F; Sestito, S; Loi, D; Angelone, M; Badalà, A; Pillon, M

2000-01-01T23:59:59.000Z

151

Estimation of critical flow velocity for collapse of booster fuel assembly  

Science Conference Proceedings (OSTI)

A Gas Test Loop (GTL) system is currently being designed to provide a high intensity fast-flux irradiation environment for testing fuels and materials for advanced concept nuclear reactors. To assess the performance of candidate reactor fuels, these fuels must be irradiated under actual fast reactor flux conditions and operating environments, preferably in an existing irradiation facility. The GTL system is being designed for operation in the northwest test lobe of the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The Technical and Functional Requirements (T&FRs) for the GTL stipulate a minimum neutron flux intensity (10{sup 15} n/cm{sup 2} {center_dot} s) and fast to thermal neutron ratio (>15) for the test environment. The incorporation of booster fuel within the test lobe is necessary to achieve these neutron flux requirements. The current design of the booster fuel assembly for the GTL calls for 3 concentric rings of 4 ft long uranium silicide fuel plates clad with 6061 aluminum.

Donna Guillen; Mark J. Russell

2005-09-01T23:59:59.000Z

152

Slag recycling of irradiated vanadium  

Science Conference Proceedings (OSTI)

An experimental inductoslag apparatus to recycle irradiated vanadium was fabricated and tested. An experimental electroslag apparatus was also used to test possible slags. The testing was carried out with slag materials that were fabricated along with impurity bearing vanadium samples. Results obtained include computer simulated thermochemical calculations and experimentally determined removal efficiencies of the transmutation impurities. Analyses of the samples before and after testing were carried out to determine if the slag did indeed remove the transmutation impurities from the irradiated vanadium.

Gorman, P.K.

1995-04-05T23:59:59.000Z

153

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

154

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

155

Irradiation Studies  

Science Conference Proceedings (OSTI)

Mar 4, 2013 ... Materials and Fuels for the Current and Advanced Nuclear Reactors II: Irradiation Studies Sponsored by: TMS Structural Materials Division, ...

156

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

Science Conference Proceedings (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

157

AFIP-4 Irradiation Summary Report  

Science Conference Proceedings (OSTI)

The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-4 was designed to evaluate the performance of monolithic uranium-molybdenum (U-Mo) fuels at a scale prototypic of research reactor fuel plates. The AFIP-4 test further examine the fuel/clad interface and its behavior under extreme conditions. After irradiation, fission gas retention measurements will be performed during post irradiation (PIE). The following report summarizes the life of the AFIP-4 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results.

Danielle M Perez; Misti A Lillo; Gray S. Chang; Glenn A Roth; Nicolas Woolstenhulme; Daniel M Wachs

2011-09-01T23:59:59.000Z

158

AFIP-4 Irradiation Summary Report  

Science Conference Proceedings (OSTI)

The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-4 was designed to evaluate the performance of monolithic uranium-molybdenum (U-Mo) fuels at a scale prototypic of research reactor fuel plates. The AFIP-4 test further examine the fuel/clad interface and its behavior under extreme conditions. After irradiation, fission gas retention measurements will be performed during post irradiation (PIE)1,2. The following report summarizes the life of the AFIP-4 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results.

Danielle M Perez; Misti A Lillo; Gray S. Chang; Glenn A Roth; Nicolas Woolstenhulme; Daniel M Wachs

2012-01-01T23:59:59.000Z

159

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

160

Fuel Reliability Program: Post-Irradiation Examination and Testing of High-Fluence Control Rod Silver-Indium-Cadmium Absorber from t he Kernkraftwerk Obrigheim Reactor  

Science Conference Proceedings (OSTI)

Within a project sponsored by the Electric Power Research Institute (EPRI), one control rod absorber silver-indium-cadmium (AgInCd) specimen, KWO395, irradiated in Kernkraftwerk Obrigheim (KWO) and one control rod absorber specimen, R035/F9, irradiated in Ringhals 2 have been examined in the hot cell laboratory at Studsvik. The objective of the examinations was to characterize the absorber material and investigate its physical, chemical, and microstructural changes due to high neutron fluence/exposure us...

2011-09-16T23:59:59.000Z

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

Sensor mount assemblies and sensor assemblies  

DOE Patents (OSTI)

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

Miller, David H. (Redondo Beach, CA)

2012-04-10T23:59:59.000Z

162

Effects of neutron irradiation on fatigue and creep-fatigue crack propagation in type 316 stainless steel at 649 degree C produced no significant effect on the crack propagation rate when compared with unirradiated steel tested at 649 degree C  

Science Conference Proceedings (OSTI)

The fatigue and creep-fatigue crack propagation performance of Type 316 stainless steel has been investigated following fast neutron (n) irradiation. The purpose was to evaluate the effects of neutron fluence and temperature on the crack propagation resistance and failure mode of the steel. Results are presented from fatigue tests of the annealed steel that were irradiated at 649 degree C Scanning electron microscope examination of the fracture surfaces of the tested specimens revealed that the failure mode of the specimens which exhibited increased crack propagation rates was primarily intergranular while a transgranular mode was observed for specimens with lower crack propagation rates. The results point toward a synergistic relationship between thermomechanical history, precipitate formation, and hold time effects as the responsible mechanism for the crack propagation performance.

Michel, D.J.; Smith, H.H.

1981-01-01T23:59:59.000Z

163

Control rod assembly for liquid metal fast breeder reactors  

SciTech Connect

This standard establishes the requirements for fabrication, testing, and inspection of control rod assemblies for use in liquid metal fast breeder reactors.

1978-09-08T23:59:59.000Z

164

Test Request LTFY-2  

SciTech Connect

This test request defines pre-test data for the irradiation of enriched yttrium-uranium hydride samples in the LITR C-48 facility.

1964-12-01T23:59:59.000Z

165

TEST  

Science Conference Proceedings (OSTI)

This is an abstract. TEST Lorem ipsum dolor sit amet, consectetur adipiscing elit. Cras lacinia dui et est venenatis lacinia. Vestibulum lacus dolor, adipiscing id mattis sit amet, ultricies sed purus. Nulla consectetur aliquet feugiat. Maecenas ips

166

Sandia National Laboratories: Research: Facilities: Gamma Irradiation  

NLE Websites -- All DOE Office Websites (Extended Search)

Gamma Irradiation Facility Gamma Irradiation Facility Photo of Gamma Irradiation Facility The Gamma Irradiation Facility (GIF) provides high-fidelity simulation of nuclear radiation environments for materials and component testing. The low-dose irradiation facility also offers an environment for long-duration testing of materials and electronic components. Such testing may take place over a number of months or even years. Research and other activities The single-structure GIF can house a wide variety of gamma irradiation experiments with various test configurations and at different dose and dose rate levels. Radiation fields at the GIF are produced by high-intensity gamma-ray sources. To induce ionizing radiation effects and damage in test objects, the objects are subjected to high-energy photons from gamma-source

167

Public Assembly Buildings  

U.S. Energy Information Administration (EIA) Indexed Site

buildings. Since they comprised 7 percent of commercial floorspace, this means that their energy intensity was just slightly below the commercial average. Public assembly buildings...

168

Irradiation subassembly  

DOE Patents (OSTI)

An irradiation subassembly for use in a nuclear reactor is described which includes a bundle of slender elongated irradiation -capsules or fuel elements enclosed by a coolant tube and having yieldable retaining liner between the irradiation capsules and the coolant tube. For a hexagonal bundle surrounded by a hexagonal tube the yieldable retaining liner may consist either of six segments corresponding to the six sides of the tube or three angular segments each corresponding in two adjacent sides of the tube. The sides of adjacent segments abut and are so cut that metal-tometal contact is retained when the volume enclosed by the retaining liner is varied and Springs are provided for urging the segments toward the center of the tube to hold the capsules in a closely packed configuration. (Official Gazette)

Seim, O.S.; Filewicz, E.C.; Hutter, E.

1973-10-23T23:59:59.000Z

169

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

Science Conference Proceedings (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

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

SciTech Connect

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

171

Turbine disc sealing assembly  

SciTech Connect

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

172

Material Testing - Nuclear Engineering Division (Argonne)  

NLE Websites -- All DOE Office Websites (Extended Search)

Departments involved: Engineering Development and Applications Irradiated Materials Two hot-cell test facilities are used to develop experimental data on the irradiation-assisted...

173

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

174

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

175

Californium Neutron Irradiation Facility  

Science Conference Proceedings (OSTI)

Californium Neutron Irradiation Facility. Summary: ... Cf irradiation facility (Photograph by: Neutron Physics Group). Lead Organizational Unit: pml. Staff: ...

2013-07-23T23:59:59.000Z

176

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

177

Fracture and Impact Properties of HT-9 Steel Irradiated to High Dose ...  

Science Conference Proceedings (OSTI)

Presentation Title, Fracture and Impact Properties of HT-9 Steel Irradiated to High ... 250, and the irradiation temperature in a servo-hydraulic testing machine.

178

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

DOE Patents (OSTI)

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

Bradley, John G. (Richland, WA)

1982-01-01T23:59:59.000Z

179

Selected Isotopes for Optimized Fuel Assembly Tags  

SciTech Connect

In support of our ongoing signatures project we present information on 3 isotopes selected for possible application in optimized tags that could be applied to fuel assemblies to provide an objective measure of burnup. 1. Important factors for an optimized tag are compatibility with the reactor environment (corrosion resistance), low radioactive activation, at least 2 stable isotopes, moderate neutron absorption cross-section, which gives significant changes in isotope ratios over typical fuel assembly irradiation levels, and ease of measurement in the SIMS machine 2. From the candidate isotopes presented in the 3rd FY 08 Quarterly Report, the most promising appear to be Titanium, Hafnium, and Platinum. The other candidate isotopes (Iron, Tungsten, exhibited inadequate corrosion resistance and/or had neutron capture cross-sections either too high or too low for the burnup range of interest.

Gerlach, David C.; Mitchell, Mark R.; Reid, Bruce D.; Gesh, Christopher J.; Hurley, David E.

2008-10-01T23:59:59.000Z

180

DC source assemblies  

SciTech Connect

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

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

Steam separator latch assembly  

SciTech Connect

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

182

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

183

Silicon solar cell assembly  

DOE Patents (OSTI)

A silicon solar cell assembly comprising a large, thin silicon solar cell bonded to a metal mount for use when there exists a mismatch in the thermal expansivities of the device and the mount.

Burgess, Edward L. (Albuquerque, NM); Nasby, Robert D. (Albuquerque, NM); Schueler, Donald G. (Albuquerque, NM)

1979-01-01T23:59:59.000Z

184

Results of charpy V-notch impact testing of structural steel specimens irradiated at {approximately}30{degrees}C to 1 x 10{sup 16} neutrons/cm{sup 2} in a commercial reactor cavity  

SciTech Connect

A capsule containing Charpy V-notch (CVN) and mini-tensile specimens was irradiated at {approximately} 30{degrees}C ({approximately} 85{degrees}F) in the cavity of a commercial nuclear power plant to a fluence of 1 x 10{sup 16} neutrons/cm{sup 2} (> 1MeV). The capsule included six CVN impact specimens of archival High Flux Isotope Reactor A212 grade B ferritic steel and five CVN impact specimens of a well-studied A36 structural steel. This irradiation was part of the ongoing study of neutron-induced damage effects at the low temperature and flux experienced by reactor supports. The plant operators shut down the plant before the planned exposure was reached. The exposure of these specimens produced no significant irradiation-induced embrittlement. Of interest were the data on unirradiated specimens in the L-T orientation machined from a single plate of A36 structural steel, which is the same specification for the structural steel used in some reactor supports. The average CVN energy of five unirradiated specimens obtained from one region of the plate and tested at room temperature was {approximately} 99 J, while the energy of 11 unirradiated specimens from other locations of the same plate was 45 J, a difference of {approximately} 220%. The CVN impact energies for all 18 specimens ranged from a low of 32 J to a high of 111 J. Moreover, it appears that the University of Kansas CVN impact energy data of the unirradiated specimens at the 100-J level are shifted toward higher temperatures by about 20 K. The results were an example of the extent of scatter possible in CVN impact testing. Generic values for the CVN impact energy of A36 should be used with caution in critical applications.

Iskander, S.K.; Stoller, R.E.

1997-04-01T23:59:59.000Z

185

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

186

Assembly flow simulation of a radar  

Science Conference Proceedings (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

187

Photovoltaic self-assembly.  

DOE Green Energy (OSTI)

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

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

2010-10-01T23:59:59.000Z

188

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

189

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

190

Hot-Cell Examination and Assessment Report for a Next Generation Fuel Skeleton Irradiated in Millstone-3  

Science Conference Proceedings (OSTI)

Lead test assemblies (LTAs) of the 17 x 17 Next Generation Fuel (17NGF) fuel design from Westinghouse Electric Company have been irradiated at Millstone Unit 3 for up to three cycles and have accumulated up to ~64,000 megawatt days per metric ton of uranium (MWD/MTU) of exposure. The objective of this project is to perform a full hot-cell examination of one LTA skeleton at discharge exposure, including two main activities: general characterization (for example, wear, dimensional stability, ...

2013-06-27T23:59:59.000Z

191

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

SciTech Connect

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

PITNER, A.L.

2000-02-28T23:59:59.000Z

192

Power module assembly  

SciTech Connect

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

193

Irradiation Performance - Nuclear Engineering Division (Argonne)  

NLE Websites -- All DOE Office Websites (Extended Search)

Materials Testing > Materials Testing > Irradiation Performance Capabilities Materials Testing Environmentally Assisted Cracking (EAC) of Reactor Materials Corrosion Performance/Metal Dusting Irradiated Materials Overview Light Water Reactor Materials Other Current Activities Future Directions Steam Generator Tube Integrity Other Facilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Irradiation Performance Bookmark and Share The activities of the Irradiation Performance Section (IPS) are aimed at determining and assessing normal-operation and accident behavior of neutron-irradiated material throughout the life cycle of the materials. The conditions of interest are normal in-reactor operation, design-basis accidents, intermediate storage in pools and dry casks, and ultimate

194

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

195

PROCEEDINGS OF THE AEC SYMPOSIUM FOR CHEMICAL PROCESSING OF IRRADIATED FUELS FROM POWER, TEST, AND RESEARCH REACTORS, RICHLAND, WASHINGTON, OCTOBER 20 AND 21, 1959  

SciTech Connect

A review is presented in this symposium of the technology currently available for processing spent fuels from research, test, and power reactors. Twenty-one papers are included. Separate abstracts have been prepared for each paper. (W.L.H.)

1960-01-01T23:59:59.000Z

196

The Effect Of Neutron Irradiation On The Mechanical Properties Of Welded Zircaloy-2  

DOE Green Energy (OSTI)

Zircaloy-2 tensile specimens, subsize impact bars and representative spigot welds were subjected to three NRX cycles in the X-5 loop. Average loop temperature was 260 deg C over the three cycles. One group of tensile specimens was heat-treated in vacuum at 900 deg C for 40 minutes, another group contained welded areas in the center of the gauge length and a third group was hydrided after welding. Notches of the impact specimens were located in the fusion zone of the weld. Spigot welds were made on autoclaved and unautoclaved simulated production assemblies. Neutron irradiation had no effect on the impact properties of welded Zircaloy-2. Welding decreased the uniform and total elongation at room temperature and at 260 deg C, and increased the 260 deg C PL, YS, and UTS. Hydriding to a nominal 100 ppm hydrogen had no effect on the unirradiated tensile properties at either test temperature. The heat treatment decreased the strength properties but did not affect the ductility. Neutron irradiation increased the YS of the welded and hydrided material by 20% and the heat treated YS by 40%. Irradiation also increased the 260 deg C strength properties of the as-welded material. The unautoclaved spigot welds had a generally higher tensile strength than the autoclaved and welded specimens. For specimens welded in either condition, the outer welds of the 19-element bundle had a lower average breaking load than the inner welds. Neutron irradiation had no effect on the tensile strength of these welds. It was also demonstrated that a cup-and-cone type of fracture could be produced in a bend test. The fractures were similar to those observed in irradiated fuel bundles which was damaged during transfer operations. (auth)

Evans, D.G.

1962-07-15T23:59:59.000Z

197

Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 1, Activation measurements and comparison with calculations for spent fuel assembly hardware  

Science Conference Proceedings (OSTI)

Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly remains that is also radioactive and requires disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report presents a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volumes 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1. 5 refs., 4 figs., 21 tabs.

Luksic, A.

1989-06-01T23:59:59.000Z

198

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

199

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

200

NEUTRONIC REACTOR BURIAL ASSEMBLY  

DOE Patents (OSTI)

A burial assembly is shown whereby an entire reactor core may be encased with lead shielding, withdrawn from the reactor site and buried. This is made possible by a five-piece interlocking arrangement that may be easily put together by remote control with no aligning of bolt holes or other such close adjustments being necessary.

Treshow, M.

1961-05-01T23:59:59.000Z

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

Solar collector assembly  

Science Conference Proceedings (OSTI)

A solar collector assembly includes shingles which have integral tubes projecting therefrom, and which are mounted in overlapping parallel array. Mounting brackets for the shingles are engaged on roof rafters or the like, and interlocked light transmissive plates overlie the shingles. The plates are also engaged with shingle components. A special fitting for the tube ends is provided.

Murphy, J.A.

1980-09-09T23:59:59.000Z

202

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

203

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

204

Solar central receiver heliostat reflector assembly  

SciTech Connect

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

205

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

206

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

207

Assembling and Installing LRUs for NIF  

SciTech Connect

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

Bonanno, R E

2003-12-31T23:59:59.000Z

208

Fire Tests of Single Office Workstations  

Science Conference Proceedings (OSTI)

... Trade Center Disaster: The Con Ed Substation in World ... was typically assembled shortly before a test. ... at the time of the testing was approximately ...

209

Fire Resistance Tests of Floor Truss Systems  

Science Conference Proceedings (OSTI)

... Trade Center Disaster: The Con Ed Substation in World ... 4.1.1 Span of Test Assembly ... The Underwriters Laboratories of Canada fire testing facility in ...

210

RERTR-13 Irradiation Summary Report  

Science Conference Proceedings (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

211

Radiochemistry Results from the IER-163 COMET Irradiation  

SciTech Connect

The COMET assembly at the National Criticality Experiments Research Center (NCERC) at the Nevada National Security Site (NNSS) was used to irradiate twelve foils in September 2011. The intention of this irradiation was to measure 'peak yield' fission product activities, activation products, and so-called 'endpoint R values' for different foil materials in a non-thermal neutron spectrum. After irradiation, several of the foils were shipped to Los Alamos National Laboratory (LANL) for radiochemical analysis. The results from the non-destructive and radiochemical analyses of six of these foils are presented.

Jackman, Kevin R. [Los Alamos National Laboratory; Bredeweg, Todd Allen [Los Alamos National Laboratory; Schake, Ann R. [Los Alamos National Laboratory; Oldham, Warren J. [Los Alamos National Laboratory; Bounds, John Alan [Los Alamos National Laboratory; Attrep, Moses Jr. [Los Alamos National Laboratory; Rundberg, Robert S. [Los Alamos National Laboratory

2012-03-27T23:59:59.000Z

212

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

213

Vacuum breaker valve assembly  

Science Conference Proceedings (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

214

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

215

Turbine seal assembly  

SciTech Connect

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

216

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.

1999-06-18T23:59:59.000Z

217

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

218

FLUORINE CELL ANODE ASSEMBLY  

DOE Patents (OSTI)

An improved anode assembly is deslgned for use in electrolytlc cells ln the productlon of hydrogen and fluorlne from a moIten electrolyte. The anode assembly comprises a copper post, a copper hanger supported by the post, a plurality of carbon anode members, and bolt means for clamplng half of the anode members to one slde of the hanger and for clamplng the other half of the anode members to the other slde of the hanger. The heads of the clamplng bolts are recessed withln the anode members and carbon plugs are inserted ln the recesses above the bolt heads to protect the boIts agalnst corroslon. A copper washer is provided under the head of each clamplng boIt such that the anode members can be tightly clamped to the hanger with a resultant low anode jolnt resistance. (AEC)

Cable, R.E.; Goode, W.B. Jr.; Henderson, W.K.; Montillon, G.H.

1962-06-26T23:59:59.000Z

219

RERTR-12 Insertion 2 Irradiation Summary Report  

SciTech Connect

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

220

Assembly of Energy Nanomaterials via Water Based ...  

Science Conference Proceedings (OSTI)

Page 1. Assembly of Energy Nanomaterials via Water Based Assembly Paula T. Hammond Department of Chemical Engineering ...

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

Nuclear core and fuel assemblies  

DOE Patents (OSTI)

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

Downs, Robert E. (Monroeville, PA)

1981-01-01T23:59:59.000Z

222

GTL-1 Irradiation Summary Report  

Science Conference Proceedings (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

223

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

Science Conference Proceedings (OSTI)

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

Dawn Scates

2010-10-01T23:59:59.000Z

224

Startup of the Fission Converter Epithermal Neutron Irradiation Facility at the MIT Reactor  

Science Conference Proceedings (OSTI)

A new epithermal neutron irradiation facility, based on a fission converter assembly placed in the thermal column outside the reactor core, has been put into operation at the Massachusetts Institute of Technology Research Reactor (MITR). This facility was constructed to provide a high-intensity, forward-directed beam for use in neutron capture therapy with an epithermal flux of [approximately equal to]10{sup 10} n/cm{sup 2}.s at the medical room entrance with negligible fast neutron and gamma-ray contamination. The fission converter assembly consists of 10 or 11 MITR fuel elements placed in an aluminum tank and cooled with D{sub 2}O. Thermal-hydraulic criteria were established based on heat deposition calculations. Various startup tests were performed to verify expected neutronic and thermal-hydraulic behavior. Flow testing showed an almost flat flow distribution across the fuel elements with <5% bypass flow. The total reactivity change caused by operation of the facility was measured at 0.014 {+-} 0.002% {delta}K/K. Thermal power produced by the facility was measured to be 83.1 {+-} 4.2 kW. All of these test results satisfied the thermal-hydraulic safety criteria. In addition, radiation shielding design measurements were made that verified design calculations for the neutronic performance.

Newton, Thomas H. Jr.; Riley, Kent J.; Binns, Peter J.; Kohse, Gordon E.; Hu Linwen; Harling, Otto K. [Massachusetts Institute of Technology (United States)

2002-08-15T23:59:59.000Z

225

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

226

Generation and Retention of Helium and Hydrogen in Austenitic Steels Irradiated in a Variety of LWR and Test Reactor Spectral Environments  

DOE Green Energy (OSTI)

In fission and fusion reactor environments stainless steels generate significant amounts of helium and hydrogen by transmutation. The primary sources of helium are boron and nickel, interacting with both fast and especially thermal neutrons. Hydrogen arises primarily from fast neutron reactions, but is also introduced into steels at often much higher levels by other environmental processes. Although essentially all of the helium is retained in the steel, it is commonly assumed that most of the hydrogen is not retained. It now appears that under some circumstances, significant levels of hydrogen can be retained, especially when helium-nucleated cavities become a significant part of the microstructure. A variety of stainless steel specimens have been examined from various test reactors, PWRs and BWRs. These specimens were exposed to a wide range of neutron spectra with different thermal/fast neutron ratios. Pure nickel and pure iron have also been examined. It is shown that all major features of the retention of helium and hydrogen can be explained in terms of the composition, thermal/fast neutron ratio and the presence or absence of helium-nucleated cavities. In some cases, the hydrogen retention is very large and can exceed that generated by transmutation, with the additional hydrogen arising from either environmental sources and/or previously unidentified radioisotope sources that may come into operation at high neutron exposures.

Garner, Francis A.; Oliver, Brian M.; Greenwood, Lawrence R.; Edwards, Danny J.; Bruemmer, Stephen M.; Grossbeck, Martin L.

2002-03-31T23:59:59.000Z

227

Neutron Irradiation of Hydrided Cladding Material in HFIR Summary of  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Neutron Irradiation of Hydrided Cladding Material in HFIR Summary Neutron Irradiation of Hydrided Cladding Material in HFIR Summary of Initial Activities Neutron Irradiation of Hydrided Cladding Material in HFIR Summary of Initial Activities Irradiation is known to have a significant impact on the properties and performance of Zircaloy cladding and structural materials (material degradation processes, e.g., effects of hydriding). This UFD study examines the behavior and performance of unirradiated cladding and actual irradiated cladding through testing and simulation. Three capsules containing hydrogen-charged Zircaloy-4 cladding material have been placed in the High Flux Isotope Reactor (HFIR). Irradiation of the capsules was conducted for post-irradiation examination (PIE) metallography. Neutron Irradiation of Hydrided Cladding Material in HFIR Summary of

228

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

229

Gas separation membrane module assembly  

SciTech Connect

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

230

Multi-position photovoltaic assembly  

Science Conference Proceedings (OSTI)

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

Dinwoodie, Thomas L. (Piedmont, CA)

2003-03-18T23:59:59.000Z

231

Next-generation transcriptome assembly  

Science Conference Proceedings (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

232

CGAL: computing genome assembly likelihoods  

E-Print Network (OSTI)

SM, Lei M, Li J, et al: Genome sequencing in microfabricatedDe novo bacterial genome sequencing: millions of very shortGenome assembly, evaluation, likelihood, sequencing.

Rahman, Atif; Pachter, Lior

2013-01-01T23:59:59.000Z

233

AGC-1 Post Irradiation Examination Status  

SciTech Connect

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

234

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

Science Conference Proceedings (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

235

Switchable heat pipe assembly  

SciTech Connect

The heat pipe assembly is formed into an H-shape or a Y-shape. The H-shaped configuration comprises two heat pipes, each having condenser and evaporator sections with wicking therein coupled by a tube with wick at their evaporator sections. The Y-shaped configuration utilizes a common evaporator section in place of the two evaporator sections of the H-shaped configuration. In both configurations, the connection between the vapor spaces of the two heat pipes equalizes vapor pressure within the heat pipes. Although both heat pipes have wicks, they have sufficient fluid only to saturate a single pipe. If heat is applied to the condenser section of one of the pipes, this heat pipe becomes inoperative since all the fluid is transferred to the second pipe which can operate with a lower thermal load.

Sun, T.H.; Basiulis, A.

1977-02-15T23:59:59.000Z

236

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

237

Photovoltaic cell assembly  

DOE Patents (OSTI)

A photovoltaic assembly for converting high intensity solar radiation into electrical 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. 2 figs.

Beavis, L.C.; Panitz, J.K.G.; Sharp, D.J.

1989-09-26T23:59:59.000Z

238

Rotatable seal assembly  

DOE Patents (OSTI)

An assembly is provided for rotatably supporting a rotor on a stator so that vacuum chambers in the rotor and stator remain in communication while the chambers are sealed from ambient air, which enables the use of a ball bearing or the like to support most of the weight of the rotor. The apparatus includes a seal device mounted on the rotor to rotate therewith, but shiftable in position on the rotor while being sealed to the rotor as by an O-ring. The seal device has a flat face that is biased towards a flat face on the stator, and pressurized air is pumped between the faces to prevent contact between them while spacing them a small distance apart to avoid the inflow of large amounts of air between the faces and into the vacuum chambers.

Logan, Clinton M. (Pleasanton, CA); Garibaldi, Jack L. (Livermore, CA)

1982-01-01T23:59:59.000Z

239

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

SciTech Connect

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

240

Reactivity Initiated Accident Test Series Test RIA 1-2 Experiment Operating Specification  

SciTech Connect

This document describes the experiment operating specifications for the Reactivity Initiated Accident (RIA) Test RIA 1-2 to be conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. The RIA Series I research objectives are to determine fuel failure thresholds, modes and consequences as functions of enthalpy insertion, irradiation history, and fuel design. Coolant conditions of pressure, temperature, and flow rate that are typical of hot-startup conditions in commercial boiling water reactors {BWRs) will be used. The second test in Series I, Test RIA 1-2, will be comprised of four individual rods, each surrounded by a separate flow shroud. The four rods will be preirradiated. The specific objectives of the test are to: (1) characterize the response of preirradiated fuel rods during a RIA event conducted at BWR hot-startup conditions and (2) evaluate the effect of internal rod pressure on preirradiated fuel rod transient response. The test sequence will begin with steady state power operation to condition the fuel (pellet cracking and relocation) and determine the fuel rod power calibration. The loop will then be cooled down, the test train removed from the in-pile tube, and the cobalt flux wires that are mounted on each flow shroud will be replaced. The transient fuel rod energy deposition for the Test RIA 1-2 rods will be chosen from the fuel rod response vs. energy deposition observed in the first three phases of the RIA Scoping Test and the first test of Series J, Test RIA 1-1. The design of the test fuel rods, test assembly, and instrumentation associated with Test RIA 1-2 are described. The planned experiment conduct for the test is described. The data recording and reduction requirements are provided. The posttest operations support and the postirradiation examination requirements associated with Test RIA 1-2 are described.

1978-07-01T23:59:59.000Z

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.


241

Coaxial test fixture  

DOE Patents (OSTI)

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

Praeg, Walter F. (Palos Park, IL)

1986-01-01T23:59:59.000Z

242

Fuel cell sub-assembly  

DOE Patents (OSTI)

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

Chi, Chang V. (Brookfield, CT)

1983-01-01T23:59:59.000Z

243

Moisture Research - Optimizing Wall Assemblies  

SciTech Connect

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

244

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

245

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

246

Reactivity Initiated Accident Test Series Test RIA 1-1 Experiment Operating Specification  

SciTech Connect

This document describes the experiment operating specifications for the Reactivity Initiated Accident (RIA) Test RIA 1-1 to be conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. The RIA Series I research objectives are to determine fuel failure thresholds, modes and consequences as functions of enthalpy insertion, irradiation history, and fuel design. Coolant conditions of pressure, temperature, and flow rate that are typical of hot-startup conditions in commercial boiling water reactors (BWRs) will be used. The first test in Series I, Test RIA 1-1, will be comprised of four individual rods, each surrounded by a separate flow shroud. Two rods will be preirradiated and two rods will be unirradiated. The specific objectives of the test are to: (1) characterize the response of unirradiated and preirradiated fuel rods during a RIA event conducted at BWR hot-startup conditions and (2) evaluate test instrumentation response during an RIA. The test sequence will begin with steady state power operation to condition the fuel (pellet cracking and relocation) and determine the fuel rod power calibration. The loop will then be cooled down, the test train removed from the in-pile tube, and one of the unirradiated rods will be removed for fission product analysis and replaced with an identical unirradiated rod. The transient fuel rod energy deposition for Test RIA 1-1 will be chosen from the fuel rod response vs. energy deposition data observed in the first three phases of the RIA Scoping Test. It is anticipated that a fuel pellet surface energy deposition of about 1100 J/g will be required to ensure cladding failure of all four rods. The design of the test fuel rods, test assembly, and instrumentation associated with Test RIA 1-1 are described. The experiment conduct for the test is described. The data recording and reduction requirements are provided. The posttest support and the postirradiation examination requirements associated with Test RIA 1-1 are described.

1978-08-01T23:59:59.000Z

247

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

SciTech Connect

This document has been prepared to assist research reactor operators possessing spent fuel containing enriched uranium of United States origin to prepare part of the documentation necessary to ship this fuel to the United States. Data are included on the nuclear mass inventory, photon dose rate, and thermal decay heat of spent research reactor fuel assemblies. Isotopic masses of U, Np, Pu and Am that are present in spent research reactor fuel are estimated for MTR, TRIGA and DIDO-type fuel assembly types. The isotopic masses of each fuel assembly type are given as functions of U-235 burnup in the spent fuel, and of initial U-235 enrichment and U-235 mass in the fuel assembly. Photon dose rates of spent MTR, TRIGA and DIDO-type fuel assemblies are estimated for fuel assemblies with up to 80% U-235 burnup and specific power densities between 0.089 and 2.857 MW/kg[sup 235]U, and for fission product decay times of up to 20 years. Thermal decay heat loads are estimated for spent fuel based upon the fuel assembly irradiation history (average assembly power vs. elapsed time) and the spent fuel cooling time.

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

1996-12-31T23:59:59.000Z

248

EOR databank assembled  

Science Conference Proceedings (OSTI)

Enhanced oil recovery (EOR) technology has progressed rapidly in the last few years, and a unique situation is now at hand. The Bartlesville Energy Technology Center (BETC) of the Department of Energy has supported research to survey and collect existing, publicly available data on oil reservoirs, to indentify those reservoirs amendable to EOR technology. The data from these efforts act as a broad base from which simplified models may be derived to predict the amounts of oil reserves technically and economically recoverable by EOR techniques. BETC also has been involved in 2 programs related to EOR technology - the enhanced oil recovery cost-sharing program, and the tertiary incentive crude oil program. These 2 programs have achieved the largest accumulation of data on EOR projects assembled in one place. The data will be used to improve the predictability of the simplified models; this improved predictability, it is hoped, will promote more widespread use for EOR technology and eventually reduce the risk involved in applying this technology to new areas such as the Northeast.

Ray, R.M.

1981-10-01T23:59:59.000Z

249

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

250

Public Assembly | Open Energy Information  

Open Energy Info (EERE)

Assembly Assembly Jump to: navigation, search Building Type Public Assembly Definition Buildings in which people gather for social or recreational activities, whether in private or non-private meeting halls. Sub Categories social or meeting (e.g. community center, lodge, meeting hall, convention center, senior center); recreation (e.g. gymnasium, health club, bowling alley, ice rink, field house, indoor racquet sports); entertainment or culture (e.g. museum, theater, cinema, sports arena, casino, night club); library; funeral home; student activities center; armory; exhibition hall; broadcasting studio; transportation terminal References EIA CBECS Building Types [1] References ↑ EIA CBECS Building Types U.S. Energy Information Administration (Oct 2008)

251

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

252

Fuel assembly for nuclear reactors  

DOE Patents (OSTI)

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

Creagan, Robert J. (Pitcairn, PA); Frisch, Erling (Pittsburgh, PA)

1977-01-01T23:59:59.000Z

253

Ion Irradiation Effects  

Science Conference Proceedings (OSTI)

Oct 17, 2011 ... Materials Science Challenges for Nuclear Applications: Ion Irradiation Effects Sponsored by: MS&T Organization Program Organizers: Ram ...

254

Irradiation Damage Processes  

Science Conference Proceedings (OSTI)

...R.L. Klueh, Effect of Neutron Irradiation on Properties of Steels, Properties and Selection: Irons, Steels, and High-Performance Alloys,

255

RERTR-7 Irradiation Summary Report  

Science Conference Proceedings (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

256

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

Science Conference Proceedings (OSTI)

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

Terry, W.K.; Ryskamp, J.M.; Sterbentz, J.W. [and others

1995-08-01T23:59:59.000Z

257

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

258

Nuclear fuel assembly spacer  

Science Conference Proceedings (OSTI)

In a fuel assembly for a nuclear reactor including a plurality of elongated elements, a spacer is described for retaining the elements in lateral position. The spacer consists of: an array of laterally positioned, cojoined tubular ferrules, each of the ferrules providing a passage for one of the elements, laterally oriented leaf spring members, each of the spring members spanning two adjacent ones of the ferrules and extending therein to engage and laterally support the elements extending through the adjacent ferrules, facing sides of the adjacent ferrules being formed with cutouts to receive and support the spring member. The sides of the ferrules opposite the facing sides are formed with openings to receive and restrain the ends of the spring member, the spring member being formed with a generally V-shaped central portion with an apex extending toward the adjacent sides of the adjacent ferrules whereby in the absence of elements through the adjacent ferrules the central portion contacts the adjacent sides to provide a preload on the spring member and limit the amount of projection of the spring member into the ferrules whereby the insertion of the elements through the ferrules is facilitated. The central portion of the spring member is unrestrained in the presence of the elements through the ferrules, the spring member having left and right arms extending outward from the V-shaped central portion, each of the arms including a relatively long center portion for contacting a respective one of the elements. A shorter end portion is angled toward the ferrules and a tab of reduced height at the end of each arm engaging a respective one of the openings whereby the resulting shoulders at the ends of the spring member engage the inner surface of the ferrules adjacent the openings to laterally locate and retain the spring member.

Johanssen, E.B.; Matzner, B.

1986-02-18T23:59:59.000Z

259

Wafer scale micromachine assembly method  

SciTech Connect

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

260

Micro and Macro Scale Mechanical Testing and Characterization on ...  

Science Conference Proceedings (OSTI)

Here we present micro compression testing and nanoindentation results performed on spallation source (neutron) and ion beam irradiated engineering ...

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

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

262

The Effects of Ionizing Irradiation  

Science Conference Proceedings (OSTI)

Page 1. The Effects of Ionizing Irradiation on Liquid, Dried, and Absorbed DNA Extracts ... Page 12. Study Shipped Land Carrier Irradiation ? ...

2012-02-29T23:59:59.000Z

263

Automated solar panel assembly line. LSA task: production processes and equipment. Quarterly report No. 1  

DOE Green Energy (OSTI)

The objective of this program is to design, fabricate and demonstrate an automated solar cell module production line with the ultimate goal of reducing module assembly costs. During this reporting period the automated module design was completed. The design of the solar cell assembly prototype (SCAP) was about 75% completed and the solar panel lamination prototype (SPLP) was built and tested.

Somberg, H.

1979-04-08T23:59:59.000Z

264

Irradiation-induced creep of HT-9 cladding in LMR fuel pins  

SciTech Connect

Metal-fueled liquid-metal reactors (LMRs) with their hard neutron spectrum have many desirable performance properties. To take advantage of these, design considerations call for low-swelling alloys, such as the ferritic steel HT-9, as core structural materials. The steady-state performance of the fuel pin is limited to some extent by the degree of deformation of the cladding with burnup. Since HT-9 steel does not exhibit irradiation-induced swelling to design-level fast fluences, the limiting cladding deformation is expected to be due to creep. The experimental and analysis activities in the Integral Fast Reactor (IFR) program at Argonne National Laboratory have afforded an opportunity to study the creep behavior of HT-9 cladding. The methodology consists of applying precise neutronic and thermal-hydraulic calculational capabilities to individual experimental fuel pins. This allows the creation of a rather large data base that relates the measured axial variation of the cladding deformation to the calculated local neutronic properties and cladding temperature, thereby significantly increasing the amount of available data for developing correlations. For an application of this methodology, the lead IFR test assembly X425 irradiated in Experimental Breeder Reactor II (EBR-II) was chosen.

Yacout, A.M.; Orechwa, Y. (Argonne National Lab., IL (United States))

1992-01-01T23:59:59.000Z

265

In the Vehicle Assembly Building at NASA's Kennedy Space  

E-Print Network (OSTI)

was tested March 26, 2004. The engines will burn for eight and one-half minutes as Atlantis roars into spaceIn the Vehicle Assembly Building at NASA's Kennedy Space Center, the external fuel tank for space. The external tank carries the fuel that will be used by a trio of space shuttle main engines to lift Atlantis

266

Evaluation of Neutron Irradiated Silicon Carbide and Silicon Carbide Composites  

SciTech Connect

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

267

A Monte Carlo Model for Interrogation of Thick Cargos for Clandestine Fissionable Materials; Tests with 14-MeV Neutrons  

DOE Green Energy (OSTI)

A Monte Carlo model has been developed for interrogation of fissionable material embedded in thick cargos when high-energy {beta}-delayed {gamma} rays are detected following neutron-induced fission. The model includes the principal structural components of the laboratory, the neutron source and collimator assembly in which it resides, the assembly that represents cargo of given characteristics, a target of highly-enriched uranium (HEU) and large external plastic scintillators for photon detection. The ability of this model to reproduce experimental measurements was tested by comparing simulations with measurements of the number of induced fissions and the number of detected photons when the HUE target was irradiated with 14.25-MeV neutrons in the absence of any cargo and while embedded in assemblies of plywood and iron pipes. The simulations agreed with experimental measurements within a factor of about 2 for irradiation of the bare target and when the areal density of intervening cargo was 33 g cm{sup -2} (wood) and 61 g cm{sup -2} (steel pipes). This suggests that the model can permit exploration of a large range in parameter space with reasonable fidelity.

Prussin, S; Descalle, M; Hall, J; Pruet, J; Slaughter, D; Accatino, M; Alford, O; Asztalos, S; Bernstein, A; Church, J; Gosnell, T; Loshak, A; Madden, N; Manatt, D; Mauger, G; Meyer, A; Moore, T; Norman, E; Pohl, B; Petersen, D; Rusnak, B; Sundsmo, T; Tembrook, W; Walling, R

2006-06-08T23:59:59.000Z

268

LDRD final report: Automated planning and programming of assembly of fully 3D mechanisms  

SciTech Connect

This report describes the results of assembly planning research under the LDRD. The assembly planning problem is that of finding a sequence of assembly operations, starting from individual parts, that will result in complete assembly of a device specified as a CAD model. The automated assembly programming problem is that of automatically producing a robot program that will carry out a given assembly sequence. Given solutions to both of these problems, it is possible to automatically program a robot to assemble a mechanical device given as a CAD data file. This report describes the current state of our solutions to both of these problems, and a software system called Archimedes 2 we have constructed to automate these solutions. Because Archimedes 2 can input CAD data in several standard formats, we have been able to test it on a number of industrial assembly models more complex than any before attempted by automated assembly planning systems, some having over 100 parts. A complete path from a CAD model to an automatically generated robot program for assembling the device represented by the CAD model has also been demonstrated.

Kaufman, S.G.; Wilson, R.H.; Jones, R.E.; Calton, T.L.; Ames, A.L.

1996-11-01T23:59:59.000Z

269

PHAST (PHAGE ASSEMBLY SUITE AND TUTORIAL): A WEB-BASED GENOME ASSEMBLY TEACHING TOOL  

E-Print Network (OSTI)

PHAST (PHAGE ASSEMBLY SUITE AND TUTORIAL): A WEB-BASED GENOME as genome assembly. PHAST (Phage Assembly Suite and Tutorial) is an online set small phage genomes of their own. With PHAST, entry-level biology students learn

Campbell, A. Malcolm

270

Advances in Mechanical Testing - Programmaster.org  

Science Conference Proceedings (OSTI)

Feb 15, 2010 ... Testing of reactor irradiated materials for nuclear applications (fission .... Laboratory; 2Y-12 National Security Complex; 3University of Idaho

271

The Archimedes 2 mechanical assembly planning system  

SciTech Connect

We describe the implementation and performance of Archimedes 2, an integrated mechanical assembly planning system. Archimedes 2 includes two planners, two assembly sequence animation facilities, and an associated robotic workcell. Both planners use full 3 dimensional data. A rudimentary translator from high level assembly plans to control code for the robotic workcell has also been implemented. We can translate data from a commercial CAD system into input data for the system, which has allowed us to plan assembly sequences for many industrial assemblies. Archimedes 2 has been used to plan sequences for assemblies consisting of 5 to 109 parts. We have also successfully taken a CAD model of an assembly, produced an optimized assembly sequence for it, and translated the plan into robot code, which successfully assembles the device specified in the model.

Kaufman, S.G.; Wilson, R.H.; Jones, R.E.; Calton, T.L.; Ames, A.L.

1996-03-01T23:59:59.000Z

272

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

273

Circuit breaker lock out assembly  

DOE Patents (OSTI)

A lock out assembly for a circuit breaker which consists of a generally step-shaped unitary base with an aperture in the small portion of the step-shaped base and a roughly S shaped retaining pin which loops through the large portion of the step-shaped base. The lock out assembly is adapted to fit over a circuit breaker with the handle switch projecting through the aperture, and the retaining pin projecting into an opening of the handle switch, preventing removal.

Gordy, W.T.

1983-05-18T23:59:59.000Z

274

Circuit breaker lock out assembly  

Science Conference Proceedings (OSTI)

A lock out assembly for a circuit breaker which consists of a generally step-shaped unitary base with an aperture in the small portion of the step-shaped base and a roughly "S" shaped retaining pin which loops through the large portion of the step-shaped base. The lock out assembly is adapted to fit over a circuit breaker with the handle switch projecting through the aperture, and the retaining pin projecting into an opening of the handle switch, preventing removal.

Gordy, Wade T. (Jackson, SC)

1984-01-01T23:59:59.000Z

275

Ultrasonic Assembly of Thermoplastic Parts  

SciTech Connect

Four ultrasonic methods were evaluated for assembly of experimental plastic parts for detonators: (1) welding, (2) crimping and staking, (3) insertion, and (4) reactivation of adhesives. For welding, staking and insertion, plastics with low elastic moduli, such as acrylics and polycarbonate, produced the best results. Thermosetting, hot-melt, and solution adhesives could all be activated ultrasonically to form good bonds on plastics and other materials. This evaluation indicated that thermoplastic detonator parts could be assembled ultrasonically in shorter times than by present production techniques with high bond strengths and high product acceptance rates.

Schurman, W. R.

1970-03-31T23:59:59.000Z

276

irradiance | OpenEI  

Open Energy Info (EERE)

irradiance irradiance Dataset Summary Description (Abstract): Latitude Tilt Irradiance NASA Surface meteorology and Solar Energy (SSE) Release 6.0 Data Set (Jan 2008)22-year Monthly & Annual Average (July 1983 - June 2005) Parameter: Latitude Tilt Radiation (kWh/m^2/day) Internet: http://eosweb.larc.nasa.gov/sse/ Note 1: SSE Methodology & Accuracy sections online Source U.S. National Aeronautics and Space Administration (NASA), Surface meteorology and Solar Energy (SSE) Date Released March 31st, 2009 (5 years ago) Date Updated April 01st, 2009 (5 years ago) Keywords GIS global irradiance latitude mapping NASA renewable energy solar solar PV SWERA TILT UNEP Data text/csv icon Latitude Tilt Radiation (kWh/m^2/day) (csv, 11.8 MiB) application/zip icon Download Shapefile (zip, 5 MiB)

277

NREL: Performance and Reliability R&D - Field Testing  

NLE Websites -- All DOE Office Websites (Extended Search)

testing-Long-term testing of PV arrays to evaluate degradation rates. Contact Dirk Jordan. Real-Time Meteorological and Irradiance Monitoring (RMIS)-Monitoring and recording of...

278

FY 2013 Summary Report: Post-Irradiation Examination of Zircaloy...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

In the future, baseline data generated from these activities will be used to benchmark hot-cell testing of actual high-burnup UNF cladding. FY13SumRprtPostIrradiationExaminatZir...

279

Irradiation facilities at the Los Alamos Meson Physics Facility  

Science Conference Proceedings (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

280

Program for alloy development for irradiation performance in fusion reactors  

SciTech Connect

The use of fission reactors as irradiation test facilities for structural materials for a fusion environment is discussed. A comparison is made of displacement damage and helium production in fast fission and fusion reactors for stainless steel. (MOW)

Stiegler, J.O.; Reuther, T.C.

1977-01-01T23:59:59.000Z

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

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

E-Print Network (OSTI)

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

Rasala, Beth A.

2008-01-01T23:59:59.000Z

282

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

Science Conference Proceedings (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

283

Influence of the deuteron energy on the testing volume of IFMIF and its impact on other parameters  

DOE Green Energy (OSTI)

The influence of the energy of the deuteron beam on irradiation parameters of IFMIF is analyzed. The main purpose of this paper is to identify possible positive and negative impacts on irradiation parameters that an increase in the deuteron energy of the beam can cause. Several parameters of the facility, such as neutron generation rate, number of neutrons with energy above 20 MeV at the source and in the test assembly, volume with dpa rate above a threshold value, gas production, and gradient of the atomic displacement rate, are analyzed and conclusions are drawn based on the calculated values. It is shown that an increase in the deuteron energy to 40 MeV does not produce a significant negative impact for the elements analyzed, but instead is beneficial in producing nuclear responses more similar to a fusion environment than the lower deuteron energies.

Gomes, I.C.; Smith, D.L.

1995-09-01T23:59:59.000Z

284

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

285

Radial blanket assembly orificing arrangement  

DOE Patents (OSTI)

A nuclear reactor core for a liquid metal cooled fast breeder reactor is described in which means are provided for increasing the coolant flow through the reactor fuel assemblies as the reactor ages by varying the coolant flow rate with the changing coolant requirements during the core operating lifetime. (auth)

Patterson, J.F.

1975-07-01T23:59:59.000Z

286

Valve assembly for inflatable packer  

Science Conference Proceedings (OSTI)

This patent describes a valve assembly for controlling the flow of fluid from within casing into a bladder element of an inflatable packer for use in a wellbore. The bladder element has bladder port means. The valve assembly is mountable within a single cavity in a body member in the wall of casing or a casing coupling. The body member has casing port means communicating with the cavity for communicating casing fluid to the bladder port means so that the casing fluid can flow through the bladder port means to inflate the bladder element and the body member having overpressure port means communicating with the cavity for communicating part of the fluid flowing into the bladder element back to the cavity and to the valve assembly. The valve assembly comprises: control piston means movably mounted within the cavity, restraint means for insuring that the control piston means initially moves only when the pressure of the casing fluid reaches a predetermined level, and closing piston means movably mounted within the cavity and responsive to the pressure of the fluid in the bladder element, and the control piston means and the closing piston movable about stem means disposed in the cavity.

Stringfellow, W.D.

1987-12-08T23:59:59.000Z

287

De-caf-einated : life without chromatin assembly factors  

E-Print Network (OSTI)

SACCHAROMYCES CEREVISIAE CHROMATIN ASSEMBLY FACTORS THAT ACTSaccharomyces cerevisiae chromatin- assembly factors thatSaccharomyces cerevisiae chromatin-assembly factors that act

Kats, Ellen Simona

2006-01-01T23:59:59.000Z

288

Ultraviolet Germicidal Irradiation for Preventing Infectious Disease  

NLE Websites -- All DOE Office Websites (Extended Search)

Ultraviolet Germicidal Irradiation for Preventing Infectious Disease Ultraviolet Germicidal Irradiation for Preventing Infectious Disease Transmission Speaker(s): Peng Xu Date: February 19, 2002 - 12:00pm Location: Bldg. 90 The transmission of tuberculosis (TB) and other infectious diseases in health-care buildings has been a recognized hazard for decades. Ultraviolet germicidal irradiation (UVGI) of upper room air is used as an engineering control method to prevent the spread of airborne infectious disease. Under full-scale conditions, the efficacy of UVGI for inactivating airborne bacterial spores and active cells was evaluated. A test room fitted with a modern UVGI system was used to conduct bio-aerosol inactivation experiments. UVGI efficacy can be affected by environmental factors such as relative humidity (RH), and air mixing

289

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

290

Characteristics and development report for the MC4169 double-layer capacitor assembly  

Science Conference Proceedings (OSTI)

The MC4169 Double-Layer Capacitor Assembly was developed in response to a request from the B61 Systems organization to provide interim power for the B61 Common JTA Development. The project has been successfully completed, and Lot 1 has been built by MMSC/GEND. Development testing showed that this assembly met all design requirements. This report describes the design configuration, environmental testing, and aging, reliability, and safety studies done to ensure that the design requirements were met.

Clark, N.H.; Baca, W.E.

1993-09-01T23:59:59.000Z

291

Characterization of spent fuel approved testing material: ATM-106  

Science Conference Proceedings (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

292

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

SciTech Connect

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

293

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

SciTech Connect

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

294

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

SciTech Connect

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

295

Graph theoretic approach to parallel gene assembly  

Science Conference Proceedings (OSTI)

We study parallel complexity of signed graphs motivated by the highly complex genetic recombination processes in ciliates. The molecular gene assembly operations have been modeled by operations of signed graphs, i.e., graphs where the vertices have a ... Keywords: Double-split graphs, Gene assembly, Local complement, Parallel assembly, Perfect matching, Signed graphs, Split graphs

Tero Harju; Chang Li; Ion Petre

2008-11-01T23:59:59.000Z

296

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

297

Fuel or irradiation subassembly  

DOE Patents (OSTI)

A subassembly for use in a nuclear reactor is described which incorporates a loose bundle of fuel or irradiation pins enclosed within an inner tube which in turn is enclosed within an outer coolant tube and includes a locking comb consisting of a head extending through one side of the inner sleeve and a plurality of teeth which extend through the other side of the inner sleeve while engaging annular undercut portions in the bottom portion of the fuel or irradiation pins to prevent movement of the pins.

Seim, O.S.; Hutter, E.

1975-12-23T23:59:59.000Z

298

FOOD IRRADIATION REACTOR  

DOE Patents (OSTI)

An irradiation apparatus is described. It comprises a pressure vessel, a neutronic reactor active portion having a substantially greater height than diameter in the pressure vessel, an annular tank surrounding and spaced from the pressure vessel containing an aqueous indium/sup 1//sup 1//sup 5/ sulfate solution of approximately 600 grams per liter concentration, means for circulating separate coolants through the active portion and the space between the annular tank and the pressure vessel, radiator means adapted to receive the materials to be irradiated, and means for flowing the indium/sup 1//sup 1//sup 5/ sulfate solution through the radiator means.

Leyse, C.F.; Putnam, G.E.

1961-05-01T23:59:59.000Z

299

CALUTRON ASSEMBLING AND DISASSEMBLING APPARATUS  

DOE Patents (OSTI)

The construction of a calutron tank is described, whcre the face plate of the tank carrying the ion separating mechanism may be inserted or withdrawn with a minimum of difficulty, even though the plate has considerable mass and the center of gravity of the plate assembly lies within the tank. In general, the plate is pivoted at its lower end by a specially designed hinge, whereby the weight of ths plate rests on the hinge when the plato is inserted in the tank opening. A pistoncylinder arrangement is mounted on the tank and attached at the top of the plate to produce sufficient force to pivot the plate out to a point where it withdraws by its own weight and to retard the natural tendency of the plate to close with heavy impact due to the unbalanced center of gravity of the plate assembly.

Andrews, R.E.

1959-01-27T23:59:59.000Z

300

Regenerator cross arm seal assembly  

SciTech Connect

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

Jackman, Anthony V. (Indianapolis, IN)

1988-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "test assembly irradiation" from the National Library of EnergyBeta (NLEBeta).
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they are not comprehensive nor are they the most current set.
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301

Nuclear reactor composite fuel assembly  

DOE Patents (OSTI)

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

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

1980-01-01T23:59:59.000Z

302

Peach Bottom test element program. Final report  

Science Conference Proceedings (OSTI)

Thirty-three test elements were irradiated in the Peach Bottom high-temperature gas-cooled reactor (HTGR) as part of the testing program for advanced HTGRs. Extensive postirradiation examinations and evaluations of 21 of these irradiation experiments were performed. The test element irradiations were simulated using HTGR design codes and data. Calculated fuel burnups, power profiles, fast neutron fluences, and temperatures were verified via destructive burnup measurements, gamma scanning, and in-pile thermocouple readings corrected for decalibration effects. Analytical techniques were developed to improve the quality of temperature predictions through feedback of nuclear measurements into thermal calculations. Dimensional measurements, pressure burst tests, diametral compression tests, ring-cutting tests, strip-cutting tests, and four-point bend tests were performed to measure residual stress, strain, and strength distributions in H-327 graphite structures irradiated in the test elements.

Saurwein, J.J.; Holzgraf, J.F.; MIller, C.M.; Myers, B.F.; Wallroth, C.F.

1982-11-01T23:59:59.000Z

303

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

304

Rotor assembly and assay method  

DOE Patents (OSTI)

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

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

1993-01-01T23:59:59.000Z

305

NREL: Wind Research - Structural Testing Laboratory  

NLE Websites -- All DOE Office Websites (Extended Search)

Structural Testing Laboratory Structural Testing Laboratory Photo of NREL's Wind Research User Facility. Shown in front are several test bays that protect proprietary information while companies disassemble turbines to analyze, test, and modify individual components. NREL's Structural Testing Laboratory includes office space for industry researchers, houses experimental laboratories, computer facilities, space for assembling turbines, components, and blades for testing. Credit: Patrick Corkery. NREL's Structural Testing Laboratory at the National Wind Technology Center (NWTC) provides office space for industry researchers, experimental laboratories, computer facilities for analytical work, and space for assembling components and turbines for atmospheric testing. The facility also houses two blade stands equipped with overhead cranes and

306

Irradiation hardening in F82H irradiated at 573 K in the HFIR  

Science Conference Proceedings (OSTI)

Post-irradiation tensile tests were conducted on alloy F82H and variants of this steels irradiated at 573 K up to 19 dpa in the High Flux Isotope Reactor (HFIR) in Oak Ridge National Laboratory. Post-irradiation tensile and hardness tests revealed that the strength of F82H steeply increased below 5 dpa, and the total elongation decreased. The ductility of the variants, which showed more ductility in the unirradiated condition was the same as irradiated F82H, even though the magnitude of irradiation hardening is smaller than F82H. This suggests that the softened parts of the blanket, such as heat affected zones, could show more ductility loss at this temperature. The hardening behavior of F82H with 0.09% additional tantalum (mod3), which demonstrated microstructural stability under high temperature processing, was very similar to that of F82H. Therefore mod3 can be an attractive alternate structural material for a blanket when processed above 1373 K.

Stoller, Roger E [ORNL; Sokolov, Mikhail A [ORNL; Hirose, Takanori [Japan Atomic Energy Agency (JAEA); Okubo, N. [Japan Atomic Energy Agency (JAEA); Tanigawa, Hiroyasu [ORNL; Odette, G.R. [University of California, Santa Barbara; Ando, M. [Japan Atomic Energy Agency (JAEA)

2011-01-01T23:59:59.000Z

307

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

308

Camera assembly design proposal for SRF cavity image collection  

SciTech Connect

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

309

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

310

Solar Irradiance Variability  

E-Print Network (OSTI)

The Sun has long been considered a constant star, to the extent that its total irradiance was termed the solar constant. It required radiometers in space to detect the small variations in solar irradiance on timescales of the solar rotation and the solar cycle. A part of the difficulty is that there are no other constant natural daytime sources to which the Sun's brightness can be compared. The discovery of solar irradiance variability rekindled a long-running discussion on how strongly the Sun affects our climate. A non-negligible influence is suggested by correlation studies between solar variability and climate indicators. The mechanism for solar irradiance variations that fits the observations best is that magnetic features at the solar surface, i.e. sunspots, faculae and the magnetic network, are responsible for almost all variations (although on short timescales convection and p-mode oscillations also contribute). In spite of significant progress important questions are still open. Thus there is a debat...

Solanki, Sami K

2012-01-01T23:59:59.000Z

311

NNSA, Air Force Conduct Three Successful Joint Flight Tests ...  

NLE Websites -- All DOE Office Websites (Extended Search)

National LaboratoriesCalifornia. The JTA test components are manufactured at the Kansas City Plant are then assembled at the Pantex Plant in Amarillo, Texas. The NNSA Tonopah...

312

Methodology for Mechanical Property Testing on Fuel Cladding Using an Expanded Plug Wedge Test  

SciTech Connect

To determine the tensile properties of irradiated fuel cladding in a hot cell, a simple test was developed at ORNL and is described fully in US Patent Application 20060070455, Expanded plug method for developing circumferential mechanical properties of tubular materials. This method is designed for testing fuel rod cladding ductility in a hot cell utilizing an expandable plug to stretch a small ring of irradiated cladding material. The specimen strain is determined using the measured diametrical expansion of the ring. This method removes many complexities associated with specimen preparation and testing. The advantages are the simplicity of measuring the test component assembly in the hot cell and the direct measurement of specimen strain. It was also found that cladding strength could be determined from the test results. The basic approach of this test method is to apply an axial compressive load to a cylindrical plug of polyurethane (or other materials) fitted inside a short ring of the test material to achieve radial expansion of the specimen. The diameter increase of the specimen is used to calculate the circumferential strain accrued during the test. The other two basic measurements are total applied load and amount of plug compression (extension). A simple procedure is used to convert the load circumferential strain data from the ring tests into material pseudo-stress-strain curves. However, several deficiencies exist in this expanded-plug loading ring test, which will impact accuracy of test results and introduce potential shear failure of the specimen due to inherited large axial compressive stress from the expansion plug test. First of all, the highly non-uniform stress and strain distribution resulted in the gage section of the clad. To ensure reliable testing and test repeatability, the potential for highly non-uniform stress distribution or displacement/strain deformation has to be eliminated at the gage section of the specimen. Second, significant compressive stresses were induced by clad bending deformation due to a clad bulging effect (or the barreling effect). The barreling effect caused very large localized shear stress in the clad and left testing material at a high risk of shear failure. The above combined effects will result in highly non-conservative predictions both in strength and ductility of the tested clad, and the associated mechanical properties as well. To overcome/mitigate the mentioned deficiencies associated with the current expansion plug test, systematic studies have been conducted. Through detailed parameter investigation on specific geometry designs, careful filtering of material for the expansion plug, as well as adding newly designed parts to the testing system, a method to reconcile the potential non-conservatism embedded in the expansion plug test system has been discovered. A modified expansion plug testing protocol has been developed based on the method. In order to closely resemble thin-wall theory, a general procedure was also developed to determine the hoop stress in the tested ring specimen. A scaling factor called -factor is defined to correlate the ring load P into hoop stress . , = . The generated stress-strain curve agrees very well with tensile test data in both the elastic and plastic regions.

Wang, Jy-An John [ORNL; Jiang, Hao [ORNL

2013-08-01T23:59:59.000Z

313

Definition: Global horizontal irradiance | Open Energy Information  

Open Energy Info (EERE)

Normal Irradiance (DNI) and Diffuse Horizontal Irradiance (DIF).1 Related Terms DNI, Solar radiation, Concentrating solar power, Photovoltaics References http:...

314

Post-irradiation-examination of irradiated fuel outside the hot cell  

Science Conference Proceedings (OSTI)

Because of their high radioactivity, irradiated fuels are commonly examined in a hot cell. However, the Idaho National Laboratory (INL) has recently investigated irradiated U-Mo-Al metallic fuel from the Reduced Enrichment for Research and Test Reactors (RERTR) project using a conventional unshielded scanning electron microscope outside a hot cell. This examination was possible because of a two-step sample-preparation approach in which a small volume of fuel was isolated in a hot cell and shielding was introduced during later stages of sample preparation. The resulting sample contained numerous sample-preparation artifacts but allowed analysis of microstructures from selected areas.

Dawn E. Janney; Adam B. Robinson; Thomas P. O'Holleran; R. Paul Lind; Marc Babcock; Laurence C. Brower; Julie Jacobs; Pamela K. Hoggan

2007-09-01T23:59:59.000Z

315

Experimental plan for irradiation experiment HRB-21  

SciTech Connect

Irradiation experiment HRB-21 is the first in a series of test capsules that are designed to provide a fuel-performance data base to be used for the validation of modular high-temperature gas-cooled reactor (MHTGR) coated-particle fuel performance models under MHTGR normal operating conditions and specific licensing basis events. Capsule HRB-21 will contain an advanced TRISO-P UCO/ThO{sub 2} - coated-particle fuel system with demonstrated low defective-particle fraction ({le}5 {times} 10{sup {minus}5}) and a heavy metal-contamination fraction ({le}1 {times} 10{sup {minus}5}) that meets MHTGR quality specifications. The coated particles and fuel compacts were fabricated in laboratory-scale facilities using MHTGR reference procedures at General Atomics (GA). Nearly 150,000 fissile and fertile particles will be irradiated in capsule HRB-21 at a mean volumetric fuel temperature of 975{degree}C and will achieve a peak fissile burnup of 26% fissions per initial metal atom (FIMA) while accumulating a fast neutron fluence of about 4.5 {times} 10{sup 25} neutrons/m{sup 2}. This experiment is a cooperative effort between the US Department of Energy (DOE) and the Japan Atomic Energy Research Institute (JAERI). The participants are the Oak Ridge National Laboratory (ORNL), GA, and the Tokai Research Establishment. Capsule HRB-21 will contain the US MHTGR fuel specimens, and a companion capsule, HRB-22, will contain the JAERI fuel. The irradiation will take place in the removable beryllium reflector facility of the High Flux Isotope Reactor (HFIR) at ORNL. The performance of the fuel during irradiation will be closely monitored through on-line fission gas release measurements. Detailed postirradiation examination and conduction cooldown simulation testing will be performed on the irradiated fuel compacts from both the HRB-21 and HRB-22 capsules. 5 refs., 9 figs., 6 tabs.

Goodin, D. T. [General Atomics, San Diego, CA (United States); Kania, M. J.; Patton, B. W. [Oak Ridge National Lab., TN (United States)

1989-04-01T23:59:59.000Z

316

Hydrogen isotope distillation for the Tritium Systems Test Assembly  

DOE Green Energy (OSTI)

A system of four, interlinked, cryogenic fractional distillation columns has been designed as a prototype for fuel processing for fusion power reactors. The distillation system will continuously separate a feedstream of 360 g moles/day of roughly 50-50 deuterium-tritium containing approximately 1% H into four product streams: (1) a tritium-free stream of HD for waste disposal; (2) a stream of high-purity D/sub 2/ for simulated neutral beam injection; (3) a stream of DT for simulated reactor refueling; and (4) a stream of high purity T/sub 2/ for refueling and studies on properties of tritium and effects of tritium on materials.

Bartlit, J.R.; Denton, W.H.; Sherman, R.H.

1978-01-01T23:59:59.000Z

317

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

318

Ultra-Precise Assembly of Micro-Electromechanical Systems (MEMS) Components  

SciTech Connect

This report summarizes a three year effort to develop an automated microassembly workcell for the assembly of LIGA (Lithography Galvonoforming Abforming) parts. Over the last several years, Sandia has developed processes for producing surface machined silicon and LIGA parts for use in weapons surety devices. Some of these parts have outside dimensions as small as 100 micron, and most all have submicron tolerances. Parts this small and precise are extremely difficult to assembly by hand. Therefore, in this project, we investigated the technologies required to develop a robotic workcell to assembly these parts. In particular, we concentrated on micro-grippers, visual servoing, micro-assembly planning, and parallel assembly. Three different micro-grippers were tested: a pneumatic probe, a thermally actuated polysilicon tweezer, and a LIGA fabricated tweezer. Visual servoing was used to accuracy position two parts relative to one another. Fourier optics methods were used to generate synthetic microscope images from CAD drawings. These synthetic images are used off-line to test image processing routines under varying magnifications and depths of field. They also provide reference image features which are used to visually servo the part to the desired position. We also investigated a new aspect of fine motion planning for the micro-domain. As parts approach 1-10 {micro}m or less in outside dimensions, interactive forces such as van der Waals and electrostatic forces become major factors which greatly change the assembly sequence and path plans. We developed the mathematics required to determine the goal regions for pick up, holding, and release of a micro-sphere being handled by a rectangular tool. Finally, we implemented and tested the ability to assemble an array of LIGA parts attached to two 3 inch diameter wafers. In this way, hundreds of parts can be assembled in parallel rather than assembling each part individually.

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

1999-04-01T23:59:59.000Z

319

ELECTRON IRRADIATION OF SOLIDS  

DOE Patents (OSTI)

A method is presented for altering physical properties of certain solids, such as enhancing the usefulness of solids, in which atomic interchange occurs through a vacancy mechanism, electron irradiation, and temperature control. In a centain class of metals, alloys, and semiconductors, diffusion or displacement of atoms occurs through a vacancy mechanism, i.e., an atom can only move when there exists a vacant atomic or lattice site in an adjacent position. In the process of the invention highenergy electron irradiation produces additional vacancies in a solid over those normally occurring at a given temperature and allows diffusion of the component atoms of the solid to proceed at temperatures at which it would not occur under thermal means alone in any reasonable length of time. The invention offers a precise way to increase the number of vacancies and thereby, to a controlled degree, change the physical properties of some materials, such as resistivity or hardness.

Damask, A.C.

1959-11-01T23:59:59.000Z

320

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

SciTech Connect

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

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

HAPO GRAPHITE IRRADIATION CAPSULES  

SciTech Connect

A summary is presented of the broad field of graphite irradiation capsules. The various capsule designs are considered; they include temperature- controlled and temperature-monitored capsules. The components and materials of the capsules are described. Finally, methods are given for carrying out heat trandsfer calculations in capsule design and neutron spectra calculations for correlation of radiation data from different reactors. (D.L.C.)

Helm, J.W.

1963-09-18T23:59:59.000Z

322

BIOLOGICAL IRRADIATION FACILITY  

DOE Patents (OSTI)

A facility for irradiating biological specimens with neutrons is described. It includes a reactor wherein the core is off center in a reflector. A high-exposure room is located outside the reactor on the side nearest the core while a low-exposure room is located on the opposite side. Means for converting thermal neutrons to fast neutrons are movably disposed between the reactor core and the high and low-exposure rooms. (AEC)

McCorkle, W.H.; Cern, H.S.

1962-04-24T23:59:59.000Z

323

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

324

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, J.J.

1990-12-31T23:59:59.000Z

325

Fluidic self-actuating control assembly  

DOE Patents (OSTI)

A fluidic self-actuating control assembly for use in a reactor wherein no external control inputs are required to actuate (scram) the system. The assembly is constructed to scram upon sensing either a sudden depressurization of reactor inlet flow or a sudden increase in core neutron flux. A fluidic control system senses abnormal flow or neutron flux transients and actuates the system, whereupon assembly coolant flow reverses, forcing absorber balls into the reactor core region.

Grantz, Alan L. (Santa Clara, CA)

1979-01-01T23:59:59.000Z

326

Improved nuclear fuel assembly grid spacer  

DOE Patents (OSTI)

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

Marshall, John (San Jose, CA); Kaplan, Samuel (Los Gatos, CA)

1977-01-01T23:59:59.000Z

327

Fusion Test Facilities John Sheffield  

E-Print Network (OSTI)

flexing tests - Testing nuclear fuel assemblies to meltdown--PHEBUS reactor #12;#12;Released on February REACTOR--CADARACHE · Purpose: studies of hypothetical accidents in pressurized water reactors · Type: pool.78% · The reactor was transformed into a miniature PWR (scale 1/5000) for the program Phébus PFF, a study

328

Apparatus for shearing spent nuclear fuel assemblies  

DOE Patents (OSTI)

A method and apparatus are described for shearing spent nuclear fuel assemblies of the type comprising an array of fuel pins disposed within an outer metal shell or shroud. A spent fuel assembly is first compacted in a known manner and then incrementally sheared using fixed and movable shear blades having matched laterally projecting teeth which slidably intermesh to provide the desired shearing action. Incremental advancement of the fuel assembly after each shear cycle is limited to a distance corresponding to the lateral projection of the teeth to ensure fuel assembly breakup into small uniform segments which are amenable to remote chemical processing.

Weil, Bradley S. (Knoxville, TN); Metz, III, Curtis F. (Knoxville, TN)

1980-01-01T23:59:59.000Z

329

Available Technologies: Self-Assembling Small Molecule ...  

... “Efficient Small Molecule Bulk Heterojunction Solar Cells with High Fill Factors via Pyrene-Directed Molecular Self-Assembly,” Adv. Mater. 2011, ...

330

Method for shearing spent nuclear fuel assemblies  

DOE Patents (OSTI)

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

Weil, Bradley S. (Oak Ridge, TN); Watson, Clyde D. (Knoxville, TN)

1977-01-01T23:59:59.000Z

331

Assembly for directional drilling of boreholes  

Science Conference Proceedings (OSTI)

This patent describes a drilling assembly for directional drilling of boreholes in subsurface formations. The assembly comprising a downhole drilling motor. The motor having an output shaft which is suitable to drive a rotary drill bit and a motor housing which is suitable to be arranged at the lower end of a drill string; stabilizing means for stabilizing the assembly; means in the assembly for permanently tilting the central axis of the output shaft with respect to the longitudinal axis of the drill string in the borehole. It is characterized in that the stabilizing means include a lower-most stabilizer which is secured to and rotates with the output shaft.

Steiginga, A.; Worrall, R.N.

1989-11-14T23:59:59.000Z

332

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

333

North Anna 1 Cycle 15 Poolside Examination of Mark-BW Lead Assemblies with M5 Cladding  

Science Conference Proceedings (OSTI)

North Anna-1 was among the first few plants in the United States to operate with fuel assemblies having fuel cladding made of M5™ alloy, a Zr-1 percent Nb based alloy that has shown good corrosion, growth, and creep resistance essential to ensure good fuel integrity and dimensional stability at high burnups. North Anna-1 irradiated M5 cladding in four Advanced Mark-BW lead assemblies for three cycles to a peak rod burnup of about 56 GWd/MTU in late 2001. This report documents poolside inspection of...

2002-10-02T23:59:59.000Z

334

Micro-grippers for assembly of LIGA parts  

SciTech Connect

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

335

Understanding the Irradiation Behavior of Zirconium Carbide  

SciTech Connect

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

336

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

SciTech Connect

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

337

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

338

NUCLEAR REACTOR FUEL ELEMENT ASSEMBLY  

DOE Patents (OSTI)

A method of fabricating nuclear reactor fuel element assemblies having a plurality of longitudinally extending flat fuel elements in spaced parallel relation to each other to form channels is presented. One side of a flat side plate is held contiguous to the ends of the elements and a welding means is passed along the other side of the platertransverse to the direction of the longitudinal extension of the elements. The setting and speed of travel of the welding means is set to cause penetration of the side plate with welds at bridge the gap in each channel between adjacent fuel elements with a weld-through bubble of predetermined size. The fabrication of a high strength, dependable fuel element is provided, and the reduction of distortion and high production costs are facilitated by this method. (AEC)

Stengel, F.G.

1963-12-24T23:59:59.000Z

339

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

340

Diverter assembly for radioactive material  

DOE Patents (OSTI)

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

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

1988-04-11T23:59:59.000Z

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

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

342

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

SciTech Connect

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

343

Valve stem and packing assembly  

SciTech Connect

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

344

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

345

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, J.J.

1990-01-01T23:59:59.000Z

346

Fuel rod assembly to manifold attachment  

DOE Patents (OSTI)

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

Donck, Harry A. (San Diego, CA); Veca, Anthony R. (San Diego, CA); Snyder, Jr., Harold J. (San Diego, CA)

1980-01-01T23:59:59.000Z

347

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

348

Productivity Improvement of a Manual Assembly Line  

E-Print Network (OSTI)

The current project addresses the productivity improvement of a manual assembly line by making use of operations analysis in the framework of Lean production. A methodology is proposed that helps to improve the productivity of any production process. The methodology consists of selecting a product or product family to be studied followed by current process study. Once the existing process is documented, all the assembly tasks involved must be timed using time study techniques. Operations analysis enables the reduction of non-productive tasks and results in a set of standardized work elements along with the set of standard procedures for performing the operations. Assembly line balancing along with the associated operations analysis assists in constructing or re-configuring an assembly system, which is the key step in improving the overall performance of an assembly line. Following this approach, two manual assembly line configurations (single stage parallel line and five-stage serial line) are constructed for a case study. The results show that by changing over to the single stage assembly line configuration the operator productivity is doubled when compared to the existing assembly method.

Yerasi, Pranavi

2011-08-01T23:59:59.000Z

349

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

350

Structural Materials - Irradiation Studies II  

Science Conference Proceedings (OSTI)

Mar 15, 2012 ... Materials and Fuels for the Current and Advanced Nuclear Reactors: Structural Materials - Irradiation Studies II Sponsored by: The Minerals, ...

351

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

352

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

353

Carbon Characterization Laboratory Readiness to Receive Irradiated Graphite Samples  

SciTech Connect

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

Karen A. Moore

2011-05-01T23:59:59.000Z

354

The Effect of Neutron Irradiation on the Fracture Toughness of Graphite  

SciTech Connect

As part of our irradiated graphite recycle program a small quantity of PCEA grade graphite was irradiated in the High Flux Isotope Reactor (HFIR) at ORNL. The graphite will provide the raw material for future recycle experiments. The geometry of the irradiated graphite allowed us to study the effects of neutron irradiation on the Critical Stress Intensity Factor, KIc, of graphite. The specimens where irradiated in two groups of 6 at an irradiation temperature of 900 C in rabbit capsules to doses of 6.6 and 10.2 DPA, respectively. Following a full suite of pre-and post-irradiation examination, which included dimensions, mass, electrical resistivity, elastic constants, and thermal expansion (to 800 C) the samples were notched and tested to determine their KIc using the newly approved ATSM test method for SENB fracture toughness of graphite. Here we report the irradiation induced changes in the dimensions, elastic constants, resistivity, and coefficient of thermal expansion of PCEA graphite. Moreover, irradiation induced changes in the Critical Stress Intensity Factor, KIc, or fracture toughness, are reported and discussed. Very little work on the effect of neutron irradiation on the fracture toughness of graphite has previously be performed or reported.

Burchell, Timothy D [ORNL; Strizak, Joe P [ORNL

2012-01-01T23:59:59.000Z

355

Light water reactor mixed-oxide fuel irradiation experiment  

SciTech Connect

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

356

Self-Assembly of Polymer Nano-Elements on Sapphire  

NLE Websites -- All DOE Office Websites (Extended Search)

Self-Assembly of Polymer Nano-Elements on Sapphire Self-Assembly of Polymer Nano-Elements on Sapphire Print Wednesday, 25 March 2009 00:00 Self-assembly of polymers promises to...

357

Designing effective step-by-step assembly instructions  

Science Conference Proceedings (OSTI)

We present design principles for creating effective assembly instructions and a system that is based on these principles. The principles are drawn from cognitive psychology research which investigated people's conceptual models of assembly and effective ... Keywords: assembly instructions, visualization

Maneesh Agrawala; Doantam Phan; Julie Heiser; John Haymaker; Jeff Klingner; Pat Hanrahan; Barbara Tversky

2003-07-01T23:59:59.000Z

358

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

359

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

360

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 were used in experiments, and several defect characterization techniques were applied to characterize the damage. Development of dimensional changes of carbon nanotubes in microscopes operated at accelerating voltages of 30 keV revealed that binding energy of carbon atoms in CNs is much lower than in bulk materials. Resistivity measurements during irradiation demonstrated existence of a quasi state of defect creation. Linear relationship between ID/IG ratio and increasing irradiation fluence was revealed by Raman spectroscopy study of irradiated carbon buckypapers. The deviations from linear relationship were observed for the samples irradiated to very high fluence values. Annealing of irradiated samples was able to reduce the value of ID/IG ratio and remove defects. However, annealing could not affect ID/IG ratio and remove defects in amorphized samples. The extracted value of activation energy for irradiated sample was 0.36 ±0.05 eV. The value of activation energy was in good agreement with theoretical studies.

Aitkaliyeva, Assel

2009-08-01T23:59:59.000Z

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.


361

A classification scheme for LWR fuel assemblies  

Science Conference Proceedings (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

362

Definition: Irradiance | Open Energy Information  

Open Energy Info (EERE)

Irradiance Irradiance Jump to: navigation, search Dictionary.png Irradiance The direct, diffuse, and reflected solar radiation that strikes a surface. Usually expressed in kilowatts per square meter. Irradiance multiplied by time equals insolation.[1] View on Wikipedia Wikipedia Definition Irradiance is the power of electromagnetic radiation per unit area incident on a surface. Radiant emittance or radiant exitance is the power per unit area radiated by a surface. The SI units for all of these quantities are watts per square meter (W/m), while the cgs units are ergs per square centimeter per second (erg·cm·s, often used in astronomy). These quantities are sometimes called intensity, but this usage leads to confusion with radiant intensity, which has different units. All of these

363

Deputy Secretary Poneman's Remarks at the Nuclear Energy Assembly...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Energy Assembly - As Prepared for Delivery Deputy Secretary Poneman's Remarks at the Nuclear Energy Assembly - As Prepared for Delivery May 11, 2011 - 6:01pm Addthis Deputy...

364

Los Alamos shares Nano 50 award for directed assembly  

NLE Websites -- All DOE Office Websites (Extended Search)

Nano 50 award for directed assembly Los Alamos shares Nano 50 award for directed assembly Nano 50 Awards recognize "the top 50 technologies, products, and innovators that have...

365

Self-Assembly of Polymer Nano-Elements on Sapphire  

NLE Websites -- All DOE Office Websites (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...

366

Layer-by-Layer Assembled Thin Films for Battery Electrolytes  

Science Conference Proceedings (OSTI)

Presentation Title, Layer-by-Layer Assembled Thin Films for Battery Electrolytes ... Abstract Scope, Exponential layer-by-layer (eLBL) assembled battery ...

367

Low thermal resistance power module assembly - Energy Innovation ...  

A power module assembly (400) with low thermal resistance and enhanced heat dissipation to a cooling medium. The assembly includes a heat sink or spreader plate (410 ...

368

Structural analysis of an LMFBR shield assembly duct under thermo-mechanical and seismic loads  

SciTech Connect

This paper describes the stress analysis performed to assess structural adequacy of the Clinch River Breeder Reactor (CRBR) core removable shield assemblies. Removable shield assemblies are located in the peripheral region of the core (between blanket assemblies and the fixed radial shield), and are subjected to severe cross-sectional thermal gradients and seismic loads requiring a relatively complex duct load pad design. For cost-effectiveness, the analysis was conducted in two stages. First, an elasto-plastic seismic stress analysis was performed using a detailed nonlinear finite element model (with gaps) of the load pad configuration. Next, in order to determine the total strain accumulation and the creep-fatigue damage the maximum seismic stresses combined with the ''worst'' thermal stresses from a single assembly model were used to perform a simplified inelastic analysis using two sets of material properties to bound the changing material conditions during reactor operation. This work demonstrated the necessity and applicability of the two simplified analysis techniques in elevated temperature structural design, i.e., the treatment of time-dependent degradation of material properties due to temperature and nuclear irradiation, and the use of time-independent finite element stress analysis results to perform a simplified creep-fatigue analysis.

Malik, S.N.; Sazawal, V.K.

1984-06-01T23:59:59.000Z

369

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

370

10-MWe pilot-plant-receiver panel test requirements document solar thermal test facility  

DOE Green Energy (OSTI)

Testing plans for a full-scale test receiver panel and supporting hardware which essentially duplicate both physically and functionally, the design planned for the Barstow Solar Pilot Plant are presented. Testing is to include operation during normal start and shutdown, intermittent cloud conditions, and emergencies to determine the panel's transient and steady state operating characteristics and performance under conditions equal to or exceeding those expected in the pilot plant. The effects of variations of input and output conditions on receiver operation are also to be investigated. Test hardware are described, including the pilot plant receiver, the test receiver assembly, receiver panel, flow control, electrical control and instrumentation, and structural assembly. Requirements for the Solar Thermal Test Facility for the tests are given. The safety of the system is briefly discussed, and procedures are described for assembly, installation, checkout, normal and abnormal operations, maintenance, removal and disposition. Also briefly discussed are quality assurance, contract responsibilities, and test documentation. (LEW)

Not Available

1978-08-25T23:59:59.000Z

371

Gas generation from the irradiation of mortar  

DOE Green Energy (OSTI)

A mortar formulation capable of immobilizing chloride salts with high levels of radioactivity is being developed. As part of the developmental effort, radiation effects are being investigated. The radiolytic generation of gas(es) from irradiated mortar formulations was determined for several formulations with variable salt loadings at several test temperatures. The irradiation of a mortar formulation consisting of cement, slag, fly ash, water and 0 to 10 wt % salt led to the generation of hydrogen. The rate of generation was approximately constant, steady state pressures were not attained and final pressures were comparatively high. Higher salt concentrations were correlated with higher hydrogen generation rates for experiments at ambient temperature while lower rates were observed at 120/degree/C. The irradiation of a mortar consisting of cement, fly ash, water and salt led to the radiolytic generation of both oxygen and hydrogen. The addition of 2 wt % FeS or CaS inhibited oxygen generation and changed the hydrogen production rate. 10 refs., 4 figs., 3 tabs.

Lewis, M.A.; Warren, D.W.

1989-01-01T23:59:59.000Z

372

Mesoscale assembly of NiO nanosheets into spheres  

SciTech Connect

NiO solid/hollow spheres with diameters about 100 nm have been successfully synthesized through thermal decomposition of nickel acetate in ethylene glycol at 200 deg. C. These spheres are composed of nanosheets about 3-5 nm thick. Introducing poly(vinyl pyrrolidone) (PVP) surfactant to reaction system can effectively control the products' morphology. By adjusting the quantity of PVP, we accomplish surface areas-tunable NiO assembled spheres from {approx}70 to {approx}200 m{sup 2} g{sup -1}. Electrochemical tests show that NiO hollow spheres deliver a large discharge capacity of 823 mA h g{sup -1}. Furthermore, these hollow spheres also display a slow capacity-fading rate. A series of contrastive experiments demonstrate that the surface area of NiO assembled spheres has a noticeable influence on their discharge capacity. - Graphical abstract: The mesoscale assembly of NiO nanosheets into spheres have been achieved by a solvothermal method. N{sub 2} adsorption/desorption isotherms show the S{sub BET} of NiO is tunable. NiO spheres show large discharge capacity and slow capacity-fading rate.

Zhang Meng, E-mail: meng_zhang@haut.edu.c [School of Materials Science and Engineering, Henan University of Technology, Zhengzhou, Henan 450007 (China); Yan Guojin; Hou Yonggai; Wang Chunhua [School of Materials Science and Engineering, Henan University of Technology, Zhengzhou, Henan 450007 (China)

2009-05-15T23:59:59.000Z

373

ANTIGENIC STRUCTURE OF GLOBULIN FRACTION MADE FROM PERFUSE TISSUES FROM IRRADIATED DOG  

SciTech Connect

Test data on desensitization anaphylotis in guinea pigs showed decreased antigen complexes in gamma -globulin from perfuse tissues from irradiated dogs. Immunization of healthy dogs by homologous gamma -globulin from irradiated dog perfuse tissues also indicated antigenic changes. (R.V.J.)

L' vitsina, G.M.; Balin, Yu.D.

1962-01-01T23:59:59.000Z

374

Mixed Stream Test Rig (MISTER) Startup Report  

DOE Green Energy (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

375

Irradiation Effects on Human Cortical Bone Fracture Behavior  

NLE Websites -- All DOE Office Websites (Extended Search)

Irradiation Effects on Human Irradiation Effects on Human Cortical Bone Fracture Behavior Irradiation Effects on Human Cortical Bone Fracture Behavior Print Wednesday, 28 July 2010 00:00 Human bone is strong but still fallible. To better predict fracturing in bone, researchers need a mechanistic framework to understand the changes taking place on different size scales within bone, as well as the role of sustained irradiation damage. Combining in situ mechanical testing with synchrotron x-ray diffraction imaging and/or tomography, is a popular method of investigating micrometer deformation and fracture behavior in bone. However, the role that irradiation plays in these high-exposure experiments, and how it affects the properties of bone tissue, are not yet fully understood. A team of researchers led by Robert O. Ritchie at the Lawrence Berkeley National Laboratory and the University of California, Berkeley used synchrotron radiation micro-tomography at Advanced Light Source Beamline 8.3.2 to investigate changes in crack path and toughening mechanisms in human cortical bone with increased exposure to radiation, finding that exposure to high levels of irradiation can lead to drastic losses in strength, ductility, and toughness.

376

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

SciTech Connect

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

Byun, Thak Sang [ORNL; Toloczko, M [Pacific Northwest National Laboratory (PNNL); Maloy, S [Los Alamos National Laboratory (LANL)

2012-01-01T23:59:59.000Z

377

Carbon nanotubes : a study on assembly methods  

E-Print Network (OSTI)

The urgent stipulation is to manufacture CNTs of desired properties and dimensions. The heart of this yearning lies in understanding the growth and assembly methods of CNTs, which are not yet clear. In this study, hence, ...

Quiñones, Lisandro E. (Quiñones Ortiz)

2008-01-01T23:59:59.000Z

378

Self-Assembling Efficient Organic Electronics  

NLE Websites -- All DOE Office Websites (Extended Search)

Self-Assembling Efficient Organic Electronics Self-Assembling Efficient Organic Electronics Speaker(s): Rachel Segalman Date: April 26, 2005 - 12:00pm Location: Bldg. 90 Seminar Host/Point of Contact: Venkat Srinivasan In the last decade, the use of self-assembling block copolymers to nanopattern substrates and template synthesis has made incredible gains as a primary step towards the fabrication of nanodevices. Many studies have demonstrated a sophisticated level of control over the self-assembling, coil-type polymer systems to produce long range order. The knowledge now exists to begin to pattern polymers with a much higher degree of complexity and inherent functionality. It is apparent, for instance, that the mesostructure of conductive polymers impacts their luminescence and photovoltaic efficiency. For instance, block copolymers made from

379

Rack assembly for mounting solar modules  

SciTech Connect

A rack assembly is provided for mounting solar modules over an underlying body. The rack assembly may include a plurality of rail structures that are arrangeable over the underlying body to form an overall perimeter for the rack assembly. One or more retention structures may be provided with the plurality of rail structures, where each retention structure is configured to support one or more solar modules at a given height above the underlying body. At least some of the plurality of rail structures are adapted to enable individual rail structures to be sealed over the underlying body so as to constrain air flow underneath the solar modules. Additionally, at least one of (i) one or more of the rail structures, or (ii) the one or more retention structures are adjustable so as to adapt the rack assembly to accommodate solar modules of varying forms or dimensions.

Plaisted, Joshua Reed; West, Brian

2012-09-04T23:59:59.000Z

380

Rack assembly for mounting solar modules  

SciTech Connect

A rack assembly is provided for mounting solar modules over an underlying body. The rack assembly may include a plurality of rail structures that are arrangeable over the underlying body to form an overall perimeter for the rack assembly. One or more retention structures may be provided with the plurality of rail structures, where each retention structure is configured to support one or more solar modules at a given height above the underlying body. At least some of the plurality of rail structures are adapted to enable individual rail structures o be sealed over the underlying body so as to constrain air flow underneath the solar modules. Additionally, at least one of (i) one or more of the rail structures, or (ii) the one or more retention structures are adjustable so as to adapt the rack assembly to accommodate solar modules of varying forms or dimensions.

Plaisted, Joshua Reed (Oakland, CA); West, Brian (San Francisco, CA)

2010-12-28T23:59:59.000Z

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.


381

Reactivity control assembly for nuclear reactor  

DOE Patents (OSTI)

Reactivity control assembly for nuclear reactor comprises supports stacked above reactor core for holding control rods. Couplers associated with the supports and a vertically movable drive shaft have lugs at their lower ends for engagement with the supports.

Bollinger, Lawrence R. (Schenectady, NY)

1984-01-01T23:59:59.000Z

382

Method of making a unitized electrode assembly  

DOE Patents (OSTI)

A battery assembly of the consumable metal anode type has now been constructed for ready assembly as well as disassembly. In a non-conductive and at least substantially inert cell body, space is provided for receiving an open-structured, non-consumable anode cage. The cage has an open top for facilitating insertion of an anode. A modular cathode is used, comprising a peripheral current conductor frame clamped about a grid reinforced air cathode in sheet form. The air cathode may be double gridded. The cathode frame can be sealed, during assembly, with electrolyte-resistant-sealant as well as with adhesive. The resulting cathode module can be assembled outside the cell body and readily inserted therein, or can later be easily removed therefrom.

Niksa, Marilyn J. (Painesville, OH); Pohto, Gerald R. (Mentor, OH); Lakatos, Leslie K. (Mentor, OH); Wheeler, Douglas J. (Cleveland Heights, OH); Solomon, Frank (Great Neck, NY); Niksa, Andrew J. (Painesville, OH); Schue, Thomas J. (Huntsburg, OH); Genodman, Yury (Brooklyn, NY); Turk, Thomas R. (Mentor, OH); Hagel, Daniel P. (Willoughby, OH)

1988-01-01T23:59:59.000Z

383

Nonclassical assembly pathways of anisotropic particles  

E-Print Network (OSTI)

Advances in synthetic methods have spawned an array of nanoparticles and bio-inspired molecules of diverse shapes and interaction geometries. Recent experiments indicate that such anisotropic particles exhibit a variety of 'nonclassical' self-assembly pathways, forming ordered assemblies via intermediates that do not share the architecture of the bulk material. Here we apply mean field theory to a prototypical model of interacting anisotropic particles, and find a clear thermodynamic impetus for nonclassical ordering in certain regimes of parameter space. In other parameter regimes, by contrast, assembly pathways are selected by dynamics. This approach suggests a means of predicting when anisotropic particles might assemble in a manner more complicated than that assumed by classical nucleation theory.

Stephen Whitelam

2009-12-10T23:59:59.000Z

384

Unified framework for finite element assembly  

Science Conference Proceedings (OSTI)

At the heart of any finite element simulation is the assembly of matrices and vectors from discrete variational forms. We propose a general interface between problem-specific and general-purpose components of finite element programs. This interface ...

M. S. Alnaes; A. Logg; K-A. Mardal; O. Skavhaug; H. P. Langtangen

2009-11-01T23:59:59.000Z

385

BWR Assembly Optimization for Minor Actinide Recycling  

Science Conference Proceedings (OSTI)

The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs).

G. Ivan Maldonado; John M. Christenson; J.P. Renier; T.F. Marcille; J. Casal

2010-03-22T23:59:59.000Z

386

Nozzle and shroud assembly mounting structure  

DOE Patents (OSTI)

The present nozzle and shroud assembly mounting structure configuration increases component life and reduces maintenance by reducing internal stress between the mounting structure having a preestablished rate of thermal expansion and the nozzle and shroud assembly having a preestablished rate of thermal expansion being less than that of the mounting structure. The mounting structure includes an outer sealing portion forming a cradling member in which an annular ring member is slidably positioned. The mounting structure further includes an inner mounting portion to which a hooked end of the nozzle and shroud assembly is attached. As the inner mounting portion expands and contracts, the nozzle and shroud assembly slidably moves within the outer sealing portion.

Faulder, Leslie J. (San Diego, CA); Frey, deceased, Gary A. (late of Seattle, WA); Nielsen, Engward W. (El Cajon, CA); Ridler, Kenneth J. (San Diego, CA)

1997-01-01T23:59:59.000Z

387

Method of making a unitized electrode assembly  

DOE Patents (OSTI)

A battery assembly of the consumable metal anode type has now been constructed for ready assembly as well as disassembly. In a non-conductive and at least substantially inert cell body, space is provided for receiving an open-structured, non-consumable anode cage. The cage has an open top for facilitating insertion of an anode. A modular cathode is used, comprising a peripheral current conductor frame clamped about a grid reinforced air cathode in sheet form. The air cathode may be double gridded. The cathode frame can be sealed, during assembly, with electrolyte-resistant-sealant as well as with adhesive. The resulting cathode module can be assembled outside the cell body and readily inserted therein, or can later be easily removed therefrom. 6 figs.

Niksa, M.J.; Pohto, G.R.; Lakatos, L.K.; Wheeler, D.J.; Solomon, F.; Niksa, A.J.; Schue, T.J.; Genodman, Y.; Turk, T.R.; Hagel, D.P.

1988-12-06T23:59:59.000Z

388

Car Parts Self-Assembled From  

E-Print Network (OSTI)

Car Parts Self-Assembled From DNA Next: Navigation Helmet Cr SoundTech, Clay Dillow, clean energy, desali desalinization, electricity production, micro cells, wastewater Not bad for a microbe Microbial fuel cell desalinates water while generating electricity: This microbial

389

Long-life leak standard assembly  

DOE Patents (OSTI)

The present invention is directed to a portable leak standard assembly which is capable of providing a stream of high-purity reference gas at a virtually constant flow rate over an extensive period of time. The leak assembly comprises a high pressure reservoir coupled to a metal leak valve through a valve-controlled conduit. A reproducible leak valve useful in this assembly is provided by a metal tube crimped with a selected pressure loading for forming an orifice in the tube with this orifice being of a sufficient size to provide the selected flow rate. The leak valve assembly is formed of metal so that it can be "baked-out" in a vacuum furnace to rid the reservoir and attendent components of volatile impurities which reduce the efficiency of the leak standard.

Basford, James A. (Oak Ridge, TN); Mathis, John E. (Oak Ridge, TN); Wright, Harlan C. (Oak Ridge, TN)

1982-01-01T23:59:59.000Z

390

Method of forming and assembly of parts  

DOE Patents (OSTI)

A method of assembling two or more parts together that may be metal, ceramic, metal and ceramic parts, or parts that have different CTE. Individual parts are formed and sintered from particles that leave a network of interconnecting porosity in each sintered part. The separate parts are assembled together and then a fill material is infiltrated into the assembled, sintered parts using a method such as capillary action, gravity, and/or pressure. The assembly is then cured to yield a bonded and fully or near-fully dense part that has the desired physical and mechanical properties for the part's intended purpose. Structural strength may be added to the parts by the inclusion of fibrous materials.

Ripley, Edward B. (Knoxville, TN)

2010-12-28T23:59:59.000Z

391

Automated array assembly. Final report  

DOE Green Energy (OSTI)

Three main sections are included which describe a general technology assessment and manufacturing cost analysis; a near-term (1982) factory design; and the results of an experimental production study for the large-scale production of flat-panel silicon solar-cell arrays. The results of an extensive study and detailed analysis of technologies which could be related to array module manufacturing are presented. From this study, several manufacturing sequences emerge as candidates for satisfying the ERDA/JPL cost goal of $0.50/W selling price in 1986. A minimum manufacturing cost was found in a highly automated line of $0.30/W assuming the silicon is free. The panels are of a double-glass construction and are based on round wafers. Screen-printed silver has been used as the metallization with a spray-coated antireflection (AR) layer. The least expensive junction-formation technology appears to be ion implantation;however, several other technologies also may be used with very little cost penalty as described. An interim 1982 factory is described for the large-scale production of silicon solar-cell array modules. The boundary conditions for this design are the use of Czochralski silicon crystals and $25/kg polycrystalline silicon. The objective is a large-scale production facility to meet an intermediate ERDA cost goal of $2.00/W in 1982. A 6-month experimental production study of the elements of low-cost solar-cell manufacturing sequences is described as an outgrowth of the cost and manufacturing studies. This program consisted of three parts: an experimental production line study of the major variables associated with the fabrication of 3-in.-diameter silicon solar cells; a study of thick-film screen-printed silver metallization; and panel design and assembly development. (WHK)

D'Aiello, R.V.

1977-12-01T23:59:59.000Z

392

Application of hydraulically assembled shaft coupling hubs to large agitators  

SciTech Connect

This paper describes the basis for and implementation of hydraulically assembled shaft coupling hubs for large tank-mounted agitators. This modification to the original design was intended to minimize maintenance personnel exposure to ionizing radiation and also provide for disassembly capability without damage to shafts or hubs. In addition to realizing these objectives, test confirmed that the modified couplings reduced agitator shaft end runouts approximately 65%, thereby reducing bearing loads and increasing service life, a significant enhancement for a nuclear facility. 5 refs.

Murray, W.E.; Anderson, T.D. [Bechtel National, Inc., Aiken, SC (United States); Bethmann, H.K. [Westinghouse Savannah River Co., Aiken, SC (United States)

1991-12-31T23:59:59.000Z

393

Application of hydraulically assembled shaft coupling hubs to large agitators  

SciTech Connect

This paper describes the basis for and implementation of hydraulically assembled shaft coupling hubs for large tank-mounted agitators. This modification to the original design was intended to minimize maintenance personnel exposure to ionizing radiation and also provide for disassembly capability without damage to shafts or hubs. In addition to realizing these objectives, test confirmed that the modified couplings reduced agitator shaft end runouts approximately 65%, thereby reducing bearing loads and increasing service life, a significant enhancement for a nuclear facility. 5 refs.

Murray, W.E.; Anderson, T.D. (Bechtel National, Inc., Aiken, SC (United States)); Bethmann, H.K. (Westinghouse Savannah River Co., Aiken, SC (United States))

1991-01-01T23:59:59.000Z

394

Fuel cell with electrolyte matrix assembly  

DOE Patents (OSTI)

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

Kaufman, Arthur (West Orange, NJ); Pudick, Sheldon (Sayreville, NJ); Wang, Chiu L. (Edison, NJ)

1988-01-01T23:59:59.000Z

395

THERMAL HYDRAULIC ANALYSIS OF A GAS TEST LOOP SYSTEM  

Science Conference Proceedings (OSTI)

This paper discusses thermal hydraulic calculations for a Gas Test Loop (GTL) system designed to provide a high intensity fast-flux irradiation environment for testing fuels and materials for advanced concept nuclear reactors. To assess the performance of candidate reactor fuels, these fuels must be irradiated under actual fast reactor flux conditions and operating environments, preferably in an existing irradiation facility [1]. Potential users of the GTL include the Generation IV Reactor Program, the Advanced Fuel Cycle Initiative and Space Nuclear Programs.

Donna Post Guillen; James E. Fisher

2005-11-01T23:59:59.000Z

396

Simulated Space Environmental Testing on Thin Films  

Science Conference Proceedings (OSTI)

An exploratory program has been conducted, to irradiate some mature commercial and some experimental polymer films with radiation simulating certain Earth orbits, and to obtain data about the response of each test film''s reflective and tensile properties. ...

Russell Dennis A.; Fogdall Larry B.; Bohnhoff Gail

2000-04-01T23:59:59.000Z

397

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

SciTech Connect

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

398

Focusing on RISC assembly in mammalian cells  

SciTech Connect

RISC (RNA-induced silencing complex) is a central protein complex in RNAi, into which a siRNA strand is assembled to become effective in gene silencing. By using an in vitro RNAi reaction based on Drosophila embryo extract, an asymmetric model was recently proposed for RISC assembly of siRNA strands, suggesting that the strand that is more loosely paired at its 5' end is selectively assembled into RISC and results in target gene silencing. However, in the present study, we were unable to establish such a correlation in cell-based RNAi assays, as well as in large-scale RNAi data analyses. This suggests that the thermodynamic stability of siRNA is not a major determinant of gene silencing in mammalian cells. Further studies on fork siRNAs showed that mismatch at the 5' end of the siRNA sense strand decreased RISC assembly of the antisense strand, but surprisingly did not increase RISC assembly of the sense strand. More interestingly, measurements of melting temperature showed that the terminal stability of fork siRNAs correlated with the positions of the mismatches, but not gene silencing efficacy. In summary, our data demonstrate that there is no definite correlation between siRNA stability and gene silencing in mammalian cells, which suggests that instead of thermodynamic stability, other features of the siRNA duplex contribute to RISC assembly in RNAi.

Hong Junmei; Wei Na [Institute of Molecular Medicine, Peking University, 100871 Beijing (China); Chalk, Alistair [Department of Molecular Medicine and Surgery, Karolinska Institute, 171 76 Stockholm (Sweden); Wang Jue [Institute of Molecular Medicine, Peking University, 100871 Beijing (China); Song, Yutong [Department of Woman and Child Health, Karolinska Institute, 171 76 Stockholm (Sweden); Yi Fan [Institute of Molecular Medicine, Peking University, 100871 Beijing (China); Qiao Renping [State Key Laboratory of Natural and Biomimetic Drugs, School of Pharmaceutical Science, Peking University, Beijing 100083 (China); Sonnhammer, Erik L.L. [Stockholm Bioinformatics Center, 171 74 Stockholm (Sweden); Wahlestedt, Claes [Scripps Florida, Jupiter, FL 33458 (United States); Liang Zicai [Institute of Molecular Medicine, Peking University, 100871 Beijing (China); Department of Molecular Medicine and Surgery, Karolinska Institute, 171 76 Stockholm (Sweden)], E-mail: liangz@pku.edu.cn; Du, Quan [Institute of Molecular Medicine, Peking University, 100871 Beijing (China)], E-mail: quan.du@pku.edu.cn

2008-04-11T23:59:59.000Z

399

Housing assembly for electric vehicle transaxle  

DOE Patents (OSTI)

Disclosed is a drive assembly (10) for an electrically powered vehicle (12). The assembly includes a transaxle (16) having a two-speed transmission (40) and a drive axle differential (46) disposed in a unitary housing assembly (38), an oil-cooled prime mover or electric motor (14) for driving the transmission input shaft (42), an adapter assembly (24) for supporting the prime mover on the transaxle housing assembly, and a hydraulic system (172) providing pressurized oil flow for cooling and lubricating the electric motor and transaxle and for operating a clutch (84) and a brake (86) in the transmission to shift between the two-speed ratios of the transmission. The adapter assembly allows the prime mover to be supported in several positions on the transaxle housing. The brake is spring-applied and locks the transmission in its low-speed ratio should the hydraulic system fail. The hydraulic system pump is driven by an electric motor (212) independent of the prime mover and transaxle.

Kalns, Ilmars (Northville, MI)

1981-01-01T23:59:59.000Z

400

Advancing Design-for-Assembly: The Next Generation in Assembly Planning  

SciTech Connect

At the 1995 IEEE Symposium on Assembly and Task Planning, Sandia National Laboratories introduced the Archimedes 2 Software Tool [2]. The system was described as a second-generation assembly planning system that allowed preliminmy application of awembly planning for industry, while solidly supporting further research in planning techniques. Sandia has worked closely with indust~ and academia over the last four years. The results of these working relationships have bridged a gap for the next generation in assembly planning. Zke goal of this paper is to share Sandia 's technological advancements in assembly planning over the last four years and the impact these advancements have made on the manufacturing communip.

Calton, T.L.

1998-12-09T23:59:59.000Z

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

Poolside Examination of GNF BWR Fuel from Limerick: Irradiated to 65 GWd/MTU With Variations in Cladding Process and NMCA Exposure  

Science Conference Proceedings (OSTI)

Boiling water reactor (BWR) fuel assemblies of the GE11 (9x9) design that operated to ~65-GWd/MTU average exposure in the Limerick 1 reactor were examined in the site storage pool. Irradiation of these assemblies and their examination are part of a program to quantify operating margins on BWR fuel under high-duty conditions. Fuel assemblies of the GE13 (9x9) design that operated ~51 GWd/MTU also were examined to provide data on effects of microstructure on cladding corrosion at end of life exposures. The...

2003-12-16T23:59:59.000Z

402

Reconfigurable Assembly Station for Precision Manufacture of Nuclear Fusion Ignition Targets  

SciTech Connect

This paper explores the design and testing of a reconfigurable assembly station developed for assembling the inertial confinement nuclear fusion ignition targets that will be fielded in the National Ignition Facility (NIF) laser [1]. The assembly station, referred to as the Flexible Final Assembly Machine (FlexFAM) and shown in Figure 1, is a companion system to the earlier Final Assembly Machine (FAM) [2]. Both machines consist of a manipulator system integrated with an optical coordinate measuring machine (OCMM). The manipulator system has six groups of stacked axis used to manipulate the millimeter-sized target components with submicron precision, and utilizes the same force and torque feedback sensing as the FAM. Real-time dimensional metrology is provided by the OCMM's vision system and through-the-lens (TTL) laser-based height measuring probe. The manually actuated manipulator system of the FlexFAM provides a total of thirty degrees-of-freedom to the target components being assembled predominantly in a cubic centimeter work zone.

Castro, C; Montesanti, R C; Taylor, J S; Hamza, A V; Dzenitis, E G

2009-08-11T23:59:59.000Z

403

Modeling Thermal Fatigue in CPV Cell Assemblies (Presentation)  

DOE Green Energy (OSTI)

This presentation outlines the modeling of thermal fatigue in concentrating photovoltaic (CPV) assemblies.

Bosco, N.; Panchagade, D.; Kurtz, S.

2011-02-01T23:59:59.000Z

404

Impact Properties of Irradiated HT9 from the Fuel Duct of FFTF  

SciTech Connect

This paper reports Charpy impact test data for the ACO-3 duct material (HT9) from the Fast Flux Test Facility (FFTF) and its archive material. Irradiation doses for the specimens were in the range of 3 148 dpa and irradiation temperatures in the range of 378 504 oC. The impact tests were performed for the small V-notched Charpy specimens with dimensions of 3 4 27 mm at an impact speed of 3.2 m/s in a 25J capacity machine. Irradiation lowered the upper-shelf energy (USE) and increased the transition temperatures significantly. The shift of transition temperatures was greater after relatively low temperature irradiation. The USE values were in the range of 5.5 6.7 J before irradiation and decreased to the range of 2 5 J after irradiation. Lower USEs were measured for lower irradiation temperatures and specimens with T-L orientation. For the irradiated specimens, the dose dependences of transition temperature and USE were not significant because of the radiation effect on impact behavior nearly saturated at the lowest dose of about 3 dpa. A comparison showed that the lateral expansion of specimens showed a linear correlation with absorbed impact energy, but with large scatter in the results. The size effect was also discussed to clarify the differences in the impact data of subsize and standard specimens.

Byun, Thak Sang [ORNL; Maloy, S [Los Alamos National Laboratory (LANL); Toloczko, M [Pacific Northwest National Laboratory (PNNL); Lewis, William Daniel [ORNL

2012-01-01T23:59:59.000Z

405

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

E-Print Network (OSTI)

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

Pang-Hung Liu; Patrick Guenoun; Jean Daillant

2009-02-15T23:59:59.000Z

406

Irradiation Processing Department monthly report, May 1962  

SciTech Connect

This document details activities of the Irradiation Processing Department during the month of May 1962.

1962-06-13T23:59:59.000Z

407

Self-assembly of Chiral Tubules  

E-Print Network (OSTI)

The efficient and controlled assembly of complex structures from macromolecular building blocks is a critical open question in both biological systems and nanoscience. Using molecular dynamics simulations we study the self-assembly of tubular structures from model macromolecular monomers with multiple binding sites on their surfaces [Cheng et al., Soft Matter 8, 5666-5678 (2012)]. In this work we add chirality to the model monomer and a lock-and-key interaction. The self-assembly of free monomers into tubules yields a pitch value that often does not match the chirality of the monomer (including achiral monomers). We show that this mismatch occurs because of a twist deformation that brings the lateral interaction sites into alignment when the tubule pitch differs from the monomer chirality. The energy cost for this deformation is small as the energy distributions substantially overlap for small differences in the pitch and chirality. In order to control the tubule pitch by preventing the twist deformation, the interaction between the vertical surfaces must be increased without resulting in kinetically trapped structures. For this purpose, we employ the lock-and-key interactions and obtain good control of the self-assembled tubule pitch. These results explain some fundamental features of microtubules. The vertical interaction strength is larger than the lateral in microtubules because this yields a controlled assembly of tubules with the proper pitch. We also generally find that the control of the assembly into tubules is difficult, which explains the wide range of pitch and protofilament number observed in microtubule assembly.

Shengfeng Cheng; Mark J. Stevens

2013-11-19T23:59:59.000Z

408

FY 2013 Summary Report: Post-Irradiation Examination of Zircaloy-4 Samples  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Summary Report: Post-Irradiation Examination of Zircaloy-4 Summary Report: Post-Irradiation Examination of Zircaloy-4 Samples in Target Capsules and Initiation of Bending Fatigue Testing for Used Nuclear Fuel Vibration Integrity Investigations FY 2013 Summary Report: Post-Irradiation Examination of Zircaloy-4 Samples in Target Capsules and Initiation of Bending Fatigue Testing for Used Nuclear Fuel Vibration Integrity Investigations The R&D objective for this work is to conduct the separate effects tests (SET) and small-scale tests that have been identified in the Used Nuclear Fuel Storage and Transportation Data Gap Prioritization (FCRD-USED-2012-000109). R&D activities conducted during fiscal year 2013 are provided and include information derived from: 1) irradiation of hydrogen-doped zircaloy cladding in High Flux Isotope Reactor (HFIR); 2)

409

Microhole Coiled Tubing Bottom Hole Assemblies  

Science Conference Proceedings (OSTI)

The original objective of the project, to deliver an integrated 3 1/8-inch diameter Measurement While Drilling (MWD) and Logging While Drilling (LWD) system for drilling small boreholes using coiled tubing drilling, has been achieved. Two prototype systems have been assembled and tested in the lab. One of the systems has been successfully tested downhole in a conventional rotary drilling environment. Development of the 3 1/8-inch system has also lead to development and commercialization of a slightly larger 3.5-inch diameter system. We are presently filling customer orders for the 3.5-inch system while continuing with commercialization of the 3 1/8-inch system. The equipment developed by this project will be offered for sale to multiple service providers around the world, enabling the more rapid expansion of both coiled tubing drilling and conventional small diameter drilling. The project was based on the reuse of existing technology whenever possible in order to minimize development costs, time, and risks. The project was begun initially by Ultima Labs, at the time a small company ({approx}12 employees) which had successfully developed a number of products for larger oil well service companies. In September, 2006, approximately 20 months after inception of the project, Ultima Labs was acquired by Sondex plc, a worldwide manufacturer of downhole instrumentation for cased hole and drilling applications. The acquisition provided access to proven technology for mud pulse telemetry, downhole directional and natural gamma ray measurements, and surface data acquisition and processing, as well as a global sales and support network. The acquisition accelerated commercialization through existing Sondex customers. Customer demand resulted in changes to the product specification to support hotter (150 C) and deeper drilling (20,000 psi pressure) than originally proposed. The Sondex acquisition resulted in some project delays as the resistivity collar was interfaced to a different MWD system and also as the mechanical design was revised for the new pressure requirements. However, the Sondex acquisition has resulted in a more robust system, secure funding for completion of the project, and more rapid commercialization.

Don Macune

2008-06-30T23:59:59.000Z

410

MOX Lead Assembly Fabrication at the Savannah River Site  

SciTech Connect

The U. S. Department of Energy (DOE) announced its intent to prepare an Environmental Impact Statement (EIS) under the National Environmental Policy Act (NEPA) on the disposition of the nations weapon-usable surplus plutonium.This EIS is tiered from the Storage and Disposition of Weapons-Usable Fissile Material Programmatic Environmental Impact Statement issued in December 1996,and the associated Record of Decision issued on January, 1997. The EIS will examine reasonable alternatives and potential environmental impacts for the proposed siting, construction, and operation of three types of facilities for plutonium disposition. The three types of facilities are: a pit disassembly and conversion facility, a facility to immobilize surplus plutonium in a glass or ceramic form for disposition, and a facility to fabricate plutonium oxide into mixed oxide (MOX) fuel.As an integral part of the surplus plutonium program, Oak Ridge National Laboratory (ORNL) was tasked by the DOE Office of Fissile Material Disposition(MD) as the technical lead to organize and evaluate existing facilities in the DOE complex which may meet MD`s need for a domestic MOX fuel fabrication demonstration facility. The Lead Assembly (LA) facility is to produce 1 MT of usable test fuel per year for three years. The Savannah River Site (SRS) as the only operating plutonium processing site in the DOE complex, proposes two options to carry out the fabrication of MOX fuel lead test assemblies: an all Category I facility option and a combined Category I and non-Category I facilities option.

Geddes, R.L. [Westinghouse Savannah River Company, AIKEN, SC (United States); Spiker, D.L.; Poon, A.P.

1997-12-01T23:59:59.000Z

411

Simplified flangeless unisex waveguide coupler assembly  

DOE Patents (OSTI)

A unisex coupler assembly is disclosed capable of providing a leak tight coupling for waveguides with axial alignment of the waveguides and rotational capability. The sealing means of the coupler assembly are not exposed to RF energy, and the coupler assembly does not require the provision of external flanges on the waveguides. In a preferred embodiment, O ring seals are not used and the coupler assembly is, therefore, bakeable at a temperature up to about 150{degrees}C. The coupler assembly comprises a split collar which clamps around the waveguides and a second collar which fastens to the split collar. The split collar contains an inner annular groove. Each of the waveguides is provided with an external annular groove which receives a retaining ring. The split collar is clamped around one of the waveguides with the inner annular groove of the split collar engaging the retaining ring carried in the external annular groove in the waveguide. The second collar is then slipped over the second waveguide behind the annular groove and retaining ring therein and the second collar is coaxially secured by fastening means to the split collar to draw the respective waveguides together by coaxial force exerted by the second collar against the retaining ring on the second waveguide. A sealing ring is placed against an external sealing surface at a reduced external diameter end formed on one waveguide to sealingly engage a corresponding sealing surface on the other waveguide as the waveguides are urged toward each other.

DiMartino, M.; Moeller, C.P.

1992-12-31T23:59:59.000Z

412

Displacement Damage in Silicon Carbide Irradiated in Fission Reactors  

SciTech Connect

Calculations are performed for displacement damage in SiC due to irradiation in the neutron environments of various types of nuclear reactors using the best available models and nuclear data. The displacement damage calculations use recently developed damage functions for SiC that are based on extensive molecular dynamics simulations of displacement events1. Displacements per atom (DPA) cross sections for SiC have been calculated as a function of neutron energy, and they are presented here in tabular form to facilitate their use as the standard measure of displacement damage for irradiated SiC. DPA cross sections averaged over the neutron energy spectrum are calculated for neutron spectra in the cores of typical commercial reactors and in the test sample irradiation regions of several materials test reactors used in both past and present irradiation testing. Particular attention is focused on a next-generation high-temperature gas-cooled pebble bed reactor, for which the high-temperature properties of silicon carbide fiber-reinforced silicon carbide composites are well suited. Calculated transmutations and activation levels in a pebble bed reactor are compared to those in other reactors.

Heinisch, Howard L.; Greenwood, Lawrence R.; Weber, William J.; Williford, Rick E.

2004-04-05T23:59:59.000Z

413

Materials Reliability Program: Characterizations of Type 316 Cold-Worked Stainless Steel Highly Irradiated Under PWR Operating Condi tions (MRP-73)  

Science Conference Proceedings (OSTI)

Irradiation-induced material degradations such as irradiation-assisted stress corrosion cracking (IASCC), irradiation-induced void swelling, and irradiation-caused embrittlement have been observed in core internals components in pressurized water reactors (PWRs). This report describes hot cell testing and characterization of a bottom-mounted instrument tube (flux thimble) that was exposed in an operating PWR for about 23 years, providing valuable data for assessing radiation effects in PWRs.

2002-08-26T23:59:59.000Z

414

Materials Reliability Program: Characterization of Type 316 Cold Worked Stainless Steel Highly Irradiated Under PWR Operating Conditions (International IASCC Advisory Committee Phase 3 Program Final Report) (MRP-214)  

Science Conference Proceedings (OSTI)

Various types of irradiation-induced material degradation such as irradiation-assisted stress corrosion cracking (IASCC), irradiation-induced void swelling, and irradiation-caused embrittlement have been observed in core internals components in pressurized water reactors (PWR). This report describes hot cell testing and characterization of bottom-mounted instrument tubes (flux thimble) that were exposed in operating PWRs for about 10 to 20 effective full power years (EFPY), providing valuable data for as...

2007-09-06T23:59:59.000Z

415

A STUDY OF CHEMICAL CHANGES PRODUCED BY HEAT AND BY IRRADIATION OF MEAT AND MEAT FRACTIONS. Report No. 2 (Progress) for Period December 24, 1958-March 23, 1959  

SciTech Connect

Irradiated fresh and irradiated cooked ground round of beef were separated into various fractions. Methanol extraction of the residue remaining after fractionation of irradiated fresh heef gave a substance which contained the characteristic wet dog hair irradiation odor. Results of various classification tests are given which indicate that this material contains steroids and lecithin. Efforts are being made to identify the entitv responsible for this wet dog hair odor. (auth)

Landmann, W.A.

1960-10-31T23:59:59.000Z

416

EFFECTS OF GAMMA IRRADIATION ON EPDM ELASTOMERS (REVISION 1)  

SciTech Connect

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

Clark, E.

2013-09-13T23:59:59.000Z

417

Photovoltaic assemblies and methods for transporting  

DOE Patents (OSTI)

A PV assembly including framework, PV laminate(s), and a stiffening device. The framework includes a perimeter frame at least 10 feet in length and at least 5 feet in width. The PV laminate(s) are assembled to the perimeter frame to define a receiving zone having a depth of not more than 8 inches. The stiffening device is associated with the framework and is configured to provide a first state and a second state. In the first state, an entirety of the stiffening device is maintained within the receiving zone. In the second state, at least a portion of the stiffening device projects from the receiving zone. The stiffening device enhances a stiffness of the PV assembly in a plane of the perimeter frame, and can include rods defining truss structures.

Almy, Charles; Campbell, Matt; Sandler, Reuben; Wares, Brian; Wayman, Elizabeth

2013-09-17T23:59:59.000Z

418

California State Assembly | Open Energy Information  

Open Energy Info (EERE)

Assembly Assembly Jump to: navigation, search Name California State Assembly Place Sacramento, California Zip 94249-0000 Product The body of the state of California that reviews bills, laws and acts. Coordinates 38.579065°, -121.491014° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":38.579065,"lon":-121.491014,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

419

Head assembly for multiposition borehole extensometer  

DOE Patents (OSTI)

A head assembly for a borehole extensometer and an improved extensometer for measuring subsurface subsidence. A plurality of inflatable anchors provide discrete measurement points. A metering rod is fixed to each of the anchors which are displaced when subsidence occurs, thereby translating the attached rod. The head assembly includes a sprocket wheel rotatably mounted on a standpipe and engaged by a chain which is connected at one end to the metering rod and at the other end to a counterweight. A second sprocket wheel connected to the standpipe also engages the chain and drives a connected potentiometer. The head assembly converts the linear displacement of the metering rod to the rotary motion of the second sprocket wheel, which is measured by the potentiometer, producing a continuous electrical output.

Frank, Donald N. (Livermore, CA)

1983-01-01T23:59:59.000Z

420

Low thermal resistance power module assembly  

DOE Patents (OSTI)

A power module assembly with low thermal resistance and enhanced heat dissipation to a cooling medium. The assembly includes a heat sink or spreader plate with passageways or openings for coolant that extend through the plate from a lower surface to an upper surface. A circuit substrate is provided and positioned on the spreader plate to cover the coolant passageways. The circuit substrate includes a bonding layer configured to extend about the periphery of each of the coolant passageways and is made up of a substantially nonporous material. The bonding layer may be solder material which bonds to the upper surface of the plate to provide a continuous seal around the upper edge of each opening in the plate. The assembly includes power modules mounted on the circuit substrate on a surface opposite the bonding layer. The power modules are positioned over or proximal to the coolant passageways.

Hassani, Vahab (Denver, CO); Vlahinos, Andreas (Castle Rock, CO); Bharathan, Desikan (Arvada, CO)

2007-03-13T23:59:59.000Z

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.


421

Head assembly for multiposition borehole extensometer  

DOE Patents (OSTI)

A head assembly for a borehole extensometer and an improved extensometer for measuring subsurface subsidence. A plurality of inflatable anchors provide discrete measurement points. A metering rod is fixed to each of the anchors which are displaced when subsidence occurs, thereby translating the attached rod. The head assembly includes a sprocket wheel rotatably mounted on a standpipe and engaged by a chain which is connected at one end to the metering rod and at the other end to a counterweight. A second sprocket wheel connected to the standpipe also engages the chain and drives a connected potentiometer. The head assembly converts the linear displacement of the metering rod to the rotary motion of the second sprocket wheel, which is measured by the potentiometer, producing a continuous electrical output.

Frank, D.N.

1981-06-09T23:59:59.000Z

422

Inspection system performance test procedure  

SciTech Connect

This procedure establishes requirements to administer a performance demonstration test. The test is to demonstrate that the double-shell tank inspection system (DSTIS) supplied by the contractor performs in accordance with the WHC-S-4108, Double-Shell Tank Ultrasonic Inspection Performance Specification, Rev. 2-A, January, 1995. The inspection system is intended to provide ultrasonic (UT) and visual data to determine integrity of the Westinghouse Hanford Company (WHC) site underground waste tanks. The robotic inspection system consists of the following major sub-systems (modules) and components: Mobile control center; Deployment module; Cable management assembly; Robot mechanism; Ultrasonic testing system; Visual testing system; Pneumatic system; Electrical system; and Control system.

Jensen, C.E.

1995-01-17T23:59:59.000Z

423

Solvation Effects in Self-Assembled Systems  

SciTech Connect

Many types of self-assembly can be found in nature. They include crystallization, the formation of micelles, and the folding of proteins. Recently there has been much interest in pursuing nano-to-microscopically engineered materials by way of self-assembly on imprinted or templated surfaces. In all of these diverse cases, wetting plays a critical role in the assembly process. Wetting involves the interactions of the substrate or amphiphilic molecule or macromolecule with a solvent. In many self-assembled systems we find that the critical feature of the system is a substrate! or macromolecule with a both hydrophilic and hydrophobic nature. In this paper we discuss the wetting properties of a striped surface where the stripes represent alternating chemical characteristics. We show how the chemical heterogeneity affects the wetting properties of the surface (e.g. the static contact angle), and discuss the length limitations on the soft lithography approach. In this paper, the wetting of a chemically heterogeneous surface is studied using a nonlocal Density Functional Theory (DFT). The results for the heterogeneous surface model we discuss have immediate implications for soft-lithography by self-assembly. It also lends fundamental insight into the mechanisms controlling self-assembly of macromolecules. We present the results of nonlocal 2D DFT calculations on the wetting properties of chemically heterogeneous surfaces. These calculations showed complex density distributions and phase behavior as a result of the heterogeneity. The location of the wetting transition are found to be strongly dependent on the extent and strength of the heterogeneity, and complete wetting was suppressed altogether if the hydrophobic parts of the surface were large enough. In these cases, the condensed nanophase may crystallize if the hydrophilic surface-fluid interactions are strong enough. By exploring the phase space including strength of hydrophilic interactions and extent of chemical heterogeneity, an operational phase diagram was established that could be used for designing nanoscopically tailored devices and materials.

Frink, L.J.D.

1998-11-10T23:59:59.000Z

424

W-026, acceptance test report manipulator system  

Science Conference Proceedings (OSTI)

The purpose of the WRAP Manipulator System Acceptance Test Plan (ATP) is to verify that the 4 glovebox sets of WRAP manipulator components, including rail/carriage, slave arm, master controller and auxiliary equipment, meets the requirements of the functional segments of 14590 specification. The demonstration of performance elements of the ATP are performed as a part of the Assembly specifications. Manipulator integration is integrated in the performance testing of the gloveboxes. Each requirement of the Assembly specification will be carried out in conjunction with glovebox performance tests.

Watson, T.L.

1997-04-15T23:59:59.000Z

425

pH Meter probe assembly  

DOE Patents (OSTI)

An assembly for mounting a pH probe in a flowing solution, such as a sanitary sewer line, which prevents the sensitive glass portion of the probe from becoming coated with grease, oil, and other contaminants, whereby the probe gives reliable pH indication over an extended period of time. The pH probe assembly utilizes a special filter media and a timed back-rinse feature for flushing clear surface contaminants of the filter. The flushing liquid is of a known pH and is utilized to check performance of the probe.

Hale, Charles J. (San Jose, CA)

1983-01-01T23:59:59.000Z

426

pH Meter probe assembly  

DOE Patents (OSTI)

An assembly for mounting a pH probe in a flowing solution, such as a sanitary sewer line, which prevents the sensitive glass portion of the probe from becoming coated with grease, oil, and other contaminants, whereby the probe gives reliable pH indication over an extended period of time. The pH probe assembly utilizes a special filter media and a timed back-rinse feature for flushing clear surface contaminants of the filter. The flushing liquid is of a known pH and is utilized to check performance of the probe. 1 fig.

Hale, C.J.

1983-11-15T23:59:59.000Z

427

Assembly planning at the micro scale  

SciTech Connect

This paper investigates a new aspect of fine motion planning for the micro domain. As parts approach 1--10 {micro}m or less in outside dimensions, interactive forces such as van der Waals and electrostatic forces become major factors which greatly change the assembly sequence and path plans. It has been experimentally shown that assembly plans in the micro domain are not reversible, motions required to pick up a part are not the reverse of motions required to release a part. This paper develops the mathematics required to determine the goal regions for pick up, holding, and release of a micro-sphere being handled by a rectangular tool.

Feddema, J.T.; Xavier, P.; Brown, R.

1998-05-14T23:59:59.000Z

428

Electrochemical cell assembled in discharged state  

DOE Patents (OSTI)

A secondary, electrochemical cell is assembled in a completely discharged state within a sealed containment. As assembled, the cell includes a positive electrode separated from a negative electrode by a molten salt electrolyte. The positive electrode is contained within a porous structure, permitting passage of molten electrolyte, and includes one or more layers of a metallic mesh, e.g. iron, impregnated with an intimate mixture of lithium sulfide and the electrolyte. The negative electrode is a porous plaque of aluminum metal. Prior to using the cell, an electrical charge forms lithium-aluminum alloy within the negative electrode and metal sulfide within the positive electrode.

Yao, Neng-Ping (Hinsdale, IL); Walsh, William J. (Naperville, IL)

1976-01-01T23:59:59.000Z

429

Vented Cavity Radiant Barrier Assembly And Method  

DOE Patents (OSTI)

A vented cavity radiant barrier assembly (2) includes a barrier (12), typically a PV module, having inner and outer surfaces (18, 22). A support assembly (14) is secured to the barrier and extends inwardly from the inner surface of the barrier to a building surface (14) creating a vented cavity (24) between the building surface and the barrier inner surface. A low emissivity element (20) is mounted at or between the building surface and the barrier inner surface. At least part of the cavity exit (30) is higher than the cavity entrance (28) to promote cooling air flow through the cavity.

Dinwoodie, Thomas L. (Piedmont, CA); Jackaway, Adam D. (Berkeley, CA)

2000-05-16T23:59:59.000Z

430

Irradiation Effects on Human Cortical Bone Fracture Behavior  

NLE Websites -- All DOE Office Websites (Extended Search)

Irradiation Effects on Human Cortical Bone Fracture Behavior Print Irradiation Effects on Human Cortical Bone Fracture Behavior Print Human bone is strong but still fallible. To better predict fracturing in bone, researchers need a mechanistic framework to understand the changes taking place on different size scales within bone, as well as the role of sustained irradiation damage. Combining in situ mechanical testing with synchrotron x-ray diffraction imaging and/or tomography, is a popular method of investigating micrometer deformation and fracture behavior in bone. However, the role that irradiation plays in these high-exposure experiments, and how it affects the properties of bone tissue, are not yet fully understood. A team of researchers led by Robert O. Ritchie at the Lawrence Berkeley National Laboratory and the University of California, Berkeley used synchrotron radiation micro-tomography at Advanced Light Source Beamline 8.3.2 to investigate changes in crack path and toughening mechanisms in human cortical bone with increased exposure to radiation, finding that exposure to high levels of irradiation can lead to drastic losses in strength, ductility, and toughness.

431

Irradiation Effects on Human Cortical Bone Fracture Behavior  

NLE Websites -- All DOE Office Websites (Extended Search)

Irradiation Effects on Human Cortical Bone Fracture Behavior Print Irradiation Effects on Human Cortical Bone Fracture Behavior Print Human bone is strong but still fallible. To better predict fracturing in bone, researchers need a mechanistic framework to understand the changes taking place on different size scales within bone, as well as the role of sustained irradiation damage. Combining in situ mechanical testing with synchrotron x-ray diffraction imaging and/or tomography, is a popular method of investigating micrometer deformation and fracture behavior in bone. However, the role that irradiation plays in these high-exposure experiments, and how it affects the properties of bone tissue, are not yet fully understood. A team of researchers led by Robert O. Ritchie at the Lawrence Berkeley National Laboratory and the University of California, Berkeley used synchrotron radiation micro-tomography at Advanced Light Source Beamline 8.3.2 to investigate changes in crack path and toughening mechanisms in human cortical bone with increased exposure to radiation, finding that exposure to high levels of irradiation can lead to drastic losses in strength, ductility, and toughness.

432

Irradiation Effects on Human Cortical Bone Fracture Behavior  

NLE Websites -- All DOE Office Websites (Extended Search)

Irradiation Effects on Human Cortical Bone Fracture Behavior Print Irradiation Effects on Human Cortical Bone Fracture Behavior Print Human bone is strong but still fallible. To better predict fracturing in bone, researchers need a mechanistic framework to understand the changes taking place on different size scales within bone, as well as the role of sustained irradiation damage. Combining in situ mechanical testing with synchrotron x-ray diffraction imaging and/or tomography, is a popular method of investigating micrometer deformation and fracture behavior in bone. However, the role that irradiation plays in these high-exposure experiments, and how it affects the properties of bone tissue, are not yet fully understood. A team of researchers led by Robert O. Ritchie at the Lawrence Berkeley National Laboratory and the University of California, Berkeley used synchrotron radiation micro-tomography at Advanced Light Source Beamline 8.3.2 to investigate changes in crack path and toughening mechanisms in human cortical bone with increased exposure to radiation, finding that exposure to high levels of irradiation can lead to drastic losses in strength, ductility, and toughness.

433

Irradiation Effects on Human Cortical Bone Fracture Behavior  

NLE Websites -- All DOE Office Websites (Extended Search)

Irradiation Effects on Human Cortical Bone Fracture Behavior Print Irradiation Effects on Human Cortical Bone Fracture Behavior Print Human bone is strong but still fallible. To better predict fracturing in bone, researchers need a mechanistic framework to understand the changes taking place on different size scales within bone, as well as the role of sustained irradiation damage. Combining in situ mechanical testing with synchrotron x-ray diffraction imaging and/or tomography, is a popular method of investigating micrometer deformation and fracture behavior in bone. However, the role that irradiation plays in these high-exposure experiments, and how it affects the properties of bone tissue, are not yet fully understood. A team of researchers led by Robert O. Ritchie at the Lawrence Berkeley National Laboratory and the University of California, Berkeley used synchrotron radiation micro-tomography at Advanced Light Source Beamline 8.3.2 to investigate changes in crack path and toughening mechanisms in human cortical bone with increased exposure to radiation, finding that exposure to high levels of irradiation can lead to drastic losses in strength, ductility, and toughness.

434

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

SciTech Connect

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

435

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

SciTech Connect

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