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Note: This page contains sample records for the topic "reactor testing station" 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.
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1

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

2

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

3

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...

4

High speed imager test station  

DOE Patents (OSTI)

A test station enables the performance of a solid state imager (herein called a focal plane array or FPA) to be determined at high image frame rates. A programmable waveform generator is adapted to generate clock pulses at determinable rates for clock light-induced charges from a FPA. The FPA is mounted on an imager header board for placing the imager in operable proximity to level shifters for receiving the clock pulses and outputting pulses effective to clock charge from the pixels forming the FPA. Each of the clock level shifters is driven by leading and trailing edge portions of the clock pulses to reduce power dissipation in the FPA. Analog circuits receive output charge pulses clocked from the FPA pixels. The analog circuits condition the charge pulses to cancel noise in the pulses and to determine and hold a peak value of the charge for digitizing. A high speed digitizer receives the peak signal value and outputs a digital representation of each one of the charge pulses. A video system then displays an image associated with the digital representation of the output charge pulses clocked from the FPA. In one embodiment, the FPA image is formatted to a standard video format for display on conventional video equipment. 12 figs.

Yates, G.J.; Albright, K.L.; Turko, B.T.

1995-11-14T23:59:59.000Z

5

High speed imager test station  

DOE Patents (OSTI)

A test station enables the performance of a solid state imager (herein called a focal plane array or FPA) to be determined at high image frame rates. A programmable waveform generator is adapted to generate clock pulses at determinable rates for clock light-induced charges from a FPA. The FPA is mounted on an imager header board for placing the imager in operable proximity to level shifters for receiving the clock pulses and outputting pulses effective to clock charge from the pixels forming the FPA. Each of the clock level shifters is driven by leading and trailing edge portions of the clock pulses to reduce power dissipation in the FPA. Analog circuits receive output charge pulses clocked from the FPA pixels. The analog circuits condition the charge pulses to cancel noise in the pulses and to determine and hold a peak value of the charge for digitizing. A high speed digitizer receives the peak signal value and outputs a digital representation of each one of the charge pulses. A video system then displays an image associated with the digital representation of the output charge pulses clocked from the FPA. In one embodiment, the FPA image is formatted to a standard video format for display on conventional video equipment.

Yates, George J. (Santa Fe, NM); Albright, Kevin L. (Los Alamos, NM); Turko, Bojan T. (Moraga, CA)

1995-01-01T23:59:59.000Z

6

Improving Unit Operations-Test Station Performance  

E-Print Network (OSTI)

) usage. The basic concept evaluates the varying criterias affecting these elements and their direct impact on production/test station operating costs. Second consideration explores other methods available to enhance mechanical compatibility with operator...

Filak, J. J. Jr.

7

Technology, Safety and Costs of Decommissioning Nuclear Reactors At Multiple-Reactor Stations  

SciTech Connect

Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWRs) and large (1155-MWe) boiling water reactors {BWRs) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite nuclear waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services. Five scenarios for decommissioning reactors at a multiple-reactor station are investigated. The number of reactors on a site is assumed to be either four or ten; nuclear waste disposal is varied between immediate offsite disposal, interim onsite storage, and immediate onsite disposal. It is assumed that the decommissioned reactors are not replaced in one scenario but are replaced in the other scenarios. Centralized service facilities are provided in two scenarios but are not provided in the other three. Decommissioning of a PWR or a BWR at a multiple-reactor station probably will be less costly and result in lower radiation doses than decommissioning an identical reactor at a single-reactor station. Regardless of whether the light water reactor being decommissioned is at a single- or multiple-reactor station: • the estimated occupational radiation dose for decommissioning an LWR is lowest for SAFSTOR and highest for DECON • the estimated cost of decommissioning a PWR is lowest for ENTOMB and highest for SAFSTOR • the estimated cost of decommissioning a BWR is lowest for OECON and highest for SAFSTOR. In all cases, SAFSTOR has the lowest occupational radiation dose and the highest cost.

Wittenbrock, N. G.

1982-01-01T23:59:59.000Z

8

Hydrogen Station Test Device Design and Fabrication | Department...  

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

Hydrogen Station Test Device Design and Fabrication Hydrogen Station Test Device Design and Fabrication October 2, 2014 - 3:02pm Addthis Open Date: August 26, 2014 Requesting...

9

Nuclear Rocket Development Station at the Nevada Test Site |...  

Office of Environmental Management (EM)

Nuclear Rocket Development Station at the Nevada Test Site Nuclear Rocket Development Station at the Nevada Test Site During the 1950s, the United States launched a nuclear rocket...

10

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...

11

Reactor Vessel and Reactor Vessel Internals Segmentation at Zion Nuclear Power Station - 13230  

SciTech Connect

Zion Nuclear Power Station (ZNPS) is a dual-unit Pressurized Water Reactor (PWR) nuclear power plant located on the Lake Michigan shoreline, in the city of Zion, Illinois approximately 64 km (40 miles) north of Chicago, Illinois and 67 km (42 miles) south of Milwaukee, Wisconsin. Each PWR is of the Westinghouse design and had a generation capacity of 1040 MW. Exelon Corporation operated both reactors with the first unit starting production of power in 1973 and the second unit coming on line in 1974. The operation of both reactors ceased in 1996/1997. In 2010 the Nuclear Regulatory Commission approved the transfer of Exelon Corporation's license to ZionSolutions, the Long Term Stewardship subsidiary of EnergySolutions responsible for the decommissioning of ZNPS. In October 2010, ZionSolutions awarded Siempelkamp Nuclear Services, Inc. (SNS) the contract to plan, segment, remove, and package both reactor vessels and their respective internals. This presentation discusses the tools employed by SNS to remove and segment the Reactor Vessel Internals (RVI) and Reactor Vessels (RV) and conveys the recent progress. SNS's mechanical segmentation tooling includes the C-HORCE (Circumferential Hydraulically Operated Cutting Equipment), BMT (Bolt Milling Tool), FaST (Former Attachment Severing Tool) and the VRS (Volume Reduction Station). Thermal segmentation of the reactor vessels will be accomplished using an Oxygen- Propane cutting system. The tools for internals segmentation were designed by SNS using their experience from other successful reactor and large component decommissioning and demolition (D and D) projects in the US. All of the designs allow for the mechanical segmentation of the internals remotely in the water-filled reactor cavities. The C-HORCE is designed to saw seven circumferential cuts through the Core Barrel and Thermal Shield walls with individual thicknesses up to 100 mm (4 inches). The BMT is designed to remove the bolts that fasten the Baffle Plates to the Baffle Former Plates. The FaST is designed to remove the Baffle Former Plates from the Core Barrel. The VRS further volume reduces segmented components using multiple configurations of the 38i and horizontal reciprocating saws. After the successful removal and volume reduction of the Internals, the RV will be segmented using a 'First in the US' thermal cutting process through a co-operative effort with Siempelkamp NIS Ingenieurgesellschaft mbH using their experience at the Stade NPP and Karlsruhe in Germany. SNS mobilized in the fall of 2011 to commence execution of the project in order to complete the RVI segmentation, removal and packaging activities for the first unit (Unit 2) by end of the 2012/beginning 2013 and then mobilize to the second unit, Unit 1. Parallel to the completion of the segmentation of the reactor vessel internals at Unit 1, SNS will segment the Unit 2 pressure vessel and at completion move to Unit 1. (authors)

Cooke, Conrad; Spann, Holger [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)] [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)

2013-07-01T23:59:59.000Z

12

Formation of a nuclear reactor’s molten core bath in a crucible-type corium catcher for a nuclear power station equipped with VVER reactors  

Science Journals Connector (OSTI)

Results from a calculation study on analyzing the formation of a melt bath in a crucible-type catcher for the conditions of a severe accident at a nuclear power station equipped with VVER-1000 reactors are presen...

S. V. Beshta; S. A. Vitol’; V. S. Granovskii; E. K. Kalyago…

2011-05-01T23:59:59.000Z

13

Machine for monitoring nuclear reactor equipment at its storage station  

SciTech Connect

Machine for monitoring internal equipment of a nuclear reactor, when stored outside the reactor vessel in the swimming pool of the reactor, comprising a first longitudinal carriage movable on rails on the bottom of the swimming pool, a transverse beam movable on the first carriage, a second transverse carriage bearing monitoring instruments and movable on the beam, and an arrangement for moving the beam in the predetermined steps on the first carriage, and the second carriage on the beam.

Castrec, Y. M.; Launay, J. P.

1985-02-12T23:59:59.000Z

14

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

15

Hydrogen Station & ICE Vehicle Operations and Testing  

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

Hydrogen Station & ICE Vehicle Operations and Testing Jim Francfort for Lee Slezak WestStart CALSTART Hydrogen Internal Combustion Engine Symposium - February 2006 INL/CON-06-01109 Presentation Outline * Background and Goal * Arizona Public Service (APS) Alternative Fuel (Hydrogen) Pilot Plant - design and operations * Fuel Dispensing * Prototype Dispenser Testing * Hydrogen and HCNG Internal Combustion Engine (ICE) Vehicle Testing Activities * WWW Information AVTA Background and Goal * AVTA is part of the U.S. Department of Energy's FreedomCAR and Vehicle Technologies Program * These activities are conducted by the Idaho National Laboratory (INL) and the AVTA testing partner Electric Transportation Applications * AVTA Goal - Provide benchmark data for technology

16

Developing Fully Coupled Dynamical Reactor Core Isolation System Models in RELAP-7 for Extended Station Black-Out Analysis  

SciTech Connect

The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup water to the reactor vessel for core cooling when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. It was one of the very few safety systems still available during the Fukushima Daiichi accidents after the tsunamis hit the plants and the system successfully delayed the core meltdown for a few days for unit 2 & 3. Therefore, detailed models for RCIC system components are indispensable to understand extended station black-out accidents (SBO) for BWRs. As part of the effort to develop the new generation reactor system safety analysis code RELAP-7, major components to simulate the RCIC system have been developed. This paper describes the models for those components such as turbine, pump, and wet well. Selected individual component test simulations and a simplified SBO simulation up to but before core damage is presented. The successful implementation of the simplified RCIC and wet well models paves the way to further improve the models for safety analysis by including more detailed physical processes in the near future.

Haihua Zhao; Ling Zou; Hongbin Zhang; David Andrs; Richard Martineau

2014-04-01T23:59:59.000Z

17

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

18

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

19

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...

20

AVTA: GE Energy WattStation AC Level 2 Charging System Testing...  

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

GE Energy WattStation AC Level 2 Charging System Testing Results AVTA: GE Energy WattStation AC Level 2 Charging System Testing Results The Vehicle Technologies Office's Advanced...

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

SNL Issues a Request for Quotation for a Hydrogen Station Test Device  

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

Sandia National Laboratories (SNL) has issued a Request for Quotation (RFQ) for hydrogen station equipment performance testing device fabrication.

22

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

23

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.

24

Underground nuclear power station using self-regulating heat-pipe controlled reactors  

DOE Patents (OSTI)

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working flud in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor.

Hampel, Viktor E. (Pleasanton, CA)

1989-01-01T23:59:59.000Z

25

An underground nuclear power station using self-regulating heat-pipe controlled reactors  

DOE Patents (OSTI)

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast- acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor. 5 figs.

Hampel, V.E.

1988-05-17T23:59:59.000Z

26

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

27

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

28

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

29

BWRSAR (Boiling Water Reactor Severe Accident Response) calculations of reactor vessel debris pours for Peach Bottom short-term station blackout  

SciTech Connect

This paper describes recent analyses performed by the BWR Severe Accident Technology (BWRSAT) Program at Oak Ridge National Laboratory to estimate the release of debris from the reactor vessel for the unmitigated short-term station blackout accident sequence. Calculations were performed with the BWR Severe Accident Response (BWRSAR) code and are based upon consideration of the Peach Bottom Atomic Power Station. The modeling strategies employed within BWRSAR for debris relocation within the reactor vessel are briefly discussed and the calculated events of the accident sequence, including details of the calculated debris pours, are presented. 4 refs., 13 figs., 3 tabs.

Hodge, S.A.; Ott, L.J.

1988-01-01T23:59:59.000Z

30

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

31

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

32

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

33

Trona Injection Tests: Mirant Potomac River Station, Unit 1, November 12 to  

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

Trona Injection Tests: Mirant Potomac River Station, Unit 1, Trona Injection Tests: Mirant Potomac River Station, Unit 1, November 12 to December 23, 2005, Summary Report Trona Injection Tests: Mirant Potomac River Station, Unit 1, November 12 to December 23, 2005, Summary Report Docket No. EO-05-01: Trona injection tests were conducted at Mirant's Potomac River Station on Unit 1 between November 12 and December 23, 2005. The purpose of these tests was to determine the capability of dry injection of trona to achieve substantial SO2 removal from the stack discharge, and the determination of other operating impacts from he trona injection, if any. Trona Injection Tests: Mirant Potomac River Station, Unit 1, November 12 to December 23, 2005, Summary Report More Documents & Publications Special Environmental Analysis For Actions Taken under U.S. Department of

34

Fuel Retrieval Sub (FRS) Project Decapping Station Performance Test Data Report  

SciTech Connect

This document is to provide the test data report for Decapping Station Performance Testing. These performance tests were full scale and viewed as a continuation of development testing performed earlier (SNF-2710). A prototype decapping station confinement box was tested, along with some special tools required for the process, providing assurance that the fuel handling equipment will operate as designed, allowing for release of the FRS equipment for installation.

THIELGES, J.R.

2000-01-13T23:59:59.000Z

35

System Definition Document: Reactor Data Necessary for Modeling Plutonium Disposition in Catawba Nuclear Station Units 1 and 2  

SciTech Connect

The US Department of Energy (USDOE) has contracted with Duke Engineering and Services, Cogema, Inc., and Stone and Webster (DCS) to provide mixed-oxide (MOX) fuel fabrication and reactor irradiation services in support of USDOE's mission to dispose of surplus weapons-grade plutonium. The nuclear station units currently identified as mission reactors for this project are Catawba Units 1 and 2 and McGuire Units 1 and 2. This report is specific to Catawba Nuclear Station Units 1 and 2, but the details and materials for the McGuire reactors are very similar. The purpose of this document is to present a complete set of data about the reactor materials and components to be used in modeling the Catawba reactors to predict reactor physics parameters for the Catawba site. Except where noted, Duke Power Company or DCS documents are the sources of these data. These data are being used with the ORNL computer code models of the DCS Catawba (and McGuire) pressurized-water reactors.

Ellis, R.J.

2000-11-01T23:59:59.000Z

36

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...

37

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

38

NEPA CX Determination SS-SC-10-01 for End Station Test Beam (ESTB)  

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

0-01 for End Station Test Beam (ESTB) 0-01 for End Station Test Beam (ESTB) National Environmental Policy Act (NEPA) Categorical Exclusion (CX) Determination A. SSO NEPA Control #: SS-SC-10-01 B. Brief Description of Proposed Action: The End Station Test Beam (ESTB) is a new experimental facility that will use 5Hz of the 120 hz 13.6 GeV electron beam from the existing Linac Coherent Light Source (LCLS) to restore test beam capabilities in End Station A (ESA), an existing building at SLAC. In Stage I of this proposal, four new kicker magnets will be added to the Beam Switchyard (BSY) to divert a small fraction of the existing LCLS beam pulses to the A-line for beam instrumentation and accelerator physics studies at full electron beam intensity. The Personnel Protection System in ESA will be

39

Testing of power-generating gas-turbine plants at Russian electric power stations  

Science Journals Connector (OSTI)

This paper cites results of thermal testing of various types and designs of power-generating gas-turbine plants (GTP), which have been placed in service at electric-power stations in Russia in recent years. Therm...

G. G. Ol’khovskii; A. V. Ageev; S. V. Malakhov…

2006-07-01T23:59:59.000Z

40

Technology, safety and costs of decommissioning a Reference Boiling Water Reactor Power Station. Main report. Volume 1  

SciTech Connect

Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWe.

Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

1980-06-01T23:59:59.000Z

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

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)

42

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

43

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

44

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

45

14th Annual international meeting of wind turbine test stations: Proceedings  

SciTech Connect

These proceedings are of the 14th Annual International Meeting of Test Stations. As the original charter states these meetings are intended to be an international forum for sharing wind turbine testing experiences. By sharing their experiences they can improve testing skills and techniques. As with all new industries the quality of the products is marked by how well they learn from their experiences and incorporate this learning into the next generation of products. The test station`s role in this process is to provide accurate information to the companies they serve. This information is used by designers to conform and improve their designs. It is also used by certification agencies for confirming the quality of these designs. By sharing of experiences they are able to accomplished these goals, serve these customers better and ultimately improve the international wind energy industry.

Not Available

1994-11-01T23:59:59.000Z

46

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

47

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

48

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

49

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

50

DOE - Office of Legacy Management -- Naval Ordnance Test Station - CA 06  

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

Ordnance Test Station - CA 06 Ordnance Test Station - CA 06 FUSRAP Considered Sites Site: NAVAL ORDNANCE TEST STATION (CA.06) Eliminated from consideration under FUSRAP Designated Name: Not Designated Alternate Name: China Lake Naval Weapons Center Salt Wells Pilot Plant CA.06-1 Location: Inyokern , California CA.06-1 Evaluation Year: 1987 CA.06-1 Site Operations: Naval facility; experimental development work on shape charges and quality castings on a pilot plant scale. CA.06-1 Site Disposition: Eliminated - No indication that radioactive materials were handled at the site CA.06-1 Radioactive Materials Handled: None Indicated CA.06-1 Primary Radioactive Materials Handled: None CA.06-1 Radiological Survey(s): No Site Status: Eliminated from consideration under FUSRAP Also see

51

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

52

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

53

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

54

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...

55

Investigations on optimization of accident management measures following a station blackout accident in a VVER-1000 pressurized water reactor  

SciTech Connect

The reactor safety issues are of primary importance for preserving the health of the population and ensuring no release of radioactivity and fission products into the environment. A part of the nuclear research focuses on improvement of the safety of existing nuclear power plants. Studies, research and efforts are a continuing process at improving the safety and reliability of existing and newly developed nuclear power plants at prevention of a core melt accident. Station blackout (loss of AC power supply) is one of the dominant accidents taken into consideration at performing accident analysis. In case of multiple failures of safety systems it leads to a severe accident. To prevent an accident to turn into a severe one or to mitigate the consequences, accident management measures must be performed. The present paper outlines possibilities for application and optimization of accident management measures following a station blackout accident. Assessed is the behaviour of the nuclear power plant during a station blackout accident without accident management measures and with application of primary/secondary side oriented accident management measures. Discussed are the possibilities for operators ' intervention and the influence of the performed accident management measures on the course of the accident. Special attention has been paid to the effectiveness of the passive feeding and physical phenomena having an influence on the system behaviour. The performed simulations show that the effectiveness of the secondary side feeding procedure can be limited due to an early evaporation or flashing effects in the feed water system. The analyzed cases show that the effectiveness of the accident management measures strongly depends on the initiation criteria applied for depressurization of the reactor coolant system. (authors)

Tusheva, P.; Schaefer, F.; Kliem, S. [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstrasse 400, D-01328 Dresden (Germany)

2012-07-01T23:59:59.000Z

56

Corrective Action Plan for Corrective Action Unit 490: Station 44 Burn Area, Tonopah Test Range, Nevada  

SciTech Connect

Corrective Action Unit (CAU) 490, Station 44 Burn Area is located on the Tonopah Test Range (TTR). CAU 490 is listed in the Federal Facility Agreement and Consent Order (FFACO, 1996) and includes for Corrective Action Sites (CASs): (1) Fire Training Area (CAS 03-56-001-03BA); (2) Station 44 Burn Area (CAS RG-56-001-RGBA); (3) Sandia Service Yard (CAS 03-58-001-03FN); and (4) Gun Propellant Burn Area (CAS 09-54-001-09L2).

K. B. Campbell

2002-04-01T23:59:59.000Z

57

Corrective action decision document, Second Gas Station, Tonopah test range, Nevada (Corrective Action Unit No. 403)  

SciTech Connect

This Corrective Action Decision Document (CADD) for Second Gas Station (Corrective Action Unit [CAU] No. 403) has been developed for the U.S. Department of Energy`s (DOE) Nevada Environmental Restoration Project to meet the requirements of the Federal Facility Agreement and Consent Order (FFACO) of 1996 as stated in Appendix VI, {open_quotes}Corrective Action Strategy{close_quotes} (FFACO, 1996). The Second Gas Station Corrective Action Site (CAS) No. 03-02-004-0360 is the only CAS in CAU No. 403. The Second Gas Station CAS is located within Area 3 of the Tonopah Test Range (TTR), west of the Main Road at the location of former Underground Storage Tanks (USTs) and their associated fuel dispensary stations. The TTR is approximately 225 kilometers (km) (140 miles [mi]) northwest of Las Vegas, Nevada, by air and approximately 56 km (35 mi) southeast of Tonopah, Nevada, by road. The TTR is bordered on the south, east, and west by the Nellis Air Force Range and on the north by sparsely populated public land administered by the Bureau of Land Management and the U.S. Forest Service. The Second Gas Station CAS was formerly known as the Underground Diesel Tank Site, Sandia Environmental Restoration Site Number 118. The gas station was in use from approximately 1965 to 1980. The USTs were originally thought to be located 11 meters (m) (36 feet [ft]) east of the Old Light Duty Shop, Building 0360, and consisted of one gasoline UST (southern tank) and one diesel UST (northern tank) (DOE/NV, 1996a). The two associated fuel dispensary stations were located northeast (diesel) and southeast (gasoline) of Building 0360 (CAU 423). Presently the site is used as a parking lot, Building 0360 is used for mechanical repairs of vehicles.

NONE

1997-11-01T23:59:59.000Z

58

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

59

Chernobyl doses. Volume 3. Habitat and vegetation near the Chernobyl Nuclear Reactor Station. Technical report, 29 September 1987-28 February 1992  

SciTech Connect

This volume presents a detailed exposition on the soils, climate, and vegetation of the Poles'ye region of Ukraine and Belorussia with emphasis on the area around the Chernobyl Nuclear Reactor Station. This data provides background for interpretation of multispectral satellite imagery of the area. Volume 1 uses these images and the information of this report to analyze the radiation response of the canopy of the coniferous forests in the immediate vicinity of the reactor station after the accident of 26 April 1986.... Chernobyl, Forest damage, Landsat, Change detection, Conifer stress, Fallout, Ionizing radiation, Multispectral imagery.

Painter, E.L.; Whicker, F.W.

1993-01-01T23:59:59.000Z

60

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

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

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

62

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

63

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

64

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

65

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

66

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

67

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

68

SNL Issues Notice of Intent to Release a Request for Quotation for a Hydrogen Station Test Device  

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

In support of DOE's Hydrogen Fueling Infrastructure Research and Station Technology (H2FIRST) project launched in April 2014, Sandia National Laboratories (SNL) has posted a notice of intent to issue a Request for Quotation for hydrogen station equipment performance testing device fabrication.

69

Light Water Reactor Sustainability Program Support and Modeling for the Boiling Water Reactor Station Black Out Case Study Using RELAP and RAVEN  

SciTech Connect

The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated. In order to evaluate the impact of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the impact of power uprate on the safety of a boiled water reactor system. The case study considered is a loss of off-site power followed by the loss of diesel generators, i.e., a station black out (SBO) event. Analysis is performed by using a thermo-hydraulic code, i.e. RELAP-5, and a stochastic analysis tool currently under development at INL, i.e. RAVEN. Starting from the event tree models contained in SAPHIRE, we built the input file for RELAP-5 that models in great detail system dynamics under SBO conditions. We also interfaced RAVEN with RELAP-5 so that it would be possible to run multiple RELAP-5 simulation runs by changing specific keywords of the input file. We both employed classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. We also employed advanced data analysis and visualization tools that helped us to correlate simulation outcome such as maximum core temperature with a set of input uncertain parameters. Results obtained gave a detailed overview of the issues associated to power uprate for a SBO accident scenario. We were able to quantify how timing of safety related events were impacted by a higher reactor core power. Such insights can provide useful material to the decision makers to perform risk-infomed safety margins management.

Diego Mandelli; Curtis Smith; Thomas Riley; John Schroeder; Cristian Rabiti; Aldrea Alfonsi; Joe Nielsen; Dan Maljovec; Bie Wang; Valerio Pascucci

2013-09-01T23:59:59.000Z

70

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

71

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

72

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...

73

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...

74

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...

75

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

76

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

77

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

78

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

79

Field Testing of Activated Carbon Injection Options for Mercury Control at TXU's Big Brown Station  

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

Field TesTing oF AcTivATed cArbon Field TesTing oF AcTivATed cArbon injecTion opTions For Mercury conTrol AT TXu's big brown sTATion Background The 2005 Clean Air Mercury Rule will require significant reductions in mercury emissions from coal-fired power plants. Lignite coal is unique because of its highly variable ash content (rich in alkali and alkaline-earth elements), high moisture levels, low chlorine content, and high calcium content. Unique to Texas lignite coals are relatively high iron and selenium concentrations. When combusting Texas lignite coals, up to 80 percent of the mercury in the flue gas is present as elemental mercury, which is not readily captured by downstream pollution control devices. To better understand the factors that influence mercury control at units firing

80

Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 1. Main report. Technical report, September 1977-October 1979  

SciTech Connect

Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE.

Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

1980-06-01T23:59:59.000Z

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

Analysis of molybdenum-99 production capability in the materials test station  

SciTech Connect

The United States of America currently relies on foreign suppliers to meet all of it needs for molybdenum-99 (Mo-99) used in medical diagnostic procedures. The current US demand is at least 5000 six-day curies per week. Neutronics calculations have been performed to assess whether the proposed Materials Test Station (MTS) could potentially generate Mo-99. Two target material options have been explored for Mo-99 production in the MTS: low enriched uranium (LEU) and Tc-99. For LEU, scoping calculations indicate that MTS can supply nearly half of the current US demand with only minor neutronic impact on the MTS primary mission. For the Tc-99 option, the MTS could produce about one-tenth of the US demand.

Pitcher, Eric J [Los Alamos National Laboratory

2009-01-01T23:59:59.000Z

82

Overview of New Tools to Perform Safety Analysis: BWR Station Black Out Test Case  

SciTech Connect

Dynamic Probabilistic Risk Assessment (DPRA) methodologies couple system simulator codes (e.g., RELAP, MELCOR) with simulation controller codes (e.g., RAVEN, ADAPT). While system simulator codes accurately model system dynamics deterministically, simulation controller codes introduce both deterministic (e.g., system control logic, operating procedures) and stochastic (e.g., component failures, parameter uncertainties) elements into the simulation. Typically, a DPRA is performed by: 1) sampling values of a set of parameters from the uncertainty space of interest (using the simulation controller codes), and 2) simulating the system behavior for that specific set of parameter values (using the system simulator codes). For complex systems, one of the major challenges in using DPRA methodologies is to analyze the large amount of information (i.e., large number of scenarios ) generated, where clustering techniques are typically employed to allow users to better organize and interpret the data. In this paper, we focus on the analysis of a nuclear simulation dataset that is part of the Risk Informed Safety Margin Characterization (RISMC) Boiling Water Reactor (BWR) station blackout (SBO) case study. We apply a software tool that provides the domain experts with an interactive analysis and visualization environment for understanding the structures of such high-dimensional nuclear simulation datasets. Our tool encodes traditional and topology-based clustering techniques, where the latter partitions the data points into clusters based on their uniform gradient flow behavior. We demonstrate through our case study that both types of clustering techniques complement each other in bringing enhanced structural understanding of the data.

D. Mandelli; C. Smith; T. Riley; J. Nielsen; J. Schroeder; C. Rabiti; A. Alfonsi; Cogliati; R. Kinoshita; V. Pasucci; B. Wang; D. Maljovec

2014-06-01T23:59:59.000Z

83

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...

84

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...

85

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

86

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

87

ERRATA SHEET for Corrective Action Plan for Corrective Action Unit 490: Station 44 Burn Area, Tonopah Test Range, Nevada  

SciTech Connect

Section 2.1.1.3 of the Table of Contents reference on Page v and on Page 12 of the Corrective Action Plan for Corrective Action Unit 490: Station 44 Burn Area, Tonopah Test Range, Nevada erroneously refers to the Nevada Environmental Policy Act Determination. The correct title of the referenced document is the National Environmental Policy Act Determination.

K. B. Campbell

2002-04-01T23:59:59.000Z

88

Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 2. Appendices. Technical report, September 1977-October 1979  

SciTech Connect

Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE. This volume contains the appendices.

Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

1980-06-01T23:59:59.000Z

89

50 MW X-BAND RF SYSTEM FOR A PHOTOINJECTOR TEST STATION AT LLNL  

SciTech Connect

In support of X-band photoinjector development efforts at LLNL, a 50 MW test station is being constructed to investigate structure and photocathode optimization for future upgrades. A SLAC XL-4 klystron capable of generating 50 MW, 1.5 microsecond pulses will be the high power RF source for the system. Timing of the laser pulse on the photocathode with the applied RF field places very stringent requirements on phase jitter and drift. To achieve these requirements, the klystron will be powered by a state of the art, solid-state, high voltage modulator. The 50 MW will be divided between the photoinjector and a traveling wave accelerator section. A high power phase shifter is located between the photoinjector and accelerator section to adjust the phasing of the electron bunches with respect to the accelerating field. A variable attenuator is included on the input of the photoinjector. The distribution system including the various x-band components is being designed and constructed. In this paper, we will present the design, layout, and status of the RF system.

Marsh, R A; Anderson, S G; Barty, C J; Beer, G K; Cross, R R; Ebbers, C A; Gibson, D J; Hartemann, F V; Houck, T L; Adolphsen, C; Candel, A; Chu, T S; Jongewaard, E N; Li, Z; Raubenheimer, T; Tantawi, S G; Vlieks, A; Wang, F; Wang, J W; Zhou, F; Deis, G A

2011-03-11T23:59:59.000Z

90

Automating controller tuning for gas-turbine mini power stations in computer testing  

Science Journals Connector (OSTI)

We justify the application of mathematical modeling methods for electric power systems in order to automate the tuning of controllers in control systems of gas-turbine power stations constructed on the basis of c...

A. I. Polulyakh; I. G. Lisovin; B. V. Kavalerov…

2014-07-01T23:59:59.000Z

91

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

92

Efficiency of producing additional power in units of nuclear power stations containing water-cooled-water-moderated reactors  

Science Journals Connector (OSTI)

There is a basic possibility to raise the maximum power of a unit containing the VVÉR-1000 reactor in the course of the fuel charge burn-up and with lowering the coefficient of the energy-release nonuniformity...

R. Z. Aminov; V. A. Khrustalev; A. A. Serdobintsev…

1986-12-01T23:59:59.000Z

93

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

94

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

95

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

96

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

97

Bruce Nuclear Generating Station B rapid cooldown test and validation of simulation model  

SciTech Connect

The SOPHT code was assessed against Bruce Nuclear Generating Station B commissioning data from a heat transport system rapid cooldown. It was found that (a) under a rapid upstream depressurization, the steam relief valves, like orifices, had a lower discharge coefficient than the corresponding steadystate value and (b) the flashing of water in the steam generators during depressurization causes the at-power boiling heat transfer correlations to overpredict the steam generator heat transfer.

Chang, Y.F.; Langan, M.D.; Sermer, P.; Watson, P.C.

1985-09-01T23:59:59.000Z

98

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

99

Automatic control systems for gas-turbine units in mini power stations: Testing automation at the stages of design and tuning  

Science Journals Connector (OSTI)

This paper presents the testing automation procedure for automatic control systems of gas-turbine units used as drives in small-size power stations. We substantiate the applicability of mathematical modeling...

B. V. Kavalerov

2013-11-01T23:59:59.000Z

100

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

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

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

102

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

103

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

104

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

105

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

106

AVTA: GE Energy WattStation AC Level 2 Charging System 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 results from testing done on the GE Energy Wattstation AC Level 2 charging system for plug-in electric vehicles.

107

High-heat-load synchrotron tests of room-temperature, silicon crystal monochromators at the CHESS F-2 wiggler station  

SciTech Connect

This note summarizes the results of the single crystal monochromator high-heat-load tests performed at the CHESS F-2 wiggler station. The results from two different cooling geometries are presented: (1) the ``pin-post`` crystal and (2) the ``criss-cross`` crystal. The data presented were taken in August 1993 (water-cooled pin-post) and in April 1995 (water- and gallium-cooled pin-post crystal and gallium-cooled criss-cross crystal). The motivation for trying these cooling (or heat exchanger) geometries is to improve the heat transfer efficiency over that of the conventional slotted crystals. Calculations suggest that the pin-post or the microchannel design can significantly improve the thermal performance of the crystal. The pin-post crystal used here was fabricated by Rocketdyne Albuquerque Operations. From the performance of the conventional slotted crystals, it was thought that increased turbulence in the flow pattern may also enhance the heat transfer. The criss-cross crystal was a simple attempt to achieve the increased flow turbulence. The criss-cross crystal was partly fabricated in-house (cutting, etching and polishing) and bonded by RAO. Finally, a performance comparison among all the different room temperature silicon monochromators that have been tested by the APS is presented. The data includes measurements with the slotted crystal and the core-drilled crystals. Altogether, the data presented here were taken at the CHESS F-2 wiggler station between 1991 and 1995.

Lee, W.K.; Fernandez, P.B.; Graber, T.; Assoufid, L.

1995-09-08T23:59:59.000Z

108

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

109

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

110

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

111

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

112

Experience in conducting full-scale tests when modernizing the turbines at the Plavinas hydroelectric station  

Science Journals Connector (OSTI)

Starting and adjustment tests of the units at the Plavinas hydrostation during their modernization made it possible to obtain information having great importance for reliable operation of the equipment and to ...

I. M. Nagornyi

1997-07-01T23:59:59.000Z

113

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

114

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

115

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

116

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

117

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

118

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

119

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

120

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

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

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

122

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

123

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

124

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

125

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...

126

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

127

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

128

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

129

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

130

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

131

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

132

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

133

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

134

Lunar Base Thermoelectric Power Station Study  

Science Journals Connector (OSTI)

Under NASA’s Project Prometheus the Nuclear Space Power Systems Program the Jet Propulsion Laboratory Pratt & Whitney Rocketdyne and Teledyne Energy Systems have teamed with a number of universities under the Segmented Thermoelectric Multicouple Converter (STMC) Task to develop the next generation of advanced thermoelectric converters for space reactor power systems. Work on the STMC converter assembly has progressed to the point where the lower temperature stage of the segmented multicouple converter assembly is ready for laboratory testing and promising candidates for the upper stage materials have been identified and their properties are being characterized. One aspect of the program involves mission application studies to help define the potential benefits from the use of these STMC technologies for designated NASA missions such as a lunar base power station where kilowatts of power would be required to maintain a permanent manned presence on the surface of the moon. A modular 50 kWe thermoelectric power station concept was developed to address a specific set of requirements developed for this particular mission concept. Previous lunar lander concepts had proposed the use of lunar regolith as in?situ radiation shielding material for a reactor power station with a one kilometer exclusion zone radius to minimize astronaut radiation dose rate levels. In the present concept we will examine the benefits and requirements for a hermetically?sealed reactor thermoelectric power station module suspended within a man?made lunar surface cavity. The concept appears to maximize the shielding capabilities of the lunar regolith while minimizing its handling requirements. Both thermal and nuclear radiation levels from operation of the station at its 100?m exclusion zone radius were evaluated and found to be acceptable. Site preparation activities are reviewed as well as transport issues for this concept. The goal of the study was to review the entire life cycle of the unit to assess its technical problems and technology needs in all areas to support the development deployment operation and disposal of the unit.

William Determan; Patrick Frye; Jack Mondt; Jean?Pierre Fleurial; Ken Johnson; Gerhard Stapfer; Michael Brooks; Ben Heshmatpour

2006-01-01T23:59:59.000Z

135

Lunar Base Thermoelectric Power Station Study  

SciTech Connect

Under NASA's Project Prometheus, the Nuclear Space Power Systems Program, the Jet Propulsion Laboratory, Pratt and Whitney Rocketdyne, and Teledyne Energy Systems have teamed with a number of universities, under the Segmented Thermoelectric Multicouple Converter (STMC) Task, to develop the next generation of advanced thermoelectric converters for space reactor power systems. Work on the STMC converter assembly has progressed to the point where the lower temperature stage of the segmented multicouple converter assembly is ready for laboratory testing, and promising candidates for the upper stage materials have been identified and their properties are being characterized. One aspect of the program involves mission application studies to help define the potential benefits from the use of these STMC technologies for designated NASA missions such as a lunar base power station where kilowatts of power would be required to maintain a permanent manned presence on the surface of the moon. A modular 50 kWe thermoelectric power station concept was developed to address a specific set of requirements developed for this particular mission concept. Previous lunar lander concepts had proposed the use of lunar regolith as in-situ radiation shielding material for a reactor power station with a one kilometer exclusion zone radius to minimize astronaut radiation dose rate levels. In the present concept, we will examine the benefits and requirements for a hermetically-sealed reactor thermoelectric power station module suspended within a man-made lunar surface cavity. The concept appears to maximize the shielding capabilities of the lunar regolith while minimizing its handling requirements. Both thermal and nuclear radiation levels from operation of the station, at its 100-m exclusion zone radius, were evaluated and found to be acceptable. Site preparation activities are reviewed as well as transport issues for this concept. The goal of the study was to review the entire life cycle of the unit to assess its technical problems and technology needs in all areas to support the development, deployment, operation and disposal of the unit.

Determan, William; Frye, Patrick [Jet Propulsion Laboratory, California Institute of Technology, 4800 Oak Grove Dr., Pasadena, CA 91109 (United States); Mondt, Jack; Fleurial, Jean-Pierre; Johnson, Ken; Stapfer, Gerhard [Pratt and Whitney Rocketdyne Inc., P.O. Box 7922, Canoga Park, CA 91309 (United States); Brooks, Michael; Heshmatpour, Ben [Teledyne Energy Systems, Inc., 10707 Gilroy Rd, Hunt Valley, MD 21031 (United States)

2006-01-20T23:59:59.000Z

136

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

137

LANSCE | Materials Test Station  

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

most efficient way to transmute problematic elements in high-level waste: neptunium, americium, curium. However, these elements must first be manufactured into a form of nuclear...

138

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

139

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

140

Seismic stations  

Science Journals Connector (OSTI)

In the previous chapters, the equipment used for seismic stations has been described. When putting this equipment out in the field, we have a seismic station. Unfortunately it is not as simple as just putting ...

Jens Havskov; Gerardo Alguacil

2004-01-01T23:59:59.000Z

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

Technical evaluation of the noise and isolation testing of the safety features actuation system at the Davis Besse Nuclear Power Station, Unit 1  

SciTech Connect

This report documents the technical evaluation of the noise and isolation testing of the safety features actuation system at the Davis Besse Nuclear Power Station, Unit 1. The tests were to verify that faults on the non-Class 1E circuits would not propagate to the Class 1E circuits and degrade them below acceptable levels. The tests conducted demonstrated that the safety features actuation system did not degrade below acceptable levels nor was the system's ability to perform its protective functions affected.

Selan, J.C.

1981-07-01T23:59:59.000Z

142

Addendum to the Closure Report for Corrective Action Unit 403: Second Gas Station, Tonopah Test Range, Nevada, Revision 0  

SciTech Connect

This document constitutes an addendum to the Closure Report for Corrective Action Unit 403: Second Gas Station, Tonopah Test Range, Nevada, September 1998 as described in the document Supplemental Investigation Report for FFACO Use Restrictions, Nevada Test Site, Nevada (SIR) dated November 2008. The SIR document was approved by NDEP on December 5, 2008. The approval of the SIR document constituted approval of each of the recommended UR removals. In conformance with the SIR document, this addendum consists of: • This page that refers the reader to the SIR document for additional information • The cover, title, and signature pages of the SIR document • The NDEP approval letter • The corresponding section of the SIR document This addendum provides the documentation justifying the cancellation of the UR for CAS 03-02-004-0360, Underground Storage Tanks. This UR was established as part of a Federal Facility Agreement and Consent Order (FFACO) corrective action and is based on the presence of contaminants at concentrations greater than the action levels established at the time of the initial investigation (FFACO, 1996). Since this UR was established, practices and procedures relating to the implementation of risk-based corrective actions (RBCA) have changed. Therefore, this UR was reevaluated against the current RBCA criteria as defined in the Industrial Sites Project Establishment of Final Action Levels (NNSA/NSO, 2006). This re-evaluation consisted of comparing the original data (used to define the need for the UR) to risk-based final action levels (FALs) developed using the current Industrial Sites RBCA process. The re-evaluation resulted in a recommendation to remove the UR because contamination is not present at the site above the risk-based FALs. Requirements for inspecting and maintaining this UR will be canceled, and the postings and signage at this site will be removed. Fencing and posting may be present at this site that are unrelated to the FFACO UR such as for radiological control purposes as required by the NV/YMP Radiological Control Manual (NNSA/NSO, 2004). This modification will not affect or modify any non-FFACO requirements for fencing, posting, or monitoring at this site.

Grant Evenson

2009-05-01T23:59:59.000Z

143

A comprehensive approach to selecting the water chemistry of the secondary coolant circuit in the projects of nuclear power stations equipped with VVER-1200 reactors  

Science Journals Connector (OSTI)

The paper presents the results obtained from studies on selecting the water chemistry of the secondary coolant circuit carried out for the project of a nuclear power station equipped with a new-generation VVER-12...

V. F. Tyapkov

2011-05-01T23:59:59.000Z

144

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

145

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

146

A Test Facility for the International Linear Collider at SLAC End Station A, for Prototypes of Beam Delivery and IR Components  

SciTech Connect

The SLAC Linac can deliver damped bunches with ILC parameters for bunch charge and bunch length to End Station A. A 10Hz beam at 28.5 GeV energy can be delivered there, parasitic with PEP-II operation. We plan to use this facility to test prototype components of the Beam Delivery System and Interaction Region. We discuss our plans for this ILC Test Facility and preparations for carrying out experiments related to collimator wakefields and energy spectrometers. We also plan an interaction region mockup to investigate effects from backgrounds and beam-induced electromagnetic interference.

Woods, M.; Erickson, R.; Frisch, J.; Hast, C.; Jobe, R.K.; Keller, L.; Markiewicz, T.; Maruyama, T.; McCormick, D.; Nelson, J.; Nelson, T.; Phinney, N.; Raubenheimer, T.; Ross, M.; Seryi, A.; Smith, S.; Szalata, Z.; Tenenbaum, P.; Woodley, M.; /SLAC; Angal-Kalinin, D.; Beard, C.; /Daresbury /CERN /DESY /KEK, Tsukuba /LLNL, Livermore /Lancaster U.

2005-05-23T23:59:59.000Z

147

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

148

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

149

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

150

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

151

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

152

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

153

The 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-12) Sheraton Station Square, Pittsburgh, Pennsylvania, U.S.A. September 30-October 4, 2007  

E-Print Network (OSTI)

in the unit square. Figure 1: Physical Domain and Fictitious Domain. simulation of nuclear power plants fromThe 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-12) Sheraton DOMAIN SIMULATIONS FOR THE TWO-PHASE FLOW ENERGY BALANCE OF THE CLOTAIRE STEAM GENERATOR MOCK-UP Michel

Paris-Sud XI, Université de

154

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

155

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

156

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

157

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

158

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

159

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

160

Next Generation Fast RF Interlock Module and ATCA Adapter for ILC High Availability RF Test Station Demonstration  

SciTech Connect

High availability interlocks and controls are required for the ILC (International Linear Collider) L-Band high power RF stations. A new F3 (Fast Fault Finder) VME module has been developed to process both fast and slow interlocks using FPGA logic to detect the interlock trip excursions. This combination eliminates the need for separate PLC (Programmable Logic Controller) control of slow interlocks. Modules are chained together to accommodate as many inputs as needed. In the next phase of development the F3's will be ported to the new industry standard ATCA (Advanced Telecom Computing Architecture) crate (shelf) via a specially designed VME adapter module with IPMI (Intelligent Platform Management Interface). The goal is to demonstrate auto-failover and hot-swap for future partially redundant systems.

Larsen, R

2009-10-17T23:59:59.000Z

Note: This page contains sample records for the topic "reactor testing station" from the National Library of EnergyBeta (NLEBeta).
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161

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

162

On-site field tests for study of low-rank western coal fly ash. Technical summary report, field test No. 3. Big Brown Station electrostatic precipitator  

SciTech Connect

This report describes the results of field and laboratory studies of combined NH/sub 3/ and SO/sub 3/ conditioning at the Big Brown Station of Texas Utilities Generating Company. This unusual combination of conditioning agents is used routinely at the Big Brown Station in order to improve the performance of the cold-side electrostatic precipitators. The primary objectives of this field study were to evaluate the performance of one of the Big Brown precipitators, and to obtain data on the concentration, composition, and size distribution of the fly ash, as well as the composition of the flue gas and the overall and fractional collection efficiencies of the precipitator. The laboratory studies of the Big Brown fly ash were intended to further characterize the ash both physically and chemically, and to study the attenuation of the electrical resistivity of the ash associated with the surface film produced by the dual conditioning process and by the use of SO/sub 3/ conditioning alone. 6 references, 22 figures, 9 tables.

Dahlin, R. S.; Bickelhaupt, R. E.; Marchant, Jr., G. H.; Gooch, J. P.

1984-02-01T23:59:59.000Z

163

Technology, safety and costs of decommissioning a reference boiling water reactor power station: Technical support for decommissioning matters related to preparation of the final decommissioning rule  

SciTech Connect

Preparation of the final Decommissioning Rule by the Nuclear Regulatory Commission (NRC) staff has been assisted by Pacific Northwest Laboratory (PNL) staff familiar with decommissioning matters. These efforts have included updating previous cost estimates developed during the series of studies of conceptually decommissioning reference licensed nuclear facilities for inclusion in the Final Generic Environmental Impact Statement (FGEIS) on decommissioning; documenting the cost updates; evaluating the cost and dose impacts of post-TMI-2 backfits on decommissioning; developing a revised scaling formula for estimating decommissioning costs for reactor plants different in size from the reference boiling water reactor (BWR) described in the earlier study; and defining a formula for adjusting current cost estimates to reflect future escalation in labor, materials, and waste disposal costs. This report presents the results of recent PNL studies to provide supporting information in three areas concerning decommissioning of the reference BWR: updating the previous cost estimates to January 1986 dollars; assessing the cost and dose impacts of post-TMI-2 backfits; and developing a scaling formula for plants different in size than the reference plant and an escalation formula for adjusting current cost estimates for future escalation.

Konzek, G.J.; Smith, R.I.

1988-07-01T23:59:59.000Z

164

Technology, safety and costs of decommissioning a reference pressurized water reactor power station: Technical support for decommissioning matters related to preparation of the final decommissioning rule  

SciTech Connect

Preparation of the final Decommissioning Rule by the Nuclear Regulatory Commission (NRC) staff has been assisted by Pacific Northwest Laboratory (PNL) staff familiar with decommissioning matters. These efforts have included updating previous cost estimates developed during the series of studies on conceptually decommissioning reference licensed nuclear facilities for inclusion in the Final Generic Environmental Impact Statement (FGEIS) on decommissioning; documenting the cost updates; evaluating the cost and dose impacts of post-TMI-2 backfits on decommissioning; developing a revised scaling formula for estimating decommissioning costs for reactor plants different in size from the reference pressurized water reactor (PWR) described in the earlier study; defining a formula for adjusting current cost estimates to reflect future escalation in labor, materials, and waste disposal costs; and completing a study of recent PWR steam generator replacements to determine realistic estimates for time, costs and doses associated with steam generator removal during decommissioning. This report presents the results of recent PNL studies to provide supporting information in four areas concerning decommissioning of the reference PWR: updating the previous cost estimates to January 1986 dollars; assessing the cost and dose impacts of post-TMI-2 backfits; assessing the cost and dose impacts of recent steam generator replacements; and developing a scaling formula for plants different in size than the reference plant and an escalation formula for adjusting current cost estimates for future escalation.

Konzek, G.J.; Smith, R.I.

1988-07-01T23:59:59.000Z

165

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

166

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

167

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

168

Testing the technical state of gas-transferring unit compressor cases and their repair under the conditions of compressor stations  

Science Journals Connector (OSTI)

Problems related to the use of nondestructive testing methods during engineering diagnostics of the cast cases of compressors that are used in gas-transfer units that operate in gas-main pipelines are considered....

I. I. Kryukov; N. A. Kalinin; S. A. Leont’ev…

2011-02-01T23:59:59.000Z

169

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

170

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

171

Corrective Action Decision Document for Corrective Action Unit 490: Station 44 Burn Area, Tonopah Test Range, Nevada (Rev. No.: 0, February 2001)  

SciTech Connect

This Corrective Action Decision Document identifies and rationalizes the U.S. Department of Energy, Nevada Operations Office's selection of a recommended Corrective Action Alternative (CAA) appropriate to facilitate the closure of Corrective Action Unit (CAU) 490, Station 44 Burn Area, Tonopah Test Range (TTR), Nevada, under the Federal Facility Agreement and Consent Order. Corrective Action Unit 490 is located on the Nellis Air Force Range and the Tonopah Test Range and is approximately 140 miles northwest of Las Vegas, Nevada. This CAU is comprised of four Corrective Action Sites (CASs): 03-56-001-03BA, Fire Training Area (located southwest of Area 3); RG-56-001-RGBA, Station 44 Burn Area (located west of Main Lake); 03-58-001-03FN, Sandia Service Yard (located north of the northwest corner of Area 3); and 09-54-001-09L2, Gun Propellant Burn Area (located south of the Area 9 Compound on the TTR). A Corrective Action Investigation was performed in July and August 2000, and analytes detected during the corrective action investigation were evaluated against preliminary action levels to determine contaminants of concern (COCs). There were no COCs identified in soil at the Gun Propellant Burn Area or the Station 44 Burn Area; therefore, there is no need for corrective actions at these two sites. Five soil samples at the Fire Training Area and seven at the Sandia Service Yard exceeded PALs for total petroleum hydrocarbons-diesel. Upon the identification of COCs specific to CAU 490, Corrective Action Objectives were developed based on a review of existing data, future use, and current operations at the TTR, with the following three CAAs under consideration: Alternative 1 - No Further Action, Alternative 2 - Closure In Place - No Further Action With Administrative Controls, and Alternative 3 - Clean Closure by Excavation and Disposal. These alternatives were evaluated based on four general corrective action standards and five remedy selection decision factors. Based on the results of this evaluation, the preferred choice for CAU 490 was Alternative 3. This alternative was judged to meet all requirements for the technical components evaluated, all applicable state and federal regulations for closure of the site, and will eliminate potential future exposure pathways to the contaminated soils at this site.

DOE /NV

2001-02-23T23:59:59.000Z

172

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

173

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

174

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

175

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

176

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

177

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

178

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

179

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

180

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

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

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

182

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

183

Geologic and hydrologic records of observation wells, test holes, test wells, supply wells, springs, and surface water stations in the Los Alamos area  

SciTech Connect

Hundreds of holes have been drilled into the Pajarito Plateau and surrounding test areas of the Los Alamos National Laboratory since the end of World War II. They range in depth from a few feet to more than 14,000 ft. The holes were drilled to provide geologic, hydrologic, and engineering information related to development of a water supply, to provide data on the likelihood or presence of subsurface contamination from hazardous and nuclear materials, and for engineering design for construction. The data contained in this report provide a basis for further investigations into the consequences of our past, present, and future interactions with the environment.

Purtymun, W.D.

1995-01-01T23:59:59.000Z

184

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

185

Corrective Action Investigation Plan for Corrective Action Unit 490: Station 44 Burn Area, Tonopah Test Range, Nevada (with Record of Technical Change No.1)  

SciTech Connect

This Corrective Action Investigation Plan (CAIP) contains the U.S. Department of Energy, Nevada Operations Office's approach to collect the data necessary to evaluate corrective action alternatives appropriate for the closure of Corrective Action Unit (CAU) 490 under the Federal Facility Agreement and Consent Order. Corrective Active Unit 490 consists of four Corrective Action Sites (CASs): 03-56-001-03BA, Fire Training Area (FTA); RG-56-001-RGBA, Station 44 Burn Area; 03-58-001-03FN, Sandia Service Yard; and 09-54-001-09L2, Gun Propellant Burn Area. These CASs are located at the Tonopah Test Range near Areas 3 and 9. Historically, the FTA was used for training exercises where tires and wood were ignited with diesel fuel. Records indicate that water and carbon dioxide were the only extinguishing agents used during these training exercises. The Station 44 Burn Area was used for fire training exercises and consisted of two wooden structures. The two burn areas (ignition of tires, wood, and wooden structures with diesel fuel and water) were limited to the building footprints (10 ft by 10 ft each). The Sandia Service Yard was used for storage (i.e., wood, tires, metal, electronic and office equipment, construction debris, and drums of oil/grease) from approximately 1979 to 1993. The Gun Propellant Burn Area was used from the 1960s to 1980s to burn excess artillery gun propellant, solid-fuel rocket motors, black powder, and deteriorated explosives; additionally, the area was used for the disposal of experimental explosive items. Based on site history, the focus of the field investigation activities will be to: (1) determine the presence of contaminants of potential concern (COPCs) at each CAS, (2) determine if any COPCs exceed field-screening levels and/or preliminary action levels, and (3) determine the nature and extent of contamination with enough certainty to support selection of corrective action alternatives for each CAS. The scope of this CAIP is to resolve the question of whether or not potentially hazardous wastes were generated at three of the four CASs within CAU 490, and whether or not potentially hazardous and radioactive wastes were generated at the fourth CAS in CAU 490 (CAS 09-54-001-09L2). Suspected CAS-specific COPCs include volatile organic compounds, semivolatile organic compounds, total petroleum hydrocarbons, polychlorinated biphenyls, pesticides, explosives, and uranium and plutonium isotopes. The results of this field investigation will support a defensible evaluation of corrective action alternatives in the corrective action decision document.

U.S. Department of Energy, Nevada Operations Office

2000-06-09T23:59:59.000Z

186

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

187

Modular Pebble Bed Reactor High Temperature Gas Reactor  

E-Print Network (OSTI)

Modular Pebble Bed Reactor High Temperature Gas Reactor Andrew C Kadak Massachusetts Institute For 1150 MW Combined Heat and Power Station Oil Refinery Hydrogen Production Desalinization Plant VHTR/Graphite Discrimination system Damaged Sphere ContainerGraphiteReturn FuelReturn Fresh Fuel Container Spent Fuel Tank #12

188

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

189

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

190

Operational Experience in Nuclear Power Stations [and Discussion  

Science Journals Connector (OSTI)

...Operational Experience in Nuclear Power Stations...self-sustaining nuclear reaction to the present...time large-scale generation of electrical power from nuclear energy has become...the C.E.G.B. reactors have been in service...

1974-01-01T23:59:59.000Z

191

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

192

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

193

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

194

Allowable Residual Contamination Levels in soil for decommissioning the Shippingport Atomic Power Station site  

SciTech Connect

As part of decommissioning the Shippingport Atomic Power Station, a fundamental concern is the determination of Allowable Residual Contamination Levels (ARCL) for radionuclides in the soil at the site. The ARCL method described in this report is based on a scenario/exposure-pathway analysis and compliance with an annual dose limit for unrestricted use of the land after decommissioning. In addition to naturally occurring radionuclides and fallout from weapons testing, soil contamination could potentially come from five other sources. These include operation of the Shippingport Station as a pressurized water reactor, operations of the Shippingport Station as a light-water breeder, operation of the nearby Beaver Valley reactors, releases during decommissioning, and operation of other nearby industries, including the Bruce-Mansfield coal-fired power plants. ARCL values are presented for 29 individual radionculides and a worksheet is provided so that ARCL values can be determined for any mixture of the individual radionuclides for any annual dose limit selected. In addition, a worksheet is provided for calculating present time soil concentration value that will decay to the ARCL values after any selected period of time, such as would occur during a period of restricted access. The ARCL results are presented for both unconfined (surface) and confined (subsurface) soil contamination. The ARCL method and results described in this report provide a flexible means of determining unrestricted-use site release conditions after decommissioning the Shippingport Atomic Power Station.

Kennedy, W.E. Jr.; Napier, B.A.; Soldat, J.K.

1983-09-01T23:59:59.000Z

195

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

196

Active dc filter for HVDC system--A test installation in the Konti-Skan DC link at Lindome converter station  

SciTech Connect

The purpose of introducing active dc filters is to meet the more and more stringent requirement from power utilities on limiting telephone interference caused by harmonic currents from HVdc transmission lines, without unnecessarily increasing the cost of HVdc stations. An active dc filter installed in the Konti-Skan HVdc link is described. The active dc filter is connected at the bottom of an existing passive dc filter at the Lindome station. The active dc filter includes optic harmonic current measuring unit, control system, protection and supervision system, PWM power amplifier, high-frequency transformer, surge arrester, and coupling apparatuses. The active dc filter has small physical size and occupies small ground area. The performance of the active dc filter for eliminating the disturbing harmonics is excellent. To achieve comparable results by passive filters would require something like ten times more high voltage equipment.

Zhang, Wenyan; Asplund, G. (ABB Power Systems, Ludvika (Sweden). HVDC Division); Aberg, A. (ABB Corporate Research, Lund (Sweden). Dept. of Man-Machine Communication); Jonsson, U. (Svenska Kraftnaet, Vaellingby (Sweden)); Loeoef, O. (Vattenfall AB, Trollhaettan (Sweden). Region Vaestsverige)

1993-07-01T23:59:59.000Z

197

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

198

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

199

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

200

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

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

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

202

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

203

SOUTH STATION TAmtrak, Commuter  

E-Print Network (OSTI)

SOUTH STATION TAmtrak, Commuter Rail, Bus Station, MBTA Red Line Knapp St. Kneeland St. Stuart St) T BOYLSTON (MBTA Green Line) T NEW ENGLAND MEDICAL CENTER (MBTA Orange Line) Boston Campus Accessibility Map

Dennett, Daniel

204

wind power station  

Science Journals Connector (OSTI)

wind power station [It may consist of just one wind turbine or a network of windmills] ? Windkraftanlage

2014-08-01T23:59:59.000Z

205

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

206

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

207

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

208

Early Station Costs Questionnaire  

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

Early Station Costs Questionnaire Early Station Costs Questionnaire Marc Melaina Hydrogen Technologies and Systems Center Market Readiness Workshop February 16-17th, 2011 Washington, DC Questionnaire Goals * The Early Station Costs questionnaire provides an anonymous mechanism for organizations with direct experience with hydrogen station costs to provide feedback on current costs, near-term costs, economies of scale, and R&D priorities. * This feedback serves the hydrogen community and government agencies by increasing awareness of the status of refueling infrastructure costs National Renewable Energy Laboratory Innovation for Our Energy Future Questions for Market Readiness Workshop Attendees * Are these questions the right ones to be asking?

209

Hot Plate Station  

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

temperature is limited to 200C in order to maintain temperature inside the cleanroom. A hood located over the hot plate station ensures evaporated fumes are not released...

210

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

211

Daily Reporting Rainfall Station HERBERT RIVER Manual Heavy Rainfall Station  

E-Print Network (OSTI)

Daily Reporting Rainfall Station HERBERT RIVER Manual Heavy Rainfall Station Manual River Station Central Mill AL Tung Oil AL Corsis AL Innisfail Clump Point Tide TM Mourilyan Harbour TM 0 10 kilometres

Greenslade, Diana

212

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

213

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

214

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

215

Refueling Stations | Open Energy Information  

Open Energy Info (EERE)

Jump to: navigation, search TODO: Add description List of Refueling Stations Incentives Retrieved from "http:en.openei.orgwindex.php?titleRefuelingStations&oldid267188...

216

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...

217

Timber Mountain Precipitation Monitoring Station  

SciTech Connect

A precipitation monitoring station was placed on the west flank of Timber Mountain during the year 2010. It is located in an isolated highland area near the western border of the Nevada National Security Site (NNSS), south of Pahute Mesa. The cost of the equipment, permitting, and installation was provided by the Environmental Monitoring Systems Initiative (EMSI) project. Data collection, analysis, and maintenance of the station during fiscal year 2011 was funded by the U.S. Department of Energy, National Nuclear Security Administration, Nevada Site Office Environmental Restoration, Soils Activity. The station is located near the western headwaters of Forty Mile Wash on the Nevada Test and Training Range (NTTR). Overland flows from precipitation events that occur in the Timber Mountain high elevation area cross several of the contaminated Soils project CAU (Corrective Action Unit) sites located in the Forty Mile Wash watershed. Rain-on-snow events in the early winter and spring around Timber Mountain have contributed to several significant flow events in Forty Mile Wash. The data from the new precipitation gauge at Timber Mountain will provide important information for determining runoff response to precipitation events in this area of the NNSS. Timber Mountain is also a groundwater recharge area, and estimation of recharge from precipitation was important for the EMSI project in determining groundwater flowpaths and designing effective groundwater monitoring for Yucca Mountain. Recharge estimation additionally provides benefit to the Underground Test Area Sub-project analysis of groundwater flow direction and velocity from nuclear test areas on Pahute Mesa. Additionally, this site provides data that has been used during wild fire events and provided a singular monitoring location of the extreme precipitation events during December 2010 (see data section for more details). This letter report provides a summary of the site location, equipment, and data collected in fiscal year 2011.

Lyles Brad,McCurdy Greg,Chapman Jenny,Miller Julianne

2012-01-01T23:59:59.000Z

218

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

219

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

220

INDEPENDENT CONFIRMATORY SURVEY REPORT FOR THE REACTOR BUILDING, HOT LABORATORY, PRIMARY PUMP HOUSE, AND LAND AREAS AT THE PLUM BROOK REACTOR FACILITY, SANDUSKY, OHIO  

SciTech Connect

In 1941, the War Department acquired approximately 9,000 acres of land near Sandusky, Ohio and constructed a munitions plant. The Plum Brook Ordnance Works Plant produced munitions, such as TNT, until the end of World War II. Following the war, the land remained idle until the National Advisory Committee for Aeronautics later called the National Aeronautics and Space Administration (NASA) obtained 500 acres to construct a nuclear research reactor designed to study the effects of radiation on materials used in space flight. The research reactor was put into operation in 1961 and was the first of fifteen test facilities eventually built by NASA at the Plum Brook Station. By 1963, NASA had acquired the remaining land at Plum Brook for these additional test facilities

Erika N. Bailey

2011-10-10T23:59:59.000Z

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

Revised analyses of decommissioning for the reference boiling water reactor power station. Effects of current regulatory and other considerations on the financial assurance requirements of the decommissioning rule and on estimates of occupational radiation exposure - appendices. Final report  

SciTech Connect

The NRC staff is in need of decommissioning bases documentation that will assist them in assessing the adequacy of the licensee submittals, from the viewpoint of both the planned actions, including occupational radiation exposure, and the probable costs. The purpose of this reevaluation study is to update the needed bases documentation. This report presents the results of a review and reevaluation of the PNL 1980 decommissioning study of the Washington Public Power Supply System`s Washington Nuclear Plant Two (WNP-2) located at Richland, Washington, including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the plant for the DECON, SAFSTOR, and ENTOMB decommissioning alternatives. These alternatives now include an initial 5-7 year period during which time the spent fuel is stored in the spent fuel pool prior to beginning major disassembly or extended safe storage of the plant. Included for information (but not presently part of the license termination cost) is an estimate of the cost to demolish the decontaminated and clear structures on the site and to restore the site to a {open_quotes}green field{close_quotes} condition. This report also includes consideration of the NRC requirement that decontamination and decommissioning activities leading to termination of the nuclear license be completed within 60 years of final reactor shutdown, consideration of packaging and disposal requirements for materials whose radionuclide concentrations exceed the limits for Class C low-level waste (i.e., Greater-Than-Class C), and reflects 1993 costs for labor, materials, transport, and disposal activities. Sensitivity of the total license termination cost to the disposal costs at different low-level radioactive waste disposal sites, to different depths of contaminated concrete surface removal within the facilities, and to different transport distances is also examined.

Smith, R.I.; Bierschbach, M.C.; Konzek, G.J.; McDuffie, P.N.

1996-07-01T23:59:59.000Z

222

DESIGNING AN OPPORTUNITY FUEL WITH BIOMASS AND TIRE-DERIVED FUEL FOR COFIRING AT WILLOW ISLAND GENERATING STATION  

SciTech Connect

During the period January 1, 2001-March 31, 2001, Allegheny Energy Supply Co., LLC (Allegheny) finalized the engineering of the Willow Island cofiring project, completed the fuel characterizations for both the Willow Island and Albright Generating Station projects, and initiated construction of both projects. Allegheny and its contractor, Foster Wheeler, selected appropriate fuel blends and issued purchase orders for all processing and mechanical equipment to be installed at both sites. This report summarizes the activities associated with the Designer Opportunity Fuel program, and demonstrations at Willow Island and Albright Generating Stations. The third quarter of the project involved completing the detailed designs for the Willow Island Designer Fuel project. It also included complete characterization of the coal and biomass fuels being burned, focusing upon the following characteristics: proximate and ultimate analysis; higher heating value; carbon 13 nuclear magnetic resonance testing for aromaticity, number of aromatic carbons per cluster, and the structural characteristics of oxygen in the fuel; drop tube reactor testing for high temperature devolatilization kinetics and generation of fuel chars; thermogravimetric analyses (TGA) for char oxidation kinetics; and related testing. The construction at both sites commenced during this quarter, and was largely completed at the Albright Generating Station site.

K. Payette; D. Tillman

2001-04-01T23:59:59.000Z

223

Mobile Alternative Fueling Station Locator  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

Alternative Fueling Station Locator Alternative Fueling Station Locator Fuel Type Biodiesel (B20 and above) Compressed Natural Gas Electric Ethanol (E85) Hydrogen Liquefied Natural Gas (LNG) Liquefied Petroleum Gas (Propane) Location Enter a city, postal code, or address Include private stations Not all stations are open to the public. Choose this option to also search private fueling stations. Search Caution: The AFDC recommends that users verify that stations are open, available to the public, and have the fuel prior to making a trip to that location. Some stations in our database have addresses that could not be located by the Station Locator application. This may result in the station appearing in the center of the zip code area instead of the actual location. If you're having difficulty, please contact the technical response team at

224

Princeton Station Lewis  

E-Print Network (OSTI)

Dean Mathey Princeton Hospital Forrestal AOS PPPL Millstone Apts Dean Mathey Icahn 8:00 AM Mathey Princeton Hospital Forrestal AOS PPPL Millstone Apts Dean Mathey Icahn Order of Stops: Princeton Station; Lewis Lirbary; Dean Mathey; Princeton Hospital Forrestal AOS; PPPL

Singh, Jaswinder Pal

225

Ann Arbor Seismograph Station  

Science Journals Connector (OSTI)

...MEMBERSHIP COHM ITTEE Ross R. Heinrich...n~bers and changes or address should...seismograms, (2) a world-wide...Utah a 0.3 Bonneville lake beds Scoresbysund...Alaska 1.1 2.0 Graywacke...Stores, 221 North Grand Boulevard...SEISM and the name of the station...day of month (2) ml n2 sl s2...

James T. Wilson

226

2014 Bull Evaluation Station Senior Bull Test  

E-Print Network (OSTI)

Duck River, TN 38454 2200 1 45 AN Joe Faulkner 4795 Ash Hill Road Spring Hill, TN 37174 2750 1 47, TN 37091 3800 1 10 AN W & D Farms Inc. 699 White Tail Road Holladay, TN 38341 5000 1 12 AN Douglas Grove, TN 37046 4750 1 17 AN Larry James 3495 Joe Yates Rd Obion, TN 38240 4500 1 20 AN Jason Sommers

Wang, Xiaorui "Ray"

227

A Fuel-Cell Vehicle Test Station.  

E-Print Network (OSTI)

??Due to concerns about energy security, rising oil prices, and adverse effects of internal combustion engine vehicles on the environment, the automotive industry is quickly… (more)

Thorne, Michelle I

2008-01-01T23:59:59.000Z

228

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

229

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

230

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

231

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

232

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

233

Early Site Permit Demonstration Program: Station design alternatives report  

SciTech Connect

This report provides the results of investigating the basis for including Station Design Alternatives (SDAs) in the regulatory guidance given for nuclear plant environmental reports (ERs), explains approaches or processes for evaluating SDAs at the early site permit (ESP) stage, and applies one of the processes to each of the ten systems or subsystems considered as SDAS. The key objective o this report s to demonstrate an adequate examination of alternatives can be performed without the extensive development f design data. The report discusses the Composite Suitability Approach and the Established Cutoff Approach in evaluating station design alternatives and selects one of these approaches to evaluate alternatives for each of the plant or station that were considered. Four types of ALWRs have been considered due to the availability of extensive plant data: System 80+, AP600, Advanced Boiling Reactor (ABWR), and Simplified Boiling Water Reactor (SBWR). This report demonstrates the feasibility of evaluating station design alternatives when reactor design detail has not been determined, quantitatively compares the potential ental impacts of alternatives, and focuses the ultimate selection of a alternative on cost and applicant-specific factors. The range of alternatives system is deliberately limited to a reasonable number to demonstrate the or to the three most commonly used at operating plants.

Not Available

1993-03-01T23:59:59.000Z

234

Quantitative Analysis of Station Hydrogen  

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

Analysis of Station Analysis of Station Hydrogen * Role of ENAA (Engineering Advancement Association of Japan) - To manage the construction and operation of hydrogen stations in national project, JHFC Project - To act as secretariat of ISO/TC197 (Hydrogen technologies) committee of Japan Kazuo Koseki Chief Secretary of ISO/TC197 of Japan ENAA Yokohama Daikoku Station (Desulfurized Gasoline) Yokohama Asahi Station (Naphtha) Senju Station (LPG) Kawasaki Station (Methanol) Yokohama Asahi Station Naphtha PSA Compressor Storage Tanks Dispenser Reformer Buffer Tank 25 MPa 35 MPa 1073 K 0.8 MPa Inlet : 0.6 MPa Outlet : 40 MPa Vent Stack 40 MPa Result of Quantitative Analysis Concentration. vol.ppm Min.Detect Analysis Impurity Gasoline Naphtha LPG Methanol Conc. Method CO 0.05 0.06 0.02 0.06 0.01 GC-FID

235

Development of a Renewable Hydrogen Energy Station  

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

Development of a Renewable Development of a Renewable Hydrogen Energy Station Edward C. Heydorn - Air Products and Chemicals, Inc. Pinakin Patel - FuelCell Energy, Inc. Fred Jahnke - FuelCell Energy, Inc. "Delivering Renewable Hydrogen - A Focus on Near-Term Applications" Palm Springs, CA 16 November 2009 Presentation Outline * Hydrogen Energy Station Technology Overview * Process Description * Performance and Economic Parameters * Proposed Demonstration on Renewable Feedstock * Status of Shop Validation Test * Conclusion 2 Objectives * Determine the economic and technical viability of a hydrogen energy station designed to co-produce power and hydrogen Utilize technology development roadmap to provide deliverables and go/no-go decision

236

Station Instruction Cooperative  

E-Print Network (OSTI)

$0 INSTITUTE OF MUSEUM/LIBRARY SV $0 $0 $0 $135,575 $0 $135,575 NATIONAL AERO & SPACE ADMIN $971General Research Experiment Station Instruction Public Service Cooperative Extension Total $96,610,455 23,968,233 14,198,142 ARRA 6,188,801 Public Service (PS) 33,400,117 45,534,482 34,497,464 39

Arnold, Jonathan

237

Pacific Southwest Research Station Publications  

E-Print Network (OSTI)

Bulletins 1- 28 1965-1989 Soil- Vegetation Tables -- 1965-1980 Solar Irradiation and Shadow Length Tables-1971 1938 California Forest and Range Experiment Station Annual Report 1939 California Forest and Range Experiment Station Annual Report 1940 California Forest and Range Experiment Station Annual Report 1951

Standiford, Richard B.

238

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.

239

Transport reactor development status  

SciTech Connect

This project is part of METC`s Power Systems Development Facility (PSDF) located at Wilsonville, Alabama. The primary objective of the Advanced Gasifier module is to produce vitiated gases for intermediate-term testing of Particulate Control Devices (PCDs). The Transport reactor potentially allows particle size distribution, solids loading, and particulate characteristics in the off-gas stream to be varied in a number of ways. Particulates in the hot gases from the Transport reactor will be removed in the PCDs. Two PCDs will be initially installed in the module; one a ceramic candle filter, the other a granular bed filter. After testing of the initial PCDs they will be removed and replaced with PCDs supplied by other vendors. A secondary objective is to verify the performance of a Transport reactor for use in advanced Integrated Gasification Combined Cycle (IGCC), Integrated Gasification Fuel Cell (IG-FC), and Pressurized Combustion Combined Cycle (PCCC) power generation units. This paper discusses the development of the Transport reactor design from bench-scale testing through pilot-scale testing to design of the Process Development Unit (PDU-scale) facility at Wilsonville.

Rush, R.E.; Fankhanel, M.O.; Campbell, W.M.

1994-10-01T23:59:59.000Z

240

NETL - Chemical Looping Reactor  

ScienceCinema (OSTI)

NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

None

2014-06-26T23:59:59.000Z

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

Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel  

SciTech Connect

The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (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 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±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; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

2011-03-01T23:59:59.000Z

242

Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel  

SciTech Connect

The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (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 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±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; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

2014-03-01T23:59:59.000Z

243

WCH Removes Massive Test Reactor  

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...

244

Potassium-Modified Iron Ore as Oxygen Carrier for Coal Chemical Looping Combustion: Continuous Test in 1 kW Reactor  

Science Journals Connector (OSTI)

(20) Briefly, it consists of a fast fluidized bed as an air reactor, a cyclone, a spout-fluid bed as a fuel reactor, and an external loop-seal. ... Industrial & Engineering Chemistry Research (2013), 52 (28), 9573-9585 CODEN: IECRED; ISSN:0888-5885. ... Gu, H.; Shen, L.; Xiao, J.; Zhang, S.; Song, T.; Chen, D.Iron ore as oxygen carrier improved with potassium for chemical looping combustion of anthracite coal Combust. ...

Haiming Gu; Laihong Shen; Zhaoping Zhong; Xin Niu; Huijun Ge; Yufei Zhou; Shen Xiao

2014-07-30T23:59:59.000Z

245

SUPPLEMENT ANALYSIS OF FOREIGN RESEARCH REACTOR srENT NUCLEAR FUEL  

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

FOREIGN RESEARCH REACTOR srENT NUCLEAR FUEL FOREIGN RESEARCH REACTOR srENT NUCLEAR FUEL TRANSPORTATION ALONG OTHER THAN~. PRESENTATIVE ROUTE FROM CONCORD NAVAL WEAPO~~ STATION TO IDAHO NATIONAL ENGINEERING AND ENVIRONMENTAL LADORA TORY Introduction The Department of Energy is planning to transport foreign research reactor spent nuclear fuel by rail from the Concord Naval Weapons Station (CNWS), Concord, California, to the Idaho National Engineering and Environmental Laboratory (INEEL). The environmental analysis supporting the decision to transport, by rail or truck, foreign research reactor spent nuclear fuel from CNWS to the INEEL is contained in +he Final Environmental Impact Statement on a Proposed Nuclear Weapons Nonproliftration Policy Concerning Foreign Research Reactor

246

Light Water Reactor Sustainability Constellation Pilot Project FY11 Summary Report  

SciTech Connect

Summary report for Fiscal Year 2011 activities associated with the Constellation Pilot Project. The project is a joint effor between Constellation Nuclear Energy Group (CENG), EPRI, and the DOE Light Water Reactor Sustainability Program. The project utilizes two CENG reactor stations: R.E. Ginna and Nine Point Unit 1. Included in the report are activities associate with reactor internals and concrete containments.

R. Johansen

2011-09-01T23:59:59.000Z

247

Alternative Fuels Data Center: Ethanol Fueling Stations  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

Fueling Fueling Stations to someone by E-mail Share Alternative Fuels Data Center: Ethanol Fueling Stations on Facebook Tweet about Alternative Fuels Data Center: Ethanol Fueling Stations on Twitter Bookmark Alternative Fuels Data Center: Ethanol Fueling Stations on Google Bookmark Alternative Fuels Data Center: Ethanol Fueling Stations on Delicious Rank Alternative Fuels Data Center: Ethanol Fueling Stations on Digg Find More places to share Alternative Fuels Data Center: Ethanol Fueling Stations on AddThis.com... More in this section... Ethanol Basics Benefits & Considerations Stations Locations Infrastructure Development Vehicles Laws & Incentives Ethanol Fueling Stations Photo of an ethanol fueling station. Thousands of ethanol fueling stations are available in the United States.

248

Alternative Fuels Data Center: Hydrogen Fueling Stations  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

Fueling Fueling Stations to someone by E-mail Share Alternative Fuels Data Center: Hydrogen Fueling Stations on Facebook Tweet about Alternative Fuels Data Center: Hydrogen Fueling Stations on Twitter Bookmark Alternative Fuels Data Center: Hydrogen Fueling Stations on Google Bookmark Alternative Fuels Data Center: Hydrogen Fueling Stations on Delicious Rank Alternative Fuels Data Center: Hydrogen Fueling Stations on Digg Find More places to share Alternative Fuels Data Center: Hydrogen Fueling Stations on AddThis.com... More in this section... Hydrogen Basics Benefits & Considerations Stations Locations Infrastructure Development Vehicles Laws & Incentives Hydrogen Fueling Stations Photo of a hydrogen fueling station. A handful of hydrogen fueling stations are available in the United States

249

Alternative Fuels Data Center: Biodiesel Fueling Stations  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

Fueling Fueling Stations to someone by E-mail Share Alternative Fuels Data Center: Biodiesel Fueling Stations on Facebook Tweet about Alternative Fuels Data Center: Biodiesel Fueling Stations on Twitter Bookmark Alternative Fuels Data Center: Biodiesel Fueling Stations on Google Bookmark Alternative Fuels Data Center: Biodiesel Fueling Stations on Delicious Rank Alternative Fuels Data Center: Biodiesel Fueling Stations on Digg Find More places to share Alternative Fuels Data Center: Biodiesel Fueling Stations on AddThis.com... More in this section... Biodiesel Basics Benefits & Considerations Stations Locations Infrastructure Development Vehicles Laws & Incentives Biodiesel Fueling Stations Photo of a biodiesel fueling station. Hundreds of biodiesel fueling stations are available in the United States.

250

Nuclear reactor characteristics and operational history  

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

Nuclear > U.S. reactor operation status tables Nuclear > U.S. reactor operation status tables Nuclear Reactor Operational Status Tables Release date: November 22, 2011 Next release date: November 2012 See also: Table 2. Ownership Data, Table 3. Characteristics and Operational History Table 1. Nuclear Reactor, State, Type, Net Capacity, Generation, and Capacity Factor PDF XLS Plant/Reactor Name Generator ID State Type 2009 Summer Capacity Net MW(e)1 2010 Annual Generation Net MWh2 Capacity Factor Percent3 Arkansas Nuclear One 1 AR PWR 842 6,607,090 90 Arkansas Nuclear One 2 AR PWR 993 8,415,588 97 Beaver Valley 1 PA PWR 892 7,119,413 91 Beaver Valley 2 PA PWR 885 7,874,151 102 Braidwood Generation Station 1 IL PWR 1,178 9,196,689 89

251

naval reactors  

National Nuclear Security Administration (NNSA)

After operating for 34 years and training over 14,000 sailors, the Department of Energy S1C Prototype Reactor Site in Windsor, Connecticut, was returned to "green field"...

252

Thermionic Reactor Design Studies  

SciTech Connect

Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their (thermionic reactor) 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.

Schock, Alfred

1994-08-01T23:59:59.000Z

253

REACTOR GROUT THERMAL PROPERTIES  

SciTech Connect

Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.

2011-01-28T23:59:59.000Z

254

Hydrogen Station Cost Estimates: Comparing Hydrogen Station Cost...  

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

classification) NREL National Renewable Energy Laboratory RFP request for proposals SMR steam methane reformer SOTA state of the art (referring to station classification) TAR...

255

Research reactors - an overview  

SciTech Connect

A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.

West, C.D.

1997-03-01T23:59:59.000Z

256

Station design using orifice meters  

SciTech Connect

This paper reports that proper meter station design using gas orifice meters must include consideration of a number of factors to minimize operation and maintenance problems while obtaining the best accuracy over the life of the station. A station should provide accuracy, be safe, functional, economical and free of undue maintenance. It should comply with all codes, reports and specifications but most of all it should be able to comply with terms set forth in the contract. All measuring stations should be designed with considerations of growth or reductions in volume. It should be attractive and built to last for many years.

Upp, E.L. (Daniel Industries, Inc., Houston, TX (United States))

1992-07-01T23:59:59.000Z

257

Celilo Converter Station - October 2005  

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

safer and ecologi- cally improved. The new thyristors eliminate reliance on mercury for power conversion. Asbestos has been removed from the station. Old, noisy, maintenance-in-...

258

Hanford Meteorological Station - Hanford Site  

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

Station Real Time Met Data from Around the Site Current HMS Observations Daily HMS Extremes in Met Data Met and Climate Data Summary Products Contacts Hours Current NWS...

259

Light Water Reactor Sustainability  

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

4 Light Water Reactor Sustainability ACCOMPLISHMENTS REPORT 2014 Accomplishments Report | Light Water Reactor Sustainability 2 T he mission of the Light Water Reactor...

260

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

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

Catalytic reactor  

DOE Patents (OSTI)

A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

Aaron, Timothy Mark (East Amherst, NY); Shah, Minish Mahendra (East Amherst, NY); Jibb, Richard John (Amherst, NY)

2009-03-10T23:59:59.000Z

262

An Assessment of the Near-Term Costs of Hydrogen Refueling Stations and Station Components  

E-Print Network (OSTI)

480 kg/day natural gas reformation station. The table belowReciprocating gas compressor Electrolyzer Station: Thisfor reformer-type stations (natural gas), however, is more

Lipman, T E; Weinert, Jonathan X.

2006-01-01T23:59:59.000Z

263

Fuel Station Procedure Applicability All  

E-Print Network (OSTI)

Fuel Station Procedure Applicability All Last Revised 11/20/12 Procedure Owner Andrew Grant agrant for the purchasing and distribution of fuel for vehicles owned by Bowling Green State University (BGSU). This centralization is important to ensure compliance for BGSU employees who use the centralized fuel station and fuel

Moore, Paul A.

264

Station Costs Pinch Other Programs  

Science Journals Connector (OSTI)

...station. The overruns, encountered by Boeing Co. in its role as general contractor, mean...And he promised a skeptical panel that Boeing has the overruns under control...station. The overruns, encountered by Boeing Co. in its role as general contractor...

Andrew Lawler

1997-09-26T23:59:59.000Z

265

station locations | OpenEI  

Open Energy Info (EERE)

00 00 Varnish cache server Browse Upload data GDR 429 Throttled (bot load) Error 429 Throttled (bot load) Throttled (bot load) Guru Meditation: XID: 2142288500 Varnish cache server station locations Dataset Summary Description Alternative fueling stations are located throughout the United States and their availability continues to grow. The Alternative Fuels Data Center (AFDC) maintains a website where you can find alternative fuels stations near you or on a route, obtain counts of alternative fuels stations by state, Source Alternative Fuels Data Center Date Released December 13th, 2010 (4 years ago) Date Updated December 13th, 2010 (4 years ago) Keywords alt fuel alternative fuels alternative fuels stations biodiesel CNG compressed natural gas E85 Electricity ethanol

266

High Flux Isotope Reactor system RELAP5 input model  

SciTech Connect

A thermal-hydraulic computational model of the High Flux Isotope Reactor (HFIR) has been developed using the RELAP5 program. The purpose of the model is to provide a state-of-the art thermal-hydraulic simulation tool for analyzing selected hypothetical accident scenarios for a revised HFIR Safety Analysis Report (SAR). The model includes (1) a detailed representation of the reactor core and other vessel components, (2) three heat exchanger/pump cells, (3) pressurizing pumps and letdown valves, and (4) secondary coolant system (with less detail than the primary system). Data from HFIR operation, component tests, tests in facility mockups and the HFIR, HFIR specific experiments, and other pertinent experiments performed independent of HFIR were used to construct the model and validate it to the extent permitted by the data. The detailed version of the model has been used to simulate loss-of-coolant accidents (LOCAs), while the abbreviated version has been developed for the operational transients that allow use of a less detailed nodalization. Analysis of station blackout with core long-term decay heat removal via natural convection has been performed using the core and vessel portions of the detailed model.

Morris, D.G.; Wendel, M.W.

1993-01-01T23:59:59.000Z

267

Development and validation of a real-time SAFT-UT (synthetic aperture focusing technique for ultrasonic testing) system for the inspection of light water reactor components: Annual report, October 1985-September 1986  

SciTech Connect

The Pacific Northwest Laboratory is working to design, fabricate, and evaluate a real-time flaw detection and characterization system based on the synthetic aperture focusing technique for ultrasonic testing (SAFT-UT). The system is designed to perform inservice inspection of light-water reactor components. Included objectives of this program for the Nuclear Regulatory Commission are to develop procedures for system calibration and field operation, to validate the system through laboratory and field inspections, and to generate an engineering data base to support ASME Code acceptance of the technology. This progress report covers the programmatic work from October 1985 through September 1986. 45 figs., 8 tabs.

Doctor, S.R.; Hall, T.E.; Reid, L.D.; Mart, G.A.

1987-07-01T23:59:59.000Z

268

FNS Presentation - Hydrogen Station & Hydrogen ICE Vehicles Operation  

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

Hydrogen Station & Hydrogen ICE Hydrogen Station & Hydrogen ICE Vehicle Operations Federal Network for Sustainability Idaho Falls, Idaho - July 2006 Jim Francfort INL/CON-06-11569 Presentation Outline * Background & Goal * Arizona Public Service (APS) Alternative Fuel (Hydrogen) Pilot Plant - design & operations * Fuel Dispensing * Hydrogen & HCNG Internal Combustion Engine (ICE) Vehicle Testing Activities * Briefly, other AVTA Activities * WWW Information 2 AVTA Background & Goal * Advanced Vehicle Testing Activity (AVTA) is part of the U.S. Department of Energy's (DOE) FreedomCAR and Vehicle Technologies Program * These activities are conducted by the Idaho National Laboratory (INL) & the AVTA testing partner Electric Transportation Applications (ETA) * AVTA Goal - Provide benchmark data for technology

269

Research Program of a Super Fast Reactor  

SciTech Connect

Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is not breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)

Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie; Terai, Takayuki; Nagasaki, Shinya; Muroya, Yusa; Abe, Hiroaki [Nuclear Professional School / Department of Nuclear Engineering and Management, The University of Tokyo, Tokaimura, Naka-gun, Ibaraki, 319-1188 (Japan); Mori, Hideo [Department of Mechanical Engineering, Kyushu University (Japan); Akiba, Masato; Akimoto, Hajime; Okumura, Keisuke; Akasaka, Naoaki [Japan Atomic Energy Agency (Japan); GOTO, Shoji [Tokyo Electric Power Company (Japan)

2006-07-01T23:59:59.000Z

270

The siting of UK nuclear reactors  

Science Journals Connector (OSTI)

Choosing a suitable site for a nuclear power station requires the consideration and balancing of several factors. Some 'physical' site characteristics, such as the local climate and the potential for seismic activity, will be generic to all reactors designs, while others, such as the availability of cooling water, the area of land required and geological conditions capable of sustaining the weight of the reactor and other buildings will to an extent be dependent on the particular design of reactor chosen (or alternatively the reactor design chosen may to an extent be dependent on the characteristics of an available site). However, one particularly interesting tension is a human and demographic one. On the one hand it is beneficial to place nuclear stations close to centres of population, to reduce transmission losses and other costs (including to the local environment) of transporting electricity over large distances from generator to consumer. On the other it is advantageous to place nuclear stations some distance away from such population centres in order to minimise the potential human consequences of a major release of radioactive materials in the (extremely unlikely) event of a major nuclear accident, not only in terms of direct exposure but also concerning the management of emergency planning, notably evacuation.This paper considers the emergence of policies aimed at managing this tension in the UK. In the first phase of nuclear development (roughly speaking 1945–1965) there was a highly cautious attitude, with installations being placed in remote rural locations with very low population density. The second phase (1965–1985) saw a more relaxed approach, allowing the development of AGR nuclear power stations (which with concrete pressure vessels were regarded as significantly safer) closer to population centres (in 'semi-urban' locations, notably at Hartlepool and Heysham). In the third phase (1985–2005) there was very little new nuclear development, Sizewell B (the first and so far only PWR power reactor in the UK) being colocated with an early Magnox station on the rural Suffolk coast. Renewed interest in nuclear new build from 2005 onward led to a number of sites being identified for new reactors before 2025, all having previously hosted nuclear stations and including the semi-urban locations of the 1960s and 1970s. Finally, some speculative comments are made as to what a 'fifth phase' starting in 2025 might look like.

Malcolm Grimston; William J Nuttall; Geoff Vaughan

2014-01-01T23:59:59.000Z

271

An Assessment of the Near-Term Costs of Hydrogen Refueling Stations and Station Components  

E-Print Network (OSTI)

4-12: Hydrogen Cost Comparison for Electrolysis Station Withthe hydrogen costs from the HSCM for electrolysis stations

Weinert, Jonathan X.; Lipman, Timothy

2006-01-01T23:59:59.000Z

272

A Study and Comparison of SCR Reaction Kinetics from Reactor...  

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

Data A Study and Comparison of SCR Reaction Kinetics from Reactor and Engine Experimental Data Presents experimental study of a Cu-zeolite SCR in both reactor and engine test cell,...

273

Fuzzy Synthetic Evaluation of Gas Station Safety  

Science Journals Connector (OSTI)

Based on the comprehensive analysis of hazard factors and evaluation indexes in gas stations, gas station safety is assessed in a fuzzy synthetic ... comprehensive evaluation, the specific safety level of gas stations

Xiaohua Hao; Xiao Feng

2010-01-01T23:59:59.000Z

274

Alternative Fuels Data Center: Propane Fueling Stations  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

Stations to someone by E-mail Stations to someone by E-mail Share Alternative Fuels Data Center: Propane Fueling Stations on Facebook Tweet about Alternative Fuels Data Center: Propane Fueling Stations on Twitter Bookmark Alternative Fuels Data Center: Propane Fueling Stations on Google Bookmark Alternative Fuels Data Center: Propane Fueling Stations on Delicious Rank Alternative Fuels Data Center: Propane Fueling Stations on Digg Find More places to share Alternative Fuels Data Center: Propane Fueling Stations on AddThis.com... More in this section... Propane Basics Benefits & Considerations Stations Locations Infrastructure Development Vehicles Laws & Incentives Propane Fueling Stations Photo of a liquefied petroleum gas fueling station. Thousands of liquefied petroleum gas (propane) fueling stations are

275

The IAEA Coordinated Research Program on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis: Description of the Benchmark Test Cases and Phases  

SciTech Connect

The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The uncertainties in the HTR analysis tools are today typically assessed with sensitivity analysis and then a few important input uncertainties (typically based on a PIRT process) are varied in the analysis to find a spread in the parameter of importance. However, one wish to apply a more fundamental approach to determine the predictive capability and accuracies of coupled neutronics/thermal-hydraulics and depletion simulations used for reactor design and safety assessment. Today there is a broader acceptance of the use of uncertainty analysis even in safety studies and it has been accepted by regulators in some cases to replace the traditional conservative analysis. Finally, there is also a renewed focus in supplying reliable covariance data (nuclear data uncertainties) that can then be used in uncertainty methods. Uncertainty and sensitivity studies are therefore becoming an essential component of any significant effort in data and simulation improvement. In order to address uncertainty in analysis and methods in the HTGR community the IAEA launched a Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modelling early in 2012. The project is built on the experience of the OECD/NEA Light Water Reactor (LWR) Uncertainty Analysis in Best-Estimate Modelling (UAM) benchmark activity, but focuses specifically on the peculiarities of HTGR designs and its simulation requirements. Two benchmark problems were defined with the prismatic type design represented by the MHTGR-350 design from General Atomics (GA) while a 250 MW modular pebble bed design, similar to the INET (China) and indirect-cycle PBMR (South Africa) designs are also included. In the paper more detail on the benchmark cases, the different specific phases and tasks and the latest status and plans are presented.

Frederik Reitsma; Gerhard Strydom; Bismark Tyobeka; Kostadin Ivanov

2012-10-01T23:59:59.000Z

276

Franklin Heating Station | Open Energy Information  

Open Energy Info (EERE)

Station Jump to: navigation, search Name: Franklin Heating Station Place: Minnesota References: EIA Form EIA-861 Final Data File for 2010 - File220101 EIA Form 861 Data Utility...

277

Energy Department Launches Alternative Fueling Station Locator...  

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

Launches Alternative Fueling Station Locator App Energy Department Launches Alternative Fueling Station Locator App November 7, 2013 - 11:16am Addthis As part of the Obama...

278

SOFTWARE TESTING, VERIFICATION AND RELIABILITY Softw. Test. Verif. Reliab. (2014)  

E-Print Network (OSTI)

. An industrial case study using reactor protection system software shows that the automatically generated test defined by the International Electrotechnical Commission (IEC). Industrial FBD testing relies mostlySOFTWARE TESTING, VERIFICATION AND RELIABILITY Softw. Test. Verif. Reliab. (2014) Published online

279

Light Water Reactors Technology Development - Nuclear Reactors  

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

Light Water Reactors Light Water 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 Argonne's Nuclear Science and Technology Legacy

280

Hydrogen at the Fueling Station  

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

Hydrogen) Service Stations 101 Hydrogen) Service Stations 101 Steven M. Schlasner September 22, 2004 2 DISCLAIMER Opinions expressed within are strictly those of the presenter and do not necessarily represent ConocoPhillips Company. 3 Presentation Outline * Introduction to ConocoPhillips * Introduction to Service Stations * Comparison of Conventional with Hydrogen Fueling Stations * Hydrogen Fueling Life Cycle * Practical Design Example * Concluding Observations 4 ConocoPhillips * 7 th on Fortune's list of largest companies (2003 revenues) * 3 rd largest integrated petroleum company in U.S. * 1 st (largest) petroleum refiner in U.S. * 14,000 retail outlets (350 company-owned) in 44 states * Brands: Conoco, Phillips 66, 76 * 32,800 miles pipeline, owned or interest in * 64 terminals: crude, LPG, refined products

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


281

Pilgrim Station | Open Energy Information  

Open Energy Info (EERE)

Station Station Jump to: navigation, search Name Pilgrim Station Facility Pilgrim Stage Station Sector Wind energy Facility Type Commercial Scale Wind Facility Status In Service Owner ReunionPower/Exergy Developer Exergy Location Twin Falls County ID Coordinates 42.741336°, -114.865865° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":42.741336,"lon":-114.865865,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

282

Mini cogeneration stations: Foreign experience  

Science Journals Connector (OSTI)

The prospects of using autonomous power and heat supply systems are analyzed. The economic advantages of mini cogeneration power stations equipped with gas piston, diesel, or gas turbine units are shown. Examples...

V. R. Kotler

2006-08-01T23:59:59.000Z

283

Richmond Field Station Tow Tank | Open Energy Information  

Open Energy Info (EERE)

Richmond Field Station Tow Tank Richmond Field Station Tow Tank Jump to: navigation, search Basic Specifications Facility Name Richmond Field Station Tow Tank Overseeing Organization University of California, Berkeley Hydrodynamic Testing Facility Type Tow Tank Length(m) 67.0 Beam(m) 2.4 Depth(m) 1.7 Water Type Freshwater Cost(per day) Contact POC Special Physical Features Glass observation station, suitable for optical access Towing Capabilities Towing Capabilities Yes Maximum Velocity(m/s) 5 Length of Effective Tow(m) 50.0 Wavemaking Capabilities Wavemaking Capabilities Yes Maximum Wave Height(m) 0.3 Maximum Wave Length(m) 2 Wave Period Range(s) 0.0 Current Velocity Range(m/s) 0.0 Programmable Wavemaking Yes Wavemaking Description Waveform can be programmed Wave Direction Both

284

MHK Technologies/Ocean Powered Compressed Air Stations | Open Energy  

Open Energy Info (EERE)

Powered Compressed Air Stations Powered Compressed Air Stations < MHK Technologies Jump to: navigation, search << Return to the MHK database homepage Ocean Powered Compressed Air Stations.png Technology Profile Primary Organization Wave Power Plant Inc Technology Resource Click here Wave Technology Type Click here Point Absorber - Submerged Technology Readiness Level Click here TRL 4 Proof of Concept Technology Description The Ocean Powered Compressed Air Station is a point absorber that uses an air pump to force air to a landbased generator The device only needs 4m water depth and electricity production fluctations through storing energy at a constant air pressure Technology Dimensions Device Testing Date Submitted 13:16.5 << Return to the MHK database homepage Retrieved from

285

MHK Technologies/Jiangxia Tidal Power Station | Open Energy Information  

Open Energy Info (EERE)

Jiangxia Tidal Power Station Jiangxia Tidal Power Station < MHK Technologies Jump to: navigation, search << Return to the MHK database homepage Jiangxia Tidal Power Station.jpg Technology Profile Primary Organization China Guodian Corporation Technology Type Click here Axial Flow Turbine Technology Readiness Level Click here TRL 9 Commercial Scale Production Application Technology Description There are 6 bulb turbine generator units operating in both ebb and flood tides with a total installed capacity up to 3 9 MW Technology Dimensions Technology Nameplate Capacity (MW) 3 9 Device Testing Date Submitted 14:15.7 << Return to the MHK database homepage Retrieved from "http://en.openei.org/w/index.php?title=MHK_Technologies/Jiangxia_Tidal_Power_Station&oldid=681601

286

Nuclear reactor characteristics and operational history  

Gasoline and Diesel Fuel Update (EIA)

1. Capacity and Generation, Table 3. Characteristics and Operational History 1. Capacity and Generation, Table 3. Characteristics and Operational History Table 2. U.S. Nuclear Reactor Ownership Data PDF XLS Plant/Reactor Name Generator ID Utility Name - Operator Owner Name % Owned Arkansas Nuclear One 1 Entergy Arkansas Inc Entergy Arkansas Inc 100 Arkansas Nuclear One 2 Entergy Arkansas Inc Entergy Arkansas Inc 100 Beaver Valley 1 FirstEnergy Nuclear Operating Company FirstEnergy Nuclear Generation Corp 100 Beaver Valley 2 FirstEnergy Nuclear Operating Company FirstEnergy Nuclear Generation Corp 100 Braidwood Generation Station 1 Exelon Nuclear Exelon Nuclear 100 Braidwood Generation Station 2 Exelon Nuclear Exelon Nuclear 100 Browns Ferry 1 Tennessee Valley Authority Tennessee Valley Authority 100

287

DOE Permitting Hydrogen Facilities: Hydrogen Fueling Stations  

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

Stations Stations Public-use hydrogen fueling stations are very much like gasoline ones. In fact, sometimes, hydrogen and gasoline cars can be fueled at the same station. These stations offer self-service pumps, convenience stores, and other services in high-traffic locations. Photo of a Shell fueling station showing the site convenience store and hydrogen and gasoline fuel pumps. This fueling station in Washington, D.C., provides drivers with both hydrogen and gasoline fuels Many future hydrogen fueling stations will be expansions of existing fueling stations. These facilities will offer hydrogen pumps in addition to gasoline or natural gas pumps. Other hydrogen fueling stations will be "standalone" operations. These stations will be designed and constructed to

288

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

289

Photocatalytic reactor  

DOE Patents (OSTI)

A photocatalytic reactor is described for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane. 4 figs.

Bischoff, B.L.; Fain, D.E.; Stockdale, J.A.D.

1999-01-19T23:59:59.000Z

290

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

SciTech Connect

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

NONE

1997-05-01T23:59:59.000Z

291

Hybrid adsorptive membrane reactor  

DOE Patents (OSTI)

A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

Tsotsis, Theodore T. (Huntington Beach, CA); Sahimi, Muhammad (Altadena, CA); Fayyaz-Najafi, Babak (Richmond, CA); Harale, Aadesh (Los Angeles, CA); Park, Byoung-Gi (Yeosu, KR); Liu, Paul K. T. (Lafayette Hill, PA)

2011-03-01T23:59:59.000Z

292

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

293

GEN-IV Reactors  

Science Journals Connector (OSTI)

Generation-IV reactors are a set of nuclear reactors currently being developed under international collaborations targeting ... economics, proliferation resistance, and physical protection of nuclear energy. Nuclear

Taek K. Kim

2013-01-01T23:59:59.000Z

294

The Netherlands Reactor Centre  

Science Journals Connector (OSTI)

... Two illustrated brochures in English have recently J. been issued by the Netherlands Reactor Centre ( ... Centre (Reactor Centrum Nederland). The first* gives a general survey of the ...

S. WEINTROUB

1964-04-04T23:59:59.000Z

295

Nuclear Reactor Safety Design Criteria  

Directives, Delegations, and Requirements

The order establishes nuclear safety criteria applicable to the design, fabrication, construction, testing, and performance requirements of nuclear reactor facilities and safety class structures, systems, and components (SSCs) within these facilities. Cancels paragraphs 8a and 8b of DOE 5480.6. Cancels DOE O 5480.6 in part. Certified 11-18-10.

1993-01-19T23:59:59.000Z

296

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

297

Tritium issues in commercial pressurized water reactors  

SciTech Connect

Tritium has become an important radionuclide in commercial Pressurized Water Reactors because of its mobility and tendency to concentrate in plant systems as tritiated water during the recycling of reactor coolant. Small quantities of tritium are released in routine regulated effluents as liquid water and as water vapor. Tritium has become a focus of attention at commercial nuclear power plants in recent years due to inadvertent, low-level, chronic releases arising from routine maintenance operations and from component failures. Tritium has been observed in groundwater in the vicinity of stations. The nuclear industry has undertaken strong proactive corrective measures to prevent recurrence, and continues to eliminate emission sources through its singular focus on public safety and environmental stewardship. This paper will discuss: production mechanisms for tritium, transport mechanisms from the reactor through plant, systems to the environment, examples of routine effluent releases, offsite doses, basic groundwater transport and geological issues, and recent nuclear industry environmental and legal ramifications. (authors)

Jones, G. [Constellation Energy Group, R.E. Ginna Nuclear Power Plant, Ontario, NY (United States)

2008-07-15T23:59:59.000Z

298

Consortium for Advanced Simulation of Light Water Reactors (CASL...  

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

has designed and operated 52 test reactors, including EBR-1, the world's first nuclear power plant Key Contributions System safety analysis Multiscale fuel performance...

299

An Assessment of the Near-Term Costs of Hydrogen Refueling Stations and Station Components  

E-Print Network (OSTI)

e.g. co-locate with gasoline station, bus-yard, or officeintegrated into existing gasoline stations with 8 dispensersof a liquid H 2 and gasoline station layout. Figure 4-9:

Weinert, Jonathan X.; Lipman, Timothy

2006-01-01T23:59:59.000Z

300

An Assessment of the Near-Term Costs of Hydrogen Refueling Stations and Station Components  

E-Print Network (OSTI)

located at an existing gasoline station. One could use othere.g. co-locate with gasoline station, bus-yard, or officeequivalent to about 6% of gasoline stations in California 4.

Lipman, T E; Weinert, Jonathan X.

2006-01-01T23:59:59.000Z

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

Mobile Alternative Fueling Station Locator  

SciTech Connect

The Department of Energy's Alternative Fueling Station Locator is available on-the-go via cell phones, BlackBerrys, or other personal handheld devices. The mobile locator allows users to find the five closest biodiesel, electricity, E85, hydrogen, natural gas, and propane fueling sites using Google technology.

Not Available

2009-04-01T23:59:59.000Z

302

University of Washington Seismograph Stations  

Science Journals Connector (OSTI)

...MEMBERSHIP COHM ITTEE Ross R. Heinrich...n~bers and changes or address should...seismograms, (2) a world-wide...Utah a 0.3 Bonneville lake beds Scoresbysund...Alaska 1.1 2.0 Graywacke...Stores, 221 North Grand Boulevard...SEISM and the name of the station...day of month (2) ml n2 sl s2...

303

US approves first new nuclear reactors in 30 years  

Science Journals Connector (OSTI)

... giant based in Atlanta, Georgia, can now move forward with a pair of Westinghouse AP1000 nuclear reactors at its Vogtle nuclear station near Waynesboro. But whereas many within government ... to hasten a historically slow, expensive and, some say, unpredictable regulatory process).  The AP1000 (fuzzy artistic rendering, top right) is one of a new generation of pressurized- ...

Jeff Tollefson

2012-02-10T23:59:59.000Z

304

Gas-cooled fast breeder reactor. Quarterly progress report, November 1, 1979 through January 31, 1980  

SciTech Connect

Information is presented concerning the nuclear steam supply system; reactor core; systems engineering; safety and reliability; and circulator test facility.

Not Available

1980-02-01T23:59:59.000Z

305

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

306

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

307

End-of-life destructive examination of light water breeder reactor fuel rods (LWBR Development Program)  

SciTech Connect

Destructive examination of 12 representative Light Water Breeder Reactor fuel rods was performed following successful operation in the Shippingport Atomic Power Station for 29,047 effective full power hours, about five years. Light Water Breeder Reactor fuel rods were unique in that the thorium oxide and uranium-233 oxide fuel was contained within Zircaloy-4 cladding. Destructive examinations included analysis of released fission gas; chemical analysis of the fuel to determine depletion, iodine, and cesium levels; chemical analysis of the cladding to determine hydrogen, iodine, and cesium levels; metallographic examination of the cladding, fuel, and other rod components to determine microstructural features and cladding corrosion features; and tensile testing of the irradiated cladding to determine mechanical strength. The examinations confirmed that Light Water Breeder Reactor fuel rod performance was excellent. No evidence of fuel rod failure was observed, and the fuel operating temperature was low (below 2580/sup 0/F at which an increased percentage of fission gas is released). 21 refs., 80 figs., 20 tabs.

Richardson, K.D.

1987-10-01T23:59:59.000Z

308

(Liquid metal reactor/fast breeder reactor research and development)  

SciTech Connect

The second meeting of the UJCC was held in Japan on June 6--8, 1990. The first day was devoted to presentations of the status of the US and Japanese Fast Breeder Reactor (FBR) programs and the status of specific areas of cooperative work. Briefly, the Japanese are following the FBR development program which has been in place since the 1970s. This program includes an FBR test reactor (JOYO), a pilot-scale reactor (MONJU), a demonstration-scale plant, and commercial-scale plants by about 2020. The US program has been redirected toward an actinide recycle mission using metal fuel and pyroprocessing of spent fuel to recovery both Pu and the higher actinides for return to the Liquid Metal Reactor (LMR). The second day was spent traveling from Tokyo to Tsuruga for a tour of the MONJU reactor. The tour was especially interesting. The third day was spent writing the minutes of the meeting and the return trip to Tokyo.

Homan, F.J.

1990-06-20T23:59:59.000Z

309

Nuclear reactor  

DOE Patents (OSTI)

A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

Thomson, Wallace B. (Severna Park, MD)

2004-03-16T23:59:59.000Z

310

alternative fuels stations | OpenEI  

Open Energy Info (EERE)

fuels stations fuels stations Dataset Summary Description Alternative fueling stations are located throughout the United States and their availability continues to grow. The Alternative Fuels Data Center (AFDC) maintains a website where you can find alternative fuels stations near you or on a route, obtain counts of alternative fuels stations by state, Source Alternative Fuels Data Center Date Released December 13th, 2010 (4 years ago) Date Updated December 13th, 2010 (4 years ago) Keywords alt fuel alternative fuels alternative fuels stations biodiesel CNG compressed natural gas E85 Electricity ethanol hydrogen liquefied natural gas LNG liquefied petroleum gas LPG propane station locations Data text/csv icon alt_fuel_stations_apr_4_2012.csv (csv, 2.3 MiB) Quality Metrics Level of Review Peer Reviewed

311

COMPARATIVE COSTS OF CALIFORNIA CENTRAL STATION ELECTRICITY  

E-Print Network (OSTI)

CALIFORNIA ENERGY COMMISSION COMPARATIVE COSTS OF CALIFORNIA CENTRAL STATION ELECTRICITY GENERATION and Anitha Rednam, Comparative Costs of California Central Station Electricity Generation Technologies Manager Ruben Tavares - Acting Manager ELECTRICITY ANALYSIS OFFICE Sylvia Bender Deputy Director

Laughlin, Robert B.

312

Alternative Fueling Station Locator | Department of Energy  

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

Alternative Fueling Station Locator Alternative Fueling Station Locator Alternative Fueling Station Locator Find Stations Plan a Route Location: Go Start: End: Go Fuel: All Fuels Biodiesel (B20 and above) Compressed Natural Gas Electric Ethanol (E85) Hydrogen Liquefied Natural Gas (LNG) Liquefied Petroleum Gas (Propane) more search options close × More Search Options Include private stations Include planned stations Owner All Private Federal State Local Utility Payment All American Express Discover MasterCard VISA Cash Checks CFN Clean Energy Fuel Man Gas Card PHH Services Voyager WEX Electric charger types Include level 1 Include level 2 Include DC fast Include legacy chargers Limit results to within 5 miles Limit results to within 5 miles 12,782 alternative fuel stations in the United States Excluding private stations

313

Enhancement of NRC station blackout requirements for nuclear power plants  

SciTech Connect

The U.S. Nuclear Regulatory Commission (NRC) established a Near-Term Task Force (NTTF) in response to Commission direction to conduct a systematic and methodical review of NRC processes and regulations to determine whether the agency should make additional improvements to its regulatory system and to make recommendations to the Commission for its policy direction, in light of the accident at the Fukushima Dai-ichi Nuclear Power Plant. The NTTF's review resulted in a set of recommendations that took a balanced approach to defense-in-depth as applied to low-likelihood, high-consequence events such as prolonged station blackout (SBO) resulting from severe natural phenomena. Part 50, Section 63, of Title 10 of the Code of Federal Regulations (CFR), 'Loss of All Alternating Current Power,' currently requires that each nuclear power plant must be able to cool the reactor core and maintain containment integrity for a specified duration of an SBO. The SBO duration and mitigation strategy for each nuclear power plant is site specific and is based on the robustness of the local transmission system and the transmission system operator's capability to restore offsite power to the nuclear power plant. With regard to SBO, the NTTF recommended that the NRC strengthen SBO mitigation capability at all operating and new reactors for design-basis and beyond-design-basis external events. The NTTF also recommended strengthening emergency preparedness for prolonged SBO and multi-unit events. These recommendations, taken together, are intended to clarify and strengthen US nuclear reactor safety regarding protection against and mitigation of the consequences of natural disasters and emergency preparedness during SBO. The focus of this paper is on the existing SBO requirements and NRC initiatives to strengthen SBO capability at all operating and new reactors to address prolonged SBO stemming from design-basis and beyond-design-basis external events. The NRC initiatives are intended to enhance core and spent fuel pool cooling, reactor coolant system integrity, and containment integrity. (authors)

McConnell, M. W. [United States Nuclear Regulatory Commission, Mail Stop: 012-H2, Washington, DC 20555 (United States)

2012-07-01T23:59:59.000Z

314

7.1.1. Fernbahnhof / Rail Station  

E-Print Network (OSTI)

.1.1.4.7 Gasversorgung des Fernbahnhofs / Gas Supply of Rail Station 7.1.1.4.7.1 Gasversorgung des Fernbahnhofs 77.1.1. Fernbahnhof / Rail Station 7.1.2. �PNV-Bahnhof / Public Transport Station 7.1.3. Busbahnhof / Bus Station 7.1.4. Taxianlagen / Taxi Stand 7.1.5. Sonstige Passagiergebäude / Other passenger

Berlin,Technische Universität

315

International Space Station Program Overview  

E-Print Network (OSTI)

International Space Station Program Overview 1 #12;BerthingDock MRM1 / FGB Nadir DC1 / MLM / RS Node SM Aft MRM2 / SM-Zenith Node 1 Nadir Node 2 Nadir Node 2 Zenith 2013 2014 2015 May Jun Jul Aug Sep the line (Dock on 5/1/13) HTV4 (7/20/13 ­ 8/19/13) Pre-Decisional, For Internal Use Only Date Color Key

316

The Fuel-Travel-Back Approach to Hydrogen Station Siting  

E-Print Network (OSTI)

the experience of gasoline stations. Driven by the notion "percentages of existing gasoline stations, for a successfulsubset of the existing gasoline station network [14]. These

Lin, Zhenhong; Ogden, Joan; Fan, Yueyue; Chen, Chien-Wei

2009-01-01T23:59:59.000Z

317

Solar Powered Radioactive Air Monitoring Stations  

SciTech Connect

Environmental monitoring of ambient air for radioactive material is required as stipulated in the PNNL Site radioactive air license. Sampling ambient air at identified preferred locations could not be initially accomplished because utilities were not readily available. Therefore, solar powered environmental monitoring systems were considered as a possible option. PNNL purchased two 24-V DC solar powered environmental monitoring systems which consisted of solar panels, battery banks, and sampling units. During an approximate four month performance evaluation period, the solar stations operated satisfactorily at an on-site test location. They were subsequently relocated to their preferred locations in June 2012 where they continue to function adequately under the conditions found in Richland, Washington.

Barnett, J. M.; Bisping, Lynn E.; Gervais, Todd L.

2013-10-30T23:59:59.000Z

318

Advanced Resin Cleaning System (ARCS) at Grand Gulf Nuclear Station  

SciTech Connect

Steam generation system in-core components can undergo serious material degradation by a variety of corrosion-related phenomena. These phenomena are largely controlled by boiler water (i.e. reactor water) chemistry which is strongly impacted by the performance of the condensate system mixed bed ion exchange units. In Boiling Water Reactors (BWR), the mixed bed ion exchange units not only provide protection from ionic contaminants, but also remove insoluble corrosion products by filtration/adsorption. These insoluble corrosion products removed by the ion exchange units must then be periodically cleaned from the resin bed by some process external to the BWR primary water loop. A unique resin cleaning process called the {open_quotes}Advanced Resin Cleaning System{close_quotes} (ARCS) was developed in the late 1980`s by members of CENTEC-XXI, located in Santa Clara, CA. This system, which has been successfully operated for several years at a Pressurized Water Reactor is highly efficient for removal of both insoluble corrosion products and anion/cation resin fines, and generates significantly less waste water than other cleaning methods. The ARCS was considered the most attractive method for meeting the demanding and costly resin cleaning needs of a BWR. A {open_quotes}Tailored Collaboration{close_quotes} project was initiated between EPRI, Entergy Operations (Grand Gulf Station), and CENTEC-XXI to demonstrate the {open_quotes}Advanced Resin Cleaning System{close_quotes} in a BWR.

Asay, R.H.; Earls, J.E.; Naughton, M.D. [Centec 21, Inc., Santa Clara, CA (United States)

1996-10-01T23:59:59.000Z

319

Fuel cycle facility control system for the Integral Fast Reactor Program  

SciTech Connect

As part of the Integral Fast Reactor (IFR) Fuel Demonstration, a new distributed control system designed, implemented and installed. The Fuel processes are a combination of chemical and machining processes operated remotely. To meet this special requirement, the new control system provides complete sequential logic control motion and positioning control and continuous PID loop control. Also, a centralized computer system provides near-real time nuclear material tracking, product quality control data archiving and a centralized reporting function. The control system was configured to use programmable logic controllers, small logic controllers, personal computers with touch screens, engineering work stations and interconnecting networks. By following a structured software development method the operator interface was standardized. The system has been installed and is presently being tested for operations.

Benedict, R.W.; Tate, D.A.

1993-09-01T23:59:59.000Z

320

Final Report: Particulate Emissions Testing, Unit 1, Potomac...  

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

Final Report: Particulate Emissions Testing, Unit 1, Potomac River Generating Station, Alexandria, Virginia Final Report: Particulate Emissions Testing, Unit 1, Potomac River...

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

Alternative Fueling Station Locations | OpenEI  

Open Energy Info (EERE)

Alternative Fueling Station Locations Alternative Fueling Station Locations Dataset Summary Description Alternative fueling stations are located throughout the United States and their availability continues to grow. The Alternative Fuels Data Center (AFDC) maintains a website where you can find alternative fuels stations near you or on a route, obtain counts of alternative fuels stations by state, view U.S. maps, and more. Access up-to-date fuel station data here: http://www.afdc.energy.gov/afdc/data_download The dataset available for download here provides a "snapshot" of the alternative fueling station information for: compressed natural gas (CNG), E85 (85% ethanol, 15% gasoline), propane/liquefied petroleum gas (LPG), biodiesel, electricity, hydrogen, and liquefied natural gas

322

Alternative Fueling Station Locator | Open Energy Information  

Open Energy Info (EERE)

Alternative Fueling Station Locator Alternative Fueling Station Locator Jump to: navigation, search Tool Summary LAUNCH TOOL Name: Alternative Fueling Station Locator Agency/Company /Organization: United States Department of Energy Partner: National Renewable Energy Laboratory Sector: Energy Focus Area: Fuels & Efficiency, Transportation Phase: Evaluate Options, Prepare a Plan Topics: Datasets Resource Type: Online calculator User Interface: Website Website: www.afdc.energy.gov/afdc/locator/stations/ Web Application Link: www.afdc.energy.gov/afdc/locator/stations/ Cost: Free OpenEI Keyword(s): Featured References: National Renewable Energy Laboratory Advanced Vehicles and Fuels Research: Data and Resources[1] Logo: Alternative Fueling Station Locator The alternative fuel station locator uses an address based search to find

323

Pipeline compressor station construction cost analysis  

Science Journals Connector (OSTI)

This study aims to provide a reference for pipeline compressor station construction costs by analysing individual compressor station cost components using historical compressor station cost data between 1992 and 2008. Distribution and share of these pipeline compressor station cost components are assessed based on compressor station capacity, year of completion, and locations. Average unit costs in material, labour, miscellaneous, land, and total costs are $866/hp, $466/hp, $367/hp, $13/hp, and $1,712/hp, respectively. Primary costs for compressor stations are material cost, approximately 50.6% of the total cost. This study conducts a learning curve analysis to investigate the learning rate of material and labour costs for different groups. Results show that learning rates and construction component costs vary by capacity and locations. This study also investigates the causes of pipeline compressor station construction cost differences. [Received: March 25, 2012; Accepted; 20 February 2013

Yipeng Zhao; Zhenhua Rui

2014-01-01T23:59:59.000Z

324

Taipei terminal rail station : casting an urban gateway  

E-Print Network (OSTI)

Access is a key issue in the design of railway stations. The evolution of the train station typology, has resulted in many types of stations based on the development of the stations' access. Since rail travel on a larger ...

Tsai, May Deanna

1991-01-01T23:59:59.000Z

325

Dubuque generation station, Dubuque, Iowa  

SciTech Connect

Alliant Energy's Dubuque generation station is a fine example of why small does not mean insignificant in the power generation industry. This winner of the EUCG best performer award in the small plant category shows that its operating excellence towers over that of many larger and much newer coal-fired power plants. The plant has three operating units with boilers originally designed for Illinois basin coal but now Powder River Basin coal makes up 75% of the coal consumed. The boilers can also burn natural gas. 4 photos.

Peltier, R.

2008-10-15T23:59:59.000Z

326

Attrition reactor system  

DOE Patents (OSTI)

A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

Scott, C.D.; Davison, B.H.

1993-09-28T23:59:59.000Z

327

Elementary Reactor Physics  

Science Journals Connector (OSTI)

... THERE are few subjects which have developed at the rate at which reactor physics and ... physics and reactor theory have done. This, of course, is largely due to the circumstances in ...

J. F. HILL

1962-02-10T23:59:59.000Z

328

Colliding Beam Fusion Reactors  

Science Journals Connector (OSTI)

The recirculating power for virtually all types of fusion reactors has previously been calculated [1] with the Fokker–Planck equation. The reactors involve non-Maxwellian plasmas. The calculations are ... the rec...

Norman Rostoker; Artan Qerushi; Michl Binderbauer

2003-06-01T23:59:59.000Z

329

The Vermont Yankee Nuclear Power Station will use a single-unit boilingwater  

E-Print Network (OSTI)

reactor with an initial power rating of 1593 thermal megawatts (MWt) to provide a net power output of 513 electrical megawatts (Me). The reactor will be cooled by a once-through flow of water pumped from and returned to Vernon Pond, an existing impoundment of the Connecticut River (built to serve the Vernon Hydroelectric Station) and also by means of mechanical draft cooling towers. 3. Sumnary of environmental impact including beneficial and adverse effects follows: a. Cooling water heated to about 20*F above inlet temperature will be discharged to Vernon Pond at a rate of 840 cfs when the plant operates on a total open-cycle basis. Mechanical-draft cooling towers are provided to protect Vernon Pond during low flow and critical temperature periods in the Connecticut River. b. About 150 acres of Vernon Pond in the vicinity of the station may be subjected to some thermal and biological stress from discharge of the

Vermont Yankee

330

Fluid sampling system for a nuclear reactor  

DOE Patents (OSTI)

A system of extracting fluid samples, either liquid or gas, from the interior of a nuclear reactor containment utilizes a jet pump. To extract the sample fluid, a nonradioactive motive fluid is forced through the inlet and discharge ports of a jet pump located outside the containment, creating a suction that draws the sample fluid from the containment through a sample conduit connected to the pump suction port. The mixture of motive fluid and sample fluid is discharged through a return conduit to the interior of the containment. The jet pump and means for removing a portion of the sample fluid from the sample conduit can be located in a shielded sample grab station located next to the containment. A non-nuclear grade active pump can be located outside the grab sampling station and the containment to pump the nonradioactive motive fluid through the jet pump.

Lau, Louis K. (Monroeville, PA); Alper, Naum I. (Monroeville, PA)

1994-01-01T23:59:59.000Z

331

Parallel Monte Carlo reactor neutronics  

SciTech Connect

The issues affecting implementation of parallel algorithms for large-scale engineering Monte Carlo neutron transport simulations are discussed. For nuclear reactor calculations, these include load balancing, recoding effort, reproducibility, domain decomposition techniques, I/O minimization, and strategies for different parallel architectures. Two codes were parallelized and tested for performance. The architectures employed include SIMD, MIMD-distributed memory, and workstation network with uneven interactive load. Speedups linear with the number of nodes were achieved.

Blomquist, R.N.; Brown, F.B.

1994-03-01T23:59:59.000Z

332

Prospects for spheromak fusion reactors  

Science Journals Connector (OSTI)

The reactor study of Hagenson and Krakowski demonstrated the attractiveness of the spheromak as a compact fusion reactor, based on...

T. K. Fowler; D. D. Hua

1995-06-01T23:59:59.000Z

333

Improved vortex reactor system  

DOE Patents (OSTI)

An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

Diebold, James P. (Lakewood, CO); Scahill, John W. (Evergreen, CO)

1995-01-01T23:59:59.000Z

334

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

335

Preservation of FFTF Data Related to Passive Safety Testing  

SciTech Connect

One of the goals of the Fuel Cycle Research and Development Program (FCRD) is to preserve the knowledge that has been gained in the United States on Liquid Metal Reactors (LMR). A key area deserving special attention for preservation is the data relating to passive safety testing that was conducted in FFTF and EBR-II during the 1980’s. Accidents at Unit 4 of the Chernobyl Station and Unit 2 at Three Mile Island changed the safety paradigm of the nuclear power industry. New emphasis was placed on assured safety based on intrinsic plant characteristics that protect not only the public, but the significant investment in the plant as well. Plants designated to perform in this manner are considered to be passively safe since no active sensor/alarm system or human intervention is required to bring the reactor to a safe shutdown condition. The liquid metal reactor (LMR) has several key characteristics needed for a passively safe reactor: reactor coolant with superior heat transfer capability and very high boiling point, low (atmospheric) system pressures, and reliable negative reactivity feedback. The credibility of the design for a passively safe LMR rests on two issues: the validity of analytic methods used to predict passive safety performance and the availability of relevant test data to calibrate design tools. Safety analysis methods used to analyze LMRs under the old safety paradigm were focused on calculating the source term for the Core Disruptive Accident. Passive safety design requires refined analysis methods for transient events because treatment of the detailed reactivity feedbacks is important in predicting the response of the reactor. Similarly, analytic tools should be calibrated against actual test experience in existing LMR facilities. The principal objectives of the combined FFTF natural circulation and Passive Safety Testing program were: 1) to verify natural circulation as a reliable means to safely remove decay heat, 2) to extend passive safety experience to a large-size LMR and obtain data for validating design analysis computer codes, and 3) to develop and test passive safety enhancements that might be used for future LMRs. These tests were designed to provide data sufficient to allow separation of fuel temperature effects from structural temperature effects. The data developed through this testing program were used to verify the predictive capability of passive safety analysis methods as well as provide a data base for calibrating design tools such as the SASSYS/SAS4A codes. These tests were instrumental in improving understanding of reactivity feedback mechanisms in LMRs and demonstrating passive safety margins available in an LMR. Knowledge preservation at the FFTF is focused on the areas of design, construction, startup, and operation of the reactor. This information may be of potential use for international exchanges with other LMR programs around the world. This information provides the basis for creating benchmarks for validating and testing large scale computer programs. All information preserved to date is now being stored and categorized consistent with the IAEA international standardized taxonomy. The test results information exists in several different formats depending upon the final stage of the test evaluation. Over 100 documents relevant to passive safety testing have been identified and are being recovered, scanned, and catalogued. Attempts to recover plant data tapes are also in progress. Documents related to passive safety testing are now being categorized consistent with internationally agreed upon IAEA standards. Documents are being converted to electronic format compatible with a general search engine being developed by INL. The data from the FFTF passive safety tests provides experimental verification of structural reactivity effects that should be very useful to innovative designers seeking to optimize passive safety in the design of new LMRs.

Wootan, David W.; Butner, R. Scott; Omberg, Ronald P.; Makenas, Bruce J.; Nielsen, Deborah L.

2010-10-01T23:59:59.000Z

336

High Temperature Test Laboratory Accomplishments  

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

Knudson, K. Condie, and B. Sencer, "In-situ Creep Testing Capability for the Advanced Test Reactor," Nuclear Technology, 179, 3, September 2012, pp 413-428. B. Geslot, T. Unruh,...

337

Reactor vessel support system  

DOE Patents (OSTI)

A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

Golden, Martin P. (Trafford, PA); Holley, John C. (McKeesport, PA)

1982-01-01T23:59:59.000Z

338

Bush Hydrogen Vision "Fueled" By California Station Opening | Department of  

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

Bush Hydrogen Vision "Fueled" By California Station Opening Bush Hydrogen Vision "Fueled" By California Station Opening Bush Hydrogen Vision "Fueled" By California Station Opening February 18, 2005 - 10:26am Addthis CHINO, CALIF. - In a major step toward achieving President George W. Bush's vision for a hydrogen economy, Assistant Secretary of Energy David Garman today joined representatives of ChevronTexaco, Hyundai-Kia and UTC Fuel Cells at the opening of a hydrogen fueling station in Chino, CA. The station is a major part of the Department of Energy's Hydrogen "Learning Demonstration," which brings together automobile makers and energy companies to test fuel cell vehicles and hydrogen fueling systems in real-world conditions. "Hydrogen fuel cells represent one of the most encouraging, innovative

339

Bush Hydrogen Vision "Fueled" By California Station Opening | Department of  

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

Hydrogen Vision "Fueled" By California Station Opening Hydrogen Vision "Fueled" By California Station Opening Bush Hydrogen Vision "Fueled" By California Station Opening February 18, 2005 - 10:26am Addthis CHINO, CALIF. - In a major step toward achieving President George W. Bush's vision for a hydrogen economy, Assistant Secretary of Energy David Garman today joined representatives of ChevronTexaco, Hyundai-Kia and UTC Fuel Cells at the opening of a hydrogen fueling station in Chino, CA. The station is a major part of the Department of Energy's Hydrogen "Learning Demonstration," which brings together automobile makers and energy companies to test fuel cell vehicles and hydrogen fueling systems in real-world conditions. "Hydrogen fuel cells represent one of the most encouraging, innovative

340

Testing the scaling of thermal transport models: predicted and measured temperatures in the Tokamak Fusion Test  

E-Print Network (OSTI)

in the Tokamak Fusion Test Reactor dimensionless scaling experiments D. R. Mikkelsen, S. D. Scott Princeton the Tokamak Fusion Test Reactor [D. J. Grove and D. M. Meade, Nucl. Fusion 25, 1167 (1985)] nondimensional to International Tokamak Experimental Reactor [2] (ITER) class tokamaks. This paper compares the predictions

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

Reactor water cleanup system  

DOE Patents (OSTI)

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

Gluntz, Douglas M. (San Jose, CA); Taft, William E. (Los Gatos, CA)

1994-01-01T23:59:59.000Z

342

Reactor water cleanup system  

DOE Patents (OSTI)

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

Gluntz, D.M.; Taft, W.E.

1994-12-20T23:59:59.000Z

343

A Study and Comparison of SCR Reaction Kinetics from Reactor and Engine Experimental Data  

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

Presents experimental study of a Cu-zeolite SCR in both reactor and engine test cell, and comparison of the model parameters between the SCR reactor and engine model

344

Installation for a nuclear power station with staggered swimming pools  

SciTech Connect

In an installation for a nuclear power station comprising a ''reactor building'' with a first swimming pool for handling of fuel units and a fuel building with a second swimming pool for the transfer, storage and deactivation of the units, the second swimming pool is located at a lower level than that of the first and is connected to the first by an intermediate auxiliary chamber filled with water and located under the first swimming pool. The auxiliary chamber is connected by a vertical pipeline to the first swimming pool and by a horizontal connecting pipeline to the second swimming pool. Each of the pipelines is provided with a shut-off valve, with interlocking means which prevents the simultaneous opening of the two valves. There is negligible dead space around a conveyor basket for fuel units when it is in the vertical or horizontal pipelines.

Gigou, R.

1982-12-28T23:59:59.000Z

345

Stridsvagn 122 och Remote Weapon Station.  

E-Print Network (OSTI)

?? Den här uppsatsen behandlar stridsvagn122 och Remote Weapon Station (RWS). Det finns ett verkansglapp mellan dagens kalibrar 120 mm och 7,62 mm. Observationsmöjligheterna i… (more)

Sellberg, Martin

2010-01-01T23:59:59.000Z

346

Illinois Nuclear Profile - Braidwood Generation Station  

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

Braidwood Generation Station" "Unit","Summer capacity (mw)","Net generation (thousand mwh)","Summer capacity factor (percent)","Type","Commercial operation date","License...

347

Illinois Nuclear Profile - Byron Generating Station  

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

Byron Generating Station" ,"Summer capacity (mw)","Net generation (thousand mwh)","Summer capacity factor (percent)","Type","Commercial operation date","License expiration date"...

348

Massachusetts Nuclear Profile - Pilgrim Nuclear Power Station  

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

Pilgrim Nuclear Power Station" "Unit","Summer capacity (mw)","Net generation (thousand mwh)","Summer cpacity factor (percent)","Type","Commercial operation date","License...

349

GC GUIDANCE ON ELECTRIC VEHICLE RECHARGING STATIONS  

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

Several National Laboratory contractors have asked whether appropriated funds may be used to reimburse cost of installing electric vehicle recharging stations and to pay electricity bill costs...

350

Hydrogen Fueling Infrastructure Research and Station Technology  

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

box hydrogenandfuelcells.energy.gov Hydrogen Fueling Infrastructure Research and Station Technology Chris Ainscough, Joe Pratt, Jennifer Kurtz, Brian Somerday, Danny Terlip, Terry...

351

Hydrogen Fueling Infrastructure Research and Station Technology...  

Energy Savers (EERE)

Infrastructure Research and Station Technology Download presentation slides from the DOE Fuel Cell Technologies Office webinar "An Overview of the Hydrogen Fueling Infrastructure...

352

Alternative Fueling Station Locator | Department of Energy  

Energy Savers (EERE)

your browser to a new version. U.S. Department of Energy Energy Efficiency and Renewable Energy Source: Alternative Fuels Data Center Find alternative fueling stations near an...

353

New Buildings at the Fire Research Station  

Science Journals Connector (OSTI)

... THE new buildings of the Fire Research Station A of the Joint Fire Research Organization of the Department ... chemicals will be studied in the new chemical engineering laboratory.

D. I. LAWSON

1961-01-14T23:59:59.000Z

354

Model based multivariable controller for large scale compression stations. Design and experimental validation on the LHC 18KW cryorefrigerator  

SciTech Connect

In this paper, a multivariable model-based non-linear controller for Warm Compression Stations (WCS) is proposed. The strategy is to replace all the PID loops controlling the WCS with an optimally designed model-based multivariable loop. This new strategy leads to high stability and fast disturbance rejection such as those induced by a turbine or a compressor stop, a key-aspect in the case of large scale cryogenic refrigeration. The proposed control scheme can be used to have precise control of every pressure in normal operation or to stabilize and control the cryoplant under high variation of thermal loads (such as a pulsed heat load expected to take place in future fusion reactors such as those expected in the cryogenic cooling systems of the International Thermonuclear Experimental Reactor ITER or the Japan Torus-60 Super Advanced fusion experiment JT-60SA). The paper details how to set the WCS model up to synthesize the Linear Quadratic Optimal feedback gain and how to use it. After preliminary tuning at CEA-Grenoble on the 400W@1.8K helium test facility, the controller has been implemented on a Schneider PLC and fully tested first on the CERN's real-time simulator. Then, it was experimentally validated on a real CERN cryoplant. The efficiency of the solution is experimentally assessed using a reasonable operating scenario of start and stop of compressors and cryogenic turbines. This work is partially supported through the European Fusion Development Agreement (EFDA) Goal Oriented Training Program, task agreement WP10-GOT-GIRO.

Bonne, François; Bonnay, Patrick [INAC, SBT, UMR-E 9004 CEA/UJF-Grenoble, 17 rue des Martyrs, 38054 Grenoble (France); Alamir, Mazen [Gipsa-Lab, Control Systems Department, CNRS-University of Grenoble, 11, rue des Mathématiques, BP 46, 38402 Saint Martin d'Hères (France); Bradu, Benjamin [CERN, CH-1211 Genève 23 (Switzerland)

2014-01-29T23:59:59.000Z

355

Savannah River Site Removes Dome, Opening Reactor for Recovery Act  

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

Savannah River Site Removes Dome, Opening Reactor for Recovery Act Savannah River Site Removes Dome, Opening Reactor for Recovery Act Decommissioning Savannah River Site Removes Dome, Opening Reactor for Recovery Act Decommissioning American Recovery and Reinvestment Act workers achieved a significant milestone in the decommissioning of a Cold War reactor at the Savannah River Site this month after they safely removed its rusty, orange, 75-foot-tall dome. With the help of a 660-ton crane and lifting lugs, the workers pulled the 174,000-pound dome off the Heavy Water Components Test Reactor, capping more than 16 months of preparations. Savannah River Site Removes Dome, Opening Reactor for Recovery Act Decommissioning More Documents & Publications Recovery Act Changes Hanford Skyline with Explosive Demolitions Recovery Act Workers Add Time Capsule Before Sealing Reactor for Hundreds

356

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

357

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

358

An Assessment of the Near-Term Costs of Hydrogen Refueling Stations and Station Components  

E-Print Network (OSTI)

4-12: Hydrogen Cost Comparison for Electrolysis Station WithAnalysis: Electrolysis, 30 kg/day, grid Hydrogen Cost ($/kg)the hydrogen costs from the HSCM for electrolysis stations

Lipman, T E; Weinert, Jonathan X.

2006-01-01T23:59:59.000Z

359

Daily Reporting Rainfall Station TULLY & JOHNSTONE RIVERS Manual Heavy Rainfall Station  

E-Print Network (OSTI)

Daily Reporting Rainfall Station TULLY & JOHNSTONE RIVERS Manual Heavy Rainfall Station Manual Tide TM Bulgun Ck AL Bingil Bay The Boulders TM Nerada AL Tung Oil AL Fishers Ck TM Corsis AL Russell

Greenslade, Diana

360

Daily Reporting Rainfall Station TULLY & JOHNSTONE RIVERS Manual Heavy Rainfall Station  

E-Print Network (OSTI)

Daily Reporting Rainfall Station TULLY & JOHNSTONE RIVERS Manual Heavy Rainfall Station Manual The Boulders TM Nerada AL Tung Oil AL Fishers Ck TM Corsis AL RussellR Babinda Clyde Rd AL Central Mill AL

Greenslade, Diana

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

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...

362

DESIGNING AN OPPORTUNITY FUEL WITH BIOMASS AND TIRE-DERIVED FUEL FOR COFIRING AT WILLOW ISLAND GENERATING STATION AND COFIRING SAWDUST WITH COAL AT ALBRIGHT GENERATING STATION  

SciTech Connect

During the period July 1, 2001--September 30, 2001, Allegheny Energy Supply Co., LLC (Allegheny) continued construction of the Willow Island cofiring project, completed the installation of the fuel storage facility, the fuel receiving facility, and the processing building. All mechanical equipment has been installed and electrical construction has proceeded. During this time period significant short term testing of the Albright Generating Station cofiring facility was completed, and the 100-hour test was planned for early October. The testing demonstrated that cofiring at the Albright Generating Station could contribute to a ''4P Strategy''--reduction of SO{sub 2}, NO{sub x}, mercury, and greenhouse gas emissions. This report summarizes the activities associated with the Designer Opportunity Fuel program, and demonstrations at Willow Island and Albright Generating Stations. It details the construction activities at both sites along with the combustion modeling at the Willow Island site.

K. Payette; D. Tillman

2001-10-01T23:59:59.000Z

363

An Assessment of the Near-Term Costs of Hydrogen Refueling Stations and Station Components  

E-Print Network (OSTI)

4: Energy Station Grid electricity Cogen Heat Exhaust (CO2)Recycled Reformate Grid electricity Cogen Heat Electricity

Weinert, Jonathan X.; Lipman, Timothy

2006-01-01T23:59:59.000Z

364

Daily Reporting Rainfall Station MULGRAVE-RUSSELL RIVERS Manual Heavy Rainfall Station  

E-Print Network (OSTI)

Station Telemetry Rainfall Station Telemetry River Station Revised: Nov 2011 MAP 111.1 FLOOD WARNING ulgraveR The Boulders TM RussellR Clyde Rd AL Babinda Bucklands TM Daradgee McAvoy Br AL 0 5 10 kilometres

Greenslade, Diana

365

Using mixed integer programming to schedule experiments on the Texas A&M University nuclear science center reactor  

E-Print Network (OSTI)

. The reactor core locations are modeled as three machines running at different rates; A2 at rate Y~ q, A4 and A6 at rate q. These machines collectively constitute Station 1. The storage area, in which samples are allowed to decay, constitutes Station 2.... Since there is no contention for storage space, the model considers the number of machines at Station 2 to be infinite. We use p&, , to denote the processing time (i. e. , number of hours) of job j at Station s (s=1, 2). Processing times can...

Drees, Lawrence David

2012-06-07T23:59:59.000Z

366

Systems and methods for dismantling a nuclear reactor  

DOE Patents (OSTI)

Systems and methods for dismantling a nuclear reactor are described. In one aspect the system includes a remotely controlled heavy manipulator ("manipulator") operatively coupled to a support structure, and a control station in a non-contaminated portion of a workspace. The support structure provides the manipulator with top down access into a bioshield of a nuclear reactor. At least one computing device in the control station provides remote control to perform operations including: (a) dismantling, using the manipulator, a graphite moderator, concrete walls, and a ceiling of the bioshield, the manipulator being provided with automated access to all internal portions of the bioshield; (b) loading, using the manipulator, contaminated graphite blocks from the graphite core and other components from the bioshield into one or more waste containers; and (c) dispersing, using the manipulator, dust suppression and contamination fixing spray to contaminated matter.

Heim, Robert R; Adams, Scott Ryan; Cole, Matthew Denver; Kirby, William E; Linnebur, Paul Damon

2014-10-28T23:59:59.000Z

367

Background noise spectra of global seismic stations  

SciTech Connect

Over an extended period of time station noise spectra were collected from various sources for use in estimating the detection and location performance of global networks of seismic stations. As the database of noise spectra enlarged and duplicate entries became available, an effort was mounted to more carefully select station noise spectra while discarding others. This report discusses the methodology and criteria by which the noise spectra were selected. It also identifies and illustrates the station noise spectra which survived the selection process and which currently contribute to the modeling efforts. The resulting catalog of noise statistics not only benefits those who model network performance but also those who wish to select stations on the basis of their noise level as may occur in designing networks or in selecting seismological data for analysis on the basis of station noise level. In view of the various ways by which station noise were estimated by the different contributors, it is advisable that future efforts which predict network performance have available station noise data and spectral estimation methods which are compatible with the statistics underlying seismic noise. This appropriately requires (1) averaging noise over seasonal and/or diurnal cycles, (2) averaging noise over time intervals comparable to those employed by actual detectors, and (3) using logarithmic measures of the noise.

Wada, M.M.; Claassen, J.P.

1996-08-01T23:59:59.000Z

368

Oben: Die Station im Watt bei Spiekeroog.  

E-Print Network (OSTI)

Oben: Die Station im Watt bei Spiekeroog. Unten: Ausschnitt des innen begeh- baren Pfahlrohrs mit Ossietzky Universität Oldenburg Meeresdaten rund um die Uhr: Die Station im Watt Von Rainer Reuter In autumn Rückseiten- watt und offener Nordsee. Die Umwelt zu beobachten und verläss- liche Messungen zu gewinnen, ist

Oldenburg, Carl von Ossietzky Universität

369

On-Site Hydrogen Generation & Refueling Station  

E-Print Network (OSTI)

Reforming based refueling station DOE Objectives Public education on hydrogen and fuel cells Evaluation cell & HCNG busses in commercial operation ­ Refueling fuel cell & HCNG street sweepers and cars Thermal Reforming of natural gas Demonstrate hydrogen fueling station #12;Performance goals 15 min per bus

370

Validation of an Integrated Hydrogen Energy Station  

SciTech Connect

This report presents the results of a 10-year project conducted by Air Products and Chemicals, Inc. (Air Products) to determine the feasibility of coproducing hydrogen with electricity. The primary objective was to demonstrate the technical and economic viability of a hydrogen energy station using a high-temperature fuel cell designed to produce power and hydrogen. This four-phase project had intermediate go/no-go decisions and the following specific goals: �¢���¢ Complete a technical assessment and economic analysis of the use of high-temperature fuel cells, including solid oxide and molten carbonate, for the co-production of power and hydrogen (energy park concept). �¢���¢ Build on the experience gained at the Las Vegas H2 Energy Station and compare/contrast the two approaches for co-production. �¢���¢ Determine the applicability of co-production from a high-temperature fuel cell for the existing merchant hydrogen market and for the emerging hydrogen economy. �¢���¢ Demonstrate the concept on natural gas for six months at a suitable site with demand for both hydrogen and electricity. �¢���¢ Maintain safety as the top priority in the system design and operation. �¢���¢ Obtain adequate operational data to provide the basis for future commercial activities, including hydrogen fueling stations. Work began with the execution of the cooperative agreement with DOE on 30 September 2001. During Phase 1, Air Products identified high-temperature fuel cells as having the potential to meet the coproduction targets, and the molten carbonate fuel cell system from FuelCell Energy, Inc. (FuelCell Energy) was selected by Air Products and DOE following the feasibility assessment performed during Phase 2. Detailed design, construction and shop validation testing of a system to produce 250 kW of electricity and 100 kilograms per day of hydrogen, along with site selection to include a renewable feedstock for the fuel cell, were completed in Phase 3. The system also completed six months of demonstration operation at the wastewater treatment facility operated by Orange County Sanitation District (OCSD, Fountain Valley, CA). As part of achieving the objective of operating on a renewable feedstock, Air Products secured additional funding via an award from the California Air Resources Board. The South Coast Air Quality Management District also provided cost share which supported the objectives of this project. System operation at OCSD confirmed the results from shop validation testing performed during Phase 3. Hydrogen was produced at rates and purity that met the targets from the system design basis, and coproduction efficiency exceeded the 50% target set in conjunction with input from the DOE. Hydrogen production economics, updated from the Phase 2 analysis, showed pricing of $5 to $6 per kilogram of hydrogen using current gas purification systems. Hydrogen costs under $3 per kilogram are achievable if next-generation electrochemical separation technologies become available.

Edward C. Heydorn

2012-10-26T23:59:59.000Z

371

Generating Unstructured Nuclear Reactor Core Meshes in Parallel  

Science Journals Connector (OSTI)

Abstract Recent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor core examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a ¼ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed.

Rajeev Jain; Timothy J. Tautges

2014-01-01T23:59:59.000Z

372

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),

373

Seventh annual report of the Texas Agricultural Experiment Station for the year 1894.  

E-Print Network (OSTI)

.. Agric~~lturist- .................................. J. W. CARSON, B. S.. .Assistant to Director. .............................. A. M. SOULE, B. S. A. .Assistant. Agriculturist. SUB-STATION SUPERINTENDENTS. ..................................... J... conducted for the past season at McKinney nncl TVichita Falls Sub-Stations. Bulletin No. 30, treating of glanders experiment, tuberculin tests, lumpy jaw of cattle, notes on animal parasites, Texas fever experiment, and device for destroying ticks...

Connell, J. H.; Harrington, H. H.; Francis, M.; Price, R. H.; Adriance, Duncan; Clayton, James

1895-01-01T23:59:59.000Z

374

Power Reactor Progress  

Science Journals Connector (OSTI)

Argonne kicks off EBWR; Allis-Chalmers plans power reactor using both nuclear and conventional fuels ... NUCLEAR POWER took two giant steps last week. ... Just as the first nuclear power system in the U. S. designed and built solely for the generation of electric power went into full operation at Argonne, Allis-Chalmers came up with a new twist in power reactors—a controlled recirculation boiling reactor (CRBR) using both nuclear and conventional fuels (C&EN, Feb. 18, page 7). ...

1957-02-25T23:59:59.000Z

375

Improved vortex reactor system  

DOE Patents (OSTI)

An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

Diebold, J.P.; Scahill, J.W.

1995-05-09T23:59:59.000Z

376

Search for Neutrino Oscillations at the Palo Verde Nuclear Reactors  

E-Print Network (OSTI)

We report on the initial results from a measurement of the anti-neutrino flux and spectrum at a distance of about 800 m from the three reactors of the Palo Verde Nuclear Generating Station using a segmented gadolinium-loaded scintillation detector. We find that the anti-neutrino flux agrees with that predicted in the absence of oscillations to better than 5%, excluding at 90% CL $\\rm\\bar\

F. Boehm; J. Busenitz; B. Cook; G. Gratta; H. Henrikson; J. Kornis; D. Lawrence; K. B. Lee; K. McKinny; L. Miller; V. Novikov; A. Piepke; B. Ritchie; D. Tracy; P. Vogel; Y-F. Wang; J. Wolf

1999-12-22T23:59:59.000Z

377

Alternative Fuels Data Center: Compressed Natural Gas Fueling Stations  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

Compressed Natural Gas Compressed Natural Gas Fueling Stations to someone by E-mail Share Alternative Fuels Data Center: Compressed Natural Gas Fueling Stations on Facebook Tweet about Alternative Fuels Data Center: Compressed Natural Gas Fueling Stations on Twitter Bookmark Alternative Fuels Data Center: Compressed Natural Gas Fueling Stations on Google Bookmark Alternative Fuels Data Center: Compressed Natural Gas Fueling Stations on Delicious Rank Alternative Fuels Data Center: Compressed Natural Gas Fueling Stations on Digg Find More places to share Alternative Fuels Data Center: Compressed Natural Gas Fueling Stations on AddThis.com... More in this section... Natural Gas Basics Benefits & Considerations Stations Locations Infrastructure Development Compressed Natural Gas Stations

378

Alternative Fuels Data Center: Electric Vehicle Charging Stations  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

Electric Vehicle Electric Vehicle Charging Stations to someone by E-mail Share Alternative Fuels Data Center: Electric Vehicle Charging Stations on Facebook Tweet about Alternative Fuels Data Center: Electric Vehicle Charging Stations on Twitter Bookmark Alternative Fuels Data Center: Electric Vehicle Charging Stations on Google Bookmark Alternative Fuels Data Center: Electric Vehicle Charging Stations on Delicious Rank Alternative Fuels Data Center: Electric Vehicle Charging Stations on Digg Find More places to share Alternative Fuels Data Center: Electric Vehicle Charging Stations on AddThis.com... More in this section... Electricity Basics Benefits & Considerations Stations Locations Infrastructure Development Vehicles Laws & Incentives Electric Vehicle Charging Stations

379

Assessing low power reactivity levels in subcritical CANDU reactors  

SciTech Connect

A new technique has been developed for monitoring slow reactivity changes in Canadian deuterium uranium (CANDU) nuclear reactors during low power operation, following a sustained period of high power operation. The power doubling halving worth (PDHW) test tracks slow reactivity changes by evaluating the reactivity perturbation required by the power regulating system to halve and double reactor power over time. During low power operation of a CANDU reactor, the PDHW test is used to monitor the decay of the photoneutron precursors, so that the reactor power can be lowered using a preset amount of reactivity. The PDHW test is described in this paper and is validated by using computer simulations and operating data from a CANDU reactor.

Teare, S.W. [Corporation of the City of Mississauga, Ontario (Canada)] [Corporation of the City of Mississauga, Ontario (Canada); Shanes, F.C. [Ontario Hydro, Toronto, Ontario (Canada)] [Ontario Hydro, Toronto, Ontario (Canada); Hersey, M.W. [Pickering Nuclear Generating Station, Ontario (Canada)] [Pickering Nuclear Generating Station, Ontario (Canada)

1996-08-01T23:59:59.000Z

380

AEC Pushes Fusion Reactors  

Science Journals Connector (OSTI)

AEC Pushes Fusion Reactors ... Project Sherwood, as the study program is called, began in 1951-52 soon after the first successful thermonuclear explosion in the Pacific. ...

1955-10-10T23:59:59.000Z

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

Tokamak reactor first wall  

DOE Patents (OSTI)

This invention relates to an improved first wall construction for a tokamak fusion reactor vessel, or other vessels subjected to similar pressure and thermal stresses.

Creedon, R.L.; Levine, H.E.; Wong, C.; Battaglia, J.

1984-11-20T23:59:59.000Z

382

Advanced Nuclear Research Reactor  

SciTech Connect

This report describes technical modifications implemented by INVAP to improve the safety of the Research Reactors the company designs and builds.

Lolich, J.V.

2004-10-06T23:59:59.000Z

383

LNG to CNG refueling stations  

SciTech Connect

While the fleet operator is concerned about the environment, he or she is going to make the choice based primarily on economics. Which fuel provides the lowest total operating cost? The calculation of this costing must include the price-per-gallon of the fuel delivered, as well as the tangible and intangible components of fuel delivery, such as downtime for vehicles during the refueling process, idle time for drivers during refueling, emissions costings resulting from compressor oil blow-by, inclusion of non-combustible constituents in the CNG, and energy consumption during the refueling process. Also, the upfront capital requirement of similar delivery capabilities must be compared. The use of LNG as the base resource for the delivered CNG, in conjunction with the utilization of a fully temperature-compressed LNG/CNG refueling system, eliminates many of the perceived shortfalls of CNG. An LNG/CNG refueling center designed to match the capabilities of the compressor-based station will have approximately the same initial capital requirement. However, because it derives its CNG sales product from the {minus}260 F LNG base product, thus availing itself of the natural physical properties of the cryogenic product, all other economic elements of the system favor the LNG/CNG product.

Branson, J.D. [ECOGAS Corp., Austin, TX (United States)

1995-12-31T23:59:59.000Z

384

Burner balancing Salem Harbor Station  

SciTech Connect

The traditional method of burner balancing is first to determine the fuel distribution, then to measure the economizer outlet excess oxygen distribution and to adjust the burners accordingly. Fuel distribution is typically measured by clean and dirty air probing. Coal pipe flow can then be adjusted, if necessary, through the use of coal pipe orificing or by other means. Primary air flow must be adjusted to meet the design criteria of the burner. Once coal pipe flow is balanced to within the desired criteria, secondary air flow to individual burners can be changed by adjusting windbox dampers, burner registers, shrouds or other devices in the secondary air stream. This paper discusses problems encountered in measuring excess O{sub 2} at the economizer outlet. It is important to recognize that O{sub 2} measurements at the economizer outlet, by themselves, can be very misleading. If measurement problems are suspected or encountered, an alternate approach similar to that described should be considered. The alternate method is not only useful for burner balancing but also can be used to help in calibrating the plant excess O{sub 2} instruments and provide an on line means of cross-checking excess air measurements. Balanced burners operate closer to their design stoichiometry, providing better NO{sub x} reduction. For Salem Harbor Station, this means a significant saving in urea consumption.

Sload, A.W. [New England Power Co., Salem, MA (United States); Dube, R.J. [DB Riley, Inc., Worcester, MA (United States). Fuel Equipment Design

1995-12-31T23:59:59.000Z

385

CASL Test Stands . . . Enabling the CASL mission  

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

Test Stands . . . Enabling the CASL mission In a nuclear reactor core there are many phenomena in motion at both the microscopic and macroscopic levels that ultimately affect the...

386

A TEN MEGAWATT BOILING HETEROGENEOUS PACKAGE POWER REACTOR. Reactor...  

Office of Scientific and Technical Information (OSTI)

A TEN MEGAWATT BOILING HETEROGENEOUS PACKAGE POWER REACTOR. Reactor Design and Feasibility Problem Re-direct Destination: Temp Data Fields Rosen, M. A.; Coburn, D. B.; Flynn, T....

387

Portfolio for fast reactor collaboration  

SciTech Connect

The development of the LMFBR type reactor in the United Kingdom is reviewed. Design characteristics of a commercial demonstration fast reactor are presented and compared with the Super Phenix reactor.

Rippon, S.

1981-12-01T23:59:59.000Z

388

Handbook of Reactor Physics  

Science Journals Connector (OSTI)

... THIS handbook is one volume in a series sponsored by the United States Atomic Energy Commission with ... data and reference information in the field of reactors. The volume is devoted to reactor physics and radiation shielding, the latter subject occupying approximately a quarter of the book.

PETER W. MUMMERY

1956-08-25T23:59:59.000Z

389

Fast reactor safety  

Science Journals Connector (OSTI)

... SIR, - In his article on fast reactor safety (26 July, page 270) Norman Dombey claims to introduce to non-specialists ... , page 270) Norman Dombey claims to introduce to non-specialists some features of fast reactors that are not available outside the technical literature. The non-specialist would do well ...

R.D. SMITH

1979-08-23T23:59:59.000Z

390

Instrumentation of Nuclear Reactors  

Science Journals Connector (OSTI)

... s Lecture Theatre on January 8, a symposium of papers on the instrumentation of nuclear reactors was organized, at which about five hundred members and visitors attended, including guests from ... the Institution, took the chair and introduced Sir John Cockcroft, whose lecture on "Nuclear Reactors and their Applications" provided a general background for the three specialized papers which followed. ...

1953-03-07T23:59:59.000Z

391

Nuclear Research Reactors  

Science Journals Connector (OSTI)

... their countries for the advent of nuclear power. A few countries had built large research reactors for the production of isotopes and to study the behaviour of nuclear fuel, but ... production of isotopes and to study the behaviour of nuclear fuel, but the small training reactor had not been developed. Since then, research ...

T. E. ALLIBONE

1963-07-20T23:59:59.000Z

392

23rd steam-station cost survey  

SciTech Connect

The results of the 23rd Steam Station Cost Survey covering the year 1982 are summarized. The major categories of the survey are as follows: general data; output data, 1982; fuel consumption, 1982; operation 1982 (mills/net kWh); investment ($/net kWh); energy cost, 1982 (mills/net kWh); and station performance, 1982. Thirty-one fossil-fuel steam plants and four nuclear stations were included in the survey. Fuel and operating cost increases are felt to be responsible for the moderate rise in total busbar-enery costs. 11 figures, 1 table.

Friedlander, G.D.; Going, M.C.

1983-11-01T23:59:59.000Z

393

A Near-Term Economic Analysis of Hydrogen Fueling Stations  

E-Print Network (OSTI)

0.07/kWh has on hydrogen cost for electrolysis type station.3-12: Hydrogen Cost Comparison for Electrolysis Station,3-12: Hydrogen Cost Comparison for Electrolysis Station, NAS

Weinert, Jonathan X.

2005-01-01T23:59:59.000Z

394

A Near-term Economic Analysis of Hydrogen Fueling Stations  

E-Print Network (OSTI)

0.07/kWh has on hydrogen cost for electrolysis type station.3-12: Hydrogen Cost Comparison for Electrolysis Station,3-12: Hydrogen Cost Comparison for Electrolysis Station, NAS

Weinert, Jonathan X.

2005-01-01T23:59:59.000Z

395

University Climatological Station Compiled by Erwin Wodarczak (1998)  

E-Print Network (OSTI)

University Climatological Station Committee fonds Compiled by Erwin Wodarczak (1998) #12;Fonds Description University Climatological Station Committee fonds. ­ 1961-1974. 3 cm of textual records. Administrative History In 1954 a President's Climatological Station Committee was established

Handy, Todd C.

396

A Near-Term Economic Analysis of Hydrogen Fueling Stations  

E-Print Network (OSTI)

based on industry experiences with natural gas stations.Few natural gas stations have yet to achieve a 47% capacitynts 0 .2 % of to tal gas stations. Achieving low co st hydr

Weinert, Jonathan X.

2005-01-01T23:59:59.000Z

397

A Near-term Economic Analysis of Hydrogen Fueling Stations  

E-Print Network (OSTI)

based on industry experiences with natural gas stations.Few natural gas stations have yet to achieve a 47% capacitynts 0 .2 % of to tal gas stations. Achieving low co st hydr

Weinert, Jonathan X.

2005-01-01T23:59:59.000Z

398

The Fuel-Travel-Back Approach to Hydrogen Station Siting  

E-Print Network (OSTI)

only 18% of existing gas station number is needed to achievean intersection like 4-corner gas stations in real life, butis only 708 or 18% of gas stations in the study region. This

Lin, Zhenhong; Ogden, Joan; Fan, Yueyue; Chen, Chien-Wei

2009-01-01T23:59:59.000Z

399

The integral fast reactor fuel cycle  

SciTech Connect

The liquid-metal reactor (LMR) has the potential to extend the uranium resource by a factor of 50 to 100 over current commercial light water reactors (LWRs). In the integral fast reactor (IFR) development program, the entire reactor system - reactor, fuel cycle, and waste process - is being developed and optimized at the same time as a single integral entity. A key feature of the IFR concept is the metallic fuel. The lead irradiation tests on the new U-Pu-Zr metallic fuel in the Experimental Breeder Reactor II have surpassed 185000 MWd/t burnup, and its high burnup capability has now been fully demonstrated. The metallic fuel also allows a radically improved fuel cycle technology. Pyroprocessing, which utilizes high temperatures and molten salt and molten metal solvents, can be advantageously utilized for processing metal fuels because the product is metal suitable for fabrication into new fuel elements. Direct production of a metal product avoids expensive and cumbersome chemical conversion steps that would result from use of the conventional Purex solvent extraction process. The key step in the IFR process is electrorefining, which provides for recovery of the valuable fuel constituents, uranium and plutonium, and for removal of fission products. A notable feature of the IFR process is that the actinide elements accompany plutonium through the process. This results in a major advantage in the high-level waste management.

Chang, Y.I. (Argonne National Lab., IL (United States))

1990-01-01T23:59:59.000Z

400

Canadian university research reactors  

SciTech Connect

In Canada there are seven university research reactors: one medium-power (2-MW) swimming pool reactor at McMaster University and six low-power (20-kW) SLOWPOKE reactors at Dalhousie University, Ecole Polytechnique, the Royal Military College, the University of Toronto, the University of Saskatchewan, and the University of Alberta. This paper describes primarily the McMaster Nuclear Reactor (MNR), which operates on a wider scale than the SLOWPOKE reactors. The MNR has over a hundred user groups and is a very broad-based tool. The main applications are in the following areas: (1) neutron activation analysis (NAA); (2) isotope production; (3) neutron beam research; (4) nuclear engineering; (5) neutron radiography; and (6) nuclear physics.

Ernst, P.C.; Collins, M.F.

1989-11-01T23:59:59.000Z

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

Reactor & Nuclear Systems Publications | ORNL  

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

Nuclear Science Home | Science & Discovery | Nuclear Science | Publications and Reports | Reactor and Nuclear Systems Publications SHARE Reactor and Nuclear Systems Publications...

402

Revised analyses of decommissioning for the reference boiling water reactor power station. Effects of current regulatory and other considerations on the financial assurance requirements of the decommissioning rule and on estimates of occupational radiation exposure: Main report, draft report for comment. Volume 1  

SciTech Connect

On June 27, 1988, the U.S. Nuclear Regulatory Commission (NRC) published in the Federal Register (53 FR 24018) the final rule for the General Requirements for Decommissioning Nuclear Facilities. With the issuance of the final rule, owners and operators of licensed nuclear power plants are required to prepare, and submit to the NRC for review, decommissioning plans and cost estimates. The NRC staff is in need of updated bases documentation that will assist them in assessing the adequacy of the licensee submittals, from the viewpoint of both the planned actions, including occupational radiation exposure, and the probable costs. The purpose of this reevaluation study is to update the needed bases documentation. This report presents the results of a review and reevaluation of the PNL 1980 decommissioning study of the Washington Public Power Supply System`s Washington Nuclear Plant Two (WNP-2), including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the plant for the DECON, SAFSTOR, and ENTOMB alternatives, which now include an initial 5-7 year period during which time the spent fuel is stored in the spent fuel pool prior to beginning major disassembly or extended safe storage of the plant. This report also includes consideration of the NRC requirement that decontamination and decommissioning activities leading to termination of the nuclear license be completed within 60 years of final reactor shutdown, consideration of packaging and disposal requirements for materials whose radionuclide concentrations exceed the limits for Class C low-level waste. Costs for labor, transport, and disposal activities are given in 1993 dollars. Sensitivities of the total license termination cost to the disposal costs at different low-level radioactive waste disposal sites, to different depths of contaminated concrete surface removal within the facilities, and to different transport distances are also examined.

Smith, R.I.; Bierschbach, M.C.; Konzek, G.J. [Pacific Northwest Lab., Richland, WA (United States)] [and others

1994-09-01T23:59:59.000Z

403

The CANDU Reactor System: An Appropriate Technology  

Science Journals Connector (OSTI)

...apparent. The devel-opment of Zircaloy-2 by the U.S. Bettis Laboratory in Pittsburgh, made this con-cept practicable...necessary to base a reactor de-sign on Zircaloy-2 had not Bettis work-ers been testing the material in NRX as part of a collaborative...

J. A. L. Robertson

1978-02-10T23:59:59.000Z

404

Creation of a neutrino laboratory for search for sterile neutrino at SM-3 reactor  

E-Print Network (OSTI)

In connection with the question of possible existence of sterile neutrino the laboratory on the basis of SM-3 reactor was created to search for oscillations of reactor antineutrino. A prototype of a neutrino detector with scintillator volume of 400 l can be moved at the distance of 6-11 m from the reactor core. The measurements of background conditions have been made. It is shown that the main experimental problem is associated with cosmic radiation background. Test measurements of dependence of a reactor antineutrino flux on the distance from a reactor core have been made. The prospects of search for oscillations of reactor antineutrino at short distances are discussed.

Serebrov, A P; Samoylov, R M; Fomin, A K; Zinoviev, V G; Neustroev, P V; Golovtsov, V L; Gruzinsky, N V; Solovey, V A; Cherniy, A V; Zherebtsov, O M; Martemyanov, V P; Zinoev, V G; Tarasenkov, V G; Aleshin, V I; Petelin, A L; Pavlov, S V; Izhutov, A L; Sazontov, S A; Ryazanov, D K; Gromov, M O; Afanasiev, V V; Matrosov, L N; Matrosova, M Yu

2015-01-01T23:59:59.000Z

405

Alternative Fuels Data Center: Hydrogen Fueling Station Locations  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

Hydrogen Hydrogen Printable Version Share this resource Send a link to Alternative Fuels Data Center: Hydrogen Fueling Station Locations to someone by E-mail Share Alternative Fuels Data Center: Hydrogen Fueling Station Locations on Facebook Tweet about Alternative Fuels Data Center: Hydrogen Fueling Station Locations on Twitter Bookmark Alternative Fuels Data Center: Hydrogen Fueling Station Locations on Google Bookmark Alternative Fuels Data Center: Hydrogen Fueling Station Locations on Delicious Rank Alternative Fuels Data Center: Hydrogen Fueling Station Locations on Digg Find More places to share Alternative Fuels Data Center: Hydrogen Fueling Station Locations on AddThis.com... More in this section... Hydrogen Basics Benefits & Considerations Stations Locations

406

Alternative Fuels Data Center: Propane Fueling Station Locations  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

Propane Propane Printable Version Share this resource Send a link to Alternative Fuels Data Center: Propane Fueling Station Locations to someone by E-mail Share Alternative Fuels Data Center: Propane Fueling Station Locations on Facebook Tweet about Alternative Fuels Data Center: Propane Fueling Station Locations on Twitter Bookmark Alternative Fuels Data Center: Propane Fueling Station Locations on Google Bookmark Alternative Fuels Data Center: Propane Fueling Station Locations on Delicious Rank Alternative Fuels Data Center: Propane Fueling Station Locations on Digg Find More places to share Alternative Fuels Data Center: Propane Fueling Station Locations on AddThis.com... More in this section... Propane Basics Benefits & Considerations Stations Locations Infrastructure Development

407

Fact #717: March 5, 2012 Availability of Electric Charging Stations...  

Energy Savers (EERE)

7: March 5, 2012 Availability of Electric Charging Stations Has Increased Dramatically in Recent Years Fact 717: March 5, 2012 Availability of Electric Charging Stations Has...

408

Hydrogen Fueling Station in Honolulu, Hawaii Feasibility Analysis...  

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

Station in Honolulu, Hawaii Feasibility Analysis Hydrogen Fueling Station in Honolulu, Hawaii Feasibility Analysis This feasibility report assesses the technical and economic...

409

Alternative Fuels Data Center: Alternative Fueling Station Locator  

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

Locate Stations Printable Version Share this resource Send a link to Alternative Fuels Data Center: Alternative Fueling Station Locator to someone by E-mail Share Alternative Fuels...

410

Climate analysis with satellite versus weather station data  

Science Journals Connector (OSTI)

This paper compares how well satellite versus weather station measurements of climate predict agricultural performance ... , India, and the United States. Although weather stations give accurate measures of groun...

Robert Mendelsohn; Pradeep Kurukulasuriya; Alan Basist; Felix Kogan…

2007-03-01T23:59:59.000Z

411

Alternative Fueling Station Locator App Provides Info at Your...  

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

Fueling Station Locator website. It provides information on more than 15,000 public and private alternative fueling stations throughout the United States. The app lists where...

412

November 10, 2004: First hydrogen refueling station opens in...  

Energy Savers (EERE)

refueling station in Washington, D.C. The station will be used to refuel General Motors' fuel cell vehicles in DOE's Vehicle and Infrastructure Learning Demonstration and...

413

H2FIRST: Hydrogen Fueling Infrastructure Research and Station...  

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

FIRST: Hydrogen Fueling Infrastructure Research and Station Technology H2FIRST: Hydrogen Fueling Infrastructure Research and Station Technology Hydrogen Fueling Infrastructure...

414

Alternative Fuels Data Center: Electric Vehicle Charging Station Locations  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

Electric Vehicle Electric Vehicle Charging Station Locations to someone by E-mail Share Alternative Fuels Data Center: Electric Vehicle Charging Station Locations on Facebook Tweet about Alternative Fuels Data Center: Electric Vehicle Charging Station Locations on Twitter Bookmark Alternative Fuels Data Center: Electric Vehicle Charging Station Locations on Google Bookmark Alternative Fuels Data Center: Electric Vehicle Charging Station Locations on Delicious Rank Alternative Fuels Data Center: Electric Vehicle Charging Station Locations on Digg Find More places to share Alternative Fuels Data Center: Electric Vehicle Charging Station Locations on AddThis.com... More in this section... Electricity Basics Benefits & Considerations Stations Locations Infrastructure Development

415

Schiller Station Biomass Facility | Open Energy Information  

Open Energy Info (EERE)

Station Biomass Facility Station Biomass Facility Jump to: navigation, search Name Schiller Station Biomass Facility Facility Schiller Station Sector Biomass Owner PSNH Location Portsmouth, New Hampshire Coordinates 43.0717552°, -70.7625532° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":43.0717552,"lon":-70.7625532,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

416

Boiler House and Power Station Chemistry  

Science Journals Connector (OSTI)

... and power stations". It provides a useful background of information on the properties and combustion of ... of coals, and on such subjects as the treatment of boiler feed water, types of oil ...

A. PARKER

1949-01-01T23:59:59.000Z

417

The College Station Residential Energy Compliance Code  

E-Print Network (OSTI)

The City of College Station, Texas adopted a new residential Energy Compliance Code in January, 1988. The code, which strengthens compliance requirements in several areas, has received broadly based support and acceptance from all major constituent...

Claridge, D. E.; Schrock, D.

1988-01-01T23:59:59.000Z

418

Hydrogen Refueling Station Costs in Shanghai  

E-Print Network (OSTI)

stations and vendors (e.g. Air Products and Chemicals, Inc,including Chevron and Air Products and Chemicals, Inc. , asDiesel a. Verified with Air Products representative, Feb

Weinert, Jonathan X.; Shaojun, Liu; Ogden, J; Jianxin, Ma

2006-01-01T23:59:59.000Z

419

Hydrogen Fueling Infrastructure Research and Station Technology  

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

Presentation slides from the DOE Fuel Cell Technologies Office webinar "An Overview of the Hydrogen Fueling Infrastructure Research and Station Technology (H2FIRST) Project" held on November 18, 2014.

420

Hydrogen Fueling Infrastructure Research and Station Technology  

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

contaminants from downstream sources - L2: must be integrated with <87.5 MPa and > -40C gas - L3: burden on vehicle OEM; many more cars than stations Possible HCD Location...

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

IntegratedScienceWorkingforYou Research Station  

E-Print Network (OSTI)

. . . . . . . . . . . . . . . . . . .5 Colorado's forest resources, 2002-2006 . . . . . . . . .6 Post Mountain Research Station Federal Recycling Program Printed on Recycled Paper The Rocky Mountain Research organi- zation--the most extensive natural resources research organization in the world. We maintain 14

422

Field Station Contributors, 2010 Acton, Gary  

E-Print Network (OSTI)

Field Station Contributors, 2010 Acton, Gary Allen, Deborah and Harry Ardell, Robert and Lee Baker Mead, Judson and Jane Mead, Thomas and Lenore Merritt, Andrew and Eleanor Morganwalp, David and Jill

Polly, David

423

Solar and Infrared Radiation Station (SIRS) Handbook  

SciTech Connect

The Solar Infrared Radiation Station (SIRS) provides continuous measurements of broadband shortwave (solar) and longwave (atmospheric or infrared) irradiances for downwelling and upwelling components. The following six irradiance measurements are collected from a network of stations to help determine the total radiative flux exchange within the Atmospheric Radiation Measurement (ARM) Southern Great Plains (SGP) Climate Research Facility: • Direct normal shortwave (solar beam) • Diffuse horizontal shortwave (sky) • Global horizontal shortwave (total hemispheric) • Upwelling shortwave (reflected) • Downwelling longwave (atmospheric infrared) • Upwelling longwave (surface infrared)

Stoffel, T

2005-07-01T23:59:59.000Z

424

Light Water Reactors [Corrosion and Mechanics of Materials] - Nuclear  

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

Light Water Reactors Light Water Reactors Capabilities Materials Testing Environmentally Assisted Cracking (EAC) of Reactor Materials Corrosion Performance/Metal Dusting Overview Light Water Reactors Fatigue Testing of Carbon Steels and Low-Alloy Steels Environmentally Assisted Cracking of Ni-Base Alloys Irradiation-Induced Stress Corrosion Cracking of Austenitic Stainless Steels Steam Generator Tube Integrity Program Air Oxidation Kinetics for Zr-based Alloys Fossil Energy Fusion Energy Metal Dusting Publications List Irradiated Materials Steam Generator Tube Integrity Other Facilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Corrosion and Mechanics of Materials Light Water Reactors Bookmark and Share To continue safe operation of current LWRs, the aging degradation of the

425

Gas-Cooled Fast Reactor Program. Annual progress report for period ending December 31, 1979  

SciTech Connect

Information on the GCFR reactor is presented concerning the Core Flow Test Loop; shielding and physics; pressure vessel and closure studies; and irradiation program.

Gat, U.; Kasten, P.R.

1980-11-01T23:59:59.000Z

426

Nuclear reactor control column  

DOE Patents (OSTI)

The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

Bachovchin, Dennis M. (Plum Borough, PA)

1982-01-01T23:59:59.000Z

427

Transient thermal analysis of a space reactor power system  

E-Print Network (OSTI)

Thermoelectric Power Conversion Module Heat Pipe Radiator Module . Auxiliary Modules . Flow of Calculation . Transient Test Cases Studied Summary . 10 10 CHAPTER II. ENERGY EQUATION FINITE DIFFERENCING . . 12 Energy Equation for a Solid Finite..., but this stud~ uses a generic liquid metal cooled fast reactor concept as the model to test the code. The space power svstem to be modeled consists of a liquid lithium cooled fast reactor, primarv and secondary loops svith a sell-induced thermoelectric...

Gaeta, Michael J.

1988-01-01T23:59:59.000Z

428

Alternative Fuels Data Center: Biodiesel Fueling Station Locations  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

Station Locations to someone by E-mail Station Locations to someone by E-mail Share Alternative Fuels Data Center: Biodiesel Fueling Station Locations on Facebook Tweet about Alternative Fuels Data Center: Biodiesel Fueling Station Locations on Twitter Bookmark Alternative Fuels Data Center: Biodiesel Fueling Station Locations on Google Bookmark Alternative Fuels Data Center: Biodiesel Fueling Station Locations on Delicious Rank Alternative Fuels Data Center: Biodiesel Fueling Station Locations on Digg Find More places to share Alternative Fuels Data Center: Biodiesel Fueling Station Locations on AddThis.com... More in this section... Biodiesel Basics Benefits & Considerations Stations Locations Infrastructure Development Vehicles Laws & Incentives Biodiesel Fueling Station Locations Find biodiesel (B20 and above) fueling stations near an address or ZIP code

429

Alternative Fuels Data Center: Natural Gas Fueling Stations  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

Fueling Fueling Stations to someone by E-mail Share Alternative Fuels Data Center: Natural Gas Fueling Stations on Facebook Tweet about Alternative Fuels Data Center: Natural Gas Fueling Stations on Twitter Bookmark Alternative Fuels Data Center: Natural Gas Fueling Stations on Google Bookmark Alternative Fuels Data Center: Natural Gas Fueling Stations on Delicious Rank Alternative Fuels Data Center: Natural Gas Fueling Stations on Digg Find More places to share Alternative Fuels Data Center: Natural Gas Fueling Stations on AddThis.com... More in this section... Natural Gas Basics Benefits & Considerations Stations Locations Infrastructure Development Vehicles Laws & Incentives Natural Gas Fueling Stations Photo of a compressed natural gas fueling station. Hundreds of compressed natural gas (CNG) fueling stations are available in

430

Testing the ae \\Lambda scaling of thermal transport models: predicted and measured temperatures in the Tokamak Fusion Test  

E-Print Network (OSTI)

in the Tokamak Fusion Test Reactor dimensionless scaling experiments D. R. Mikkelsen, S. D. Scott Princeton the Tokamak Fusion Test Reactor [D. J. Grove and D. M. Meade, Nucl. Fusion 25, 1167 (1985)] nondimensional to extrapo� late [1] from current experiments to International Tokamak Experimental Reactor [2] (ITER) class

431

Coal dust exposure among power station workers during normal operations at Hatfield's Ferry Power Station.  

E-Print Network (OSTI)

??Changes in coal composition could produce higher levels of coal dust exposure thanthose found in the past at Hatfield's Ferry Power Station. Air sampling was… (more)

Lewis, Christian S.

2008-01-01T23:59:59.000Z

432

Nuclear reactor reflector  

DOE Patents (OSTI)

A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.

Hopkins, Ronald J. (Pensacola, FL); Land, John T. (Pensacola, FL); Misvel, Michael C. (Pensacola, FL)

1994-01-01T23:59:59.000Z

433

Spherical torus fusion reactor  

DOE Patents (OSTI)

The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.

Martin Peng, Y.K.M.

1985-10-03T23:59:59.000Z

434

Measurements of Collimator Wakefields at End Station A  

SciTech Connect

The angular deflection of a 28.5 GeV electron beam passing off-axis between the jaws of a collimator, generating a transverse wakefield, were measured in End Station A (ESA) at SLAC. In total, fifteen different configurations of collimator geometry and material were tested: some were chosen for compatibility with previous measurements while others served to study the effect of geometry and taper angles (geometrical contribution to the wakefield) and the effect of the material resistivity (resistive contribution) to the imparted kick. This paper summarises the last update of preliminary experimental results before they are finalised. The reconstructed kick factor is compared to analytical calculations and simulations.

Fernandez-Hernando, J.L.; /Daresbury; Molloy, S.; /SLAC; Smith, J.D.A.; /Cockcroft Inst. Accel. Sci. Tech.; Watson, Nigel Keith; /Birmingham U.

2011-11-01T23:59:59.000Z

435

Field Testing of a Wet FGD Additive for Enhanced Mercury Control - Pilot-Scale Test Results  

SciTech Connect

This Topical Report summarizes progress on Cooperative Agreement DE-FC26-04NT42309, ''Field Testing of a Wet FGD Additive.'' The objective of the project is to demonstrate the use of a flue gas desulfurization (FGD) additive, Degussa Corporation's TMT-15, to prevent the reemissions of elemental mercury (Hg{sup 0}) in flue gas exiting wet FGD systems on coal-fired boilers. Furthermore, the project intends to demonstrate that the additive can be used to precipitate most of the mercury (Hg) removed in the wet FGD system as a fine TMT salt that can be separated from the FGD liquor and bulk solid byproducts for separate disposal. The project will conduct pilot and full-scale tests of the TMT-15 additive in wet FGD absorbers. The tests are intended to determine required additive dosage requirements to prevent Hg{sup 0} reemissions and to separate mercury from the normal FGD byproducts for three coal types: Texas lignite/Power River Basin (PRB) coal blend, high-sulfur Eastern bituminous coal, and low-sulfur Eastern bituminous coal. The project team consists of URS Group, Inc., EPRI, TXU Generation Company LP, Southern Company, and Degussa Corporation. TXU Generation has provided the Texas lignite/PRB co-fired test site for pilot FGD tests, Monticello Steam Electric Station Unit 3. Southern Company is providing the low-sulfur Eastern bituminous coal host site for wet scrubbing tests, as well as the pilot and full-scale jet bubbling reactor (JBR) FGD systems to be tested. A third utility, to be named later, will provide the high-sulfur Eastern bituminous coal full-scale FGD test site. Degussa Corporation is providing the TMT-15 additive and technical support to the test program. The project is being conducted in six tasks. Of the six project tasks, Task 1 involves project planning and Task 6 involves management and reporting. The other four tasks involve field testing on FGD systems, either at pilot or full scale. The four tasks include: Task 2 - Pilot Additive Testing in Texas Lignite Flue Gas; Task 3 - Full-scale FGD Additive Testing in High Sulfur Eastern Bituminous Flue Gas; Task 4 - Pilot Wet Scrubber Additive Tests at Yates; and Task 5 - Full-scale Additive Tests at Plant Yates. This topical report presents the results from the Task 2 and Task 4 pilot-scale additive tests. The Task 3 and Task 5 full-scale additive tests will be conducted later in calendar year 2006.

Gary M. Blythe

2006-03-01T23:59:59.000Z

436

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

437

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

438

Reactor hot spot analysis  

SciTech Connect

The principle methods for performing reactor hot spot analysis are reviewed and examined for potential use in the Applied Physics Division. The semistatistical horizontal method is recommended for future work and is now available as an option in the SE2-ANL core thermal hydraulic code. The semistatistical horizontal method is applied to a small LMR to illustrate the calculation of cladding midwall and fuel centerline hot spot temperatures. The example includes a listing of uncertainties, estimates for their magnitudes, computation of hot spot subfactor values and calculation of two sigma temperatures. A review of the uncertainties that affect liquid metal fast reactors is also presented. It was found that hot spot subfactor magnitudes are strongly dependent on the reactor design and therefore reactor specific details must be carefully studied. 13 refs., 1 fig., 5 tabs.

Vilim, R.B.

1985-08-01T23:59:59.000Z

439

Molten metal reactors  

DOE Patents (OSTI)

A molten metal reactor for converting a carbon material and steam into a gas comprising hydrogen, carbon monoxide, and carbon dioxide is disclosed. The reactor includes an interior crucible having a portion contained within an exterior crucible. The interior crucible includes an inlet and an outlet; the outlet leads to the exterior crucible and may comprise a diffuser. The exterior crucible may contain a molten alkaline metal compound. Contained between the exterior crucible and the interior crucible is at least one baffle.

Bingham, Dennis N; Klingler, Kerry M; Turner, Terry D; Wilding, Bruce M

2013-11-05T23:59:59.000Z

440

Testing WIMS-D4M cross sections and the ANL ENDF/B-V 69 group library. Results from global diffusion and Monte Carlo calculations compared with measurements in the Romanian 14-MW TRIGA reactor  

SciTech Connect

The WIMS-D4 code has been modified (WIMS-D4M) to produce microscopic isotopic cross sections in ISOTXS format for use in diffusion and transport calculations. Beginning with 69-group libraries based on ENDF/B-V data, numerous cell calculations have been made to prepare a set of broad group cross sections for use in diffusion calculations. Global calculations have been made for two control rod states of the Romanian steady state TRIGA reactor with 29 fresh HEU fuel clusters. Detailed Monte Carlo calculations also have been performed for the same reactor configurations using data based on ENDF/B-V. Results from these global calculations are compared with each other and with the measured excess reactivities. Although region-averaged macroscopic principal cross sections obtained from WIMS-D4M are in good agreement with the corresponding Monte Carlo values, problems exist with the high energy (E > 10 keV) microscopic hydrogen transport cross sections.

Bretscher, M.M.

1993-12-31T23:59:59.000Z

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

Hydrogen Fueling - Coming Soon to a Station Near You (Brochure)  

SciTech Connect

Fact sheet providing information useful to local permitting officials facing hydrogen fueling station proposals.

Not Available

2009-04-01T23:59:59.000Z

442

The LOFAR Super Station concept : an input for discussion  

E-Print Network (OSTI)

throughout Europe : several stations in Germany1 , UK2 , soon a French LOFAR station in Nançay3 , and further contacts in several other countries4 FLOW context · Official decision for funding the French station taken MHz) o Incoherent addition of the same stations to synthesize a broad instantaneous beam (maximum

Demoulin, Pascal

443

F Reactor Inspection  

ScienceCinema (OSTI)

Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2014-11-24T23:59:59.000Z

444

F Reactor Inspection  

SciTech Connect

Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2014-10-29T23:59:59.000Z

445

Test Automation Test Automation  

E-Print Network (OSTI)

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

Mousavi, Mohammad

446

Alternative Fueling Station Locator - Mobile | Open Energy Information  

Open Energy Info (EERE)

Fueling Station Locator - Mobile Fueling Station Locator - Mobile Jump to: navigation, search Tool Summary LAUNCH TOOL Name: Alternative Fueling Station Locator - Mobile Agency/Company /Organization: United States Department of Energy Partner: National Renewable Energy Laboratory Sector: Energy Focus Area: Transportation Phase: Evaluate Options, Prepare a Plan Resource Type: Online calculator User Interface: Mobile Device Website: www.afdc.energy.gov/afdc/locator/m/stations/ Web Application Link: www.afdc.energy.gov/afdc/locator/m/stations/ Cost: Free References: National Renewable Energy Laboratory Advanced Vehicles and Fuels Research: Data and Resources[1] Logo: Alternative Fueling Station Locator - Mobile Find fueling stations for your alternative fuel vehicle on-the-go with the

447

Alternative Fuels Data Center: Utility District Natural Gas Fueling Station  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

Utility District Utility District Natural Gas Fueling Station Regulation to someone by E-mail Share Alternative Fuels Data Center: Utility District Natural Gas Fueling Station Regulation on Facebook Tweet about Alternative Fuels Data Center: Utility District Natural Gas Fueling Station Regulation on Twitter Bookmark Alternative Fuels Data Center: Utility District Natural Gas Fueling Station Regulation on Google Bookmark Alternative Fuels Data Center: Utility District Natural Gas Fueling Station Regulation on Delicious Rank Alternative Fuels Data Center: Utility District Natural Gas Fueling Station Regulation on Digg Find More places to share Alternative Fuels Data Center: Utility District Natural Gas Fueling Station Regulation on AddThis.com... More in this section... Federal

448

OPTIMIZING PERFORMANCE OF THE HESKETT STATION  

SciTech Connect

The overall conclusion from this work is that a switch from river sand bed material to limestone at the R.M. Heskett Station would provide substantial benefits to MDU. A switch to limestone would increase the fuel flexibility of the unit, allowing fuels higher in both sodium and sulfur to be burned. The limestone bed can tolerate a much higher buildup of sodium in the bed without agglomeration, allowing either the bed turnover rate to be reduced to half the current sand feed rate for a fuel with equivalent sodium or allow a higher sodium fuel to be burned with limestone feed rates equivalent to the current sand feed rate. Both stack and ambient SO{sub 2} emissions can be controlled. A small improvement in boiler efficiency should be achievable by operating at lower excess oxygen levels at low load. This reduction in oxygen will also lower NO{sub x} emissions, providing a margin of safety for meeting emission standards. No detrimental effects of using limestone at the Heskett Station were uncovered as a result of the test burn. Some specific conclusions from this work include the following: The bed material feed rate can be reduced from the current rate of 5.4% of the coal feed rate (57.4 tons of sand/day) to 2.5% of the coal feed rate (27 tons of limestone/day). This will result in an annual savings of approximately $200,000. (1) SO{sub 2} emissions at the recommended feed rate would be approximately 250 ppm (0.82 lb/MMBtu) using a similar lignite. Based on the cost of the limestones, SO{sub 2} allowances could be generated at a cost of $60/ton SO{sub 2} , leaving a large profit margin for the sale of allowances. The addition of limestone at the same rate currently used for sand feed could generate $455,000 net income if allowances are sold at $200/ton SO2 . (2) At full-load operation, unburned carbon losses increase significantly at excess oxygen levels below 2.8%. No efficiency gains are expected at high-load operation by switching from sand to limestone. By reducing the oxygen level at low load to 8.5%, an efficiency gain of approximately 1.2% could be realized, equating to $25,000 to $30,000 in annual savings. (3) A reduction of 25 tons/day total ash (bed material plus fly ash) will be realized by using limestone at the recommended feed rate compared to the current sand feed rate. No measurable change in volume would be realized because of the lower bulk density of the limestone-derived material.

Michael D. Mann; Ann K. Henderson

1999-03-01T23:59:59.000Z

449

Alternative Fuels Data Center: Ethanol Fueling Station Locations  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

Fueling Fueling Station Locations to someone by E-mail Share Alternative Fuels Data Center: Ethanol Fueling Station Locations on Facebook Tweet about Alternative Fuels Data Center: Ethanol Fueling Station Locations on Twitter Bookmark Alternative Fuels Data Center: Ethanol Fueling Station Locations on Google Bookmark Alternative Fuels Data Center: Ethanol Fueling Station Locations on Delicious Rank Alternative Fuels Data Center: Ethanol Fueling Station Locations on Digg Find More places to share Alternative Fuels Data Center: Ethanol Fueling Station Locations on AddThis.com... More in this section... Ethanol Basics Benefits & Considerations Stations Locations Infrastructure Development Vehicles Laws & Incentives Ethanol Fueling Station Locations Find ethanol (E85) fueling stations near an address or ZIP code or along a

450

Alternative Fuels Data Center: About the Alternative Fueling Station Data  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

Locate Stations Locate Stations Printable Version Share this resource Send a link to Alternative Fuels Data Center: About the Alternative Fueling Station Data to someone by E-mail Share Alternative Fuels Data Center: About the Alternative Fueling Station Data on Facebook Tweet about Alternative Fuels Data Center: About the Alternative Fueling Station Data on Twitter Bookmark Alternative Fuels Data Center: About the Alternative Fueling Station Data on Google Bookmark Alternative Fuels Data Center: About the Alternative Fueling Station Data on Delicious Rank Alternative Fuels Data Center: About the Alternative Fueling Station Data on Digg Find More places to share Alternative Fuels Data Center: About the Alternative Fueling Station Data on AddThis.com... About the Alternative Fueling Station Data

451

Alternative Fuels Data Center: Alternative Fueling Station Counts by State  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

Locate Stations Locate Stations Printable Version Share this resource Send a link to Alternative Fuels Data Center: Alternative Fueling Station Counts by State to someone by E-mail Share Alternative Fuels Data Center: Alternative Fueling Station Counts by State on Facebook Tweet about Alternative Fuels Data Center: Alternative Fueling Station Counts by State on Twitter Bookmark Alternative Fuels Data Center: Alternative Fueling Station Counts by State on Google Bookmark Alternative Fuels Data Center: Alternative Fueling Station Counts by State on Delicious Rank Alternative Fuels Data Center: Alternative Fueling Station Counts by State on Digg Find More places to share Alternative Fuels Data Center: Alternative Fueling Station Counts by State on AddThis.com... Alternative Fueling Station Counts by State

452

E-Print Network 3.0 - anna power station Sample Search Results  

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

station Search Powered by Explorit Topic List Advanced Search Sample search results for: anna power station...

453

E-Print Network 3.0 - anna power stations Sample Search Results  

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

stations Search Powered by Explorit Topic List Advanced Search Sample search results for: anna power stations...

454

GASIFICATION TEST RUN TC06  

SciTech Connect

This report discusses test campaign TC06 of the Kellogg Brown & Root, Inc. (KBR) Transport Reactor train with a Siemens Westinghouse Power Corporation (Siemens Westinghouse) particle filter system at the Power Systems Development Facility (PSDF) located in Wilsonville, Alabama. The Transport Reactor is an advanced circulating fluidized-bed reactor designed to operate as either a combustor or a gasifier using a particulate control device (PCD). The Transport Reactor was operated as a pressurized gasifier during TC06. Test run TC06 was started on July 4, 2001, and completed on September 24, 2001, with an interruption in service between July 25, 2001, and August 19, 2001, due to a filter element failure in the PCD caused by abnormal operating conditions while tuning the main air compressor. The reactor temperature was varied between 1,725 and 1,825 F at pressures from 190 to 230 psig. In TC06, 1,214 hours of solid circulation and 1,025 hours of coal feed were attained with 797 hours of coal feed after the filter element failure. Both reactor and PCD operations were stable during the test run with a stable baseline pressure drop. Due to its length and stability, the TC06 test run provided valuable data necessary to analyze long-term reactor operations and to identify necessary modifications to improve equipment and process performance as well as progressing the goal of many thousands of hours of filter element exposure.

Southern Company Services, Inc.

2003-08-01T23:59:59.000Z

455

FUSION POWER Tokamak Test a Big Success  

Science Journals Connector (OSTI)

FUSION POWER Tokamak Test a Big Success ... Successful plasma production in the tokamak fusion test reactor at Princeton University's Plasma Physics Laboratory has set the stage for possible sustained fusion reactions in the device by 1990. ...

1983-01-10T23:59:59.000Z

456

Methods for quantifying uncertainty in fast reactor analyses.  

SciTech Connect

Liquid-metal-cooled fast reactors in the form of sodium-cooled fast reactors have been successfully built and tested in the U.S. and throughout the world. However, no fast reactor has operated in the U.S. for nearly fourteen years. More importantly, the U.S. has not constructed a fast reactor in nearly 30 years. In addition to reestablishing the necessary industrial infrastructure, the development, testing, and licensing of a new, advanced fast reactor concept will likely require a significant base technology program that will rely more heavily on modeling and simulation than has been done in the past. The ability to quantify uncertainty in modeling and simulations will be an important part of any experimental program and can provide added confidence that established design limits and safety margins are appropriate. In addition, there is an increasing demand from the nuclear industry for best-estimate analysis methods to provide confidence bounds along with their results. The ability to quantify uncertainty will be an important component of modeling that is used to support design, testing, and experimental programs. Three avenues of UQ investigation are proposed. Two relatively new approaches are described which can be directly coupled to simulation codes currently being developed under the Advanced Simulation and Modeling program within the Reactor Campaign. A third approach, based on robust Monte Carlo methods, can be used in conjunction with existing reactor analysis codes as a means of verification and validation of the more detailed approaches.

Fanning, T. H.; Fischer, P. F.

2008-04-07T23:59:59.000Z

457

POST 10/Truck Inspection Station (Map 3  

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

POST 10/Truck Inspection Station (Map 3) POST 10/Truck Inspection Station (Map 3) Changes Effective January 11, 2010 Pajarito Corridor Deliveries: Drivers of commercial delivery trucks headed to the Pajarito Corridor (Pajarito Road bounded by NM Highway 4 and Diamond Drive) must stop at Post 10 for truck inspections. Drivers will then need to present time-stamped inspection passes from Post 10 to protective force officers stationed at the Pajarito Corridor. (Drivers exiting Post 10 should (1) turn right and proceed west on the Truck Route; (2) turn left onto West Jemez Road; (3) proceed to Lane 7; (4) STOP and present the inspection pass to the protective force officer; (5) turn left onto Diamond

458

Rawhide Energy Station, Fort Collins, Colorado  

SciTech Connect

The staff of Platte River Power Authority's Rawhide Energy Station have been racking up operating stats and an environmental performance record that is the envy of other plant managers. In the past decade Rawhide has enjoyed an equivalent availability factor in the mid to high 90s and an average capacity factor approaching 90%. Still not content with this performance, Rawhide invested in new technology and equipment upgrades to further optimise performance, reduce emissions, and keep cost competitive. The Energy Station includes four GE France 7EA natural gas-fired turbines totalling 260 MW and a 274 MW coal-fired unit located in northeastern Colorado. 7 figs.

Peltier, R.

2008-10-15T23:59:59.000Z

459

B Reactor | Department of Energy  

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

Operational Management » History » Manhattan Project » Signature Operational Management » History » Manhattan Project » Signature Facilities » B Reactor B Reactor B Reactor Completed in September 1944, the B Reactor was the world's first large-scale plutonium production reactor. As at Oak Ridge, the need for labor turned Hanford into an atomic boomtown, with the population reaching 50,000 by summer 1944. Similar to the X-10 Graphite Reactor at Oak Ridge in terms of loading and unloading fuel, the B Reactor was built on a much larger scale and used water rather than air as a coolant. Whereas the X-10 had an initial design output of 1,000 kilowatts, the B Reactor was designed to operate at 250,000 kilowatts. Consisting of a 28- by 36-foot, 1,200-ton graphite cylinder lying on its side, the reactor was penetrated through its

460

Development of Materials for Supercritical-Water-Cooled Reactor |  

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

Development of Materials for Supercritical-Water-Cooled Reactor Development of Materials for Supercritical-Water-Cooled Reactor Development of Materials for Supercritical-Water-Cooled Reactor Supercritical-Water-Cooled Reactor (SCWR) was selected as one of the promising candidates in Generation IV reactors for its prominent advantages; those are the high thermal efficiency, the system simplification, the R&D cost minimization and the flexibility for core design. As the demand for advanced nuclear system increases, Japanese R&D project started in 1999 aiming to provide technical information essential to demonstration of SCPR technologies through three sub-themes of 1. Plant conceptual design, 2. Thermal-hydraulics, and 3. Material. Although the material development is critical issue of SCWR development, previous studies were limited for the screening tests on commercial alloys

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461

High Flux Isotope Reactor (HFIR) | Nuclear Science | ORNL  

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

High Flux Isotope Reactor High Flux Isotope Reactor May 30, 2013 The High Flux Isotope Reactor (HFIR) first achieved criticality on August 25, 1965, and achieved full power in August 1966. It is a versatile 85-MW isotope production, research, and test reactor with the capability and facilities for performing a wide variety of irradiation experiments and a world-class neutron scattering science program. HFIR is a beryllium-reflected, light water-cooled and moderated flux-trap type swimming pool reactor that uses highly enriched uranium-235 as fuel. HFIR typically operates seven 23-to-27 day cycles per year. Irradiation facility capabilities include Flux trap positions: Peak thermal flux of 2.5X1015 n/cm2/s with similar epithermal and fast fluxes (Highest thermal flux available in the

462

JAPANESE REACTOR SAGA CONTINUES  

Science Journals Connector (OSTI)

NUCLEAR POWER: Japanese engineers fight to regain control; Congress reviews U.S. nuclear plant practices ... RADIATION CONCERNS continued to heighten last week at Japan’s Fukushima Daiichi Nuclear Power Station, which suffered catastrophic damage from the March 11 earthquake and tsunami. ...

JEFF JOHNSON

2011-04-04T23:59:59.000Z

463

Shipppingport Station Decommissioning Project decommissioning plan. Volume 10  

SciTech Connect

Appendices are presented concerning activation analysis; reactor and intervals removal; reactor vessel removal; escalation effects; removal of contaminated concrete; and removal of contaminated chanbers.

Not Available

1983-01-01T23:59:59.000Z

464

SPERT Destructive Test - I on Aluminum, Highly Enriched Plate Type Core  

ScienceCinema (OSTI)

SPERT - Special Power Excursion Reactor Tests Destructive Test number 1 On Aluminum, Highly Enriched Plate Type Core. A test studying the behavior of the reactor under destructive conditions on a light water moderated pool-type reactor with a plate-type core.

None

2014-05-07T23:59:59.000Z

465

Alternative Fuels Data Center: Natural Gas Station Property Tax Reduction  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

Natural Gas Station Natural Gas Station Property Tax Reduction to someone by E-mail Share Alternative Fuels Data Center: Natural Gas Station Property Tax Reduction on Facebook Tweet about Alternative Fuels Data Center: Natural Gas Station Property Tax Reduction on Twitter Bookmark Alternative Fuels Data Center: Natural Gas Station Property Tax Reduction on Google Bookmark Alternative Fuels Data Center: Natural Gas Station Property Tax Reduction on Delicious Rank Alternative Fuels Data Center: Natural Gas Station Property Tax Reduction on Digg Find More places to share Alternative Fuels Data Center: Natural Gas Station Property Tax Reduction on AddThis.com... More in this section... Federal State Advanced Search All Laws & Incentives Sorted by Type Natural Gas Station Property Tax Reduction

466

Alternative Fuels Data Center: Alternative Fueling Station Grant Program  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

Fueling Fueling Station Grant Program to someone by E-mail Share Alternative Fuels Data Center: Alternative Fueling Station Grant Program on Facebook Tweet about Alternative F