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1

PIA - Advanced Test Reactor National Scientific User Facility...  

Energy Savers (EERE)

Advanced Test Reactor National Scientific User Facility Users Week 2009 PIA - Advanced Test Reactor National Scientific User Facility Users Week 2009 PIA - Advanced Test Reactor...

2

Advanced Test Reactor National Scientific User Facility  

SciTech Connect

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

2011-08-01T23:59:59.000Z

3

Idaho National Laboratory Advanced Test Reactor Probabilistic Risk  

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

Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment September 19, 2012 Presenter: Bentley Harwood, Advanced Test Reactor Nuclear Safety Engineer Battelle Energy Alliance Idaho National Laboratory Topics covered: PRA studies began in the late 1980s 1989, ATR PRA published as a summary report 1991, ATR PRA full report 1994 and 2004 various model changes 2011, Consolidation, update and improvement of previous PRA work 2012/2013, PRA risk monitor implementation Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment More Documents & Publications DOE's Approach to Nuclear Facility Safety Analysis and Management Nuclear Regulatory Commission Handling of Beyond Design Basis Events for

4

Advanced Test Reactor National Scientific User Facility Partnerships  

SciTech Connect

In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of Wisconsin-Madison; (8) Illinois Institute of Technology (IIT) Materials Research Collaborative Access Team (MRCAT) beamline at Argonne National Laboratory's Advanced Photon Source; and (9) Nanoindenter in the University of California at Berkeley (UCB) Nuclear Engineering laboratory Materials have been analyzed for ATR NSUF users at the Advanced Photon Source at the MRCAT beam, the NIST Center for Neutron Research in Gaithersburg, MD, the Los Alamos Neutron Science Center, and the SHaRE user facility at Oak Ridge National Laboratory (ORNL). Additionally, ORNL has been accepted as a partner facility to enable ATR NSUF users to access the facilities at the High Flux Isotope Reactor and related facilities.

Frances M. Marshall; Todd R. Allen; Jeff B. Benson; James I. Cole; Mary Catherine Thelen

2012-03-01T23:59:59.000Z

5

The Advanced Test Reactor National Scientific User Facility  

SciTech Connect

In 2007, the Advanced Test Reactor (ATR), located at Idaho National Laboratory (INL), was designated by the Department of Energy (DOE) as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by approved researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide those researchers with the best ideas access to the most advanced test capability, regardless of the proposer’s physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, obtained access to additional PIE equipment, taken steps to enable the most advanced post-irradiation analysis possible, and initiated an educational program and digital learning library to help potential users better understand the critical issues in reactor technology and how a test reactor facility could be used to address this critical research. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program invited universities to nominate their capability to become part of a broader user facility. Any university is eligible to self-nominate. Any nomination is then peer reviewed to ensure that the addition of the university facilities adds useful capability to the NSUF. Once added to the NSUF team, the university capability is then integral to the NSUF operations and is available to all users via the proposal process. So far, six universities have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these university capabilities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user’s technical needs. The current NSUF partners are shown in Figure 1. This article describes the ATR as well as the expanded capabilities, partnerships, and services that allow researchers to take full advantage of this national resource.

Todd R. Allen; Collin J. Knight; Jeff B. Benson; Frances M. Marshall; Mitchell K. Meyer; Mary Catherine Thelen

2011-08-01T23:59:59.000Z

6

Advanced Test Reactor Tour  

SciTech Connect

The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

Miley, Don

2011-01-01T23:59:59.000Z

7

Advanced Test Reactor Tour  

ScienceCinema (OSTI)

The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

Miley, Don

2013-05-28T23:59:59.000Z

8

Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research  

SciTech Connect

The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials with Particles and Components Testing (IMPACT) facility and the Pacific Northwest Nuclear Laboratory (PNNL) Radiochemistry Processing Laboratory (RPL) and PIE facilities were added. The ATR NSUF annually hosts a weeklong event called User’s Week in which students and faculty from universities as well as other interested parties from regulatory agencies or industry convene in Idaho Falls, Idaho to see presentations from ATR NSUF staff as well as select researchers from the materials research field. User’s week provides an overview of current materials research topics of interest and an opportunity for young researchers to understand the process of performing work through ATR NSUF. Additionally, to increase the number of researchers engaged in LWR materials issues, a series of workshops are in progress to introduce research staff to stress corrosion cracking, zirconium alloy degradation, and uranium dioxide degradation during in-reactor use.

John Jackson; Todd Allen; Frances Marshall; Jim Cole

2013-03-01T23:59:59.000Z

9

THE COMPONENT TEST FACILITY – A NATIONAL USER FACILITY FOR TESTING OF HIGH TEMPERATURE GAS-COOLED REACTOR (HTGR) COMPONENTS AND SYSTEMS  

SciTech Connect

The Next Generation Nuclear Plant (NGNP) and other High-Temperature Gas-cooled Reactor (HTGR) Projects require research, development, design, construction, and operation of a nuclear plant intended for both high-efficiency electricity production and high-temperature industrial applications, including hydrogen production. During the life cycle stages of an HTGR, plant systems, structures and components (SSCs) will be developed to support this reactor technology. To mitigate technical, schedule, and project risk associated with development of these SSCs, a large-scale test facility is required to support design verification and qualification prior to operational implementation. As a full-scale helium test facility, the Component Test facility (CTF) will provide prototype testing and qualification of heat transfer system components (e.g., Intermediate Heat Exchanger, valves, hot gas ducts), reactor internals, and hydrogen generation processing. It will perform confirmation tests for large-scale effects, validate component performance requirements, perform transient effects tests, and provide production demonstration of hydrogen and other high-temperature applications. Sponsored wholly or in part by the U.S. Department of Energy, the CTF will support NGNP and will also act as a National User Facility to support worldwide development of High-Temperature Gas-cooled Reactor technologies.

David S. Duncan; Vondell J. Balls; Stephanie L. Austad

2008-09-01T23:59:59.000Z

10

Massive Hanford Test Reactor Removed - Plutonium Recycle Test...  

Office of Environmental Management (EM)

Massive Hanford Test Reactor Removed - Plutonium Recycle Test Reactor removed from Hanford's 300 Area Massive Hanford Test Reactor Removed - Plutonium Recycle Test Reactor removed...

11

Experiment Operations Plan for a Loss-of-Coolant Accident Simulation in the National Research Universal Reactor Materials Tests 1 and 2  

SciTech Connect

A loss of Coolant Accident (LOCA) simulation program is evaluating the thermal-hydraulic and mechanical effects of LOCA conditions on pressurized water reactor test fuel bundles. This experiment operation plan for the second and third experiments of the program will provide peak fuel cladding temperatures of up to 1172K (1650{degree}F) and 1061K (1450{degree}) respectively. for a long enough time to cause test fuel cladding deformation and rupture in both. Reflood coolant delay times and the reflooding rates for the experiments were selected from thermal-hydraulic data measured in the National Research Universal (NRU) reactor facilities and test train assembly during the first experiment.

Russcher, G. E.; Wilson, C. L.; Marshall, R, K.; King, L. L.; Parchen, L. J.; Pilger, J. P.; Hesson, G. M.; Mohr, C. L.

1981-09-01T23:59:59.000Z

12

2011 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond  

SciTech Connect

This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond from November 1, 2010 through October 31, 2011. The report contains the following information: Facility and system description Permit required effluent monitoring data and loading rates Groundwater monitoring data Status of compliance activities Noncompliance and other issues Discussion of the facility's environmental impacts During the 2011 permit year, approximately 166 million gallons of wastewater were discharged to the Cold Waste Pond. This is well below the maximum annual permit limit of 375 million gallons. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters were below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

Mike Lewis

2012-02-01T23:59:59.000Z

13

Preliminary Results of an On-Line, Multi-Spectrometer Fission Product Monitoring System to Support Advanced Gas Reactor Fuel Testing and Qualification in the Advanced Test Reactor at the Idaho National Laboratory  

SciTech Connect

The Advanced Gas Reactor -1 (AGR-1) experiment is the first experiment in a series of eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments scheduled for placement in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The experiment began irradiation in the ATR with a cycle that reached full power on December 26, 2006 and will continue irradiation for about 2.5 years. During this time six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The goals of the irradiation experiment is to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. This paper presents the preliminary test details of the fuel performance, as measured by the control and acquisition software.

Dawn M. Scates; John K. Hartwell; John B. Walter; Mark W. Drigert

2007-10-01T23:59:59.000Z

14

Overview of Sandia National Laboratories pulse nuclear reactors  

SciTech Connect

Sandia National Laboratories has designed, constructed and operated bare metal Godiva-type and pool-type pulse reactors since 1961. The reactor facilities were designed to support a wide spectrum of research, development, and testing activities associated with weapon and reactor systems.

Schmidt, T.R. [Sandia National Labs., Albuquerque, NM (United States); Reuscher, J.A. [Texas A& M Univ., College Station, TX (United States)

1994-10-01T23:59:59.000Z

15

RAMI Analysis Program Design and Research for CFETR (Chinese Fusion Engineering Testing Reactor) Tokamak Machine  

Science Journals Connector (OSTI)

Chinese Fusion Engineering Testing Reactor (CFETR) is a test reactor which shall be constructed by National Integration Design Group for Magnetic Confinement Fusion Reactor of China with an ambitious scientific ...

Shijun Qin; Yuntao Song; Damao Yao; Yuanxi Wan; Songtao Wu…

2014-10-01T23:59:59.000Z

16

Experiment Operations Plan for a Loss-of-Coolant Accident Simulation in the National Research Universal Reactor Materials Test 2  

SciTech Connect

A loss-of-coolant accident (LOCA) simulation program is evaluating the thermal-hydraulic and mechanical effects on pressurized water reactor (PWR) test fuel bundles. This Experiment Operation Plan (EOP) Addendum 2, together with the referenced EOP, describes the desired operating conditions and additional hazards review associated with the four-part MT-2 experiment. The primary portions of the experiment, MT-2.2 and MT-2.3, will evaluate the following: 1) the mechanical deformation of pressurized fuel rods subjected to a slow LOCA, using reflood water for temperature control, that is designed to produce cladding temperatures in the range from 1033 to 1089K (1400 to 1500°F) for an extended time, and 2) the effects of the deformed and possibly failed cladding on the thermal-hydraulic performance of the test assembly during simulated LOCA heating and reflooding. The secondary portions of the experiment, MT-2.1 and MT-2.4, are intended to provide thermal-hydraulic calibration information during two-stage reflood conditions for 1) relatively low cladding temperatures, <839K (1050°F), on nondeformed rods, and 2) moderately high cladding temperatures, <1089K (1500°F), on deformed rods.

Russcher, G. E.; Barner, J. O.; Hesson, G. M.; Wilson, C. L.; Parchen, L. J.; Cunningham, M. E.; Marshall, R. K.; Mohr, C. L.

1981-09-01T23:59:59.000Z

17

STATEMENT OF CONSIDERATIONS Advance Test Reactor Class Waiver  

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

Advance Test Reactor Class Waiver Advance Test Reactor Class Waiver W(C)-2008-004 The Advanced Test Reactor (A TR) is a pressurized water test reactor at the Idaho National Laboratory (INL) that operates at low pressure and temperature. The ATR was originally designed to study the effects of intense radiation on reactor material and fuels . It has a "Four Leaf Clover" design that allows a diverse array of testing locations. The unique design allows for different flux in various locations and specialized systems also allow for certain experiments to be run at their own temperature and pressure. The U.S. Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007. This designation will allow the ATR to

18

A Transient Numerical Simulation of Perched Ground-Water Flow at the Test Reactor Area, Idaho National Engineering and Environmental Laboratory, Idaho, 1952-94  

SciTech Connect

Studies of flow through the unsaturated zone and perched ground-water zones above the Snake River Plain aquifer are part of the overall assessment of ground-water flow and determination of the fate and transport of contaminants in the subsurface at the Idaho National Engineering and Environmental Laboratory (INEEL). These studies include definition of the hydrologic controls on the formation of perched ground-water zones and description of the transport and fate of wastewater constituents as they moved through the unsaturated zone. The definition of hydrologic controls requires stratigraphic correlation of basalt flows and sedimentary interbeds within the saturated zone, analysis of hydraulic properties of unsaturated-zone rocks, numerical modeling of the formation of perched ground-water zones, and batch and column experiments to determine rock-water geochemical processes. This report describes the development of a transient numerical simulation that was used to evaluate a conceptual model of flow through perched ground-water zones beneath wastewater infiltration ponds at the Test Reactor Area (TRA).

B. R. Orr (USGS)

1999-11-01T23:59:59.000Z

19

2010 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond  

SciTech Connect

This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (#LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond from November 1, 2009 through October 31, 2010. The report contains the following information: • Facility and system description • Permit required effluent monitoring data and loading rates • Groundwater monitoring data • Status of compliance activities • Discussion of the facility’s environmental impacts During the 2010 permit year, approximately 164 million gallons of wastewater were discharged to the Cold Waste Pond. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters were below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

mike lewis

2011-02-01T23:59:59.000Z

20

2013 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond  

SciTech Connect

This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (#LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond from November 1, 2012–October 31, 2013. The report contains the following information: • Facility and system description • Permit required effluent monitoring data and loading rates • Groundwater monitoring data • Status of compliance activities • Noncompliance issues • Discussion of the facility’s environmental impacts. During the 2013 permit year, approximately 238 million gallons of wastewater was discharged to the Cold Waste Pond. This is well below the maximum annual permit limit of 375 million gallons. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters are below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

Mike Lewis

2014-02-01T23:59:59.000Z

Note: This page contains sample records for the topic "national reactor testing" 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

Department of Energy Designates the Idaho National Laboratory Advanced Test  

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

Designates the Idaho National Laboratory Designates the Idaho National Laboratory Advanced Test Reactor as a National Scientific User Facility Department of Energy Designates the Idaho National Laboratory Advanced Test Reactor as a National Scientific User Facility April 23, 2007 - 12:36pm Addthis WASHINGTON, DC - The U.S. Department of Energy (DOE) today designated the Idaho National Laboratory's (INL) Advanced Test Reactor (ATR) as a National Scientific User Facility. Establishing the ATR as a National Scientific User Facility will help assert U.S. leadership in nuclear science and technology, and will attract new users - universities, laboratories and industry - to conduct research at the ATR. This facility will support basic and applied nuclear research and development (R&D), furthering

22

Achievements: Nuclear Reactors designed/built by Argonne National  

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

Achievements > Achievements > Argonne National Laboratory Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy

23

RERTR 2009 (Reduced Enrichment for Research and Test Reactors)  

SciTech Connect

The U.S. Department of Energy/National Nuclear Security Administration's Office of Global Threat Reduction in cooperation with the China Atomic Energy Authority and International Atomic Energy Agency hosted the 'RERTR 2009 International Meeting on Reduced Enrichment for Research and Test Reactors.' The meeting was organized by Argonne National Laboratory, China Institute of Atomic Energy and Idaho National Laboratory and was held in Beijing, China from November 1-5, 2009. This was the 31st annual meeting in a series on the same general subject regarding the conversion of reactors within the Global Threat Reduction Initiative (GTRI). The Reduced Enrichment for Research and Test Reactors (RERTR) Program develops technology necessary to enable the conversion of civilian facilities using high enriched uranium (HEU) to low enriched uranium (LEU) fuels and targets.

Totev, T.; Stevens, J.; Kim, Y. S.; Hofman, G.; Matos, J.; Hanan, N.; Garner, P.; Dionne, B.; Olson, A.; Feldman, E.; Dunn, F.; Nuclear Engineering Division; Atomic Research Center; Inst. of Nuclear Physics; LLNL; INL; Korea Atomic Energy Research Inst.; Comisi?n Nacional de Energ?a At?mica; Nuclear Reactor Lab.; Inst. of Atomic Energy-Poland; AECL-Canada; Hungarian Academy of Sciences KFKI Atomic Energy Research Inst.; Japan Atomic Energy Agency; Nuclear Power Inst. of China; Kyoto Univ. Research Reactor Inst.

2010-03-01T23:59:59.000Z

24

The Argonaut Reactor - Reactors designed/built by Argonne National  

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

Achievements > Achievements > Argonne Reactors > Training Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy

25

CRAD, Training - Oak Ridge National Laboratory High Flux Isotope Reactor |  

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

Reactor Reactor CRAD, Training - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Training Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor. RADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Training - Oak Ridge National Laboratory High Flux Isotope Reactor More Documents & Publications CRAD, Nuclear Safety - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Training - Oak Ridge National Laboratory High Flux Isotope Reactor

26

CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor  

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

Reactor Reactor CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Maintenance Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor More Documents & Publications CRAD, Safety Basis - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor

27

In-Situ Creep Testing Capability for the Advanced Test Reactor  

SciTech Connect

An instrumented creep testing capability is being developed for specimens irradiated in Pressurized Water Reactor (PWR) coolant conditions at the Advanced Test Reactor (ATR). The test rig has been developed such that samples will be subjected to stresses ranging from 92 to 350 MPa at temperatures between 290 and 370 °C up to at least 2 dpa (displacement per atom). The status of Idaho National Laboratory (INL) efforts to develop the test rig in-situ creep testing capability for the ATR is described. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper reports efforts by INL to evaluate a prototype test rig in an autoclave at INL’s High Temperature Test Laboratory (HTTL). Initial data from autoclave tests with 304 stainless steel (304 SS) specimens are reported.

B. G. Kim; J. L. Rempe; D. L. Knudson; K. G. Condie; B. H. Sencer

2012-09-01T23:59:59.000Z

28

Electric Power Produced from Nuclear Reactor | National Nuclear...  

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

Electric Power Produced from Nuclear Reactor | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing...

29

Action Memorandum for the Engineering Test Reactor under the Idaho Cleanup Project  

SciTech Connect

This Action Memorandum documents the selected alternative for decommissioning of the Engineering Test Reactor at the Idaho National Laboratory under the Idaho Cleanup Project. Since the missions of the Engineering Test Reactor Complex have been completed, an engineering evaluation/cost analysis that evaluated alternatives to accomplish the decommissioning of the Engineering Test Reactor Complex was prepared adn released for public comment. The scope of this Action Memorandum is to encompass the final end state of the Complex and disposal of the Engineering Test Reactor vessol. The selected removal action includes removing and disposing of the vessel at the Idaho CERCLA Disposal Facility and demolishing the reactor building to ground surface.

A. B. Culp

2007-01-26T23:59:59.000Z

30

CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor  

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

Engineering - Oak Ridge National Laboratory High Flux Isotope Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Engineering Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor More Documents & Publications CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor

31

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

SciTech Connect

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

32

CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor  

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

Reactor Contractor ORR Reactor Contractor ORR CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Maintenance Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR More Documents & Publications CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor

33

National Spill Test Technology Database  

DOE Data Explorer (OSTI)

Western Research Institute established, and ACRC continues to maintain, the National Spill Technology database to provide support to the Liquified Gaseous Fuels Spill Test Facility (now called the National HAZMAT Spill Center) as directed by Congress in Section 118(n) of the Superfund Amendments and Reauthorization Act of 1986 (SARA). The Albany County Research Corporation (ACRC) was established to make publicly funded data developed from research projects available to benefit public safety. The founders since 1987 have been investigating the behavior of toxic chemicals that are deliberately or accidentally spilled, educating emergency response organizations, and maintaining funding to conduct the research at the DOEÆs HAZMAT Spill Center (HSC) located on the Nevada Test Site. ACRC also supports DOE in collaborative research and development efforts mandated by Congress in the Clean Air Act Amendments. The data files are results of spill tests conducted at various times by the Silicones Environmental Health and Safety Council (SEHSC) and DOE, ANSUL, Dow Chemical, the Center for Chemical Process Safety (CCPS) and DOE, Lawrence Livermore National Laboratory (LLNL), OSHA, and DOT; DuPont, and the Western Research Institute (WRI), Desert Research Institute (DRI), and EPA. Each test data page contains one executable file for each test in the test series as well as a file named DOC.EXE that contains information documenting the test series. These executable files are actually self-extracting zip files that, when executed, create one or more comma separated value (CSV) text files containing the actual test data or other test information.

Sheesley, David (Western Research Institute)

34

National Geothermal Data System Architecture Design, Testing...  

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

Architecture Design, Testing and Maintenance National Geothermal Data System Architecture Design, Testing and Maintenance Project objective: To create the National Geothermal Data...

35

CRAD, Management- Oak Ridge National Laboratory High Flux Isotope Reactor |  

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

Management- Oak Ridge National Laboratory High Flux Isotope Management- Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Management- Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Management in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Management- Oak Ridge National Laboratory High Flux Isotope Reactor More Documents & Publications CRAD, Nuclear Safety - Oak Ridge National Laboratory High Flux Isotope

36

Sandia National Laboratories: National Solar Thermal Test Facility  

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

Test Facility Air Force Research Laboratory Testing On August 17, 2012, in Concentrating Solar Power, Energy, Facilities, National Solar Thermal Test Facility, News, Renewable...

37

Sandia National Laboratories: National Solar Thermal Test Facility  

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

Test Facility (NSTTF) Operated by Sandia National Laboratories for the U.S. Department of Energy (DOE), the National Solar Thermal Test Facility (NSTTF) is the only test facility...

38

CRAD, Training - Oak Ridge National Laboratory High Flux Isotope Reactor  

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

Reactor Contractor ORR Reactor Contractor ORR CRAD, Training - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Training Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Training - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR More Documents & Publications CRAD, Conduct of Operations - Oak Ridge National Laboratory High Flux

39

Enhanced in-pile instrumentation at the advanced test reactor  

SciTech Connect

Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper reports results from this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and realtime flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted. (authors)

Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T.; Chase, B. M.; Palmer, J.; Condie, K. G.; Davis, K. L. [Idaho National Laboratory, MS 3840, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

2011-07-01T23:59:59.000Z

40

Sandia National Laboratories: National Solar Thermal Test Facility  

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

Visit to NSTTF On September 10, 2012, in Concentrating Solar Power, EC, National Solar Thermal Test Facility, Renewable Energy Dr. David Danielson visited Sandia National...

Note: This page contains sample records for the topic "national reactor testing" 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

CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor  

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

Oak Ridge National Laboratory High Flux Isotope Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Engineering Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory, High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR More Documents & Publications

42

CRAD, Management - Oak Ridge National Laboratory High Flux Isotope Reactor  

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

Oak Ridge National Laboratory High Flux Isotope Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Management - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Management portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Management - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR More Documents & Publications

43

Laminar Entrained Flow Reactor (Fact Sheet), National Bioenergy...  

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

Flow Reactor Investigating the core principles of in situ and ex situ catalytic fast pyrolysis of biomass NREL is a national laboratory of the U.S. Department of Energy,...

44

Using reactor operating experience to improve the design of a new Broad Application Test Reactor  

SciTech Connect

Increasing regulatory demands and effects of plant aging are limiting the operation of existing test reactors. Additionally, these reactors have limited capacities and capabilities for supporting future testing missions. A multidisciplinary team of experts developed sets of preliminary safety requirements, facility user needs, and reactor design concepts for a new Broad Application Test Reactor (BATR). Anticipated missions for the new reactor include fuels and materials irradiation testing, isotope production, space testing, medical research, fusion testing, intense positron research, and transmutation doping. The early BATR design decisions have benefited from operating experiences with existing reactors. This paper discusses these experiences and highlights their significance for the design of a new BATR.

Fletcher, C.D.; Ryskamp, J.M.; Drexler, R.L.; Leyse, C.F.

1993-07-01T23:59:59.000Z

45

Electric Power Produced from Nuclear Reactor | National Nuclear Security  

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

Electric Power Produced from Nuclear Reactor | National Nuclear Security Electric Power Produced from Nuclear Reactor | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > About Us > Our History > NNSA Timeline > Electric Power Produced from Nuclear Reactor Electric Power Produced from Nuclear Reactor December 20, 1951 Arco, ID Electric Power Produced from Nuclear Reactor

46

Maria Research Reactor loaded with LEU - Otwock, Poland | National Nuclear  

National Nuclear Security Administration (NNSA)

Maria Research Reactor loaded with LEU - Otwock, Poland | National Nuclear Maria Research Reactor loaded with LEU - Otwock, Poland | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Video Gallery > Maria Research Reactor loaded with LEU - ... Maria Research Reactor loaded with LEU - Otwock, Poland Maria Research Reactor loaded with LEU - Otwock, Poland

47

Automated Test Coverage Measurement for Reactor Protection System Software  

E-Print Network (OSTI)

Automated Test Coverage Measurement for Reactor Protection System Software Implemented in Function- ing a case study using test cases prepared by domain experts for reactor protection system software) are widely used to implement safety- critical systems such as nuclear reactor protection systems, testing

48

Plutonium Recycle Test Reactor 309 B-Roll | Department of Energy  

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

Plutonium Recycle Test Reactor 309 B-Roll Plutonium Recycle Test Reactor 309 B-Roll Addthis Description Plutonium Recycle Test Reactor 309 B-Roll...

49

A review of experiments and results from the transient reactor test (TREAT) facility.  

SciTech Connect

The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop.

Deitrich, L. W.

1998-07-28T23:59:59.000Z

50

Sandia National Laboratories: Research: Facilities: Sandia Pulsed Reactor  

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

Sandia Pulsed Reactor Facility - Critical Experiments Sandia Pulsed Reactor Facility - Critical Experiments Sandia scientist John Ford places fuel rods in the Seven Percent Critical Experiment (7uPCX) at the Sandia Pulsed Reactor Facility Critical Experiments (SPRF/CX) test reactor - a reactor stripped down to its simplest form. The Sandia Pulsed Reactor Facility - Critical Experiments (SPRF/CX) provides a flexible, shielded location for performing critical experiments that employ different reactor core configurations and fuel types. The facility is also available for hands-on nuclear criticality safety training. Research and other activities The 7% series, an evaluation of various core characteristics for higher commercial-fuel enrichment, is currently under way at the SPRF/CX. Past critical experiments at the SPRF/CX have included the Burnup Credit

51

Dual Axis Radiographic Hydrodynamic Test Facility | National...  

National Nuclear Security Administration (NNSA)

Dual Axis Radiographic Hydrodynamic Test Facility | National Nuclear Security Administration People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear...

52

test and evaluation | National Nuclear Security Administration  

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

test and evaluation | National Nuclear Security Administration People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response...

53

First Thermonuclear Device Successfully Tested | National Nuclear...  

National Nuclear Security Administration (NNSA)

Thermonuclear Device Successfully Tested | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing...

54

Sandia National Laboratories: accelerated lifetime testing  

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

accelerated lifetime testing Sandia Solar Energy Test System Cited in National Engineering Competition On May 16, 2013, in Concentrating Solar Power, Energy, Energy Storage,...

55

Decommissioning of the Tokamak Fusion Test Reactor  

SciTech Connect

The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D&D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D&D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D&D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget.

E. Perry; J. Chrzanowski; C. Gentile; R. Parsells; K. Rule; R. Strykowsky; M. Viola

2003-10-28T23:59:59.000Z

56

Reference site selection report for the advanced liquid metal reactor at the Idaho National Engineering Laboratory  

SciTech Connect

This Reference Site Selection Report was prepared by EG G, Idaho Inc., for General Electric (GE) to provide information for use by the Department of Energy (DOE) in selecting a Safety Test Site for an Advanced Liquid Metal Reactor. Similar Evaluation studies are planned to be conducted at other potential DOE sites. The Power Reactor Innovative Small Module (PRISM) Concept was developed for ALMR by GE. A ALMR Safety Test is planned to be performed on a DOE site to demonstrate features and meet Nuclear Regulatory Commission Requirements. This study considered possible locations at the Idaho National Engineering Laboratory that met the ALMR Prototype Site Selection Methodology and Criteria. Four sites were identified, after further evaluation one site was eliminated. Each of the remaining three sites satisfied the criteria and was graded. The results were relatively close. Thus concluding that the Idaho National Engineering Laboratory is a suitable location for an Advanced Liquid Metal Reactor Safety Test. 23 refs., 13 figs., 9 tabs.

Sivill, R.L.

1990-03-01T23:59:59.000Z

57

Multi-physics Reactor Performance and Safety Simulations - Argonne National  

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

Engineering Computation Engineering Computation and Design > Multi-physics Reactor Performance and Safety Simulations Capabilities Engineering Computation and Design Engineering and Structural Mechanics Systems/Component Design, Engineering and Drafting Heat Transfer and Fluid Mechanics Multi-physics Reactor Performance and Safety Simulations Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Multi-physics Reactor Performance and Safety Simulations Bookmark and Share Contact Keith S. Bradley, Ph.D. Technical Director, Nuclear Engineering Division Argonne National Laboratory Email address protected by JavaScript. Please enable JavaScript The SHARP simulation suite development team, led by Argonne National Laboratory, includes other leading national laboratories and research universities. SHARP is developed under the auspices of the U.S. Department of Energy, Office of Nuclear Energy, Nuclear Energy Advanced Modeling and Simulation Program (NEAMS).

58

Sandia National Laboratories: National Solar Thermal Test Facility  

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

Center in Vermont Achieves Milestone Installation On September 23, 2014, in Concentrating Solar Power, Energy, Facilities, National Solar Thermal Test Facility, News, News &...

59

Sandia National Laboratories: National Solar Thermal Test Facility  

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

Sandia Wins Three R&D100 Awards On July 24, 2013, in Concentrating Solar Power, Energy, Facilities, National Solar Thermal Test Facility, News, News & Events, Photovoltaic,...

60

Sandia National Laboratories: National Solar Thermal Test Facility  

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

Solar Power Technical Management Position On July 12, 2012, in Concentrating Solar Power, Energy, Facilities, Job Listing, National Solar Thermal Test Facility, News,...

Note: This page contains sample records for the topic "national reactor testing" 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

Sandia National Laboratories: National Solar Thermal Test Facility  

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

Funding Award On June 4, 2014, in Advanced Materials Laboratory, Concentrating Solar Power, Energy, Energy Storage, Facilities, National Solar Thermal Test Facility,...

62

Sandia National Laboratories: National Solar Thermal Test Facility  

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

Better Sandia Capabilities to Support Power Industry On January 8, 2013, in Concentrating Solar Power, Energy, Energy Storage, Facilities, National Solar Thermal Test Facility,...

63

Sandia National Laboratories: Mechanical Testing  

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

EnergyNuclear Energy Systems Laboratory (NESL) Brayton LabMechanical Testing Mechanical Testing Mechanical Testing Overview Mechanical 1-2 (2008). Standard Test Methods for...

64

Massive Hanford Test Reactor Removed- Plutonium Recycle Test Reactor removed from Hanford’s 300 Area  

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

RICHLAND, WA – Hanford’s River Corridor contractor, Washington Closure Hanford, has met a significant cleanup challenge on the U.S. Department of Energy’s (DOE) Hanford Site by removing a 1,082-ton nuclear test reactor from the 300 Area.

65

Sandia National Laboratories: National Solar Thermal Test Facility  

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

ECISEnergyRenewable EnergySolar EnergyConcentrating Solar Power ECISEnergyRenewable EnergySolar EnergyConcentrating Solar Power (CSP)National Solar Thermal Test Facility National Solar Thermal Test Facility NSTTF Interactive Tour National Solar Thermal Test Facility (NSTTF) Operated by Sandia National Laboratories for the U.S. Department of Energy (DOE), the National Solar Thermal Test Facility (NSTTF) is the only test facility of this type in the United States. The NSTTF's primary goal is to provide experimental engineering data for the design, construction, and operation of unique components and systems in proposed solar thermal electrical plants planned for large-scale power generation. In addition, the site was built and instrumented to provide test facilities for a variety of solar and nonsolar applications. The facility can provide

66

CRAD, Engineering- Oak Ridge National Laboratory High Flux Isotope Reactor  

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

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Engineering Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

67

CRAD, Management- Oak Ridge National Laboratory High Flux Isotope Reactor  

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

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Management in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

68

CRAD, Training- Oak Ridge National Laboratory High Flux Isotope Reactor  

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

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Training Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

69

Sandia National Laboratories: Solar Regional Test Center  

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

Center in Vermont Achieves Milestone Installation On September 23, 2014, in Concentrating Solar Power, Energy, Facilities, National Solar Thermal Test Facility, News, News &...

70

Sandia National Laboratories Test Capabilities Revitalization...  

National Nuclear Security Administration (NNSA)

Test Capabilities Revitalization Phase 2 Project Completed On Time, Under Budget | National Nuclear Security Administration People Mission Managing the Stockpile Preventing...

71

Sandia National Laboratories: Reactor Component Testing  

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

Doppler Velocimeter EC Top Publications A Comparison of Platform Options for Deep-water Floating Offshore Vertical Axis Wind Turbines: An Initial Study Nonlinear Time-Domain...

72

EPR/PTFE dosimetry for test reactor environments  

SciTech Connect

The use of Electron Paramagnetic Resonance (EPR) spectroscopy with materials such as alanine is well established as a technique for measurement of ionizing radiation absorbed dose in photon and electron fields such as Co-60, high-energy bremsstrahlung and electron-beam fields [1]. In fact, EPR/Alanine dosimetry has become a routine transfer standard for national standards bodies such as NIST and NPL. In 1992 the Radiation Metrology Laboratory (RML) at Sandia National Laboratories implemented EPR/Alanine capabilities for use in routine and calibration activities at its Co-60 and pulsed-power facilities. At that time it also investigated the usefulness of the system for measurement of absorbed dose in the mixed neutron/photon environments of reactors such as the Sandia Pulsed Reactor and the Annular Core Research Reactor used for hardness testing of electronics. The RML concluded that the neutron response of alanine was a sufficiently high fraction of the overall dosimeter response that the resulting uncertainties in the photon dose would be unacceptably large for silicon-device testing. However, it also suggested that non-hydrogenous materials such as polytetrafluoroethylene (PTFE) would exhibit smaller neutron response and might be useful in mixed environments. Preliminary research with PTFE in photon environments indicated considerable promise, but further development was not pursued at that time. Because of renewed interest in absorbed dose measurements that could better define the individual contributions of photon and neutron components to the overall dose delivered to a test object, the RML has re-initiated the development of an EPR/PTFE dosimetry system. This effort consists of three stages: 1) Identification of PTFE materials that may be suitable for dosimetry applications. It was speculated that the inconsistency of EPR signatures in the earlier samples may have been due to variability in PTFE manufacturing processes. 2) Characterization of dosimetry in photon-only environments. This is necessary to establish requirements for sample preparation, operating parameters and limitations for use in well-defined and predictable environments prior to deployment in the less well-defined mixed environments of test reactors. 3) Characterization of the EPR responses obtained with PTFE in mixed neutron/photon fields. This includes evaluation of the neutron and photon contributions to response, determination of applicable of neutron fluence and photon dose ranges. This paper presents a summary of the research, a description of the EPR/PTFE dosimetry system, and recommendations for preparation and fielding of the dosimetry in photon and mixed neutron/photon environments. (authors)

Vehar, D.W.; Griffin, P.J.; Quirk, T.J. [Sandia National Laboratories, Albuquerque, NM 87185-1146 (United States)

2011-07-01T23:59:59.000Z

73

National SCADA Test Bed | Department of Energy  

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

Cybersecurity » Energy Delivery Systems Cybersecurity Cybersecurity » Energy Delivery Systems Cybersecurity » National SCADA Test Bed National SCADA Test Bed Created in 2003, the National SCADA Test Bed (NSTB) is a one-of-a-kind national resource that draws on the integrated expertise and capabilities of the Argonne, Idaho, Lawrence Berkeley, Los Alamos, Oak Ridge, Pacific Northwest, and Sandia National Laboratories to address the cybersecurity challenges of energy delivery systems. Core and Frontier Research The NSTB core capabilities combine a network of the national labs' state-of-the-art operational system testing facilities with expert research, development, analysis, and training to discover and address critical security vulnerabilities and threats the energy sector faces. NSTB offers testing and research facilities, encompassing field-scale control

74

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

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

Printer Friendly Printer Friendly Engineering Test Reactor (ETR) Vessel Relocated Engineering Test Reactor Vessel Pre-startup 1957 Click on image to enlarge. Image 1 of 5 Gantry jacks attached to ETR vessel. Initial lift starts. - Click on image to enlarge. Image 2 of 5 ETR vessel removed from substructure. Vessel lifted approximately 40 ft. - Click on image to enlarge. On Monday, September 24, 2007 the Engineering Test Reactor (ETR) vessel was removed from its location and delivered to the Idaho CERCLA Disposal Facility (ICDF). The long history of the ETR began for this water-cooled reactor with its start up in 1957, after taking only 2 years to build. According to "Proving the Principles," by Susan M. Stacy: When the Engineering Test Reactor started up at the Test Reactor Area in

75

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

76

Blade Testing at NREL's National Wind Technology Center (NWTC) (Presentation)  

SciTech Connect

Presentation of Blade Testing at NREL's National Wind Technology Center for the 2010 Sandia National Laboratories Blade Testing Workshop.

Hughes, S.

2010-07-20T23:59:59.000Z

77

Risk management at the Oak Ridge National Laboratory research reactors  

SciTech Connect

In November of 1986, the High Flux Isotope Reactor (HFIR) was shut down by Oak Ridge National Laboratory (ORNL) due to a concern regarding embrittlement of the reactor vessel. A massive review effort was undertaken by ORNL and the Department of Energy (DOE). This review resulted in an extensive list of analyses and design modifications to be completed before restart could take place. The review also focused on the improvement of management practices including implementation of several of the Institute of Nuclear Power Operations (INPO) requirements. One of the early items identified was the need to perform a Probabilistic Risk Assessment (PRA) on the reactor. It was decided by ORNL management that this PRA would not be just an exercise to assess the ``bottom`` line in order to restart, but would be used to improve the overall safety of the reactor, especially since resources (both manpower and dollars) were severely limited. The PRA would become a basic safety tool to be used instead of a more standard deterministic approach to safety used in commercial reactor power plants. This approach was further reinforced, because the reactor was nearly 25 years old at this time, and the design standards and regulations had changed significantly since the original design, and many of the safety issues could not be addressed by compliance to codes and standards.

Flanagan, G.F.; Linn, M.A.; Proctor, L.D.; Cook, D.H.

1994-12-31T23:59:59.000Z

78

Project Profile: National Solar Thermal Test Facility  

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

The first solar receivers ever tested in the world were tested at the National Solar Thermal Test Facility (NSTTF). The receivers were each rated up to 5 megawatts thermal (MWt). Receivers with various working fluids have been tested here over the years, including air, water-steam, molten salt, liquid sodium, and solid particles. The NSTTF has also been used for a large variety of other tests, including materials tests, simulation of thermal nuclear pulses and aerodynamic heating, and ablator testing for NASA.

79

Sandia National Laboratories: Sandia National Laboratories: Tonopah Test  

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

Tonopah Test Range Tonopah Test Range Tonopah Tonopah Test Range (TTR) is the testing range of choice for all national security missions. Sandia conducts operations at TTR in support of the Department of Energy/National Nuclear Security Administration's weapons programs. Principal DOE activities at TTR include stockpile reliability testing; arming, fusing, and firing systems testing; and the testing of nuclear weapon delivery systems. The range also offers a unique test environment for use by other U.S. government agencies and their contractors. Located about 160 miles northwest of Las Vegas, TTR is an immense area of flat terrain ideal for rockets and low-altitude, high-speed aircraft operations. Situated between two mountain ranges, TTR's remote location and restricted airspace ensure that tests can be conducted with a high degree

80

Reactor and Material Supply | Y-12 National Security Complex  

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

Reactor and Reactor and Material Supply Reactor and Material Supply Y-12 has processed highly enriched uranium for more than 60 years in support of the nation's defense. The end of the Cold War and ensuing strategic arms control treaties have resulted in an excess of HEU materials. In 1994, approximately 174 metric tons of weapons-usable HEU was declared surplus to defense needs. The HEU disposition program was established to make the surplus HEU unsuitable for use in weapons by blending it down to low-enriched uranium and to recover the economic value of the materials to the extent practical. In 2005, the Secretary of Energy announced that an additional 200 metric tons of HEU would be removed from further use as fissile material in U.S. nuclear weapons. Approximately 20 metric tons of this material will

Note: This page contains sample records for the topic "national reactor testing" 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.


81

I. INTRODUCTION The Tokamak Fusion Test Reactor (TFTR) is a  

E-Print Network (OSTI)

I. INTRODUCTION The Tokamak Fusion Test Reactor (TFTR) is a one­of­a­kind, tritium­fueled fusion of the Tokamak present a unique and challenging task for dismantling. Figure 1 ­ TFTR II. Project Overview The Decommissioning and Decontamination of the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics

82

TR-EDB: Test Reactor Embrittlement Data Base, Version 1  

SciTech Connect

The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.

Stallmann, F.W.; Wang, J.A.; Kam, F.B.K. [Oak Ridge National Lab., TN (United States)] [Oak Ridge National Lab., TN (United States)

1994-01-01T23:59:59.000Z

83

Relay testing at Brookhaven National Laboratory  

SciTech Connect

Brookhaven National Laboratory (BNL) is conducting a seismic test program on relays. The purpose of the test program is to investigate the influence of various designs, electrical and vibration parameters on the seismic capacity levels. The first series of testing has been completed and performed at Wyle Laboratories. The major part of the test program consisted of single axis, single frequency sine dwell tests. Random multiaxis, multifrequency tests were also performed. Highlights of the test results as well as a description of the testing methods are presented in this paper. 10 figs.

Bandyopadhyay, K.; Hofmayer, C.

1989-01-01T23:59:59.000Z

84

Hydrogen loops in existing reactors for testing fuel elements for nuclear propulsion  

Science Journals Connector (OSTI)

The Space Exploration Initiative (SEI) has revitalized interest in adapting nuclear energy for power and propulsion. Prior to the selection of a nuclear thermal propulsion (NTP) system extensive testing of the various proposed concepts will be required. In today’s environmental safety and health culture full size rocket engine tests as were done under the Rover/NERVA program will be extremely difficult and expensive to perform and meet NASA’s schedules. A different test strategy uses a hydrogen loop in an existing reactor to test a wide variety of single elements or clusters of elements for fuel qualification. This approach is expected to reduce operating and capital costs and expedite the testing schedule. This paper examines the potential of performing subscale tests in a hydrogen loop in an existing reactor such as the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory or the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory. The HFIR is expected to achieve power densities comparable to those achieved in ATR because of the 85 MWt power level and the high thermal and fast flux levels. The available length and diameter of the test region of FHIR are 60 cm and 10 cm whereas the available length and diameter of the test region of ATR are 120 cm and 12 cm respectively.

Charles S. Olsen; Henry Welland; James Abraschoff; Kenneth Thoms

1993-01-01T23:59:59.000Z

85

National Geothermal Data System Design and Testing | Department...  

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

Design and Testing National Geothermal Data System Design and Testing National Geothermal Data System Design and Testing presentation at the April 2013 peer review meeting held in...

86

2013 Radiological Monitoring Results Associated with the Advanced Test Reactor Complex Cold Waste Pond  

SciTech Connect

This report summarizes radiological monitoring performed of the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste wastewater prior to discharge into the Cold Waste Pond and of specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit (#LA-000161-01, Modification B). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

Mike Lewis

2014-02-01T23:59:59.000Z

87

E-Print Network 3.0 - application test reactor Sample Search...  

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

test reactor Search Powered by Explorit Topic List Advanced Search Sample search results for: application test reactor Page: << < 1 2 3 4 5 > >> 1 Engineers at Western are...

88

Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)  

SciTech Connect

This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

Bradley K. Heath

2014-03-01T23:59:59.000Z

89

DOI Designates B Reactor at DOE's Hanford Site as a National Historic  

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

DOI Designates B Reactor at DOE's Hanford Site as a National DOI Designates B Reactor at DOE's Hanford Site as a National Historic Landmark DOI Designates B Reactor at DOE's Hanford Site as a National Historic Landmark August 25, 2008 - 3:20pm Addthis DOE to offer regular public tours in 2009 WASHINGTON, DC - U.S. Department of the Interior (DOI) Deputy Secretary Lynn Scarlett and U.S. Department of Energy (DOE) Acting Deputy Secretary Jeffrey F. Kupfer today announced the designation of DOE's B Reactor as a National Historic Landmark and unveiled DOE's plan for a new public access program to enable American citizens to visit B Reactor during the 2009 tourist season. The B Reactor at DOE's Hanford Site in southeast Washington State was the world's first industrial-scale nuclear reactor and produced plutonium for the atomic weapon that was dropped on Nagasaki,

90

DOI Designates B Reactor at DOE's Hanford Site as a National Historic  

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

DOI Designates B Reactor at DOE's Hanford Site as a National DOI Designates B Reactor at DOE's Hanford Site as a National Historic Landmark DOI Designates B Reactor at DOE's Hanford Site as a National Historic Landmark August 25, 2008 - 3:20pm Addthis DOE to offer regular public tours in 2009 WASHINGTON, DC - U.S. Department of the Interior (DOI) Deputy Secretary Lynn Scarlett and U.S. Department of Energy (DOE) Acting Deputy Secretary Jeffrey F. Kupfer today announced the designation of DOE's B Reactor as a National Historic Landmark and unveiled DOE's plan for a new public access program to enable American citizens to visit B Reactor during the 2009 tourist season. The B Reactor at DOE's Hanford Site in southeast Washington State was the world's first industrial-scale nuclear reactor and produced plutonium for the atomic weapon that was dropped on Nagasaki,

91

Early Exploration - Reactors designed/built by Argonne National Laboratory  

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

Early Exploration Early Exploration About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

92

Independent Oversight Inspection, Idaho National Laboratory- June 2005  

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

Inspection of Environment, Safety, and Health Programs at the Idaho National Laboratory Advanced Test Reactor

93

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor  

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

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC)

94

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor  

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

(RISMC) Advanced Test (RISMC) Advanced Test Reactor Demonstration Case Study Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for

95

Department of Energy Designates the Idaho National Laboratory...  

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

Designates the Idaho National Laboratory Advanced Test Reactor as a National Scientific User Facility Department of Energy Designates the Idaho National Laboratory Advanced Test...

96

Materials Test-2 LOCA Simulation in the NRU Reactor  

SciTech Connect

A simulated loss-of-coolant accident was performed with a full-length test bundle of pressurized water reactor fuel rods. This third experiment of the program produced fuel cladding temperatures exceeding 1033 K (1400°F) for 155 s and resulted in eight ruptured fuel rods. Experiment data and initial results are presented in the form of photographs and graphical summaries.

Barner, J. O.; Hesson, G. M.; King, I. L.; Marshall, R. K.; Parchen, L. J.; Pilger, J. P.; Rausch, W. N.; Russcher, G. E.; Webb, B. J.; Wildung, N. J.; Wilson, C. L.; Wismer, M. D.; Mohr, C. L.

1982-03-01T23:59:59.000Z

97

Reactor protection system with automatic self-testing and diagnostic  

DOE Patents (OSTI)

A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically ``identical`` values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic. 16 figs.

Gaubatz, D.C.

1996-12-17T23:59:59.000Z

98

Why Nuclear Energy? - Reactors designed/built by Argonne National  

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

Nuclear Energy: Nuclear Energy: Why Nuclear Energy? About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

99

LOCA simulation in the NRU reactor: materials test-1  

SciTech Connect

A simulated loss-of-coolant accident was performed with a full-length test bundle of pressurized water reactor fuel rods. This second experiment of the program produced peak fuel cladding temperatures of 1148K (1607/sup 0/F) and resulted in six ruptured fuel rods. Test data and initial results from the experiment are presented here in the form of photographs and graphical summaries. These results are also compared with the preceding prototypic thermal-hydraulic test results and with computer model test predictions.

Russcher, G.E.; Marshall, R.K.; Hesson, G.M.; Wildung, N.J.; Rausch, W.N.

1981-10-01T23:59:59.000Z

100

Nuclear Weapons Testing Resumes | National Nuclear Security Administra...  

National Nuclear Security Administration (NNSA)

Testing Resumes | National Nuclear Security Administration People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response...

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


101

Office of Test and Evaluation | National Nuclear Security Administrati...  

National Nuclear Security Administration (NNSA)

Test and Evaluation | National Nuclear Security Administration People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response...

102

The High Flux Isotope Reactor at Oak Ridge National Laboratory  

NLE Websites -- All DOE Office Websites

The High Flux Isotope Reactor at ORNL The High Flux Isotope Reactor at ORNL Aerial of the High Flux Isotope Reactor Site The High Flux Isotope Reactor site is located on the south side of the ORNL campus and is about a three-minute drive from her sister neutron facility, the Spallation Neutron Source. Operating at 85 MW, HFIR is the highest flux reactor-based source of neutrons for research in the United States, and it provides one of the highest steady-state neutron fluxes of any research reactor in the world. The thermal and cold neutrons produced by HFIR are used to study physics, chemistry, materials science, engineering, and biology. The intense neutron flux, constant power density, and constant-length fuel cycles are used by more than 500 researchers each year for neutron scattering research into

103

National SCADA Test Bed - Enhancing control systems security...  

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

SCADA Test Bed - Enhancing control systems security in the energy sector (September 2009) National SCADA Test Bed - Enhancing control systems security in the energy sector...

104

E-Print Network 3.0 - aircraft shield test reactor Sample Search...  

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

shield test reactor Search Powered by Explorit Topic List Advanced Search Sample search results for: aircraft shield test reactor Page: << < 1 2 3 4 5 > >> 1 A' Brief. History of...

105

Results of the DF-4 BWR (boiling water reactor) control blade-channel box test  

SciTech Connect

The DF-4 in-pile fuel damage experiment investigated the behavior of boiling water reactor (BWR) fuel canisters and control blades in the high temperature environment of an unrecovered reactor accident. This experiment, which was carried out in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories, was performed under the USNRC's internationally sponsored severe fuel damage (SFD) program. The DF-4 test is described herein and results from the experiment are presented. Important findings from the DF-4 test include the low temperature melting of the stainless steel control blade caused by reaction with the B{sub 4}C, and the subsequent low temperature attack of the Zr-4 channel box by the relocating molten blade components. Hydrogen generation was found to continue throughout the experiment, diminishing slightly following the relocation of molten oxidizing zircaloy to the lower extreme of the test bundle. A large blockage which was formed from this material continued to oxidize while steam was being fed into the the test bundle. The results of this test have provided information on the initial stages of core melt progression in BWR geometry involving the heatup and cladding oxidation stages of a severe accident and terminating at the point of melting and relocation of the metallic core components. The information is useful in modeling melt progression in BWR core geometry, and provides engineering insight into the key phenomena controlling these processes. 12 refs., 12 figs.

Gauntt, R.O.; Gasser, R.D.

1990-10-01T23:59:59.000Z

106

Proceedings of the 1990 International Meeting on Reduced Enrichment for Research and Test Reactors  

SciTech Connect

The global effort to reduce, and possibly, eliminate the international traffic in highly-enriched uranium caused by its use in research reactors requires extensive cooperation and free exchange of information among all participants. To foster this free exchange of information, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the thirteenth of a series which began in 1978. The common effort brought together, past, a large number of specialists from many countries. On hundred twenty-three participants from 26 countries, including scientists, reactor operators, and personnel from commercial fuel suppliers, research centers, and government organizations, convened in Newport, Rhode Island to discuss their results, their activities, and their plans relative to converting research reactors to low-enriched fuels. As more and more reactors convert to the use of low-enriched uranium, the emphasis of our effort has begun to shift from research and development to tasks more directly related to implementation of the new fuels and technologies that have been developed, and to refinements of those fuels and technologies. It is appropriate, for this reason, that the emphasis of this meeting was placed on safety and on conversion experiences. This individual papers in this report have been cataloged separately.

Not Available

1993-07-01T23:59:59.000Z

107

Thermal effects testing at the National Solar Thermal Test Facility  

SciTech Connect

The National Solar Thermal Test Facility is operated by Sandia National Laboratories and located on Kirkland Air Force Base in Albuquerque, New Mexico. The permanent features of the facility include a heliostat field and associated receiver tower, two solar furnaces, two point-focus parabolic concentrators, and Engine Test Facility. The heliostat field contains 220 computer-controlled mirrors, which reflect concentrated solar energy to test stations on a 61-m tower. The field produces a peak flux density of 250 W/cm{sup 2} that is uniform over a 15-cm diameter with a total beam power of over 5 MW{sub t}. The solar beam has been used to simulate aerodynamic heating for several customers. Thermal nuclear blasts have also been simulated using a high-speed shutter in combination with heliostat control. The shutter can accommodate samples up to 1 m {times} 1 m and it has been used by several US and Canadian agencies. A glass-windowed wind tunnel is also available in the Solar Tower. It provides simultaneous exposure to the thermal flux and air flow. Each solar furnace at the facility includes a heliostat, an attenuator, and a parabolic concentrator. One solar furnace produces flux levels of 270 W/cm{sup 2} over and delivers a 6-mm diameter and total power of 16 kW{sub t}. A second furnace produces flux levels up to 1000 W/cm{sup 2} over a 4 cm diameter and total power of 60 kW{sub t}. Both furnaces include shutters and attenuators that can provide square or shaped pulses. The two 11 m diameter tracking parabolic point-focusing concentrators at the facility can each produce peak flux levels of 1500 W/cm{sup 2} over a 2.5 cm diameter and total power of 75 kW{sub t}. High-speed shutters have been used to produce square pulses.

Ralph, M.E.; Cameron, C.P. [Sandia National Labs., Albuquerque, NM (United States); Ghanbari, C.M. [Technadyne Engineering Consultants, Inc., Albuquerque, NM (United States)

1992-12-31T23:59:59.000Z

108

Thermal effects testing at the National Solar Thermal Test Facility  

SciTech Connect

The National Solar Thermal Test Facility is operated by Sandia National Laboratories and located on Kirkland Air Force Base in Albuquerque, New Mexico. The permanent features of the facility include a heliostat field and associated receiver tower, two solar furnaces, two point-focus parabolic concentrators, and Engine Test Facility. The heliostat field contains 220 computer-controlled mirrors, which reflect concentrated solar energy to test stations on a 61-m tower. The field produces a peak flux density of 250 W/cm[sup 2] that is uniform over a 15-cm diameter with a total beam power of over 5 MW[sub t]. The solar beam has been used to simulate aerodynamic heating for several customers. Thermal nuclear blasts have also been simulated using a high-speed shutter in combination with heliostat control. The shutter can accommodate samples up to 1 m [times] 1 m and it has been used by several US and Canadian agencies. A glass-windowed wind tunnel is also available in the Solar Tower. It provides simultaneous exposure to the thermal flux and air flow. Each solar furnace at the facility includes a heliostat, an attenuator, and a parabolic concentrator. One solar furnace produces flux levels of 270 W/cm[sup 2] over and delivers a 6-mm diameter and total power of 16 kW[sub t]. A second furnace produces flux levels up to 1000 W/cm[sup 2] over a 4 cm diameter and total power of 60 kW[sub t]. Both furnaces include shutters and attenuators that can provide square or shaped pulses. The two 11 m diameter tracking parabolic point-focusing concentrators at the facility can each produce peak flux levels of 1500 W/cm[sup 2] over a 2.5 cm diameter and total power of 75 kW[sub t]. High-speed shutters have been used to produce square pulses.

Ralph, M.E.; Cameron, C.P. (Sandia National Labs., Albuquerque, NM (United States)); Ghanbari, C.M. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (United States))

1992-01-01T23:59:59.000Z

109

LOSS-OF-COOLANT ACIDENT SIMULATIONS IN THE NATIONAL RESEARCH UNIVERSAL REACTOR  

SciTech Connect

Pressurized water reactor loss-of-coolant accident (LOCA) phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship among the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. Subsequent experiments establish the fuel rod failure characteristics at selected peak cladding temperatures. Fuel rod cladding pressurization simulates high burnup fission gas pressure levels of modern PWRs. This document contains both an experiment overview of the LOCA simulation program and a review of the safety analyses performed by Pacific Northwest Laboratory (PNL) to define the expected operating conditions as well as to evaluate the worst case operating conditions. The primary intent of this document is to supply safety information required by the Chalk River Nuclear Laboratories (CRNL), to establish readiness to proceed from one test phase to the next and to establish the overall safety of the experiment. A hazards review summarizes safety issues, normal operation and three worst case accidents that have been addressed during the development of the experiment plan.

Bennett, W D; Goodman, R L; Heaberlin, S W; Hesson, G M; Nealley, C; Kirg, L L; Marshall, R K; McNair, G W; Meitzler, W D; Neally, G W; Parchen, L J; Pilger, J P; Rausch, W N; Russcher, G E; Schreiber, R E; Wildung, N J

1981-02-01T23:59:59.000Z

110

Scaling analysis for a reactor vessel mixing test  

SciTech Connect

The Westinghouse AP600 advanced pressurized water reactor design uses a gravity-forced safety injection system with two nozzles in the reactor vessel downcomer. In the event of a severe overcooling transient such as a steam-line break, this system delivers boron to the core to offset positive reactivity introduced by the negative moderator defect. To determine if the system design is capable of successfully terminating this type of reactivity transient, a test of the system has been initiated. The test will utilize a 1:9 scale model of the reactor vessel and cold legs. The coolant will be modeled with air, while the safety injection fluid will be simulated with a dense gas. To determine the necessary parameters for this model, a scaling analysis was performed. The continuity, diffusion, and axial Navier-Stokes equations for the injected fluid were converted into dimensionless form. A Boussinesq formulation for turbulent viscosity was applied in these equations. This procedure identified the Richardson, mixing Reynolds, diffusion Fourier, and Euler numbers as dimensionless groups of interest. Order-of-magnitude evaluation was used to determine that the Richardson and mixing Reynolds numbers were the most significant parameters to match for a similar experiment.

Radcliff, T.D.; Parsons, J.R.; Johnson, W.S. (Univ. of Tennessee, Knoxville, TN (United States)); Ekeroth, D.E. (Westinghouse Electric Corp., Pittsburgh, PA (United States))

1993-01-01T23:59:59.000Z

111

Sandia National Laboratories: National Solar Thermal Test Facility  

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

in Concentrating Solar Power, Customers & Partners, Energy, News, Partnership, Renewable Energy, Solar Areva Solar is collaborating with Sandia National Laboratories on a new...

112

Sandia National Laboratories: Engine Test Facility  

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

FacilityEngine Test Facility Engine Test Facility Test Cell 1 Test Cell 2 DataControl Room Maintenance Assembly Bay Test Cell 1 This testing area is primarily configured to...

113

Vertical Pretreatment Reactor System (Poster), NREL (National Renewable Energy Laboratory)  

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

Vertical Pretreatment Reactor System Vertical Pretreatment Reactor System Two-vessel system for primary and secondary pretreatment at diff erent temperatures * Biomass is heated by steam injection to temperatures of 120°C to 210°C in the pressurized mixing tube * Preheated, premixed biomass is retained for specified residence time in vertical holding vessel; material continuously moves by gravity from top to bottom of reactor in plug-fl ow fashion * Residence time is adjusted by changing amount of material held in vertical vessel relative to continuous fl ow of material entering and exiting vessel * Optional additional reactor vessel allows for secondary pretreatment at lower temperatures-120°C to 180°C-with potential to add other chemical catalysts * First vessel can operate at residence

114

Domestic U.S. Reactor Conversions: Fact Sheet | National Nuclear Security  

National Nuclear Security Administration (NNSA)

Domestic U.S. Reactor Conversions: Fact Sheet | National Nuclear Security Domestic U.S. Reactor Conversions: Fact Sheet | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Fact Sheets > Domestic U.S. Reactor Conversions: Fact Sheet Fact Sheet Domestic U.S. Reactor Conversions: Fact Sheet Mar 23, 2012 The National Nuclear Security Administration (NNSA) helps convert research

115

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

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

2014-01-01T23:59:59.000Z

116

An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory  

SciTech Connect

Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs.

Boing, L.E.; Henley, D.R. (Argonne National Lab., IL (USA)); Manion, W.J.; Gordon, J.W. (Nuclear Energy Services, Inc., Danbury, CT (USA))

1989-12-01T23:59:59.000Z

117

Sandia National Laboratories: National Solar Thermal Test Facility  

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

Salt Initial Flow Testing is a Tremendous Success On November 2, 2012, in Concentrating Solar Power, News, Renewable Energy, Solar The Molten Salt Test Loop (MSTL ) system at...

118

Alpha Particle Physics Experiments in the Tokamak Fusion Test Reactor  

SciTech Connect

Alpha particle physics experiments were done on the Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium (DT) run from 1993-1997. These experiments utilized several new alpha particle diagnostics and hundreds of DT discharges to characterize the alpha particle confinement and wave-particle interactions. In general, the results from the alpha particle diagnostics agreed with the classical single-particle confinement model in magnetohydrodynamic (MHD) quiescent discharges. Also, the observed alpha particle interactions with sawteeth, toroidal Alfvén eigenmodes (TAE), and ion cyclotron resonant frequency (ICRF) waves were roughly consistent with theoretical modeling. This paper reviews what was learned and identifies what remains to be understood.

Budny, R.V.; Darrow, D.S.; Medley, S.S.; Nazikian, R.; Zweben, S.J.; et al.

1998-12-14T23:59:59.000Z

119

Sandia National Laboratories: Dish Test Facility  

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

Engine Test Facility Central Receiver Test Facility Power Towers for Utilities Solar Furnace Dish Test Facility Optics Lab Parabolic Dishes Work For Others (WFO) User...

120

Sandia National Laboratories: Test and Evaluation  

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

Engine Test Facility Central Receiver Test Facility Power Towers for Utilities Solar Furnace Dish Test Facility Optics Lab Parabolic Dishes Work For Others (WFO) User...

Note: This page contains sample records for the topic "national reactor testing" 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

Sandia National Laboratories: Regional Test Facility  

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

Engine Test Facility Central Receiver Test Facility Power Towers for Utilities Solar Furnace Dish Test Facility Optics Lab Parabolic Dishes Work For Others (WFO) User...

122

Sandia National Laboratories: Test Site Operations & Maintenance...  

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

Engine Test Facility Central Receiver Test Facility Power Towers for Utilities Solar Furnace Dish Test Facility Optics Lab Parabolic Dishes Work For Others (WFO) User...

123

Sandia National Laboratories: Molten Salt Test Loop  

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

Engine Test Facility Central Receiver Test Facility Power Towers for Utilities Solar Furnace Dish Test Facility Optics Lab Parabolic Dishes Work For Others (WFO) User...

124

Sandia National Laboratories: Central Receiver Test Facility  

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

Engine Test Facility Central Receiver Test Facility Power Towers for Utilities Solar Furnace Dish Test Facility Optics Lab Parabolic Dishes Work For Others (WFO) User...

125

Tag: Naval Reactors | Y-12 National Security Complex  

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

Naval Reactors Naval Reactors Tag: Naval Reactors Displaying 1 - 7 of 7... Category: Employees & Retirees "Cook"ing at Y-12 for 70 years We have an enduring mission. Y-12 plays a key role in it. And a nuclear deterrent remains the ultimate insurance policy for America. More... Category: News Y-12 Knows Uranium Y-12 produces many forms of uranium. More... Category: News A Rich Resource Requires Recovery Given the value and scarcity of enriched uranium, Y-12 recycles and reuses as much of it as possible. More... Category: News Seawolf Manufacturing Challenge For decades, attack submarines were either fast or quiet - but never both. The fast subs were so loud that an enemy could hear them long before they were within striking distance. More... Category: News Reliable fuel source

126

Horizontal Pretreatment Reactor System (Poster), NREL (National Renewable Energy Laboratory)  

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

Diff Diff erent pretreatment chemistry/ residence time combinations are possible using these multiple horizontal-tube reactors * Each tube is indirectly and directly steam heated to temperatures of 150 0 C to 210 0 C * Residence time is varied by changing the speed of the auger that moves the biomass through each tube reactor * Tubes are used individually or in combination to achieve diff erent pretreatment residence times * Smaller tubes made from Hastelloy, an acid-resistant material, are used with more corrosive chemicals and residence times from 3 to 20 minutes * Larger tubes made from 316 stainless steel are used for residence times from 20 to 120 minutes Horizontal Pretreatment Reactor System Versatile pretreatment system for a wide range of pretreatment chemistries

127

Vortex Diode Analysis and Testing for Fluoride Salt-Cooled High-Temperature Reactors  

SciTech Connect

Fluidic diodes are presently being considered for use in several fluoride salt-cooled high-temperature reactor designs. A fluidic diode is a passive device that acts as a leaky check valve. These devices are installed in emergency heat removal systems that are designed to passively remove reactor decay heat using natural circulation. The direct reactor auxiliary cooling system (DRACS) uses DRACS salt-to-salt heat exchangers (DHXs) that operate in a path parallel to the core flow. Because of this geometry, under normal operating conditions some flow bypasses the core and flows through the DHX. A flow diode, operating in reverse direction, is-used to minimize this flow when the primary coolant pumps are in operation, while allowing forward flow through the DHX under natural circulation conditions. The DRACSs reject the core decay heat to the environment under loss-of-flow accident conditions and as such are a reactor safety feature. Fluidic diodes have not previously been used in an operating reactor system, and therefore their characteristics must be quantified to ensure successful operation. This report parametrically examines multiple design parameters of a vortex-type fluidic diode to determine the size of diode needed to reject a particular amount of decay heat. Additional calculations were performed to size a scaled diode that could be tested in the Oak Ridge National Laboratory Liquid Salt Flow Loop. These parametric studies have shown that a 152.4 mm diode could be used as a test article in that facility. A design for this diode is developed, and changes to the loop that will be necessary to test the diode are discussed. Initial testing of a scaled flow diode has been carried out in a water loop. The 150 mm diode design discussed above was modified to improve performance, and the final design tested was a 171.45 mm diameter vortex diode. The results of this testing indicate that diodicities of about 20 can be obtained for diodes of this size. Experimental results show similar trends as the computational fluid dynamics (CFD) results presented in this report; however, some differences exist that will need to be assessed in future studies. The results of this testing will be used to improve the diode design to be tested in the liquid salt loop system.

Yoder Jr, Graydon L [ORNL; Elkassabgi, Yousri M. [Texas A& M University, Kingsville; De Leon, Gerardo I. [Texas A& M University, Kingsville; Fetterly, Caitlin N. [Texas A& M University, Kingsville; Ramos, Jorge A. [Texas A& M University, Kingsville; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK)

2012-02-01T23:59:59.000Z

128

Tritium production analysis and management strategies for a Fluoride-salt-cooled high-temperature test reactor (FHTR)  

E-Print Network (OSTI)

The Fluoride-salt-cooled High-temperature Test Reactor (FHTR) is a test reactor concept that aims to demonstrate the neutronics, thermal-hydraulics, materials, tritium management, and to address other reactor operational ...

Rodriguez, Judy N

2013-01-01T23:59:59.000Z

129

Sandia National Laboratories: Solar Regional Test Center in Vermont...  

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

Center in Vermont Achieves Milestone Installation On September 23, 2014, in Concentrating Solar Power, Energy, Facilities, National Solar Thermal Test Facility, News, News &...

130

Sandia National Laboratories: NASA's Solar Tower Test of the...  

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

* National Solar Thermal Test Facility * NSTTF * Partnership * Renewable Energy * Solar Energy Comments are closed. Last Updated: September 29, 2014 Go To Top Exceptional...

131

Idaho National Laboratory Testing of Advanced Technology Vehicles...  

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

vss021francfort2012o.pdf More Documents & Publications Idaho National Laboratory Testing of Advanced Technology Vehicles Vehicle Technologies Office Merit Review 2014: Idaho...

132

Idaho National Laboratory Testing of Advanced Technology Vehicles...  

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

Publications Vehicle Technologies Office Merit Review 2014: Idaho National Laboratory Testing of Advanced Technology Vehicles AVTA HEV, NEV, BEV and HICEV Demonstrations and...

133

Limited Test Ban Treaty Signed | National Nuclear Security Administrat...  

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

Limited Test Ban Treaty Signed | National Nuclear Security Administration People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency...

134

Senate Rejects Test Ban Treaty | National Nuclear Security Administrat...  

National Nuclear Security Administration (NNSA)

Senate Rejects Test Ban Treaty | National Nuclear Security Administration People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency...

135

Reducing emissions to improve nuclear test detection | National...  

National Nuclear Security Administration (NNSA)

Reducing emissions to improve nuclear test detection | National Nuclear Security Administration People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear...

136

Clinton Extends Moratorium on Nuclear Weapons Testing | National...  

National Nuclear Security Administration (NNSA)

Clinton Extends Moratorium on Nuclear Weapons Testing | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile...

137

Los Alamos National Laboratory Omega West Reactor restart  

SciTech Connect

This report is a critical evaluation of the effort for the restart of the Omega West reactor. It is divided into the following areas: progress made; difficulties in restart effort; current needs; and suggested detailed steps for improvement. A brief discussion is given for each area of study.

NONE

1993-08-27T23:59:59.000Z

138

Research, Development, Test, and Evaluation | National Nuclear...  

National Nuclear Security Administration (NNSA)

the long term. Among our ongoing efforts are the following activities: Providing a thermonuclear ignition platform at the National Ignition Facility (NIF) to investigate physics...

139

EXPERIMENT OPERATIONS PLAN FOR A LOSS-OF-COOLANT ACCIDENT SIMULATION IN THE NATIONAL RESEARCH UNIVERSAL REACTOR  

SciTech Connect

Pressurized water reactor loss-of-coolant accident phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship between the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. This document contains both experiment proposal and assembly proposal information. The intent of this document is to supply information required by the Chalk River Nuclear Laboratories (CRNL), and to identify the planned procedures and data that will be used both to establish readiness to proceed from one test phase to the next and to operate the experiment. Operating control settings and limits are provided for both experimenter systems and CRNL systems. A hazards review summarizes safety issues that have been addressed during the development of the experiment plan.

Russcher, G. E.; Cannon, L. W.; Goodman, R. L.; Hesson, G. M.; King, L. L.; McDuffie, P. N.; Marshall, R. K.; Nealley, C.; Pilger, J. P.; Mohr, C. L.

1981-04-01T23:59:59.000Z

140

Plan for decommissioning the Tokamak Fusion Test Reactor  

SciTech Connect

The Tokamak Fusion Test Reactor (TFTR) Project is in the planning phase of developing a decommissioning project. A Preliminary Decontamination and Decommissioning (D D) Plan has been developed which provides a framework for the baseline approach, and the cost and schedule estimates. TFTR will become activated and contaminated with tritium after completion of the deuterium-tritium (D-T) experiments. Hence some of the D D operations will require remote handling. It is expected that all of the waste generated will be low level radioactive waste (LLW). The objective of the D D Project is to make TFTR Test Cell available for use by a new fusion experiment. This paper discusses the D D objectives, the facility to be decommissioned, estimates of activation, the technical (baseline) approach, and the assumptions used to develop cost and schedule estimates.

Spampinato, P.T.; Walton, G.R. (Princeton Univ., NJ (United States). Plasma Physics Lab.); Commander, J.C. (EG and G Idaho, Inc., Idaho Falls, ID (United States))

1993-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "national reactor testing" 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

Plan for decommissioning the Tokamak Fusion Test Reactor  

SciTech Connect

The Tokamak Fusion Test Reactor (TFTR) Project is in the planning phase of developing a decommissioning project. A Preliminary Decontamination and Decommissioning (D&D) Plan has been developed which provides a framework for the baseline approach, and the cost and schedule estimates. TFTR will become activated and contaminated with tritium after completion of the deuterium-tritium (D-T) experiments. Hence some of the D&D operations will require remote handling. It is expected that all of the waste generated will be low level radioactive waste (LLW). The objective of the D&D Project is to make TFTR Test Cell available for use by a new fusion experiment. This paper discusses the D&D objectives, the facility to be decommissioned, estimates of activation, the technical (baseline) approach, and the assumptions used to develop cost and schedule estimates.

Spampinato, P.T.; Walton, G.R. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Commander, J.C. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

1993-12-31T23:59:59.000Z

142

EIS-0291: High Flux Beam Reactor (HFBR) Transition Project at the Brookhaven National Laboratory, Upton, New York  

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

The EIS evaluates the range of reasonable alternatives and their impacts regarding the future management of the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL).

143

INDEPENDENT VERIFICATION SURVEY OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT OUTSIDE AREAS BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK  

SciTech Connect

5098-SR-03-0 FINAL REPORT- INDEPENDENT VERIFICATION SURVEY OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT OUTSIDE AREAS, BROOKHAVEN NATIONAL LABORATORY

P.C. Weaver

2010-12-15T23:59:59.000Z

144

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

SciTech Connect

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

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

2010-03-01T23:59:59.000Z

145

Sandia National Laboratories: Photovoltaic (PV) Regional Test...  

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

ClimateECEnergyPhotovoltaic (PV) Regional Test Center (RTC) Website Goes Live Photovoltaic (PV) Regional Test Center (RTC) Website Goes Live Hope Michelsen named to Alameda County...

146

Sandia National Laboratories: Pratt Whitney Rocketdyne Testing  

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

ClimateECEnergyRenewable EnergySolarConcentrating Solar PowerPratt Whitney Rocketdyne Testing Pratt Whitney Rocketdyne Testing Excellence Award in the 2012 Facilities...

147

Sandia National Laboratories: Vermont Photovoltaic Regional Test...  

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

Photovoltaic Regional Test Center Launch of Solar Testing Site in Vermont On November 27, 2013, in Energy, Facilities, News, News & Events, Partnership, Photovoltaic, Photovoltaic...

148

Brookhaven National Laboratory | Accelerator Test Facility  

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

physics, BNL will provide Program Development funding totaling 2M over the 3 years for upgrading the CO 2 laser to the level of 100 TW. Brookhaven National Laboratory |...

149

Recent results on the RIA test in IGR reactor  

SciTech Connect

At the 23d WRSM meeting the data base characterizing results of VVER high burnup fuel rods tests under reactivity-initiated accident (RIA) conditions was presented. Comparison of PWR and VVER failure thresholds was given also. Additional analysis of the obtained results was being carried out during 1996. The results of analysis show that the two different failure mechanisms were observed for PWR and VVER fuel rods. Some factors which can be as the possible reasons of these differences are presented. First of them is the state of preirradiated cladding. Published test data for PWR high burnup fuel rods demonstrated that the PWR high burnup fuel rods failed at the RIA test are characterized by very high level of oxidation and hydriding for the claddings. Corresponding researches were performed at Institute of Atomic Reactors (RLAR, Dimitrovgrad, Russia) for large set of VVER high burnup fuel rods. Results of these investigations show that preirradiated commercial Zr-1%Nb claddings practically keep their initial levels of oxidation and H{sub 2} concentration. Consequently the VVER preirradiated cladding must keep the high level of mechanical properties. The second reason leading to differences between failure mechanisms for two types of high burnup fuel rods can be the test conditions. Now such kind of analysis have been performed by two methods.

Asmolov, V.; Yegorova, L. [Nuclear Safety Institute, Moscow (Russian Federation)

1997-01-01T23:59:59.000Z

150

National Wind Tecnology Center Provides Dual Axis Resonant Blade Testing  

ScienceCinema (OSTI)

NREL's Structural Testing Laboratory at the National Wind Technology Center (NWTC) provides experimental laboratories, computer facilities for analytical work, space for assembling components and turbines for atmospheric testing as well as office space for industry researchers. Fort Felker, center director at the NWTC, discusses NREL's state-of-the-art structural testing capabilities and shows a flapwise and edgewise blade test in progress.

Felker, Fort

2014-06-10T23:59:59.000Z

151

AVTA: Idaho National Laboratory Experimental Hybrid Shuttle Bus testing results  

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

The Vehicle Technologies Office's Advanced Vehicle Testing Activity carries out testing on a wide range of advanced vehicles and technologies on dynamometers, closed test tracks, and on-the-road. These results provide benchmark data that researchers can use to develop technology models and guide future research and development. The following report describes testing results of the Idaho National Laboratory's demonstration hybrid shuttle bus.

152

High conduction neutron absorber to simulate fast reactor environment in an existing test reactor  

SciTech Connect

A new metal matrix composite material has been developed to serve as a thermal neutron absorber for testing fast reactor fuels and materials in an existing pressurized water reactor. The performance of this material was evaluated by placing neutron fluence monitors within shrouded and unshrouded holders and irradiating for up to four cycles. The monitor wires were analyzed by gamma and X-ray spectrometry to determine the activities of the activation products. Adjusted neutron fluences were calculated and grouped into three bins—thermal, epithermal, and fast—to evaluate the spectral shift created by the new material. A comparison of shrouded and unshrouded fluence monitors shows a thermal fluence decrease of ~11 % for the shielded monitors. Radioisotope activity and mass for each of the major activation products is given to provide insight into the evolution of thermal absorption cross-section during irradiation. The thermal neutron absorption capability of the composite material appears to diminish at total neutron fluence levels of ~8 × 1025 n/m2. Calculated values for dpa in excess of 2.0 were obtained for two common structural materials (iron and nickel) of interest for future fast flux experiments.

Donna Post Guillen; Larry R. Greenwood; James R. Parry

2014-10-01T23:59:59.000Z

153

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

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

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

154

E-Print Network 3.0 - advanced test reactor critical facility...  

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

Powered by Explorit Topic List Advanced Search Sample search results for: advanced test reactor critical facility Page: << < 1 2 3 4 5 > >> 1 Engineers at Western are...

155

Microsoft Word - 911136_0 SSC-4b Reactor Graphite Test Plan_rel...  

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

Services for the Next Generation Nuclear Plant (NGNP) with Hydrogen Production Test Plan for Reactor Graphite Elements Prepared by General Atomics for the Battelle Energy...

156

Office of Research, Development, Test, and Evaluation | National Nuclear  

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

Research, Development, Test, and Evaluation | National Nuclear Research, Development, Test, and Evaluation | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog The National Nuclear Security Administration Office of Research, Development, Test, and Evaluation Home > About Us > Our Programs > Defense Programs > Office of Research, Development, Test, and Evaluation

157

Recovery Act Funds Test Reactor Dome Removal in Historic D&D Project |  

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

Recovery Act Funds Test Reactor Dome Removal in Historic D&D Recovery Act Funds Test Reactor Dome Removal in Historic D&D Project Recovery Act Funds Test Reactor Dome Removal in Historic D&D Project February 1, 2011 - 12:00pm Addthis Media Contacts Jim Giusti, DOE (803) 952-7697 james-r.giusti@srs.gov Paivi Nettamo, SRNS (803) 646-6075 paivi.nettamo@srs.gov AIKEN, S.C. - The landscape of the Savannah River Site (SRS) is a little flatter and a little less colorful with the removal today of the 75-foot-tall rusty-orange dome from the Cold War-era test reactor. This $25-million reactor decommissioning and deactivation project is funded By the American Recovery and Reinvestment Act. Affectionately known by SRS employees as "Hector," the iconic Heavy Water Components Test Reactor (HWCTR) has stood in the Site's B Area since 1959

158

Testing of an advanced thermochemical conversion reactor system  

SciTech Connect

This report presents the results of work conducted by MTCI to verify and confirm experimentally the ability of the MTCI gasification process to effectively generate a high-quality, medium-Btu gas from a wider variety of feedstock and waste than that attainable in air-blown, direct gasification systems. The system's overall simplicity, due to the compact nature of the pulse combustor, and the high heat transfer rates attainable within the pulsating flow resonance tubes, provide a decided and near-term potential economic advantage for the MTCI indirect gasification system. The primary objective of this project was the design, construction, and testing of a Process Design Verification System for an indirectly heated, thermochemical fluid-bed reactor and a pulse combustor an an integrated system that can process alternative renewable sources of energy such as biomass, black liquor, municipal solid waste and waste hydrocarbons, including heavy oils into a useful product gas. The test objectives for the biomass portion of this program were to establish definitive performance data on biomass feedstocks covering a wide range of feedstock qualities and characteristics. The test objectives for the black liquor portion of this program were to verify the operation of the indirect gasifier on commercial black liquor containing 65 percent solids at several temperature levels and to characterize the bed carbon content, bed solids particle size and sulfur distribution as a function of gasification conditions. 6 refs., 59 figs., 29 tabs.

Not Available

1990-01-01T23:59:59.000Z

159

Research reactor usage at the Idaho National Engineering Laboratory in support of university research and education  

SciTech Connect

The Idaho National Engineering Laboratory is a US Department of Energy laboratory which has a substantial history of research and development in nuclear reactor technologies. There are a number of available nuclear reactor facilities which have been incorporated into the research and training needs of university nuclear engineering programs. This paper addresses the utilization of the Advanced Reactivity Measurement Facility (ARMF) and the Coupled Fast Reactivity Measurement Facility (CFRMF) for thesis and dissertation research in the PhD program in Nuclear Science and Engineering by the University of Idaho and Idaho State University. Other reactors at the INEL are also being used by various members of the academic community for thesis and dissertation research, as well as for research to advance the state of knowledge in innovative nuclear technologies, with the EBR-II facility playing an essential role in liquid metal breeder reactor research. 3 refs.

Woodall, D.M.; Dolan, T.J.; Stephens, A.G. (Idaho National Engineering Lab., Idaho Falls, ID (USA))

1990-01-01T23:59:59.000Z

160

Blade Testing Trends (Presentation), NREL (National Renewable...  

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

industry support. * Field Testing o Small to megawatt-scale turbines (more than 10 MW installed) o Demonstrates advances in control systems and innovative technologies. *...

Note: This page contains sample records for the topic "national reactor testing" 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

Sandia National Laboratories: Photovoltaic Regional Testing Center...  

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

Regional Testing Center (PV RTC) Sandia Participation in the 39th IEEE Photovoltaic Specialists (PVSC) Conference On August 14, 2013, in DETL, Distribution Grid Integration,...

162

Sandia National Laboratories: Photovoltaic Regional Testing Center...  

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

Photovoltaic Regional Test Center (RTC). The RTC will enable research on integrating solar panels into the statewide smart grid and help reduce the cost of solar power. The...

163

CRAD, Management- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR  

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

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Management portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor.

164

CRAD, Environmental Protection- Oak Ridge National Laboratory High Flux Isotope Reactor  

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

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Environmental Compliance Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

165

CRAD, Configuration Management- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR  

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

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Configuration Management Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory, High Flux Isotope Reactor.

166

CRAD, Occupational Safety & Health- Oak Ridge National Laboratory High Flux Isotope Reactor  

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

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Industrial Safety and Hygiene Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

167

CRAD, Configuration Management- Oak Ridge National Laboratory High Flux Isotope Reactor  

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

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Configuration Management Program in preparation for restart of the Oak Ridge National Laboratory, High Flux Isotope Reactor.

168

CRAD, Emergency Management- Oak Ridge National Laboratory High Flux Isotope Reactor  

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

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Emergency Management Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

169

CRAD, Conduct of Operations- Oak Ridge National Laboratory High Flux Isotope Reactor  

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

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February, 2007 assessment of the Conduct of Operations Program in preparation for restart of the Oak Ridge National Laboratory, High Flux Isotope Reactor.

170

CRAD, Conduct of Operations- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR  

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

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February, 2007 assessment of the Conduct of Operations Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory, High Flux Isotope Reactor.

171

CRAD, Training- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR  

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

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Training Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor.

172

CRAD, Radiological Controls- Oak Ridge National Laboratory High Flux Isotope Reactor  

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

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Radiation Protection Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

173

CRAD, Safety Basis- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR  

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

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Safety Basis portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor.

174

CRAD, Safety Basis- Oak Ridge National Laboratory High Flux Isotope Reactor  

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

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Safety Basis in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

175

CRAD, Maintenance- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR  

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

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Maintenance Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor.

176

CRAD, Occupational Safety & Health- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR  

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

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Occupational Safety and Health Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor.

177

CRAD, Nuclear Safety- Oak Ridge National Laboratory High Flux Isotope Reactor  

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

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Nuclear Safety Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

178

CRAD, Quality Assurance- Oak Ridge National Laboratory High Flux Isotope Reactor  

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

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Quality Assurance Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

179

Dynamic Impregnator Reactor System (Poster), NREL (National Renewable Energy Laboratory)  

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

Several unit operations are combined into Several unit operations are combined into one robust system, off ering fl exible and staged process confi gurations in one vessel. Spraying, soaking, low-severity pretreat- ment, enzymatic hydrolysis, fermentation, concentration/evaporation, and distillation are amongst its many capabilities. * 1,900 L Horizontal Paddle Blender Vessel with Sidewall Liquid Drains * 6-60 rpm / 50 HP Tri-Directional Agitator * 3.4 bar & Vacuum ASME Design, 316L Stainless Steel * Heating/Cooling Jacket using Water or Steam * 150 L Chemical Mix Tank & Pump with Spray Injectors * Vent Condenser with Collection Tank and Vacuum Pump Dynamic Impregnator Reactor System Multifaceted system designed for complex feedstock impregnation and processing Integrated Biorefi nery Research Facility | NREL * Golden, Colorado | December 15, 2011 | NREL/PO-5100-56156

180

First Plutonium Bomb Successfully Tested | National Nuclear Security  

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

Plutonium Bomb Successfully Tested | National Nuclear Security Plutonium Bomb Successfully Tested | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > About Us > Our History > NNSA Timeline > First Plutonium Bomb Successfully Tested First Plutonium Bomb Successfully Tested July 16, 1945 Los Alamos, NM First Plutonium Bomb Successfully Tested

Note: This page contains sample records for the topic "national reactor testing" 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

Eisenhower Halts Nuclear Weapons Testing | National Nuclear Security  

National Nuclear Security Administration (NNSA)

Eisenhower Halts Nuclear Weapons Testing | National Nuclear Security Eisenhower Halts Nuclear Weapons Testing | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > About Us > Our History > NNSA Timeline > Eisenhower Halts Nuclear Weapons Testing Eisenhower Halts Nuclear Weapons Testing August 22, 1958 Washington, DC Eisenhower Halts Nuclear Weapons Testing

182

First Plutonium Bomb Successfully Tested | National Nuclear Security  

National Nuclear Security Administration (NNSA)

Plutonium Bomb Successfully Tested | National Nuclear Security Plutonium Bomb Successfully Tested | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > About Us > Our History > NNSA Timeline > First Plutonium Bomb Successfully Tested First Plutonium Bomb Successfully Tested July 16, 1945 Los Alamos, NM First Plutonium Bomb Successfully Tested

183

First Thermonuclear Device Successfully Tested | National Nuclear Security  

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

Thermonuclear Device Successfully Tested | National Nuclear Security Thermonuclear Device Successfully Tested | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > About Us > Our History > NNSA Timeline > First Thermonuclear Device Successfully Tested First Thermonuclear Device Successfully Tested December 31, 1952 Enewetak Atoll First Thermonuclear Device Successfully Tested

184

Advanced Test Reactor (ATR) Facility 10CFR830 Safety Basis Related to Facility Experiments  

SciTech Connect

The Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Test Reactor (ATR), a DOE Category A reactor, was designed to provide an irradiation test environment for conducting a variety of experiments. The ATR Safety Analysis Report, determined by DOE to meet the requirements of 10 CFR 830, Subpart B, provides versatility in types of experiments that may be conducted. This paper addresses two general types of experiments in the ATR facility and how 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 with more detail in the safety basis. This allows individual safety analyses for these experiments to be more routine and repetitive. The second type of experiment is less defined and is permitted under more general controls. Therefore, individual safety analyses for the second type of experiment tend to be more unique from experiment to experiment. Experiments are also discussed relative to "major modifications" and DOE-STD-1027-92. Application of the USQ process to ATR experiments is also discussed.

Tomberlin, Terry Alan

2002-06-01T23:59:59.000Z

185

Advanced Test Reactor (ATR) Facility 10CFR830 Safety Basis Related to Facility Experiments  

SciTech Connect

The Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Test Reactor (ATR), a DOE Category A reactor, was designed to provide an irradiation test environment for conducting a variety of experiments. The ATR Safety Analysis Report, determined by DOE to meet the requirements of 10 CFR 830, Subpart B, provides versatility in types of experiments that may be conducted. This paper addresses two general types of experiments in the ATR facility and how 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 with more detail in the safety basis. This allows individual safety analyses for these experiments to be more routine and repetitive. The second type of experiment is less defined and is permitted under more general controls. Therefore, individual safety analyses for the second type of experiment tend to be more unique from experiment to experiment. Experiments are also discussed relative to ''major modifications'' and DOE-STD-1027-92. Application of the USQ process to ATR experiments is also discussed.

Tomberlin, T.A.

2002-06-19T23:59:59.000Z

186

NREL: Wind Research - National Wind Technology Center Blade Testing Video  

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

Center Blade Testing Video (Text Version) Center Blade Testing Video (Text Version) Below is the text version for the National Wind Technology Center Blade Testing Video. The video opens with the NREL and NWTC logos, surrounded by black screen and including the title: "NWTC Test Facility Introduction, Dr. Fort Felker, Director of the National Wind Technology Center, TRT 1:42, May 29, 2013." Fort Felker is in a yellow helmet and vest, standing in the NWTC's testing facility. There is a railing to his left, construction cones behind him, and a ladder to his right. Fort Felker: "I'm Fort Felker, I'm the director at the Department of Energy's National Wind Technology Center." Fort's name and title cut in on the right. Fort walks toward the camera while talking. Fort Felker: "Here at the NWTC, we have been conducting structural testing

187

National Aeronautics and Space Administration Analog Missions and Field Tests  

E-Print Network (OSTI)

the habitat while wearing spacesuits. Testing in the antarctic simulated what it would be like for astronauts. The Florida Keys National Marine Sanctuary, home of the National Oceanic and Atmospheric Administration Marine Sanctuary, about 62 feet beneath the surface. A surface buoy provides connections for power, life

188

Inverter Testing at Sandia National Laboratories* Jerry W. Ginn  

Office of Scientific and Technical Information (OSTI)

Inverter Testing at Sandia National Inverter Testing at Sandia National Laboratories* Jerry W. Ginn Russell H. Bonn Photovoltaic System Components Department Sandia National Laboratories PO Box 5800 Albuquerque, NM 87185-0752 Abstract. Inverters are key building blocks of photovoltaic (PV) systems that produce ac power. The balance of systems @OS) portion of a PV system can account for up to 50% of the system cost, and its reliable operation is essential for a successful PV system. As part of its BOS program, Sandia National Laboratories (SNL) maintains a laboratory wherein accurate electrical measurements of power systems can be made under a variety of conditions. This paper outlines the work that is done in that laboratory. TESTING ACTIVITIES Inverter testing at SNL thus far

189

South Carolina Opens Nation's Largest Wind Drivetrain Testing Facility  

Office of Energy Efficiency and Renewable Energy (EERE)

Today, U.S. Deputy Secretary of Energy Daniel Poneman joined with officials from Clemson University to dedicate the nation's largest and one of the world's most advanced wind energy testing facilities in North Charleston, S.C.

190

The integral fast reactor fuels reprocessing laboratory at Argonne National Laboratory, Illinois  

SciTech Connect

The processing of Integral Fast Reactor (IFR) metal fuel utilizes pyrochemical fuel reprocessing steps. These steps include separation of the fission products from uranium and plutonium by electrorefining in a fused salt, subsequent concentration of uranium and plutonium for reuse, removal, concentration, and packaging of the waste material. Approximately two years ago a facility became operational at Argonne National Laboratory-Illinois to establish the chemical feasibility of proposed reprocessing and consolidation processes. Sensitivity of the pyroprocessing melts to air oxidation necessitated operation in atmosphere-controlled enclosures. The Integral Fast Reactor Fuels Reprocessing Laboratory is described.

Wolson, R.D.; Tomczuk, Z.; Fischer, D.F.; Slawecki, M.A.; Miller, W.E.

1986-09-01T23:59:59.000Z

191

National SCADA Test Bed Fact Sheet  

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

PROTECTING ENERGY INFRASTRUCTURE BY IMPROVING THE SECURITY OF CONTROL SYSTEMS PROTECTING ENERGY INFRASTRUCTURE BY IMPROVING THE SECURITY OF CONTROL SYSTEMS Improving the security of energy control systems has become a national priority. Since the mid-1990's, security experts have become increasingly concerned about the threat of malicious cyber attacks on the vital supervisory control and data acquisition (SCADA) and distributed control systems (DCS) used to monitor and manage our energy infrastructure. Many of the systems still in use today were designed to operate in closed, proprietary networks. Increasing use of common software and operating systems and connection to public telecommunication networks and the Internet have made systems more reliable and efficient-but also more

192

HISTORICAL AMERICAN ENGINEERING RECORD - IDAHO NATIONAL ENGINEERING AND ENVIRONMENTAL LABORATORY, TEST AREA NORTH, HAER NO. ID-33-E  

SciTech Connect

Test Area North (TAN) was a site of the Aircraft Nuclear Propulsion (ANP) Project of the U.S. Air Force and the Atomic Energy Commission. Its Cold War mission was to develop a turbojet bomber propelled by nuclear power. The project was part of an arms race. Test activities took place in five areas at TAN. The Assembly & Maintenance area was a shop and hot cell complex. Nuclear tests ran at the Initial Engine Test area. Low-power test reactors operated at a third cluster. The fourth area was for Administration. A Flight Engine Test facility (hangar) was built to house the anticipated nuclear-powered aircraft. Experiments between 1955-1961 proved that a nuclear reactor could power a jet engine, but President John F. Kennedy canceled the project in March 1961. ANP facilities were adapted for new reactor projects, the most important of which were Loss of Fluid Tests (LOFT), part of an international safety program for commercial power reactors. Other projects included NASA's Systems for Nuclear Auxiliary Power and storage of Three Mile Island meltdown debris. National missions for TAN in reactor research and safety research have expired; demolition of historic TAN buildings is underway.

Susan Stacy; Hollie K. Gilbert

2005-02-01T23:59:59.000Z

193

National Carbon Capture Center Launches Post-Combustion Test Center |  

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

National Carbon Capture Center Launches Post-Combustion Test Center National Carbon Capture Center Launches Post-Combustion Test Center National Carbon Capture Center Launches Post-Combustion Test Center June 7, 2011 - 1:00pm Addthis Washington, D.C. - The recent successful commissioning of an Alabama-based test facility is another step forward in research that will speed deployment of innovative post-combustion carbon dioxide (CO2) capture technologies for coal-based power plants, according to the U.S. Department of Energy (DOE). Technologies tested at the Post-Combustion Carbon Capture Center (or PC4) are an important component of Carbon Capture and Storage, whose commercial deployment is considered by many experts as essential for helping to reduce human-generated CO2 emissions that contribute to potential climate change.

194

South Carolina Opens Nation’s Largest Wind Drivetrain Testing Facility  

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

Today, U.S. Deputy Secretary of Energy Daniel Poneman joined with officials from Clemson University to dedicate the nation's largest and one of the world's most advanced wind energy testing facilities in North Charleston, S.C.

195

Review of the Oak Ridge National Laboratory High Flux Isotope Reactor Implementation Verification Review Processes  

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

Independent Oversight Review of the Independent Oversight Review of the Oak Ridge National Laboratory High Flux Isotope Reactor Implementation Verification Review Processes May 2011 January 2013 Office of Safety and Emergency Management Evaluations Office of Enforcement and Oversight Office of Health, Safety and Security U. S. Department of Energy Table of Contents 1.0 Purpose ................................................................................................................................................. 1 2.0 Background........................................................................................................................................... 1 3.0 Scope..................................................................................................................................................... 2

196

Review of the Oak Ridge National Laboratory High Flux Isotope Reactor Implementation Verification Review Processes  

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

Independent Oversight Review of the Independent Oversight Review of the Oak Ridge National Laboratory High Flux Isotope Reactor Implementation Verification Review Processes May 2011 January 2013 Office of Safety and Emergency Management Evaluations Office of Enforcement and Oversight Office of Health, Safety and Security U. S. Department of Energy Table of Contents 1.0 Purpose ................................................................................................................................................. 1 2.0 Background........................................................................................................................................... 1 3.0 Scope..................................................................................................................................................... 2

197

Nuclear Weapons Testing Resumes | National Nuclear Security Administration  

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

Testing Resumes | National Nuclear Security Administration Testing Resumes | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > About Us > Our History > NNSA Timeline > Nuclear Weapons Testing Resumes Nuclear Weapons Testing Resumes September 01, 1961 Washington, DC Nuclear Weapons Testing Resumes The Soviet Union breaks the nuclear test moratorium and the United States

198

DOE - Office of Legacy Management -- Idaho National Engineering...  

Office of Legacy Management (LM)

that supports the Departments missions of environmental quality, energy resources, science, and national security. Originally named the National Reactor Testing Station, the...

199

Testing, Training, and Signature Devices | Y-12 National Security Complex  

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

Testing, Testing, Training, and ... Testing, Training, and Signature Devices Y-12 manufactures specialized uranium testing, training, and signature devices to support the nuclear detection community. As part of our national security mission, and in partnership with Oak Ridge National Laboratory, we are producing unique test objects for passive gamma ray signature analysis. Y-12 is fabricating new Highly Enriched Uranium Equivalent Radiological Signature Training Devices, tools that use an innovative method to replicate a much larger mass of uranium. These objects contain small amounts of U-235 embedded in an aluminum alloy. When seen by a detector, however, the gamma ray signature is nearly equivalent to a much larger amount of U-235, due to the alloying effect that minimizes the uranium

200

High heat flux testing capabilities at Sandia National Laboratories - New Mexico  

SciTech Connect

High heat flux testing for the United States fusion power program is the primary mission of the Plasma Materials Test Facility (PMTF) located at Sandia National Laboratories - New Mexico. This facility, which is owned by the United States Department of Energy, has been in operation for over 17 years and has provided much of the high heat flux data used in the design and evaluation of plasma facing components for many of the world`s magnetic fusion, tokamak experiments. In addition to domestic tokamaks such as Tokamak Fusion Test Reactor (TFTR) at Princeton and the DIII-D tokamak at General Atomics, components for international experiments like TEXTOR, Tore-Supra, and JET also have been tested at the PMTF. High heat flux testing spans a wide spectrum including thermal shock tests on passively cooled materials, thermal response and thermal fatigue tests on actively cooled components, critical heat flux-burnout tests, braze reliability tests and safety related tests. The objective of this article is to provide a brief overview of the high heat flux testing capabilities at the PMTF and describe a few of the experiments performed over the last year.

Youchison, D.L.; McDonald, J.M.; Wold, L.S.

1994-12-31T23:59:59.000Z

Note: This page contains sample records for the topic "national reactor testing" 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

Environmental Assessment for Decontamination and Decommissioning of the Juggernaut Reactor at Argonne National Laboratory Â… East Argonne, Illinois  

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

DOE/EA-1483 DOE/EA-1483 Environmental Assessment for Decontamination and Decommissioning of the Juggernaut Reactor at Argonne National Laboratory - East Argonne, Illinois March 2004 U.S. Department of Energy Chicago Operations Office Argonne Area Office Argonne, Illinois Environmental Assessment for Decontamination and Decommissioning of the Juggernaut Reactor at Argonne National Laboratory - East Argonne, Illinois Table of Contents Acronyms....................................................................................................................................... iii 1.0 Background ..........................................................................................................................1 1.1 Facility History ........................................................................................................1

202

Materials Transportation Testing & Analysis at Sandia National Laboratories  

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

Testing Testing Carlos Lopez, (505) 845-9545 Packages transporting the larger "Type B" quantities of radioactive materials must be qualified and certified under Title 10, Code of Federal Regulations, Part 71, or under the equivalent international standard ST-1 issued by the International Atomic Energy Agency. The principal thermal qualification test is the 30 minute pool fire. As part of the National Transportation Program, the Transportation Risk & Packaging Program at Sandia can plan and conduct these tests for DOE and other package suppliers. Test Plans, QA plans and other necessary test documents can be prepared for customer and regulatory approval. Tests may be conducted with a variety of available facilities at Sandia, including large pools, an indoor fire facility, and a radiant heat test

203

E-Print Network 3.0 - advanced test idaho reactor Sample Search...  

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

Powered by Explorit Topic List Advanced Search Sample search results for: advanced test idaho reactor Page: << < 1 2 3 4 5 > >> 1 Point of Contact: Doug Kothe CASL Director...

204

E-Print Network 3.0 - advanced test reactor Sample Search Results  

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

20 03012006 09:51 AMLoading "People's Daily Online --Chinese experimental thermonuclear reactor on discharge test in July" Page 1 of 1http:english.people.com.cn200603...

205

Preliminary Cost Assessment and Compare of China Fusion Engineering Test Reactor  

Science Journals Connector (OSTI)

Full superconducting tokamak and water-cooling Cu magnets tokamak are two options proposed for China Fusion Engineering Test Reactor (CFETR). Based on the concept design ... Program for Parameters Optimization an...

Dehong Chen; Jieqiong Jiang; Yawei Hou; Wenxue Duan; Muyi Ni…

2014-09-01T23:59:59.000Z

206

Materials Transportation Testing & Analysis at Sandia National Laboratories  

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

Unique Solutions] Unique Solutions] [Working With Us] [Contacting Us] [News Center] [Search] [Home] [navigation panel] Materials Transportation Testing & Analysis Our Mission Our Contacts Write to Us Package Development Risk Assessment RADTRAN GIS Mapping Structural Analysis Thermal Analysis Structural Testing Thermal Testing MIDAS Data Aquisition System Concepts Materials Characterization Regulatory Development Certification Support RMIR Data Base Scientific Visualization Mobile Instrumentation Data Acquisition System (MIDAS) Doug Ammerman, (505) 845-8158 The Mobile Instrumentation Data Acquisition System (MIDAS), developed by Sandia National Laboratories for the U.S. Department of Energy, provides on-site data acquisition of containers that transport radioactive materials during impact, puncture, fire, and immersion tests.

207

Preliminary requirements for a Fluoride Salt-Cooled High-Temperature Test Reactor (FHTR)  

SciTech Connect

A Fluoride Salt-Cooled High-Temperature Test Reactor (FHTR) design is being developed at MIT to provide the first demonstration and test of a salt-cooled reactor using high-temperature fuel. The first step is to define the requirements. The top level requirements are (1) provide the confidence that a larger demonstration reactor is warranted and (2) develop the necessary data for a larger-scale reactor. Because requirements will drive the design of the FHTR, a significant effort is being undertaken to define requirements and understand the tradeoffs that will be required for a practical design. The preliminary requirements include specifications for design parameters and necessary tests of major reactor systems. Testing requirements include demonstration of components, systems, and procedures for refueling, instrumentation, salt temperature control to avoid coolant freezing, salt chemistry and volume control, tritium monitoring and control, and in-service inspection. Safety tests include thermal hydraulics, neutronics - including intrinsic core shutdown mechanisms such as Doppler feedback - and decay heat removal systems. Materials and coolant testing includes fuels (including mechanical wear and fatigue) and system corrosion behavior. Preliminary analysis indicates a thermal power output below 30 MW, an initial core using pebble-bed or prismatic-block fuel, peak outlet temperatures of at least 700 deg. C, and use of FLi{sup 7}Be ({sup 7}LiF-BeF{sub 2}) coolant. The option to change-out the reactor core, fuel type, and major components is being investigated. While the FHTR will be used for materials testing, its primary mission is as a reactor system performance test to enable the design and licensing of a FHR demonstration power reactor. (authors)

Massie, M.; Forsberg, C.; Forget, B. [Dept. of Nuclear Science and Engineering, Massachusetts Inst. of Technology, Cambridge, MA 02139 (United States); Hu, L. W. [Nuclear Reactor Laboratory, Massachusetts Inst. of Technology, 77 Massachusetts Avenue, Cambridge, MA 02139 (United States)

2012-07-01T23:59:59.000Z

208

Clinton Extends Moratorium on Nuclear Weapons Testing | National Nuclear  

National Nuclear Security Administration (NNSA)

Clinton Extends Moratorium on Nuclear Weapons Testing | National Nuclear Clinton Extends Moratorium on Nuclear Weapons Testing | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > About Us > Our History > NNSA Timeline > Clinton Extends Moratorium on Nuclear Weapons Testing Clinton Extends Moratorium on Nuclear Weapons Testing July 03, 1993 Washington, DC

209

Reducing emissions to improve nuclear test detection | National Nuclear  

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

Reducing emissions to improve nuclear test detection | National Nuclear Reducing emissions to improve nuclear test detection | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > NNSA Blog > Reducing emissions to improve nuclear test detection Reducing emissions to improve nuclear test detection Posted By Office of Public Affairs In early November, medical isotope producers met with nuclear explosion

210

National Carbon Capture Center Launches Post-Combustion Test Center |  

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

Carbon Capture Center Launches Post-Combustion Test Center Carbon Capture Center Launches Post-Combustion Test Center National Carbon Capture Center Launches Post-Combustion Test Center June 6, 2011 - 2:32pm Addthis Jenny Hakun What does this mean for me? Commercial deployment of the processes tested here could cut carbon pollution. Innovation is important to finding ways to make energy cleaner. And testing the ideas and processes that researchers come up with is critical to moving ideas from the lab to the marketplace. That's why the Department of Energy recently commissioned an Alabama testing facility that will help move research forward and speed up deployment of innovative post-combustion carbon dioxide (CO2) capture technologies for coal-based power plants. The Post-Combustion Carbon Capture Center (or PC4) facility tests new

211

Results of Sandia National Laboratories grid-tied inverter testing  

SciTech Connect

This paper proposes a definition for a Non-Islanding Inverter. This paper also presents methods that can be used to implement such an inverter, along with references to prior work on the subject. Justification for the definition is provided on both a theoretical basis and results from tests conducted at Sandia National Laboratories and Ascension Technology, Inc.

Kern, G.A. [Ascension Technology, Inc., Boulder, CO (United States); Bonn, R.H.; Ginn, J.; Gonzalez, S. [Sandia National Labs., Albuquerque, NM (United States)

1998-07-01T23:59:59.000Z

212

Prototype dish testing and analysis at Sandia National Laboratories  

SciTech Connect

During the past year, Sandia National Laboratories performed on-sun testing of several dish concentrator concepts. These tests were undertaken at the National Solar Thermal Test Facility (NSTTF). Two of the tests were performed in support of the DOE Concentrator Receiver Development Program. The first was on-sun testing of the single-element stretched-membrane dish; this 7-meter diameter dish uses a single preformed metal membrane with an aluminized polyester optical surface and shows potential for future dish-Stirling systems. The next involved two prototype facets from the Faceted Stretched-Membrane Dish Program. These facets, representing competitive design concepts, are closest to commercialization. Five 1-meter triangular facets were tested on-sun as part of the development program for a solar dynamic system on Space Station Freedom. While unique in character, all the tests utilized the Beam Characterization System (BCS) as the main measurement tool and all were analyzed using the Sandia-developed CIRCE2 computer code. The BCS is used to capture and digitize an image of the reflected concentrator beam that is incident on a target surface. The CIRCE2 program provides a computational tool, which when given the geometry of the concentrator and target as well as other design parameters will predict the flux distribution of the reflected beam. One of these parameters, slope error, is the variable that has a major effect in determining the quality of the reflected beam. The methodology used to combine these two tools to predict uniform slope errors for the dishes is discussed in this document. As the Concentrator Development Programs continue, Sandia will test and evaluate two prototype dish systems. The first, the faceted stretched-membrane dish, is expected to be tested in 1992, followed by the full-scale single-element stretched-membrane dish in 1993. These tests will use the tools and methodology discussed in this document. 14 refs., 10 figs., 5 tabs.

Grossman, J.W.; Houser, R.M.; Erdman, W.W.

1991-01-01T23:59:59.000Z

213

Materials Transportation Testing & Analysis at Sandia National Laboratories  

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

Materials Characterization Materials Characterization Paul McConnell, (505) 844-8361 The purpose of hazardous and radioactive materials, i.e., mixed waste, packaging is to enable this waste type to be transported without posing a threat to the health or property of the general public. To achieve this goal, regulations have been written establishing general design requirement for such packagings. Based on these regulatory requirements, a Mixed Waste Chemical Compatibility Testing Program is intended to assure regulatory bodies that the issue of packaging compatibility towards hazardous and radioactive materials has been addressed. Such a testing program has been developed in the Transportation Systems Department at Sandia National Laboratories. Materials Characterization Capabilities

214

PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 5, BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK  

SciTech Connect

5098-SR-04-0 PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 5, BROOKHAVEN NATIONAL LABORATORY

P.C. Weaver

2010-11-03T23:59:59.000Z

215

PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 1, BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK  

SciTech Connect

5098-SR-05-0 PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 1 BROOKHAVEN NATIONAL LABORATORY

E.M. Harpenau

2010-12-15T23:59:59.000Z

216

Light Water Reactor Fuel Cladding Research and Testing | ornl.gov  

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

Light Water Reactor Fuel Cladding Research Light Water Reactor Fuel Cladding Research June 01, 2013 Severe Accident Test Station ORNL is the focus point for Light Water Reactor (LWR) fuel cladding research and testing. The purpose of this research is to furnish U.S. industry (EPRI, Areva, Westinghouse), and regulators (NRC) with much-needed data supporting safe and economical nuclear power generation and used fuel management. LWR fuel cladding work is tightly integrated with ORNL accident tolerant fuel development and used fuel disposition programs thereby providing a powerful capability that couples basic materials science research with the nuclear applications research and development. The ORNL LWR fuel cladding program consists of five complementary areas of research: Accident tolerant fuel and cladding material testing under design

217

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

SciTech Connect

This document contains the analysis details and summary of analyses conducted to evaluate the environmental impacts for the Resumption of Transient Fuel and Materials Testing Program. It provides an assessment of the impacts for the two action alternatives being evaluated in the environmental assessment. These alternatives are (1) resumption of transient testing using the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) and (2) conducting transient testing using the Annular Core Research Reactor (ACRR) at Sandia National Laboratory in New Mexico (SNL/NM). Analyses are provided for radiologic emissions, other air emissions, soil contamination, and groundwater contamination that could occur (1) during normal operations, (2) as a result of accidents in one of the facilities, and (3) during transport. It does not include an assessment of the biotic, cultural resources, waste generation, or other impacts that could result from the resumption of transient testing. Analyses were conducted by technical professionals at INL and SNL/NM as noted throughout this report. The analyses are based on bounding radionuclide inventories, with the same inventories used for test materials by both alternatives and different inventories for the TREAT Reactor and ACRR. An upper value on the number of tests was assumed, with a test frequency determined by the realistic turn-around times required between experiments. The estimates provided for impacts during normal operations are based on historical emission rates and projected usage rates; therefore, they are bounding. Estimated doses for members of the public, collocated workers, and facility workers that could be incurred as a result of an accident are very conservative. They do not credit safety systems or administrative procedures (such as evacuation plans or use of personal protective equipment) that could be used to limit worker doses. Doses estimated for transportation are conservative and are based on transport of the bounding radiologic inventory that will be contained in any given test. The transportation analysis assumes all transports will contain the bounding inventory.

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

2014-02-01T23:59:59.000Z

218

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

219

Initial Confinement Studies of Ohmically Heated Plasmas in the Tokamak Fusion Test Reactor  

Science Journals Connector (OSTI)

Initial operation of the tokamak fusion test reactor has concentrated upon confinement studies of Ohmically heated hydrogen and deuterium plasmas. Total energy confinement times (?E) are 0.1-0.2 s for a line-average density range (n¯e) of (1-2.5)×1019 m-3 with electron temperatures of Te(0)?1.2-2.2 keV, ion temperatures of Ti(0)?0.9-1.5 keV, and Zeff?3. A comparison of Princeton large torus, poloidal divertor experiment, and tokamak fusion test reactor plasma confinement supports a dimension-cubed scaling law.

P. C. Efthimion et al.

1984-04-23T23:59:59.000Z

220

TESTING OF THE RADBALL TECHNOLOGY AT SAVANNAH RIVER NATIONAL LABORATORY  

SciTech Connect

The United Kingdom's National Nuclear Laboratory (NNL) has developed a remote, nonelectrical, radiation-mapping device known as RadBall (patent pending), which offers a means to locate and quantify radiation hazards and sources within contaminated areas of the nuclear industry. Positive results from initial deployment trials in nuclear waste reprocessing plants at Sellafield in the United Kingdom and the anticipated future potential use of RadBall throughout the U.S. Department of Energy Complex have led to the NNL partnering with the Savannah River National Laboratory (SRNL) to further test, underpin, and strengthen the technical performance of the technology. The study completed at SRNL addresses key aspects of the testing of the RadBall technology. The first set of tests was performed at Savannah River Nuclear Solutions Health Physics Instrument Calibration Laboratory (HPICL) using various gamma-ray sources and an x-ray machine with known radiological characteristics. The objective of these preliminary tests was to identify the optimal dose and collimator thickness. The second set of tests involved a highly contaminated hot cell. The objective of this testing was to characterize a hot cell with unknown radiation sources. The RadBall calibration experiments and hot cell deployment were successful in that for each trial radiation tracks were visible. The deployment of RadBall can be accomplished in different ways depending on the size and characteristics of the contaminated area (e.g., a hot cell that already has a crane/manipulator available or highly contaminated room that requires the use of a remote control device with sensor and video equipment to position RadBall). This report also presents SRNL-designed RadBall accessories for future RadBall deployment (a harness, PODS, and robot).

Farfan, E.; Foley, T.

2010-02-10T23:59:59.000Z

Note: This page contains sample records for the topic "national reactor testing" 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

Hot-Gas Filter Testing with a Transport Reactor Gasifier  

SciTech Connect

Today, coal supplies over 55% of the electricity consumed in the United States and will continue to do so well into the next century. One of the technologies being developed for advanced electric power generation is an integrated gasification combined cycle (IGCC) system that converts coal to a combustible gas, cleans the gas of pollutants, and combusts the gas in a gas turbine to generate electricity. The hot exhaust from the gas turbine is used to produce steam to generate more electricity from a steam turbine cycle. The utilization of advanced hot-gas particulate and sulfur control technologies together with the combined power generation cycles make IGCC one of the cleanest and most efficient ways available to generate electric power from coal. One of the strategic objectives for U.S. Department of Energy (DOE) IGCC research and development program is to develop and demonstrate advanced gasifiers and second-generation IGCC systems. Another objective is to develop advanced hot-gas cleanup and trace contaminant control technologies. One of the more recent gasification concepts to be investigated is that of the transport reactor gasifier, which functions as a circulating fluid-bed gasifier while operating in the pneumatic transport regime of solid particle flow. This gasifier concept provides excellent solid-gas contacting of relatively small particles to promote high gasification rates and also provides the highest coal throughput per unit cross-sectional area of any other gasifier, thereby reducing capital cost of the gasification island.

Swanson, M.L.; Hajicek, D.R.

2002-09-18T23:59:59.000Z

222

Continuous-flow stirred-tank reactor 20-L demonstration test: Final report  

SciTech Connect

One of the proposed methods of removing the cesium, strontium, and transuranics from the radioactive waste storage tanks at Savannah River is the small-tank tetraphenylborate (TPB) precipitation process. A two-reactor-in-series (15-L working volume each) continuous-flow stirred-tank reactor (CSTR) system was designed, constructed, and installed in a hot cell to test the Savannah River process. The system also includes two cross-flow filtration systems to concentrate and wash the slurry produced in the process, which contains the bulk of radioactivity from the supernatant processed through the system. Installation, operational readiness reviews, and system preparation and testing were completed. The first test using the filtration systems, two CSTRs, and the slurry concentration system was conducted over a 61-h period with design removal of Cs, Sr, and U achieved. With the successful completion of Test 1a, the following tests, 1b and 1c, were not required.

Lee, D.D.; Collins, J.L.

2000-02-01T23:59:59.000Z

223

Review of the International Atomic Energy Agency International database on reactor pressure vessel materials and US Nuclear Regulatory Commission/Oak Ridge National Laboratory embrittlement data base  

SciTech Connect

The International Atomic Energy Agency (IAEA) has supported neutron radiation effects information exchange through meetings and conferences since the mid-1960s. Through an International Working Group on Reliability of Reactor Pressure Components, information exchange and research activities were fostered through the Coordinated Research Program (CRP) sponsored by the IAEA. The final CRP meeting was held in November 1993, where it was recommended that the IAEA coordinate the development of an International Database on Reactor Pressure Vessel Material (IDRPVM) as the first step in generating an International Database on Aging Management. The purpose of this study was to provide special technical assistance to the NRC in monitoring and evaluating the IAEA activities in developing the IAEA IDRPVM, and to compare the IDRPVM with the Nuclear Regulatory Commission (NRC) - Oak Ridge National Laboratory (ORNL) Power Reactor Embrittlement Data Base (PR-EDB) and provide recommendations for improving the PR-EDB. A first test version of the IDRPVM was distributed at the First Meeting of Liaison Officers to the IAEA IDRPVM, in November 1996. No power reactor surveillance data were included in this version; the testing data were mainly from CRP Phase III data. Therefore, because of insufficient data and a lack of power reactor surveillance data received from the IAEA IDRPVM, the comparison is made based only on the structure of the IDRPVM. In general, the IDRPVM and the EDB have very similar data structure and data format. One anticipates that because the IDRPVM data will be collected from so many different sources, quality assurance of the data will be a difficult task. The consistency of experimental test results will be an important issue. A very wide spectrum of material characteristics of RPV steels and irradiation environments exists among the various countries. Hence the development of embrittlement prediction models will be a formidable task. 4 refs., 2 figs., 4 tabs.

Wang, J.A.; Kam, F.B.K.

1998-02-01T23:59:59.000Z

224

DOE/OE National SCADA Test Bed Fiscal Year 2009 Work Plan | Department...  

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

OE National SCADA Test Bed Fiscal Year 2009 Work Plan DOEOE National SCADA Test Bed Fiscal Year 2009 Work Plan This document is designed to help guide and strengthen the DOEOE...

225

Argonne National Laboratory Terahertz- and Millimeter-Wave Test Facility  

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

PROFILE: PROFILE: Argonne Homeland Security Technologies APPLICATIONS A R G O N N E N A T I O N A L L A B O R A T O R Y Terahertz- and Millimeter-Wave Test Facility B E N E F I T S Detect Terrorist-Related Contraband with Terahertz Technology * Spectral "fingerprints" uniquely identify materials * Can identify the factory where explosives and other chemicals were manufactured * Detects minute amounts of chemicals from a distance * Identifies materials in seconds Companies that develop or manufacture instruments to detect terrorist contraband can benefit by using a unique facility at the U.S. Department of Energy's Argonne National Laboratory. Called the Terahertz Test Facility, its sensitive, new instruments - developed at Argonne and available nowhere else in the world - can obtain spectral "fingerprints" that uniquely

226

National RF Test Facility as a multipurpose development tool  

SciTech Connect

Additions and modifications to the National RF Test Facility design have been made that (1) focus its use for technology development for future large systems in the ion cyclotron range of frequencies (ICRF), (2) expand its applicability to technology development in the electron cyclotron range of frequencies (ECRF) at 60 GHz, (3) provide a facility for ELMO Bumpy Torus (EBT) 60-GHz ring physics studies, and (4) permit engineering studies of steady-state plasma systems, including superconducting magnet performance, vacuum vessel heat flux removal, and microwave protection. The facility will continue to function as a test bed for generic technology developments for ICRF and the lower hybrid range of frequencies (LHRF). The upgraded facility is also suitable for mirror halo physics experiments.

McManamy, T.J.; Becraft, W.R.; Berry, L.A.; Blue, C.W.; Gardner, W.L.; Haselton, H.H.; Hoffman, D.J.; Loring, C.M. Jr.; Moeller, F.A.; Ponte, N.S.

1983-01-01T23:59:59.000Z

227

Advanced Test Reactor Safety Basis Upgrade Lessons Learned Relative to Design Basis Verification and Safety Basis Management  

SciTech Connect

The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The reactor also provides other irradiation services such as radioisotope production. The ATR and its support facilities are located at the Test Reactor Area of the Idaho National Engineering and Environmental Laboratory (INEEL). An audit conducted by the Department of Energy's Office of Independent Oversight and Performance Assurance (DOE OA) raised concerns that design conditions at the ATR were not adequately analyzed in the safety analysis and that legacy design basis management practices had the potential to further impact safe operation of the facility.1 The concerns identified by the audit team, and issues raised during additional reviews performed by ATR safety analysts, were evaluated through the unreviewed safety question process resulting in shutdown of the ATR for more than three months while these concerns were resolved. Past management of the ATR safety basis, relative to facility design basis management and change control, led to concerns that discrepancies in the safety basis may have developed. Although not required by DOE orders or regulations, not performing design basis verification in conjunction with development of the 10 CFR 830 Subpart B upgraded safety basis allowed these potential weaknesses to be carried forward. Configuration management and a clear definition of the existing facility design basis have a direct relation to developing and maintaining a high quality safety basis which properly identifies and mitigates all hazards and postulated accident conditions. These relations and the impact of past safety basis management practices have been reviewed in order to identify lessons learned from the safety basis upgrade process and appropriate actions to resolve possible concerns with respect to the current ATR safety basis. The need for a design basis reconstitution program for the ATR has been identified along with the use of sound configuration management principles in order to support safe and efficient facility operation.

G. L. Sharp; R. T. McCracken

2004-05-01T23:59:59.000Z

228

Environmental Assessment for Decontamination and Decommissioning of the Juggernaut Reactor at Argonne National Laboratory Â… East Argonne, Illinois  

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

Finding of No Significant Impact Finding of No Significant Impact Proposed Decontamination and Decommissioning of the Juggernaut Reactor at Argonne National Laboratory - East Argonne, Illinois AGENCY: U. S. Department of Energy (DOE) ACTION: Finding of No Significant Impact (FONSI) SUMMARY: DOE has prepared an Environmental Assessment (EA), DOE/EA-1483, evaluating the decontamination and decommissioning of the Juggernaut Reactor at Argonne National Laboratory-East (ANL-E), in Argonne, Illinois. The decontamination and decommissioning of the reactor is needed to ensure the protection of the health and safety of the public, DOE and contractor employees, and the environment, consistent with DOE Order 5400.5, Radiation Protection of the Public and the Environment. Based on the analysis in the EA, DOE has determined that the proposed action does not

229

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

SciTech Connect

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

230

RECENT ADVANCES IN HIGH TEMPERATURE ELECTROLYSIS AT IDAHO NATIONAL LABORATORY: STACK TESTS  

SciTech Connect

High temperature steam electrolysis is a promising technology for efficient sustainable large-scale hydrogen production. Solid oxide electrolysis cells (SOECs) are able to utilize high temperature heat and electric power from advanced high-temperature nuclear reactors or renewable sources to generate carbon-free hydrogen at large scale. However, long term durability of SOECs needs to be improved significantly before commercialization of this technology. A degradation rate of 1%/khr or lower is proposed as a threshold value for commercialization of this technology. Solid oxide electrolysis stack tests have been conducted at Idaho National Laboratory to demonstrate recent improvements in long-term durability of SOECs. Electrolytesupported and electrode-supported SOEC stacks were provided by Ceramatec Inc., Materials and Systems Research Inc. (MSRI), and Saint Gobain Advanced Materials (St. Gobain), respectively for these tests. Long-term durability tests were generally operated for a duration of 1000 hours or more. Stack tests based on technology developed at Ceramatec and MSRI have shown significant improvement in durability in the electrolysis mode. Long-term degradation rates of 3.2%/khr and 4.6%/khr were observed for MSRI and Ceramatec stacks, respectively. One recent Ceramatec stack even showed negative degradation (performance improvement) over 1900 hours of operation. A three-cell short stack provided by St. Gobain, however, showed rapid degradation in the electrolysis mode. Improvements on electrode materials, interconnect coatings, and electrolyteelectrode interface microstructures contribute to better durability of SOEC stacks.

X, Zhang; J. E. O'Brien; R. C. O'Brien; J. J. Hartvigsen; G. Tao; N. Petigny

2012-07-01T23:59:59.000Z

231

Acoustic emission monitoring of hot functional testing: Watts Bar Unit 1 Nuclear Reactor  

SciTech Connect

Acoustic emission (AE) monitoring of selected pressure boundary areas at TVA's Watts Bar, Unit 1 Nuclear Power Plant during hot functional preservice testing is described in this report. The report deals with background, methodology, and results. The work discussed here is a major milestone in a program supported by NRC to develop and demonstrate application of AE monitoring for continuous surveillance of reactor pressure boundaries to detect and evaluate growing flaws. The subject work demonstrated that anticipated problem areas can be overcome. Work is continuing toward AE monitoring during reactor operation.

Hutton, P.H.; Dawson, J.F.; Friesel, M.A.; Harris, J.C.; Pappas, R.A.

1984-06-01T23:59:59.000Z

232

Development and Testing of a High Capacity Plasma Chemical Reactor in the Ukraine  

SciTech Connect

This project, Development and Testing of a High Capacity Plasma Chemical Reactor in the Ukraine was established at the Kharkiv Institute of Physics and Technology (KIPT). The associated CRADA was established with Campbell Applied Physics (CAP) located in El Dorado Hills, California. This project extends an earlier project involving both CAP and KIPT conducted under a separate CRADA. The initial project developed the basic Plasma Chemical Reactor (PCR) for generation of ozone gas. This project built upon the technology developed in the first project, greatly enhancing the output of the PCR while also improving reliability and system control.

Reilly, Raymond W.

2012-07-30T23:59:59.000Z

233

The results of systems tests of the 500 kV busbar controllable shunting reactor in the Tavricheskaya substation  

SciTech Connect

The results of systems tests of the 500 kV busbar magnetization-controllable shunting reactor (CSR), set up in the Tavricheskaya substation, including measurements of the quality of the electric power, the harmonic composition of the network currents of the reactor for different values of the reactive power consumed, the determination of the regulating characteristics of the reactor, the speed of response of the shunting reactor in the current and voltage stabilization modes, and also the operation of the reactor under dynamic conditions for different perturbations, are presented. The results obtained are analyzed.

Gusev, S. I. [JSC 'FSK EES' (Russian Federation); Karpov, V. N.; Kiselev, A. N.; Kochkin, V. I. [Scientific-Research Institute of Electric Power Engineering (VNIIE) - Branch of the JSC 'NTTs Elektroenergetiki', (Russian Federation)

2009-09-15T23:59:59.000Z

234

RERTR Program: goals, progress and plans. [Reduced Enrichment Research and Test Reactor  

SciTech Connect

The status of the US Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a brief outline of RERTR Program objectives and goals, program accomplishments are discussed with emphasis on the development, demonstration and application of new LEU fuels. Most program activities have proceeded as planned, and a combination of two silicide fuels (U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al) holds excellent promise for achieving the long-term program goals. Current plans and schedules project the uranium density of qualified RERTR fuels for plate-type reactors to grow by approximately 1 g U/cm/sup 3/ each year, from the nearly null value of 1982 to the 7.0 g U/cm/sup 3/ which will be reached in early 1989. The technical needs of research reactors for HEU exports are also estimated to undergo a gradual but dramatic decline in the coming years.

Travelli, A.

1984-09-25T23:59:59.000Z

235

Advanced Test Reactor Complex Facilities Radioactive Waste Management Basis and DOE Manual 435.1-1 Compliance Tables  

SciTech Connect

U.S. Department of Energy Order 435.1, 'Radioactive Waste Management,' along with its associated manual and guidance, requires development and maintenance of a radioactive waste management basis for each radioactive waste management facility, operation, and activity. This document presents a radioactive waste management basis for Idaho National Laboratory's Advanced Test Reactor Complex facilities that manage radioactive waste. The radioactive waste management basis for a facility comprises existing laboratory-wide and facility-specific documents. U.S. Department of Energy Manual 435.1-1, 'Radioactive Waste Management Manual,' facility compliance tables also are presented for the facilities. The tables serve as a tool to develop the radioactive waste management basis.

Lisa Harvego; Brion Bennett

2011-11-01T23:59:59.000Z

236

Investigation of the November 8, 2011, Plutonium Contamination in the Zero Power Physics Reactor Facility, at the Idaho National Laboratory  

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

On November 8, 2011, workers at the Idaho National Laboratory (INL) Materials and Fuels Complex (MFC) Zero Power Physics Reactor (ZPPR) Facility were packaging plutonium (Pu) reactor fuel plates. Two of the fuel storage containers had atypical labels indicating potential abnormalities with the fuel plates located inside. Upon opening one of the storage containers, the workers discovered a Pu fuel plate wrapped in plastic and tape. When the workers attempted to remove the wrapping material, an uncontrolled release of radioactive contaminants occurred, resulting in the contamination of 16 workers and the facility

237

SRS Small Modular Reactors  

SciTech Connect

The small modular reactor program at the Savannah River Site and the Savannah River National Laboratory.

None

2012-04-27T23:59:59.000Z

238

SRS Small Modular Reactors  

ScienceCinema (OSTI)

The small modular reactor program at the Savannah River Site and the Savannah River National Laboratory.

None

2014-05-21T23:59:59.000Z

239

10 CFR 830 Major Modification Determination for Advanced Test Reactor RDAS and LPCIS Replacement  

SciTech Connect

The replacement of the ATR Control Complex's obsolete computer based Reactor Data Acquisition System (RDAS) and its safety-related Lobe Power Calculation and Indication System (LPCIS) software application is vitally important to ensure the ATR remains available to support this national mission. The RDAS supports safe operation of the reactor by providing 'real-time' plant status information (indications and alarms) for use by the reactor operators via the Console Display System (CDS). The RDAS is a computer support system that acquires analog and digital information from various reactor and reactor support systems. The RDAS information is used to display quadrant and lobe powers via a display interface more user friendly than that provided by the recorders and the Control Room upright panels. RDAS provides input to the Nuclear Engineering ATR Surveillance Data System (ASUDAS) for fuel burn-up analysis and the production of cycle data for experiment sponsors and the generation of the Core Safety Assurance Package (CSAP). RDAS also archives and provides for retrieval of historical plant data which may be used for event reconstruction, data analysis, training and safety analysis. The RDAS, LPCIS and ASUDAS need to be replaced with state-of-the-art technology in order to eliminate problems of aged computer systems, and difficulty in obtaining software upgrades, spare parts, and technical support. The major modification criteria evaluation of the project design did not lead to the conclusion that the project is a major modification. The negative major modification determination is driven by the fact that the project requires a one-for-one equivalent replacement of existing systems that protects and maintains functional and operational requirements as credited in the safety basis.

David E. Korns

2012-05-01T23:59:59.000Z

240

Certification testing at the National Wind Technology Center  

SciTech Connect

The International Electrotechnical Commission is developing a new standard that defines power performance measurement techniques. The standard will provide the basis for international recognition of a wind turbine`s performance primarily for certification, but also for qualification for tax and investment incentives, and for contracts. According to the standard, the power performance characteristics are defined by a measured power curve and by projections of annual energy production for a range of wind conditions. The National Wind Technology Center (NWTC) has adopted these power performance measurement techniques. This paper reviews the results of the NWTC`s first test conducted under the new protocol on the Atlantic Orient Corporation`s AOC 15/50 wind turbine at the NWTC. The test required collecting sufficient data to establish a statistically significant database over a range of wind speeds and conditions. From the data, the power curve was calculated. Then the results from a site calibration procedure determined the flow distortion between winds measured at the turbine location and those measured at the meteorological tower. Finally, this paper discusses the uncertainty analysis that was performed in accordance with the standard. Use of these procedures resulted in the definition of the AOC 15/50`s power curve within about 3 kW.

Huskey, A.; Link, H.

1996-11-01T23:59:59.000Z

Note: This page contains sample records for the topic "national reactor testing" 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

DOE National SCADA Test Bed Program Multi-Year Plan | Department of Energy  

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

National SCADA Test Bed Program Multi-Year Plan National SCADA Test Bed Program Multi-Year Plan DOE National SCADA Test Bed Program Multi-Year Plan This document presents the National SCADA Test Bed Program Multi-Year Plan, a coherent strategy for improving the cyber security of control systems in the energy sector. The NSTB Program is conducted within DOE's Office of Electricity Delivery and Energy Reliability (OE), which leads national efforts to modernize the electric grid, enhance the security and reliability of the energy infrastructure, and facilitate recovery from disruptions to the energy supply. The Plan covers the planning period of fiscal year 2008 to 2013. DOE National SCADA Test Bed Program Multi-Year Plan More Documents & Publications DOE/OE National SCADA Test Bed Fiscal Year 2009 Work Plan

242

Pap test use and cervical cancer incidence in First Nations women living in Manitoba  

Science Journals Connector (OSTI)

...special-property>author-choice Pap test use and cervical cancer incidence in First Nations...Sciences, University of Manitoba This study examined Pap test utilization, Pap test results, and cervical cancer incidence among First...

Kathleen M. Decker; Alain A. Demers; Erich V. Kliewer; Natalie Biswanger; Grace Musto; Brenda Elias; Jane Griffith; Donna Turner

243

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

SciTech Connect

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

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

1988-01-01T23:59:59.000Z

244

National Library of Energy (BETA): the Department of Energy's...  

Office of Scientific and Technical Information (OSTI)

Advanced Light Source Advanced Photon Source Advanced Research Projects Agency - Energy Advanced Test Reactor National Scientific User Facility Albuquerque Complex Alcator...

245

Decontamination and decommissioning of the Experimental Boiling Water Reactor (EBWR): Project final report, Argonne National Laboratory  

SciTech Connect

The Final Report for the Decontamination and Decommissioning (D&D) of the Argonne National Laboratory - East (ANL-E) Experimental Boiling Water Reactor (EBWR) facility contains the descriptions and evaluations of the activities and the results of the EBWR D&D project. It provides the following information: (1) An overall description of the ANL-E site and EBWR facility. (2) The history of the EBWR facility. (3) A description of the D&D activities conducted during the EBWR project. (4) A summary of the final status of the facility, including the final and confirmation surveys. (5) A summary of the final cost, schedule, and personnel exposure associated with the project, including a summary of the total waste generated. This project report covers the entire EBWR D&D project, from the initiation of Phase I activities to final project closeout. After the confirmation survey, the EBWR facility was released as a {open_quotes}Radiologically Controlled Area,{close_quotes} noting residual elevated activity remains in inaccessible areas. However, exposure levels in accessible areas are at background levels. Personnel working in accessible areas do not need Radiation Work Permits, radiation monitors, or other radiological controls. Planned use for the containment structure is as an interim transuranic waste storage facility (after conversion).

Fellhauer, C.R.; Boing, L.E. [Argonne National Lab., IL (United States); Aldana, J. [NES, Inc., Danbury, CT (United States)

1997-03-01T23:59:59.000Z

246

Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Main report  

SciTech Connect

Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). The study results are presented in two volumes. Volume 1 (Main Report) contains the results in summary form.

Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

1982-03-01T23:59:59.000Z

247

Light Water Reactor Sustainability Program Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study  

SciTech Connect

Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about LWR design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the RISMC Pathway R&D is to support plant decisions for risk-informed margins management with the aim to improve economics, reliability, and sustain safety of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. This report describes the RISMC methodology demonstration where the Advanced Test Reactor (ATR) was used as a test-bed for purposes of determining safety margins. As part of the demonstration, we describe how both the thermal-hydraulics and probabilistic safety calculations are integrated and used to quantify margin management strategies.

Curtis Smith; David Schwieder; Cherie Phelan; Anh Bui; Paul Bayless

2012-08-01T23:59:59.000Z

248

CONTROL TESTING OF THE UK NATIONAL NUCLEAR LABORATORY'S RADBALL TECHNOLOGY AT SAVANNAH RIVER NATIONAL LABORATORY  

SciTech Connect

The UK National Nuclear Laboratory (NNL) has developed a remote, non-electrical, radiation-mapping device known as RadBall (patent pending), which offers a means to locate and quantify radiation hazards and sources within contaminated areas of the nuclear industry. To date, the RadBall has been deployed in a number of technology trials in nuclear waste reprocessing plants at Sellafield in the UK. The trials have demonstrated the successful ability of the RadBall technology to be deployed and retrieved from active areas. The positive results from these initial deployment trials and the anticipated future potential of RadBall have led to the NNL partnering with the Savannah River National Laboratory (SRNL) to further underpin and strengthen the technical performance of the technology. RadBall consists of a colander-like outer shell that houses a radiation-sensitive polymer sphere. It has no power requirements and can be positioned in tight or hard-to reach places. The outer shell works to collimate radiation sources and those areas of the polymer sphere that are exposed react, becoming increasingly less transparent, in proportion to the absorbed dose. The polymer sphere is imaged in an optical-CT scanner which produces a high resolution 3D map of optical attenuation coefficients. Subsequent analysis of the optical attenuation maps provides information on the spatial distribution and strength of the sources in a given area forming a 3D characterization of the area of interest. This study completed at SRNL addresses key aspects of the testing of the RadBall technology. The first set of tests was performed at Savannah River Nuclear Solutions Health Physics Instrument Calibration Laboratory (HPICL) using various gamma-ray sources and an x-ray machine with known radiological characteristics. The objective of these preliminary tests was to identify the optimal dose and collimator thickness. The second set of tests involved a highly contaminated hot cell. The objective of this part of the testing was to characterize a hot cell with unknown radiation sources. The RadBall calibration experiments and hot cell deployment completed at SRNL were successful in that for each trial, the technology was able to locate the radiation sources. The NNL believe that the ability of RadBall to be remotely deployed with no electrical supplies into difficult to access areas of plant and locate and quantify radiation hazards is a unique radiation mapping service. The NNL consider there to be significant business potential associated with this innovative technology.

Farfan, E.

2009-11-23T23:59:59.000Z

249

Acoustic Emission Monitoring of ASME Section III Hydrostatic Test: Watts Bar Unit 1 Nuclear Reactor  

SciTech Connect

Through the cooperation of the Tennessee Valley Authority, Pacific Northwest Laboratory has installed instrumentation on Watts Bar Nuclear Power Plant Unit 1 for the purpose of test and evaluation of acoustic emission (AE) monitoring of nuclear reactor pressure vessels and piping for flaw detection. This report describes the acoustic emission monitoring performed during the ASME Section III hydrostatic testing of Watts Bar Nuclear Power Plant Unit 1 and the results obtained. Highlights of the results are: • Spontaneous AE was detected from a nozzle area during final pressurization. • Evaluation of the apparent source of the spontaneous AE using an empirically derived AE/fracture mechanics relationship agreed within a factor of two with an evaluation by ASME Section XI Code procedures. • AE was detected from a fracture specimen which was pressure coupled to the 10-inch accumulator nozzle. This provided reassurance of adequate system sensitivity. • High background noise was observed when all four reactor coolant pumps were operating. Work is continuing at Watts Bar Unit 1 toward AE monitoring hot functional testing and subsequently monitoring during reactor operation.

Hutton,, P. H.; Taylor,, T. T.; Dawson,, J. F.; Pappas,, R. A.; Kurtz,, R. J.

1982-06-01T23:59:59.000Z

250

IDAHO NATIONAL LABORATORY  

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

History of the Idaho National Laboratory (INL) History of the Idaho National Laboratory (INL) You are here: DOE-ID Home > Inside ID > Brief History Site History The Idaho National Laboratory (INL), an 890-square-mile section of desert in southeast Idaho, was established in 1949 as the National Reactor Testing Station. Initially, the missions at the INL were the development of civilian and defense nuclear reactor technologies and management of spent nuclear fuel. Fifty-two reactors—most of them first-of-a-kind—were built, including the Navy’s first prototype nuclear propulsion plant. Of the 52 reactors, three remain in operation at the site. In 1951, the INL achieved one of the most significant scientific accomplishments of the century—the first use of nuclear fission to produce a usable quantity of electricity at the Experimental Breeder Reactor No.

251

Axial effects of xenon-samarium poisoning in the advanced test reactor  

SciTech Connect

The paper details an analytical study of the time-dependent behavior in the spatial distributions of xenon and samarium fission product poisons in the Advanced Test Reactor (ATR) during operation and after shutdown. The results of this study provide insight into the behavior and significance of the changing spatial distributions of fission product poisons with respect to the prediction of shim positions at critical for reactor restart after a xenon shutdown. The study was performed with the PDQ neutron diffusion theory code and ENDF/B-V cross sections using a one-dimensional radial model of an ATR lobe and a two-dimensional radial-axial (RZ) model of an ATR lobe. The PDQ results were supported by a review of the basic differential equations, which describe the buildup and decay of the xenon and samarium fission product poisons and precursors. The ATR is a 250-MW, uranium-aluminum-fueled reactor used to study the effects of irradiation on reactor materials. Forty highly enriched uranium fuel elements are arranged in a serpentine configuration within the compact core resulting in a very high power density of (1.0 MW/[ell] of core).

Auslander, D.J.; Smith, A.C.; McCracken, R.T. (Idaho National Engineering Lab., Idaho Falls (United States))

1990-01-01T23:59:59.000Z

252

Testing of Passive Safety System Performance for Higher Power Advanced Reactors  

SciTech Connect

This report describes the results of NERI research on the testing of advanced passive safety performance for the Westinghouse AP1000 design. The objectives of this research were: (a) to assess the AP1000 passive safety system core cooling performance under high decay power conditions for a spectrum of breaks located at a variety of locations, (b) to compare advanced thermal hydraulic computer code predictions to the APEX high decay power test data and (c) to develop new passive safety system concepts that could be used for Generation IV higher power reactors.

brian G. Woods; Jose Reyes, Jr.; John Woods; John Groome; Richard Wright

2004-12-31T23:59:59.000Z

253

An Experimental Test Facility to Support Development of the Fluoride Salt Cooled High Temperature Reactor  

SciTech Connect

The need for high-temperature (greater than 600 C) energy exchange and delivery systems is significantly increasing as the world strives to improve energy efficiency and develop alternatives to petroleum-based fuels. Liquid fluoride salts are one of the few energy transport fluids that have the capability of operating at high temperatures in combination with low system pressures. The Fluoride Salt-Cooled High-Temperature Reactor design uses fluoride salt to remove core heat and interface with a power conversion system. Although a significant amount of experimentation has been performed with these salts, specific aspects of this reactor concept will require experimental confirmation during the development process. The experimental facility described here has been constructed to support the development of the Fluoride Salt Cooled High Temperature Reactor concept. The facility is capable of operating at up to 700 C and incorporates a centrifugal pump to circulate FLiNaK salt through a removable test section. A unique inductive heating technique is used to apply heat to the test section, allowing heat transfer testing to be performed. An air-cooled heat exchanger removes added heat. Supporting loop infrastructure includes a pressure control system; trace heating system; and a complement of instrumentation to measure salt flow, temperatures, and pressures around the loop. The initial experiment is aimed at measuring fluoride salt heat transfer inside a heated pebble bed similar to that used for the core of the pebble bed advanced high-temperature reactor. This document describes the details of the loop design, auxiliary systems used to support the facility, the inductive heating system, and facility capabilities.

Yoder Jr, Graydon L [ORNL] [ORNL; Aaron, Adam M [ORNL] [ORNL; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK)] [University of Tennessee, Knoxville (UTK); Fugate, David L [ORNL] [ORNL; Holcomb, David Eugene [ORNL] [ORNL; Kisner, Roger A [ORNL] [ORNL; Peretz, Fred J [ORNL] [ORNL; Robb, Kevin R [ORNL] [ORNL; Wilgen, John B [ORNL] [ORNL; Wilson, Dane F [ORNL] [ORNL

2014-01-01T23:59:59.000Z

254

Testing, Training, and Signature Devices | Y-12 National Security...  

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

Testing, Training, and ... Testing, Training, and Signature Devices Y-12 manufactures specialized uranium testing, training, and signature devices to support the nuclear detection...

255

Ultracold neutron source at the PULSTAR reactor: Engineering design and cryogenic testing  

Science Journals Connector (OSTI)

Abstract Construction is completed and commissioning is in progress for an ultracold neutron (UCN) source at the PULSTAR reactor on the campus of North Carolina State University. The source utilizes two stages of neutron moderation, one in heavy water at room temperature and the other in solid methane at ~ 40 K , followed by a converter stage, solid deuterium at 5 K, that allows a single down scattering of cold neutrons to provide UCN. The UCN source rolls into the thermal column enclosure of the PULSTAR reactor, where neutrons will be delivered from a bare face of the reactor core by streaming through a graphite-lined assembly. The source infrastructure, i.e., graphite-lined assembly, heavy-water system, gas handling system, and helium liquefier cooling system, has been tested and all systems operate as predicted. The research program being considered for the PULSTAR UCN source includes the physics of UCN production, fundamental particle physics, and material surface studies of nanolayers containing hydrogen. In the present paper we report details of the engineering and cryogenic design of the facility as well as results of critical commissioning tests without neutrons.

E. Korobkina; G. Medlin; B. Wehring; A.I. Hawari; P.R. Huffman; A.R. Young; B. Beaumont; G. Palmquist

2014-01-01T23:59:59.000Z

256

The RERTR (Reduced Enrichment Research and Test Reactor) program: A progress report  

SciTech Connect

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1985, the activities, results, and new developments which occurred in 1986 are reviewed. The second miniplate series, concentrating on U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al fuels, was expanded and its irradiation continued. Postirradiation examinations of several of these miniplates and of six previously irradiated U/sub 3/Si/sub 2/-Al full-size elements were completed with excellent results. The whole-core ORR demonstration with U/sub 3/Si/sub 2/-Al fuel at 4.8 g U/cm/sup 3/ is well under way and due for completion before the end of 1987. DOE removed an important barrier to conversions by announcing that the new LEU fuels will be accepted for reprocessing. New DOE prices for enrichment and reprocessing services were calculated to have minimal effect on HEU reactors, and to reduce by about 8 to 10% the total fuel cycle costs of LEU reactors. New program activities include preliminary feasibility studies of LEU use in DOE reactors, evaluation of the feasibility to use LEU targets for the production of fission-product /sup 99/Mo, and responsibility for coordinating safety evaluations related to LEU conversions of US university reactors, as required by NRC. Achievement of the final program goals is projected for 1990. This progress could not have been achieved without close international cooperation, whose continuation and intensification are essential to the achievement of the ultimate goals of the RERTR Program.

Travelli, A.

1986-11-01T23:59:59.000Z

257

The results of systems tests of the 500 kV busbar controllable shunting reactor in the Tavricheskaya substation  

Science Journals Connector (OSTI)

The results of systems tests of the 500 kV busbar magnetization-controllable shunting reactor (CSR), set up in the Tavricheskaya substation, including measurements of the quality of the...

S. I. Gusev; V. N. Karpov; A. N. Kiselev; V. I. Kochkin

2009-09-01T23:59:59.000Z

258

DOE - Office of Legacy Management -- Idaho National Engineering and  

Office of Legacy Management (LM)

Idaho National Engineering and Idaho National Engineering and Environmental Laboratory - 015 FUSRAP Considered Sites Site: Idaho National Engineering and Environmental Laboratory (015) Designated Name: Alternate Name: Location: Evaluation Year: Site Operations: Site Disposition: Radioactive Materials Handled: Primary Radioactive Materials Handled: Radiological Survey(s): Site Status: In operation since 1949, the Idaho National Engineering and Environmental Laboratory (INEEL) is a Department of Energy multiprogram national laboratory that supports the Department¿s missions of environmental quality, energy resources, science, and national security. Originally named the National Reactor Testing Station, the INEEL was once the site of the world¿s largest concentration of nuclear reactors. 52 test reactors most

259

LANL's ChemCam conducts first laser test over the weekend | National...  

National Nuclear Security Administration (NNSA)

ChemCam conducts first laser test over the weekend | National Nuclear Security Administration People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear...

260

Modeling and Experimental Tests on the Hydraulically Driven Control Rod option for IRIS Reactor  

SciTech Connect

The adoption of Internal Control Rod Drive Mechanisms (ICRDMs) represents a valuable alternative to classical, external CRDMs based on electro-magnetic devices, as adopted in current PWRs. The advantages on the safety features of the reactor are apparent: inherent elimination of the Rod Ejection accidents and of possible concerns about the vessel head penetrations. A further positive feedback on the design is the reduction of the primary system overall dimensions. Within the frame of the ICRDM concepts, the Hydraulically Driven Control Rod solution is investigated as a possible option for the IRIS integral reactor. After a brief comparison of the solutions currently proposed for integral reactors, the configuration of the Hydraulic Control Rod device for IRIS, made up by an external movable piston and an internal fixed cylinder, is described. A description of the whole control system is reported as well. Particular attention is devoted to the Control Rod profile characterization, performed by means of a Computational Fluid Dynamics (CFD) analysis. The investigation of the system behavior has been carried out, including the dynamic equilibrium and its stability properties, the withdrawal and insertion step movement and the sensitivity study on command time periods. A suitable dynamic model has been set up for the mentioned purposes: the models corresponding to the various Control Rod system devices have been written in an Object-Oriented language (Modelica), thus allowing an easy implementation of such a system into the simulator for the whole reactor. Finally, a preliminary low pressure, low temperature, reduced length experimental facility has been built. Tests on HDCR stability and operational transients have been performed. The results are compared with the dynamic system model and CFD simulation model, showing good agreement between simulations and experimental data. During these preliminary tests, the control system performed correctly, allowing stable dynamic equilibrium positions for the Control Rod and stable behavior during withdrawal and insertion steps. (authors)

Cammi, Antonio; Ricotti, Marco E.; Vitulo, Alessia [Department of Nuclear Engineering, Politecnico di Milano, Via Ponzio, 34/3, 20133 Milano (Italy)

2004-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "national reactor testing" 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

2007 international meeting on Reduced Enrichment for Research and Test Reactors (RERTR). Abstracts and available papers presented at the meeting  

SciTech Connect

The Meeting papers discuss research and test reactor fuel performance, manufacturing and testing. Some of the main topics are: conversion from HEU to LEU in different reactors and corresponding problems and activities; flux performance and core lifetime analysis with HEU and LEU fuels; physics and safety characteristics; measurement of gamma field parameters in core with LEU fuel; nondestructive analysis of RERTR fuel; thermal hydraulic analysis; fuel interactions; transient analyses and thermal hydraulics for HEU and LEU cores; microstructure research reactor fuels; post irradiation analysis and performance; computer codes and other related problems.

NONE

2008-07-15T23:59:59.000Z

262

Sandia National Laboratories: Sunshine to Petrol  

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

in the lab's hand-built precision prototype reactor currently undergoing tests at the solar furnace at the National Solar Thermal Test Facility . The team has made skillful use...

263

Thomas Wallner | Argonne National Laboratory  

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

Argonne National Laboratory's Omnivorous Engine Argonne National Laboratory's Omnivorous Engine Argonne National Laboratory's Omnivorous Engine Argonne National Laboratory's Omnivorous Engine Browse by Topic Energy Energy efficiency Vehicles Alternative fuels Automotive engineering Biofuels Diesel Fuel economy Fuel injection Heavy-duty vehicles Hybrid & electric vehicles Hydrogen & fuel cells Internal combustion Powertrain research Vehicle testing Building design Manufacturing Energy sources Renewable energy Bioenergy Solar energy Wind energy Fossil fuels Oil Nuclear energy Nuclear energy modeling & simulation Nuclear fuel cycle Geology & disposal Reactors Nuclear reactor safety Nuclear reactor materials Energy usage Energy life-cycle analysis Energy storage Batteries Lithium-ion batteries Lithium-air batteries Smart Grid

264

Idaho National Laboratory Testing of Advanced Technology Vehicles...  

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

& Publications Vehicle Technologies Office: 2010 Vehicle and Systems Simulation and Testing R&D Annual Progress Report AVTA HEV, NEV, BEV and HICEV Demonstrations and Testing...

265

Limited Test Ban Treaty | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Photo Gallery Jobs Apply for Our Jobs Our Jobs Working at NNSA Blog Home Limited Test Ban Treaty Limited Test Ban Treaty US Air Force Launches Satellite Carrying...

266

Waste Heat Recovery from the Advanced Test Reactor Secondary Coolant Loop  

SciTech Connect

This study investigated the feasibility of using a waste heat recovery system (WHRS) to recover heat from the Advanced Test Reactor (ATR) secondary coolant system (SCS). This heat would be used to preheat air for space heating of the reactor building, thus reducing energy consumption, carbon footprint, and energy costs. Currently, the waste heat from the reactor is rejected to the atmosphere via a four-cell, induced-draft cooling tower. Potential energy and cost savings are 929 kW and $285K/yr. The WHRS would extract a tertiary coolant stream from the SCS loop and pump it to a new plate and frame heat exchanger, from which the heat would be transferred to a glycol loop for preheating outdoor air supplied to the heating and ventilation system. The use of glycol was proposed to avoid the freezing issues that plagued and ultimately caused the failure of a WHRS installed at the ATR in the 1980s. This study assessed the potential installation of a new WHRS for technical, logistical, and economic feasibility.

Donna Post Guillen

2012-11-01T23:59:59.000Z

267

Progress of the RERTR (Reduced Enrichment Research and Test Reactor) Program in 1989  

SciTech Connect

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1988, the major events, findings, and activities of 1989 are reviewed. The scope of the RERTR Program activities was curtailed, in 1989, by an unexpected legislative restriction which limited the ability of the Arms Control and Disarmament Agency to adequately fund the program. Nevertheless, the thrust of the major planned program activities was maintained, and meaningful results were obtained in several areas of great significance for future work. 15 refs., 12 figs.

Travelli, A.

1989-01-01T23:59:59.000Z

268

Methods for nondestructive testing of austenitic high-temperature gas-cooled reactor components  

SciTech Connect

Safety-relevant components of high-temperature gas-cooled reactor components are mostly fabricated in nickel-based alloys and austenitic materials like Inconel-617, Hastelloy-X, Nimonic-86, or Incoloy-800H. Compared to ferritic steels, these austenitic materials can have a coarse-grained microstructure, especially in weldments and castings. Coarse-grained or elastic anisotropic materials are difficult to inspect with ultrasonics due to strong attenuation, high noise level (scattering, ''grass'' indications), and sound beam distortions (skewing, splitting, and mode conversion). Only few results dealing with the nondestructive testing of nickel-based alloys are known. The problem area, solutions, and first experiences are reported.

Gobbels, K.; Kapitza, H.

1984-09-01T23:59:59.000Z

269

E-Print Network 3.0 - aerospace system test reactor Sample Search...  

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

Engineering 15 www.eprg.group.cam.ac.uk EPRGWORKINGPAPER Summary: Accelerator-Driven Subcritical Reactor Park Michel-Alexandre Cardin1 Engineering Systems Division... and reactor...

270

National Geothermal Data System Architecture Design, Testing and Maintenance  

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

Project objective: To create the National Geothermal Data System (NGDS) comprised of a core and distributed network of databases and data sites that will comprise a federated system for acquisition, management, maintenance, and dissemination of geothermal and related data.

271

The RERTR (Reduced Enrichment Research and Test Reactor) Program: Progress and plans  

SciTech Connect

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1986, the activities, results, and new developments which occurred in 1987 are reviewed. Irradiation of the second miniplate series, concentrating on U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al fuels, was completed and postirradiation examinations were performed on many of its miniplates. The whole-core ORR demonstration with U/sub 3/Si/sub 2/-Al fuel at 4.8 g U/cm/sup 3/ was completed at the end of March with excellent results and with 29 elements estimated to have reached at least 40% average burnup. Good progress was made in the area of LEU usage for the production of fission /sup 99/Mo, and in the coordination of safety evaluations related to LEU conversions of US university reactors. Planned activities include testing and demonstrating advanced fuels intended to allow use of reduced enrichment uranium in very-high-performance reactors. Two candidate fuels are U/sub 3/Si-Al with 19.75% enrichment and U/sub 3/Si/sub 2/-Al with 45% enrichment. Demonstration of these fuels will include irradiation of full-size elements and, possibly, a full-core demonstration. Achievement of the final program goals is still projected for 1990. This progress could not have been possible without the close international cooperation which has existed from the beginning, and which is essential to the ultimate success of the RERTR Program.

Travelli, A.

1987-01-01T23:59:59.000Z

272

Status of the RERTR (Reduced Enrichment Research and Test Reactor) Program  

SciTech Connect

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1987, the major events, findings and activities of 1988 are reviewed. The US Nuclear Regulatory Commission issued a formal and generic approval of the use of U3Si2-Al dispersion fuel in research and test reactors, with densities up to 4.8 g U/cmT. New significant findings from postirradiation examinations, from ion-beam irradiations, and from analytical modeling, have raised serious doubts about the potential of LEU U3Si-Al dispersion fuel for applications requiring very high uranium densities and high burnups (>6 g U/cmT, >50% burnup). As a result of these findings, the fuel development efforts have been redirected towards three new initiatives: (1) a systematic application of ion-beam irradiations to screen new materials; (2) application of Hot Isostatic Pressing (HIP) procedures to produce U3Si2-Al plates with high uranium densities and thin uniform cladding; and (3) application of HIP procedures to produce plates with U3Si wires imbedded in an aluminum matrix, achieving stability, high uranium density, and thin uniform cladding. The new fuel concepts hold the promise of extraordinary performance potential and require approximately five years to develop.

Travelli, A.

1988-01-01T23:59:59.000Z

273

Reactor Safety Research Programs  

SciTech Connect

This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

Edler, S. K.

1981-07-01T23:59:59.000Z

274

Verification Survey of the Building 315 Zero Power Reactor-6 Facility, Argonne National Laboratory-East, Argonne, Illinois  

SciTech Connect

Oak Ridge Institute for Science and Education (ORISE) conducted independent verification radiological survey activities at Argonne National Laboratory’s Building 315, Zero Power Reactor-6 facility in Argonne, Illinois. Independent verification survey activities included document and data reviews, alpha plus beta and gamma surface scans, alpha and beta surface activity measurements, and instrumentation comparisons. An interim letter report and a draft report, documenting the verification survey findings, were submitted to the DOE on November 8, 2006 and February 22, 2007, respectively (ORISE 2006b and 2007).

W. C. Adams

2007-05-25T23:59:59.000Z

275

Materials Transportation Testing & Analysis at Sandia National Laboratories  

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

Testing Testing Doug Ammerman, (505) 845-8158 Type B packages that transport radioactive materials must survive a sequence of full-scale (actual physical size) impact, puncture, fire, and immersion tests designed to replicate transportation accident conditions. The Hypothetical Accident Conditions (six tests as defined in 10 CFR Part 71.73) tests 1 through 4 (Drop, Crush, Puncture and Fire) are sequential, test 5 (Immersion) is performed on either a previously tested or untested package. Free Drop Test Crush Test Puncture Test Thermal Test Immersion Test [drop] Click to view picture [crush] Click to view picture [puncture] Click to view picture [thermal] Click to view picture [immersion] Click to view picture Dropping a package from 30 feet onto an unyielding target. (the unyielding target forces all of the deformation to be in the package, none in the target). The speed on impact is 44 feet per second or 30 miles per hour. Dropping a 1100 pound steel plate from 30 feet onto a package. This test is only required for packages weighing less than 1100 pounds. The speed on impact is 44 feet per second or 30 miles per hour. Dropping a package from 40 inches onto a welded, 6 inch diameter, steel spike. The speed on impact is 14.6 feet per second or 10 miles per hour. Placing a package 40 inches above a pool of burning fuel for 30 minutes at 800 degrees Celsius (1475 degrees Fahrenheit). Placing a package under 50 feet of water for 8 hours. Fissile material packages are also immersed under 3 feet of water for 8 hours sequentially after tests 1 through 4

276

DOE/EA-1557; Final Envrionmental Assessment for the National Security Test Range  

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

Environmental Assessment for the National Security Test Range DOE/EA-1557 April, 2007 CONTENTS 1. PURPOSE AND NEED ..................................................................................................................... 1 2. ALTERNATIVES .............................................................................................................................. 3 2.1. Consolidate Testing on a New National Security Test Range at the INL (Preferred Alternative)3 2.1.1 Construction Activities ................................................................................................. 6 2.1.2 Operational Activities................................................................................................... 6 2.2 Alternatives Considered, but Eliminated from Detailed Analysis................................................

277

Flow Test At Lassen Volcanic National Park Area (Janik & Mclaren, 2010) |  

Open Energy Info (EERE)

Lassen Volcanic National Park Area (Janik & Mclaren, 2010) Lassen Volcanic National Park Area (Janik & Mclaren, 2010) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Lassen Volcanic National Park Area (Janik & Mclaren, 2010) Exploration Activity Details Location Lassen Volcanic National Park Area Exploration Technique Flow Test Activity Date Usefulness not indicated DOE-funding Unknown Notes Water samples were collected during nitrogen-stimulated flow tests in 1978, but no information was provided on sampling conditions. The well was flowed again for the last time in 1982, but the flow test lasted only 1 h (Thompson, 1985). References Cathy J. Janik, Marcia K. McLaren (2010) Seismicity And Fluid Geochemistry At Lassen Volcanic National Park, California- Evidence For Two

278

U.S. DOE/OE National SCADA Test Bed Supports | Department of Energy  

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

U.S. DOE/OE National SCADA Test Bed Supports U.S. DOE/OE National SCADA Test Bed Supports U.S. DOE/OE National SCADA Test Bed Supports To help advance the U.S. Department of Energy (DOE) National SCADA Test Bed's (NSTB) efforts to enhance control system security in the energy sector, DOE's Office of Electricity Delivery and Energy Reliability (OE) recently awarded a total of nearly $8 million to fund five industry-led projects: Hallmark Project. (PDF 789 KB) Will commercialize the Secure SCADA Communications Protocol (SSCP), which marks SCADA messages with a unique identifier that must be authenticated before the function is carried out, ensuring message integrity. (Lead: Schweitzer Engineering Laboratories; Partners: Pacific Northwest National Laboratories, CenterPoint Energy) Detection and Analysis of Threats to the Energy Sector (DATES) (PDF

279

Nevada National Security Site Nuclear Testing Artifacts Become Part of U.S.  

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

Nevada National Security Site Nuclear Testing Artifacts Become Part Nevada National Security Site Nuclear Testing Artifacts Become Part of U.S. Cultural Archive Nevada National Security Site Nuclear Testing Artifacts Become Part of U.S. Cultural Archive April 1, 2012 - 12:00pm Addthis Stanchions are among the remnants of Smoky Tower. Stanchions are among the remnants of Smoky Tower. LAS VEGAS, NV - The Nevada National Security Site's (NNSS) historic Smoky site may soon join a long list of former nuclear testing locations eligible for inclusion in the National Register of Historic Places. The Desert Research Institute (DRI) is currently working alongside the Nevada Site Office (NSO) to determine the eligibility of Smoky and a number of other EM sites slated for cleanup and closure. "In the last year, we've conducted assessments at over 30 EM sites,"

280

Fabrication and Testing of Full-Length Single-Cell Externally Fueled Converters for Thermionic Reactors  

SciTech Connect

The preceding paper described designs and analyses of thermionic reactors employing full-core-length single-cell converters with their heated emitters located on the outside of their internally cooled collectors, and it presented results of detailed parametric analyses which illustrate the benefits of this unconventional design. The present paper describes the fabrication and testing of full-length prototypical converters, both unfueled and fueled, and presents parametric results of electrically heated tests. The unfueled converter tests demonstrated the practicality of operating such long converters without shorting across a 0.3-mm interelectrode gap. They produced a measured peak output of 751 watts(e) from a single diode and a peak efficiency of 15.4%. The fueled converter tests measured the parametric performance of prototypic UO(subscript 2)-fueled converters designed for subsequent in-pile testing. They employed revolver-shaped tungsten elements with a central emitter hole surrounded by six fuel chambers. The full-length converters were heated by a water-cooled RF-induction coil inside an ion-pumped vacuum chamber. This required development of high-vacuum coaxial RF feedthroughs. In-pile test rules required multiple containment of the UO (subscript 2)-fuel, which complicated the fabrication of the test article and required successful development of techniques for welding tungsten and other refractory components. The test measured a peak power output of 530 watts(e) or 7.1 watts/cm (superscript 2) at an efficiency of 11.5%. There are three copies in the file. Cross-Reference a copy FSC-ESD-217-94-529 in the ESD files with a CID #8574.

Schock, Alfred

1994-06-01T23:59:59.000Z

Note: This page contains sample records for the topic "national reactor testing" 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
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281

Sandia National Laboratories: Joint Sandia-DOE-HMRC Testing of...  

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

and Exhibition (EU PVSC) EC Top Publications Reference Model 5 (RM5): Oscillating Surge Wave Energy Converter Experimental Wave Tank Test for Reference Model 3 Floating- Point...

282

Sandia National Laboratories: Molten Salt Test Loop Pump Installed  

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

and Exhibition (EU PVSC) EC Top Publications Reference Model 5 (RM5): Oscillating Surge Wave Energy Converter Experimental Wave Tank Test for Reference Model 3 Floating- Point...

283

Sandia National Laboratories: Sandia Solar Energy Test System...  

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

and Exhibition (EU PVSC) EC Top Publications Reference Model 5 (RM5): Oscillating Surge Wave Energy Converter Experimental Wave Tank Test for Reference Model 3 Floating- Point...

284

Sandia National Laboratories: Thermal Pulses for Boeing Test...  

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

and Exhibition (EU PVSC) EC Top Publications Reference Model 5 (RM5): Oscillating Surge Wave Energy Converter Experimental Wave Tank Test for Reference Model 3 Floating- Point...

285

Sandia National Laboratories: Launch of Solar Testing Site in...  

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

and Exhibition (EU PVSC) EC Top Publications Reference Model 5 (RM5): Oscillating Surge Wave Energy Converter Experimental Wave Tank Test for Reference Model 3 Floating- Point...

286

Sandia National Laboratories: Molten Salt Test Loop Commissioning  

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

and Exhibition (EU PVSC) EC Top Publications Reference Model 5 (RM5): Oscillating Surge Wave Energy Converter Experimental Wave Tank Test for Reference Model 3 Floating- Point...

287

South Carolina Opens Nation's Largest Wind Drivetrain Testing...  

Office of Environmental Management (EM)

makes it ideal for American and international companies to testing larger offshore wind turbines. Supported by a 47 million Energy Department investment as well as about 60...

288

Idaho National Laboratory Testing of Advanced Technology Vehicles  

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

Timeline The Advanced Vehicle Testing Activity (AVTA) is an annually funded DOE activity Barriers Barriers addressed * High risk to develop and purchase plug-in...

289

Idaho National Laboratory Testing of Advanced Technology Vehicles  

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

Vehicle Testing Activity (AVTA) is to support DOE's goal of petroleum reduction and energy security by: - Providing benchmarked field-based vehicle performance and system...

290

Idaho National Laboratory Testing of Advanced Technology Vehicles  

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

(not modeled) instrumentation and data collection of vehicle charging demand and energy costs at Tacoma Power, in Tacoma Washington * Tested PHEVs with lithium batteries...

291

Nuclear Detection and Sensor Testing Center | Y-12 National Security...  

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

Detection and ... Nuclear Detection and Sensor Testing Center As part of our increased global nuclear nonproliferation efforts, Y-12 commissioned the Nuclear Detection and Sensor...

292

Materials Transportation Testing & Analysis at Sandia National Laboratories  

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

RMIR (Radioactive Materials Incident Report) Database Transportation RMIR (Radioactive Materials Incident Report) Database Transportation Accident and Incident Experience,1971-1999 Access Hazardous Materials Information System (HMIS) the primary source of national data for the Federal, state, and local governmental agencies responsible for the safety of hazardous materials transportation. Rail Transport Highway Transport Air Transport The Radioactive Material Incident Report (RMIR) Database was developed in 1981 at the Transportation Technology Center of Sandia National Laboratories (SNL) to support its research and development activities for the U.S. Department of Energy (DOE). This database contains information about radioactive materials transportation incidents that have occurred in the U.S. from 1971 through 1999. These data were drawn from the U.S.

293

Nuclear Detection and Sensor Testing Center | Y-12 National Security  

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

Detection and ... Detection and ... Nuclear Detection and Sensor Testing Center As part of our increased global nuclear nonproliferation efforts, Y-12 commissioned the Nuclear Detection and Sensor Testing Center, which offers dedicated facilities for the testing of radiation detection capabilities using enriched and highly enriched uranium. In addition to supporting measurements of instrumentation for detecting ionizing radiation, non-destructive measurements of both fissile and non-fissile materials may be deployed at NDSTC. The NDSTC supports proliferation detection, nuclear safeguards, emergency response, treaty verification, and university research. We can test devices with various forms and quantities of HEU, and the Center offers subject matter experts to assist in planning measurements, safely deploy material,

294

Idaho National Laboratory Testing of Advanced Technology Vehicles  

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

* Development of codes and standards for products and testing is required Budget FY 2013 project funding *1.8 million FY 2014 project funding *1.05M Partners * Intertek...

295

Sandia National Laboratories: acceler-ated lifetime test  

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

acceler-ated lifetime test Sandia R&D Funded under New DOE SunShot Program On November 27, 2013, in Energy, News, News & Events, Partnership, Photovoltaic, Renewable Energy, Solar,...

296

Safety evaluation for packaging (onsite) plutonium recycle test reactor graphite cask  

SciTech Connect

This safety evaluation for packaging (SEP) provides the evaluation necessary to demonstrate that the Plutonium Recycle Test Reactor (PRTR) Graphite Cask meets the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B, fissile, non-highway route controlled quantities of radioactive material within the 300 Area of the Hanford Site. The scope of this SEP includes risk, shieldling, criticality, and.tiedown analyses to demonstrate that onsite transportation safety requirements are satisfied. This SEP also establishes operational and maintenance guidelines to ensure that transport of the PRTR Graphite Cask is performed safely in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required.

Romano, T.

1997-09-29T23:59:59.000Z

297

Vehicle Technologies Office Merit Review 2014: Idaho National Laboratory Testing of Advanced Technology Vehicles  

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

Presentation given by Idaho National Laboratory at 2014 DOE Hydrogen and Fuel Cells Program and Vehicle Technologies Office Annual Merit Review and Peer Evaluation Meeting about testing of advanced...

298

South Carolina Opens Nation’s Largest Wind Drivetrain Testing Facility  

Office of Energy Efficiency and Renewable Energy (EERE)

Clemson University Project Converted Former Navy Warehouse to First-of-its-Kind Testing Facility for Land-Based and Offshore Wind Turbines

299

Final Site Specific Decommissioning Inspection Report #2 for the University of Washington Research and Test Reactor, Seattle, Washington  

SciTech Connect

During the period of August through November 2006, ORISE performed a comprehensive IV at the University of Washington Research and Test Reactor Facility. The objective of the ORISE IV was to validate the licensee’s final status survey processes and data, and to assure the requirements of the DP and FSSP were met.

S.J. Roberts

2007-03-20T23:59:59.000Z

300

Materials Transportation Testing & Analysis at Sandia National Laboratories  

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

Analysis Analysis Doug Ammerman, (505) 845-8158 Structural analysis utilizes computer design and analysis tools to provide package designers and certifiers with the most accurate method of determining package response to transportation environments. Computer analysis is an application of known engineering principles that take advantage of high-power computing capabilities in solving the response of computer models to various environments with complex mathematical calculations. It can be used for package certification by generating a computer model of a test object (package) and subjecting it to an accident environment to understand its response. A computer model must be constructed with the same weights, dimensions, hardnesses, specific heat, conduction, etc. as an

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


301

Status of the RERTR program: overview, progress and plans. [Reduced Enrighment Research and Test Reactor  

SciTech Connect

The status of the US Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a summary of the accomplishments which the RERTR Program had achieved by the end of 1984 with its many international partners, emphasis is placed on the progress achieved during 1985 and on current plans and schedules. A new miniplate series, concentrating on U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al fuels, was fabricated and is well into irradiation. The whole-core ORR demonstration is scheduled to begin in November 1985, with U/sub 3/Si/sub 2/-Al fuel at 4.8 g U/cm/sup 3/. Altogether, 921 full-size test and prototype elements have been ordered for fabrication with reduced enrichment and the new technologies. Qualification of U/sub 3/Si-Al fuel with approx.7 g U/cm/sup 3/ is still projected for 1989. This progress could not have been achieved without the close international cooperation which has existed since the beginning, and whose continuation and intensification will be essential to the achievement of the long-term RERTR goals.

Travelli, A.

1985-01-01T23:59:59.000Z

302

Technical Letter Report, An Evaluation of Ultrasonic Phased Array Testing for Reactor Piping System Components Containing Dissimilar Metal Welds, JCN N6398, Task 2A  

SciTech Connect

Research is being conducted for the U.S. Nuclear Regulatory Commission at the Pacific Northwest National Laboratory to assess the effectiveness and reliability of advanced nondestructive examination (NDE) methods for the inspection of light-water reactor components. The scope of this research encom¬passes primary system pressure boundary materials including dissimilar metal welds (DMWs), cast austenitic stainless steels (CASS), piping with corrosion-resistant cladding, weld overlays, inlays and onlays, and far-side examinations of austenitic piping welds. A primary objective of this work is to evaluate various NDE methods to assess their ability to detect, localize, and size cracks in steel components that challenge standard and/or conventional inspection methodologies. This interim technical letter report provides a summary of a technical evaluation aimed at assessing the capabilities of phased-array (PA) ultrasonic testing (UT) methods as applied to the inspection of small-bore DMW components that exist in the reactor coolant systems (RCS) of pressurized water reactors (PWRs). Operating experience and events such as the circumferential cracking in the reactor vessel nozzle-to-RCS hot leg pipe at V.C. Summer nuclear power station, identified in 2000, show that in PWRs where primary coolant water (or steam) are present under normal operation, Alloy 82/182 materials are susceptible to pressurized water stress corrosion cracking. The extent and number of occurrences of DMW cracking in nuclear power plants (domestically and internationally) indicate the necessity for reliable and effective inspection techniques. The work described herein was performed to provide insights for evaluating the utility of advanced NDE approaches for the inspection of DMW components such as a pressurizer surge nozzle DMW, a shutdown cooling pipe DMW, and a ferritic (low-alloy carbon steel)-to-CASS pipe DMW configuration.

Diaz, Aaron A.; Cinson, Anthony D.; Crawford, Susan L.; Anderson, Michael T.

2009-11-30T23:59:59.000Z

303

High-Temperature Superconducting Cable Testing Gregory S. Boebinger, National High Magnetic Field Laboratory  

E-Print Network (OSTI)

High-Temperature Superconducting Cable Testing Gregory S. Boebinger, National High Magnetic Field-Temperature Superconducting (HTS) Cables are desirable for application in large high-field magnets (>20 T), especially when). Of the three HTS magnet cable concepts emerging, the Conductor On Round Core was the first that was tested

Weston, Ken

304

Relap5-3d model validation and benchmark exercises for advanced gas cooled reactor application  

E-Print Network (OSTI)

HTTR High Temperature engineering Test Reactor INET Institute of Nuclear Energy Technology LWR Light Water Reactor OKBM Test Design Bureau for Machine Building ORNL Oak Ridge National Laboratory RCCS Reactor Cavity Cooling System... to be at right angles to each other, ignoring an angular distribution of radiant heat.7 MORECA, used by ORNL, simulates accident scenarios for certain gas-cooled reactor types.7 INET conducts their analysis using Thermix, which performs two...

Moore, Eugene James Thomas

2006-08-16T23:59:59.000Z

305

National SCADA Test Bed Enhancing control systems security in the energy  

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

SCADA Test Bed Enhancing control systems security in the SCADA Test Bed Enhancing control systems security in the energy sector National SCADA Test Bed Enhancing control systems security in the energy sector Improving the security of energy control systems has become a national priority. Since the mid-1990's, security experts have become increasingly concerned about the threat of malicious cyber attacks on the vital supervisory control and data acquisition (SCADA) and distributed control systems (DCS) used to monitor and manage our energy infrastructure. Many of the systems still in use today were designed to operate in closed, proprietary networks. National SCADA Test Bed Enhancing control systems security in the energy sector More Documents & Publications NSTB Summarizes Vulnerable Areas Transmission and Distribution World March 2007: DOE Focuses on Cyber

306

CRAD, DOE Oversight- Oak Ridge National Laboratory High Flux Isotope Reactor  

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

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a DOE independent oversight assessment of the Oak Ridge National Laboratory programs for oversight of its contractors.

307

EU, U.S., Russia, Asian States Sign Nuclear-Fusion Reactor May 24 (Bloomberg) --The European Union, the U.S., Russia and Asian nations  

E-Print Network (OSTI)

Europe EU, U.S., Russia, Asian States Sign Nuclear-Fusion Reactor Pact May 24 (Bloomberg) -- The European Union, the U.S., Russia and Asian nations including China signed a treaty to build the first percent, China, India, South Korea and Russia 10 percent each and France 8 percent, #12;according

308

Neutron Scattering Science User Office, neutronusers@ornl.gov or (865) 574-4600. Proposals for beam time at Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR)  

E-Print Network (OSTI)

Neutron Scattering Science User Office, neutronusers@ornl.gov or (865) 574-4600. Proposals for beam Wildgruber, wildgrubercu@ornl.gov. VISION CallforProposals neutrons.ornl.gov Neutron Scattering Science - Oak time at Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR) and Spallation Neutron Source

Pennycook, Steve

309

Certification testing of the Los Alamos National Laboratory Heat Source/Radioisotopic Thermoelectric Generator shipping container  

SciTech Connect

The Heat Source/Radioisotopic Thermoelectric Generator shipping counter is a Type B packaging currently under development by Los Alamos National Laboratory. Type B packaging for transporting radioactive material is required to maintain containment and shielding after being exposed to normal and hypothetical accident environments defined in Title 10 of the Code of Federal Regulations Part 71. A combination of testing and analysis is used to verify the adequacy of this packaging design. This report documents the testing portion of the design verification. Six tests were conducted on a prototype package: a water spray test, a 4-foot normal conditions drop test, a 30-foot drop test, a 40-inch puncture test, a 30-minute thermal test, and an 8-hour immersion test.

Bronowski, D.R.; Madsen, M.M.

1991-09-01T23:59:59.000Z

310

Baseline risk assessment of the perched water system at the INEL test reactor area  

SciTech Connect

A baseline health risk assessment (HRA) was prepared to evaluate potential risks to human health and the environment posed by the Perched Water System (PWS) at the Test Reactor Area (TRA). The PWS has been designated Operable Unit 2-12, one of the 13 operable units identified at TRA. During the period from 1962 to 1990, a total of 6770 million gal of water were discharged from the TRA to unlined surface ponds. Wastewater discharged to the surface ponds at TRA percolates downward through the surficial alluvium and the underlying basalt bedrock. A resulting shallow perched water zone has formed at the interface between the surficial sediments and the underlying basalt. Further downward movement of groundwater is again impeded by a low-permeability layer of silt, clay, and sand encountered at a depth of [approximately]150 ft. The deep perched water zone occurs on top of this low-permeability interbed. An evaluation was made as to whether potential risks for the PWS could justify implementing a remedial action. The risk evaluation consisted of two parts, the human health evaluation and the ecological evaluation.

Gordon, J.W.; Sinton, P.O. (Dames Moore, Denver, CO (United States)); Jensen, N. (DOE, Idaho Falls, ID (United States)); McCormick, S. (Idaho National Engineering Lab., Idaho Falls, ID (United States))

1993-01-01T23:59:59.000Z

311

Safety Design Strategy for the Advanced Test Reactor Primary Coolant Pump and Motor Replacement Project  

SciTech Connect

In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

Noel Duckwitz

2011-06-01T23:59:59.000Z

312

Safety Design Strategy for the Advanced Test Reactor Emergency Firewater Injection System Replacement Project  

SciTech Connect

In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

Noel Duckwitz

2011-06-01T23:59:59.000Z

313

Safety Design Strategy for the Advanced Test Reactor Diesel Bus (E-3) and Switchgear Replacement Project  

SciTech Connect

In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

Noel Duckwitz

2011-06-01T23:59:59.000Z

314

Last U.S. Underground Nuclear Test Conducted | National Nuclear Security  

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

U.S. Underground Nuclear Test Conducted | National Nuclear Security U.S. Underground Nuclear Test Conducted | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > About Us > Our History > NNSA Timeline > Last U.S. Underground Nuclear Test Conducted Last U.S. Underground Nuclear Test Conducted September 23, 1992 USA Last U.S. Underground Nuclear Test Conducted

315

Last U.S. Underground Nuclear Test Conducted | National Nuclear Security  

National Nuclear Security Administration (NNSA)

U.S. Underground Nuclear Test Conducted | National Nuclear Security U.S. Underground Nuclear Test Conducted | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > About Us > Our History > NNSA Timeline > Last U.S. Underground Nuclear Test Conducted Last U.S. Underground Nuclear Test Conducted September 23, 1992 USA Last U.S. Underground Nuclear Test Conducted

316

In-vessel Retention Strategy for High Power Reactors - K-INERI Final Report (includes SBLB Test Results for Task 3 on External Reactor Vessel Cooling (ERVC) Boiling Data and CHF Enhancement Correlations)  

SciTech Connect

In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe PWR (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing LWRs. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a collaborative, three-year, U.S. - Korean International Nuclear Energy Research Initiative (INERI) project was completed in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) investigated the performance of ERVC and an in-vessel core catcher (IVCC) to determine if IVR is feasible for reactors up to 1500 MWe.

F. B. Cheung; J. Yang; M. B. Dizon; J. Rempe

2005-01-01T23:59:59.000Z

317

The technique and preliminary results of LEU U-Mo full-size IRT type fuel testing in the MIR reactor  

SciTech Connect

In March 2007 in-pile testing of LEU U-Mo full-size IRT type fuel elements was started in the MIR reactor. Four prototype fuel elements for Uzbekistan WWR SM reactor are being tested simultaneously - two of tube type design and two of pin type design. The dismountable irradiation devices were constructed for intermediate reloading and inspection of fuel elements during reactor testing. The objective of the test is to obtain the experimental results for determination of more reliable design and licensing fuel elements for conversion of the WWR SM reactor. The heat power of fuel elements is measured on-line by thermal balance method. The distribution of fission density and burn-up of uranium in the volume of elements are calculated by using the MIR reactor MCU code (Monte-Carlo) model. In this paper the design of fuel elements, the technique, main parameters and preliminary results are described. (author)

Izhutov, A.L.; Starkov, V.A.; Pimenov, V.V.; Fedoseev, V.Ye. [Research Reactor Complex, RIAR, 433510, Dimitrovgrad-10, Ulyanovsk Region (Russian Federation); Dobrikova, I.V.; Vatulin, A.V.; Suprun, V.B. [A.A. Bochvar All-Russian Scientific Research Institute of Inorganic Materials, P. O. Box 369, 123060, Moscow (Russian Federation); Kartashov, Ye.F.; Lukichev, V.A. [Research and Development Institute of Nuclear Energy and Industry, P. O. Box 788, 107014, Moscow (Russian Federation); Troyanov, V.M.; Enin, A.A.; Tkachev, A.A. [OAO 'TVEL' 119017, ul. B. Ordinka 24/26, Moscow (Russian Federation)

2008-07-15T23:59:59.000Z

318

Design and Testing of a Boron Carbide Capsule for Spectral Tailoring in Mixed-Spectrum Reactors  

SciTech Connect

A boron carbide capsule has been designed and used for spectral-tailoring experiments at the TRIGA reactor at Washington State University. Irradiations were conducted in pulsed mode and in continuous operation for up to 4 hours. A cadmium cover was used to reduce thermal heating. The neutron spectrum calculated with MCNP was found to be in good agreement with reactor dosimetry measurements using the STAY'SL computer code. The neutron spectrum resembles that of a fast reactor. Design of a capsule using boron carbide enriched in {sup 10}B shows that it is possible to produce a neutron spectrum similar to {sup 235}U fission.

Greenwood, Lawrence R.; Wittman, Richard S.; Pierson, Bruce D.; Metz, Lori A.; Payne, Rosara F.; Finn, Erin C.; Friese, Judah I.

2012-03-01T23:59:59.000Z

319

Basic Test of A 3-Phase Superconducting Fault Current Limiting Reactor  

Science Journals Connector (OSTI)

A novel 3-phase superconducting fault current limiting reactor (SCFCLR) for power system ... some experiments, the fundamental behavior of the limiter is confirmed. In the experiment, two ... exhibits very small ...

H. Kado; T. Ishigohka

1990-01-01T23:59:59.000Z

320

Safety of power transformers, power supplies, reactors and similar products - Part 1: General requirements and tests  

E-Print Network (OSTI)

This International Standard deals with safety aspects of power transformers, power supplies, reactors and similar products such as electrical, thermal and mechanical safety. This standard covers the following types of dry-type transformers, power supplies, including switch mode power supplies, and reactors, the windings of which may be encapsulated or non-encapsulated. It has the status of a group safety publication in accordance with IEC Guide 104.

International Electrotechnical Commission. Geneva

2005-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "national reactor testing" 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

Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1  

SciTech Connect

This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future.

Owen, M.B.

1997-04-01T23:59:59.000Z

322

Cyber Security Audit and Attack Detection Toolkit: National SCADA Test Bed  

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

Audit and Attack Detection Toolkit: National SCADA Audit and Attack Detection Toolkit: National SCADA Test Bed May 2008 Cyber Security Audit and Attack Detection Toolkit: National SCADA Test Bed May 2008 This project of the cyber security audit and attack detection toolkit is adding control system intelligence to widely deployed enterprise vulnerability scanners and security event managers While many energy utilities employ vulnerability scanners and security event managers (SEM) on their enterprise systems, these tools often lack the intelligence necessary to be effective in control systems. This two-year project aims to integrate control system intelligence into widely deployed vulnerability scanners and SEM, and to integrate security incident detection intelligence into control system historians. These upgrades will

323

Full-length U-xPu-10Zr (x=0, 8, 19 wt%) Fast Reactor Fuel Test in FFTF  

SciTech Connect

The Integral Fast Reactor-1 (IFR-1) experiment performed in the Fast Flux Test Facility (FFTF) was the only U-Pu-10Zr (Pu-0, 8 and 19 wt%) metallic fast reactor test with commercial-length (91.4 cm active fuel column length) conducted to date. With few remaining test reactors there is little opportunity for performing another test with a long active fuel column. The assembly was irradiated to the goal burnup of 10 at.%. The beginning of life (BOL) peak cladding temperature of the hottest pin was 608?C, cooling to 522?C at end of life (EOL). Selected fuel pins were examined non destructively using neutron radiography, precision axial gamma scanning, and both laser and spiral contact cladding profilometry. Destructive exams included plenum gas pressure, volume, and gas composition determinations on a number of pins followed by optical metallography, electron probe microanalysis (EPMA), and alpha and beta gamma autoradiography on a single U-19Pu-10Zr pin. The post-irradiation examinations (PIEs) showed very few differences compared to the short-pin (34.3 cm fuel column) testing performed on fuels of similar composition in Experimental Breeder Reactor-II (EBR-II). The fuel column grew axially slightly less than observed in the short pins, but with the same pattern of decreasing growth with increasing Pu content. There was a difference in the fuel-cladding chemical interaction (FCCI) in that the maximum cladding penetration by interdiffusion with fuel/fission products did not occur at the top of the fuel column where the cladding temperature is highest, as observed in EBR-II tests. Instead, the more exaggerated fission-rate profile of the FFTF pins resulted in a peak FCCI at ~0.7 X/L axial location along the fuel column. This resulted from a lower production of rare earth fission products higher in the fuel column as well as a much smaller delta-T between fuel center and cladding, and therefore less FCCI, despite the higher cladding temperature. This behavior could actually help extend the life of a fuel pin in a “long pin” reactor design to a higher peak fuel burnup.

D. L. Porter; H.C. Tsai

2012-08-01T23:59:59.000Z

324

Fission product transport and behavior during two postulated loss-of-flow transients in the Advanced Test Reactor  

SciTech Connect

The fission product behavior during two postulated loss-of-flow accidents (leading to high- and low-pressure core degradations) in the Advanced Test Reactor (ATR) has been analyzed. These transients are designated ATR transients LCP 15 (high pressure) and LPP9 (low pressure). Normally, transients of this nature would be easily mitigated using existing safety systems and procedures. In these analyses, failure of these safety systems was assumed so that core degradation and fission product release could be studied. A probabilistic risk analysis was performed that indicated that the probability of occurrence for these two transients is on the order of 10[sup [minus]5] and 10[sup [minus]7] per reactor year for LCP15 and LPP9, respectively. The fission product behavior analysis included calculations of the gaseous and highly volatile fission product (xenon, krypton, cesium, iodine, and tellurium) inventories in the fuel before accident initiation, release of the fission products from the fuel into the reactor vessel during core melt, the probable chemical forms, and transport of the fission products from the core through the reactor vessel and existing piping to the confinement. In addition to a base-case analysis of fission product behavior, a series of analyses was performed to determine the sensitivity of fission product release to several parameters including steam flow rate, (structural) aluminum oxidation, and initial aerosol size. The base-case analyses indicate that the volatile fission products (excluding the noble gases) will be transported as condensed species on zinc aerosols.

Adams, J.P.; Carboneau, M.L.; Hagrman, D.L. (Idaho National Engineering Lab. EG and G Idaho, Idaho Falls, ID (United States))

1993-07-01T23:59:59.000Z

325

Scoping analysis of the Advanced Test Reactor using SN2ND  

SciTech Connect

A detailed set of calculations was carried out for the Advanced Test Reactor (ATR) using the SN2ND solver of the UNIC code which is part of the SHARP multi-physics code being developed under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in DOE-NE. The primary motivation of this work is to assess whether high fidelity deterministic transport codes can tackle coupled dynamics simulations of the ATR. The successful use of such codes in a coupled dynamics simulation can impact what experiments are performed and what power levels are permitted during those experiments at the ATR. The advantages of the SN2ND solver over comparable neutronics tools are its superior parallel performance and demonstrated accuracy on large scale homogeneous and heterogeneous reactor geometries. However, it should be noted that virtually no effort from this project was spent constructing a proper cross section generation methodology for the ATR usable in the SN2ND solver. While attempts were made to use cross section data derived from SCALE, the minimal number of compositional cross section sets were generated to be consistent with the reference Monte Carlo input specification. The accuracy of any deterministic transport solver is impacted by such an approach and clearly it causes substantial errors in this work. The reasoning behind this decision is justified given the overall funding dedicated to the task (two months) and the real focus of the work: can modern deterministic tools actually treat complex facilities like the ATR with heterogeneous geometry modeling. SN2ND has been demonstrated to solve problems with upwards of one trillion degrees of freedom which translates to tens of millions of finite elements, hundreds of angles, and hundreds of energy groups, resulting in a very high-fidelity model of the system unachievable by most deterministic transport codes today. A space-angle convergence study was conducted to determine the meshing and angular cubature requirements for the ATR, and also to demonstrate the feasibility of performing this analysis with a deterministic transport code capable of modeling heterogeneous geometries. The work performed indicates that a minimum of 260,000 linear finite elements combined with a L3T11 cubature (96 angles on the sphere) is required for both eigenvalue and flux convergence of the ATR. A critical finding was that the fuel meat and water channels must each be meshed with at least 3 'radial zones' for accurate flux convergence. A small number of 3D calculations were also performed to show axial mesh and eigenvalue convergence for a full core problem. Finally, a brief analysis was performed with different cross sections sets generated from DRAGON and SCALE, and the findings show that more effort will be required to improve the multigroup cross section generation process. The total number of degrees of freedom for a converged 27 group, 2D ATR problem is {approx}340 million. This number increases to {approx}25 billion for a 3D ATR problem. This scoping study shows that both 2D and 3D calculations are well within the capabilities of the current SN2ND solver, given the availability of a large-scale computing center such as BlueGene/P. However, dynamics calculations are not realistic without the implementation of improvements in the solver.

Wolters, E.; Smith, M. (NE NEAMS PROGRAM); ( SC)

2012-07-26T23:59:59.000Z

326

SITEWIDE CATEGORICAL EXCLUSION FOR OUTDOOR TESTS ON MATERIALS AND COMPONENTS, PACIFIC NORTHWEST NATIONAL  

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

SITEWIDE CATEGORICAL EXCLUSION FOR OUTDOOR TESTS ON SITEWIDE CATEGORICAL EXCLUSION FOR OUTDOOR TESTS ON MATERIALS AND COMPONENTS, PACIFIC NORTHWEST NATIONAL LABORATORY, RICHLAND, WASHINGTON Proposed AetioD: The U.S. Department of Energy (DOE) Pacific Northwest Site Office (PNSO) proposes to conduct outdoor tests and experiments on materials and equipment components under controlled conditions. No source, special nuclear, or byproduct materials would be involved, but encapsulated radioactive sources manufactured to applicable standards or other radiological materials could be used in activities under this categorical exclusion (eX). LoeatioD of Action: The locations would include DOE property at the Pacific Northwest National Laboratory (PNNL) Site and other offsite outdoor locations. Description of the Proposed Action:

327

Test Results From The Idaho National Laboratory Of The NASA Bi-Supported Cell Design  

SciTech Connect

The Idaho National Laboratory has been researching the application of solid-oxide fuel cell technology for large-scale hydrogen production. As a result, the Idaho National Laboratory has been testing various cell designs to characterize electrolytic performance. NASA, in conjunction with the University of Toledo, has developed a new cell concept with the goals of reduced weight and high power density. This paper presents results of the INL's testing of this new solid oxide cell design as an electrolyzer. Gas composition, operating voltage, and other parameters were varied during testing. Results to date show the NASA cell to be a promising design for both high power-to-weight fuel cell and electrolyzer applications.

C Stoots; J O'Brien; T Cable

2009-11-01T23:59:59.000Z

328

Development of a national spill test facility data base. Topical report, February 1994--February 1995  

SciTech Connect

In the United States, the production of gas, liquid and solid fuels and the associated chemical use accounts for significant volumes of material with the potential of becoming hazardous. Accidental spills or releases of these hazardous materials do occur, and action must be taken to minimize damage to life, property, and the environment. Because of the hazards of testing with chemical spills, a national spill test facility (STF) and an associated testing program have been established to systematically develop new data on the effects and mitigation of hazardous chemical spills Western Research Institute (WRI), in conjunction with the DOE, is developing a comprehensive national spill test data base. I The data base will be easily accessible by industry and the public on the Spill Research Bulletin Board System and will allow users to download spill test data and test descriptions, as well as an extensive bibliography. The 1990 Clean Air Act and Amendments (CAAA) requires that at least two chemicals be field tested at the STF and at least 10 chemicals be studied each year. The chemicals to be studied are chosen with priority given to those that present the greatest risk to human health. The National Spill Test Facility Data Base will include a common chemical data base covering the overlap of federal chemical lists and significant information from other sources. Also, the (CAAA) directs the DOE and EPA to work together with the STF and industry to provide a scientific and engineering basis for writing regulations for implementation of the (CAAA). The data base will be a primary resource in this effort.

NONE

1995-02-01T23:59:59.000Z

329

MANAGEMENT OF RESEARCH AND TEST REACTOR ALUMINUM SPENT NUCLEAR FUEL - A TECHNOLOGY ASSESSMENT  

SciTech Connect

The Department of Energy's Environmental Management (DOE-EM) Program is responsible for the receipt and storage of aluminum research reactor spent nuclear fuel or used fuel until ultimate disposition. Aluminum research reactor used fuel is currently being stored or is anticipated to be returned to the U.S. and stored at DOE-EM storage facilities at the Savannah River Site and the Idaho Nuclear Technology and Engineering Center. This paper assesses the technologies and the options for safe transportation/receipt and interim storage of aluminum research reactor spent fuel and reviews the comprehensive strategy for its management. The U.S. Department of Energy uses the Appendix A, Spent Nuclear Fuel Acceptance Criteria, to identify the physical, chemical, and isotopic characteristics of spent nuclear fuel to be returned to the United States under the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The fuel is further evaluated for acceptance through assessments of the fuel at the foreign sites that include corrosion damage and handleability. Transport involves use of commercial shipping casks with defined leakage rates that can provide containment of the fuel, some of which are breached. Options for safe storage include wet storage and dry storage. Both options must fully address potential degradation of the aluminum during the storage period. This paper focuses on the various options for safe transport and storage with respect to technology maturity and application.

Vinson, D.

2010-07-11T23:59:59.000Z

330

E-Print Network 3.0 - argonne tank research and test reactor...  

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

for Large-Scale Problems Summary: Argonne National Laboratory, Argonne, IL 60439 3 Praxair Inc., 39 Old Ridgebury Road, Danbury, CT 06810... November, 2010 To be submitted to...

331

New Interpretation of Alpha-Particle-Driven Instabilities in Deuterium-Tritium Experiments on the Tokamak Fusion Test Reactor  

Science Journals Connector (OSTI)

The original description of alpha particle driven instabilities in the Tokamak Fusion Test Reactor in terms of toroidal Alfvén eigenmodes (TAEs) remained inconsistent with three fundamental characteristics of the observations: (i) the variation of the mode frequency with toroidal mode number, (ii) the chirping of the mode frequency for a given toroidal mode number, and (iii) the antiballooning density perturbation of the modes. It is now shown that these characteristics can be explained by observing that cylindrical-like modes can exist in the weak magnetic shear region of the plasma that then make a transition to TAEs as the central safety factor decreases in time.

R. Nazikian, G. J. Kramer, C. Z. Cheng, N. N. Gorelenkov, H. L. Berk, and S. E. Sharapov

2003-09-19T23:59:59.000Z

332

Observations of the boiling process from a downward-facing torispherical surface: Confirmatory testing of the heavy water new production reactor flooded cavity design  

SciTech Connect

Reactor-scale ex-vessel boiling experiments were performed in the CYBL facility at Sandia National Laboratories. The boiling flow pattern outside the RPV bottom head shows a center pulsating region and an outer steady two-phase boundary layer region. The local heat transfer data can be correlated in terms of a modified Rohsenow correlation.

Chu, T.Y.; Bentz, J.H.; Simpson, R.B.

1995-06-01T23:59:59.000Z

333

CRAD, Configuration Management - Oak Ridge National Laboratory...  

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

Configuration Management - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Configuration Management - Oak Ridge National Laboratory High Flux Isotope Reactor February...

334

Technology, safety, and costs of decommissioning reference nuclear research and test reactors: sensitivity of decommissioning radiation exposure and costs to selected parameters  

SciTech Connect

Additional analyses of decommissioning at the reference research and test (R and T) reactors and analyses of five recent reactor decommissionings are made that examine some parameters not covered in the initial study report (NUREG/CR-1756). The parameters examined for decommissioning are: (1) the effect on costs and radiation exposure of plant size and/or type; (2) the effects on costs of increasing disposal charges and of unavailability of waste disposal capacity at licensed waste disposal facilities; and (3) the costs of and the available alternatives for the disposal of nuclear R and T reactor fuel assemblies.

Konzek, G.J.

1983-07-01T23:59:59.000Z

335

Fabrication and Testing of Full-Length Single-Cell Externally Fueled Converters for Thermionic Reactors  

SciTech Connect

Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes the fabrication and testing of full-length prototypcial converters, both unfueled and fueled, and presents parametric results of electrically heated tests.

Schock, Alfred

1995-08-01T23:59:59.000Z

336

Idaho National Laboratory Lead or Lead-Bismuth Eutectic (LBE) Test Facility - R&D Requirements, Design Criteria, Design Concept, and Concept Guidance  

SciTech Connect

The Idaho National Laboratory Lead-Bismuth Eutectic Test Facility will advance the state of nuclear technology relative to heavy-metal coolants (primarily Pb and Pb-Bi), thereby allowing the U.S. to maintain the pre-eminent position in overseas markets and a future domestic market. The end results will be a better qualitative understanding and quantitative measure of the thermal physics and chemistry conditions in the molten metal systems for varied flow conditions (single and multiphase), flow regime transitions, heat input methods, pumping requirements for varied conditions and geometries, and corrosion performance. Furthering INL knowledge in these areas is crucial to sustaining a competitive global position. This fundamental heavy-metal research supports the National Energy Policy Development Group’s stated need for energy systems to support electrical generation.1 The project will also assist the Department of Energy in achieving goals outlined in the Nuclear Energy Research Advisory Committee Long Term Nuclear Technology Research and Development Plan,2 the Generation IV Roadmap for Lead Fast Reactor development, and Advanced Fuel Cycle Initiative research and development. This multi-unit Lead-Bismuth Eutectic Test Facility with its flexible and reconfigurable apparatus will maintain and extend the U.S. nuclear knowledge base, while educating young scientists and engineers. The uniqueness of the Lead-Bismuth Eutectic Test Facility is its integrated Pool Unit and Storage Unit. This combination will support large-scale investigation of structural and fuel cladding material compatibility issues with heavy-metal coolants, oxygen chemistry control, and thermal hydraulic physics properties. Its ability to reconfigure flow conditions and piping configurations to more accurately approximate prototypical reactor designs will provide a key resource for Lead Fast Reactor research and development. The other principal elements of the Lead-Bismuth Eutectic Test Facility (in addition to the Pool Unit and Storage Unit) are the Bench Scale Unit and Supporting Systems, principal of which are the O2 Sensor/Calibration System, Feed System, Transfer System, Off- Gas System, Purge and Evacuation System, Oxygen Sensor and Control System, Data Acquisition and Control System, and the Safety Systems. Parallel and/or independent corrosion studies and convective heat transfer experiments for cylindrical and annular geometries will support investigation of heat transfer phenomena into the secondary side. In addition, molten metal pumping concepts and power requirements will be measured for future design use.

Eric P. Loewen; Paul Demkowicz

2005-05-01T23:59:59.000Z

337

The development and operational testing of an experimental reactor for gas-liquid-solid reaction systems at high temperatures and pressures  

E-Print Network (OSTI)

shaft. With the impeller in place and rotating, gas was drawn into the top port and ejected at the impeller mount. The reactor pressure was monitored via the transducer port. The transducer was a Viatran Pressure Transducer, model 103. The liquid...THE DEVELOPMENT AND OPERATIONAL TESTING OF AN EXPERIMENTAL REACTOR FOR GAS-LIQUID-SOLID REACTION SYSTEMS AT HIGH TEMPERATURES AND PRESSURES A Thesis by RICHARD KENNETH HESS Submitted to the Graduate College of Texas A&M University in partial...

Hess, Richard Kenneth

2012-06-07T23:59:59.000Z

338

Identification and evaluation of alternatives for the disposition of fluoride fuel and flush salts from the molten salt reactor experiment at Oak Ridge National Laboratory, Oak Ridge, Tennessee  

SciTech Connect

This document presents an initial identification and evaluation of the alternatives for disposition of the fluoride fuel and flush salts stored in the drain tanks at the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL). It will serve as a resource for the U.S. Department of Energy contractor preparing the feasibility study for this activity under the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA). This document will also facilitate further discussion on the range of credible alternatives, and the relative merits of alternatives, throughout the time that a final alternative is selected under the CERCLA process.

NONE

1996-08-15T23:59:59.000Z

339

PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE BROOKHAVEN GRAPHITE RESEARCH REACTOR ENGINEERED CAP, BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK DCN 5098-SR-07-0  

SciTech Connect

The Oak Ridge Institute for Science and Education (ORISE) has reviewed the project documentation and data for the Brookhaven Graphite Research Reactor (BGRR) Engineered Cap at Brookhaven National Laboratory (BNL) in Upton, New York. The Brookhaven Science Associates (BSA) have completed removal of affected soils and performed as-left surveys by BSA associated with the BGRR Engineered Cap. Sample results have been submitted, as required, to demonstrate that remediation efforts comply with the cleanup goal of {approx}15 mrem/yr above background to a resident in 50 years (BNL 2011a).

Evan Harpenau

2011-07-15T23:59:59.000Z

340

Direct Test of the Time-Independence of Fundamental Nuclear Constants Using the Oklo Natural Reactor  

E-Print Network (OSTI)

[NOTE: This 1983 preprint is being uploaded to arXiv.org after the death of its author, who supported online distribution of his work. Contact info of the submitter is at http://ilya.cc .] The positions of neutron resonances have been shown to be highly sensitive to the variation of fundamental nuclear constants. The analysis of the measured isotopic shifts in the natural fossil reactor at Oklo gives the following restrictions on the possible rates of the interaction constants variation: strong ~2x10^-19 yr^-1, electromagnetic ~5x10^-18 yr^-1, weak ~10^-12 yr^-1. These limits permit to exclude all the versions of nuclear constants contemporary variation discussed in the literature. URL: http://alexonline.info >. For more recent analyses see hep-ph/9606486, hep-ph/0205206 and astro-ph/0204069 .

Alexander I. Shlyakhter

2003-07-03T23:59:59.000Z

Note: This page contains sample records for the topic "national reactor testing" 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

Vehicle Technologies Office Merit Review 2014: Post-Test Analysis of Lithium-Ion Battery Materials at Argonne National Laboratory  

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

Presentation given by Argonne National Laboratory at 2014 DOE Hydrogen and Fuel Cells Program and Vehicle Technologies Office Annual Merit Review and Peer Evaluation Meeting about post-test...

342

High Temperature Solid-Oxide Electrolyzer 2500 Hour Test Results At The Idaho National Laboratory  

SciTech Connect

The Idaho National Laboratory (INL) has been developing the concept of using solid oxide fuel cells as electrolyzers for large-scale, high-temperature (efficient), hydrogen production. This program is sponsored by the U.S. Department of Energy under the Nuclear Hydrogen Initiative. Utilizing a fuel cell as an electrolyzer introduces some inherent differences in cell operating conditions. In particular, the performance of fuel cells operated as electrolyzers degrades with time faster. This issue of electrolyzer cell and stack performance degradation over time has been identified as a major barrier to technology development. Consequently, the INL has been working together with Ceramatec, Inc. (Salt Lake City, Utah) to improve the long-term performance of high temperature electrolyzers. As part of this research partnership, the INL conducted a 2500 hour test of a Ceramatec designed and produced stack operated in the electrolysis mode. This paper will provide a summary of experimental results to date for this ongoing test.

Carl Stoots; James O'Brien; Stephen Herring; Keith Condie; Lisa Moore-McAteer; Joseph J. Hartvigsen; Dennis Larsen

2009-11-01T23:59:59.000Z

343

Results of the analysis of the blood lymphocyte proliferation test data from the National Jewish Center  

SciTech Connect

A new approach to the analysis of the blood beryllium lymphocyte proliferation test (LPT) was presented to the Committee to Accredit Beryllium Sensitization Testing-Beryllium Industry Scientific Advisory Committee in April, 1994. Two new outlier resistant methods were proposed for the analysis of the blood LPT and compared with the approach then in use by most labs. The National Jewish Center (NJC) agreed to provide data from a study that was underway at that time. Three groups of LPT data are considered: (1) a sample of 168 beryllium exposed (BE) workers and 20 nonexposed (NE) persons; (2) 25 unacceptable LPTs, and (3) 32 abnormal LPTs for individuals known to have chronic beryllium disease (CBD). The LAV method described in ORNL-6818 was applied to each LPT. Graphical and numerical summaries similar to those presented for the ORISE data are given. Three methods were used to identify abnormal LPTs. All three methods correctly identified the 32 known CBD cases as abnormal.

From, E.L. [Oak Ridge National Lab., TN (United States). Mathematical Sciences Section; Newman, L.S.; Mroz, M.M. [National Jewish Center for Immunology and Respiratory Medicine, Denver, CO (United States)

1997-03-01T23:59:59.000Z

344

Final Environmental Assessment for the Test Capabilities Revitalization at Sandia National Laboratories/New Mexico  

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

6 6 Final Environmental Assessment for the Test Capabilities Revitalization at Sandia National Laboratories/New Mexico D E P A R T M E N T O F E N E R G Y U N I T E D S T A T E S O F A M E R I C A January 2003 Department of Energy, Office of Kirtland Site Operations Kirtland Air Force Base, Albuquerque New Mexico Test Capabilities Revitalization Environmental Assessment January 2003 Department of Energy Office of Kirtland Site Operations i TABLE OF CONTENTS 1.0 PURPOSE AND NEED FOR AGENCY ACTION ........................................................... 1 2.0 NO ACTION AND PROPOSED ACTION ALTERNATIVES........................................ 2 2.1 EXISTING FACILITIES ...........................................................................................................

345

Applicability of RELAP5-3D for Thermal-Hydraulic Analyses of a Sodium-Cooled Actinide Burner Test Reactor  

SciTech Connect

The Actinide Burner Test Reactor (ABTR) is envisioned as a sodium-cooled, fast reactor that will burn the actinides generated in light water reactors to reduce nuclear waste and ease proliferation concerns. The RELAP5-3D computer code is being considered as the thermal-hydraulic system code to support the development of the ABTR. An evaluation was performed to determine the applicability of RELAP5-3D for the analysis of a sodium-cooled fast reactor. The applicability evaluation consisted of several steps, including identifying the important transients and phenomena expected in the ABTR, identifying the models and correlations that affect the code’s calculation of the important phenomena, and evaluating the applicability of the important models and correlations for calculating the important phenomena expected in the ABTR. The applicability evaluation identified code improvements and additional models needed to simulate the ABTR. The accuracy of the calculated thermodynamic and transport properties for sodium was also evaluated.

C. B. Davis

2006-07-01T23:59:59.000Z

346

RECENT ADVANCES IN HIGH TEMPERATURE ELECTROLYSIS AT IDAHO NATIONAL LABORATORY: SINGLE CELL TESTS  

SciTech Connect

An experimental investigation on the performance and durability of single solid oxide electrolysis cells (SOECs) is under way at the Idaho National Laboratory. In order to understand and mitigate the degradation issues in high temperature electrolysis, single SOECs with different configurations from several manufacturers have been evaluated for initial performance and long-term durability. A new test apparatus has been developed for single cell and small stack tests from different vendors. Single cells from Ceramatec Inc. show improved durability compared to our previous stack tests. Single cells from Materials and Systems Research Inc. (MSRI) demonstrate low degradation both in fuel cell and electrolysis modes. Single cells from Saint Gobain Advanced Materials (St. Gobain) show stable performance in fuel cell mode, but rapid degradation in the electrolysis mode. Electrolyte-electrode delamination is found to have significant impact on degradation in some cases. Enhanced bonding between electrolyte and electrode and modification of the microstructure help to mitigate degradation. Polarization scans and AC impedance measurements are performed during the tests to characterize the cell performance and degradation.

X. Zhang; J. E. O'Brien; R. C. O'Brien

2012-07-01T23:59:59.000Z

347

Calculated and measured gas formation in beryllium samples irradiated in the high flux materials testing reactor BR2  

SciTech Connect

Beryllium samples have been irradiated in BR2, the materials testing reactor of the Nuclear Research Centre SCK/CEN at Mol, Belgium, up to fission fluence values of 5.2 10{sup 22} n/cm{sup 2} at low temperature. The gas formation (helium, tritium), as measured by SCK/CEN, as well as the induced swelling of the beryllium samples and the enhancement of the swelling due to annealing have been presented at the 17th SOFT Conference (Rome, 14--18 Sept., 1992). Since this conference, helium measurements on the same samples have been carried out at RI and calculations of the gas production have been performed, taking into account the various formation schemes. The experimental results from SCK/CEN and from RI are compared with the calculated gas formations.

De Raedt, C.M.; Sannen, L.F.; Vanmechelen, P.J. [SCK/CEN, Mol (Belgium); Oliver, B.M. [Rockwell International Corp., Canoga Park, CA (United States). Rocketdyne Div.

1994-12-31T23:59:59.000Z

348

Modeling of divertor geometry effects in China fusion engineering testing reactor by SOLPS/B2-Eirene  

SciTech Connect

The China Fusion Engineering Testing Reactor (CFETR) is currently under design. The SOLPS/B2-Eirene code package is utilized for the design and optimization of the divertor geometry for CFETR. Detailed modeling is carried out for an ITER-like divertor configuration and one with relatively open inner divertor structure, to assess, in particular, peak power loading on the divertor target, which is a key issue for the operation of a next-step fusion machine, such as ITER and CFETR. As expected, the divertor peak heat flux greatly exceeds the maximum steady-state heat load of 10?MW/m{sup 2}, which is a limit dictated by engineering, for both divertor configurations with a wide range of edge plasma conditions. Ar puffing is effective at reducing divertor peak heat fluxes below 10?MW/m{sup 2} even at relatively low densities for both cases, favoring the divertor configuration with more open inner divertor structure.

Zhao, M. L., E-mail: zml812@mail.ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027 (China); Chen, Y. P.; Li, G. Q.; Luo, Z. P. [Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 (China)] [Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 (China); Guo, H. Y. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027 (China) [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027 (China); Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 (China); General Atomics, P.O. Box 85608, San Diego, California 92186 (United States); Ye, M. Y. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027 (China) [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027 (China); Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 (China); Tendler, M. [Alfven Laboratory, Royal Institute of Technology, Stockholm (Sweden)] [Alfven Laboratory, Royal Institute of Technology, Stockholm (Sweden)

2014-05-15T23:59:59.000Z

349

Nondestructive Testing of NITE-SiC Ceramics for Fusion Reactor Application  

Science Journals Connector (OSTI)

Ultrasonic test (UT) method is a suitable one for examining the inherent structural defects such as pores and cracks of SiC/SiC composites. In this study, the examination sensitivity limit of UT method based o...

Yun-Seok Shin; Yi-Hyun Park; Tatsuya Hinoki

2010-01-01T23:59:59.000Z

350

Graphite Materials Testing in the ATR for Lifetime Management of Magnox Reactors  

SciTech Connect

A major feature of the Magnox gas cooled reactor design is the graphite core, which acts as the moderator but also provides the physical structure for fuel, control rods, instrumentation and coolant gas channels. The lifetime of a graphite core is dependent upon two principal aging processes: irradiation damage and radiolytic oxidation. Irradiation damage from fast neutrons creates lattice defects leading to changes in physical and mechanical properties and the accumulation of stresses. Radiolytic oxidation is caused by the reaction of oxidizing species from the carbon dioxide coolant gas with the graphite, these species being produced by gamma radiation. Radiolytic oxidation reduces the density and hence the moderating capability of the graphite, but also reduces strength affecting the integrity of core components. In order to manage continued operation over the planned lifetimes of their power stations, BNFL needed to extend their database of the effects of these two phenomena on the ir graphite cores through an irradiation experiment. This paper will discuss the background, purpose, and the processes taken and planned (i.e. post irradiation examination) to ensure meaningful data on the graphite core material is obtained from the irradiation experiment.

Grover, S.B. (INEEL); Metcalfe, M.P. (BNFL, United Kingdom)

2002-03-14T23:59:59.000Z

351

Graphite Materials Testing in the ATR for Lifetime Management of Magnox Reactors  

SciTech Connect

A major feature of the Magnox gas cooled reactor design is the graphite core, which acts as the moderator but also provides the physical structure for fuel, control rods, instrumentation and coolant gas channels. The lifetime of a graphite core is dependent upon two principal aging processes: irradiation damage and radiolytic oxidation. Irradiation damage from fast neutrons creates lattice defects leading to changes in physical and mechanical properties and the accumulation of stresses. Radiolytic oxidation is caused by the reaction of oxidizing species from the carbon dioxide coolant gas with the graphite, these species being produced by gamma radiation. Radiolytic oxidation reduces the density and hence the moderating capability of the graphite, but also reduces strength affecting the integrity of core components. In order to manage continued operation over the planned lifetimes of their power stations, BNFL needed to extend their database of the effects of these two phenomena on their graphite cores through an irradiation experiment. This paper will discuss the background, purpose, and the processes taken and planned (i.e. post irradiation examination) to ensure meaningful data on the graphite core material is obtained from the irradiation experiment.

Grover, Stanley Blaine; Metcalfe, M. P.

2002-04-01T23:59:59.000Z

352

Special Analysis for the Disposal of the Idaho National Laboratory Unirradiated Light Water Breeder Reactor Rods and Pellets Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada  

SciTech Connect

The purpose of this special analysis (SA) is to determine if the Idaho National Laboratory (INL) Unirradiated Light Water Breeder Reactor (LWBR) Rods and Pellets waste stream (INEL103597TR2, Revision 2) is suitable for disposal by shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS). The INL Unirradiated LWBR Rods and Pellets waste stream consists of 24 containers with unirradiated fabricated rods and pellets composed of uranium oxide (UO2) and thorium oxide (ThO2) fuel in zirconium cladding. The INL Unirradiated LWBR Rods and Pellets waste stream requires an SA because the 229Th, 230Th, 232U, 233U, and 234U activity concentrations exceed the Nevada National Security Site (NNSS) Waste Acceptance Criteria (WAC) Action Levels.

Shott, Gregory [NSTec

2014-08-31T23:59:59.000Z

353

A FEASIBILITY AND OPTIMIZATION STUDY TO DETERMINE COOLING TIME AND BURNUP OF ADVANCED TEST REACTOR FUELS USING A NONDESTRUCTIVE TECHNIQUE  

SciTech Connect

The goal of this study presented is to determine the best available non-destructive technique necessary to collect validation data as well as to determine burn-up and cooling time of the fuel elements onsite at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads3 to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements non-destructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed was used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however in order to enhance the quality of the spectra collected using this scintillator a deconvolution method was developed. Following the development of the deconvolution method for ATR applications the technique was tested using one-isotope, multi-isotope and fuel simulated sources. Burnup calibrations were perfomed using convoluted and deconvoluted data. The calibrations results showed burnup prediction by this method improves using deconvolution. The final stage of the deconvolution method development was to perform an irradiation experiment in order to create a surrogate fuel source to test the deconvolution method using experimental data. A conceptual design of the fuel scan system is path forward using the rugged LaBr3 detector in an above the water configuration and deconvolution algorithms.

Jorge Navarro

2013-12-01T23:59:59.000Z

354

Cold test plan for the Old Hydrofracture Facility tank contents removal project, Oak Ridge National Laboratory, Oak Ridge, Tennessee  

SciTech Connect

This Old Hydrofracture Facility (OHF) Tanks Contents Removal Project Cold Test Plan describes the activities to be conducted during the cold test of the OHF sluicing and pumping system at the Tank Technology Cold Test Facility (TTCTF). The TTCTF is located at the Robotics and Process Systems Complex at the Oak Ridge National Laboratory (ORNL). The cold test will demonstrate performance of the pumping and sluicing system, fine-tune operating instructions, and train the personnel in the actual work to be performed. After completion of the cold test a Technical Memorandum will be prepared documenting completion of the cold test, and the equipment will be relocated to the OHF site.

NONE

1997-11-01T23:59:59.000Z

355

Testing the Reactor and Gallium Anomalies with Intense (Anti)Neutrino Emitters  

E-Print Network (OSTI)

Several observed anomalies in neutrino oscillation data could be explained by a hypothetical fourth neutrino separated from the three standard neutrinos by a squared mass difference of a few 0.1 eV$^2$ or more. This hypothesis can be tested with MCi neutrino electron capture sources ($^{51}$Cr) or kCi antineutrino $\\beta$-source ($^{144}$Ce) deployed inside or next to a large low background neutrino detector. In particular, the compact size of this source coupled with the localization of the interaction vertex lead to an oscillating pattern in event spatial (and possibly energy) distributions that would unambiguously determine neutrino mass differences and mixing angles.

Th. Lasserre

2012-09-23T23:59:59.000Z

356

Type A verification report for the high flux beam reactor stack and grounds, Brookhaven National Laboratory, Upton, New York  

SciTech Connect

The U.S. Department of Energy (DOE) Order 458.1 requires independent verification (IV) of DOE cleanup projects (DOE 2011). The Oak Ridge Institute for Science and Education (ORISE) has been designated as the responsible organization for IV of the High Flux Beam Reactor (HFBR) Stack and Grounds area at Brookhaven National Laboratory (BNL) in Upton, New York. The IV evaluation may consist of an in-process inspection with document and data reviews (Type A Verification) or a confirmatory survey of the site (Type B Verification). DOE and ORISE determined that a Type A verification of the documents and data for the HFBR Stack and Grounds: Survey Units (SU) 6, 7, and 8 was appropriate based on the initial survey unit classification, the walkover surveys, and the final analytical results provided by the Brookhaven Science Associates (BSA). The HFBR Stack and Grounds surveys began in June 2011 and were completed in September 2011. Survey activities by BSA included gamma walkover scans and sampling of the as-left soils in accordance with the BSA Work Procedure (BNL 2010a). The Field Sampling Plan - Stack and Remaining HFBR Outside Areas (FSP) stated that gamma walk-over surveys would be conducted with a bare sodium iodide (NaI) detector, and a collimated detector would be used to check areas with elevated count rates to locate the source of the high readings (BNL 2010b). BSA used the Mult- Agency Radiation Survey and Site Investigation Manual (MARSSIM) principles for determining the classifications of each survey unit. Therefore, SUs 6 and 7 were identified as Class 1 and SU 8 was deemed Class 2 (BNL 2010b). Gamma walkover surveys of SUs 6, 7, and 8 were completed using a 2?2 NaI detector coupled to a data-logger with a global positioning system (GPS). The 100% scan surveys conducted prior to the final status survey (FSS) sampling identified two general soil areas and two isolated soil locations with elevated radioactivity. The general areas of elevated activity identified were investigated further with a collimated NaI detector. The uncollimated average gamma count rate was less than 15,000 counts per minute (cpm) for the SU 6, 7, and 8 composite area (BNL 2011a). Elevated count rates were observed in portions of each survey unit. The general areas of elevated counts near the Building 801 ventilation and operations and the entry to the Stack were determined to be directly related to the radioactive processes in those structures. To compensate for this radioactive shine, a collimated or shielded detector was used to lower the background count rate (BNL 2011b and c). This allowed the surveyor(s) to distinguish between background and actual radioactive contamination. Collimated gamma survey count rates in these shine affected areas were below 9,000 cpm (BNL 2011a). The average background count rate of 7,500 cpm was reported by BSA for uncollimated NaI detectors (BNL 2011d). The average collimated background ranged from 4,500-6,500 cpm in the westernmost part of SU 8 and from 2,000-3,500 cpm in all other areas (BNL 2011e). Based on these data, no further investigations were necessary for these general areas. SU 8 was the only survey unit that exhibited verified elevated radioactivity levels. The first of two isolated locations of elevated radioactivity had an uncollimated direct measurement of 50,000 cpm with an area background of 7,500 cpm (BNL 2011f). The second small area exhibiting elevated radiation levels was identified at a depth of 6 inches from the surface. The maximum reported count rate of 28,000 cpm was observed during scanning (BNL 2011g). The affected areas were remediated, and the contaminated soils were placed in an intermodal container for disposal. BSA's post-remediation walkover surveys were expanded to include a 10-foot radius around the excavated locations, and it was determined that further investigation was not required for these areas (BNL 2011 f and g). The post-remediation soil samples were collected and analyzed with onsite gamma spectroscopy equipment. These samples were also included with the FSS s

Harpenau, Evan M.

2012-01-13T23:59:59.000Z

357

TYPE A VERIFICATION REPORT FOR THE HIGH FLUX BEAM REACTOR STACK AND GROUNDS, BROOKHAVEN NATIONAL LABORATORY, UPTON, NEW YORK DCN 5098-SR-08-0  

SciTech Connect

The U.S. Department of Energy (DOE) Order 458.1 requires independent verification (IV) of DOE cleanup projects (DOE 2011). The Oak Ridge Institute for Science and Education (ORISE) has been designated as the responsible organization for IV of the High Flux Beam Reactor (HFBR) Stack and Grounds area at Brookhaven National Laboratory (BNL) in Upton, New York. The IV evaluation may consist of an in-process inspection with document and data reviews (Type A Verification) or a confirmatory survey of the site (Type B Verification). DOE and ORISE determined that a Type A verification of the documents and data for the HFBR Stack and Grounds: Survey Units (SU) 6, 7, and 8 was appropriate based on the initial survey unit classification, the walkover surveys, and the final analytical results provided by the Brookhaven Science Associates (BSA).

Evan Harpenau

2011-11-30T23:59:59.000Z

358

Project Management Plan for the Idaho National Engineering Laboratory Waste Isolation Pilot Plant Experimental Test Program  

SciTech Connect

EG&G Idaho, Inc. and Argonne National Laboratory-West (ANL-W) are participating in the Idaho National Engineering Laboratory`s (INEL`s) Waste Isolation Pilot Plant (WIPP) Experimental Test Program (WETP). The purpose of the INEL WET is to provide chemical, physical, and radiochemical data on transuranic (TRU) waste to be stored at WIPP. The waste characterization data collected will be used to support the WIPP Performance Assessment (PA), development of the disposal No-Migration Variance Petition (NMVP), and to support the WIPP disposal decision. The PA is an analysis required by the Code of Federal Regulations (CFR), Title 40, Part 191 (40 CFR 191), which identifies the processes and events that may affect the disposal system (WIPP) and examines the effects of those processes and events on the performance of WIPP. A NMVP is required for the WIPP by 40 CFR 268 in order to dispose of land disposal restriction (LDR) mixed TRU waste in WIPP. It is anticipated that the detailed Resource Conservation and Recovery Act (RCRA) waste characterization data of all INEL retrievably-stored TRU waste to be stored in WIPP will be required for the NMVP. Waste characterization requirements for PA and RCRA may not necessarily be identical. Waste characterization requirements for the PA will be defined by Sandia National Laboratories. The requirements for RCRA are defined in 40 CFR 268, WIPP RCRA Part B Application Waste Analysis Plan (WAP), and WIPP Waste Characterization Program Plan (WWCP). This Project Management Plan (PMP) addresses only the characterization of the contact handled (CH) TRU waste at the INEL. This document will address all work in which EG&G Idaho is responsible concerning the INEL WETP. Even though EG&G Idaho has no responsibility for the work that ANL-W is performing, EG&G Idaho will keep a current status and provide a project coordination effort with ANL-W to ensure that the INEL, as a whole, is effectively and efficiently completing the requirements for WETP.

Connolly, M.J.; Sayer, D.L.

1993-11-01T23:59:59.000Z

359

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

SciTech Connect

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

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

2013-07-01T23:59:59.000Z

360

CRAD, Fire Protection - Oak Ridge National Laboratory High Flux...  

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

Fire Protection - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Fire Protection - Oak Ridge National Laboratory High Flux Isotope Reactor February 2006 A section of...

Note: This page contains sample records for the topic "national reactor testing" 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

CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope...  

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

Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G...

362

CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope...  

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

Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of...

363

CRAD, Emergency Management - Oak Ridge National Laboratory High...  

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

Emergency Management - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Emergency Management - Oak Ridge National Laboratory High Flux Isotope Reactor...

364

CRAD, Emergency Management - Oak Ridge National Laboratory High...  

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

Emergency Management - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Emergency Management - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A...

365

CRAD, Nuclear Safety - Oak Ridge National Laboratory High Flux...  

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

Nuclear Safety - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Nuclear Safety - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of...

366

CRAD, Safety Basis - Oak Ridge National Laboratory High Flux...  

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

Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Safety Basis - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR February 2007 A...

367

CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope...  

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

Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR...

368

CRAD, Radiological Controls - Oak Ridge National Laboratory High...  

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

Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Radiological Controls - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C...

369

CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope...  

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

Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR...

370

CRAD, Management- Oak Ridge National Laboratory High Flux Isotope...  

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

Management- Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Management- Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C...

371

CRAD, Conduct of Operations - Oak Ridge National Laboratory High...  

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

Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Conduct of Operations - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR February...

372

CRAD, Occupational Safety & Health - Oak Ridge National Laboratory...  

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

Occupational Safety & Health - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Occupational Safety & Health - Oak Ridge National Laboratory High Flux Isotope Reactor...

373

Generation -IV Reactor Concepts  

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

Generation-IV Reactor Concepts Generation-IV Reactor Concepts Thomas H. Fanning Argonne National Laboratory 9700 South Cass Avenue Argonne, Illinois 60439, USA The Generation-IV International Forum (GIF) is a multi-national research and development (R&D) collaboration. The GIF pursues the development of advanced, next generation reactor technology with goals to improve: a) sustainability (effective fuel utilization and minimization of waste) b) economics (competitiveness with respect to other energy sources) c) safety and reliability (e.g., no need for offsite emergency response), and d) proliferation resistance and physical protection The GIF Technology Roadmap exercise selected six generic systems for further study: the Gas- cooled Fast Reactor (GFR), the Lead-cooled Fast Reactor (LFR), the Molten Salt Reactor (MSR),

374

Nevada Test Site National Emission Standards for Hazardous Air Pollutants Calendar Year 2008  

SciTech Connect

The Nevada Test Site (NTS) is operated by the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office. From 1951 through 1992, the NTS was the continental testing location for U.S. nuclear weapons. The release of radionuclides from NTS activities has been monitored since the initiation of atmospheric testing. Limitation to under-ground detonations after 1962 greatly reduced radiation exposure to the public surrounding the NTS. After nuclear testing ended in 1992, NTS radiation monitoring focused on detecting airborne radionuclides from historically contaminated soils. These radionuclides are derived from re-suspension of soil (primarily by winds) and emission of tritium-contaminated soil moisture through evapotranspiration. Low amounts of tritium were also emitted to air at the North Las Vegas Facility (NLVF), an NTS support complex in the city of North Las Vegas. To protect the public from harmful levels of man-made radiation, the Clean Air Act, National Emission Standards for Hazardous Air Pollutants (NESHAP) (Title 40 Code of Federal Regulations [CFR] Part 61 Subpart H) (CFR, 2008a) limits the release of radioactivity from a U.S. Department of Energy facility (e.g., the NTS) to 10 millirem per year (mrem/yr) effective dose equivalent to any member of the public. This limit does not include radiation not related to NTS activities. Unrelated doses could come from naturally occurring radioactive elements or from other man-made sources such as medical treatments. The NTS demonstrates compliance with the NESHAP limit by using environmental measurements of radionuclide air concentrations at critical receptor locations. This method was approved by the U.S. Environmental Protection Agency for use on the NTS in 2001 and has been the sole method used since 2005. Six locations on the NTS have been established to act as critical receptor locations to demonstrate compliance with the NESHAP limit. These locations are actually pseudo-critical receptor stations, because no member of the public actually resides at these onsite locations. Compliance is demonstrated if the measured annual average concentration of each detected radionuclide at each of these locations is less than the NESHAP Concentration Levels (CLs) for Environmental Compliance listed in 40 CFR 61, Appendix E, Table 2 (CFR, 2008a). At any one location, if multiple radionuclides are detected then compliance with NESHAP is demonstrated when the sum of the fractions (determined by dividing each radionuclide's concentration by its CL and then adding the fractions together) is less than 1.0. In 2008, the potential dose from radiological emissions to air, from both current and past NTS activities, at onsite compliance monitoring stations was a maximum of 1.9 mrem/yr; well below the 10 mrem/yr dose limit. Air sampling data collected at all six pseudo-critical receptor stations had average concentrations of radioactivity that were a fraction of the CL values listed in Table 2 in Appendix E of 40 CFR 61 (CFR, 2008a). Concentrations ranged from less than 1 percent to a maximum of 19 percent of the allowed NESHAP limit. Because the nearest member of the public resides approximately 20 kilometers (12 miles) from the NTS boundary, concentrations at this location would be only a small fraction of that measured on the NTS. Potential dose to the public from NLVF was also very low at 0.00006 mrem/yr; more than 160,000 times lower than the 10 mrem/yr limit.

Ronald Warren and Robert F. Grossman

2009-06-30T23:59:59.000Z

375

Testing of a 50-kW wind-diesel hybrid system at the National Wind Technology Center  

SciTech Connect

To further the development of commercial hybrid power systems, the National Renewable Energy Laboratory (NREL), in collaboration with the New World Village Power Corporation (NWVP), tested a NWVP 50-kW wind-diesel hybrid system connected to a 15/50 Atlantic Orient Corporation (AOC) wind turbine. Testing was conducted from October 1995 through March 1996 at the National Wind Technology Center (NWTC). A main objective of the testing was to better understand the application of wind turbines to weak grids typical of small villages. Performance results contained in this paper include component characterization, such as power conversion losses for the rotary converter systems and battery round trip efficiencies. In addition, systems operation over this period is discussed with special attention given to dynamic issues. Finally, future plans for continued testing and research are discussed.

Corbus, D.A.; Green, J.; Allderdice, A.; Rand, K.; Bianchi, J. [National Renewable Energy Lab., Golden, CO (United States); Linton, E. [New World Village Power, Waitsfield, VT (United States)

1996-07-01T23:59:59.000Z

376

Nuclear Reactor (atomic reactor)  

Science Journals Connector (OSTI)

A nuclear reactor splits Uranium or Plutonium nuclei, and the...235 is fissionable but more than 99% of the naturally occurring Uranium is U238 that makes enrichment mandatory. In some reactors U238 and Thorium23...

2008-01-01T23:59:59.000Z

377

Program management plan for the Molten Salt Reactor Experiment Remediation Project at Oak Ridge National Laboratory, Oak Ridge, Tennessee  

SciTech Connect

The primary mission of the Molten Salt Reactor Experiment (MSRE) Remediation Project is to effectively implement the risk-reduction strategies and technical plans to stabilize and prevent further migration of uranium within the MSRE facility, remove the uranium and fuel salts from the system, and dispose of the fuel and flush salts by storage in appropriate depositories to bring the facility to a surveillance and maintenance condition before decontamination and decommissioning. This Project Management Plan (PMP) for the MSRE Remediation Project details project purpose; technical objectives, milestones, and cost objectives; work plan; work breakdown structure (WBS); schedule; management organization and responsibilities; project management performance measurement planning, and control; conduct of operations; configuration management; environmental, safety, and health compliance; quality assurance; operational readiness reviews; and training.

NONE

1996-09-01T23:59:59.000Z

378

Underground Test Area Fiscal Year 2013 Annual Quality Assurance Report Nevada National Security Site, Nevada, Revision 0  

SciTech Connect

This report is required by the Underground Test Area (UGTA) Quality Assurance Plan (QAP) and identifies the UGTA quality assurance (QA) activities for fiscal year (FY) 2013. All UGTA organizations—U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Field Office (NNSA/NFO); Desert Research Institute (DRI); Lawrence Livermore National Laboratory (LLNL); Los Alamos National Laboratory (LANL); Navarro-Intera, LLC (N-I); National Security Technologies, LLC (NSTec); and the U.S. Geological Survey (USGS)—conducted QA activities in FY 2013. The activities included conducting assessments, identifying findings and completing corrective actions, evaluating laboratory performance, and publishing documents. In addition, integrated UGTA required reading and corrective action tracking was instituted.

Krenzien, Susan; Marutzky, Sam

2014-01-01T23:59:59.000Z

379

Evaluation of Cavity Collapse and Surface Crater Formation for Selected Lawrence Livermore National Laboratory Underground Nuclear Tests - 2010  

SciTech Connect

This report evaluates collapse evolution for selected Lawrence Livermore National Laboratory (LLNL) underground nuclear tests at the Nevada National Security Site (NNSS, formerly called the Nevada Test Site). The work is being done at the request of Navarro-Interra LLC, and supports environmental restoration efforts by the Department of Energy, National Nuclear Security Administration for the Nevada Site Office. Safety decisions must be made before a surface crater area, or potential surface crater area, can be reentered for any work. Our statements on cavity collapse and surface crater formation are input into their safety decisions. These statements do not include the effects of erosion that may modify the surface collapse craters over time. They also do not address possible radiation dangers that may be present. Subject matter experts from the LLNL Containment Program who had been active in weapons testing activities performed these evaluations. Information used included drilling and hole construction, emplacement and stemming, timing and sequence of the selected test and nearby tests, geology, yield, depth of burial, collapse times, surface crater sizes, cavity and crater volume estimations, and ground motion. Both classified and unclassified data were reviewed. Various amounts of information are available for these tests, depending on their age and other associated activities. Lack of data can hamper evaluations and introduce uncertainty. We make no attempt to quantify this uncertainty.

Pawloski, G A

2011-01-03T23:59:59.000Z

380

Strengthening the nuclear-reactor fuel cycle against proliferation  

SciTech Connect

Argonne National Laboratory (ANL) conducts several research programs that serve to reduce the risks of fissile-material diversion from the nuclear-reactor fuel cycle. The objectives are to provide economical and efficient neutron or power generation with the minimum of inherent risks, and to further minimize risks by utilizing sophisticated techniques to detect attempts at material diversion. This paper will discuss the Reduced Enrichment Research and Test Reactor (RERTR) Program, the Isotope Correlation Technique (ICT), and Proliferation-Resistant Closed-Cycle Reactors. The first two are sponsored by the DOE Office of Arms Control and Nonproliferation.

Travelli, A.; Snelgrove, J.; Persiani, P. [Argonne National Lab., IL (United States). Arms Control and Nonproliferation Program

1992-12-31T23:59:59.000Z

Note: This page contains sample records for the topic "national reactor testing" 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

Code validation with EBR-II test data  

SciTech Connect

An extensive system of computer codes is used at Argonne National Laboratory to analyze whole-plant transient behavior of the Experimental Breeder Reactor 2. Three of these codes, NATDEMO/HOTCHAN, SASSYS, and DSNP have been validated with data from reactor transient tests. The validated codes are the foundation of safety analyses and pretest predictions for the continuing design improvements and experimental programs in EBR-II, and are also valuable tools for the analysis of innovative reactor designs.

Herzog, J.P.; Chang, L.K.; Dean, E.M.; Feldman, E.E.; Hill, D.J.; Mohr, D.; Planchon, H.P.

1992-01-01T23:59:59.000Z

382

Code validation with EBR-II test data  

SciTech Connect

An extensive system of computer codes is used at Argonne National Laboratory to analyze whole-plant transient behavior of the Experimental Breeder Reactor 2. Three of these codes, NATDEMO/HOTCHAN, SASSYS, and DSNP have been validated with data from reactor transient tests. The validated codes are the foundation of safety analyses and pretest predictions for the continuing design improvements and experimental programs in EBR-II, and are also valuable tools for the analysis of innovative reactor designs.

Herzog, J.P.; Chang, L.K.; Dean, E.M.; Feldman, E.E.; Hill, D.J.; Mohr, D.; Planchon, H.P.

1992-07-01T23:59:59.000Z

383

Code validation with EBR-II test data  

SciTech Connect

An extensive system of computer codes is used at Argonne National Laboratory to analyze whole-plant transient behavior of the Experiment Breeder Reactor 2. Three of these codes, NATDEMO/HOTCHAN, SASSYS, and DSNP have been validated with data from reactor transient tests. The validated codes are the foundation of safety analyses and pretest predictions for the continuing design improvements and experimental programs in EBR-2, and are also valuable tools for the analysis of innovative reactor designs. 29 refs., 6 figs.

Herzog, J.P.; Chang, L.K.; Dean, E.M.; Feldman, E.E.; Hill, D.J.; Mohr, D.; Planchon, H.P.

1991-01-01T23:59:59.000Z

384

Powerline Conductor Accelerated Testing Facility (PCAT) The Powerline Conductor Accelerated Testing facility (PCAT) at Oak Ridge National  

E-Print Network (OSTI)

as simultaneous measuring of conductor tension, sag, and environmental conditions (e.g., wind, solar, ambient environmental conditions. The tests provide both the manufacturer and utilities with conductor performance data under accelerated field-like operating conditions. These tests short-circuit the need for utilities

385

Design, Assembly, and Testing of the Neutron Imaging Lens for the National Ignition Facility  

SciTech Connect

The National Ignition Facility will begin testing DT fuel capsules yielding greater than 10^13 neutrons during 2010. Neutron imaging is an important diagnostic for understanding capsule behavior. Neutrons are imaged at a scintillator after passing through a pinhole. The pixelated, 160-mm square scintillator is made up of ¼ mm diameter rods 50 mm long. Shielding and distance (28 m) are used to preserve the recording diagnostic hardware. Neutron imaging is light starved. We designed a large nine-element collecting lens to relay as much scintillator light as reasonable onto a 75 mm gated microchannel plate (MCP) intensifier. The image from the intensifier’s phosphor passes through a fiber taper onto a CCD camera for digital storage. Alignment of the pinhole and tilting of the scintillator is performed before the relay lens and MCP can be aligned. Careful tilting of the scintillator is done so that each neutron only passes through one rod (no crosstalk allowed). The 3.2 ns decay time scintillator emits light in the deep blue, requiring special glass materials. The glass within the lens housing weighs 26 lbs, with the largest element being 7.7 inches in diameter. The distance between the scintillator and the MCP is only 27 inches. The scintillator emits light with 0.56 NA and the lens collects light at 0.15 NA. Thus, the MCP collects only 7% of the available light. Baffling the stray light is a major concern in the design of the optics. Glass cost considerations, tolerancing, and alignment of this lens system will be discussed.

Malone, Robert M; Fatherley, Valerie E; Frogget, Brent C; Grim, Gary P; Kaufman, Morris I; McGillivray, Kevin D; Oertel, John A; Palagi, Martin J; Skarda, William K; Tibbitts, Aric; Wilde, Carl H

2010-09-01T23:59:59.000Z

386

03/01/2006 09:51 AMLoading "People's Daily Online --Chinese experimental thermonuclear reactor on discharge test in July" Page 1 of 1http://english.people.com.cn/200603/01/print20060301_247035.html  

E-Print Network (OSTI)

03/01/2006 09:51 AMLoading "People's Daily Online -- Chinese experimental thermonuclear reactor experimental thermonuclear reactor on discharge test in July China's new generation experimental Tokamak fusion and the former Soviet Union launched a 10 billion- euro ambitious plan, the International Thermonuclear

387

Thermionic Reactor Design Studies  

SciTech Connect

During the 1960's and early 70's the author performed extensive design studies, analyses, and tests aimed at thermionic reactor concepts that differed significantly from those pursued by other investigators. Those studies, like most others under Atomic Energy Commission (AEC and DOE) and the National Aeronautics and Space Administration (NASA) sponsorship, were terminated in the early 1970's. Some of this work was previously published, but much of it was never made available in the open literature. U.S. interest in thermionic reactors resumed in the early 80's, and was greatly intensified by reports about Soviet ground and flight tests in the late 80's. This recent interest resulted in renewed U.S. thermionic reactor development programs, primarily under Department of Defense (DOD) and Department of Energy (DOE) sponsorship. Since most current investigators have not had an opportunity to study all of the author's previous work, a review of the highlights of that work may be of value to them. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling. Where the author's concepts differed from the later Topaz-2 design was in the relative location of the emitter and the collector. Placing the fueled emitter on the outside of the cylindrical diodes permits much higher axial conductances to reduce ohmic losses in the electrodes of full-core-height diodes. Moreover, placing the fuel on the outside of the diode makes possible reactors with much higher fuel volume fractions, which enable power-flattened fast reactors scalable to very low power levels without the need for life-limiting hydride moderators or the use of efficiency-limiting driver fuel. In addition, with the fuel on the outside its swelling does not increase the emitter diameter or reduce the interelectrode gap. This should permit long lifetimes even with closer spacings, which can significantly improve the system efficiences. This was confirmed by coupled neutronic, thermal, thermionic, and electrical system analyses - some of which are presented in this paper - and by subsequent experiments. A companion paper presented next describes the fabrication and testing of full-scale converter elements, both fueled and unfueled, and summarizes the test results obtained. There is a duplicate copy in the file.

Schock, Alfred

1994-06-01T23:59:59.000Z

388

In situ vitrification application to buried waste: Final report of intermediate field tests at Idaho National Engineering Laboratory  

SciTech Connect

This report describes two in situ vitrification field tests conducted on simulated buried waste pits during June and July 1990 at the Idaho National Engineering Laboratory. In situ vitrification, an emerging technology for in place conversion of contaminated soils into a durable glass and crystalline waste form, is being investigated as a potential remediation technology for buried waste. The overall objective of the two tests was to access the general suitability of the process to remediate waste structures representative of buried waste found at Idaho National Engineering Laboratory. In particular, these tests, as part of a treatability study, were designed to provide essential information on the field performance of the process under conditions of significant combustible and metal wastes and to test a newly developed electrode feed technology. The tests were successfully completed, and the electrode feed technology successfully processed the high metal content waste. Test results indicate the process is a feasible technology for application to buried waste. 33 refs., 109 figs., 39 tabs.

Callow, R.A.; Weidner, J.R.; Loehr, C.A.; Bates, S.O. (EG and G Idaho, Inc., Idaho Falls, ID (United States)); Thompson, L.E.; McGrail, B.P. (Pacific Northwest Lab., Richland, WA (United States))

1991-08-01T23:59:59.000Z

389

National SCADA Test Bed- Enhancing control systems security in the energy sector (September 2009)  

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

Improving the security of energy control systems has become a national priority. Since the mid-1990’s, security experts have become increasingly concerned about the threat of malicious cyber...

390

RERTR program activities related to the development and application of new LEU fuels. [Reduced Enrichment Research and Test Reactor; low-enriched uranium  

SciTech Connect

The statue of the U.S. Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a brief outline of RERTR Program objectives and goals, program accomplishments are discussed with emphasis on the development, demonstration and application of new LEU fuels. Most program activities have proceeded as planned, and a combination of two silicide fuels (U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al) holds excellent promise for achieving the long-term program goals. Current plans and schedules project the uranium density of qualified RERTR fuels for plate-type reactors to grow by approximately 1 g U/cm/sup 3/ each year, from the current 1.7 g U/cm/sup 3/ to the 7.0 g U/cm/sup 3/ which will be reached in late 1988. The technical needs of research and test reactors for HEU exports are also forecasted to undergo a gradual but dramatic decline in the coming years.

Travelli, A.

1983-01-01T23:59:59.000Z

391

Design and testing of a boron carbide capsule for spectral-tailoring in mixed-spectrum reactors  

SciTech Connect

A boron carbide capsule has been designed and used for spectral-tailoring experiments at the TRIGA reactor at Washington State Univ.. Irradiations were conducted in pulsed mode and in continuous operation for up to 4 h. A cadmium cover was used to reduce thermal heating. The neutron spectrum calculated with the Monte Carlo N-particle transport code was found to be in good agreement with reactor dosimetry measurements using the STAY'SL computer code. The neutron spectrum resembles that of a fast reactor. The design of a capsule using boron carbide fully enriched in {sup 10}B shows that it is possible to produce a neutron spectrum similar to that of {sup 235}U fission. (authors)

Greenwood, L.R.; Wittman, R. [Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Pierson, B.P. [Univ. of Michigan, Ann Arbor, MI 48109 (United States); Metz, L.A.; Payne, R.; Finn, E.C.; Friese, J.I. [Pacific Northwest National Laboratory, Richland, WA 99352 (United States)

2011-07-01T23:59:59.000Z

392

Development, Implementation, and Testing of Fault Detection Strategies on the National Wind Technology Center's Controls Advanced Research Turbines  

SciTech Connect

The National Renewable Energy Laboratory's National Wind Technology Center dedicates two 600 kW turbines for advanced control systems research. A fault detection system for both turbines has been developed, analyzed, and improved across years of experiments to protect the turbines as each new controller is tested. Analysis of field data and ongoing fault detection strategy improvements have resulted in a system of sensors, fault definitions, and detection strategies that have thus far been effective at protecting the turbines. In this paper, we document this fault detection system and provide field data illustrating its operation while detecting a range of failures. In some cases, we discuss the refinement process over time as fault detection strategies were improved. The purpose of this article is to share field experience obtained during the development and field testing of the existing fault detection system, and to offer a possible baseline for comparison with more advanced turbine fault detection controllers.

Johnson, K. E.; Fleming, P. A.

2011-06-01T23:59:59.000Z

393

Concepts and Tests for the Remote-Controlled Dismantling of the Biological Shield and Form work of the KNK Reactor - 13425  

SciTech Connect

The compact sodium-cooled nuclear reactor facility Karlsruhe (KNK), a prototype Fast Breeder, is currently in an advanced stage of dismantling. Complete dismantling is based on 10 partial licensing steps. In the frame of the 9. decommissioning permit, which is currently ongoing, the dismantling of the biological shield is foreseen. The biological shield consists of heavy reinforced concrete with built-in steel fitments, such as form-work of the reactor tank, pipe sleeves, ventilation channels, and measuring devices. Due to the activation of the inner part of the biological shield, dismantling has to be done remote-controlled. During a comprehensive basic design phase a practical dismantling strategy was developed. Necessary equipment and tools were defined. Preliminary tests revealed that hot wire plasma cutting is the most favorable cutting technology due to the geometrical boundary conditions, the varying distance between cutter and material, and the heavy concrete behind the steel form-work. The cutting devices will be operated remotely via a carrier system with an industrial manipulator. The carrier system has expandable claws to adjust to the varying diameter of the reactor shaft during dismantling progress. For design approval of this prototype development, interaction between manipulator and hot wire plasma cutting was tested in a real configuration. For the demolition of the concrete structure, an excavator with appropriate tools, such as a hydraulic hammer, was selected. Other mechanical cutting devices, such as a grinder or rope saw, were eliminated because of concrete containing steel spheres added to increase the shielding factor of the heavy concrete. Dismantling of the biological shield will be done in a ring-wise manner due to static reasons. During the demolition process, the excavator is positioned on its tripod in three concrete recesses made prior to the dismantling of the separate concrete rings. The excavator and the manipulator carrier system will be operated alternately. Main boundary condition for all the newly designed equipment is the decommissioning housing of limited space within the reactor building containment. To allow for a continuous removal of the concrete rubble, an additional opening on the lowest level of the reactor shaft will be made. All equipment and the interaction of the tools have to be tested before use in the controlled area. Therefore a full-scale model of the biological shield will be provided in a mock-up. The tests will be performed in early 2014. The dismantling of the biological shield is scheduled for 2015. (authors)

Neff, Sylvia; Graf, Anja; Petrick, Holger; Rothschmitt, Stefan [WAK Rueckbau- und Entsorgungs- GmbH, P.O.Box 12 63, 76339 Eggenstein- Leopoldshafen (Germany)] [WAK Rueckbau- und Entsorgungs- GmbH, P.O.Box 12 63, 76339 Eggenstein- Leopoldshafen (Germany); Klute, Stefan [Siempelkamp Nukleartechnik GmbH, Am Taubenfeld 25/1, 69123 Heidelberg (Germany)] [Siempelkamp Nukleartechnik GmbH, Am Taubenfeld 25/1, 69123 Heidelberg (Germany); Stanke, Dieter [Siempelkamp NIS Ingenieurgesellschaft mbH, Industriestrasse 13, 63755 Alzenau (Germany)] [Siempelkamp NIS Ingenieurgesellschaft mbH, Industriestrasse 13, 63755 Alzenau (Germany)

2013-07-01T23:59:59.000Z

394

Validation of the U.S. NRC coupled code system TRITON/TRACE/PARCS with the special power excursion reactor test III (SPERT III)  

SciTech Connect

The Special Power Excursion Reactor Test III (SPERT III) was a series of reactivity insertion experiments conducted in the 1950's. This paper describes the validation of the U.S. NRC Coupled Code system TRITON/PARCS/TRACE to simulate reactivity insertion accidents (RIA) by using several of the SPERT III tests. The work here used the SPERT III E-core configuration tests in which the RIA was initiated by ejecting a control rod. The resulting super-prompt reactivity excursion and negative reactivity feedback produced the familiar bell shaped power increase and decrease. The energy deposition during such a power peak has important safety consequences and provides validation basis for core coupled multi-physics codes. The transients of five separate tests are used to benchmark the PARCS/TRACE coupled code. The models were thoroughly validated using the original experiment documentation. (authors)

Wang, R. C.; Xu, Y.; Downar, T. [Dept. of Nuclear Engineering and Radiological Sciences, Univ. of Michigan, Ann Arbor, MI 48104 (United States); Hudson, N. [RES Div., U.S. NRC, Rockville, MD (United States)

2012-07-01T23:59:59.000Z

395

Reactor Safety Planning for Prometheus Project, for Naval Reactors Information  

SciTech Connect

The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

P. Delmolino

2005-05-06T23:59:59.000Z

396

Modeling and analysis framework for core damage propagation during flow-blockage-initiated accidents in the Advanced Neutron Source Reactor at Oak Ridge National Laboratory  

SciTech Connect

This paper describes modeling and analysis to evaluate the extent of core damage during flow blockage events in the Advanced Neutron Source (ANS) reactor planned to be built at the Oak Ridge National Laboratory (ORNL). Damage propagation is postulated to occur from thermal conduction between damaged and undamaged plates due to direct thermal contact. Such direct thermal contact may occur because of fuel plate swelling during fission product vapor release or plate buckling. Complex phenomena of damage propagation were modeled using a one-dimensional heat transfer model. A scoping study was conducted to learn what parameters are important for core damage propagation, and to obtain initial estimates of core melt mass for addressing recriticality and steam explosion events. The study included investigating the effects of the plate contact area, the convective heat transfer coefficient, thermal conductivity upon fuel swelling, and the initial temperature of the plate being contacted by the damaged plate. Also, the side support plates were modeled to account for their effects on damage propagation. The results provide useful insights into how various uncertain parameters affect damage propagation.

Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V.

1995-09-01T23:59:59.000Z

397

Improving Building Energy Simulation Programs Through Diagnostic Testing (Fact Sheet), NREL Highlights, Research & Development, NREL (National Renewable Energy Laboratory)  

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

test procedure evaluates quality and accuracy of energy test procedure evaluates quality and accuracy of energy analysis tools for the residential building retrofit market. Reducing the energy use of existing homes in the United States offers significant energy-saving opportunities, which can be identified through building simulation software tools that calculate optimal packages of efficiency measures. To improve the accuracy of energy analysis for residential buildings, the National Renewable Energy Laboratory's (NREL) Buildings Research team developed the Building Energy Simulation Test for Existing Homes (BESTEST-EX), a method for diagnosing and correcting errors in building energy audit software and calibration procedures. BESTEST-EX consists of building physics and utility bill calibration test cases, which soft-

398

EA-1965: Florida Atlantic University Southeast National Marine Renewable Energy Center’s Offshore Marine Hydrokinetic Technology Testing Project, Florida  

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

The Department of Energy (DOE), through its Wind and Water Power Technologies Office (WWPTO), is proposing to provide federal funding to Florida Atlantic University’s South-East National Marine Renewable Energy Center (FAU SNMREC) to support the at sea testing of FAU SNMREC’s experimental current generation turbine and the deployment and operation of their Small-Scale Ocean Current Turbine Test Berth, sited on the outer continental shelf (OCS) in waters off the coast of Ft Lauderdale, Florida. SNMREC would demonstrate the test berth site readiness by testing their pilot-scale experimental ocean current turbine unit at that location. The Bureau of Ocean Energy Management (BOEM) conducted an Environmental Assessment to analyze the impacts associated with leasing OCS lands to FAU SNMREC, per their jurisdictional responsibilities under the Outer Continental Shelf Lands Act. DOE was a cooperating agency in this process and based on the EA, DOE issued a Finding of No Significant Impact.

399

Testing of a 50-kW Wind-Diesel Hybrid System at the National Wind Technology Center  

SciTech Connect

In remote off-grid villages and communities, a reliable power source is important in improving the local quality of life. Villages often use a diesel generator for their power, but fuel can be expensive and maintenance burdensome. Including a wind turbine in a diesel system can reduce fuel consumption and lower maintenance, thereby reducing energy costs. However, integrating the various components of a wind-diesel system, including wind turbine, power conversion system, and battery storage (if applicable), is a challenging task. To further the development of commercial hybrid power systems, the National Renewable Energy Laboratory (NREL), in collaboration with the New World Village Power Corporation (NWVP), tested a NWVP 50-kW wind-diesel hybrid system connected to a 15/50 Atlantic Orient Corporation (AOC) wind turbine. Testing was conducted from October 1995 through March 1996 at the National Wind Technology Center (NWTC). A main objective of the testing was to better understand the application of wind turbines to weak grids typical of small villages. Performance results contained in this report include component characterization, such as power conversion losses for the rotary converter system and battery round trip efficiencies. In addition, system operation over the test period is discussed with special attention given to dynamic issues. Finally, future plans for continued testing and research are discussed.

Corbus, D. A.; Green, H. J.; Allderdice, A.; Rand, K.; Bianchi, J.; Linton, E.

1996-07-01T23:59:59.000Z

400

Independent Oversight Review, Oak Ridge National Laboratory- January 2013  

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

Review of the Oak Ridge National Laboratory High Flux Isotope Reactor Implementation Verification Review Processes

Note: This page contains sample records for the topic "national reactor testing" 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

Mirror Advanced Reactor Study (MARS) final report summary  

SciTech Connect

The Mirror Advanced Reactor Study (MARS) has resulted in an overview of a first-generation tandem mirror reactor. The central cell fusion plasma is self-sustained by alpha heating (ignition), while electron-cyclotron resonance heating and negative ion beams maintain the electrostatic confining potentials in the end plugs. Plug injection power is reduced by the use of high-field choke coils and thermal barriers, concepts to be tested in the Tandem Mirror Experiment-Upgrade (TMX-U) and Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory.

Henning, C.D.; Logan, B.G.; Carlson, G.A.; Perkins, L.J.; Werner, R.W.; Gordon, J.D.; Parmer, J.F.; Bilton, J.R.; Glancy, J.E.; Kulcinski, G.L.

1983-12-08T23:59:59.000Z

402

Advanced Reactor Technologies | Department of Energy  

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

Advanced Reactor Advanced Reactor Technologies Advanced Reactor Technologies Advanced Reactor Technologies Advanced Reactor Technologies The Office of Advanced Reactor Technologies (ART) sponsors research, development and deployment (RD&D) activities through its Next Generation Nuclear Plant (NGNP), Advanced Reactor Concepts (ARC), and Advanced Small Modular Reactor (aSMR) programs to promote safety, technical, economical, and environmental advancements of innovative Generation IV nuclear energy technologies. The Office of Nuclear Energy (NE) will pursue these advancements through RD&D activities at the Department of Energy (DOE) national laboratories and U.S. universities, as well as through collaboration with industry and international partners. These activities will focus on advancing scientific

403

Experimental investigations on decay heat removal in advanced nuclear reactors using single heater rod test facility: Air alone in the annular gap  

SciTech Connect

During a loss of coolant accident in nuclear reactors, radiation heat transfer accounts for a significant amount of the total heat transfer in the fuel bundle. In case of heavy water moderator nuclear reactors, the decay heat of a fuel bundle enclosed in the pressure tube and outer concentric calandria tube can be transferred to the moderator. Radiation heat transfer plays a significant role in removal of decay heat from the fuel rods to the moderator, which is available outside the calandria tube. A single heater rod test facility is designed and fabricated as a part of preliminary investigations. The objective is to anticipate the capability of moderator to remove decay heat, from the reactor core, generated after shut down. The present paper focuses mainly on the role of moderator in removal of decay heat, for situation with air alone in the annular gap of pressure tube and calandria tube. It is seen that the naturally aspirated air is capable of removing the heat generated in the system compared to the standstill air or stagnant water situations. It is also seen that the flowing moderator is capable of removing a greater fraction of heat generated by the heater rod compared to a stagnant pool of boiling moderator. (author)

Bopche, Santosh B.; Sridharan, Arunkumar [Department of Mechanical Engineering, Indian Institute of Technology Bombay, Powai, Mumbai 400 076 (India)

2010-11-15T23:59:59.000Z

404

Status of Cryogenic System for Spallation Neutron Source's Superconducting Radiofrequency Test Facility at Oak Ridge National Lab  

SciTech Connect

Spallation Neutron Source (SNS) at Oak Ridge National Lab (ORNL) is building an independent cryogenic system for its Superconducting Radiofrequency Test Facility (SRFTF). The scope of the system is to support the SNS cryomodule test and cavity test at 2-K (using vacuum pump) and 4.5K for the maintenance purpose and Power Upgrade Project of SNS, and to provide the part of the cooling power needed to backup the current CHL to keep Linac at 4.5-K during CHL maintenance period in the future. The system is constructed in multiple phases. The first phase is to construct an independent 4K helium refrigeration system with helium Dewar and distribution box as load interface. It is schedule to be commissioned in 2013. Here we report the concept design of the system and the status of the first phase of this project.

Xu, Ting [ORNL; Casagrande, Fabio [ORNL; Ganni, Venkatarao [ORNL; Knudsen, Peter N [ORNL; Strong, William Herb [ORNL

2011-01-01T23:59:59.000Z

405

Enforcement Documents - Idaho National Laboratory | Department of Energy  

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

Idaho National Laboratory Idaho National Laboratory Enforcement Documents - Idaho National Laboratory October 4, 2012 Preliminary Notice of Violation, Battelle Energy Alliance, LLC - NEA-2012-01 Issued to Battelle Energy Alliance, LLC related to an elevated extremity dose at the Hot Fuel Examination Facility and the plutonium contamination at the Zero Power Physics Reactor facility. January 13, 2012 Consent Order, Battelle Energy Alliance - NCO-2012-01 Issued to Battelle Energy Alliance, LLC, related to the Reactor Coolant Draining Event at the Idaho National Laboratory Advanced Test Reactor October 13, 2011 Consent Order, URS Energy & Construction, Inc. - NCO-2011-02 Issued to URS Energy & Construction, Inc., related to Quality Improvement and Work Control Program Deficiencies associated with Construction of the

406

Gas Test Loop Facilities Alternatives Assessment Report Rev 1  

SciTech Connect

An important task in the Gas Test Loop (GTL) conceptual design was to determine the best facility to serve as host for this apparatus, which will allow fast-flux neutron testing in an existing nuclear facility. A survey was undertaken of domestic and foreign nuclear reactors and accelerator facilities to arrive at that determination. Two major research reactors in the U.S. were considered in detail, the Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR), each with sufficient power to attain the required neutron fluxes. HFIR routinely operates near its design power limit of 100 MW. ATR has traditionally operated at less than half its design power limit of 250 MW. Both of these reactors should be available for at least the next 30 years. The other major U.S. research reactor, the Missouri University Research Reactor, does not have sufficient power to reach the required neutron flux nor do the smaller research reactors. Of the foreign reactors investigated, BOR-60 is perhaps the most attractive. Monju and BN 600 are power reactors for their respective electrical grids. Although the Joyo reactor is vigorously campaigning for customers, local laws regarding transport of radioactive material mean it would be very difficult to retrieve test articles from either Japanese reactor for post irradiation examination. PHENIX is scheduled to close in 2008 and is fully booked until then. FBTR is limited to domestic (Indian) users only. Data quality is often suspect in Russia. The only accelerator seriously considered was the Fuel and Material Test Station (FMTS) currently proposed for operation at Los Alamos National Laboratory. The neutron spectrum in FMTS is similar to that found in a fast reactor, but it has a pronounced high-energy tail that is atypical of fast fission reactor spectra. First irradiation in the FMTS is being contemplated for 2008. Detailed review of these facilities resulted in the recommendation that the ATR would be the best host for the GTL.

William J. Skerjanc; William F. Skerjanc

2005-07-01T23:59:59.000Z

407

Sandia National Laboratories is a multi-program laboratory operated by Sandia Co  

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

Sandia National Laboratories is a multi-program laboratory operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Sandia National Laboratories is a multi-program laboratory operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin company, for the U.S. Department of Energy's National Nuclear Security Administration. With main facilities in Albuquerque, N.M., and Livermore, Calif., Sandia has major R&D responsibilities in national security, energy and environmental technologies and economic competitiveness. Annular Core Research Reactor Facility At the Annular Core Research Reactor (ACRR) facility, Sandia researchers can subject various test objects to a

408

Cultural Resource Assessment of the Test Area North Demolition Landfill at the Idaho National Engineering and Environmental Laboratory  

SciTech Connect

The proposed new demolition landfill at Test Area North on the Idaho National Engineering and Environmental Laboratory (INEEL) will support ongoing demolition and decontamination within the facilities on the north end of the INEEL. In June of 2003, the INEEL Cultural Resource Management Office conducted archival searches, field surveys, and coordination with the Shoshone-Bannock Tribes to identify all cultural resources that might be adversely affected by the project and to provide recommendations to protect those listed or eligible for listing on the National Register of Historic Places. These investigations showed that landfill construction and operation would affect two significant cultural resources. This report outlines protective measures to ensure that these effects are not adverse.

Brenda R. Pace

2003-07-01T23:59:59.000Z

409

In-service Inspection Ultrasonic Testing of Reactor Pressure Vessel Welds for Assessing Flaw Density and Size Distribution per 10 CFR 50.61a, Alternate Fracture Toughness Requirements  

SciTech Connect

Pressurized thermal shock (PTS) events are system transients in a pressurized water reactor (PWR) in which there is a rapid operating temperature cool-down that results in cold vessel temperatures with or without repressurization of the vessel. The rapid cooling of the inside surface of the reactor pressure vessel (RPV) causes thermal stresses that can combine with stresses caused by high pressure. The aggregate effect of these stresses is an increase in the potential for fracture if a pre-existing flaw is present in a material susceptible to brittle failure. The ferritic, low alloy steel of the reactor vessel beltline adjacent to the core, where neutron radiation gradually embrittles the material over the lifetime of the plant, can be susceptible to brittle fracture. The PTS rule, described in the Code of Federal Regulations, Title 10, Section 50.61 (§50.61), “Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events,” adopted on July 23, 1985, establishes screening criteria to ensure that the potential for a reactor vessel to fail due to a PTS event is deemed to be acceptably low. The U.S. Nuclear Regulatory Commission (NRC) completed a research program that concluded that the risk of through-wall cracking due to a PTS event is much lower than previously estimated. The NRC subsequently developed a rule, §50.61a, published on January 4, 2010, entitled “Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events” (75 FR 13). Use of the new rule by licensees is optional. The §50.61a rule differs from §50.61 in that it requires licensees who choose to follow this alternate method to analyze the results from periodic volumetric examinations required by the ASME Code, Section XI, Rules for Inservice Inspection (ISI) of Nuclear Power Plants. These analyses are intended to determine if the actual flaw density and size distribution in the licensee’s reactor vessel beltline welds are bounded by the flaw density and size distribution values used in the PTS technical basis. Under a contract with the NRC, Pacific Northwest National Laboratory (PNNL) has been working on a program to assess the ability of current inservice inspection (ISI)-ultrasonic testing (UT) techniques, as qualified through ASME Code, Appendix VIII, Supplements 4 and 6, to detect small fabrication or inservice-induced flaws located in RPV welds and adjacent base materials. As part of this effort, the investigators have pursued an evaluation, based on the available information, of the capability of UT to provide flaw density/distribution inputs for making RPV weld assessments in accordance with §50.61a. This paper presents the results of an evaluation of data from the 1993 Browns Ferry Nuclear Plant, Unit 3, Spirit of Appendix VIII reactor vessel examination, a comparison of the flaw density/distribution from this data with the distribution in §50.61a, possible reasons for differences, and plans and recommendations for further work in this area.

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

2012-09-17T23:59:59.000Z

410

Testing Small Wind Turbines at the National Renewable Energy Laboratory (NREL) (Poster)  

SciTech Connect

WindPower 2008 conference sponsored by AWEA held in Houston, Texas on June 1-4, 2008. This poster describes four small wind electric systems that were tested to IEC and AWEA standards at NREL's NWTC.

Sinclair, K.; Bowen, A.

2008-06-01T23:59:59.000Z

411

Laser damage testing of small optics for the National Ignition Facility  

Science Journals Connector (OSTI)

A damage test procedure was established for optical components that have large incident beam footprints. The procedure was applied on coated samples for a high powered 1053 nm, 3-ns...

Chow, Robert; Runkel, Mike; Taylor, John R

412

Laser damage testing of small optics for the National Ignition Facility  

Science Journals Connector (OSTI)

A damage test procedure was established for optical components that have large incident beam footprints. The procedure was applied on coated samples for a high-powered 1053-nm, 3-ns...

Chow, Robert; Runkel, Mike; Taylor, John R

2005-01-01T23:59:59.000Z

413

Sandia National Laboratories: solar  

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

Interactive Tour Operated by Sandia National Laboratories for the U.S. Department of Energy (DOE), the National Solar Thermal Test Facility (NSTTF) is the only test facility...

414

CRAD, Configuration Management - Oak Ridge National Laboratory...  

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

Configuration Management - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Configuration Management - Oak Ridge National Laboratory High Flux Isotope...

415

Design and Testing of a 10B4C Capsule for Spectral-Tailoring in Mixed-Spectrum Reactors  

SciTech Connect

A boron carbide capsule highly enriched in 10B has been designed and used for spectral-tailoring experiments at the TRIGA reactor at Washington State University. New experiments show that enriching the boron to 96% B-10 results in additional absorption of neutrons in the resonance region thereby producing a neutron spectrum that is much closer to a pure 235U fission spectrum. A cadmium outer cover was used to reduce thermal heating. The neutron spectrum calculated with MCNP was found to be in very good agreement with measured activation rates from neutron fluence monitors.

Greenwood, Lawrence R.; Wittman, Richard S.; Metz, Lori A.; Finn, Erin C.; Friese, Judah I.

2014-04-11T23:59:59.000Z

416

Policing the peace: How nations will monitor a nuclear test ban  

SciTech Connect

Steve R. Bratt`s belt started chirping as if on cue, precisely as he was boasting about a new global watchdog system for detecting nuclear tests. Bratt, a seismologist with the US Department of Defense, retrieved the beeper from his hip and studied it for a few seconds. {open_quotes}I`ve got an alert. It`s from Lop Nor. Lop Nor is the Chinese test site,{close_quotes} he explained. Two stations in a worldwide network of seismometers had just picked up vibrations emanating from central Asia, near China`s known nuclear facility. The shock was small, about magnitude 3.5. In bomb equivalents, it would correspond to less than a half kiloton explosion. In this case, however, Bratt suspected the alert was just a minor earthquake. Timing provided an important clue: The shock had originated at 12:19 Greenwich Mean Time, which is not the kind of round, on-the-hour time that countries usually choose for performing a major weapon test. Seismic analysts would later confirm Bratt`s hunch when they determined that the Chinese vibrations actually originated at an unlikely place to stage a test, hundreds of kilometers away from the Lop Nor site. The impromptu demonstration nonetheless made a good advertisement for the new international monitoring system-an ever-figilant network of sensors strung around the globe, listening, sniffing, and waiting. This article describes the problems involved in test ban monitoring and the possibilities for solving them.

Monastersky, R.

1996-05-11T23:59:59.000Z

417

Building Energy Simulation Test for Existing Homes (BESTEST-EX) (Presentation), NREL (National Renewable Energy Laboratory)  

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

Building Energy Simulation Test Building Energy Simulation Test for Existing Homes (BESTEST-EX) Ron Judkoff Joel Neymark Ben Polly Updated: December 2011 NREL/PR-5500-53701 2 Goals of NREL Analysis Accuracy R&D * Provide industry with the tools and technical information needed to improve the accuracy and consistency of analysis methods * Reduce the risks associated with purchasing, financing, and selling energy efficiency upgrades * Enhance software and input collection methods considering impacts on accuracy, cost, and time of energy assessments 3 BESTEST-EX Goals * Test software predictions of retrofit energy savings in existing homes * Ensure building physics calculations and utility bill calibration procedures perform up to a minimum standard

418

Reactor Safety Research Programs Quarterly Report October - December 1981  

SciTech Connect

This document summarizes the work performed by Pacific Northwest laboratory (PNL) from October 1 through December 31, 1981, for the Division of Accident Evaluation, U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where serviceinduced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and post accident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

Edler, S. K.

1982-03-01T23:59:59.000Z

419

NREL Tests Dehumidifiers, Defines Simplified Simulation Model (Fact Sheet), NREL Highlights, Research & Development, NREL (National Renewable Energy Laboratory)  

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

residential dehumidifiers residential dehumidifiers results in practical performance curves for use in whole-building simulation tools. Dehumidifiers remove moisture from a home's indoor environment, thereby increasing occupant comfort, improving air quality, and reducing the likelihood of mold, rot, and dust mites. To help energy professionals more easily evaluate this technology for the market, National Renewable Energy Laboratory (NREL) researchers tested the efficiency and capacity of a variety of dehumidifiers and developed a generalized approach to simulate any residential dehumidifier. The test results and modeling method are documented in a new report. Typically, dehumidifiers are only rated at a single temperature and humidity, so rating data alone cannot determine whether a product will meet the moisture removal

420

Natural Convection Shutdown Heat Removal Test Facility (NSTF)  

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

Natural Convection Natural Convection Shutdown Heat Removal Test Facility Scaling Basis Full Scale Half Scale NSTF Argonne National Laboratory's Natural Convection Shutdown Heat Removal Test Facility (NSTF) - one of the world's largest facilities for ex-vessel passive decay heat removal testing-confirms the performance of reactor cavity cooling systems (RCCS) and similar passive confinement or containment decay heat removal systems in modern Small Modular Reactors. Originally built to aid in the development of General Electric's Power Reactor Innovative Small Module (PRISM) Reactor Vessel Auxiliary Cooling System (RVACS), the NSTF has a long history of providing confirmatory data for the airside of the RVACS. Argonne National Laboratory's NSTF is a state-of-the-art, large-scale facility for evaluating performance

Note: This page contains sample records for the topic "national reactor testing" 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

Plasma–wall interaction in laser inertial fusion reactors: novel proposals for radiation tests of first wall materials  

Science Journals Connector (OSTI)

Dry-wall laser inertial fusion (LIF) chambers will have to withstand strong bursts of fast charged particles which will deposit tens of kJ m?2 and implant more than 1018 particles m?2 in a few microseconds at a repetition rate of some Hz. Large chamber dimensions and resistant plasma-facing materials must be combined to guarantee the chamber performance as long as possible under the expected threats: heating, fatigue, cracking, formation of defects, retention of light species, swelling and erosion. Current and novel radiation resistant materials for the first wall need to be validated under realistic conditions. However, at present there is a lack of facilities which can reproduce such ion environments.This contribution proposes the use of ultra-intense lasers and high-intense pulsed ion beams (HIPIB) to recreate the plasma conditions in LIF reactors. By target normal sheath acceleration, ultra-intense lasers can generate very short and energetic ion pulses with a spectral distribution similar to that of the inertial fusion ion bursts, suitable to validate fusion materials and to investigate the barely known propagation of those bursts through background plasmas/gases present in the reactor chamber. HIPIB technologies, initially developed for inertial fusion driver systems, provide huge intensity pulses which meet the irradiation conditions expected in the first wall of LIF chambers and thus can be used for the validation of materials too.

J Alvarez Ruiz; A Rivera; K Mima; D Garoz; R Gonzalez-Arrabal; N Gordillo; J Fuchs; K Tanaka; I Fernández; F Briones; J Perlado

2012-01-01T23:59:59.000Z

422

DEVELOPMENT OF A HIGH BRIGHTNESS ELECTRON GUN FOR THE ACCELERATOR TEST FACILITY AT BROOKHAVEN NATIONAL LABORATORY*  

E-Print Network (OSTI)

954 DEVELOPMENT OF A HIGH BRIGHTNESS ELECTRON GUN FOR THE ACCELERATOR TEST FACILITY AT BROOKHAVEN, New York 11973 and K. McDonald Princeton [Jniversity Abstract An electron gun utilizing a radio). Here we report on the de;$n of the electron gun which will provide r.f. bunches of up to 10 electrons

McDonald, Kirk

423

Solar Thermal Reactor Materials Characterization  

SciTech Connect

Current research into hydrogen production through high temperature metal oxide water splitting cycles has created a need for robust high temperature materials. Such cycles are further enhanced by the use of concentrated solar energy as a power source. However, samples subjected to concentrated solar radiation exhibited lifetimes much shorter than expected. Characterization of the power and flux distributions representative of the High Flux Solar Furnace(HFSF) at the National Renewable Energy Laboratory(NREL) were compared to ray trace modeling of the facility. In addition, samples of candidate reactor materials were thermally cycled at the HFSF and tensile failure testing was performed to quantify material degradation. Thermal cycling tests have been completed on super alloy Haynes 214 samples and results indicate that maximum temperature plays a significant role in reduction of strength. The number of cycles was too small to establish long term failure trends for this material due to the high ductility of the material.

Lichty, P. R.; Scott, A. M.; Perkins, C. M.; Bingham, C.; Weimer, A. W.

2008-03-01T23:59:59.000Z

424

Development of advanced strain diagnostic techniques for reactor environments.  

SciTech Connect

The following research is operated as a Laboratory Directed Research and Development (LDRD) initiative at Sandia National Laboratories. The long-term goals of the program include sophisticated diagnostics of advanced fuels testing for nuclear reactors for the Department of Energy (DOE) Gen IV program, with the future capability to provide real-time measurement of strain in fuel rod cladding during operation in situ at any research or power reactor in the United States. By quantifying the stress and strain in fuel rods, it is possible to significantly improve fuel rod design, and consequently, to improve the performance and lifetime of the cladding. During the past year of this program, two sets of experiments were performed: small-scale tests to ensure reliability of the gages, and reactor pulse experiments involving the most viable samples in the Annulated Core Research Reactor (ACRR), located onsite at Sandia. Strain measurement techniques that can provide useful data in the extreme environment of a nuclear reactor core are needed to characterize nuclear fuel rods. This report documents the progression of solutions to this issue that were explored for feasibility in FY12 at Sandia National Laboratories, Albuquerque, NM.

Fleming, Darryn D.; Holschuh, Thomas Vernon,; Miller, Timothy J.; Hall, Aaron Christopher; Urrea, David Anthony,; Parma, Edward J.,

2013-02-01T23:59:59.000Z

425

A versatile elevated-pressure reactor combined with an ultrahigh vacuum surface setup for efficient testing of model and powder catalysts under clean gas-phase conditions  

SciTech Connect

A small-volume reaction cell for catalytic or photocatalytic testing of solid materials at pressures up to 1000 Torr has been coupled to a surface-science setup used for standard sample preparation and characterization under ultrahigh vacuum (UHV). The reactor and sample holder designs allow easy sample transfer from/to the UHV chamber, and investigation of both planar and small amounts of powder catalysts under the same conditions. The sample is heated with an infrared laser beam and its temperature is measured with a compact pyrometer. Combined in a regulation loop, this system ensures fast and accurate temperature control as well as clean heating. The reaction products are automatically sampled and analyzed by mass spectrometry and/or gas chromatography (GC). Unlike previous systems, our GC apparatus does not use a recirculation loop and allows working in clean conditions at pressures as low as 1 Torr while detecting partial pressures smaller than 10{sup ?4} Torr. The efficiency and versatility of the reactor are demonstrated in the study of two catalytic systems: butadiene hydrogenation on Pd(100) and CO oxidation over an AuRh/TiO{sub 2} powder catalyst.

Morfin, Franck; Piccolo, Laurent [Institut de recherches sur la catalyse et l'environnement de Lyon (IRCELYON), UMR 5256 CNRS and Université Lyon 1, 2 avenue Albert Einstein, F-69626 Villeurbanne (France)] [Institut de recherches sur la catalyse et l'environnement de Lyon (IRCELYON), UMR 5256 CNRS and Université Lyon 1, 2 avenue Albert Einstein, F-69626 Villeurbanne (France)

2013-09-15T23:59:59.000Z

426

Idaho National Laboratory Speakers Bureau  

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

Working with INL Community Outreach Visitor Information Calendar of Events ATR National Scientific User Facility Center for Advanced Energy Studies Light Water Reactor...

427

Idaho National Laboratory Safety Education  

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

Working with INL Community Outreach Visitor Information Calendar of Events ATR National Scientific User Facility Center for Advanced Energy Studies Light Water Reactor...

428

Idaho National Laboratory Other Programs  

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

Working with INL Community Outreach Visitor Information Calendar of Events ATR National Scientific User Facility Center for Advanced Energy Studies Light Water Reactor...

429

GLOBAL THREAT REDUCTION INITIATIVE REACTOR CONVERSION PROGRAM: STATUS AND CURRENT PLANS  

SciTech Connect

The U.S. Department of Energy’s National Nuclear Security Administration (NNSA) Reactor Conversion Program supports the minimization, and to the extent possible, elimination of the use of high enriched uranium (HEU) in civilian nuclear applications by working to convert research and test reactors and radioisotope production processes to the use of low enriched uranium (LEU). The Reactor Conversion Program is a technical pillar of the NNSA Global Threat Reduction Initiative (GTRI) which is a key organization for implementing U.S. HEU minimization policy and works to reduce and protect vulnerable nuclear and radiological material domestically and abroad.

Staples, Parrish A.; Leach, Wayne; Lacey, Jennifer M.

2009-10-07T23:59:59.000Z

430

Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop  

SciTech Connect

The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370/sup 0/C.

McCulloch, R.W.; MacPherson, R.E.

1983-03-01T23:59:59.000Z

431

Maximizing Thermal Efficiency and Optimizing Energy Management (Fact Sheet), Thermal Test Facility (TTF), NREL (National Renewable Energy Laboratory)  

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

Maximizing Thermal Efficiency and Maximizing Thermal Efficiency and Optimizing Energy Management Scientists at this living laboratory develop optimal solutions for managing energy flows within buildings and transportation systems. The built environment is stressing the utility grid to a greater degree than ever before. Growing demand for electric vehicles, space conditioning, and plug loads presents a critical opportunity for more effective energy management and development of efficiency technologies. Researchers at the Thermal Test Facility (TTF) on the campus of the U.S. Department of Energy's National Renewable Energy Laboratory (NREL) in Golden, Colorado, are addressing this opportunity. Through analysis of efficient heating, ventilating, and air conditioning (HVAC) strategies, automated home energy management (AHEM), and energy storage systems,

432

Production-scale LLW and RMW solidification system operational testing at Argonne National Laboratory-East (ANL-E)  

SciTech Connect

Argonne National Laboratory-East (ANL-E) has begun production-scale testing of a low-level waste and radioactive mixed waste solidification system. This system will be used to treat low-level and mixed radioactive waste to meet land burial requirements. The system can use any of several types of solidification media, including a chemically bonded phosphate ceramic developed by ANL-E scientists. The final waste product will consist of a solidified mass in a standard 208-liter drum. The system uses commercial equipment and incorporates several unique process control features to ensure proper treatment. This paper will discuss the waste types requiring treatment, the system configuration, and operation results for these waste streams.

Wescott, J.; Wagh, A.; Singh, D. [Argonne National Lab., IL (United States); Nelson, R. [Sargent and Lundy, Chicago, IL (United States); No, H. [H and P, Inc., Vienna, VA (United States)

1997-04-01T23:59:59.000Z

433

Reactor Thermal-Hydraulics  

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

Thermal-Hydraulics Thermal-Hydraulics Dr. Tanju Sofu, Argonne National Laboratory In a power reactor, the energy produced in fission reaction manifests itself as heat to be removed by a coolant and utilized in a thermodynamic energy conversion cycle to produce electricity. A simplified schematic of a typical nuclear power plant is shown in the diagram below. Primary coolant loop Steam Reactor Heat exchanger Primary pump Secondary pump Condenser Turbine Water Although this process is essentially the same as in any other steam plant configuration, the power density in a nuclear reactor core is typically four orders of magnitude higher than a fossil fueled plant and therefore it poses significant heat transfer challenges. Maximum power that can be obtained from a nuclear reactor is often limited by the

434

Results of the Sandia National Laboratories MOSAIK cask drop test program  

SciTech Connect

There has been a significant international effort over the past ten years to qualify structural materials for construction of radioactive material (RAM) transportation casks. As total life cycle cost analyses argue the necessity for more efficient casks, new candidate structural materials are evaluated relative to the historically accepted austenitic stainless steels. New candidate cask containment materials include ferritic steels, ductile iron, depleted uranium, and titanium. Another material, borated stainless steel is being considered for structural cask internals because of its neutron absorption properties. The mechanical performance of the borated stainless steels is a function of the boron content and metallurgical processing conditions. A separate paper in this symposium (Stephens et al. 1992) deals with the properties of a range of borated stainless steels. A major technical issue involved with the qualification of afl these candidate materials is that they may, under certain combinations of mechanical and environmental loading, fail in a brittle fashion. Such a failure would of course not be acceptable for a RAM transport cask involved in an accident. The cask designer must assure cask owners, regulators as well as the general public that the cask will not undergo brittle fracture for all regulatory loading conditions. This paper summarizes the drop tests that were conducted using the MOSAIK casks to verify the fracture mechanics cask design approach and to demonstrate that ductile iron could be subjected to severe loading conditions without failing in a brittle manner.

Sorenson, K.; Salzbrenner, R.; Wellman, G.; Bobbe, J.

1991-01-01T23:59:59.000Z

435

National Environmental Policy Act Hazards Assessment for the TREAT Alternative  

SciTech Connect

This document provides an assessment of hazards as required by the National Environmental Policy Act for the alternative of restarting the reactor at the Transient Reactor Test (TREAT) facility by the Resumption of Transient Testing Program. Potential hazards have been identified and screening level calculations have been conducted to provide estimates of unmitigated dose consequences that could be incurred through this alternative. Consequences considered include those related to use of the TREAT Reactor, experiment assembly handling, and combined events involving both the reactor and experiments. In addition, potential safety structures, systems, and components for processes associated with operating TREAT and onsite handling of nuclear fuels and experiments are listed. If this alternative is selected, a safety basis will be prepared in accordance with 10 CFR 830, “Nuclear Safety Management,” Subpart B, “Safety Basis Requirements.”

Boyd D. Christensen; Annette L. Schafer

2014-02-01T23:59:59.000Z

436

Consortium for Advanced Simulation of Light Water Reactors (CASL...  

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

behavior in AP1000 reactor core Test run signals emergence of the next generation in nuclear power reactor analysis tools OAK RIDGE, Tenn., Feb. 18, 2014 - Scientists and...

437

E-Print Network 3.0 - almaty wwr-k reactor Sample Search Results  

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

High Flux Isotope Reactor Center for Nanophase Materials Sciences... International Thermonuclear Experimental Reactor Center for Computational Sciences National Security 0 0 61 1...

438

EA-0907: Idaho National Engineering Laboratory Sewer System Upgrade  

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

7: Idaho National Engineering Laboratory Sewer System Upgrade 7: Idaho National Engineering Laboratory Sewer System Upgrade Project, Idaho Falls, Idaho EA-0907: Idaho National Engineering Laboratory Sewer System Upgrade Project, Idaho Falls, Idaho SUMMARY This EA evaluates the environmental impacts of a proposal to upgrade the Sewer System at the U.S. Department of Energy's Idaho National Engineering Laboratory (INEL) near Idaho Falls, Idaho. The proposed action would include activities conducted at the Central Facilities Area, Test Reactor Area, and the Containment Test Facility at the Test Area North at INEL. PUBLIC COMMENT OPPORTUNITIES None available at this time. DOCUMENTS AVAILABLE FOR DOWNLOAD April 1, 1994 EA-0907: Finding of No Significant Impact Idaho National Engineering Laboratory Sewer System Upgrade Project

439

Idaho National Laboratory DOE-NE's National Nuclear Capability-  

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

4-2023 4-2023 Idaho National Laboratory DOE-NE's National Nuclear Capability- Developing and Maintaining the INL Infrastructure TEN-YEAR SITE PLAN DOE/ID-11474 Final June 2012 Sustainable INL continues to exceed DOE goals for reduction in the use of petroleum fuels - running its entire bus fleet on biodiesel while converting 75% of its light-duty fleet to E85 fuel. The Energy Systems Laboratory (ESL), slated for completion this year, will be a state-of-the-art laboratory with high-bay lab space where leading bioenergy feedstock processing, advanced battery testing, and hybrid energy systems integration research will be conducted. The Advanced Test Reactor is the world's most advanced nuclear research capability - crucial to (1) the ongoing development of safe, efficient

440

MONOLITHIC FUEL FABRICATION PROCESS DEVELOPMENT AT THE IDAHO NATIONAL LABORATORY  

SciTech Connect

Within the Reduced Enrichment for Research and Test Reactors (RERTR) program directed by the US Department of Energy (DOE), UMo fuel-foils are being developed in an effort to realize high density monolithic fuel plates for use in high-flux research and test reactors. Namely, targeted are reactors that are not amenable to Low Enriched Uranium (LEU) fuel conversion via utilization of high density dispersion-based fuels, i.e. 8-9 gU/cc. LEU conversion of reactors having a need for >8-9 gU/cc fuel density will only be possible by way of monolithic fuel forms. The UMo fuel foils under development afford fuel meat density of ~16 gU/cc and thus have the potential to facilitate LEU conversions without any significant reactor-performance penalty. Two primary challenges have been established with respect to UMo monolithic fuel development; namely, fuel element fabrication and in-reactor fuel element performance. Both issues are being addressed concurrently at the Idaho National Laboratory. An overview is provided of the ongoing monolithic UMo fuel development effort at the Idaho National Laboratory (INL); including development of complex/graded fuel foils. Fabrication processes to be discussed include: UMo alloying and casting, foil fabrication via hot rolling, fuel-clad interlayer application via co-rolling and thermal spray processes, clad bonding via Hot Isostatic Pressing (HIP) and Friction Bonding (FB), and fuel plate finishing.

Glenn A. Moore; Francine J. Rice; Nicolas E. Woolstenhulme; W. David SwanK; DeLon C. Haggard; Jan-Fong Jue; Blair H. Park; Steven E. Steffler; N. Pat Hallinan; Michael D. Chapple; Douglas E. Burkes

2008-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "national reactor testing" 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.


441

Radiation effects on reactor pressure vessel supports  

SciTech Connect

The purpose of this report is to present the findings from the work done in accordance with the Task Action Plan developed to resolve the Nuclear Regulatory Commission (NRC) Generic Safety Issue No. 15, (GSI-15). GSI-15 was established to evaluate the potential for low-temperature, low-flux-level neutron irradiation to embrittle reactor pressure vessel (RPV) supports to the point of compromising plant safety. An evaluation of surveillance samples from the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) had suggested that some materials used for RPV supports in pressurized-water reactors could exhibit higher than expected embrittlement rates. However, further tests designed to evaluate the applicability of the HFIR data to reactor RPV supports under operating conditions led to the conclusion that RPV supports could be evaluated using traditional method. It was found that the unique HFIR radiation environment allowed the gamma radiation to contribute significantly to the embrittlement. The shielding provided by the thick steel RPV shell ensures that degradation of RPV supports from gamma irradiation is improbable or minimal. The findings reported herein were used, in part, as the basis for technical resolution of the issue.

Johnson, R.E. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering Technology; Lipinski, R.E. [Idaho National Engineering Lab., Rockville, MD (United States)

1996-05-01T23:59:59.000Z

442

Nuclear Energy Research Brookhaven National  

E-Print Network (OSTI)

Nuclear Energy Research Brookhaven National Laboratory William C. Horak, Chair Nuclear Science and Technology Department #12;BNL Nuclear Energy Research Brookhaven Graphite Research Reactor - 1948 National&T Department #12;Nuclear Energy Today 435 Operable Power Reactors, 12% electrical generation (100 in US, 19

Ohta, Shigemi

443

R t f N l C t T ti Di i GReport of Nuclear Component Testing Discussion Group National Spherical Torus ProgramNational Spherical Torus Program  

E-Print Network (OSTI)

Office of Science R t f N l C t T ti Di i GReport of Nuclear Component Testing Discussion Group nuclear technology, SG1 leader UCLA DOE contact: Eckstrand, Steve, OFES #12;Nuclear Component Testing (NCT) aims to complement ITER mission and fill many DEMO R&D gaps · Mission of the Nuclear Component Testing

444

Sandia National Laboratories: ACEC  

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

ACEC Sandia Solar Energy Test System Cited in National Engineering Competition On May 16, 2013, in Concentrating Solar Power, Energy, Energy Storage, Facilities, National Solar...

445

NEUTRON RADIOGRAPHY (NRAD) REACTOR 64-ELEMENT CORE UPGRADE  

SciTech Connect

The neutron radiography (NRAD) reactor is a 250 kW TRIGA (registered) (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The interim critical configuration developed during the core upgrade, which contains only 62 fuel elements, has been evaluated as an acceptable benchmark experiment. The final 64-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (approximately +/-1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

John D. Bess

2014-03-01T23:59:59.000Z

446

CRAD, Safety Basis - Oak Ridge National Laboratory High Flux...  

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

Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Safety Basis - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G...

447

Sorbent Testing For Solidification of Process Waste streams from the Radiochemical Engineering Development Center at Oak Ridge National Laboratory  

SciTech Connect

The U.S. Department of Energy (DOE) tasked MSE Technology Applications, Inc. (MSE) to evaluate sorbents identified by Oak Ridge National Laboratory (ORNL) to solidify the radioactive liquid organic waste from the Radiochemical Engineering Development Center (REDC) at ORNL. REDC recovers and purifies heavy elements (berkelium, californium, einsteinium, and fermium) from irradiated targets for research and industrial applications. Both organic and aqueous waste streams are discharged from REDC. The organic waste is generated from the plutonium/uranium extraction (Purex), Cleanex, and Pubex processes. The Purex waste derives from an organic-aqueous isotope separation process for plutonium and uranium fission products, the Cleanex waste derives from the removal of fission products and other impurities from the americium/curium product, and the Pubex waste is derived from the separation process of plutonium from dissolved targets. MSE had also been tasked to test a grouting formula for the aqueous waste stream that includes radioactive shielding material. The aqueous waste is a mixture of the raffinate streams from the various extraction processes plus the caustic solution that is used to dissolve the aluminum cladding from the irradiated targets. (authors)

Bickford, J. [MSE Technology Applications, Inc., MT (United States); Taylor, P. [Oak Ridge National Laboratory, Oak Ridge, TN (United States)

2007-07-01T23:59:59.000Z

448

National Securities Technologies _NSTec_ Livermore Operations...  

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

National Nuclear Security Administration NRTL Nationally Recognized Testing Laboratory NSTec National Security Technologies, LLC NTS Nevada Test Site OSHA Occupational Safety and...

449

Sandia National Laboratories: solar power  

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

Interactive Tour Operated by Sandia National Laboratories for the U.S. Department of Energy (DOE), the National Solar Thermal Test Facility (NSTTF) is the only test facility...

450

Sandia National Laboratories: Solar Energy  

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

Interactive Tour Operated by Sandia National Laboratories for the U.S. Department of Energy (DOE), the National Solar Thermal Test Facility (NSTTF) is the only test facility...

451

Sandia National Laboratories: Solar Research  

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

Interactive Tour Operated by Sandia National Laboratories for the U.S. Department of Energy (DOE), the National Solar Thermal Test Facility (NSTTF) is the only test facility...

452

CRAD, Occupational Safety & Health - Oak Ridge National Laboratory...  

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

Occupational Safety & Health - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Occupational Safety & Health - Oak Ridge National Laboratory High Flux...

453

EA-0813; Environmental Assessment and (FONSI) The Tokamak Fusion Test Reactor Decontamination and Decommissioning Project and The Tokamak Physics Experiment at the PPPL  

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

13; Environmental Assessment and (FONSI) The Tokamak Fusion 13; Environmental Assessment and (FONSI) The Tokamak Fusion Test Reactor Decontamination and Decommissioning Project and The Tokamak Physics Experiment at the PPPL Table of Contents EXECUTIVE SUMMARY ACRONYMS Glossary of Radiological Terms SCIENTIFIC NOTATION 1.0 PURPOSE AND NEED FOR THE PROPOSED ACTIONS 1.1 TFTR D&D Project 1.2 TPX Project 1.3 Scope of Document 1.4 Local Community Relations Program 1.5 References 2.0 DESCRIPTION OF THE PROPOSED ACTIONS AND ALTERNATIVES 2.1 TFTR D&D Project 2.2 TPX Project 2.3 Environmental Monitoring 2.4 References 3.0 DESCRIPTION OF THE AFFECTED ENVIRONMENT 3.1 PPPL Proposed Site 3.2 ORR Alternative Site 3.3 References 4.0 ENVIRONMENTAL CONSEQUENCES OF THE PROPOSED ACTIONS AND ALTERNATIVES 4.1 TFTR D&D Project 4.1.1 Impacts of Normal D&D Operations

454

Test plan for the Parallex CANDU-MOX irradiation  

SciTech Connect

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

455

Researchers test novel power system for space travel  

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

Power system for space travel Power system for space travel Researchers test novel power system for space travel The research team recently demonstrated the first use of a heat pipe to cool a small nuclear reactor and power a Stirling engine. November 26, 2012 John Bounds of Los Alamos National Laboratory's Advanced Nuclear Technology Division makes final adjustments on the DUFF experiment, a demonstration of a simple, robust fission reactor prototype that could be used as a power system for space travel. DUFF is the first demonstration of a space nuclear reactor system to produce electricity in the United States since 1965. John Bounds of Los Alamos National Laboratory's Advanced Nuclear Technology Division makes final adjustments on the DUFF experiment, a demonstration of a simple, robust fission reactor prototype that could be used as a power

456

NUCLEAR REACTORS.  

E-Print Network (OSTI)

??Nuclear reactors are devices containing fissionable material in sufficient quantity and so arranged as to be capable of maintaining a controlled, self-sustaining NUCLEAR FISSION chain… (more)

Belachew, Dessalegn

2010-01-01T23:59:59.000Z

457

Nuclear Energy Enabling Technologies (NEET) Reactor Materials  

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

Enabling Technologies (NEET) Reactor Materials Enabling Technologies (NEET) Reactor Materials Award Recipient Estimated Award Amount* Award Location Supporting Organizations Project Description University of Nebraska $979,978 Lincoln, NE Massachusetts Institute of Technology (Cambridge, MA), Texas A&M (College Station, TX) Project will explore the development of advanced metal/ceramic composites. These improvements could lead to more efficient production of electricity in advanced reactors. Oak Ridge National Laboratory $849,000 Oak Ridge, TN University of Wisconsin-Madison (Madison, WI) Project will develop novel high-temperature high-strength steels with the help of computational modeling, which could lead to increased efficiency in advanced reactors. Pacific Northwest National Laboratory

458

SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary  

SciTech Connect

U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in a remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.

None

2013-09-25T23:59:59.000Z

459

University Reactor Conversion Lessons Learned Workshop for Purdue University Reactor  

SciTech Connect

The Department of Energy’s Idaho National Laboratory, under its programmatic responsibility for managing the University Research Reactor Conversions, has completed the conversion of the reactor at Purdue University Reactor. With this work completed and in anticipation of other impending conversion projects, the INL convened and engaged the project participants in a structured discussion to capture the lessons learned. The lessons learned process has allowed us to capture gaps, opportunities, and good practices, drawing from the project team’s experiences. These lessons will be used to raise the standard of excellence, effectiveness, and efficiency in all future conversion projects.

Eric C. Woolstenhulme; Dana M. Hewit

2008-09-01T23:59:59.000Z

460

Reactor and Nuclear Systems Division (RNSD)  

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

RNSD Home RNSD Home Research Groups Advanced Reactor Systems & Safety Nuclear Data & Criticality Safety Nuclear Security Modeling Radiation Safety Information Computational Center Radiation Transport Reactor Physics Thermal Hydraulics & Irradiation Engineering Used Fuel Systems Staff Details (CV/Bios) Publications Org Chart Contact Us ORNL Staff Only Research Groups Advanced Reactor Systems & Safety Nuclear Data & Criticality Safety Nuclear Security Modeling Radiation Safety Information Computational Center Radiation Transport Reactor Physics Thermal Hydraulics & Irradiation Engineering Used Fuel Systems Reactor and Nuclear Systems Division News Highlights U.S. Rep. Fleischmann touts ORNL as national energy treasure Martin Peng wins Fusion Power Associates Leadership Award

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461

Idaho National Laboratory/Nuclear Power Industry Strategic Plan for Light Water Reactor Research and Development An Industry-Government Partnership to Address Climate Change and Energy Security  

SciTech Connect

The dual issues of energy security and climate change mitigation are driving a renewed debate over how to best provide safe, secure, reliable and environmentally responsible electricity to our nation. The combination of growing energy demand and aging electricity generation infrastructure suggests major new capacity additions will be required in the years ahead.

Electric Power Research

2007-11-01T23:59:59.000Z

462

EIS-0144: Siting, Construction, and Operation of New Production Reactor Capacity; Hanford Site, Idaho National Engineering Laboratory, and Savannah River Site  

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

The U.S. Department of Energy developed this statement to assess the potential environmental impacts, both on a broad programmatic level and on a project-specific level, concerning a proposed action to provide new tritium production capacity to meet the nation's nuclear defense requirements well into the 21st century.

463

DOE/NBS (Department of Energy/National Bureau of Standards) forum on testing and rating procedures for consumer products, October 2-3, 1985. Final report  

SciTech Connect

One hundred thirty-four persons participated in a Forum on Testing and Rating Procedures for Consumer Products held at the National Bureau of Standards (NBS), Gaithersburg, Maryland, on October 2-3, 1985. The objectives of the forum, planned in cooperation with various industry associations, were: (1) to provide a line of communication between test procedure users and test-procedure developers; (2) to provide an opportunity for participants to present technical and research issues concerning Department of Energy (DOE) test procedures that need to be addressed; and (3) to assist DOE and NBS in establishing a future agenda for the development and/or revision of testing and rating procedures. The report summarizes discussions, conclusions and recommendations developed by the forum participants for the following consumer products: heat pumps and air conditioners; furnaces, boilers, and household heaters; water heaters; refrigerators, refrigerator-freezers and freezers.

Dikkers, R.D.

1986-07-01T23:59:59.000Z

464

Nuclear Reactor Severe Accident Experiments  

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

Nuclear Reactor Severe Accident Experiments Nuclear Reactor Severe Accident Experiments Capabilities Engineering Experimentation Reactor Safety Testing and Analysis Overview Nuclear Reactor Severe Accident Experiments MAX NSTF SNAKE Aerosol Experiments System Components Laser Applications Robots Applications Other Facilities Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Nuclear Reactor Severe Accident Experiments 1 2 3 4 5 6 7 We perform experiments simulating reactor core melt phenomena in which molten core debris ("corium") erodes the concrete floor of a containment building. This occurred during the Fukushima nuclear power plant accident though the extent of concrete damage is yet unknown. This video shows the top view of a churning molten pool of uranium oxide at 2000°C (3600°F) seen during an experiment at Argonne. Corium behaves much like lava.