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Sample records for northern states pwr

  1. PP-45 Northern States Power Company | Department of Energy

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

    Northern States Power Company PP-45 Northern States Power Company Presidential Permit ... PDF icon PP-45 Northern States Power Company More Documents & Publications PP-45-2 ...

  2. EA-63-C Northern States Power | Department of Energy

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

    3-C Northern States Power EA-63-C Northern States Power Order authorizing Northern States Power to export electric energy to Canada PDF icon EA-63-C Northern States Power More ...

  3. Northern States Metals Company | Open Energy Information

    Open Energy Info (EERE)

    Jump to: navigation, search Name: Northern States Metals Company Address: 3207 Innovation Place Place: Youngstown, Ohio Zip: 44509 Sector: Solar Product: Manufacturing Phone...

  4. PP-63 Northern States Power Company (NSP) | Department of Energy

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

    Northern States Power Company (NSP) PP-63 Northern States Power Company (NSP) ... at the U.S. - Canada Border. PDF icon PP-63 Northern States Power Company (NSP) More ...

  5. PP-231 Northern States Power Company | Department of Energy

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

    Northern States Power Company to construct, operate, and maintain electric transmission facilities at the U.S-Canadian border. PDF icon PP-231 Northern States Power Company More ...

  6. EA-282 Northern States Power Company | Department of Energy

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

    energy to Canada. PDF icon EA-282 Northern States Power Company More Documents & Publications EA-63-C Northern States Power EA-196-B Minnesota Power EA-196-C Minnesota Power

  7. Joint Motion to Intervene of Northern States Power Company (Minnesota...

    Energy Savers [EERE]

    States Power Company (Minnesota) et al. on the Proposed Open Access Requirements Joint Motion to Intervene of Northern States Power Company (Minnesota) et al. on the Proposed ...

  8. PP-63-3 Northern States Power Company (NSP) | Department of Energy

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

    3 Northern States Power Company (NSP) More Documents & Publications PP-63 Northern States Power Company (NSP) PP-63-4

  9. PP-45-2 Northern States Power Company

    Broader source: Energy.gov [DOE]

    Presidential Permit authorizing Northern States Power Company to construct, operate, and maintain elextric transmission facilities at the U.S. - Canada Border.

  10. PP-45-1 Northern States Power Company

    Broader source: Energy.gov [DOE]

    Presidential Permit authorizing Northern States Power Company to construct, operate, and maintain elextric transmission facilities at the U.S. - Canada Border.

  11. Northern Mariana Islands Recovery Act State Memo | Department...

    Office of Environmental Management (EM)

    Northern Mariana Islands Recovery Act State Memo The American Recovery & Reinvestment Act (ARRA) is making a meaningful down payment on the nation's energy and environmental ...

  12. PRESIDENTIAL PERMIT NORTHERN STATES POWER COMPANY ORDER NO. PP-231

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

    NORTHERN STATES POWER COMPANY ORDER NO. PP-231 I. BACKGROUND The Office of Fossil Energy (FE) of the Department of Energy (DOE) has the responsibility for implementing Executive Order (EO) 10485, as amended by EO 12038, which requires the issuance of Presidential permits for the construction, operation, maintenance, and connection of electric transmission facilities at the United States international border. On November 2, 2000, Northern States Power Company (NSP) filed an application with the

  13. PP-63-4 Northern States Power Company (NSP) | Department of Energy

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

    -4 Northern States Power Company (NSP) PP-63-4 Northern States Power Company (NSP) ... at the U.S. - Canada Border. PDF icon PP-63-4 Northern States Power Company (NSP) ...

  14. Joint Motion to Intervene of Northern States Power Company (Minnesota) et

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

    al. on the Proposed Open Access Requirements | Department of Energy Motion to Intervene of Northern States Power Company (Minnesota) et al. on the Proposed Open Access Requirements Joint Motion to Intervene of Northern States Power Company (Minnesota) et al. on the Proposed Open Access Requirements Joint motion to intervene of the Northern States Power Company (Minnesota), the Northern States Power Company (Wisconsin), and NRG Energy, Incl on the Proposed Open Access Requirements for

  15. RELAP-7 Level 2 Milestone Report: Demonstration of a Steady State Single Phase PWR Simulation with RELAP-7

    SciTech Connect (OSTI)

    David Andrs; Ray Berry; Derek Gaston; Richard Martineau; John Peterson; Hongbin Zhang; Haihua Zhao; Ling Zou

    2012-05-01

    The document contains the simulation results of a steady state model PWR problem with the RELAP-7 code. The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on INL's modern scientific software development framework - MOOSE (Multi-Physics Object-Oriented Simulation Environment). This report summarizes the initial results of simulating a model steady-state single phase PWR problem using the current version of the RELAP-7 code. The major purpose of this demonstration simulation is to show that RELAP-7 code can be rapidly developed to simulate single-phase reactor problems. RELAP-7 is a new project started on October 1st, 2011. It will become the main reactor systems simulation toolkit for RISMC (Risk Informed Safety Margin Characterization) and the next generation tool in the RELAP reactor safety/systems analysis application series (the replacement for RELAP5). The key to the success of RELAP-7 is the simultaneous advancement of physical models, numerical methods, and software design while maintaining a solid user perspective. Physical models include both PDEs (Partial Differential Equations) and ODEs (Ordinary Differential Equations) and experimental based closure models. RELAP-7 will eventually utilize well posed governing equations for multiphase flow, which can be strictly verified. Closure models used in RELAP5 and newly developed models will be reviewed and selected to reflect the progress made during the past three decades. RELAP-7 uses modern numerical methods, which allow implicit time integration, higher order schemes in both time and space, and strongly coupled multi-physics simulations. RELAP-7 is written with object oriented programming language C++. Its development follows modern software design paradigms. The code is easy to read, develop, maintain, and couple with other codes. Most importantly, the modern software design allows the RELAP-7 code to

  16. Origin Basin Destination State STB EIA STB EIA Northern Appalachian...

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    Delaware W 28.49 W 131.87 21.6% 59 W 100.0% Northern Appalachian Basin Florida W - - - - - - - Northern Appalachian Basin Indiana W 20.35 W 64.82 31.4% 1,715 W 75.9% Northern...

  17. Origin Basin Destination State STB EIA STB EIA Northern Appalachian...

    Gasoline and Diesel Fuel Update (EIA)

    Florida W 38.51 W 140.84 27.3% 134 W 100.0% Northern Appalachian Basin Georgia - W - W W W - W Northern Appalachian Basin Indiana W 16.14 W 63.35 25.5% 1,681 W 88.5% Northern...

  18. PP-63-1 Northern States Power Company (NSP) | Department of Energy

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

    authorizing Northern States Power Company (NSP) to construct, operate, and maintain transmission facilities at the U.S. - Canada Border PDF icon PP-63-1

  19. PP-63-2 Northern States Power Company (NSP) | Department of Energy

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

    Authorizing Northern States Power Company (NSP) to construct, operate, and maintain transmission facilities at the U.S. - Canada Border. PDF icon PP-63-2

  20. 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...

  1. 1,"Sherburne County","Coal","Northern States Power Co - Minnesota",2222

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

    Minnesota" ,"Plant","Primary energy source","Operating company","Net summer capacity (MW)" 1,"Sherburne County","Coal","Northern States Power Co - Minnesota",2222 2,"Clay Boswell","Coal","ALLETE, Inc.",1082.4 3,"Prairie Island","Nuclear","Northern States Power Co - Minnesota",1040 4,"Monticello Nuclear Facility","Nuclear","Northern States

  2. 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 ...

  3. 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).

  4. 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, ...

  5. Dispersed-flow film boiling in rod-bundle geometry: steady-state heat-transfer data and correlation comparisons. [PWR; BWR

    SciTech Connect (OSTI)

    Yoder, G. L.; Morris, D. G.; Mullins, C. B.; Ott, L. J.; Reed, D. A.

    1982-03-01

    Assessment of six film boiling correlations and one single-phase vapor correlation has been made using data from 22 steady state upflow rod bundle tests (series 3.07.9). Bundle fluid conditions were calculated using energy and mass conservation considerations. Results of the steady state film boiling tests support the conclusions reached in the analysis of prior transient tests 3.03.6AR, 3.06.6B, and 3.08.6C. Comparisons between experimentally determined and correlation-predicted heat transfer coefficients, are presented.

  6. Steady-state axial pressure losses along the exterior of deformed fuel cladding: Multirod Burst Test (MRBT) bundles B-1 and B-2. [PWR; BWR

    SciTech Connect (OSTI)

    Mincey, J.F.

    1980-01-01

    The experimental and COBRA-IV computational data presented in this report confirm that increased pressure losses, induced by the steady-state axial flow of water exterior to deformed Multirod Burst Test (MRBT) bundles B-1 and B-2, may be closely predicted using a bundle-averaged approach for describing flow channel restrictions. One anomaly that was encountered using this technique occurred while modeling the B-2 flow test data near a severe channel restriction: the COBRA-IV results tended to underestimate experimental pressure losses.

  7. The mixing state of carbonaceous aerosol particles in northern and southern California measured during CARES and CalNex 2010

    SciTech Connect (OSTI)

    Cahill, John F.; Suski, Kaitlyn; Seinfeld, John H.; Zaveri, Rahul A.; Prather, Kimberly A.

    2012-11-21

    Carbonaceous aerosols impact climate directly by scattering and absorbing radiation, and hence play a major, although highly uncertain, role in global radiative forcing. Commonly, ambient carbonaceous aerosols are internally mixed with secondary species such as nitrate, sulfate, and ammonium, which influence their climate impacts through optical properties, hygroscopicity, and atmospheric lifetime. Aircraft-aerosol time-of-flight mass spectrometry (A-ATOFMS), which measures single-particle mixing state, was used to determine the fraction of organic and soot aerosols that were internally mixed and the variability of their mixing state in California during the Carbonaceous Aerosol and Radiative Effects Study (CARES) and the Research at the Nexus of Air Quality and Climate Change (CalNex) field campaigns in the late spring and early summer of 2010. Nearly 88% of all A-ATOFMS measured particles (100-1000 nm in diameter) were internally mixed with secondary species, with 96% and 75% of particles internally mixed with nitrate and/or sulfate in southern and northern California, respectively. Even though atmospheric particle composition in both regions was primarily influenced by urban sources, the mixing state was found to vary greatly, with nitrate and soot being the dominant species in southern California, and sulfate and organic carbon in northern California. Furthermore, mixing state varied temporally in northern California, with soot becoming the prevalent particle type towards the end of the study as regional pollution levels increased. The results from these studies demonstrate that the majority of ambient carbonaceous particles are internally mixed and are heavily influenced by secondary species that are most predominant in each region. Based on these findings, considerations of regionally dominant sources and secondary species, as well as temporal variations of aerosol physical and optical properties, will be required to obtain more accurate predictions of the

  8. 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.

  9. EIS-0032: 500 kV International Transmission Line NSP-TR-1, Forbes, Minnesota to Manitoba, Canada, Northern States Power Company

    Broader source: Energy.gov [DOE]

    The Economic Regulatory Administration developed this EIS to evaluate the environmental impacts of a 500-kilovolt transmission line proposed by the Northern States Power Company to provide a transmission facility for the exchange of electrical energy between Canada and the United States.

  10. 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...

  11. 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...

  12. 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...

  13. 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:...

  14. 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...

  15. 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...

  16. 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)

  17. 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

  18. 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.

  19. 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.

  20. 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...

  1. 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 --...

  2. 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...

  3. 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 ...

  4. 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)...

  5. 1993index.PDF

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

    Northern States Pwr Co (Minn) ORD874.FE 102993 93-114-NG Northern States Pwr Co (Wisc) ORD874.FE 102993 93-114-NG Peoples Natural Gas Co, Div of Utilicorp United, Inc ---...

  6. 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

  7. 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.

  8. 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...

  9. 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; ...

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. Application for Presidential Permit OE Docket No. PP-371 Northern...

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

    State Senator Jeanie Forrester Application for Presidential Permit OE Docket No. PP-371 Northern Pass Transmission: Comments from State Senator Jeanie Forrester Application from ...

  15. 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.

  16. 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.

  17. 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.

  18. 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

  19. 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

  20. 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.

  1. 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)

  2. 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

  3. Best-Estimate Analysis PWR LOCA.

    Energy Science and Technology Software Center (OSTI)

    2005-11-11

    Version: 00 TRAC‑PF1 performs best estimate analyses of loss of coolant accidents and other transients in pressurized light water reactors. The program can also be used to model a wide range of thermal hydraulic experiments in reduced scale facilities. Models employed include reflood, multi‑dimensional two‑phase flow, nonequilibrium thermodynamics, generalized heat transfer, and reactor kinetics. Automatic steady‑state and dump/restart capabilities are provided. The changes reported in TRACNEWS issues through Number 7 are incorporated in this release.more » TRAC-PF1 was developed on a CDC computer at Los Alamos National Laboratory. The PC version of TRAC‑PF1 was converted at CNEN in 1989 and has not been updated since that time. The NRC no longer supports the TRAC codes. They currently develop and maintain the TRACE code system, which is the TRAC/RELAP Advanced Computational Engine. TRACE is a modernized thermal-hydraulics code designed to consolidate the capabilities of NRC's 3 legacy safety codes - TRAC-P, TRAC-B and RELAP. This is NRC's flagship thermal-hydraulics analysis tool. See the website for more information http://www.nrccodes.com/.« less

  4. 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.

  5. 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.

  6. 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.

  7. 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).

  8. 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.

  9. 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.

  10. 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)

  11. Proposed Amendment to Presidential Permit PP-63 and Associated Modifications to 500-kV International Transmission Line: Forbes, Minnesota to Manitoba, Canada, Northern States Power Company. Addendum to the final Environmental Assessment

    SciTech Connect (OSTI)

    Not Available

    1992-10-01

    This Addendum to the Final Environmental Assessment for the Proposed Amendment to Presidential Permit PP-63 and Associated Modifications to 500 kV International Transmission Line: Forbes, Minnesota to Manitoba, Canada (DOE/EA-587) addresses Northern States Power Company`s (NSP) proposed expansion of the Forbes Substation. The applicant has requested that the expansion take place on the west side of the substation, within the existing property line, instead of on the north side as originally proposed. All of the proposed construction would take place on property already owned by NSP. DOE has reviewed the environmental impacts associated with this minor modification and has determined that the conclusions reached in the environmental assessment and Finding of No Significant Impact prepared in connection with NSP`s original amendment request remain valid.

  12. 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 ...

  13. 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 ...

  14. State Nuclear Profiles 2010

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

    Data for 2010 PWR Pressurized Light Water Reactor. Source: Form EIA-860, "Annual ... Data for 2010 PWR Pressurized Light Water Reactor. Source: Form EIA-860, "Annual ...

  15. 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.

  16. Northern New Mexico

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

    3 million for education, economic development, charitable giving in Northern New Mexico September 23, 2014 LOS ALAMOS, N.M., Sept. 23, 2014-The Los Alamos National Security,...

  17. Northern New Mexico Citizens' Advisory...

    Office of Environmental Management (EM)

    which was approved by the Northern New Mexico Citizens' Advisory Board during its ... the May 18, 2016 Board Meeting 1 NORTHERN NEW MEXICO CITIZENS' ADVISORY BOARD 1 ...

  18. Amended Application for Presidential Permit OE Docket No: PP–371 Northern Pass Transmission LLC

    Broader source: Energy.gov [DOE]

    Northern Pass Transmission LLC (Northern Pass) has submitted an amended application for a Presidential permit to construct, operate, maintain, and connect an electric transmission line across the United States border with Canada.

  19. Proposed amendment for Presidential Permit PP-63 and associated modifications to 500 kV international transmission line, Forbes, Minnesota to Manitoba, Canada Northern States Power Company. Final Environmental Assessment

    SciTech Connect (OSTI)

    Not Available

    1992-02-01

    Northern States Power Company, (NSP), a Minnesota investor owned utility has applied to the Office of Fossil Energy, United States Department of Energy, to amend Presidential Permit PP-63 to allow for alterations to the 500 kV transmission line and as sedated facilities currently regulated by this permit. The alterations proposed for the 500 kV line owned by NSP are part of a long term effort sponsored by NSP to upgrade the existing NSP transmission system to allow for increased exchange of electricity with the Manitoba Hydro-Electric Board. Presidential Permit PP-63 authorized NSP to construct, connect, operate and maintain a 500 kV line at the United States/Canadian border approximately seven-and-a-half miles west of Warroad in Roseau County, Minnesota. This line connects with a 500 kV line owned and operated by the Manitoba Hydro-Electric Board (MHEB), which extends from Dorsey, Manitoba, Canada to the United States/Canadian border. NSP proposes to increase the electricity transfer capability of this transmission facility by constructing a new 80-acre substation on the existing 500 kV line in Roseau County, Minnesota, and upgrading the existing substation at Forbes, Minnesota. The proposed Roseau substation would contain two 41.5 ohm series capacitor banks. In addition, static VAR compensators are to be installed at the existing Forbes Substation. Approximately 5 acres would be added to the 30-acre Forbes site to house the additional equipment. No new lines would enter or exit the facility. NSP proposes to place the new Roseau Substation in service in May 1993 and to complete the upgrading of the Forbes Substation in March 1994. The primary, initial purpose of these modifications is to enable NSP to import 400 megawatts of electric power from MHEB during the summer months to meet peak electrical demand in the Minneapolis-St. Paul area. It is expected that this power transfer would begin in 1993.

  20. 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)

  1. 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.

  2. Northern New Mexico Living

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

    Northern New Mexico Living Northern New Mexico Living The north end of the Land of Enchantment offers deserts, forests, and mountains, vibrant cities, and a rich culture and history Contact Us Email Living in Los Alamos Los Alamos, New Mexico is a community with a population of about 18,000. It sits on the Pajarito Plateau in Northern New Mexico between the city of White Rock and the Valles Caldera National Preserve. Lab employees who live in Los Alamos enjoy a 10-15 minute commute to work,

  3. Memorandum of Understanding among the United States Department...

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

    PP-371 Northern Pass Transmission LLC: Letter of MOU Cancellation Memorandum of Understanding among the United States Department of Energy and Northern Pass Transmission LLC and SE ...

  4. 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.

  5. 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.

  6. 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.

  7. Application for Presidential Permit OE Docket No. PP-371 Northern Pass

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

    Transmission: Comments from State Senator Jeanie Forrester | Department of Energy State Senator Jeanie Forrester Application for Presidential Permit OE Docket No. PP-371 Northern Pass Transmission: Comments from State Senator Jeanie Forrester Application from Northern Pass Transmission to construct, operate and maintain electric transmission facilities at the U.S. - Canada Border. State Senator Forrester_Comments.pdf (978.86

  8. 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.

  9. 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

  10. PP-45_Northern_States_Power.pdf

    Office of Environmental Management (EM)

  11. Basin Destination State

    Gasoline and Diesel Fuel Update (EIA)

    4. Estimated rail transportation rates for coal, basin to state, EIA data Basin Destination State 2008 2009 2010 2008-2010 2009-2010 Northern Appalachian Basin Delaware 26.24 - W...

  12. Basin Destination State

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    3. Estimated rail transportation rates for coal, basin to state, EIA data Basin Destination State 2008 2009 2010 2008-2010 2009-2010 Northern Appalachian Basin Delaware 28.49 - W...

  13. 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.

  14. 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.

  15. A guidebook to alternative energy projects on American Samoa, The Commonwealth of the Northern Mariana Islands, The Federated States of Micronesia, Guam, and The Republics of the Marshall Islands and Palau. [Contains bibliography

    SciTech Connect (OSTI)

    Case, C.W.

    1987-05-01

    The purpose of this guidebook is to help transfer information concerning alternative energy projects that have been tried on the Pacific islands affiliated with the US. These islands include those in American Samoa, the Commonwealth of the Northern Mariana Islands, the Federated States of Micronesia (Kosrae, Pohnpei, Truk, and Yap), Guam, and the Republics of the Marshall Islands and Palau. Distances are long between islands and populations are sparse, making communication and the transfer of information particularly difficult. A project that works on American Samoa might be appropriate for Yap, but to get this information to the proper people on Yap in a reasonable period of time is extremely difficult. This book describes 100 alternative energy projects that have been tried on the islands since the mid-1970's. This description and record of what has been done to date should be a source of ideas for energy workers, reduce duplication of work, and help encourage successes by describing other successes and failures. Alternative energy projects are projects that use indigenous, renewable resources in order to reduce local dependency on imported petroleum for electricity or liquid fuels. The islands have an apparent abundance of natural resources for this purpose such as the sun, rivers, vegetation, the ocean, and wind; and, ideally, it should be relatively simple to convert these resources to electricity or fuel. However, there are problems unique to the remote, tropical Pacific that often appear insurmountable, and successes to date are the results of unusual persistence, hard work, and ingenuity of those on the islands. Projects are confined to those that actually develop or demonstrate hardware. These projects use the complete spectrum of alternative technologies such as biomass conversion, wind electric, solar water heating, photovoltaics, wind water pumping, hydroelectric, water desalination, and integrated systems. 381 refs., 85 figs.

  16. Performance Spec. for Fuel Drying and Canister Inerting System for PWR Core 2 Blanket Fuel Assemblies Stored within Shipping Port Spent Fuel Canisters

    SciTech Connect (OSTI)

    JOHNSON, D.M.

    2000-03-14

    This specification establishes the performance requirements and basic design requirements imposed on the fuel drying and canister inerting system for Shippingport Pressurized Water Reactor (PWR) Core 2 blanket fuel assemblies (BFAs) stored within Shippingport spent fuel (SSFCs) canisters (fuel drying and canister inerting system). This fuel drying and canister inerting system is a component of the U.S. Department of Energy, Richland Operations Office (RL) Spent Nuclear Fuels Project at the Hanford Site. The fuel drying and canister inerting system provides for removing water and establishing an inert environment for Shippingport PWR Core 2 BFAs stored within SSFCs. A policy established by the U.S. Department of Energy (DOE) states that new SNF facilities (this is interpreted to include structures, systems and components) shall achieve nuclear safety equivalence to comparable U.S. Nuclear Regulatory Commission (NRC)-licensed facilities. This will be accomplished in part by applying appropriate NRC requirements for comparable NRC-licensed facilities to the fuel drying and canister inerting system, in addition to applicable DOE regulations and orders.

  17. State

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

    Created on: 8/26/2016 3:22:30 PM Table 2. Natural gas consumption in the United States, 2011-2016 (billion cubic feet, or as indicated) Year and Month Lease and Plant Fuel a Pipeline and Distribution Use b Delivered to Consumers Total Consumption Heating Value c (Btu per cubic foot) Residential Commercial Industrial Electric Power Vehicle Fuel Total 2011 Total 1,323 688 4,714 3,155 6,994 7,574 30 22,467 24,477 1,022 2012 Total 1,396 731 4,150 2,895 7,226 9,111 30 23,411 25,538 1,024 2013 Total

  18. 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.

  19. 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.

  20. EIA - State Nuclear Profiles

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

    ...171988 12182027 2,330 19,200 94.1 Data for 2010 PWR Pressurized Light Water Reactor. ... 811987 1162026 2,300 19,856 98.5 Data for 2010 PWR Pressurized Light Water Reactor. ...

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. In-plant test and evaluation of the neutron collar for verification of PWR fuel assemblies at Resende, Brazil

    SciTech Connect (OSTI)

    Menlove, H.O.; Marzo, M.A.S.; de Almeida, S.G.; de Almeida, M.C.; Moitta, L.P.M.; Conti, L.F.; de Paiva, J.R.T.

    1985-11-01

    The neutron-coincidence collar has been evaluated for the measurement of pressurized-water reactor (PWR) fuel assemblies at the Fabrica de Elementos Combustiveis plant in Resende, Brazil. This evaluation was part of the cooperative-bilateral-safeguards technical-exchange program between the United States and Brazil. The neutron collar measures the STVU content per unit length of full fuel assemblies using neutron interrogation and coincidence counting. The STYU content is measured in the passive mode without the AmLi neutron-interrogation source. The extended evaluation took place over a period of 6 months with both scanning and single-zone measurements. The results of the tests gave a coincidence-response standard deviation of 0.7% (sigma = 1.49% for mass) for the active case and 2.5% for the passive case in 1000-s measurement times. The length measurement in the scanning mode was accurate to 0.77%. The accuracies of different calibration methods were evaluated and compared.

  6. Improvement of the thermal margins in the Swedish Ringhals-3 PWR by introducing new fuel assemblies with thorium

    SciTech Connect (OSTI)

    Lau, C. W.; Demaziere, C.; Nylen, H.; Sandberg, U.

    2012-07-01

    Thorium is a fertile material and most of the past research has focused on breeding thorium to fissile material. In this paper, the focus is on using thorium to improve the thermal margins by homogeneously distributing thorium in the fuel pellets. A proposed uranium-thorium-based fuel assembly is simulated for the Swedish Ringhals-3 PWR core in a realistic demonstration. All the key safety parameters, such as isothermal temperature coefficient of reactivity, Doppler temperature of reactivity, boron worth, shutdown margins and fraction of delayed neutrons are studied in this paper, and are within safety limits for the new core design using the uranium-thorium-based fuel assemblies. The calculations were performed by the two-dimensional transport code CASMO-4E and the two group steady-state three dimensional nodal code SIMULATE-3 from Studsvik Scandpower. The results showed that the uranium-thorium-based fuel assembly improves the thermal margins, both in the pin peak power and the local power (Fq). The improved thermal margins would allow more flexible core designs with less neutron leakage or could be used in power uprates to offer efficient safety margins. (authors)

  7. State Nuclear Profiles 2010

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

    2010 57 Comanche Peak 1 1,209 9,677 91.4 PWR 8131990 282030 2 1,197 10,532 100.4 PWR 831993 222033 2,406 20,208 95.9 Data for 2010 PWR Pressurized Light Water Reactor. ...

  8. Northern Biodiesel | Open Energy Information

    Open Energy Info (EERE)

    Northern Biodiesel Place: Ontario, New York Product: Biodiesel producer. Coordinates: 34.06457, -117.647809 Show Map Loading map... "minzoom":false,"mappingservice":"googlemap...

  9. NORTHERN NEW MEXICO CITIZENS' ADVISORY

    Office of Environmental Management (EM)

    NORTHERN NEW MEXICO CITIZENS' ADVISORY BOARD NEWSLETTER NNMCAB Hosts Fall Meeting in Santa ... NNMCAB Newsletter Spring 2016 In This Issue Meet the New Board Members, Page 2. ...

  10. Northern New Mexico Citizens' Advisory...

    Office of Environmental Management (EM)

    15 Northern New Mexico Citizens' Advisory Board 94 Cities of Gold Road, Santa Fe, New Mexico 87506 Environmental Monitoring & Remediation Committee Roster 1. Stephen Schmelling,...

  11. NORTHERN NEW MEXICO CITIZENS' ADVISORY...

    Office of Environmental Management (EM)

    NORTHERN NEW MEXICO CITIZENS' ADVISORY BOARD (NNMCAB) Environmental Monitoring, Surveillance and Remediation Committee Recommendation to the Department of Energy No. 2009-12...

  12. Study of the state of design for pipe whip. Final report. [PWR; BWR

    SciTech Connect (OSTI)

    Evans, P.A.; Neely, B.B.; Wilson, D.M.; Enis, R.O.

    1980-01-01

    Design methods and parameters are described which are addressed when considering consequences of a postulated pipe rupture event in a nuclear plant design. Parameters discussed are break opening time and size, resultant blowdown characteristics of the effluent from the broken pipe, jet reaction and impingement loading, pipe motion, and pipe impact loading on steel and concrete structures. The impact the various parameters have on overall plant designs and conservatisms inherent in each consideration are evaluated in a qualitative nature. Finally, recommendations are provided for each parameter discussed for further evaluation and study.

  13. Steady-state pressure losses for Multirod Burst Test (MRBT) bundle B-5. [PWR

    SciTech Connect (OSTI)

    Bailey, R.T.

    1982-04-01

    The Oak Ridge National Laboratory (ORNL) has undertaken the Multirod Burst Test (MRBT) program for the purpose of characterizing the deformation behavior of unirradiated fuel cladding. As part of this program, ORNL contracted with the Babcock and Wilcox company (B and W) to obtain experimental hydraulic data for one of the MRBT bundles. This report presents the data that describe the pressure loss characteristics of Multirod Burst Test Bundle B-5 and a reference or pre-burst geometry bundle. The 8 x 8-rod bundles were flow tested at Reynolds numbers between 17,700 nd 177,000. For each of the five test flow rates, the static pressures at 480 points on the perimeter of the bundles were measured. The total pressure loss for the B-5 bundle showed about a fourfold increase over that for the reference geometry bundle. The shape of the axial pressure loss profile for the B-5 bundle agreed with the observed distribution of the clad deformations. The experimental data presented in this report will be used as one of essential inputs to the continuing analytical work at ORNL.

  14. Northern Great Plains

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

    Northern G reat P lains Climate C hange a nd t he U .S. E nergy S ector: Regional v ulnerabilities a nd r esilience s olutions Summary i n B rief The N orthern G reat P lains i s h ome t o l ess t han 2 % o f t he U .S. population b ut i s a m ajor s upplier o f c ritical e nergy r esources used t hroughout t he n ation. T hese r esources i nclude c oal f rom the P owder R iver B asin, e lectricity e xported v ia i nterstate transmission l ines, a nd r apidly g rowing o il p roduction f rom t he

  15. RoboRave Rally Northern New Mexico

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

    RoboRave Rally Northern New Mexico RoboRave Rally Northern New Mexico WHEN: Mar 07, 2015 9:00 AM - 3:00 PM WHERE: Northern New Mexico University 921 North Paseo De Onate, Espaola...

  16. Northern Rockies Geothermal Region | Open Energy Information

    Open Energy Info (EERE)

    ENERGYGeothermal Home Northern Rockies Geothermal Region Details Areas (0) Power Plants (0) Projects (0) Techniques (0) Map: Name Province is situated in northern Idaho...

  17. Northern Growers LLC | Open Energy Information

    Open Energy Info (EERE)

    Farmer cooperative that provides corn to Northern Lights Ethanol LLC (a 77% owned joint venture with Broin Companies). References: Northern Growers LLC1 This article is a...

  18. 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.

  19. Fuel axial relocation in ballooning fuel rods. [PWR; BWR

    SciTech Connect (OSTI)

    Siefken, L.J.

    1983-01-01

    Fuel movement, in the longitudinal direction in ballooning fuel rods, shifts the position of heat generation and may cause an increase in cladding temperature in the ballooning region. This paper summarizes the axial fuel relocation data obtained in fuel rod tests conducted in the United States and the Federal Republic of Germany, describes a model for calculating fuel axial relocation, and gives a quantitative analysis of the impact of fuel relocation on cladding temperature. The amount of fuel relocation in 18 ballooned fuel rods was determined from neutron radiographs, niobium gamma decay counts, and photomicrographs. The fuel rods had burnups in the range of 0 to 35,000 MWd/t and cladding hoop strains varying from 0 to 72%.

  20. Temperature estimates from zircaloy oxidation kinetics and microstructures. [PWR

    SciTech Connect (OSTI)

    Olsen, C.S.

    1982-10-01

    This report reviews state-of-the-art capability to determine peak zircaloy fuel rod cladding temperatures following an abnormal temperature excursion in a nuclear reactor, based on postirradiation metallographic analysis of zircaloy microstructural and oxidation characteristics. Results of a comprehensive literature search are presented to evaluate the suitability of available zircaloy microstructural and oxidation data for estimating anticipated reactor fuel rod cladding temperatures. Additional oxidation experiments were conducted to evaluate low-temperature zircaloy oxidation characteristics for postirradiation estimation of cladding temperature by metallographic examination. Results of these experiments were used to calculate peak cladding temperatures of electrical heater rods and nuclear fuel rods that had been subjected to reactor temperature transients. Comparison of the calculated and measured peak cladding temperatures for these rods indicates that oxidation kinetics is a viable technique for estimating peak cladding temperatures over a broad temperature range. However, further improvement in zircaloy microstructure technology is necessary for precise estimation of peak cladding temperatures by microstructural examination.

  1. Northern Cheyenne Tribe- 2002 Project

    Broader source: Energy.gov [DOE]

    The Northern Cheyenne Indian Tribe is a Federally Recognized Sovereign Nation, located in Big Horn and Rosebud counties in southeastern Montana. The study will assess the feasibility of a commercial wind facility on lands selected and owned by the Northern Cheyenne Nation and will examine the potential for the development of solar and biomass resources located on tribal lands. The feasibility study will focus on analyzing, qualifying, and quantifying the opportunity for the Northern Cheyenne Nation to develop, own, and operate a commercial wind facility on tribal lands.

  2. 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

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. Microsoft Word - Northern Pass Amended Application - FINAL

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

    UNITED STATES OF AMERICA BEFORE THE DEPARTMENT OF ENERGY OFFICE OF ELECTRICITY DELIVERY AND ENERGY RELIABILITY NORTHERN PASS TRANSMISSION LLC DOCKET NO. PP-371 AMENDED APPLICATION JULY 1, 2013 i TABLE OF CONTENTS Page No. LIST OF EXHIBITS iii LIST OF ABBREVIATIONS iv INTRODUCTION 1 OVERVIEW OF AMENDMENTS TO APPLICATION 1 SECTION 1 - INFORMATION REGARDING THE APPLICANT 1.1 Legal Name of the Applicant 6 1.2 Legal Names of All Partners 6 1.3 Communications and Correspondence 7 1.4 Foreign Ownership

  8. 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.

  9. Northern New Mexico Citizens' Advisory...

    Office of Environmental Management (EM)

    October 20, 2015 Northern New Mexico Citizens' Advisory Board 94 Cities of Gold Road, Santa Fe, New Mexico 87506 Executive Committee Roster 1. Doug Sayre, NNMCAB Chair 2. Gerard...

  10. Northern New Mexico Citizens' Advisory...

    Office of Environmental Management (EM)

    Committee October 20, 2015 Northern New Mexico Citizens' Advisory Board 94 Cities of Gold Road, Santa Fe, New Mexico 87506 The NNMCAB is comprised of citizens appointed by the...

  11. 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

  12. 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.

  13. 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.

  14. 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)

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. Northern Virginia Residents Improve Their Homes' Energy With...

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

    Northern Virginia Residents Improve Their Homes' Energy With A Funding Boost Northern Virginia Residents Improve Their Homes' Energy With A Funding Boost The Northern Virginia Home ...

  20. Application for Presidential Permit OE Docket No. PP-044 Northern...

    Energy Savers [EERE]

    44 Northern Electric Cooperative Association Application for Presidential Permit OE Docket No. PP-044 Northern Electric Cooperative Association Application from Northern Electric ...

  1. Northern Arizona University Wind Projects | Open Energy Information

    Open Energy Info (EERE)

    Northern Arizona University Wind Projects (Redirected from Northern Arizona University Wind Project) Jump to: navigation, search Northern Arizona University ARD Wind Project...

  2. State Nuclear Profiles 2010

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

    Data for 2010 PWR Pressurized Light Water Reactor. Source: Form EIA-860, "Annual Electric Generator Report," and Form EIA-923, "Power Plant Operations Report." Type Commercial ...

  3. State Nuclear Profiles 2010

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

    Data for 2010 PWR Pressurized Light Water Reactor. Source: Form EIA-860, "Annual ... Data for 2010 BWR Boiling Water Reactor. Source: Form EIA-860, "Annual Electric ...

  4. State Nuclear Profiles 2010

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

    Data for 2010 BWR Boiling Water Reactor. Source: Form EIA-860, "Annual Electric ... Data for 2010 PWR Pressurized Light Water Reactor. Source: Form EIA-860, "Annual ...

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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)

  16. Northern New Mexico regional airport market feasibility

    SciTech Connect (OSTI)

    Drake, R.H.; Williams, D.S.

    1998-06-01

    This report is about the market for airline travel in northern New Mexico. Interest in developing a northern New Mexico regional airport has periodically surfaced for a number of years. The New Mexico State Legislature passed a memorial during the 1998 Second Session calling for the conduct of a study to determine the feasibility of building a new regional airport in NNM. This report is a study of the passenger market feasibility of such an airport. In addition to commercial passenger market feasibility, there are other feasibility issues dealing with siting, environmental impact, noise, economic impact, intermodal transportation integration, region-wide transportation services, airport engineering requirements, and others. These other feasibility issues are not analyzed in any depth in this report although none were discovered to be show-stoppers as a by-product of the authors doing research on the passenger market itself. Preceding the need for a detailed study of these other issues is the determination of the basic market need for an airport with regular commercial airline service in the first place. This report is restricted to an in-depth look at the market for commercial passenger air service in NNM. 20 figs., 8 tabs.

  17. Tundra Bushes Add Fuel to Northern Thaw

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

    Tundra Bushes Add Fuel to Northern Thaw Tundra Bushes Add Fuel to Northern Thaw Bonfils-1.jpg Why it Matters: Simulations at NERSC are the first to investigate long-term climate...

  18. 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.

  19. EERE Success Story-Northern California: Innovative Exploration...

    Office of Environmental Management (EM)

    Northern California: Innovative Exploration Technologies Yield Geothermal Potential EERE Success Story-Northern California: Innovative Exploration Technologies Yield Geothermal ...

  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. RoboRAVE Rally Northern New Mexico

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

    RoboRAVE Rally Northern New Mexico RoboRAVE Rally Northern New Mexico WHEN: Mar 05, 2016 9:00 AM - 3:00 PM WHERE: Northern New Mexico College, Eagle Memomorial Gymnasium Espanola, NM CONTACT: Janelle Vigil-Maestas 505 665-4329 CATEGORY: Community INTERNAL: Calendar Login Event Description Numerous teams of students from Northern New Mexico area schools are testing their critical thinking and teamwork skills to build autonomous robots at RoboRAVE Rally Northern New Mexico. Additionally, Explora

  3. 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.

  4. 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

  5. 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

  6. 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.

  7. 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.

  8. 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.

  9. Northern Cheyenne Tribe- 2003 Project

    Broader source: Energy.gov [DOE]

    The Northern Cheyenne Tribe, located in Rosebud and Bighorn counties in southeastern Montana, plans to conduct preconstruction activities for a 30-MW wind facility. The tribe was awarded a feasibility study grant in FY2002 for wind resource monitoring, and is accelerating the study and proceeding with development after the collected resource data was correlated to long-term wind resource data from a nearby site, and the resources confirmed. Activities include permitting, avian and cultural assessments, and the transmission and interconnection studies needed to obtain financing and power purchase agreements.

  10. Northern Nevada Geothermal Exploration Strategy Analysis | Open...

    Open Energy Info (EERE)

    Nevada Geothermal Exploration Strategy Analysis Jump to: navigation, search OpenEI Reference LibraryAdd to library Journal Article: Northern Nevada Geothermal Exploration Strategy...

  11. Northern Pass Transmission Line Project Environmental Impact...

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

    Northern Pass Transmission Line Project Environmental Impact Statement: Announcement of Change in Public Meeting Location: Federal Register Notice Volume 78, No. 181 - September ...

  12. Northern Colorado Clean Cities | Open Energy Information

    Open Energy Info (EERE)

    Cities Jump to: navigation, search Name: Northern Colorado Clean Cities Address: PO Box 759 Place: Johnstown, Colorado Zip: 80534 Region: Rockies Area Number of Employees:...

  13. Northern Ireland Electricity | Open Energy Information

    Open Energy Info (EERE)

    Ireland Electricity Jump to: navigation, search Name: Northern Ireland Electricity Place: Belfast, United Kingdom Zip: BT9 5HT Product: NIE is reponsible for the regulated...

  14. RoboRave Rally tests Northern New

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

    RoboRave Rally tests Northern New Mexico students' critical thinking, teamwork skills March 5, 2015 Event sparks, nurtures interest in science and technology LOS ALAMOS, N.M., March 5, 2015-Eighty-four teams of students from Northern New Mexico area schools test their critical thinking and teamwork skills building autonomous robots at RoboRave Rally Northern New Mexico on Saturday, March 7 at Northern New Mexico University in Española. - 2 - "Introducing students at an early age to

  15. Northern Excellence Seed LLC | Open Energy Information

    Open Energy Info (EERE)

    Seed LLC Jump to: navigation, search Name: Northern Excellence Seed LLC Place: Williams, Minnesota Sector: Biomass Product: Producer-owned cooperative focused on...

  16. Northern California Power Agny | Open Energy Information

    Open Energy Info (EERE)

    California Power Agny Jump to: navigation, search Name: Northern California Power Agny Place: California Website: www.ncpa.com Outage Hotline: (916) 781-3636 References: EIA Form...

  17. Northern Arizona University | Open Energy Information

    Open Energy Info (EERE)

    University Jump to: navigation, search Name: Northern Arizona University Place: Flagstaff, AZ Zip: 86011 Phone Number: 928-523-0715 Website: nau.edu Coordinates: 35.1905403,...

  18. Northern New Mexico Math & Science Academy

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

    Northern New Mexico Math & Science Academy for Teachers (MSA) Program Description MSA is an intensive and comprehensive professional development program for K-12 teachers....

  19. 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.

  20. 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.

  1. 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.

  2. 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)

  3. 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.

  4. Memorandum of Understanding among the United States Department of Energy

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

    and Northern Pass Transmission LLC and SE Group: August 2011 | Department of Energy Transmission LLC and SE Group: August 2011 Memorandum of Understanding among the United States Department of Energy and Northern Pass Transmission LLC and SE Group: August 2011 Memorandum of Understanding among the United States Department of Energy and Northern Pass Transmission LLC and SE Group regarding independent third party preparation of a NEPA compliance document. It is the purpose of this Memorandum

  5. 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.

  6. 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.

  7. Decay Heat Calculations for PWR and BWR Assemblies Fueled with Uranium and Plutonium Mixed Oxide Fuel using SCALE

    SciTech Connect (OSTI)

    Ade, Brian J; Gauld, Ian C

    2011-10-01

    In currently operating commercial nuclear power plants (NPP), there are two main types of nuclear fuel, low enriched uranium (LEU) fuel, and mixed-oxide uranium-plutonium (MOX) fuel. The LEU fuel is made of pure uranium dioxide (UO{sub 2} or UOX) and has been the fuel of choice in commercial light water reactors (LWRs) for a number of years. Naturally occurring uranium contains a mixture of different uranium isotopes, primarily, {sup 235}U and {sup 238}U. {sup 235}U is a fissile isotope, and will readily undergo a fission reaction upon interaction with a thermal neutron. {sup 235}U has an isotopic concentration of 0.71% in naturally occurring uranium. For most reactors to maintain a fission chain reaction, the natural isotopic concentration of {sup 235}U must be increased (enriched) to a level greater than 0.71%. Modern nuclear reactor fuel assemblies contain a number of fuel pins potentially having different {sup 235}U enrichments varying from {approx}2.0% to {approx}5% enriched in {sup 235}U. Currently in the United States (US), all commercial nuclear power plants use UO{sub 2} fuel. In the rest of the world, UO{sub 2} fuel is still commonly used, but MOX fuel is also used in a number of reactors. MOX fuel contains a mixture of both UO{sub 2} and PuO{sub 2}. Because the plutonium provides the fissile content of the fuel, the uranium used in MOX is either natural or depleted uranium. PuO{sub 2} is added to effectively replace the fissile content of {sup 235}U so that the level of fissile content is sufficiently high to maintain the chain reaction in an LWR. Both reactor-grade and weapons-grade plutonium contains a number of fissile and non-fissile plutonium isotopes, with the fraction of fissile and non-fissile plutonium isotopes being dependent on the source of the plutonium. While only RG plutonium is currently used in MOX, there is the possibility that WG plutonium from dismantled weapons will be used to make MOX for use in US reactors. Reactor-grade plutonium

  8. PP-45-2_Northern_States.pdf

    Office of Environmental Management (EM)

  9. Northern States Power Co - Minnesota | Open Energy Information

    Open Energy Info (EERE)

    1,208,753 2008-07 110,758 966,980 1,076,117 124,248 1,312,516 129,926 63,601 839,607 486 143 1,816 1 298,750 3,120,919 1,206,530 2008-06 71,189 660,385 1,075,872 102,798...

  10. Northern States Power Co - Minnesota | Open Energy Information

    Open Energy Info (EERE)

    o Green Button Access: Implemented Green Button Landing Page: www.xcelenergy.com Green Button Reference Page: www.whitehouse.govblog2 References: EIA Form EIA-861 Final...

  11. Northern States Power Co - Minnesota | Open Energy Information

    Open Energy Info (EERE)

    129,786 42,878 755,947 495 149 1,880 1 176,260 2,415,001 1,208,224 2008-03 69,949 709,960 1,077,504 84,807 1,151,326 129,654 45,683 773,628 505 158 1,865 1 200,597 2,636,779...

  12. Northern States Power Co - Wisconsin | Open Energy Information

    Open Energy Info (EERE)

    1,440 126.007 2,120 2 817.367 9,796 9,178 2008-08 346.652 3,666 7,731 404.85 4,896 1,438 132.25 2,057 2 883.752 10,619 9,171 2008-07 365.14 3,868 7,711 364.545 4,416 1,439 148.28...

  13. EIS-0463: Northern Pass Transmission Line Project, New Hampshire...

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

    permit to Northern Pass Transmission, LLC, to construct, operate, maintain, and connect a new electric transmission line across the U.S.-Canada border in northern New Hampshire. ...

  14. Video: Energy Savings Performance Contracts Success Story-Northern...

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

    Energy Savings Performance Contracts Success Story-Northern California TRACON Video: Energy Savings Performance Contracts Success Story-Northern California TRACON See how Federal ...

  15. EA-1945: Northern Mid-Columbia Joint Project; Grant, Douglas...

    Office of Environmental Management (EM)

    5: Northern Mid-Columbia Joint Project; Grant, Douglas, and Chelan Counties, Washington EA-1945: Northern Mid-Columbia Joint Project; Grant, Douglas, and Chelan Counties, ...

  16. EA-1945: Northern Mid-Columbia Joint Project; Grant, Douglas...

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

    EA-1945: Northern Mid-Columbia Joint Project; Grant, Douglas, and Chelan Counties, Washington EA-1945: Northern Mid-Columbia Joint Project; Grant, Douglas, and Chelan Counties, ...

  17. University of Minnesota (NorthernSTAR Building America Partnership...

    Open Energy Info (EERE)

    (NorthernSTAR Building America Partnership) Jump to: navigation, search Name: University of Minnesota (NorthernSTAR Building America Partnership) Place: St. Paul, MN Information...

  18. Application for Presidential Permit OE Docket No. PP-371 Northern...

    Energy Savers [EERE]

    Vickie Bedard Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from Vickie Bedard Application from Northern Pass to construct, operate and maintain ...

  19. Application for Presidential Permit OE Docket No. PP-371 Northern...

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

    Erick Berglund, Jr. Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from Erick Berglund, Jr. Application from Northern Pass to construct, operate ...

  20. Application for Presidential Permit OE Docket No. PP-371 Northern...

    Energy Savers [EERE]

    Hanglin Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from Pamela Hanglin Application from Northern Pass to construct, operate and maintain ...

  1. Application for Presidential Permit OE Docket No. PP-371 Northern...

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

    Lee Ann Moulder Application for Presidential Permit OE Docket No. PP-371 Northern Pass Transmission: Comments from Lee Ann Moulder Application from Northern Pass Transmission to ...

  2. Application for Presidential Permit OE Docket No. PP-371 Northern...

    Office of Environmental Management (EM)

    Courtney Kearley Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from Courtney Kearley Application from Northern Pass to construct, operate and ...

  3. Application for Presidential Permit OE Docket No. PP-371 Northern...

    Office of Environmental Management (EM)

    Maureen Quinn Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from Maureen Quinn Application from Northern Pass to construct, operate and maintain ...

  4. Application for Presidential Permit OE Docket No. PP-371 Northern...

    Office of Environmental Management (EM)

    Roy Stever Application for Presidential Permit OE Docket No. PP-371 Northern Pass Transmission: Comments from Roy Stever Application from Northern Pass Transmission to construct, ...

  5. Application for Presidential Permit OE Docket No. PP-371 Northern...

    Office of Environmental Management (EM)

    John Doane Sr. Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from John Doane Sr. Application from Northern Pass to construct, operate and ...

  6. Application for Presidential Permit OE Docket No. PP-371 Northern...

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

    Robert Cote Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from Robert Cote Application from Northern Pass to construct, operate and maintain ...

  7. PP-60 Northern Electric Cooperative Association (NEC) | Department...

    Office of Environmental Management (EM)

    PP-60 Northern Electric Cooperative Association (NEC) Presidential Permit authorizing Northern Electric Cooperative Association (NEC) to construct, operate, and to maintain ...

  8. Application for Presidential Permit OE Docket No. PP-371 Northern...

    Office of Environmental Management (EM)

    Taras Kucman Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from Taras Kucman Application from Northern Pass to construct, operate and maintain ...

  9. Application for Presidential Permit OE Docket No. PP-371 Northern...

    Energy Savers [EERE]

    The Weeks Lancaster Trust Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from The Weeks Lancaster Trust Application from Northern Pass to ...

  10. Application for Presidential Permit OE Docket No. PP-371 Northern...

    Energy Savers [EERE]

    Susan Seitz Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from Susan Seitz Application from Northern Pass to construct, operate and maintain ...

  11. Application for Presidential Permit OE Docket No. PP-371 Northern...

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

    Nancy Rheinhardt Application for Presidential Permit OE Docket No. PP-371 Northern Pass Transmission: Comments from Nancy Rheinhardt Application from Northern Pass Transmission to ...

  12. Application for Presidential Permit OE Docket No. PP-371 Northern...

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

    Transmission LLC Application for Presidential Permit OE Docket No. PP-371 Northern Pass Transmission LLC Amended application from Northern Pass to construct, operate and maintain ...

  13. Application for Presidential Permit OE Docket No. PP-371 Northern...

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

    Mary Bearor Application for Presidential Permit OE Docket No. PP-371 Northern Pass Transmission: Comments from Mary Bearor Application from Northern Pass Transmission to construct, ...

  14. Application for Presidential Permit OE Docket No. PP-371 Northern...

    Office of Environmental Management (EM)

    Linda Upham Application for Presidential Permit OE Docket No. PP-371 Northern Pass Transmission: Comments from Linda Upham Application from Northern Pass Transmission to construct, ...

  15. Application for Presidential Permit OE Docket No. PP-371 Northern...

    Office of Environmental Management (EM)

    Michael Marino Application for Presidential Permit OE Docket No. PP-371 Northern Pass Transmission: Comments from Michael Marino Application from Northern Pass Transmission to ...

  16. Application for Presidential Permit OE Docket No. PP-371 Northern...

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

    John Brodeer Application for Presidential Permit OE Docket No. PP-371 Northern Pass Transmission: Comments from John Brodeer Application from Northern Pass Transmission to ...

  17. Application for Presidential Permit OE Docket No. PP-371 Northern...

    Office of Environmental Management (EM)

    Jim Cannon Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from Jim Cannon Application from Northern Pass to construct, operate and maintain ...

  18. Application for Presidential Permit OE Docket No. PP-371 Northern...

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

    Alice Peatman Application for Presidential Permit OE Docket No. PP-371 Northern Pass Transmission: Comments from Alice Peatman Application from Northern Pass Transmission to ...

  19. Application for Presidential Permit OE Docket No. PP-371 Northern...

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

    Fred Brownson Application for Presidential Permit OE Docket No. PP-371 Northern Pass Transmission: Comments from Fred Brownson Application from Northern Pass Transmission to ...

  20. Application for Presidential Permit OE Docket No: PP-371 Northern...

    Office of Environmental Management (EM)

    No: PP-371 Northern Pass Transmission: Comments from William and Michelle Shoemaker Application for Presidential Permit OE Docket No: PP-371 Northern Pass Transmission: Comments ...

  1. Application for Presidential Permit OE Docket No. PP-371 Northern...

    Energy Savers [EERE]

    Lindsey Coombs Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from Lindsey Coombs Application from Northern Pass to construct, operate and ...

  2. Application for Presidential Permit OE Docket No. PP-371 Northern...

    Office of Environmental Management (EM)

    Hayes Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from Pamela Hayes Application from Northern Pass to construct, operate and maintain electric ...

  3. Application for Presidential Permit OE Docket No. PP-371 Northern...

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

    Ann Vennerbeck Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from Ann Vennerbeck Application from Northern Pass to construct, operate and ...

  4. Application for Presidential Permit OE Docket No. PP-371 Northern...

    Energy Savers [EERE]

    Karen Skurka Application for Presidential Permit OE Docket No. PP-371 Northern Pass Transmission: Comments from Karen Skurka Application from Northern Pass Transmission to ...

  5. Application for Presidential Permit OE Docket No. PP-371 Northern...

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

    Serita Frey Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from Serita Frey Application from Northern Pass to construct, operate and maintain ...

  6. Application for Presidential Permit OE Docket No. PP-371 Northern...

    Energy Savers [EERE]

    James Kennedy Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from James Kennedy Application from Northern Pass to construct, operate and maintain ...

  7. Application for Presidential Permit OE Docket No. PP-371 Northern...

    Office of Environmental Management (EM)

    Bruce Adami Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from Bruce Adami Application from Northern Pass to construct, operate and maintain ...

  8. Application for Presidential Permit OE Docket No. PP-371 Northern...

    Energy Savers [EERE]

    Roy Kjendal Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from Roy Kjendal Application from Northern Pass to construct, operate and maintain ...

  9. Application for Presidential Permit OE Docket No. PP-371 Northern...

    Energy Savers [EERE]

    Lorna Rose Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from Lorna Rose Application from Northern Pass to construct, operate and maintain ...

  10. Application for Presidential Permit OE Docket No. PP-371 Northern...

    Energy Savers [EERE]

    Rana Klug Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from Rana Klug Application from Northern Pass to construct, operate and maintain electric ...

  11. Application for Presidential Permit OE Docket No. PP-371 Northern...

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

    Steve Nogueira Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from Steve Nogueira Application from Northern Pass to construct, operate and ...

  12. Application for Presidential Permit OE Docket No. PP-371 Northern...

    Energy Savers [EERE]

    Anne Moschella Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from Anne Moschella Application from Northern Pass to construct, operate and ...

  13. Application for Presidential Permit OE Docket No. PP-371 Northern...

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

    Paula VandeWerken Application for Presidential Permit OE Docket No. PP-371 Northern Pass Transmission: Comments from Paula VandeWerken Application from Northern Pass Transmission ...

  14. Application for Presidential Permit OE Docket No. PP-371 Northern...

    Office of Environmental Management (EM)

    Fran Buteau Application for Presidential Permit OE Docket No. PP-371 Northern Pass Transmission: Comments from Fran Buteau Application from Northern Pass Transmission to construct, ...

  15. Application for Presidential Permit OE Docket No. PP-371 Northern...

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

    Nicholas Karakoudas Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from Nicholas Karakoudas Application from Northern Pass to construct, operate ...

  16. Application for Presidential Permit OE Docket No. PP-371 Northern...

    Energy Savers [EERE]

    Gina Neily Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from Gina Neily Application from Northern Pass to construct, operate and maintain ...

  17. Application for Presidential Permit OE Docket No. PP-371 Northern...

    Energy Savers [EERE]

    Larry Rappaport Application for Presidential Permit OE Docket No. PP-371 Northern Pass Transmission: Comments from Larry Rappaport Application from Northern Pass Transmission to ...

  18. Application for Presidential Permit OE Docket No. PP-371 Northern...

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

    The Nature Conservancy Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from The Nature Conservancy Application from Northern Pass to construct, ...

  19. Northern Colorado Wind Energy Center (Siemens) | Open Energy...

    Open Energy Info (EERE)

    Siemens) Jump to: navigation, search Name Northern Colorado Wind Energy Center (Siemens) Facility Northern Colorado Wind Energy Center (Siemens) Sector Wind energy Facility Type...

  20. Northern Colorado Wind Energy Center (GE) | Open Energy Information

    Open Energy Info (EERE)

    GE) Jump to: navigation, search Name Northern Colorado Wind Energy Center (GE) Facility Northern Colorado Wind Energy Center (GE) Sector Wind energy Facility Type Commercial Scale...

  1. Application for presidential permit OE Docket No. PP-371 Northern...

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

    PP-371 Northern Pass Transmission LLC: Letter of MOU Cancellation March 7, 2011 Northern Pass Transmission LLC appreciates the cooperation of the Department of Energy ("DOE") in ...

  2. Application for Presidential Permit OE Docket No. PP-371 Northern...

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

    PP-371 Northern Pass: Comments from Michelle Kleindienst Application for Presidential Permit OE Docket No. PP-371 Northern Pass: Comments from Stephen Buzzell and Lelah Sullivan

  3. Venture Acceleration Fund awards spur investment in Northern...

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

    VAF awards spur investment in Northern New Mexico Venture Acceleration Fund awards spur investment in Northern New Mexico Proposals are being accepted for Venture Acceleration Fund...

  4. Application for Presidential Permit OE Docket No. PP-371 Northern...

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

    Elisha Gray Application for Presidential Permit OE Docket No. PP-371 Northern Pass Transmission: Comments from Elisha Gray Application from Northern Pass Transmission to construct, ...

  5. Application for Presidential Permit OE Docket No. PP-371 Northern...

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

    Application for Presidential Permit OE Docket No. PP-371 Northern Pass Transmission: Comments from Elisha Gray Application for Presidential Permit OE Docket No. PP-371 Northern ...

  6. Application for Presidential Permit OE Docket No. PP-371 Northern...

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

    No. PP-371 Northern Pass Transmission: Comments from Larry Rappaport Application for Presidential Permit OE Docket No. PP-371 Northern Pass Transmission: Comments from Elisha Gray

  7. Application for Presidential Permit OE Docket No. PP-371 Northern...

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

    No. PP-371 Northern Pass Transmission: Comments from Pamela Martin Application for Presidential Permit OE Docket No. PP-371 Northern Pass Transmission: Comments from Elisha Gray

  8. Northern New Mexico College Foundation honors LANL's Richard...

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

    LANL's Richard Marquez honored with leadership, service award Northern New Mexico College ... to the advancement of educational and economic opportunities for Northern New Mexico." ...

  9. Northern New Mexico Citizens' Advisory Board BYLAWS I. MISSION

    Office of Environmental Management (EM)

    Northern New Mexico Citizens' Advisory Board BYLAWS I. MISSION The responsibility of the Environmental Management Site Specific Advisory Board (EM SSAB) at Northern New Mexico ...

  10. Five Northern New Mexico businesses awarded funds to spur economic...

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

    Five Northern New Mexico businesses awarded funds to spur economic growth Five Northern New Mexico businesses awarded funds to spur economic growth The 2016 awardees are ...