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Sample records for rural pub pwr

  1. Polk County Rural Pub Pwr Dist | Open Energy Information

    Open Energy Info (EERE)

    Polk County Rural Pub Pwr Dist Jump to: navigation, search Name: Polk County Rural Pub Pwr Dist Place: Nebraska Phone Number: (888) 242-5265 Website: www.pcrppd.com Outage...

  2. Northwest Rural Pub Pwr Dist | Open Energy Information

    Open Energy Info (EERE)

    Data Utility Id 13805 Utility Location Yes Ownership P NERC Location WECC NERC MRO Yes RTO SPP Yes Activity Distribution Yes This article is a stub. You can help OpenEI by...

  3. Seward County Rrl Pub Pwr Dist | Open Energy Information

    Open Energy Info (EERE)

    Data Utility Id 16954 Utility Location Yes Ownership P NERC Location SPP NERC SPP Yes RTO SPP Yes Activity Distribution Yes This article is a stub. You can help OpenEI by...

  4. MICROBOONE-NOTE-1017-PUB

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

    7-PUB v1.0 A Method to Extract the Charge Distribution Arriving at the TPC Wire Planes in MicroBooNE The MicroBooNE Collaboration (Dated: June 29, 2016) In this technote, we describe the concept and general strategy of LArTPC drifted-charge extraction, which converts the raw digitized TPC waveform to the number of ionized electrons passing through the wire plane at a given time. The proper recovery of the number of ionized electrons from all wire planes is important to the success of the

  5. Keosauqua Municipal Light & Pwr | Open Energy Information

    Open Energy Info (EERE)

    Keosauqua Municipal Light & Pwr Jump to: navigation, search Name: Keosauqua Municipal Light & Pwr Place: Iowa Phone Number: 319-293-3406 Website: villagesofvanburen.comdirecto...

  6. Cal. Pub. Res. Code 30330 | Open Energy Information

    Open Energy Info (EERE)

    Pub. Res. Code 30330Legal Abstract Cal. Pub. Res. Code 30330, current through August 5, 2014. Published NA Year Signed or Took Effect 1976 Legal Citation Cal. Pub. Res....

  7. Cal. Pub. Res. Code 30600 | Open Energy Information

    Open Energy Info (EERE)

    Pub. Res. Code 30600Legal Abstract Cal. Pub. Res. Code 30600, current through August 5, 2014. Published NA Year Signed or Took Effect 1976 Legal Citation Cal. Pub. Res....

  8. Cal. Pub. Res. Code 30413 | Open Energy Information

    Open Energy Info (EERE)

    Pub. Res. Code 30413Legal Abstract Cal. Pub. Res. Code 30413, current through August 5, 2014. Published NA Year Signed or Took Effect 1991 Legal Citation Cal. Pub. Res....

  9. Cal. Pub. Res. Code 30264 | Open Energy Information

    Open Energy Info (EERE)

    Pub. Res. Code 30264Legal Abstract Cal. Pub. Res. Code 30264, current through August 5, 2014. Published NA Year Signed or Took Effect 1976 Legal Citation Cal. Pub. Res....

  10. Central Illinois Pub Serv Co (Illinois) EIA Revenue and Sales...

    Open Energy Info (EERE)

    Central Illinois Pub Serv Co (Illinois) EIA Revenue and Sales - June 2008 Jump to: navigation, search EIA Monthly Electric Utility Sales and Revenue Data for Central Illinois Pub...

  11. Preliminary study on direct recycling of spent PWR fuel in PWR...

    Office of Scientific and Technical Information (OSTI)

    Preliminary study on direct recycling of spent PWR fuel in PWR system Citation Details ... conference on advances in nuclear science and engineering, Bali (Indonesia), 14-17 ...

  12. MICROBOONE-NOTE-1005-PUB May

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

    5-PUB May 30, 2016 Cosmic Shielding Studies at MicroBooNE The MicroBooNE Collaboration MicroBooNE, a 89 ton active volume liquid argon time projection chamber (LArTPC) located on the Fermilab's Booster Neutrino Beamline, is designed to both probe neutrino physics phenomena, and further the development of LArTPC detector technology and event reconstruction. Since MicroBooNE is located only a few meters below the ground level, without significant shielding, a large flux of cosmic rays are expected

  13. MICROBOONE-NOTE-1006-PUB Study Towards

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

    6-PUB Study Towards an Event Selection for Neutral Current Inclusive Single π 0 Production in MicroBooNE The MicroBooNE Collaboration May 3, 2016 Abstract This note describes a Monte Carlo study towards an event selection for the neutrino induced Neutral Current (NC) single π 0 cross section using the MicroBooNE detector. The NC single π 0 channel is especially important in neutrino oscillation experiments as it is one of the main backgrounds in any ν e or ν e appearance search. This

  14. MICROBOONE-NOTE-1016-PUB Noise Characterization

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

    6-PUB Noise Characterization and Filtering in the MicroBooNE TPC The MicroBooNE Collaboration (Dated: July 5, 2016) In large liquid argon time projection chambers (LArTPCs), TPC signal processing, which recovers the number of ionized electrons arriving at the anode plane from the raw digitized wire signals, is a crucial step towards automated event reconstruction. The first stage of signal processing is the identification and removal of any excess TPC noise with minimal impact on the true

  15. Cal. Pub. Res. Code 30103 | Open Energy Information

    Open Energy Info (EERE)

    search OpenEI Reference LibraryAdd to library Legal Document- StatuteStatute: Cal. Pub. Res. Code 30103Legal Abstract Cal. Pub. Res. Code 30103, current through...

  16. Cal. Pub. Util. Code 1001 | Open Energy Information

    Open Energy Info (EERE)

    search OpenEI Reference LibraryAdd to library Legal Document- StatuteStatute: Cal. Pub. Util. Code 1001Legal Abstract Cal. Pub. Util. Code 1001, current through...

  17. Cal. Pub. Res. Code 6370 | Open Energy Information

    Open Energy Info (EERE)

    Cal. Pub. Res. Code 6370 Jump to: navigation, search OpenEI Reference LibraryAdd to library Legal Document- StatuteStatute: Cal. Pub. Res. Code 6370Legal Abstract Statutory...

  18. Cal. Pub. Util. Code 1001 | Open Energy Information

    Open Energy Info (EERE)

    Util. Code 1001 (Redirected from Cal. Pub. Res. Code 1001) Jump to: navigation, search OpenEI Reference LibraryAdd to library Legal Document- StatuteStatute: Cal. Pub. Util....

  19. Cal. Pub. Res. Code 25309 | Open Energy Information

    Open Energy Info (EERE)

    search OpenEI Reference LibraryAdd to library Legal Document- StatuteStatute: Cal. Pub. Res. Code 25309Legal Abstract Cal. Pub. Res. Code 25309, current through...

  20. Cal. Pub. Res. Code 6826 | Open Energy Information

    Open Energy Info (EERE)

    search OpenEI Reference LibraryAdd to library Legal Document- StatuteStatute: Cal. Pub. Res. Code 6826Legal Abstract Cal. Pub. Res. Code 6826, current through...

  1. Cal. Pub. Res. Code 25541 | Open Energy Information

    Open Energy Info (EERE)

    search OpenEI Reference LibraryAdd to library Legal Document- StatuteStatute: Cal. Pub. Res. Code 25541Legal Abstract Cal. Pub. Res. Code 25541, current through...

  2. Cal. Pub. Res. Code 25120 | Open Energy Information

    Open Energy Info (EERE)

    search OpenEI Reference LibraryAdd to library Legal Document- StatuteStatute: Cal. Pub. Res. Code 25120Legal Abstract Cal. Pub. Res. Code 25120, current through...

  3. Cal. Pub. Res. 25500 et seq | Open Energy Information

    Open Energy Info (EERE)

    search OpenEI Reference LibraryAdd to library Legal Document- StatuteStatute: Cal. Pub. Res. 25500 et seqLegal Abstract Cal. Pub. Res. Code 25531, current through...

  4. Central Nebraska Pub P&I Dist | Open Energy Information

    Open Energy Info (EERE)

    Pub P&I Dist Jump to: navigation, search Name: Central Nebraska Pub P&I Dist Place: Nebraska Phone Number: 308.995.8601 Website: www.nppd.com Twitter: @nppdnews Facebook: https:...

  5. PWR AXIAL BURNUP PROFILE ANALYSIS

    SciTech Connect (OSTI)

    J.M. Acaglione

    2003-09-17

    The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B&W 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001).

  6. Cal. Pub. Res. Code 21001 | Open Energy Information

    Open Energy Info (EERE)

    Cal. Pub. Res. Code 21001Legal Abstract Sets forth California's policy on maintenance of environmental quality. Published NA Year Signed or Took Effect 1970 Legal...

  7. Cal. Pub. Res. Code 21080 | Open Energy Information

    Open Energy Info (EERE)

    library Legal Document- StatuteStatute: Cal. Pub. Res. Code 21080Legal Abstract Sets forth general statutory provisions for California's environmental quality programs....

  8. Central Illinois Pub Serv Co (Illinois) EIA Revenue and Sales...

    Open Energy Info (EERE)

    March 2008 Jump to: navigation, search EIA Monthly Electric Utility Sales and Revenue Data for Central Illinois Pub Serv Co for March 2008. Monthly Electric Utility Sales and...

  9. Central Illinois Pub Serv Co (Illinois) EIA Revenue and Sales...

    Open Energy Info (EERE)

    Central Illinois Pub Serv Co (Illinois) EIA Revenue and Sales - November 2008 Jump to: navigation, search EIA Monthly Electric Utility Sales and Revenue Data for Central Illinois...

  10. Central Illinois Pub Serv Co (Illinois) EIA Revenue and Sales...

    Open Energy Info (EERE)

    Central Illinois Pub Serv Co (Illinois) EIA Revenue and Sales - September 2008 Jump to: navigation, search EIA Monthly Electric Utility Sales and Revenue Data for Central Illinois...

  11. Central Illinois Pub Serv Co (Illinois) EIA Revenue and Sales...

    Open Energy Info (EERE)

    April 2008 Jump to: navigation, search EIA Monthly Electric Utility Sales and Revenue Data for Central Illinois Pub Serv Co for April 2008. Monthly Electric Utility Sales and...

  12. Central Illinois Pub Serv Co (Illinois) EIA Revenue and Sales...

    Open Energy Info (EERE)

    Central Illinois Pub Serv Co (Illinois) EIA Revenue and Sales - January 2009 Jump to: navigation, search EIA Monthly Electric Utility Sales and Revenue Data for Central Illinois...

  13. Central Illinois Pub Serv Co (Illinois) EIA Revenue and Sales...

    Open Energy Info (EERE)

    January 2008 Jump to: navigation, search EIA Monthly Electric Utility Sales and Revenue Data for Central Illinois Pub Serv Co for January 2008. Monthly Electric Utility Sales and...

  14. Central Illinois Pub Serv Co (Illinois) EIA Revenue and Sales...

    Open Energy Info (EERE)

    October 2008 Jump to: navigation, search EIA Monthly Electric Utility Sales and Revenue Data for Central Illinois Pub Serv Co for October 2008. Monthly Electric Utility Sales and...

  15. Central Illinois Pub Serv Co (Illinois) EIA Revenue and Sales...

    Open Energy Info (EERE)

    Central Illinois Pub Serv Co (Illinois) EIA Revenue and Sales - December 2008 Jump to: navigation, search EIA Monthly Electric Utility Sales and Revenue Data for Central Illinois...

  16. Central Illinois Pub Serv Co (Illinois) EIA Revenue and Sales...

    Open Energy Info (EERE)

    9 Jump to: navigation, search EIA Monthly Electric Utility Sales and Revenue Data for Central Illinois Pub Serv Co for February 2009. Monthly Electric Utility Sales and Revenue...

  17. Central Illinois Pub Serv Co (Illinois) EIA Revenue and Sales...

    Open Energy Info (EERE)

    8 Jump to: navigation, search EIA Monthly Electric Utility Sales and Revenue Data for Central Illinois Pub Serv Co for February 2008. Monthly Electric Utility Sales and Revenue...

  18. Cal. Pub. Res. Code 21067 | Open Energy Information

    Open Energy Info (EERE)

    search OpenEI Reference LibraryAdd to library Legal Document- StatuteStatute: Cal. Pub. Res. Code 21067Legal Abstract Definitions section for California's...

  19. VERA Core Simulator Methodology for PWR Cycle Depletion (Conference...

    Office of Scientific and Technical Information (OSTI)

    VERA Core Simulator Methodology for PWR Cycle Depletion Citation Details In-Document Search Title: VERA Core Simulator Methodology for PWR Cycle Depletion Authors: Kochunas, ...

  20. Central Illinois Pub Serv Co (Illinois) EIA Revenue and Sales...

    Open Energy Info (EERE)

    May 2008 Jump to: navigation, search EIA Monthly Electric Utility Sales and Revenue Data for Central Illinois Pub Serv Co for May 2008. Monthly Electric Utility Sales and Revenue...

  1. PUB-3000 | BERKELEY LAB HEALTH AND SAFETY MANUAL

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

    ES&H MANUAL (PUB-3000) Berkeley Lab Table of Contents Guide to Using the ES&H Manual Responsible Authors Log of ES&H Manual Changes Requesting a Change to the ES&H Manual Search...

  2. Cal. Pub. Res. Code 6301 | Open Energy Information

    Open Energy Info (EERE)

    search OpenEI Reference LibraryAdd to library Legal Document- StatuteStatute: Cal. Pub. Res. Code 6301Legal Abstract This section grants the California State Lands...

  3. Cal. Pub. Res. Code 6502 | Open Energy Information

    Open Energy Info (EERE)

    search OpenEI Reference LibraryAdd to library Legal Document- StatuteStatute: Cal. Pub. Res. Code 6502Legal Abstract Statutory chapter providing for leasing of public...

  4. Cal. Pub. Res. Code 5020 et seq.: Historical Resources | Open...

    Open Energy Info (EERE)

    search OpenEI Reference LibraryAdd to library Legal Document- StatuteStatute: Cal. Pub. Res. Code 5020 et seq.: Historical ResourcesLegal Abstract This section...

  5. Cal. Pub. Res. Code 30603 | Open Energy Information

    Open Energy Info (EERE)

    navigation, search OpenEI Reference LibraryAdd to library Legal Document- OtherOther: Cal. Pub. Res. Code 30603Legal Abstract Delegation of local authority for Coastal Zone...

  6. Spring Valley Pub Utils Comm | Open Energy Information

    Open Energy Info (EERE)

    Name: Spring Valley Pub Utils Comm Place: Minnesota Website: www.smmpa.orgmembersspring-v Outage Hotline: 507.346.7622 References: EIA Form EIA-861 Final Data File for 2010 -...

  7. Red River Valley Coop Pwr Assn | Open Energy Information

    Open Energy Info (EERE)

    Red River Valley Coop Pwr Assn Jump to: navigation, search Name: Red River Valley Coop Pwr Assn Place: Minnesota Website: www.rrvcoop.com Facebook: https:www.facebook.comRRVCPA...

  8. Central Montana E Pwr Coop Inc | Open Energy Information

    Open Energy Info (EERE)

    E Pwr Coop Inc Jump to: navigation, search Name: Central Montana E Pwr Coop Inc Place: Montana Phone Number: 406-268-1211 Website: www.cmepc.org Outage Hotline: 406-268-1211...

  9. Pearl River Valley El Pwr Assn | Open Energy Information

    Open Energy Info (EERE)

    Valley El Pwr Assn Jump to: navigation, search Name: Pearl River Valley El Pwr Assn Place: Mississippi Phone Number: Columbia: 601-736-2666 -- Hattiesburg: 601-264-2458 -- Purvis:...

  10. Northeast Missouri El Pwr Coop | Open Energy Information

    Open Energy Info (EERE)

    Pwr Coop Jump to: navigation, search Name: Northeast Missouri El Pwr Coop Place: Missouri Phone Number: 573-769-2107 Website: www.northeast-power.coop Outage Hotline: 573-769-2107...

  11. Sam Rayburn Municipal Pwr Agny | Open Energy Information

    Open Energy Info (EERE)

    Municipal Pwr Agny Jump to: navigation, search Name: Sam Rayburn Municipal Pwr Agny Place: Texas Phone Number: 936-336-3684 or 936-336-5666 Website: www.cityofliberty.orgGOVERNME...

  12. MICROBOONE-NOTE-1008-PUB Michel Electron Reconstruction Using

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

    8-PUB Michel Electron Reconstruction Using the MicroBooNE LArTPC Cosmic Data The MicroBooNE Collaboration August 25, 2016 Abstract MicroBooNE is a Liquid Argon Time Projection Chamber (LArTPC) neutrino detector located in the Booster Neutrino Beamline at Fermilab which began collecting neutrino data in October 2015. MicroBooNE aims to explore the low-energy excess in the ν e spectrum reported by MiniBooNE as well as perform ν-Ar cross-section measurements. In this note, we present the current

  13. Improving fuel-rod performance. [PWR; BWR

    SciTech Connect (OSTI)

    Ocken, H.; Knott, S.

    1981-03-01

    To reduce the risk of fuel-rod failures, utilities operate their nuclear reactors within conservative limits on power increases proposed by nuclear-fuel vendors. Of particular concern to US utilities is that adopting these limits results in an industrywide average plant capacity loss of 3% in BWR designs and 0.3% in PWR designs. To replace lost BWR capacity by other generating means currently costs the utilities $150 million annually, and losses for PWRs are about $20 million. Efforts are therefore being made to identify the factors responsible for Zircaloy degradation under PCI condition and to improve nuclear-fuel-rod design and reactor operation.

  14. Cal. Pub. Util. Code 201-248 - Public Utilities Act | Open...

    Open Energy Info (EERE)

    from "http:en.openei.orgwindex.php?titleCal.Pub.Util.Code201-248-PublicUtilitiesAct&oldid801671" Feedback Contact needs updating Image needs updating...

  15. South Mississippi El Pwr Assn | Open Energy Information

    Open Energy Info (EERE)

    South Mississippi El Pwr Assn Place: Mississippi Phone Number: 601.268.2083 Website: www.smepa.coop Outage Hotline: 601.268.2083 References: EIA Form EIA-861 Final Data File for...

  16. East Mississippi Elec Pwr Assn | Open Energy Information

    Open Energy Info (EERE)

    search Name: East Mississippi Elec Pwr Assn Place: Mississippi Phone Number: Meridian Office: 601-581-8600 -- Quitman Office: 601-776-6271 -- DeKalb Office: 601-743-2641 --...

  17. Grand Valley Rrl Pwr Line, Inc | Open Energy Information

    Open Energy Info (EERE)

    Valley Rrl Pwr Line, Inc Place: Colorado Website: www.gvp.org Twitter: @GVRuralPower Outage Hotline: 970-242-0040 Outage Map: www.gvp.orgcontentoutage-map References: EIA Form...

  18. Impact of High Burnup on PWR Spent Fuel Characteristics (Journal...

    Office of Scientific and Technical Information (OSTI)

    Reducing the burden of management of spent nuclear fuel is important to the future of nuclear energy. The impact of higher pressurized water reactor (PWR) fuel burnup is examined ...

  19. Effects of Multiple Drying Cycles on HBU PWR Cladding Alloys

    Broader source: Energy.gov [DOE]

    The purpose of this research effort is to determine the effects of canister/cask vacuum drying and storage on radial hydride precipitation in high‐burnup (HBU) pressurized water reactor (PWR)...

  20. PWR representative behavior during a LOCA

    SciTech Connect (OSTI)

    Allison, C.M.

    1981-01-01

    To date, there has been substantial analytical and experimental effort to define the margins between design basis loss-of-coolant accident (LOCA) behavior and regulatory limits on maximum fuel rod cladding temperature and deformation. As a result, there is extensive documentation on the modeling of fuel rod behavior in test reactors and design basis LOCA's. However, modeling of that behavior using representative, non-conservative, operating histories is not nearly as well documented in the public literature. Therefore, the objective of this paper is (a) to present calculations of LOCA induced behavior for Pressurized Water Reactor (PWR) core representative fuel rods, and (b) to discuss the variability in those calculations given the variability in fuel rod condition at the initiation of the LOCA. This analysis was limited to the study of changes in fuel rod behavior due to different power operating histories. The other two important parameters which affect that behavior, initial fuel rod design and LOCA coolant conditions were held invarient for all of the representative rods analyzed.

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

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

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

  2. Microsoft Word - aac2012_Li_1_WG4-SLAC-PUB-15212.doc

    Office of Scientific and Technical Information (OSTI)

    et al., "Results from Plasma Wakefield Experiments at FACET", IPAC'11, San Sebastian, Spain, 2011, SLAC-PUB-14560. 5. E. Adli et al., to be published. 6. S.Z. Li and M.J. Hogan,...

  3. FY 2007 Total System Life Cycle Cost, Pub 2008 | Department of Energy

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

    FY 2007 Total System Life Cycle Cost, Pub 2008 FY 2007 Total System Life Cycle Cost, Pub 2008 The Analysis of the Total System Life Cycle Cost (TSLCC) of the Civilian Radioactive Waste Management Program presents the Office of Civilian Radioactive Waste Management's (OCRWM) May 2007 total system cost estimate for the disposal of the Nation's spent nuclear fuel (SNF) and high-level radioactive waste (HLW). The TSLCC analysis provides a basis for assessing the adequacy of the Nuclear Waste Fund

  4. A Combined Nonfertile and UO{sub 2} PWR Fuel Assembly for Actinide...

    Office of Scientific and Technical Information (OSTI)

    the CONFU assembly exhibits negative reactivity feedback coefficients comparable in ... NUCLEAR FUELS; PWR TYPE REACTORS; REACTIVITY COEFFICIENTS; REPROCESSING; SIMULATION; ...

  5. Swing-Down of 21-PWR Waste Package

    SciTech Connect (OSTI)

    A.K. Scheider

    2001-05-04

    The objective of this calculation is to determine the structural response of the waste package (WP) swinging down from a horizontally suspended height. The WP used for that purpose is the 21-Pressurized Water Reactor (PWR) WP. The scope of this document is limited to reporting the calculation results in terms of stress intensities. This calculation is associated with the WP design and was performed by the Waste Package Design group in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 13). AP-3.12Q, ''Calculations'' (Ref. 18) is used to perform the calculation and develop the document. The information provided by the sketches attached to this calculation is that of the potential design of the type of 21-PWR WP design considered in this calculation and provides the potential dimensions and materials for the 21-PWR WP design.

  6. Integrated rural energy planning

    SciTech Connect (OSTI)

    El Mahgary, Y.; Biswas, A.K.

    1985-01-01

    This book presents papers on integrated community energy systems in developing countries. Topics considered include an integrated rural energy system in Sri Lanka, rural energy systems in Indonesia, integrated rural food-energy systems and technology diffusion in India, bringing energy to the rural sector in the Philippines, the development of a new energy village in China, the Niaga Wolof experimental rural energy center, designing a model rural energy system for Nigeria, the Basaisa village integrated field project, a rural energy project in Tanzania, rural energy development in Columbia, and guidelines for the planning, development and operation of integrated rural energy projects.

  7. Leak before break application in French PWR plants under operation

    SciTech Connect (OSTI)

    Faidy, C.

    1997-04-01

    Practical applications of the leak-before break concept are presently limited in French Pressurized Water Reactors (PWR) compared to Fast Breeder Reactors. Neithertheless, different fracture mechanic demonstrations have been done on different primary, auxiliary and secondary PWR piping systems based on similar requirements that the American NUREG 1061 specifications. The consequences of the success in different demonstrations are still in discussion to be included in the global safety assessment of the plants, such as the consequences on in-service inspections, leak detection systems, support optimization,.... A large research and development program, realized in different co-operative agreements, completes the general approach.

  8. Analysis of PWR RCS Injection Strategy During Severe Accident

    SciTech Connect (OSTI)

    Wang, S.-J. [Institute of Nuclear Energy Research, Taiwan (China); Chiang, K.-S. [Institute of Nuclear Energy Research, Taiwan (China); Chiang, S.-C. [Taiwan Power Company, Taiwan (China)

    2004-05-15

    Reactor coolant system (RCS) injection is an important strategy for severe accident management of a pressurized water reactor (PWR) system. Maanshan is a typical Westinghouse PWR nuclear power plant (NPP) with large, dry containment. The severe accident management guideline (SAMG) of Maanshan NPP is developed based on the Westinghouse Owners Group (WOG) SAMG.The purpose of this work is to analyze the RCS injection strategy of PWR system in an overheated core condition. Power is assumed recovered as the vessel water level drops to the bottom of active fuel. The Modular Accident Analysis Program version 4.0.4 (MAAP4) code is chosen as a tool for analysis. A postulated station blackout sequence for Maanshan NPP is cited as a reference case for this analysis. The hot leg creep rupture occurs during the mitigation action with immediate injection after power recovery according to WOG SAMG, which is not desired. This phenomenon is not considered while developing the WOG SAMG. Two other RCS injection methods are analyzed by using MAAP4. The RCS injection strategy is modified in the Maanshan SAMG. These results can be applied for typical PWR NPPs.

  9. WINDExchange: Rural Communities

    Wind Powering America (EERE)

    Rural Communities Printable Version Bookmark and Share Wind for Homeowners, Farmers, & Businesses Resources & Tools Rural Communities Agricultural lands in the United States are...

  10. Robotic inspection of PWR coolant pump casing welds

    SciTech Connect (OSTI)

    Pratt, W.R.; Alford, J.W.; Davis, J.B.

    1997-12-01

    As of January 1, 1995, the Swedish Nuclear Inspectorate began requiring more thorough inspections of cast stainless-steel components in nuclear power plants, including pressurized water reactor (PWR) reactor coolant pump (RCP) casings. The examination requirements are established by fracture mechanics analyses of component weldments and demonstrated test system detection capabilities. This may include full volumetric inspection or some portion thereof. Ringhals station is a four-unit nuclear power plant, owned and operated by the Swedish State Power Board, Vattenfall. Unit 1 is a boiling water reactor. Units 2, 3, and 4 are Westinghouse-designed PWRs, ranging in size from 795 to 925 MW. The RCP casings at the PWR units are made of cast stainless steel and contain four circumferential welds that require inspection. Due to the thickness of the casings at the weld locations and configuration and surface conditions on the outside diameter of the casings, remote inspection from the inside diameter of the pump casing was mandated.

  11. Design study of long-life PWR using thorium cycle

    SciTech Connect (OSTI)

    Subkhi, Moh. Nurul; Su'ud, Zaki; Waris, Abdul

    2012-06-06

    Design study of long-life Pressurized Water Reactor (PWR) using thorium cycle has been performed. Thorium cycle in general has higher conversion ratio in the thermal spectrum domain than uranium cycle. Cell calculation, Burn-up and multigroup diffusion calculation was performed by PIJ-CITATION-SRAC code using libraries based on JENDL 3.2. The neutronic analysis result of infinite cell calculation shows that {sup 231}Pa better than {sup 237}Np as burnable poisons in thorium fuel system. Thorium oxide system with 8%{sup 233}U enrichment and 7.6{approx} 8%{sup 231}Pa is the most suitable fuel for small-long life PWR core because it gives reactivity swing less than 1%{Delta}k/k and longer burn up period (more than 20 year). By using this result, small long-life PWR core can be designed for long time operation with reduced excess reactivity as low as 0.53%{Delta}k/k and reduced power peaking during its operation.

  12. Chemical behavior of fission products in the ORNL fission product release program. Supplement. [PWR; BWR

    SciTech Connect (OSTI)

    Collins, J.L.; Osborne, M.F.; Lorenz, R.A.

    1983-01-01

    Tests data are presented for BWR and PWR rods in test HI-4 and test HI-5. Operating conditions fission product release data are included.

  13. ACHILLES: Heat Transfer in PWR Core During LOCA Reflood Phase

    Energy Science and Technology Software Center (OSTI)

    2013-11-01

    1. NAME AND TITLE OF DATA LIBRARY ACHILLES -Heat Transfer in PWR Core During LOCA Reflood Phase. 2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS N/A 3. CONTRIBUTOR AEA Technology, Winfrith Technology Centre, Dorchester DT2 8DH United Kingdom through the OECD Nuclear Energy Agency Data Bank, Issy-les-Moulineaux, France. 4. DESCRIPTION OF TEST FACILITY The most important features of the Achilles rig were the shroud vessel, which contained the test section, and the downcomer. These maymore » be thought of as representing the core barrel and the annular downcomer in the reactor pressure vessel. The test section comprises a cluster of 69 rods in a square array within a circular shroud vessel. The rod diameter and pitch (9.5 mm and 12.6 mm) were typical of PWR dimensions. The internal diameter of the shroud vessel was 128 mm. Each rod was electrically heated over a length of 3.66 m, which is typical of the nuclear heated length in a PWR fuel rod, and each contained 6 internal thermocouples. These were arranged in one of 8 groupings which concentrated the thermocouples in different axial zones. The spacer grids were at prototypic PWR locations. Each grid had two thermocouples attached to its trailing edge at radial locations. The axial power profile along the rods was an 11 step approximation to a "chopped cosine". The shroud vessel had 5 heating zones whose power could be independently controlled. 5. DESCRIPTION OF TESTS The Achilles experiments investigated the heat transfer in the core of a Pressurized Water Reactor during the re-flood phase of a postulated large break loss of coolant accident. The results provided data to validate codes and to improve modeling. Different types of experiments were carried out which included single phase cooling, re-flood under low flow conditions, level swell and re-flood under high flow conditions. Three series of experiments were performed. The first and the third used the same test section but the second used another test section

  14. ACHILLES: Heat Transfer in PWR Core During LOCA Reflood Phase

    SciTech Connect (OSTI)

    2013-11-01

    1. NAME AND TITLE OF DATA LIBRARY ACHILLES -Heat Transfer in PWR Core During LOCA Reflood Phase. 2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS N/A 3. CONTRIBUTOR AEA Technology, Winfrith Technology Centre, Dorchester DT2 8DH United Kingdom through the OECD Nuclear Energy Agency Data Bank, Issy-les-Moulineaux, France. 4. DESCRIPTION OF TEST FACILITY The most important features of the Achilles rig were the shroud vessel, which contained the test section, and the downcomer. These may be thought of as representing the core barrel and the annular downcomer in the reactor pressure vessel. The test section comprises a cluster of 69 rods in a square array within a circular shroud vessel. The rod diameter and pitch (9.5 mm and 12.6 mm) were typical of PWR dimensions. The internal diameter of the shroud vessel was 128 mm. Each rod was electrically heated over a length of 3.66 m, which is typical of the nuclear heated length in a PWR fuel rod, and each contained 6 internal thermocouples. These were arranged in one of 8 groupings which concentrated the thermocouples in different axial zones. The spacer grids were at prototypic PWR locations. Each grid had two thermocouples attached to its trailing edge at radial locations. The axial power profile along the rods was an 11 step approximation to a "chopped cosine". The shroud vessel had 5 heating zones whose power could be independently controlled. 5. DESCRIPTION OF TESTS The Achilles experiments investigated the heat transfer in the core of a Pressurized Water Reactor during the re-flood phase of a postulated large break loss of coolant accident. The results provided data to validate codes and to improve modeling. Different types of experiments were carried out which included single phase cooling, re-flood under low flow conditions, level swell and re-flood under high flow conditions. Three series of experiments were performed. The first and the third used the same test section but the second used another test section, similar in

  15. 21-PWR WASTE PACKAGE WITH ABSORBER PLATES LOADING CURVE EVALUATION

    SciTech Connect (OSTI)

    J.M. Scaglione

    2004-12-17

    The objective of this calculation is to evaluate the required minimum burnup as a function of initial pressurized water reactor (PWR) assembly enrichment that would permit loading of spent nuclear fuel into the 21 PWR waste package with absorber plates design as provided in Attachment IV. This calculation is an example of the application of the methodology presented in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003). The scope of this calculation covers a range of enrichments from 0 through 5.0 weight percent U-235, and a burnup range of 0 through 45 GWd/MTU. Higher burnups were not necessary because 45 GWd/MTU was high enough for the loading curve determination. This activity supports the validation of the use of burnup credit for commercial spent nuclear fuel applications. The intended use of these results will be in establishing PWR waste package configuration loading specifications. Limitations of this evaluation are as follows: (1) The results are based on burnup credit for actinides and selected fission products as proposed in YMP (2003, Table 3-1) and referred to as the ''Principal Isotopes''. Any change to the isotope listing will have a direct impact on the results of this report. (2) The results are based on 1.5 wt% Gd in the Ni-Gd Alloy material and having no tuff inside the waste package. If the Gd loading is reduced or a process to introduce tuff inside the waste package is defined, then this report would need to be reevaluated based on the alternative materials. This calculation is subject to the ''Quality Assurance Requirements and Description'' (QARD) (DOE 2004) because it concerns engineered barriers that are included in the ''Q-List'' (BSC 2004k, Appendix A) as items important to safety and waste isolation.

  16. Comparison of PWR-IMF and FR fuel cycles

    SciTech Connect (OSTI)

    Darilek, Petr; Zajac, Radoslav; Breza, Juraj |; Necas, Vladimir

    2007-07-01

    The paper gives a comparison of PWR (Russia origin VVER-440) cycle with improved micro-heterogeneous inert matrix fuel assemblies and FR cycle. Micro-heterogeneous combined assembly contains transmutation pins with Pu and MAs from burned uranium reprocessing and standard uranium pins. Cycle analyses were performed by HELIOS spectral code and SCALE code system. Comparison is based on fuel cycle indicators, used in the project RED-IMPACT - part of EU FP6. Advantages of both closed cycles are pointed out. (authors)

  17. CASL - PWR Reactor Vessel Multi-Physics CFD Model

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

    PWR Reactor Vessel Multi-Physics CFD Model Jin Yan*1, Yiban Xu1, Andrew Petrarca1, Zeses Karoutas1, Emre Tatli1, Emilio Baglietto2, Jess Gehin3 1Westinghouse Electric Company LLC 2Massachusetts Institute of Technology 3Oak Ridge National Lab *Correspondence to: yan3j@westinghouse.com A complete 3D SolidWorks CAD model of Watts Bar Unit 1 was constructed based on drawings. A single fuel assembly CAD model including all geometrical details was created based on the Westinghouse V5H 17x17 fuel

  18. MicroBooNE Detector Stability MICROBOONE-NOTE-1013-PUB The MicroBooNE

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

    MicroBooNE Detector Stability MICROBOONE-NOTE-1013-PUB The MicroBooNE Collaboration June 30, 2016 Abstract The Micro Booster Neutrino Experiment (MicroBooNE) is designed to explore the low- energy excess in the ν e event spectrum reported by the MiniBooNE experiment [1] and to measure ν-Ar cross sections in the 1 GeV energy range. The detector is a liquid argon time projection chamber with wire readout, supplemented with a light detection system based on photo-multiplier tubes (PMTs). The

  19. 21-PWR Waste Package Side and End Impacts

    SciTech Connect (OSTI)

    V. Delabrosse

    2003-02-27

    The objective of this calculation is to determine the structural response of a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities and initial angles between the waste package and the unyielding surface is studied. The scope of this calculation is limited to estimating the area of the outer shell (OS) where the residual stress exceeds a given limit (hereafter ''damaged area''). The stress limit is defined as a fraction of the yield strength of the OS material, Alloy 22 (SB-575 N06022), at the appropriate temperature. The design of the 21-PWR waste package used in this calculation is that defined in Reference 8. However, a value of 4 mm was used for the gap between the inner shell and the OS, and the thickness of the OS was reduced by 2 mm. The sketch in Attachment I provides additional information not included in Reference 8. All obtained results are valid for this design only. This calculation is associated with the waste package design and was performed by the Specialty Analyses and Waste Package Design Section. The waste package (i.e. uncanistered spent nuclear fuel disposal container) is classified as Quality Level 1.

  20. 21-PWR Waste Package Side and End Impacts

    SciTech Connect (OSTI)

    T. Schmitt

    2005-08-29

    The objective of this calculation is to determine the structural response of a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities and initial angles between the waste package and the unyielding surface is studied. The scope of this calculation is limited to estimating the area of the outer shell (OS) where the residual stress exceeds a given limit (hereafter ''damaged area''). The stress limit is defined as a fraction of the yield strength of the OS material, Alloy 22 (SB-575 N06022), at the appropriate temperature. The design of the 21-PWR waste package used in this calculation is that defined in Reference 8. However, a value of 4 mm was used for the gap between the inner shell and the OS, and the thickness of the OS was reduced by 2 mm. The sketch in Attachment I provides additional information not included in Reference 8. All obtained results are valid for this design only. This calculation is associated with the waste package design and was performed by the Specialty Analyses and Waste Package Design Section. The waste package (i.e. uncanistered spent nuclear fuel disposal container) is classified as Quality Level 1.

  1. Waterside corrosion of Zircaloy fuel rods. Final report. [PWR

    SciTech Connect (OSTI)

    Garzarolli, F.; Jung, W.; Schoenfeld, H.; Garde, A.M.; Parry, G.W.; Smerd, P.G.

    1982-12-01

    There is an economic incentive to extend average fuel-rod-discharge burnup to about 50 GWd/t. For these higher burnups it is necessary to know if increased waterside corrosion of the cladding will influence fuel-rod performance. For this reason, EPRI sponsored a joint program with C-E and KWU with the objective of investigating PWR waterside corrosion. This final report presents and discusses the results of various subtasks that comprised this project. In the review of corrosion data and models in the literature it was concluded that the PWR environment enhances the corrosion rate by about three times that expected from ex-reactor tests. A large number of fuel rods were characterized in both spent-fuel-pool and hot-cell campaigns. Chemical, physical and microstructural attributes of irradiated and unirradiated oxide films were measured. These included determinations of chemical composition, crystal structure, microstructure, density, specific heat, thermal conductivity, and post-irradiation autoclave corrosion behavior. Procedures used to calculate the fuel-rod surface temperature were reviewed. A model has been developed to predict in-reactor corrosion behavior.

  2. Westinghouse VANTAGE+ fuel assembly to meet future PWR operating requirements

    SciTech Connect (OSTI)

    Doshi, P.K.; Chapin, D.L.; Scherpereel, L.R.

    1988-01-01

    Many utilities operating pressurized water reactors (PWRs) are implementing longer reload cycles. Westinghouse is addressing this trend with fuel products that increase fuel utilization through higher discharge burnups. Higher burnup helps to offset added enriched uranium costs necessary to enable the higher energy output of longer cycles. Current fuel products have burnup capabilities in the area of 40,000 MWd/tonne U or more. There are three main phenomena that must be addressed to achieve even higher burnup levels: accelerated cladding, waterside corrosion, and hydriding; increased fission gas production; and fuel rod growth. Long cycle lengths also require efficient burnable absorbers to control the excess reactivity associated with increased fuel enrichment while maintaining a low residual absorber penalty at the end of cycle. Westinghouse VANTAGE + PWR fuel incorporates features intended to enhance fuel performance at very high burnups, including advances in the three basic elements of the fuel assembly: fuel cladding, fuel rod, and fuel assembly skeleton. ZIRLO {sup TM} cladding, an advanced Zircaloy cladding that contains niobium, offers a significant improvement in corrosion resistance relative to Zircaloy-4. Another important Westinghouse PWR fuel feature that facilitates long cycles is the zirconium diboride integral fuel burnable absorber (ZrB{sub 2}IFBA).

  3. PWR loss of feedwater ATWS: analysis and sensitivity study

    SciTech Connect (OSTI)

    Shier, W.G.; Lu, M.S.; Levine, M.M.; Diamond, D.J.

    1983-01-01

    The incident at the Salem Nuclear plant has presented a renewed interest in the analysis of the consequences of anticipated transients without scram (ATWS). This paper presents the results of an analysis of a complete loss of feedwater ATWS for a typical 4-loop PWR. The loss of feedwater transient was selected since previous analyses have shown that this transient produces one of the more limiting overpressure conditions in the primary system. These results provide a detailed analysis of this transient using current analytical techniques and show the sensitivity to several important parameters and plant modeling techniques. The RELAP5/MOD1 computer code has been used for this analysis. The code version is designated as Cycle 13 with additional modifications provided by both INEL and BNL.

  4. Fracture mechanics evaluation for at typical PWR primary coolant pipe

    SciTech Connect (OSTI)

    Tanaka, T.; Shimizu, S.; Ogata, Y.

    1997-04-01

    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years.

  5. Study of a Station Blackout Event in the PWR Plant

    SciTech Connect (OSTI)

    Ching-Hui Wu; Tsu-Jen Lin; Tsu-Mu Kao [Institute of Nuclear Energy Research P.O. Box 3-3, Longtan, 32500, Taiwan (China)

    2002-07-01

    On March 18, 2001, a PWR nuclear power plant located in the Southern Taiwan occurred a Station Blackout (SBO) event. Monsoon seawater mist caused the instability of offsite power grids. High salt-contained mist caused offsite power supply to the nuclear power plant very unstable, and forced the plant to be shutdown. Around 24 hours later, when both units in the plant were shutdown, several inadequate high cycles of bus transfer between 345 kV and 161 kV startup transformers degraded the emergency 4.16 kV switchgears. Then, in the Train-A switchgear room of Unit 1 occurred a fire explosion, when the degraded switchgear was hot shorted at the in-coming 345 kV breaker. Inadequate configuration arrangement of the offsite power supply to the emergency 4.16 kV switchgears led to loss of offsite power (LOOP) events to both units in the plant. Both emergency diesel generators (EDG) of Unit 1 could not be in service in time, but those of Unit 2 were running well. The SBO event of Unit 1 lasted for about two hours till the fifth EDG (DG-5) was lined-up to the Train-B switchgear. This study investigated the scenario of the SBO event and evaluated a risk profile for the SBO period. Guidelines in the SBO event, suggested by probabilistic risk assessment (PRA) procedures were also reviewed. Many related topics such as the re-configuration of offsite power supply, the addition of isolation breakers of the emergency 4.16 kV switchgears, the betterment of DG-5 lineup design, and enhancement of the reliability of offsite power supply to the PWR plant, etc., will be in further studies. (authors)

  6. PWR and BWR spent fuel assembly gamma spectra measurements

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

    Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea; Grogan, Brandon R.; Jansson, Peter; Liljenfeldt, Henrik; Mozin, Vladimir; Hu, Jianwei; Schwalbach, P.; Sjoland, A.; et al

    2016-07-17

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

  7. Design study of long-life PWR using thorium cycle (Journal Article...

    Office of Scientific and Technical Information (OSTI)

    life PWR core because it gives reactivity swing less than 1%Deltakk and longer ... long time operation with reduced excess reactivity as low as 0.53%Deltakk and reduced ...

  8. Conceptual design study of small long-life PWR based on thorium...

    Office of Scientific and Technical Information (OSTI)

    The optimization of 350 MWt small long life PWR result small excess reactivity and reduced ... on advances in nuclear science and engineering, Denpasar, Bali (Indonesia), 16-19 Sep ...

  9. Identification and evaluation of PWR in-vessel severe accident management strategies

    SciTech Connect (OSTI)

    Dukelow, J S [Pacific Northwest Lab., Richland, WA (United States); Harrison, D G [Jason Associates, Idaho Falls, ID (United States); Morgenstern, M [Battelle Human Affairs Research Center, Seattle, WA (United States)

    1992-03-01

    This reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core degradation and prior to RPV failure. A parallel project at BNL has been considering strategies applicable to the ex-vessel phase of PWR severe accidents.

  10. Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario

    SciTech Connect (OSTI)

    Rempe, J. L.; Knudson, D. L.; Lutz, R. J.

    2015-09-01

    The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 nuclear power plants demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data. While progress in these areas has been made since TMI-2, the events at Fukushima suggests that there may still be a potential need to ensure that critical plant information is available to plant operators. Recognizing the significant technical and economic challenges associated with plant modifications, it is important to focus on instrumentation that can address these information critical needs. As part of a program initiated by the Department of Energy, Office of Nuclear Energy (DOE-NE), a scoping effort was initiated to assess critical information needs identified for severe accident management and mitigation in commercial Light Water Reactors (LWRs), to quantify the environment instruments monitoring this data would have to survive, and to identify gaps where predicted environments exceed instrumentation qualification envelop (QE) limits. Results from the Pressurized Water Reactor (PWR) scoping evaluations are documented in this report. The PWR evaluations were limited in this scoping evaluation to quantifying the environmental conditions for an unmitigated Short-Term Station BlackOut (STSBO) sequence in one unit at the Surry nuclear power station. Results were obtained using the MELCOR models developed for the US Nuclear Regulatory Commission (NRC)-sponsored State of the Art Consequence Assessment (SOARCA) program project. Results from this scoping evaluation indicate that some instrumentation identified to provide critical information would be exposed to conditions that

  11. USDA Rural Development Energy Program

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

    Development Energy Programs Tedd Buelow Native American Coordinator DOE Tribal Energy Program Review November 16, 2009 USDA Rural Development 4 Types of Program Delivery - Guaranteed Loans - Direct Loans - Grants - Payments 3 Program Areas - Rural Business and Cooperative - Rural Housing and Community Facilities - Rural Utilities DOE Tribal Energy Program Review November 16, 2009 USDA Rural Development Organizational Structure National Office State Directors Area Directors Program Directors

  12. Alaska Rural Energy Conference

    Broader source: Energy.gov [DOE]

    The Rural Energy Conference is a three-day event offering a large variety of technical sessions covering new and ongoing energy projects in Alaska, as well as new technologies and needs for Alaska's remote communities.

  13. Rural Energy Savings Program

    Office of Energy Efficiency and Renewable Energy (EERE)

    This program helps rural families and small businesses achieve cost savings by providing loans to qualified consumers to implement durable, cost-effective energy efficiency measures, including on- or off-grid renewable energy.

  14. Analysis of Potential Hydrogen Risk in the PWR Containment

    SciTech Connect (OSTI)

    Deng Jian; Xuewu Cao [Shanghai Jiaotong University, Shanghai (China)

    2006-07-01

    Various studies have shown that hydrogen combustion is one of major risk contributors to threaten the integrity of the containment in a nuclear power plant. That hydrogen risk should be considered in severe accident strategies in current and future NPPs has been emphasized in the latest policies issued by the National Nuclear Safety Administration of China (NNSA). According to a deterministic approach, three typical severe accident sequences for a PWR large dry containment, such as the large break loss-of-coolant (LLOCA), the station blackout (SBO), and the small break loss-of-coolant (SLOCA) are analyzed in this paper with MELCOR code. Hydrogen concentrations in different compartments are observed to evaluate the potential hydrogen risk. The results show that there is a great amount of hydrogen released into the containment, which causes the containment pressure to increase and some potential in-consecutive burning. Therefore, certain hydrogen management strategies should be considered to reduce the risk to threaten the containment integrity. (authors)

  15. Assessment of PWR waterside corrosion models and data. Final report

    SciTech Connect (OSTI)

    Cox, B.

    1985-10-01

    The published data on waterside corrosion of PWR fuel cladding and unfuelled components have been reviewed, and the models used to assess the data have been studied. All corrosion models use too simplified a view of the corrosion process to obtain other than a general trend for the actual oxidation data. The in-reactor post-transition oxidation of the Zircaloys appears to be heavily dependent on water chemistry variations both between reactors, and along the length of an individual fuel rod. Crud deposition may be one primary cause of this, perhaps by allowing the independent development of the water chemistry within the crud layer, as much as by its effect on cladding surface temperatures. However, the effect of the thickening of the oxide film, which permits the development of an independent water chemistry inside the oxide, leading to an accelerating oxidation rate at large oxide thicknesses, seems to be the most important factor. It is concluded that a spectrum of results ranging from essentially no in-reactor enhancement of the oxidation rate to a sizeable enhancement (>10) may be seen depending upon the thickness of the oxide films, the water chemistry of the reactor, and crud deposition. A post-irradiation test that may help to distinguish between the factors involved has been suggested. 105 refs., 38 figs.

  16. Coupled Neutronics Thermal-Hydraulic Solution of a Full-Core PWR Using VERA-CS

    SciTech Connect (OSTI)

    Clarno, Kevin T; Palmtag, Scott; Davidson, Gregory G; Salko, Robert K; Evans, Thomas M; Turner, John A; Belcourt, Kenneth; Hooper, Russell; Schmidt, Rodney

    2014-01-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing a core simulator called VERA-CS to model operating PWR reactors with high resolution. This paper describes how the development of VERA-CS is being driven by a set of progression benchmark problems that specify the delivery of useful capability in discrete steps. As part of this development, this paper will describe the current capability of VERA-CS to perform a multiphysics simulation of an operating PWR at Hot Full Power (HFP) conditions using a set of existing computer codes coupled together in a novel method. Results for several single-assembly cases are shown that demonstrate coupling for different boron concentrations and power levels. Finally, high-resolution results are shown for a full-core PWR reactor modeled in quarter-symmetry.

  17. Assessment of PWR Steam Generator modelling in RELAP5/MOD2. International Agreement Report

    SciTech Connect (OSTI)

    Putney, J.M.; Preece, R.J.

    1993-06-01

    An assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD2 is presented. The assessment is based on a review of code assessment calculations performed in the UK and elsewhere, detailed calculations against a series of commissioning tests carried out on the Wolf Creek PWR and analytical investigations of the phenomena involved in normal and abnormal SG operation. A number of modelling deficiencies are identified and their implications for PWR safety analysis are discussed -- including methods for compensating for the deficiencies through changes to the input deck. Consideration is also given as to whether the deficiencies will still be present in the successor code RELAP5/MOD3.

  18. Study on Equilibrium Characteristics of Thorium-Plutonium-Minor Actinides Mixed Oxides Fuel in PWR

    SciTech Connect (OSTI)

    Waris, A.; Permana, S.; Kurniadi, R.; Su'ud, Z.; Sekimoto, H.

    2010-06-22

    A study on characteristics of thorium-plutonium-minor actinides utilization in the pressurized water reactor (PWR) with the equilibrium burnup model has been conducted. For a comprehensive evaluation, several fuel cycles scenario have been included in the present study with the variation of moderator-to-fuel volume ratio (MFR) of PWR core design. The results obviously exhibit that the neutron spectra grow to be harder with decreasing of the MFR. Moreover, the neutron spectra also turn into harder with the rising number of confined heavy nuclides. The required {sup 233}U concentration for criticality of reactor augments with the increasing of MFR for all heavy nuclides confinement and thorium and uranium confinement in PWR.

  19. Alaska Rural Small Business Conference

    Broader source: Energy.gov [DOE]

    Hosted by the Alaska Village Initiatives, the Alaska Rural Small Business Conference is a three-day conference to bring together rural businesses and leaders and provide them with networking opportunities, training, and technical information.

  20. In-core and ex-core calculations of the VENUS simulated PWR benchmark experiment

    SciTech Connect (OSTI)

    Williams, M.L.; Chowdhury, P.; Landesman, M.; Kam, F.B.K.

    1985-01-01

    The VENUS PWR engineering mockup experiment was established to simulate a beginning-of-life, generic PWR configuration at the zero-power VENUS critical facility located at CEN/SCK, Mol, Belgium. The experimental measurement program consists of (1) gamma scans to determine the core power distribution, (2) in-core and ex-core foil activations, (3) neutron spectrometer measurements, and (4) gamma heating measurements with TLD's. Analysis of the VENUS benchmark has been performed with two-dimensional discrete ordinates transport theory, using the DOT-IV code.

  1. Rural energy and development

    SciTech Connect (OSTI)

    Stern, R.

    1997-12-01

    The author discusses the worldwide problem and need for rural electrification to support development. He points out that rural areas will pay high rates to receive such services, but cannot afford the capital cost for conventional services. The author looks at this problem from the point of energy choices, subsides, initial costs, financing, investors, local involvement, and governmental actions. In particular he is concerned with ways to make better use of biofuels, to promote sustainable harvesting, and to encourage development of more modern fuels.

  2. Radionuclide release from PWR spent fuel specimens with induced cladding defects

    SciTech Connect (OSTI)

    Wilson, C.N.; Oversby, V.M.

    1984-03-01

    Radionuclide releases from pressurized water reactor (PWR) spent fuel rod specimens containing various artificially induced cladding defects were compared by leach testing. The study was conducted in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package Task to evaluate the effectiveness of failed cladding as a barrier to radionuclide release. Test description and results are presented. 6 references, 4 figures.

  3. Radionuclide release from PWR spent fuel specimens with induced cladding defects

    SciTech Connect (OSTI)

    Wilson, C.N.; Oversby, V.M.

    1984-03-01

    Radionuclide releases from pressurized water reactor (PWR) spent fuel rod specimens containing various artificially induced cladding defects were compared by leach testing. The study was conducted in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package Task to evaluate the effectiveness of failed cladding as a barrier to radionuclide release. Test description and results are presented.

  4. Performance evaluation of two-stage fuel cycle from SFR to PWR

    SciTech Connect (OSTI)

    Fei, T.; Hoffman, E.A.; Kim, T.K.; Taiwo, T.A.

    2013-07-01

    One potential fuel cycle option being considered is a two-stage fuel cycle system involving the continuous recycle of transuranics in a fast reactor and the use of bred plutonium in a thermal reactor. The first stage is a Sodium-cooled Fast Reactor (SFR) fuel cycle with metallic U-TRU-Zr fuel. The SFRs need to have a breeding ratio greater than 1.0 in order to produce fissile material for use in the second stage. The second stage is a PWR fuel cycle with uranium and plutonium mixed oxide fuel based on the design and performance of the current state-of-the-art commercial PWRs with an average discharge burnup of 50 MWd/kgHM. This paper evaluates the possibility of this fuel cycle option and discusses its fuel cycle performance characteristics. The study focuses on an equilibrium stage of the fuel cycle. Results indicate that, in order to avoid a positive coolant void reactivity feedback in the stage-2 PWR, the reactor requires high quality of plutonium from the first stage and minor actinides in the discharge fuel of the PWR needs to be separated and sent back to the stage-1 SFR. The electricity-sharing ratio between the 2 stages is 87.0% (SFR) to 13.0% (PWR) for a TRU inventory ratio (the mass of TRU in the discharge fuel divided by the mass of TRU in the fresh fuel) of 1.06. A sensitivity study indicated that by increasing the TRU inventory ratio to 1.13, The electricity generation fraction of stage-2 PWR is increased to 28.9%. The two-stage fuel cycle system considered in this study was found to provide a high uranium utilization (>80%). (authors)

  5. Z:\Professor Perl\Tau Discovery\Floppy THREE\SLAC-PUB-10150.prn.pdf

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

    SLAC-PUB-10150 October 2003 Submitted to Physics in Perspective *Work supported by Department of Energy contract DE-AC03-76SF00515. Tau Discovery THE DISCOVERY OF THE TAU LEPTON AND THE CHANGES IN ELEMENTARY PARTICLE PHYSICS IN 40 YEARS Martin L. Perl Stanford Linear Accelerator Center and Stanford University, Stanford, CA 94309 Phone: 650-926-4286 Fax: 650-926-4001 Email: martin@slac.stanford.edu Introduction This is a history of my discovery of the tau lepton in the 1970s for which I was

  6. Rural health clinics infrastructure

    SciTech Connect (OSTI)

    Olson, K.

    1997-12-01

    The author discusses programs which were directed at the installation of photovoltaic power systems in rural health clinics. The objectives included: vaccine refrigeration; ice pack freezing; lighting; communications; medical appliances; sterilization; water purification; and income generation. The paper discusses two case histories, one in the Dominican Republic and one in Colombia. The author summarizes the results of the programs, both successes and failures, and offers an array of conclusions with regard to the implementation of future programs of this general nature.

  7. Rural Energy Conference Project

    SciTech Connect (OSTI)

    Dennis Witmer; Shannon Watson

    2008-12-31

    Alaska remains, even at the beginning of the 21st century, a place with many widely scattered, small, remote communities, well beyond the end of both the road system and the power grid. These communities have the highest energy costs of any place in the United States, despite the best efforts of the utilities that service them. This is due to the widespread dependence on diesel electric generators, which require small capital investments, but recent increases in crude oil prices have resulted in dramatic increases in the cost of power. In the enabling legislation for the Arctic Energy Office in 2001, specific inclusion was made for the study of ways of reducing the cost of electrical power in these remote communities. As part of this mandate, the University of Alaska has, in conjunction with the US Department of Energy, the Denali Commission and the Alaska Energy Authority, organized a series of rural energy conferences, held approximately every 18 months. The goal of these meeting was to bring together rural utility operators, rural community leaders, government agency representatives, equipment suppliers, and researchers from universities and national laboratories to discuss the current state of the art in rural power generation, to discuss current projects, including successes as well as near successes. Many of the conference presenters were from industry and not accustomed to writing technical papers, so the typical method of organizing a conference by requesting abstracts and publishing proceedings was not considered viable. Instead, the organizing committee solicited presentations from appropriate individuals, and requested that (if they were comfortable with computers) prepare Power point presentations that were collected and posted on the web. This has become a repository of many presentations, and may be the best single source of information about current projects in the state of Alaska.

  8. Proceedings: 1983 Workshop on Secondary-Side Stress Corrosion Cracking and Intergranular Corrosion of PWR Steam Generator Tubing

    SciTech Connect (OSTI)

    1986-03-01

    Participants in this international workshop discussed research investigating mechanisms and propagation rates of intergranular corrosion in PWR steam generators. Laboratory test results, which have been consistent with power plant experience, permitted preliminary definition of corrosion rates in alloy 600 tubing.

  9. DOSE RATES FOR WESTINGHOUSE 17X17 MOX PWR SNF IN A WASTE PACKAGE (SCPB: N/A)

    SciTech Connect (OSTI)

    T.L. Lotz

    1997-01-29

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to estimate the dose rate on and near the surface a Multi-Purpose Canister (MPC) PWR waste package (WP) which is loaded with Westinghouse 17 x 17 mixed oxide (MOX) PWR fuel. The 21 PWR MPC WP is used to provide an upper bound for waste package designs since the 12 PWR MPC WP will have a smaller source term and an equivalent amount of shielding. the objectives of this evaluation are to calculate the requested dose rate(s) and document the calculation in a fashion to allow comparisons to other waste forms and WP designs at a future time.

  10. ARM - SGP Rural Driving Hazards

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

    Rural Driving Hazards SGP Related Links Virtual Tour Facilities and Instruments Central Facility Boundary Facility Extended Facility Intermediate Facility Radiometric Calibration Facility Geographic Information ES&H Guidance Statement Operations Science Field Campaigns Visiting the Site Summer Training SGP Fact Sheet Images Information for Guest Scientists Contacts SGP Rural Driving Hazards The rural location of the Southern Great Plains (SGP) site facilities requires that visitors travel on

  11. Rural Utilities Service Electric Program

    Broader source: Energy.gov [DOE]

    The Rural Utilities Service Electric Program’s loans and loan guarantees finance the construction of electric distribution, transmission, and generation facilities, including system improvements...

  12. Conceptual design study of small long-life PWR based on thorium cycle fuel

    SciTech Connect (OSTI)

    Subkhi, M. Nurul; Su'ud, Zaki; Waris, Abdul; Permana, Sidik

    2014-09-30

    A neutronic performance of small long-life Pressurized Water Reactor (PWR) using thorium cycle based fuel has been investigated. Thorium cycle which has higher conversion ratio in thermal region compared to uranium cycle produce some significant of {sup 233}U during burn up time. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.3, while the multi-energy-group diffusion calculations were optimized in whole core cylindrical two-dimension R-Z geometry by SRAC-CITATION. this study would be introduced thorium nitride fuel system which ZIRLO is the cladding material. The optimization of 350 MWt small long life PWR result small excess reactivity and reduced power peaking during its operation.

  13. Optimization of small long-life PWR based on thorium fuel

    SciTech Connect (OSTI)

    Subkhi, Moh Nurul; Suud, Zaki Waris, Abdul; Permana, Sidik

    2015-09-30

    A conceptual design of small long-life Pressurized Water Reactor (PWR) using thorium fuel has been investigated in neutronic aspect. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.2, while the multi-energy-group diffusion calculations were optimized in three-dimension X-Y-Z geometry of core by COREBN. The excess reactivity of thorium nitride with ZIRLO cladding is considered during 5 years of burnup without refueling. Optimization of 350 MWe long life PWR based on 5% {sup 233}U & 2.8% {sup 231}Pa, 6% {sup 233}U & 2.8% {sup 231}Pa and 7% {sup 233}U & 6% {sup 231}Pa give low excess reactivity.

  14. Preliminary assessment of PWR Steam Generator modelling in RELAP5/MOD3. International Agreeement Report

    SciTech Connect (OSTI)

    Preece, R.J.; Putney, J.M.

    1993-07-01

    A preliminary assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD3 is presented. The study is based on calculations against a series of steady-state commissioning tests carried out on the Wolf Creek PWR over a range of load conditions. Data from the tests are used to assess the modelling of primary to secondary side heat transfer and, in particular, to examine the effect of reverting to the standard form of the Chen heat transfer correlation in place of the modified form applied in RELAP5/MOD2. Comparisons between the two versions of the code are also used to show how the new interphase drag model in RELAP5/MOD3 affects the calculation of SG liquid inventory and the void fraction profile in the riser.

  15. WINDExchange: Agricultural and Rural Resources and Tools

    Wind Powering America (EERE)

    Rural Communities Printable Version Bookmark and Share Wind for Homeowners, Farmers, & Businesses Resources & Tools Agricultural and Rural Resources and Tools This page lists...

  16. Alliance for Rural Electrification | Open Energy Information

    Open Energy Info (EERE)

    Alliance for Rural Electrification1 The Alliance for Rural Electrification is the only international business association in the world focusing on the promotion and the...

  17. Alaska Village Initiatives Rural Small Business Conference

    Broader source: Energy.gov [DOE]

    The Alaska Rural Business Conference brings together rural businesses and leaders to provide them with networking opportunities, trainings, and technical information.

  18. Whitewater Valley Rural EMC | Open Energy Information

    Open Energy Info (EERE)

    Valley Rural EMC Jump to: navigation, search Name: Whitewater Valley Rural EMC Address: P.O. Box 349 Place: Liberty, Indiana Zip: 47353 Sector: Transmission Phone Number: (765)...

  19. MELCOR model for an experimental 17x17 spent fuel PWR assembly.

    SciTech Connect (OSTI)

    Cardoni, Jeffrey

    2010-11-01

    A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

  20. PCI-related cladding failures during off-normal events - draft. [PWR; BWR

    SciTech Connect (OSTI)

    Van Houten, R.; Tokar, M.; MacDonald, P.E.

    1984-05-01

    Pellet-cladding interaction (PCI) has long been identified as a fuel rod failure mechanism during power increases in both pressurized and boiling water reactors, and commercial guidelines have practically eliminated such failures during standard operations. A question remains regarding the possible formation of through-wall cladding cracks during several types of postulated off-normal reactor events involving power increases. This report includes preliminary findings for reactor events of the type addressed by Chapter 15 of the NRC Standard Review Plan. Specifically, the BWR turbine trip without bypass, PWR control rod withdrawal error, subcritical PWR control rod withdrawal error, BWR control blade withdrawal error, and the PWR steamline break are analyzed on the joint bases of peak rod power, power increase, ramp rate, and duration at elevated power. These Chapter 15 events are compared to numerous test reactor results and to other relevant investigations, and tentative conclusions on transient severity and data base adequacy are presented. Progress in developing computer codes for predicting PCI-induced fuel rod failures is also discussed. 49 references.

  1. Chile rural electrification cooperation

    SciTech Connect (OSTI)

    Flowers, L.

    1997-12-01

    The author describes a joint program to use renewables for rural electrification projects in Chile. The initial focus was in a limited part of the country, involving wind mapping, pilot project planning, training, and development of methodologies for comparative evaluations of resources. To this point three wind hybrid systems have been installed in one region, as a part of the regional private utility, and three additional projects are being designed. Additional resource assessment and training is ongoing. The author points out the difficulties in working with utilities, the importance of signed documentation, and the need to look at these programs as long term because of the time involved in introducing such new technologies.

  2. Physical Uncertainty Bounds (PUB)

    SciTech Connect (OSTI)

    Vaughan, Diane Elizabeth; Preston, Dean L.

    2015-03-19

    This paper introduces and motivates the need for a new methodology for determining upper bounds on the uncertainties in simulations of engineered systems due to limited fidelity in the composite continuum-level physics models needed to simulate the systems. We show that traditional uncertainty quantification methods provide, at best, a lower bound on this uncertainty. We propose to obtain bounds on the simulation uncertainties by first determining bounds on the physical quantities or processes relevant to system performance. By bounding these physics processes, as opposed to carrying out statistical analyses of the parameter sets of specific physics models or simply switching out the available physics models, one can obtain upper bounds on the uncertainties in simulated quantities of interest.

  3. Pub-3000: CH45

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

    C. Transporting Hazardous Materials Work Process D. Berkeley Lab Chemical Management System (Chemical Inventory) Work Process E. Chemical Hazard Descriptions Work Process F....

  4. SLAC-PUB-8640

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

    ... of the time as a proton and part of the time as a neutron surrounded by a positive meson cloud." and also: "...it is to be expected that the magnetic field associated with the ...

  5. IG-0550.PUB

    Energy Savers [EERE]

    were not prioritized to balance mission requirements, reduce risks, and minimize life-cycle costs. In some cases, disposition plans were in conflict with requirements for new...

  6. IG-0570.pub

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

    ... of Energy's Pit Production Project (DOEIG-0551, ... Our current findings regarding depleted uranium operations at Y-12 are consistent with our earlier observations. MANAGEMENT ...

  7. ASCI Pub (Scott)

    National Nuclear Security Administration (NNSA)

    ACADEMIC STRATEGIC ALLIANCES PROGRAM NA-ASC-107R-05-Vol.1-Rev.0 STANFORD UNIVERSITY Center for Integrated Turbulence Simulations CALIFORNIA INSTITUTE OF TECHNOLOGY Center for Simulating the Dynamic Response of Materials PAGE 4 UNIVERSITY OF UTAH Center for the Simulation of Accidental Fires & Explosions PAGE 20 PAGE 8 UNIVERSITY OF CHICAGO Center for Astrophysical Thermonuclear Flashes PAGE 12 UNIVERSITY OF ILLINOIS at Urbana-Champaign Center for Simulation of Advanced Rockets PAGE 16

  8. IG-0568.pub

    Energy Savers [EERE]

    ... critical information technology resources. ... 11 REMOTE ACCESS TO UNCLASSIFIED INFORMATION SYSTEMS TABLE OF CONTENTS ... Management waived an exit conference. SCOPE ...

  9. IG-0573.pub

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

    ... at the INEEL reservation, the Office of Nuclear Energy, Science and Technology (Nuclear Energy) is responsible for waste generated by past and on-going operations at Argonne West. ...

  10. SCANNED DARHT.pub

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

    5 Appendices 1. Objective, Scope, and Methodology ...... costs, similar to the funding methodology used for the commissioning of the first axis. ...

  11. IG-0564.pub

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

    4 AUDIT REPORT U.S. DEPARTMENT OF ENERGY OFFICE OF INSPECTOR GENERAL OFFICE OF AUDIT SERVICES ADVANCED VITRIFICATION SYSTEM AUGUST 2002 U. S. DEPARTMENT OF ENERGY Washington, DC 20585 August 20, 2002 MEMORANDUM FOR THE SECRETARY FROM: Gregory H. Friedman (Signed) Inspector General SUBJECT: INFORMATION: Audit Report on "Advanced Vitrification System" BACKGROUND The Department of Energy (Department) faces an environmental remediation task of unprecedented scope and technical complexity.

  12. IG-0565.pub

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

    5 AUDIT REPORT U.S. DEPARTMENT OF ENERGY OFFICE OF INSPECTOR GENERAL OFFICE OF AUDIT SERVICES SALT PROCESSING PROJECT AT THE SAVANNAH RIVER SITE AUGUST 2002 MEMORANDUM FOR THE SECRETARY FROM: Gregory H. Friedman (Signed) Inspector General SUBJECT: INFORMATION: Audit Report on "Salt Processing Project at the Savannah River Site" BACKGROUND As a result of Cold War weapons production, the Department of Energy accumulated millions of gallons of high-level waste that now require treatment

  13. IG-0566.pub

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

    6 AUDIT REPORT U.S. DEPARTMENT OF ENERGY OFFICE OF INSPECTOR GENERAL OFFICE OF AUDIT SERVICES NATIONAL NUCLEAR SECURITY ADMINISTRATION'S TEST READINESS PROGRAM SEPTEMBER 2002 MEMORANDUM FOR THE SECRETARY FROM: Gregory H. Friedman (Signed) Inspector General SUBJECT: INFORMATION: Audit Report on "National Nuclear Security Administration's Test Readiness Program" BACKGROUND Prior to 1992, the Department of Energy relied on underground testing at its Nevada Test Site to ensure the safety,

  14. IG-0571.pub

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

    1 AUDIT REPORT U.S. DEPARTMENT OF ENERGY OFFICE OF INSPECTOR GENERAL OFFICE OF AUDIT SERVICES IDAHO SETTLEMENT AGREEMENT ACTIVITIES OCTOBER 2002 MEMORANDUM FOR THE SECRETARY FROM: Gregory H. Friedman (Signed) Inspector General SUBJECT: INFORMATION: Audit Report on "Idaho Settlement Agreement Activities" BACKGROUND The task of cleaning up contaminated sites and disposing of radioactive waste, as reported in the Department of Energy's recent Performance and Accountability Report, is one

  15. IG-0574.pub

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

    Calutron Isotope Production Capabilities DOE/IG-0574 November 2002 MEMORANDUM FOR THE SECRETARY FROM: Gregory H. Friedman (Signed) Inspector General SUBJECT: INFORMATION: Audit Report on the "Calutron Isotope Production Capabilities" BACKGROUND In August 2002, the Department of Energy (Department) formalized its plan to permanently disable the only domestic facility capable of producing electromagnetically enriched stable isotopes. The Department intended to dismantle the equipment in

  16. IG-0572.pub

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

    ... Details of Finding Strategic Approach Page 3 Details of ... for monitoring BMIS implementation status or ensuring development ... areas were not present; and performance results ...

  17. SLAC-PUB-16075

    Office of Scientific and Technical Information (OSTI)

    Contract No. DE-AC02-76SF00515 and HEP. 1 Introduction The linear optics in ... For high-energy physics, a circular Higgs factory 14 re- quires an extremely large momentum ...

  18. IG-0569.pub

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

    ... Weak access controls and poor password management also increased the risk of unauthorized access. For instance, the Commission did not always employ strong password controls to ...

  19. scanned pma.pub

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

    Infrastructure Series Power Marketing Administration Infrastructure Protection ... INFORMATION: Audit Report on "Power Marketing Administration Infrastructure ...

  20. IG-0549.pub

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

    ... DOE preferred vitrification because it is assumed that borosilicate glass, or vitrified waste, would be the only waste form acceptable in the high-level waste repository. ...

  1. China-NREL Rural Electrification Projects | Open Energy Information

    Open Energy Info (EERE)

    Rural Electrification Projects Jump to: navigation, search Logo: China Rural Electrification Name China Rural Electrification AgencyCompany Organization National Renewable Energy...

  2. Korea Rural Community Corp KRC | Open Energy Information

    Open Energy Info (EERE)

    search Name: Korea Rural Community Corp (KRC) Place: Korea (Republic) Product: South Korea-based rural development company. References: Korea Rural Community Corp...

  3. USDA- Rural Energy for America Program (REAP) Loan Guarantees

    Broader source: Energy.gov [DOE]

    The Rural Energy for America Program (REAP) provides financial assistance to agricultural producers and rural small businesses in rural America to purchase, install, and construct renewable energ...

  4. USDA Rural Development Energy Programs

    Office of Environmental Management (EM)

    rural small businesses in purchasing renewable energy systems and making energy efficiency ... New Farm Bill was passed and signed into law on 2- 7- 14. Changes to funding levels ...

  5. 2016 Alaska Rural Energy Conference

    Broader source: Energy.gov [DOE]

    The 2016 Alaska Rural Energy Conference is a three-day event that offers a large variety of technical sessions covering new and ongoing energy projects in Alaska, as well as new technologies and needs for Alaska's remote communities.

  6. Analysis of a double-ended cold-leg break simulation: THTF Test 3. 05. 5B. [PWR

    SciTech Connect (OSTI)

    Craddick, W.G.; Pevey, R.E.

    1982-09-01

    On July 3, 1980, an experiment was performed in the Oak Ridge National Laboratory Thermal-Hydraulic Test Facility that simulated a double-ended cold-leg break pressurized-water reactor (PWR) accident. Analysis of the experiment revealed that nuclear fuel rods exposed to the same hydrodynamic environment as that which existed in the experiment would have departed from nucleate boiling both earlier and later than the fuel rod simulator (FRS), depending on the size of the gap between the nuclear fuel pellets and cladding and on the initial power of the nuclear fuel rod. Comparison of the results of the current experiment, which used an FRS bundle with geometry similar to 17 x 17 PWR fuel assemblies, to the results of earlier experiments, which used an FRS bundle with geometry similar to 15 x 15 PWR fuel assemblies, revealed no differences that can be attributed to the difference in geometries.

  7. VERA Modeling and Simulation of the AP1000 PWR Cycle 1 Depletion

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

    CASL-U-2015-0302-000 VERA Modeling and Simulation of the AP1000 PWR Cycle 1 Depletion L3:VMA.AMA.P11.06 David Salazar, Westinghouse Fausto Franceschini, Westinghouse September 30, 2015 L3:VMA.AMA.P11.06 Official Use Only ii Protected under CASL Master NDA CASL-U-2015-0302-000 REVISION LOG Revision Date Affected Pages Revision Description 0 09/30/2015 All Initial issuance Document pages that are: Export Controlled ____________No______________________________________ IP/Proprietary/NDA

  8. Thermal Response of the 21-PWR Waste Package to a Fire Accident

    SciTech Connect (OSTI)

    F.P. Faucher; H. Marr; M.J. Anderson

    2000-10-03

    The objective of this calculation is to evaluate the thermal response of the 21-PWR WP (pressurized water reactor waste package) to the regulatory fire event. The scope of this calculation is limited to the two-dimensional waste package temperature calculations to support the waste package design. The information provided by the sketches attached to this calculation (Attachment IV) is that of the potential design of the type of waste package considered in this calculation. The procedure AP-3.12Q.Calculations (Reference 1), and the Development Plan (Reference 24) are used to develop this calculation.

  9. SCALE 5.1 Predictions of PWR Spent Nuclear Fuel Isotopic Compositions

    SciTech Connect (OSTI)

    Radulescu, Georgeta; Gauld, Ian C; Ilas, Germina

    2010-03-01

    The purpose of this calculation report is to document the comparison to measurement of the isotopic concentrations for pressurized water reactor (PWR) spent nuclear fuel determined with the Standardized Computer Analysis for Licensing Evaluation (SCALE) 5.1 (Ref. ) epletion calculation method. Specifically, the depletion computer code and the cross-section library being evaluated are the twodimensional (2-D) transport and depletion module, TRITON/NEWT,2, 3 and the 44GROUPNDF5 (Ref. 4) cross-section library, respectively, in the SCALE .1 code system.

  10. Neutronics and safety characteristics of a 100% MOX fueled PWR using weapons grade plutonium

    SciTech Connect (OSTI)

    Biswas, D.; Rathbun, R.; Lee, Si Young; Rosenthal, P.

    1993-12-31

    Preliminary neutronics and safety studies, pertaining to the feasibility of using 100% weapons grade mixed-oxide (MOX) fuel in an advanced PWR Westinghouse design are presented in this paper. The preliminary results include information on boron concentration, power distribution, reactivity coefficients and xenon and control rode worth for the initial and the equilibrium cycle. Important safety issues related to rod ejection and steam line break accidents and shutdown margin requirements are also discussed. No significant change from the commercial design is needed to denature weapons-grade plutonium under the current safety and licensing criteria.

  11. Application of LBB to high energy piping systems in operating PWR

    SciTech Connect (OSTI)

    Swamy, S.A.; Bhowmick, D.C.

    1997-04-01

    The amendment to General Design Criterion 4 allows exclusion, from the design basis, of dynamic effects associated with high energy pipe rupture by application of leak-before-break (LBB) technology. This new approach has resulted in substantial financial savings to utilities when applied to the Pressurized Water Reactor (PWR) primary loop piping and auxiliary piping systems made of stainless steel material. To date majority of applications pertain to piping systems in operating plants. Various steps of evaluation associated with the LBB application to an operating plant are described in this paper.

  12. Secondary Startup Neutron Sources as a Source of Tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS)

    SciTech Connect (OSTI)

    Shaver, Mark W.; Lanning, Donald D.

    2010-02-01

    The hypothesis of this paper is that the Zircaloy clad fuel source is minimal and that secondary startup neutron sources are the significant contributors of the tritium in the RCS that was previously assigned to release from fuel. Currently there are large uncertainties in the attribution of tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS). The measured amount of tritium in the coolant cannot be separated out empirically into its individual sources. Therefore, to quantify individual contributors, all sources of tritium in the RCS of a PWR must be understood theoretically and verified by the sum of the individual components equaling the measured values.

  13. Implementation of Division F, Title I, Title II, and Title III and Division G, Consolidated and Further Continuing Appropriations Act, 2013, Pub. L. No.113-6

    Broader source: Energy.gov [DOE]

    Acquisition Letter (AL) 2013-06 and Financial Assistance Letter (FAL) 2013-04 provides implementing instructions and guidance for Division F, Title I, Title II, and Title III and Division G, Consolidated and Further Continuing Appropriations Act, 2013, Pub. L. No.113-6.

  14. Implementation of Division B, Title III, Title V and Division C Title VII, Consolidated Appropriations Act, 2012, Pub. L. No.112-74 and Related Conference Report

    Broader source: Energy.gov [DOE]

    Acquisition Letter (AL) 2012-08 and Financial Assistance Letter (FAL) 2012-01 provides implementing instructions and guidance for Division B, Title III, Title V and Division C Title VII, Consolidated Appropriations Act, 2012, Pub. L. No.112-74 and Related Conference Report.

  15. Code System for PWR & BWR Multicompartment Containment Analysis, Versions MOD5

    Energy Science and Technology Software Center (OSTI)

    1999-06-02

    CONTEMPT4/MOD6 describes the response of multicompartment containment systems subjected to postulated loss-of-coolant accident (LOCA) conditions. The program can accommodate both pressurized water reactor (PWR) and boiling water reactor (BWR) containment systems. Also, both design basis accident (DBA) and degraded core type LOCA conditions can be analyzed. The program calculates the time variation of compartment pressures, temperatures, and mass and energy inventories due to inter-compartment mass and energy exchange taking into account user-supplied descriptions of compartments,more » inter-compartment junction flow areas, LOCA source terms, and user-selected problem features. Analytical models available to describe containment systems include models for containment fans and pumps, cooling sprays, heat conducting structures, sump drains, PWR ice condensers, and BWR pressure suppression systems. CONTEMPT4/MOD6 also provides analytical models for hydrogen and carbon monoxide combustion within compartments and energy transfer due to gas radiation to accommodate degraded core type accidents.« less

  16. Phenomenon analysis of stress corrosion cracking in the vessel head penetrations of French PWR`s

    SciTech Connect (OSTI)

    Pichon, C.; Buisine, D.; Faidy, C.; Gelpi, A.; Vaindirlis, M.

    1995-12-31

    During a hydrotest in 1991, a leak was detected on,a reactor vessel head (RVH) penetration of a French PWR. This leak was due to a phenomenon of Primary Water Stress Corrosion Cracking (PWSCC) affecting these penetrations in Alloy 600. The destructive and non-destructive examinations undertaken during the following months highlighted the generic nature of the degradations. In order to well understand this phenomenon and implement the most suitable maintenance policy, a large scale scientific program was decided and performed jointly by Electricite de France and FRAMATOME. The paper will present all the results obtained in this program concerning the parameters governing the PWSCC. In particular the following fields will be developed: (1) the material, its microstructure in line with the manufacturing and its susceptibility to PWSCC; (2) the stresses and their evaluations by measurements, mock up corrosion tests and Finite Element Analysis (FEA); (3) the effect of surface finish on crack initiation; and (4) the crack growth rate. This phenomenon analysis will be useful for evaluating the risk of PWSCC on other Alloy 600 areas in PWR`s primary system.

  17. Recommendations for Addressing Axial Burnup in the PWR Burnup Credit Analyses

    SciTech Connect (OSTI)

    Wagner, J.C.

    2002-10-23

    This report presents studies performed to support the development of a technically justifiable approach for addressing the axial-burnup distribution in pressurized-water reactor (PWR) burnup-credit criticality safety analyses. The effect of the axial-burnup distribution on reactivity and proposed approaches for addressing the axial-burnup distribution are briefly reviewed. A publicly available database of profiles is examined in detail to identify profiles that maximize the neutron multiplication factor, k{sub eff}, assess its adequacy for PWR burnup credit analyses, and investigate the existence of trends with fuel type and/or reactor operations. A statistical evaluation of the k{sub eff} values associated with the profiles in the axial-burnup-profile database was performed, and the most reactive (bounding) profiles were identified as statistical outliers. The impact of these bounding profiles on k{sub eff} is quantified for a high-density burnup credit cask. Analyses are also presented to quantify the potential reactivity consequence of loading assemblies with axial-burnup profiles that are not bounded by the database. The report concludes with a discussion on the issues for consideration and recommendations for addressing axial burnup in criticality safety analyses using burnup credit for dry cask storage and transportation.

  18. Validation of the new code package APOLLO2.8 for accurate PWR neutronics calculations

    SciTech Connect (OSTI)

    Santamarina, A.; Bernard, D.; Blaise, P.; Leconte, P.; Palau, J. M.; Roque, B.; Vaglio, C.; Vidal, J. F.

    2013-07-01

    This paper summarizes the Qualification work performed to demonstrate the accuracy of the new APOLLO2.S/SHEM-MOC package based on JEFF3.1.1 nuclear data file for the prediction of PWR neutronics parameters. This experimental validation is based on PWR mock-up critical experiments performed in the EOLE/MINERVE zero-power reactors and on P.I. Es on spent fuel assemblies from the French PWRs. The Calculation-Experiment comparison for the main design parameters is presented: reactivity of UOX and MOX lattices, depletion calculation and fuel inventory, reactivity loss with burnup, pin-by-pin power maps, Doppler coefficient, Moderator Temperature Coefficient, Void coefficient, UO{sub 2}-Gd{sub 2}O{sub 3} poisoning worth, Efficiency of Ag-In-Cd and B4C control rods, Reflector Saving for both standard 2-cm baffle and GEN3 advanced thick SS reflector. From this qualification process, calculation biases and associated uncertainties are derived. This code package APOLLO2.8 is already implemented in the ARCADIA new AREVA calculation chain for core physics and is currently under implementation in the future neutronics package of the French utility Electricite de France. (authors)

  19. Application of the MELCOR code to design basis PWR large dry containment analysis.

    SciTech Connect (OSTI)

    Phillips, Jesse; Notafrancesco, Allen (USNRC, Office of Nuclear Regulatory Research, Rockville, MD); Tills, Jack Lee (Jack Tills & Associates, Inc., Sandia Park, NM)

    2009-05-01

    The MELCOR computer code has been developed by Sandia National Laboratories under USNRC sponsorship to provide capability for independently auditing analyses submitted by reactor manufactures and utilities. MELCOR is a fully integrated code (encompassing the reactor coolant system and the containment building) that models the progression of postulated accidents in light water reactor power plants. To assess the adequacy of containment thermal-hydraulic modeling incorporated in the MELCOR code for application to PWR large dry containments, several selected demonstration designs were analyzed. This report documents MELCOR code demonstration calculations performed for postulated design basis accident (DBA) analysis (LOCA and MSLB) inside containment, which are compared to other code results. The key processes when analyzing the containment loads inside PWR large dry containments are (1) expansion and transport of high mass/energy releases, (2) heat and mass transfer to structural passive heat sinks, and (3) containment pressure reduction due to engineered safety features. A code-to-code benchmarking for DBA events showed that MELCOR predictions of maximum containment loads were equivalent to similar predictions using a qualified containment code known as CONTAIN. This equivalency was found to apply for both single- and multi-cell containment models.

  20. Study of a transient identification system using a neural network for a PWR plant

    SciTech Connect (OSTI)

    Ishihara, Yoshinao; Kasai, Masao; Kambara, Masayuki [Mitsubishi Heavy Industries, Ltd., Yokohama (Japan); Mitsuda, Hiromichi; Kurata, Toshikazu; Shirosaki, Hidekazu [Inst. of Nuclear Safety System, Inc., Kyoto (Japan)

    1996-08-01

    This paper presents the procedure and results of a system for identifying PWR plant abnormal events, which uses neural network techniques. The neural network recognizes the abnormal event from the patterns of the transient changes of analog data from plant parameters when they deport from their normal state. For the identification of abnormal events in this study, events that cause a reactor to scram during power operation were selected as the design base events. The test data were prepared by simulating the transients on a compact PWR simulator. The simulation data were analyzed to determine how the plant parameters respond after the occurrence of a transient. A method of converting the pattern of the transient changes into characteristic parameters by fitting the data to pre-determined functions was developed. These characteristic parameters were used as the input data to the neural network. The neural network learning procedure used a generalized delta rule, namely a back-propagation algorithm. The neural network can identify the type of an abnormal event from a limited set of events by using these characteristic parameters obtained from the pattern of the changes in the analog data. From the results of this application of a neural network, it was concluded that it would be possible to use the method to identify abnormal events in a nuclear power plant.

  1. Lyon Rural Electric Coop | Open Energy Information

    Open Energy Info (EERE)

    Lyon Rural Electric Coop Jump to: navigation, search Name: Lyon Rural Electric Coop Place: Iowa Phone Number: 712-472-2506 Website: www.lyonrec.coop Outage Hotline: 1-800-292-8989...

  2. Alaska Village Initiatives Rural Business Conference

    Broader source: Energy.gov [DOE]

    Hosted by the Alaska Village Initiative, the 24th Annual Rural Small Business Conference brings together rural businesses and leaders to provide them with networking opportunities, training, and technical information.

  3. Development of a model for predicting intergranular stress corrosion cracking of Alloy 600 tubes in PWR primary water. Final report

    SciTech Connect (OSTI)

    Garud, Y.S.

    1985-01-01

    A preliminary mathematical model developed in this study may make it possible to predict stress corrosion cracking on the primary side of PWR steam generator tubing. The study outlines a comprehensive testing program that will provide the operational and experimental data to further develop and verify the model.

  4. Maximim Accelerations On The Fuel Assemblies Of a 21-PWR Waste Package During End Impacts 

    SciTech Connect (OSTI)

    V. DeLa Brosse

    2003-03-27

    The objective of this calculation is to determine the acceleration of the fuel assemblies contained in a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities of the waste package is studied. The scope of this calculation is limited to estimating the acceleration of the fuel assemblies during the impact.

  5. Maximim Accelerations On The Fuel Assemblies Of a 21-PWR Waste Package During End Impacts 

    SciTech Connect (OSTI)

    T. Schmitt

    2005-08-17

    The objective of this calculation is to determine the acceleration of the fuel assemblies contained in a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities of the waste package is studied. The scope of this calculation is limited to estimating the acceleration of the fuel assemblies during the impact.

  6. USDA Rural Business Enterprise Grant (RBEG) Program

    Broader source: Energy.gov [DOE]

    Request for Proposals for funding to support the development of small and emerging private business enterprises in rural Alaska

  7. A comparison of the CHF between tubes and annuli under PWR thermal-hydraulic conditions

    SciTech Connect (OSTI)

    Herer, C.

    1995-09-01

    Critical Heat Flux (CHF) tests were carried out in three tubes with inside diameters of 8, 13, and 19.2 mm and in two annuli with an inner tube of 9.5 mm and an outer tube of 13 or 19.2 mm. All axial heat flux distributions in the test sections were uniform. The coolant fluid was Refrigerant 12 (Freon-12) under PWR thermal-hydraulic conditions (equivalent water conditions - Pressure: 7 to 20 MPa, Mass Velocity: 1000 to 6000 kg/m2/s, Local Quality: -75% to +45%). The effect of tube diameter is correlated for qualities under 15%. The change from the tube to the annulus configuration is correctly taken into account by the equivalent hydraulic diameter. Useful information is also provided concerning the effect of a cold wall in an annulus.

  8. IMPACT OF FISSION PRODUCTS IMPURITY ON THE PLUTONIUM CONTENT IN PWR MOX FUELS

    SciTech Connect (OSTI)

    Gilles Youinou; Andrea Alfonsi

    2012-03-01

    This report presents the results of a neutronics analysis done in response to the charter IFCA-SAT-2 entitled 'Fuel impurity physics calculations'. This charter specifies that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies (UOX SNF) is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate PWR MOX fuel assemblies. Only non-gaseous FP have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1). This mixture of Pu and FP is called PuFP. Note that, in this preliminary analysis, the FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.

  9. On the explanation and calculation of anomalous reflood hydrodynamics in large PWR cores

    SciTech Connect (OSTI)

    Rodriguez, S.E.

    1985-01-01

    Reflood hydrodynamics from large-scale (1:20) test facilities in Japan have yielded apparently anomalous behavior relative to FLECHT tests. Namely, even at reflooding rates below one inch per second, very large liquid volume fractions (10-15%) exist above the quench fronts shortly after flood begins; thus cladding temperature excursions are terminated early in the reflood phase. This paper discusses an explanation for this behavior: liquid films on the core's unheated rods. The experimental findings are shown to be correctly simulated with a new four-field (vapor, films, droplets) version of the best-estimate TRAC-PF1 computer code, TRAC-FF. These experimental and analytical findings have important implications for PWR large-break LOCA licensing.

  10. Effect of coolant chemistry on PWR radiation transport processes. Progress report on reactor loop studies

    SciTech Connect (OSTI)

    Brown, D.J.; Flynn, G.; Haynes, J.W.; Kitt, G.P.; Large, N.R.; Lawson, D.; Mead, A.P.; Nichols, J.L.; Woodwark, D.R.

    1986-05-01

    The effect of various PWR-type coolant chemistry regimes on the behavior of corrosion products has been studied in the DIDO Water Loop at Harwell. There are strong indications that the in-core deposition behavior of corrosion product species is not fully accounted for by the solubility model based on nickel ferrite; boric acid plays a role apart from its influence on pH, and corrosion products are adsorbed to some extent in the zirconium oxide film on the fuel cladding. In DWL, soluble species appear to be dominant in deposition processes. A most important factor governing deposition behavior is surface condition; the influence of weld regions and the effect of varying pretreatment conditions have both been demonstrated. 13 figs.

  11. Fuel-rod response during the large-break LOCA Test LOC-6. [PWR

    SciTech Connect (OSTI)

    Vinjamuri, K.; Cook, B.A.; Hobbins, R.R.

    1981-01-01

    The large break Loss of Coolant Accident (LOCA) Test LOC-6 was conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory by EG and G Idaho, Inc. The objectives of the PBF LOCA tests are to obtain in-pile cladding ballooning data under blowdown and reflood conditions and assess how well out-of-pile ballooning data represent in-pile fuel rod behavior. The primary objective of the LOC-6 test was to determine the effects of internal rod pressures and prior irradiation on the deformation behavior of fuel rods that reached cladding temperatures high in the alpha phase of zircaloy. Test LOC-6 was conducted with four rods of PWR 15 x 15 design with the exception of fuel stack length (89 cm) and enrichment (12.5 W% /sup 235/U). Each rod was surrounded by an individual flow shroud.

  12. LOCA rupture strains and coolability of full-length PWR fuel bundles

    SciTech Connect (OSTI)

    Mohr, C.L.; Hesson, G.M.

    1983-03-01

    The LOCA Simulation Program tests sponsored by the United States Nuclear Regulatory Commission are the first full-length nuclear-heated experiments designed to investigate the deformation and rupture characteristics as well as the coolability of nuclear-heated fuel under accident conditions. The results of the seven tests preformed in the program using 32-rod full-length PWR fuel bundles have shown that for a wide range of flow blockage condtions no significant reduction in coolability of the fuel bundle could be found. These results have been confirmed by data from out-of-pile electrically-heated experiments. Although there is a difference between nuclear and electrically-heated test data, the conclusion is still the same. Coolability of a deformed bundle during reflood is dominated by the dispersion of droplets in the deformed zone which provides adequate cooling and which is not reduced by the deformation of the fuel rod cladding.

  13. Source term experiment STEP-3 simulating a PWR severe station blackout

    SciTech Connect (OSTI)

    Simms, R.; Baker, L. Jr.; Ritzman, R.L.

    1987-05-21

    For a severe PWR accident that leads to a loss of feedwater to the steam generators, such as might occur in a station blackout, fission product decay heating will cause a water boiloff. Without effective cooling of the core, steam will begin to oxidize the Zircaloy cladding. The noble gases and volatile fission products, such as Cs and I, that are major contributors to the radiological source term, will be released from the damaged fuel shortly after cladding failure. The accident environment when these volatile fission products escape was simulated in STEP-3 using four fuel elements from the Belgonucleaire BR3 reactor. The primary objective was to examine the releases in samples collected as close to the test zone as possible. In this paper, an analysis of the temperatures and hydrogen generation is compared with the measurements. The analysis is needed to estimate releases and characterize conditions at the source for studies of fission product transport.

  14. Electrically heated ex-reactor pellet-cladding interaction (PCI) simulations utilizing irradiated Zircaloy cladding. [PWR

    SciTech Connect (OSTI)

    Barner, J.O.; Fitzsimmons, D.E.

    1985-02-01

    In a program sponsored by the Fuel Systems Research Branch of the US Nuclear Regulatory Commission, a series of six electrically heated fuel rod simulation tests were conducted at Pacific Northwest Laboratory. The primary objective of these tests was to determine the susceptibility of irradiated pressurized-water reactor (PWR) Zircaloy-4 cladding to failures caused by pellet-cladding mechanical interaction (PCMI). A secondary objective was to acquire kinetic data (e.g., ridge growth or relaxation rates) that might be helpful in the interpretation of in-reactor performance results and/or the modeling of PCMI. No cladding failures attributable to PCMI occurred during the six tests. This report describes the testing methods, testing apparatus, fuel rod diametral strain-measuring device, and test matrix. Test results are presented and discussed.

  15. Decay Heat of Major Radionuclides for PWR Spent Fuels to 10,000 Years

    SciTech Connect (OSTI)

    J.S. Tang

    2001-12-20

    The objective of this calculation is to determine decay heat of a pressurized-water reactor (PWR) spent nuclear fuel (SNF) assembly with four different initial-enrichment and burnup characteristics. The major contributing radionuclides to the decay heat are also identified and graphically presented. The scope of this calculation is limited to the time period of the first 10,000 years after discharge from reactors. The results of this calculation will be used to evaluate the effects of the projected commercial spent nuclear fuel (CSNF) inventory on the repository design based on revised nuclear energy forecasts. This calculation was performed in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (BSC (Bechtel SAIC Company) 2001). AP-3.12Q, Calculations, is used to perform the calculation and develop the document. This calculation is associated with the repository design activity.

  16. Probability and consequences of a rapid boron dilution sequence in a PWR

    SciTech Connect (OSTI)

    Diamond, D.J.; Kohut, P.; Nourbakhsh, H.; Valtonen, K.; Secker, P.

    1995-11-01

    The reactor restart scenario is one of several beyond-design-basis events in a pressurized water reactor (PWR) which can lead to rapid boron dilution in the core. This in turn can lead to a power excursion and the potential for fuel damage. A probabilistic analysis had been done for this event for a European PWR. The estimated core damage frequency was found to be high partially because of a high frequency for a LOOP and assumptions regarding operator actions. As a result, a program of analysis and experiment was initiated and corrective actions were taken. A system was installed so that the suction of the charging pumps would switch to the highly borated refueling water storage tank (RWST) when there was a trip of the RCPs. This was felt to reduce the estimated core damage frequency to an acceptable level. In the US, this original study prompted the Nuclear Regulatory Commission to issue an information notice to follow work being done in this area and to initiate studies such as the work at BNL reported herein. In order to see if the core damage frequency might be as high in US plants, a probabilistic assessment of this scenario was done for three plants. Two important conservative assumptions in this analysis were that (1) the mixing of the injectant was insignificant and (2) fuel damage occurs when the slug passes through the core. In order to study the first assumption, analysis was carried out for two of the plants using a mixing model. The second assumption was studied by calculating the neutronic response of the core to a slug of deborated water for one of the plants. All three types of analyses are summarized below. More information is available in the original report.

  17. The CASTOR-V/21 PWR spent-fuel storage cask: Testing and analyses: Interim report

    SciTech Connect (OSTI)

    Dziadosz, D.; Moore, E.V.; Creer, J.M.; McCann, R.A.; McKinnon, M.A.; Tanner, J.E.; Gilbert, E.R.; Goodman, R.L.; Schoonen, D.H.; Jensen, M.

    1986-11-01

    A performance test of a Gesellschaft fuer Nuklear Service CASTOR-V/21 pressurized water reactor (PWR) spent fuel storage cask was performed. The test was the first of a series of cask performance tests planned under a cooperative agreement between Virginia Power and the US Department of Energy. The performance test consisted of loading the CASTOR-V/21 cask with 21 PWR spent fuel assemblies from Virginia Power's Surry reactor. Cask surface and fuel assembly guide tube temperatures, and cask surface gamma and neutron dose rates were measured. Testing was performed with vacuum, nitrogen, and helium backfill environments in both vertical and horizontal cask orientations. Limited spent fuel integrity data were also obtained. Results of the performance test indicate the CASTOR-V/21 cask exhibited exceptionally good heat transfer performance which exceeded design expectations. Peak cladding temperatures with helium and nitrogen backfills in a vertical cast orientation and with helium in a horizontal orientation were less than the allowable of 380/sup 0/C with a total cask heat load of 28 kW. Significant convection heat transfer was present in vertical nitrogen and helium test runs as indicated by peak temperatures occurring in the upper regions of the fuel assemblies. Pretest temperature predictions of the HYDRA heat transfer computer program were in good agreement with test data, and post-test predictions agreed exceptionally well (25/sup 0/C) with data. Cask surface gamma and neutron dose rates were measured to be less than the design goal of 200 mrem/h. Localized peaks as high as 163 mrem/h were measured on the side of the cask, but peak dose rates of <75 mrem/h can easily be achieved with minor refinements to the gamma shielding design. From both heat transfer and shielding perspectives, the CASTOR-V/21 cask can, with minor refinements, be effectively implemented at reactor sites and central storage facilities for safe storage of spent fuel.

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

    SciTech Connect (OSTI)

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

    2007-07-01

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

  19. Radionuclide release from PWR fuels in a reference tuff repository groundwater subsquently changed to Radionuclide release from PWR fuels in J-13 well water

    SciTech Connect (OSTI)

    Wilson, C.N.; Oversby, V.M.

    1985-04-01

    The Nevada Nuclear Waste Storage Investigations Project (NNWSI) is studying the suitability of the welded devitrified Topopah Spring tuff at Yucca Mountain, Nye County, Nevada, for potential use as a high level nuclear waste repository. In support of the Waste Package task of NNWSI, tests have been conducted under ambient air environment to measure radionuclide release from two pressurized water reactor (PWR) spent fuels in water obtained from the J-13 well near the Yucca Mountain site. Four specimen types, representing a range of fuel physical conditions that may exist in a failed waste canister containing a limited amount of water were tested. The specimen types were: (1) fuel rod sections split open to expose bare fuel particles; (2) rod sections with water-tight end fittings with a 2.5-cm long by 150-{mu}m wide slit through the cladding; (3) rod sections with water-tight end fittings and two 200-{mu}m diameter holes through the cladding; and (4) undefected rod segments with water-tight end fittings. Radionuclide release results from the first 223-day test runs on H.B. Robinson spent fuel specimens in J-13 water are reported and compared to results from a previous test series in which similar Turkey Point reactor spent fuel specimens were tested in deionized water. Selected initial results are also given for Turkey Point fuel specimens tested in J-13 water. Results suggest that the actinides Pu, Am, Cm and Np are released congruently with U as the UO{sub 2} spent fuel matrix dissolves. Fractional release of {sup 137}Cs and {sup 99}Tc was greater than that measured for the actinides. Generally, lower radionuclide releases were measured for the H.B. Robinson fuel in J-13 water than for Turkey Point Fuel in deionized water.

  20. Analysis of results from a loss-of-offsite-power-initiated ATWS experiment in the LOFT facility. [PWR

    SciTech Connect (OSTI)

    Varacalle, D.J. Jr.; Koizumi, Y.; Giri, A.H.; Koske, J.E.; Sanchez-Pope, A.E.

    1983-01-01

    An anticipated transient without scram (ATWS), initiated by loss-of-offsite power, was experimentally simulated in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR). Primary system pressure was controlled using a scaled safety relief valve (SRV) representative of those in a commercial PWR, while reactor power was reduced by moderator reactivity feedback in a natural circulation mode. The experiment showed that reactor power decreases more rapidly when the primary pumps are tripped in a loss-of-offsite-power ATWS than in a loss-of-feedwater induced ATWS when the primary pumps are left on. During the experiment, the SRV had sufficient relief capacity to control primary system pressure. Natural circulation was effective in removing core heat at high temperature, pressure, and core power. The system transient response predicted using the RELAP5/MOD1 computer code showed good agreement with the experimental data.

  1. Results and analysis of a loss-of-feedwater induced ATWS experiment in the LOFT facility. [PWR

    SciTech Connect (OSTI)

    Grush, W.H.; Woerth, S.C.; Koizumi, Y.

    1983-01-01

    An anticipated transient without scram (ATWS), initiated by a loss of feedwater, was experimentally simulated in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR). Primary system pressure was controlled using a two-position actuator relief valve to simulate a scaled power-operated relief valve (PORV) and safety relief valve (SRV) representative of those in a commercial PWR. Auxiliary feedwater injection was delayed during the experiment until the plant recovery phase where long-term shutdown was achieved by an operator-controlled plant recovery procedure without inserting the control rods. The system transient response predicted by the RELAP5/MOD1 computer code showed good agreement with the experimental data.

  2. SAS2H Generated Isotopic Concentrations For B&W 15X15 PWR Assembly (SCPB:N/A)

    SciTech Connect (OSTI)

    J.W. Davis

    1996-08-29

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide pressurized water reactor (PWR) isotopic composition data as a function of time for use in criticality analyses. The objectives of this evaluation are to generate burnup and decay dependant isotopic inventories and to provide these inventories in a form which can easily be utilized in subsequent criticality calculations.

  3. Rural Cooperative Geothermal Development Electric & Agriculture |

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

    Department of Energy Rural Cooperative Geothermal Development Electric & Agriculture Rural Cooperative Geothermal Development Electric & Agriculture DOE 2010 Geothermal Program Peer Review; Low Temperature Demonstration Projects low_silveria_rural_electric_coop.pdf (557.69 KB) More Documents & Publications Southwest Alaska Regional Geothermal Energy Project District Wide Geothermal Heating Conversion Blaine County School District Novel Energy Conversion Equipment for Low

  4. Implementation of Division D, Titles III and V, and Division E, Title VII of the Consolidated Appropriations Act, 2014, Pub. L. No. 113-76.

    Broader source: Energy.gov [DOE]

    Acquisition Letter (AL) 2014-04 and Financial Assistance Letter (FAL) 2014-01 have been revised to remove language from Section 502 that was not carried forward from previous appropriation acts. FAL 2014-01 was also revised to update the Corporate Felony Conviction and Federal Tax Liability Representations and Assurances and the Conference Spending term. As a result, AL 2014-04 (Rev 1) and FAL 2014-01 (Rev 1) provide implementing instructions and guidance for Division D, Titles III and V, and Division E, Title VII of the Consolidated Appropriations Act, 2014, Pub. L. No. 113-76.

  5. Club for Rural Electrification | Open Energy Information

    Open Energy Info (EERE)

    Electrification Jump to: navigation, search Name: Club for Rural Electrification Place: Freiburg, Germany Zip: 79114 Sector: Solar Product: An industry association of German...

  6. Linn County Rural Electric Cooperative - Agricultural Energy...

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

    water heater installed, additional 25 bonus if electric dryer installed Energy Star Television: 50 Summary Linn County Rural Electric Cooperative Association (Linn County RECA)...

  7. Intermountain Rural Elec Assn | Open Energy Information

    Open Energy Info (EERE)

    Rural Elec Assn Place: Colorado Website: www.irea.coop Twitter: @IREAColorado Facebook: https:www.facebook.comIntermountainREA Outage Hotline: 1-800-332-9540 References:...

  8. Modular Biomass Systems Could Boost Rural Areas

    Broader source: Energy.gov [DOE]

    Increased ethanol production will help revitalize the rural economy and decrease America's dependence on foreign oil, but there are other ways to create opportunities in the farmlands.

  9. 2016 Kawerak/Rural Providers' Conference

    Broader source: Energy.gov [DOE]

    The Kawerak/Rural Providers' Conference is an annual gathering and cultural celebration featuring heads-on cultural events, moderated discussions, workshops, and keynote speakers.

  10. USDA Rural Development Washington State Workshop

    Broader source: Energy.gov [DOE]

    The U.S. Department of Agriculture is hosting a Washington Rural Development Workshop. Speakers will cover local and regional broadband initiatives program and broadband success stories,...

  11. Rural Electric Coop, Inc | Open Energy Information

    Open Energy Info (EERE)

    Facebook: https:www.facebook.compagesRural-Electric-Cooperative-Inc116401145072740?refts Outage Hotline: 855-399-2683 References: EIA Form EIA-861 Final Data File for 2010 -...

  12. Building Energy Efficiency in Rural China

    SciTech Connect (OSTI)

    Evans, Meredydd; Yu, Sha; Song, Bo; Deng, Qinqin; Liu, Jing; Delgado, Alison

    2014-04-01

    Rural buildings in China now account for more than half of Chinas total building energy use. Forty percent of the floorspace in China is in rural villages and towns. Most of these buildings are very energy inefficient, and may struggle to meet basic needs. They are cold in the winter, and often experience indoor air pollution from fuel use. The Chinese government plans to adopt a voluntary building energy code, or design standard, for rural homes. The goal is to build on Chinas success with codes in urban areas to improve efficiency and comfort in rural homes. The Chinese government recognizes rural buildings represent a major opportunity for improving national building energy efficiency. The challenges of rural China are also greater than those of urban areas in many ways because of the limited local capacity and low income levels. The Chinese government wants to expand on new programs to subsidize energy efficiency improvements in rural homes to build capacity for larger-scale improvement. This article summarizes the trends and status of rural building energy use in China. It then provides an overview of the new rural building design standard, and describes options and issues to move forward with implementation.

  13. Rural Innovations Network | Open Energy Information

    Open Energy Info (EERE)

    Network Jump to: navigation, search Name: Rural Innovations Network Place: India Sector: Services Product: General Financial & Legal Services ( Charity Non-profit Association...

  14. Viet Nam Rural Electrification | Open Energy Information

    Open Energy Info (EERE)

    Rural Electrification Location of project Vietnam Energy Services Lighting, Cooking and water heating, Space heating, Cooling Year initiated 2009 Organization Asian Development...

  15. Renewable Energy Technologies for Rural Electrification - The...

    Open Energy Info (EERE)

    Sector Jump to: navigation, search Tool Summary LAUNCH TOOL Name: Renewable Energy Technologies for Rural Electrification - The Role of the Private Sector AgencyCompany...

  16. Examination of spent PWR fuel rods after 15 years in dry storage.

    SciTech Connect (OSTI)

    Einziger, R.E.; Tsai, H.C.; Billone, M.C.; Hilton, B.A.

    2002-02-11

    Virginia Power Surry Nuclear Station Pressurized Water Reactor (PWR) fuel was stored in a dry inert atmosphere Castor V/21 cask at the Idaho National Environmental and Engineering Laboratory (INEEL) for 15 years at peak cladding temperatures decreasing from about 350 to 150 C. Prior to the storage, the loaded cask was subjected to extensive thermal benchmark tests. The cask was opened to examine the fuel for degradation and to determine if it was suitable for extended storage. No rod breaches had occurred and no visible degradation or crud/oxide spallation were observed. Twelve rods were removed from the center of the T11 assembly and shipped from INEEL to the Argonne-West HFEF for profilometric scans. Four of these rods were punctured to determine the fission gas release from the fuel matrix and internal pressure in the rods. Three of the four rods were cut into five segments each, then shipped to the Argonne-East AGHCF for detailed examination. The test plan calls for metallographic examination of six samples from two of the rods, microhardness and hydrogen content measurements at or near the six metallographic sample locations, tensile testing of six samples from the two rods, and thermal creep testing of eight samples from the two rods to determine the extent of residual creep life. The results from the profilometry (12 rods), gas release measurements (4 rods), metallographic examinations (2 samples from 1 rod), and microhardness and hydrogen content characterization (2 samples from 1 rod) are reported here. The tensile and creep studies are just starting and will be reported at a later date, along with the additional characterization work to be performed. Although only limited prestorage characterization is available, a number of preliminary conclusions can be drawn based on comparison with characterization of Florida Power Turkey Point rods of a similar vintage. Based on this comparison, it appears that little or no cladding thermal creep and fission gas

  17. Examination of Spent PWR Fuel Rods After 15 Years in Dry Storage

    SciTech Connect (OSTI)

    Einziger, R.E.; Tsai, H.C.; Billone, M.C.; Hilton, B.A.

    2002-07-01

    Virginia Power Surry Nuclear Station Pressurized Water Reactor (PWR) fuel was stored in a dry inert atmosphere Castor V/21 cask at the Idaho National Environmental and Engineering Laboratory (INEEL) for 15 years at peak cladding temperatures decreasing from about 350 to 150 deg. C. Prior to the storage, the loaded cask was subjected to extensive thermal benchmark tests. The cask was opened to examine the fuel for degradation and to determine if it was suitable for extended storage. No rod breaches had occurred and no visible degradation or crud/oxide spallation were observed. Twelve rods were removed from the center of the T11 assembly and shipped from INEEL to the Argonne-West HFEF for profilometric scans. Four of these rods were punctured to determine the fission gas release from the fuel matrix and internal pressure in the rods. Three of the four rods were cut into five segments each, then shipped to the Argonne-East AGHCF for detailed examination. The test plan calls for metallographic examination of six samples from two of the rods, microhardness and hydrogen content measurements at or near the six metallographic sample locations, tensile testing of six samples from the two rods, and thermal creep testing of eight samples from the two rods to determine the extent of residual creep life. The results from the profilometry (12 rods), gas release measurements (4 rods), metallographic examinations (2 samples from 1 rod), and microhardness and hydrogen content characterization (2 samples from 1 rod) are reported here. The tensile and creep studies are just starting and will be reported at a later date, along with the additional characterization work to be performed. Although only limited pre-storage characterization is available, a number of preliminary conclusions can be drawn based on comparison with characterization of Florida Power Turkey Point rods of a similar vintage. Based on this comparison, it appears that little or no cladding thermal creep and fission

  18. Continuing Appropriations Act, 2014-- Implementation of Division F, Title I, Title II, and Title III, and Division G, Consolidated and Further Continuing Appropriations Act, 2013, Pub. L. No. 113-6

    Broader source: Energy.gov [DOE]

    Section 101(a)(6) of the Continuing Appropriations Act, 2014, Pub. L No. 113-46, makes appropriations available through January 15, 2014 for continuing projects or activities that were conducted...

  19. Mozambique-Biofuels, Land Access and Rural Livelihoods | Open...

    Open Energy Info (EERE)

    Biofuels, Land Access and Rural Livelihoods Jump to: navigation, search Name Mozambique-Biofuels, Land Access and Rural Livelihoods AgencyCompany Organization International...

  20. Tanzania-Biofuels, Land Access and Rural Livelihoods | Open Energy...

    Open Energy Info (EERE)

    Tanzania-Biofuels, Land Access and Rural Livelihoods Jump to: navigation, search Name Tanzania-Biofuels, Land Access and Rural Livelihoods AgencyCompany Organization...

  1. Farmers Rural Electric Coop Corp | Open Energy Information

    Open Energy Info (EERE)

    Rural Electric Coop Corp Jump to: navigation, search Name: Farmers Rural Electric Coop Corp Place: Kentucky Website: farmersrecc.com Facebook: https:www.facebook.comFarmersRECC...

  2. Jianghe Rural Electricity Development Co Ltd | Open Energy Information

    Open Energy Info (EERE)

    Jianghe Rural Electricity Development Co Ltd Jump to: navigation, search Name: Jianghe Rural Electricity Development Co Ltd Place: Fuzhou, Fujian Province, China Product:...

  3. Electric Cooperatives Channel Solar Resources to Rural American...

    Office of Environmental Management (EM)

    Electric Cooperatives Channel Solar Resources to Rural American Communities Electric Cooperatives Channel Solar Resources to Rural American Communities February 4, 2016 - 12:07pm ...

  4. 2014 HAC Rural Housing Conference: Retool, Rebuild, Renew

    Broader source: Energy.gov [DOE]

    The biennial HAC Rural Housing Conference brings together stakeholders in the field of rural affordable housing from local nonprofits, federal agencies, Congress, state and local governments, and...

  5. Modified Microgrid Concept for Rural Electrification in Africa...

    Open Energy Info (EERE)

    Modified Microgrid Concept for Rural Electrification in Africa Jump to: navigation, search Tool Summary LAUNCH TOOL Name: Modified Microgrid Concept for Rural Electrification in...

  6. USDA- Rural Energy for America Program (REAP) Grants

    Broader source: Energy.gov [DOE]

    The Rural Energy for America Program (REAP) provides financial assistance to agricultural producers and rural small businesses in America to purchase, install, and construct renewable energy syst...

  7. Climate Change: building the resilience of poor rural communities...

    Open Energy Info (EERE)

    of poor rural communities Jump to: navigation, search Tool Summary LAUNCH TOOL Name: Climate Change: building the resilience of poor rural communities AgencyCompany...

  8. The Gut Microbiota of Rural Papua New Guineans: Composition,...

    Office of Scientific and Technical Information (OSTI)

    The Gut Microbiota of Rural Papua New Guineans: Composition, Diversity Patterns, and Ecological Processes Prev Next Title: The Gut Microbiota of Rural Papua New Guineans: ...

  9. The Other 15%: Expanding Energy Efficiency to Rural Populations...

    Energy Savers [EERE]

    The Other 15%: Expanding Energy Efficiency to Rural Populations The Other 15%: Expanding Energy Efficiency to Rural Populations Better Buildings Residential Network Peer Exchange ...

  10. Rural Electrification Act of 1936 | Open Energy Information

    Open Energy Info (EERE)

    901 et seq. DOI Not Provided Check for DOI availability: http:crossref.org Online Internet link for Rural Electrification Act of 1936 Citation Rural Electrification Act of...

  11. The stress corrosion cracking behavior of alloys 690 and 152 WELD in a PWR environment.

    SciTech Connect (OSTI)

    Alexandreanu, B.; Chopra, O. K.; Shack, W. J.

    2009-01-01

    Alloys 690 and 152 are the replacement materials of choice for Alloys 600 and 182, respectively. The latter two alloys are used as structural materials in pressurized water reactors (PWRs) and have been found to undergo stress corrosion cracking (SCC). The objective of this work is to determine the crack growth rates (CGRs) in a simulated PWR water environment for the replacement alloys. The study involved Alloy 690 cold-rolled by 26% and a laboratory-prepared Alloy 152 double-J weld in the as-welded condition. The experimental approach involved pre-cracking in a primary water environment and monitoring the cyclic CGRs to determine the optimum conditions for transitioning from the fatigue transgranular to intergranular SCC fracture mode. The cyclic CGRs of cold-rolled Alloy 690 showed significant environmental enhancement, while those for Alloy 152 were minimal. Both materials exhibited SCC of 10{sup -11} m/s under constant loading at moderate stress intensity factors. The paper also presents tensile property data for Alloy 690TT and Alloy 152 weld in the temperature range 25--870 C.

  12. TREAT source-term experiment STEP-1 simulating a PWR LOCA

    SciTech Connect (OSTI)

    Simms, R.; Baker, L. Jr.; Blomquist, C.A.; Ritzman, R.L.

    1986-01-01

    In a hypothetical pressurized water reactor (PWR) large-break loss-of-coolant accident (LOCA) in which the emergency core cooling system fails, fission product decay heating causes water boil-off and reduced heat removal. Zircaloy cladding is oxidized by the steam. The noble gases and volatile fission products such as cesium and iodine that constitute a principal part of the source term will be released from the damaged fuel at or shortly after the time of cladding failure. TREAT test STEP-1 simulated the LOCA environment when the volatile fission products would be released using four fuel elements from the Belgonucleaire BR3 reactor. The principal objective was to collect a portion of the releases carried by the flow stream in a region as close as possible to the test zone. In this paper, the test is described and the results of an analysis of the thermal and steam/hydrogen environment are compared with the test measurements in order to provide a characterization for analysis of fission product releases and aerosol formation. The results of extensive sample examinations are reported separately.

  13. LBB evaluation for a typical Japanese PWR primary loop by using the US NRC approved methods

    SciTech Connect (OSTI)

    Swamy, S.A.; Bhowmick, D.C.; Prager, D.E.

    1997-04-01

    The regulatory requirements for postulated pipe ruptures have changed significantly since the first nuclear plants were designed. The Leak-Before-Break (LBB) methodology is now accepted as a technically justifiable approach for eliminating postulation of double-ended guillotine breaks (DEGB) in high energy piping systems. The previous pipe rupture design requirements for nuclear power plant applications are responsible for all the numerous and massive pipe whip restraints and jet shields installed for each plant. This results in significant plant congestion, increased labor costs and radiation dosage for normal maintenance and inspection. Also the restraints increase the probability of interference between the piping and supporting structures during plant heatup, thereby potentially impacting overall plant reliability. The LBB approach to eliminate postulating ruptures in high energy piping systems is a significant improvement to former regulatory methodologies, and therefore, the LBB approach to design is gaining worldwide acceptance. However, the methods and criteria for LBB evaluation depend upon the policy of individual country and significant effort continues towards accomplishing uniformity on a global basis. In this paper the historical development of the U.S. LBB criteria will be traced and the results of an LBB evaluation for a typical Japanese PWR primary loop applying U.S. NRC approved methods will be presented. In addition, another approach using the Japanese LBB criteria will be shown and compared with the U.S. criteria. The comparison will be highlighted in this paper with detailed discussion.

  14. Alaska Rural Manager Panelists Call for Nominations

    Broader source: Energy.gov [DOE]

    The Alaska Rural Managers are seeking nominations for city, tribal, and utility managers to participate in several Anchorage focus group/workshops this April. Selected panelists will represent their profession and will help develop guidelines for the training and education of Alaska's Rural Managers.

  15. CEPAN method of analyzing creep collapse of oval cladding. Volume 5. Evaluation of interpellet gap formation and clad collapse in modern PWR fuel rods

    SciTech Connect (OSTI)

    Adams, W.M.; Fisher, H.D.; Litke, H.J.; Mordarski, W.J.

    1985-04-01

    This report presents the results from a review of interpellet-gap formation, ovality, creepdown and clad collapse data in modern PWR fuel rods. Conclusions are reached regarding the propensity of modern PWR fuel to form such gaps and to undergo clad collapse. CEPAN, a creep-collapse predictor approved by the NRC in 1976, has been reformulated to include the creep analysis of cladding with finite interpellet gaps. The basis for this reformulation is discussed in detail. The model previously used in the calculation of the augmentation factor, a peak linear heat rate penalty due to the presence of interpellet gaps within the fuel rod, has been modified to incorporate gap-formation statistics from modern fuel. Finnally, the benefits of the limited gap formation and the CEPAN reformulation for the licensing of modern PWR fuel rods are evaluated.

  16. Analyses of High Pressure Molten Debris Dispersion for a Typical PWR Plant

    SciTech Connect (OSTI)

    Osamu KAawabata; Mitsuhiro Kajimoto [Japan Nuclear Energy Safety Organization (Japan)

    2006-07-01

    In such severe core damage accident, as small LOCAs with no ECCS injection or station blackout, in which the primary reactor system remains pressurized during core melt down, certain modes of vessel failure would lead to a high pressure ejection of molten core material. In case of a local failure of the lower head, the molten materials would initially be ejected into the cavity beneath the pressure vessel may subsequently be swept out from the cavity to the containment atmosphere and it might cause the early containment failure by direct contact of containment steel liner with core debris. When the contribution of a high-pressure scenario in a core damage frequency increases, early conditional containment failure probability may become large. In the present study, the verification analysis of PHOENICS code and the combining analysis with MELCOR and PHOENICS codes were performed to examine the debris dispersion behavior during high pressure melt ejection. The PHOENICS code which can treat thermal hydraulic phenomena, was applied to the verification analysis for melt dispersion experiments conducted by the Purdue university in the United States. A low pressure melt dispersion experiment at initial pressure 1.4 MPas used metal woods as a molten material was simulated. The analytical results with molten debris dispersion mostly from the model reactor cavity compartment showed an agreement with the experimental result, but the analysis result of a volumetric median diameter of the airborne debris droplets was estimated about 1.5 times of the experimental result. The injection rates of molten debris and steam after reactor vessel failure for a typical PWR plant were analyzed using the MELCOR code. In addition, PHOENICS was applied to a 3D analysis for debris dispersion with low primary pressure at the reactor vessel failure. The analysis result showed that almost all the molten debris were dispersed from the reactor vessel cavity compartment by about 45 seconds after the

  17. On-line PWR RHR pump performance testing following motor and impeller replacement

    SciTech Connect (OSTI)

    DiMarzo, J.T.

    1996-12-01

    On-line maintenance and replacement of safety-related pumps requires the performance of an inservice test to determine and confirm the operational readiness of the pumps. In 1995, major maintenance was performed on two Pressurized Water Reactor (PWR) Residual Heat Removal (RHR) Pumps. A refurbished spare motor was overhauled with a new mechanical seal, new motor bearings and equipped with pump`s `B` impeller. The spare was installed into the `B` train. The motor had never been run in the system before. A pump performance test was developed to verify it`s operational readiness and determine the in-situ pump performance curve. Since the unit was operating, emphasis was placed on conducting a highly accurate pump performance test that would ensure that it satisfied the NSSS vendors accident analysis minimum acceptance curve. The design of the RHR System allowed testing of one train while the other was aligned for normal operation. A test flow path was established from the Refueling Water Storage Tank (RWST) through the pump (under test) and back to the RWST. This allowed staff to conduct a full flow range pump performance test. Each train was analyzed and an expression developed that included an error vector term for the TDH (ft), pressure (psig), and flow rate (gpm) using the variance error vector methodology. This method allowed the engineers to select a test instrumentation system that would yield accurate readings and minimal measurement errors, for data taken in the measurement of TDH (P,Q) versus Pump Flow Rate (Q). Test results for the `B` Train showed performance well in excess of the minimum required. The motor that was originally in the `B` train was similarly overhauled and equipped with `A` pump`s original impeller, re-installed in the `A` train, and tested. Analysis of the `A` train results indicate that the RHR pump`s performance was also well in excess of the vendors requirements.

  18. Radionuclide release from PWR fuels in a reference tuff repository groundwater

    SciTech Connect (OSTI)

    Wilson, C.N.; Oversby, V.M.

    1985-03-01

    The Nevada Nuclear Waste Storage Investigations Project (NNWSI) is studying the suitability of the welded devitrified Topopah Spring tuff at Yucca Mountain, Nye County, Nevada, for potential use as a high-level nuclear waste repository. In support of the Waste Package task of NNWSI, tests have been conducted under ambient air environment to measure radionuclide release from two pressurized water reactor (PWR) spent fuels in water obtained from the J-13 well near the Yucca Mountain site. Four specimen types, representing a range of fuel physical conditions that may exist in a failed waste canister containing a limited amount of water were tested. The specimen types were: fuel rod sections split open to expose bare fuel particles; rod sections with water-tight end fittings with a 2.5-cm long by 150-{mu}m wide slit through the cladding; rod sections with water-tight end fittings and two 200-{mu}m-diameter holes through the cladding; and undefected rod segments with water-tight end fittings. Radionuclide release results from the first 223-day test runs on H.B. Robinson spent fuel specimens in J-13 water are reported and compared to results from a previous test series in which similar Turkey Point reactor spent fuel specimens were tested on deionized water. Selected initial results are also given for Turkey Point fuel specimens tested on J-13 water. Results suggest that the actinides Pu, Am, Cm and Np are released congruently with U as the UO{sub 2} spent fuel matrix dissolves. Fractional release of {sup 137}Cs and {sup 99}Tc was greater than that measured for the actinides. Generally, lower radionuclide releases were measured for the H.B. Robinson fuel in J-13 water than for Turkey Point Fuel in deionized water. 8 references, 7 figures, 9 tables.

  19. Severe accident modeling of a PWR core with different cladding materials

    SciTech Connect (OSTI)

    Johnson, S. C.; Henry, R. E.; Paik, C. Y.

    2012-07-01

    The MAAP v.4 software has been used to model two severe accident scenarios in nuclear power reactors with three different materials as fuel cladding. The TMI-2 severe accident was modeled with Zircaloy-2 and SiC as clad material and a SBO accident in a Zion-like, 4-loop, Westinghouse PWR was modeled with Zircaloy-2, SiC, and 304 stainless steel as clad material. TMI-2 modeling results indicate that lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would result if SiC was substituted for Zircaloy-2 as cladding. SBO modeling results indicate that the calculated time to RCS rupture would increase by approximately 20 minutes if SiC was substituted for Zircaloy-2. Additionally, when an extended SBO accident (RCS creep rupture failure disabled) was modeled, significantly lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would be generated by substituting SiC for Zircaloy-2 or stainless steel cladding. Because the rate of SiC oxidation reaction with elevated temperature H{sub 2}O (g) was set to 0 for this work, these results should be considered preliminary. However, the benefits of SiC as a more accident tolerant clad material have been shown and additional investigation of SiC as an LWR core material are warranted, specifically investigations of the oxidation kinetics of SiC in H{sub 2}O (g) over the range of temperatures and pressures relevant to severe accidents in LWR 's. (authors)

  20. A safety and regulatory assessment of generic BWR and PWR permanently shutdown nuclear power plants

    SciTech Connect (OSTI)

    Travis, R.J.; Davis, R.E.; Grove, E.J.; Azarm, M.A.

    1997-08-01

    The long-term availability of less expensive power and the increasing plant modification and maintenance costs have caused some utilities to re-examine the economics of nuclear power. As a result, several utilities have opted to permanently shutdown their plants. Each licensee of these permanently shutdown (PSD) plants has submitted plant-specific exemption requests for those regulations that they believe are no longer applicable to their facility. This report presents a regulatory assessment for generic BWR and PWR plants that have permanently ceased operation in support of NRC rulemaking activities in this area. After the reactor vessel is defueled, the traditional accident sequences that dominate the operating plant risk are no longer applicable. The remaining source of public risk is associated with the accidents that involve the spent fuel. Previous studies have indicated that complete spent fuel pool drainage is an accident of potential concern. Certain combinations of spent fuel storage configurations and decay times, could cause freshly discharged fuel assemblies to self heat to a temperature where the self sustained oxidation of the zircaloy fuel cladding may cause cladding failure. This study has defined four spent fuel configurations which encompass all of the anticipated spent fuel characteristics and storage modes following permanent shutdown. A representative accident sequence was chosen for each configuration. Consequence analyses were performed using these sequences to estimate onsite and boundary doses, population doses and economic costs. A list of candidate regulations was identified from a screening of 10 CFR Parts 0 to 199. The continued applicability of each regulation was assessed within the context of each spent fuel storage configuration and the results of the consequence analyses.

  1. Using Clinical Data, Hypothesis Generation Tools and PubMed Trends to Discover the Association between Diabetic Retinopathy and Antihypertensive Drugs

    SciTech Connect (OSTI)

    Senter, Katherine G; Sukumar, Sreenivas R; Patton, Robert M; Chaum, Ed

    2015-01-01

    Diabetic retinopathy (DR) is a leading cause of blindness and common complication of diabetes. Many diabetic patients take antihypertensive drugs to prevent cardiovascular problems, but these drugs may have unintended consequences on eyesight. Six common classes of antihypertensive drug are angiotensin converting enzyme (ACE) inhibitors, alpha blockers, angiotensin receptor blockers (ARBs), -blockers, calcium channel blockers, and diuretics. Analysis of medical history data might indicate which of these drugs provide safe blood pressure control, and a literature review is often used to guide such analyses. Beyond manual reading of relevant publications, we sought to identify quantitative trends in literature from the biomedical database PubMed to compare with quantitative trends in the clinical data. By recording and analyzing PubMed search results, we found wide variation in the prevalence of each antihypertensive drug in DR literature. Drug classes developed more recently such as ACE inhibitors and ARBs were most prevalent. We also identified instances of change-over-time in publication patterns. We then compared these literature trends to a dataset of 500 diabetic patients from the UT Hamilton Eye Institute. Data for each patient included class of antihypertensive drug, presence and severity of DR. Graphical comparison revealed that older drug classes such as diuretics, calcium channel blockers, and -blockers were much more prevalent in the clinical data than in the DR and antihypertensive literature. Finally, quantitative analysis of the dataset revealed that patients taking -blockers were statistically more likely to have DR than patients taking other medications, controlling for presence of hypertension and year of diabetes onset. This finding was concerning given the prevalence of -blockers in the clinical data. We determined that clinical use of -blockers should be minimized in diabetic patients to prevent retinal damage.

  2. Containment pressurization and burning of combustible gases in a large, dry PWR containment during a station blackout sequence

    SciTech Connect (OSTI)

    Lee, M.; Fan, C.T. (National Tsing-Hua Univ., Dept. of Nuclear Engineering, Hsinchu (TW))

    1992-07-01

    In this paper, responses of a large, dry pressurized water reactor (PWR) containment in a station blackout sequence are analyzed with the CONTAIN, MARCH3, and MAAP codes. Results show that the predicted containment responses in a station blackout sequence of these three codes are substantially different. Among these predictions, the MAAP code predicts the highest containment pressure because of the large amount of water made available to quench the debris upon vessel failure. The gradual water boiloff by debris pressurizes the containment. The combustible gas burning models in these codes are briefly described and compared.

  3. In-pile post-DNB behavior of a nine-rod PWR-type fuel bundle

    SciTech Connect (OSTI)

    Gunnerson, F.S.; MacDonald, P.E.

    1980-01-01

    The results of an in-pile power-cooling-mismatch (PCM) test designed to investigate the behavior of a nine-rod, PWR-type fuel bundle under intermittent and sustained periods of high temperature film boiling operation are presented. Primary emphasis is placed on the DNB and post-DNB events including rod-to-rod interactions, return to nucleate boiling (RNB), and fuel rod failure. A comparison of the DNB behavior of the individual bundle rods with single-rod data obtained from previous PCM tests is also made.

  4. Rural electric cooperatives IRP survey

    SciTech Connect (OSTI)

    Garrick, C.

    1995-11-01

    This report summarizes the integrated resource planning (IRP) practices of US rural electric cooperatives and the IRP policies which influence these practices. It was prepared by the National Renewable Energy Laboratory (NREL) and its subcontractor Garrick and Associates to assist the US Department of Energy (DOE) in satisfying the reporting requirements of Title 1, Subtitle B, Section 111(e)(3) of the Energy Policy Act of 1992 (EPAct), which states: (e) Report--Not later than 2 years after the date of the enactment of this Act, the Secretary (of the US Department of Energy) shall transmit a report to the President and to the Congress containing--(the findings from several surveys and evaluations, including:); (3) a survey of practices and policies under which electric cooperatives prepare IRPs, submit such plans to REA, and the extent to which such integrated resource planning is reflected in rates charged to customers.

  5. Sustainable Energy Solutions for Rural Alaska | Department of Energy

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

    Sustainable Energy Solutions for Rural Alaska Sustainable Energy Solutions for Rural Alaska Photo of the Sustainable Energy Solutions for Rural Alaska report. The report, "Sustainable Energy Solutions for Rural Alaska," provides recommendations from a study conducted over the course of 18 months that involved in-person interviews with utility staff and community members from more than 30 Alaska rural communities. The purpose of the study was to understand the current challenges and

  6. Rural energy - ODA`s perspective

    SciTech Connect (OSTI)

    Woolnough, D.

    1997-12-01

    The Overseas Development Administration has as a goal `to improve the quality of life of people in poorer countries by contributing to sustainable development and reducing poverty and suffering.` Rural energy fits into this goal as a means to an end. The emphasis is firmly on the service provided, with the aim being provision of basic needs as a part of rural development. ODA plays a role in this task on a number of fronts: research and development; support for NGO`s; aid in a bilateral or multilateral form. The view of ODA is that even rural energy projects must emphasize the service provided and must be economically sustainable. Within its sphere of influence, there is a clearly growing position for the employment of rural energy programs.

  7. Alaska Rural Energy Conference- Federal Energy Track

    Broader source: Energy.gov [DOE]

    On the first day of the Alaska Rural Energy Conference, the U.S. Department of Energy (DOE) Office of Indian Energy is hosting a federal energy track to cover federal programs and opportunities for Alaska Native villages.

  8. USDA Rural Small Business Connection Event

    Broader source: Energy.gov [DOE]

    The U.S. Department of Agriculture (USDA) will host Rural Small Business Connections, providing small businesses with networking sessions and opportunities to build capacity and do business with USDA and other Federal agencies.

  9. BABAR-PUB-14/002 SLAC-PUB-15979

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

    ab , 55 G. Piredda a , 55 C. B unger, 56 S. Dittrich, 56 O. Gr unberg, 56 T. Hartmann, 56 M. Hess, 56 T. Leddig, 56 C. Vo, 56 R. Waldi, 56 T. Adye, 57 E. O. Olaiya, 57...

  10. LOFTRAN/RETRAN comparison calculations for a postulated loss-of-feedwater ATWS in the Sizewell 'B' PWR

    SciTech Connect (OSTI)

    Papez, K.L.; Risher, D.H.

    1983-05-01

    The loss-of-main-feedwater transient without reactor trip (scram) has received particular attention in pressurized water reactor (PWR) anticipated transient without scram (ATWS) analysis primarily due to the potential for reactor coolant system over pressurization. To assist in the licensing of the U.K. PWR, Sizewell 'B', comparative calculations of a loss-of-feedwater ATWS have been performed using the Westinghouse-developed LOFTRAN loop analysis code and the Electric Power Research Institute/ Energy Incorporated-developed RETRAN-01 code. The calculations were performed with and without the emergency boration system (EBS), which is included in the Sizewell reference design. Initial results showed good agreement between the codes for the major features of the transient, but also a time shift in the transient profiles at the time of the pressurizer pressure peak. This was found to be due to differences in the steam generator modeling, which resulted in a difference in the onset of the very rapid degradation in heat transfer as the steam generators approach dryout. When the same model was used in both codes, very good agreement was obtained. Remaining differences in the results are attributed primarily to differences in the boron injection models, which resulted in an over-prediction of the core boron concentration in the RETRAN calculation. The results with an EBS indicate that the peak pressurizer pressure is relatively insensitive to variations in modeling.

  11. First interim examination of defected BWR and PWR rods tested in unlimited air at 229/sup 0/C

    SciTech Connect (OSTI)

    Einziger, R.E.; Cook, J.A.

    1983-01-01

    A five-year whole rod test was initiated to investigate the long-term stability of spent fuel rods under a variety of possible dry storage conditions. Both PWR and BWR rods were included in the test. The first interim examination was conducted after three months of testing to determine if there was any degradation in those defected rods stored in an unlimited air atmosphere. Visual observations, diametral measurements and radiographic smears were used to assess the degree of cladding deformation and particulate dispersal. The PWR rod showed no measurable change from the pre-test condition. The two original artificial defects had not changed in appearance and there was no diametral growth of the cladding. One of the defects in BWR rod showed significant deformation. There was approximately 10% cladding strain at the defect site and a small axial crack had formed. The fuel in the defect did not appear to be friable. The second defect showed no visible change and no cladding strain. Following examination, the test was continued at 230/sup 0/C. Another interim examination is planned during the summer of 1983. This paper discusses the details and meaning of the data from the first interim examination.

  12. Experiment data report for LOFT large-break loss-of-coolant experiment L2-5. [PWR

    SciTech Connect (OSTI)

    Bayless, P.D.; Divine, J.M.

    1982-08-01

    Selected pertinent and uninterpreted data from the third nuclear large break loss-of-coolant experiment (Experiment L2-5) conducted in the Loss-of-Fluid Test (LOFT) facility are presented. The LOFT facility is a 50-MW(t) pressurized water reactor (PWR) system with instruments that measure and provide data on the system thermal-hydraulic and nuclear conditions. The operation of the LOFT system is typical of large (approx. 1000 MW(e)) commercial PWR operations. Experiment L2-5 simulated a double-ended offset shear of a cold leg in the primary coolant system. The primary coolant pumps were tripped within 1 s after the break initiation, simulating a loss of site power. Consistent with the loss of power, the starting of the high- and low-pressure injection systems was delayed. The peak fuel rod cladding temperature achieved was 1078 +- 13 K. The emergency core cooling system re-covered the core and quenched the cladding. No evidence of core damage was detected.

  13. Impact of fuel cladding failure events on occupational radiation exposures at nuclear power plants. Case study: PWR during routine operations

    SciTech Connect (OSTI)

    Moeller, M.P.; Martin, G.F.; Haggard, D.L.

    1986-01-01

    The purpose of this report is to present data in support of evaluating the impact of fuel cladding failure events on occupational radiation exposure. To determine quantitatively whether fuel cladding failure contributes significantly to occupational radiation exposure, radiation exposure measurements were taken at comparable locations in two mirror-image pressurized-water reactors (PWRs) and their common auxiliary building. One reactor, Unit B, was experiencing degraded fuel characterized as 0.125% fuel pin-hole leakers and was operating at approximately 55% of the reactor's licensed maximum core power, while the other reactor, Unit A, was operating under normal conditions with less than 0.01% fuel pin-hole leakers at 100% of the reactor's licensed maximum core power. Measurements consisted of gamma spectral analyses, radiation exposure rates and airborne radionuclide concentrations. In addition, data from primary coolant sample results for the previous 20 months on both reactor coolant systems were analyzed. The results of the measurements and coolant sample analyses suggest that a 3560-megawatt-thermal (1100 MWe) PWR operating at full power with 0.125% failed fuel can experience an increase of 540% in radiation exposure rates as compared to a PWR operating with normal fuel. In specific plant areas, the degraded fuel may elevate radiation exposure rates even more.

  14. Rural Productivity Zones (RPZs) for microenterprises

    SciTech Connect (OSTI)

    Hansen, R.D.

    1997-12-01

    In this paper the authors discuss the concept of rural productivity zones (RPZs) which are defined as a business incubator to foster income-producing opportunities for the rural poor. The essential ingredients of such a program include: electric power; business development assistance; office services; and quality work space. The electric power source must be a good quality system, consisting of a diesel/wind/photovoltaic hybrid type system, providing reliable service, with a local maintenance program and a functional load management program.

  15. Continuing Appropriations Resolution, 2013-- Implementation of Division B, Title III, Title V and Division C Title VII, Consolidated Appropriations Act, 2012, Pub. L. No.112-74 and Related Conference Report

    Broader source: Energy.gov [DOE]

    Section 101(a)(4) of the Continuing Appropriations Resolution, 2013, Pub. L No. 112-175, makes appropriations available through March 27, 2013 for continuing projects or activities that were conducted under Consolidated Appropriations Act, 2012, Pub. L. No.112-74, under the same authority and conditions. Section 104 provides that no appropriation under the continuing resolution shall be used to initiate or resume any project or activity for which appropriations, funds, or other authority were not available during fiscal year 2012. Therefore, the Acquisition Letter (AL) 2012-08 and Financial Assistance Letter (FAL) 2012-01 providing implementing instructions and guidance for Division B, Title III, Title V and Division C Title VII, Consolidated Appropriations Act, 2012, Pub. L. No.112-74 and Related Conference Report remains in effect until rescinded or superseded.

  16. SLAC-PUB-15193 Compact

    Office of Scientific and Technical Information (OSTI)

    10 decompression (black), and for a transverse gradient undulator without decompression (red). Another method to reduce the gain length of a large energy spread beam is by decom-...

  17. SLAC-PUB-15178 July

    Office of Scientific and Technical Information (OSTI)

    D. Zep- penfeld, Phys. Rev. D 70, 113009 (2004) hep-ph0406323, hep-ph0407190. 7 R. Lafaye, T. Plehn, M. Rauch, D. Zerwas and M. D uhrssen, JHEP 08 (2009) 009...

  18. SLAC-PUB-15416 April

    Office of Scientific and Technical Information (OSTI)

    Accelerator Laboratory Stanford University Menlo Park, CA 94025, USA Frank Zimmermann CERN, Geneva, Switzerland Abstract A ring-based Higgs factory with a center-of-mass energy...

  19. SLAC-PUB-15224 August

    Office of Scientific and Technical Information (OSTI)

    24 August 25, 2012 Theoretical Summary Lecture for Higgs Hunting 2012 Michael E. Peskin 1 SLAC, Stanford University, Menlo Park, California 94025 USA ABSTRACT In this lecture, I...

  20. Newsletter Dec draft 120211.pub

    Energy Savers [EERE]

    About Us » Newsletters » Indian Energy Beat » Newsletter Archives Newsletter Archives The Office of Indian Energy's Indian Energy Beat newsletter highlights opportunities and actions to accelerate energy development in Indian Country. View past newsletters below, or read the current issue. October 22, 2015 Office of Indian Energy Fall 2015 Newsletter The Office of Indian Energy Indian Energy Beat Fall 2015 newsletter highlights opportunities and actions to accelerate energy development in

  1. Newsletter Jan draft 010512.pub

    Energy Savers [EERE]

    Green Holidays Fuel Cell Lights Up Festivities at Secretary Chu's Holiday Party For the first time ever, DOE's annual holiday party featured lights powered by a clean, efficient fuel cell. The portable fuel cell, made by Trulite, generates roughly 150 watts of power and is a clean energy solution for recharging laptops, cell phones, and other everyday low-energy demand appliances. The use of a clean, renewable technology at DOE's headquarters was appreciated by DOE employees. "It's great to

  2. Newsletter June draft 060512.pub

    Energy Savers [EERE]

    Oak Ridge Na onal Laboratory hosted its second annual regional Sustainability Summit on May 2-3, 2012. Over 110 a endees represented ci es, coun es, states, universi es, industry, power providers and DOE labs. The Summit covered three major subject areas: low- carbon power genera on, sustainable transporta on, and energy efficiency in buildings. Addi onal informa on on ORNL's Sustainable Campus Ini a ve may be found at h p://sustainability-ornl.org. Send us your champion! SUSTAINABILITY

  3. Newsletter Mar draft 030612.pub

    Energy Savers [EERE]

    STORIES Retrofits under way at LANL's Occupational Medicine Building Los Alamos NaƟonal Laboratory (LANL) is saving 764 million BTUs from steam and electricity and $11,900 in costs each year at the "Occ Med" building, one of LANL's Įrst High Performance Sustainable Buildings (HPSB). The LANL HPSB Team selected 31 buildings for reͲ commissioning to review energy usage and implement eĸciency improvements and system upgrades. Improvements include: establishing a nightͲsetback

  4. Newsletter May Final 050112.pub

    Energy Savers [EERE]

    Send us your champion! Two members of DOE were recently honored at the Federal 100 Awards which recognize government and industry leaders in the Federal IT community . DOE CIO, Michael Loca s, was honored for his impact on the collabora ve cybersecurity approach he has taken across the department and throughout the federal government, in par cular the crea on of the Joint Cybersecurity Coordina on Center. Peter Tseronis, Chief Technology Office for the CIO won for his keen eye on ensuring that

  5. Newsletter Oct draft 100611.pub

    Energy Savers [EERE]

    Reporting Milestones Metering Per DOE's 2010 SSPP, DOE sites are submitting new or updated metering plans. Best practices include:  Analysis of practicality and cost-effectiveness;  Summary tables of scheduled metering by building;  Plans on how to use the information generated and;  Financial plans, such as the use of ESPC, third party funding, and internal funds. DOE's Forrestal Building installed electrical sub-meters to measure ligh ng and plug-load electricity

  6. SLAC-PUB-15832 November

    Office of Scientific and Technical Information (OSTI)

    AdSCFT correspondence is a powerful tool for computing observables in strongly coupled systems with conformal symmetry by mapping them to weakly coupled dual gravitational...

  7. LicWk3.pub

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

    LICENSE APPLICATION GUIDELINES- 1 License Application Guidelines We're pleased you've chosen to submit a request for a Savannah River Nuclear Solutions license. We look forward to helping you launch a new product. Our License Application comes in two parts. You may submit the Intent to License first, followed by the more detailed Business Plan. Or, you may submit both at the same time. The following are guidelines, which may or may not apply in all circumstances, describing the in- formation we

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

    Office of Energy Efficiency and Renewable Energy (EERE)

    Results of testing employing surrogate instrumented rods (non-high-burnup, 17 x 17 PWR fuel assembly) to capture the response to the loadings experienced during normal conditions of transport indicate that strain- or stress-based failure of fuel rods seems unlikely; performance of high-burnup fuels continues to be assessed.

  9. PWR core design, neutronics evaluation and fuel cycle analysis for thorium-uranium breeding recycle

    SciTech Connect (OSTI)

    Bi, G.; Liu, C.; Si, S.

    2012-07-01

    This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade {sup 233}U-Thorium (U{sub 3}ThOX) FAs on the basis of reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade {sup 233}U extracted from burnt PuThOX fuel was used to fabrication of U{sub 3}ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U{sub 3}ThOX mixed core, the well designed U{sub 3}ThOX FAs with 1.94 w/o fissile uranium (mainly {sup 233}U) were located on the periphery of core as a blanket region. U{sub 3}ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U{sub 3}ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U{sub 3}ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U{sub 3}ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U{sub 3}ThOX loading on the periphery of core has no

  10. Harrison Rural Elec Assn, Inc | Open Energy Information

    Open Energy Info (EERE)

    Harrison Rural Elec Assn, Inc Jump to: navigation, search Name: Harrison Rural Elec Assn, Inc Place: West Virginia Phone Number: 304.624.6365 Website: www.harrisonrea.com...