National Library of Energy BETA

Sample records for bwr general electric

  1. Electrical equipment performance under severe accident conditions (BWR/Mark 1 plant analysis): Summary report

    SciTech Connect (OSTI)

    Bennett, P.R.; Kolaczkowski, A.M.; Medford, G.T.

    1986-09-01

    The purpose of the Performance Evaluation of Electrical Equipment during Severe Accident States Program is to determine the performance of electrical equipment, important to safety, under severe accident conditions. In FY85, a method was devised to identify important electrical equipment and the severe accident environments in which the equipment was likely to fail. This method was used to evaluate the equipment and severe accident environments for Browns Ferry Unit 1, a BWR/Mark I. Following this work, a test plan was written in FY86 to experimentally determine the performance of one selected component to two severe accident environments.

  2. SPEAR fuel reliability code system. General description. [PWR; BWR

    SciTech Connect (OSTI)

    Christensen, R.

    1980-03-01

    A general description is presented for the SPEAR fuel reliability code system. Included is a discussion of the methodology employed and the structure of the code system, as well as discussion of the major components: the data preparation routines, the mechanistic fuel performance model, the mechanistic cladding failure model, and the statistical failure model.

  3. General Electric | Open Energy Information

    Open Energy Info (EERE)

    General Electric Place: Fairfield, Connecticut Zip: 06828 Region: Northeast - NY NJ CT PA Area Year Founded: 1892 Website: www.ge.com Coordinates: 41.1758333, -73.2719444...

  4. Application of TRAC-BD1/MOD1 to a BWR/4 feedwater control failure ATWS

    SciTech Connect (OSTI)

    Rouhani, S.Z.; Giles, M.M.; Mohr, C.M. Jr.; Weaver, W.L. III

    1984-01-01

    This paper begins with a short description of the Transient Reactor Analysis Code for Boiling Water Reactors (TRAC-BWR), briefly mentioning some of its main features such as specific BWR models and input structure. Next, an input model of a BWR/4 is described, and, the assumptions used in performing an analysis of the loss of a feedwater controller without scram are listed. The important features of the calculated trends in flows, pressure, reactivity, and power are shown graphically and commented in the text. A comparison of some of the main predicted trends with the calculated results from a similar study by General Electric is also presented.

  5. EA-97-C Portland General Electric | Department of Energy

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

    C Portland General Electric EA-97-C Portland General Electric Order authorizing Portland General Electric to export electic energy to Canada PDF icon EA-97-C Portland General ...

  6. SPEAR-BETA fuel performance code system. Volume 1. General description. Final report. [BWR; PWR

    SciTech Connect (OSTI)

    Christensen, R.

    1982-04-01

    This document provides a general description of the SPEAR-BETA fuel reliability code system. Included is a discussion of the methodology employed and the structure of the code system, as well as discussion of the major components: the data preparation routines, the mechanistic fuel performance model, the mechanistic cladding failure model, and the statistical failure model.

  7. Portland General Electric- Heat Pump Rebate Program

    Office of Energy Efficiency and Renewable Energy (EERE)

    Portland General Electric's (PGE) Heat Pump Rebate Program offers residential customers a $200 rebate for an energy-efficient heat pump installed to PGE’s standards by a PGE-approved contractor....

  8. EA-97-B Portland General Electric Company | Department of Energy

    Energy Savers [EERE]

    icon EA-97-B Portland General Electric Company More Documents & Publications EA-380 Freeport Commodities EA-97-D Portland General Electric Company EA-196-A Minnesota Power, Sales

  9. NRC (Nuclear Regulatory Commission) staff evaluation of the General Electric Company Nuclear Reactor Study (''Reed Report'')

    SciTech Connect (OSTI)

    1987-07-01

    In 1975, the General Electric Company (GE) published a Nuclear Reactor Study, also referred to as ''the Reed Report,'' an internal product-improvement study. GE considered the document ''proprietary'' and thus, under the regulations of the Nuclear Regulatory Commission (NRC), exempt from mandatory public disclosure. Nonetheless, members of the NRC staff reviewed the document in 1976 and determined that it did not raise any significant new safety issues. The staff also reached the same conclusion in subsequent reviews. However, in response to recent inquiries about the report, the staff reevaluated the Reed Report from a 1987 perspective. This re-evaluation, documented in this staff report, concluded that: (1) there are no issues raised in the Reed Report that support a need to curtail the operation of any GE boiling water reactor (BWR); (2) there are no new safety issues raised in the Reed Report of which the staff was unaware; and (3) although certain issues addressed by the Reed Report are still being studied by the NRC and the industry, there is no basis for suspending licensing and operation of GE BWR plants while these issues are being resolved.

  10. EA-97-D Portland General Electric Company | Department of Energy

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

    D Portland General Electric Company EA-97-D Portland General Electric Company Order authorizing PGE to export electric energy to Canada. PDF icon EA-97-D PGE (CN).pdf More ...

  11. Current and anticipated use of thermal-hydraulic codes for BWR transient and accident analyses in Japan

    SciTech Connect (OSTI)

    Arai, Kenji; Ebata, Shigeo

    1997-07-01

    This paper summarizes the current and anticipated use of the thermal-hydraulic and neutronic codes for the BWR transient and accident analyses in Japan. The codes may be categorized into the licensing codes and the best estimate codes for the BWR transient and accident analyses. Most of the licensing codes have been originally developed by General Electric. Some codes have been updated based on the technical knowledge obtained in the thermal hydraulic study in Japan, and according to the BWR design changes. The best estimates codes have been used to support the licensing calculations and to obtain the phenomenological understanding of the thermal hydraulic phenomena during a BWR transient or accident. The best estimate codes can be also applied to a design study for a next generation BWR to which the current licensing model may not be directly applied. In order to rationalize the margin included in the current BWR design and develop a next generation reactor with appropriate design margin, it will be required to improve the accuracy of the thermal-hydraulic and neutronic model. In addition, regarding the current best estimate codes, the improvement in the user interface and the numerics will be needed.

  12. Testimonials - Partnerships in R&D - General Electric | Department...

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

    EERE Partnership Testimonials," appear on the screen, followed by "Monte Atwell, General Manager, General Electric" and footage of a man wearing protective glasses in a laboratory. ...

  13. Contacts for the Assistant General Counsel for Electricity and Fossil

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

    Energy | Department of Energy Electricity and Fossil Energy Contacts for the Assistant General Counsel for Electricity and Fossil Energy Office of the Assistant General Counsel for Electricity & Fossil Energy (GC-76) The attorneys in the GC-76 office provide counsel to the Department of Energy Office of Electricity Delivery and Energy Reliability, Office of Energy Policy and Systems Analysis, Office of Fossil Energy, and the Office of Indian Energy Policy and Programs. GC-76 handles the

  14. DOE - Office of Legacy Management -- Pinellas Plant General Electric Co -

    Office of Legacy Management (LM)

    FL 07 Pinellas Plant General Electric Co - FL 07 FUSRAP Considered Sites Site: Pinellas Plant General Electric Co. (FL.07) Designated Name: Alternate Name: Location: Evaluation Year: Site Operations: Site Disposition: Radioactive Materials Handled: Primary Radioactive Materials Handled: Radiological Survey(s): Site Status: Also see Pinellas, Florida, Site Documents Related to Pinellas Plant General Electric Co. Building 100 Area Corrective Measures Study Report Addendum; DOE-LM/GJ1241-2006;

  15. General Electric in India GE | Open Energy Information

    Open Energy Info (EERE)

    navigation, search Name: General Electric in India (GE) Place: New Delhi, Delhi (NCT), India Zip: 110015 Sector: Services, Wind energy Product: String representation...

  16. Inventors Behind General Electric | GE Global Research

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

    Inventors GE Global Research Centers are home to many of the world's brightest, most inquisitive minds in science and technology. Home > Invention > Inventors Sort: Random First Name Last Name Filter: All Aero-Thermal & Mechanical Systems Chemistry & Chemical Engineering Diagnostics, Imaging & Biomedical Technologies Electrical Technologies & Systems Manufacturing & Materials Technologies Software Sciences & Analytics Joseph Vinciquerra Senior Engineer & Manager

  17. Inventors Behind General Electric | GE Global Research

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

    Pierluigi Tenca Pierluigi Tenca Senior Engineer Electric Power Click to email this to a friend (Opens in new window) Share on Facebook (Opens in new window) Click to share (Opens in new window) Click to share on LinkedIn (Opens in new window) Click to share on Tumblr (Opens in new window) ""Because of its unique portfolio of technologies, GE is a fantastic place for multidisciplinary research. This is about learning and exploring."" -Pierluigi Tenca Pierluigi has three

  18. Inventors Behind General Electric | GE Global Research

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

    Betoni Parodi Bruno Betoni Parodi Senior Electrical Engineer Offshore & Subsea Click to email this to a friend (Opens in new window) Share on Facebook (Opens in new window) Click to share (Opens in new window) Click to share on LinkedIn (Opens in new window) Click to share on Tumblr (Opens in new window) "The sea has the potential to bring the world energy from multiple sources, and the technology required to master the space will be more advanced than anything we have seen

  19. Inventors Behind General Electric | GE Global Research

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

    Luiz Felipe Willcox Luiz Felipe Willcox Senior Engineer Smart Systems CoE Click to email this to a friend (Opens in new window) Share on Facebook (Opens in new window) Click to share (Opens in new window) Click to share on LinkedIn (Opens in new window) Click to share on Tumblr (Opens in new window) "I love working in a field that aims to make energy available in all places." -Luiz Felipe Willcox Nothing motivates Luiz Felipe like watching his team succeed. An electrical engineer by

  20. General Electric: ENERGY STAR Referral (PFSF5NFZ****)

    Broader source: Energy.gov [DOE]

    DOE referred the matter of General Electric refrigerator-freezer basic model PFSF5NFZ**** to the EPA for appropriate action after DOE testing showed that the model does not meet the ENERGY STAR specification.

  1. DOE - Office of Legacy Management -- General Electric Co - Shelbyville - IN

    Office of Legacy Management (LM)

    07 Shelbyville - IN 07 FUSRAP Considered Sites Site: General Electric Co - Shelbyville (IN.07 ) Eliminated from consideration under FUSRAP Designated Name: Not Designated Alternate Name: General Electric Plant IN.07-1 Location: Shelbyville , Indiana IN.07-1 Evaluation Year: 1994 IN.07-2 Site Operations: Compacted approximately 500 pounds of thorium (small pieces) into electrodes on 25 and 26 June, 1956. IN.07-1 Site Disposition: Eliminated - Potential for contamination considered remote due

  2. Compilation of corrosion data on CAN-DECON. Volume 1. General, galvanic, crevice, and pitting corrosion data from CANDU and BWR tests. Final report

    SciTech Connect (OSTI)

    Michalko, J.P.; Bonnici, P.J.; Smee, J.L.

    1985-10-01

    Nuclear power station ALARA radiation exposure criteria require, in many cases, decontamination of specific equipment or systems before maintenance, inspection, or work in an adjacent high radiation area. Chemical decontamination, which can be performed away from the high radiation fields, can often best satisfy these ALARA exposure criteria. CAN-DECON, a dilute chemical decontamination process was developed to meet the needs of the Canadian CANDU reactors. It was found to be effective in dissolving BWR oxide films that contain the entrapped radioactive species contributing to high radiation fields. During the development phase of the process and during subsequent field application, CAN-DECON has undergone extensive testing to determine the extent of oxide film dissolution and the degree of corrosion of materials used in construction of reactor components. This has been accomplished on many of the various materials of construction found in the components of the systems decontaminated. Materials tested include carbon steels with range of carbon content 0.1 to 0.4 wt %, 300 series, 400 series, and specialty stainless steels, low alloy steels, and gasket and seal materials. CAN-DECON caused little or no significant corrosion or deterioration on any of the materials tested when applied under conditions appropriate to that class of material. 2 figs., 63 tabs.

  3. BWR containment failure analysis during degraded-core accidents

    SciTech Connect (OSTI)

    Yue, D.D.

    1982-06-06

    This paper presents a containment failure mode analysis during a spectrum of postulated degraded core accident sequences in a typical 1000-MW(e) boiling water reactor (BWR) with a Mark-I wetwell containment. Overtemperature failure of containment electric penetration assemblies (CEPAs) has been found to be the major failure mode during such accidents.

  4. Application to Export Electric Energy OE Docket No. EA-376 Societe Generale

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

    Energy Corp. | Department of Energy 6 Societe Generale Energy Corp. Application to Export Electric Energy OE Docket No. EA-376 Societe Generale Energy Corp. Application from Societe Generale Energy Corp to export electric energy to Canada Application to Export Electric Energy OE Docket No. EA-376 Societe Generale Energy Corp. (2.13 MB) More Documents & Publications EA-376 Societe Generale Energy Corp Application to Export Electric Energy OE Docket No. EA-171-D Powerex Corp. Application

  5. Development of BWR plant analyzer

    SciTech Connect (OSTI)

    Wulff, W.; Cheng, H.S.; Lekach, S.V.; Stritar, A.; Mallen, A.N.

    1984-01-01

    The BWR Plant Analyzer has been developed for realistic and accurate simulations of normal and severe abnormal transients in BWR power plants at high simulation speeds, low capital and operating costs and with outstanding user conveniences. The simulation encompasses neutron kinetics, heat conduction in fuel structures, nonequilibrium, nonhomogeneous coolant dynamics, steam line acoustics, and the dynamics of turbines, condensers, feedwater pumps and heaters, of the suppression pool, the control systems and the plant protection systems. These objectives have been achieved. Advanced modeling, using extensively analytical integration and dynamic evaluation of analytical solutions, has been combined with modern minicomputer technology for high-speed simulation of complex systems. The High-Speed Interactive Plant Analyzer code HIPA-BWR has been implemented on the AD10 peripheral parallel processor.

  6. PPPL lends General Electric a hand in developing an advanced power switch |

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

    Princeton Plasma Physics Lab PPPL lends General Electric a hand in developing an advanced power switch By John Greenwald August 28, 2014 Tweet Widget Google Plus One Share on Facebook Laboratory test of a liquid-metal cathode. (Photo by General Electric Co.) Laboratory test of a liquid-metal cathode. Scientists at the U.S. Department of Energy's (DOE) Princeton Plasma Physics Laboratory (PPPL) are assisting General Electric Co. in developing an electrical switch that could help lower utility

  7. PPPL lends General Electric a hand in developing an advanced power switch |

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

    Princeton Plasma Physics Lab lends General Electric a hand in developing an advanced power switch By John Greenwald August 28, 2014 Tweet Widget Google Plus One Share on Facebook Laboratory test of a liquid-metal cathode. (Photo by General Electric Co. ) Laboratory test of a liquid-metal cathode. Scientists at the U.S. Department of Energy's (DOE) Princeton Plasma Physics Laboratory (PPPL) are assisting General Electric Co. in developing an electrical switch that could help lower utility

  8. Office of the Assistant General Counsel for Electricity and Fossil Energy

    Broader source: Energy.gov [DOE]

    The Office of the Assistant General Counsel for Electricity and Fossil Energy (GC-76) provides legal support and advice, and policy guidance, to the Department on electricity, fossil energy, energy...

  9. Office of the Assistant General Counsel for Electricity and Fossil...

    Office of Environmental Management (EM)

    Further, the office represents the consumer interests of the United States, including national laboratories, military bases, and certain NNSA facilities, in electric rate ...

  10. Contacts for the Assistant General Counsel for Electricity and...

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

    Subject Matter Attorney Contacts Electricity Delivery Energy Reliability Energy Emergency Lot Cooke 202-586-0503 Critical Infrastructure Protection Becca Smith 202-586-6335 ...

  11. DOE to Develop Multi-Megawatt Offshore Wind Turbine with General Electric |

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

    Department of Energy to Develop Multi-Megawatt Offshore Wind Turbine with General Electric DOE to Develop Multi-Megawatt Offshore Wind Turbine with General Electric March 9, 2006 - 11:44am Addthis Contract Valued at $27 million, supports President Bush's Advanced Energy Initiative WASHINGTON, D.C. - The U.S. Department of Energy's (DOE) National Renewable Energy Laboratory (NREL) in Golden, Colorado, has signed a $27 million, multi-year contract with the General Electric Company (GE) to

  12. Low Wind Speed Technology Phase II: Integrated Wind Energy/Desalination System; General Electric Global Research

    SciTech Connect (OSTI)

    Not Available

    2006-03-01

    This fact sheet describes a subcontract with General Electric Global Research to explore wind power as a desirable option for integration with desalination technologies.

  13. Synergistic failure of BWR internals

    SciTech Connect (OSTI)

    A. G. Ware; T. Y. Chang

    1999-10-25

    Boiling Water Reactor (BWR) core shrouds and other reactor internals important to safety are experiencing intergranular stress corrosion cracking (IGSCC). The United States Nuclear Regulatory Commission has followed the problem, and as part of its investigations, contracted with the Idaho National Engineering and Environmental Laboratory to conduct a risk assessment. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, with specific consideration given to potential cascading and common mode effects. An initial phase has been completed in which background material was gathered and evaluated, and potential accident sequences were identified. A second phase is underway to perform a simplified, quantitative probabilistic risk assessment on a representative high-power BWR/4. Results of the initial study conducted on the jet pumps show that any cascading failures would not result in a significant increase in the core damage frequency. The methodology is currently being extended to other major reactor internals components.

  14. Synergistic Failure of BWR Internals

    SciTech Connect (OSTI)

    Ware, Arthur Gates; Chang, T-Y

    1999-10-01

    Boiling Water Reactor (BWR) core shrouds and other reactor internals important to safety are experiencing intergranular stress corrosion cracking (IGSCC). The United States Nuclear Regulatory Commission has followed the problem, and as part of its investigations, contracted with the Idaho National Engineering and Environmental Laboratory to conduct a risk assessment. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, with specific consideration given to potential cascading and common mode effects. An initial phase has been completed in which background material was gathered and evaluated, and potential accident sequences were identified. A second phase is underway to perform a simplified, quantitative probabilistic risk assessment on a representative high-power BWR/4. Results of the initial study conducted on the jet pumps show that any cascading failures would not result in a significant increase in the core damage frequency. The methodology is currently being extended to other major reactor internals components.

  15. PPPL lends General Electric a hand in developing an advanced...

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

    a smarter, more advanced, more reliable, and more secure electric grid," according to the DOE's Advanced Research Projects Agency-Energy (ARPA-E), which is funding the GE project. ...

  16. (Boiling water reactor (BWR) CORA experiments)

    SciTech Connect (OSTI)

    Ott, L.J.

    1990-10-16

    To participate in the 1990 CORA Workshop at Kernforschungszentrum Karlsruhe (KfK) GmbH, Karlsruhe, FRG, on October 1--4, and to participate in detailed discussions on October 5 with the KfK CORA Boiling Water Reactor (BWR) experiments. The traveler attended the 1990 CORA Workshop at KfK, FRG. Participation included the presentation of a paper on work performed by the Boiling Water Reactor Core Melt Progression Phenomena Program at Oak Ridge National Laboratory (ORNL) on posttest analyses of CORA BWR experiments. The Statement of Work (November 1989) for the BWR Core Melt Progression Phenomena Program provides for pretest and posttest analyses of the BWR CORA experiments performed at KfK. Additionally, it is intended that ORNL personnel participate in the planning process for future CORA BWR experiments. For these purposes, meetings were held with KfK staff to discuss such topics as (1) experimental test schedule, (2) BWR test conduct, (3) perceived BWR experimental needs, and (4) KfK operational staff needs with respect to ORNL support. 19 refs.

  17. Halbach array motor/generators: A novel generalized electric machine

    SciTech Connect (OSTI)

    Merritt, B.T.; Post, R.F.; Dreifuerst, G.R.; Bender, D.A.

    1995-02-01

    For many years Klaus Halbach has been investigating novel designs for permanent magnet arrays, using advanced analytical approaches and employing a keen insight into such systems. One of his motivations for this research was to find more efficient means for the utilization of permanent magnets for use in particle accelerators and in the control of particle beams. As a result of his pioneering work, high power free-electron laser systems, such as the ones built at the Lawrence Livermore Laboratory, became feasible, and his arrays have been incorporated into other particle-focusing systems of various types. This paper reports another, quite different, application of Klaus` work, in the design of high power, high efficiency, electric generators and motors. When tested, these motor/generator systems display some rather remarkable properties. Their success derives from the special properties which these arrays, which the authors choose to call {open_quotes}Halbach arrays,{close_quotes} possess.

  18. Status update of the BWR cask simulator

    SciTech Connect (OSTI)

    Lindgren, Eric R.; Durbin, Samuel G.

    2015-09-01

    The performance of commercial nuclear spent fuel dry storage casks are typically evaluated through detailed numerical analysis of the system's thermal performance. These modeling efforts are performed by the vendor to demonstrate the performance and regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have moved the storage location from above ground to below ground and significantly increased the maximum thermal load allowed in a cask in part by increasing the canister helium pressure. Previous cask performance validation testing did not capture these parameters. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern dry casks. These modern cask designs utilize elevated helium pressure in the sealed canister or are intended for subsurface storage. The BWR cask simulator (BCS) has been designed in detail for both the above ground and below ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below ground storage configurations of

  19. BWR Assembly Optimization for Minor Actinide Recycling

    SciTech Connect (OSTI)

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

    2010-03-22

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

  20. Validation of SCALE (SAS2H) isotopic predictions for BWR spent fuel

    SciTech Connect (OSTI)

    Hermann, O.W.; DeHart, M.D.

    1998-09-01

    Thirty spent fuel samples obtained from boiling-water-reactor (BWR) fuel pins have been modeled at Oak Ridge National Laboratory using the SAS2H sequence of the SCALE code system. The SAS2H sequence uses transport methods combined with the depletion and decay capabilities of the ORIGEN-S code to estimate the isotopic composition of fuel as a function of its burnup history. Results of these calculations are compared with chemical assay measurements of spent fuel inventories for each sample. Results show reasonable agreement between measured and predicted isotopic concentrations for important actinides; however, little data are available for most fission products considered to be important for spent fuel concerns (e.g., burnup credit, shielding, source-term calculations, etc.). This work is a follow-up to earlier works that studied the ability to predict spent fuel compositions in pressurized-water-reactor (PWR) fuel pins. Biases and uncertainties associated with BWR isotopic predictions are found to be larger than those of PWR calculations. Such behavior is expected, as the operation of a BWR is significantly more complex than that of a PWR plant, and in general the design of a BWR has a more heterogeneous configuration than that of a PWR. Nevertheless, this work shows that the simple models employed using SAS2H to represent such complexities result in agreement to within 5% (and often less than 1%) or less for most nuclides important for spent fuel applications. On the other hand, however, the set of fuel samples analyzed represent a small subset of the BWR fuel population, and results reported herein may not be representative of the full population of BWR spent fuel.

  1. Validation of SCALE (SAS2H) Isotopic Predictions for BWR Spent Fuel

    SciTech Connect (OSTI)

    Hermann, O.W.

    1998-01-01

    Thirty spent fuel samples obtained from boiling-water-reactor (BWR) fuel pins have been modeled at Oak Ridge National Laboratory using the SAS2H sequence of the SCALE code system. The SAS2H sequence uses transport methods combined with the depletion and decay capabilities of the ORIGEN-S code to estimate the isotopic composition of fuel as a function of its burnup history. Results of these calculations are compared with chemical assay measurements of spent fuel inventories for each sample. Results show reasonable agreement between measured and predicted isotopic concentrations for important actinides; however, little data are available for most fission products considered to be important for spent fuel concerns (e.g., burnup credit, shielding, source-term calculations, etc.). This work is a follow-up to earlier works that studied the ability to predict spent fuel compositions in pressurized-water-reactor (PWR) fuel pins. Biases and uncertainties associated with BWR isotopic predictions are found to be larger than those of PWR calculations. Such behavior is expected, as the operation of a BWR is significantly more complex than that of a PWR plant, and in general the design of a BWR has a more heterogeneous configuration than that of a PWR. Nevertheless, this work shows that the simple models employed using SAS2H to represent such complexities result in agreement to within 5% (and often less than 1%) or less for most nuclides important for spent fuel applications. On the other hand, however, the set of fuel samples analyzed represent a small subset of the BWR fuel population, and results reported herein may not be representative of the full population of BWR spent fuel.

  2. Correlation for predicting reactor power during a BWR ATWS

    SciTech Connect (OSTI)

    Chexal, B.; Layman, W.

    1986-01-01

    An anticipated transient without scram (ATWS), while of low probability, has received much attention because of its potentially serious consequences. Under certain ATWS sequences for a boiling water reactor (BWR), it would be desirable to reduce system power, particularly where the primary system has been isolated by closure of all main steam isolation valves and is discharging steam through its safety/relief valve system to the suppression pool. Reducing reactor power increases the time available to shut down the reactor by minimizing the heat dumped to the suppression pool and thereby helping to keep the suppression pool temperature and the containment stresses with limits. This paper describes the correlation developed to provide the degree of power reduction which can be achieved for a BWR during ATWS conditions by intentionally reducing the system coolant inventory and pressure. Under proposed emergency procedure guidelines for the ATWS event, the reactor water level would be lowered to reduce system power. The correlation is based on three-dimensional quasi-static analysis based on Electric Power Research Institute computer codes NATBWR and verified by the transient code RETRAN-02.

  3. Results from the Operational Testing of the General Electric Smart Grid Capable Electric Vehicle Supply Equipment (EVSE)

    SciTech Connect (OSTI)

    Richard Barney Carlson; Don Scoffield; Brion Bennett

    2013-12-01

    The Idaho National Laboratory conducted testing and analysis of the General Electric (GE) smart grid capable electric vehicle supply equipment (EVSE), which was a deliverable from GE for the U.S. Department of Energy FOA-554. The Idaho National Laboratory has extensive knowledge and experience in testing advanced conductive and wireless charging systems though INL’s support of the U.S. Department of Energy’s Advanced Vehicle Testing Activity. This document details the findings from the EVSE operational testing conducted at the Idaho National Laboratory on the GE smart grid capable EVSE. The testing conducted on the EVSE included energy efficiency testing, SAE J1772 functionality testing, abnormal conditions testing, and charging of a plug-in vehicle.

  4. BWR plant analyzer development at BNL

    SciTech Connect (OSTI)

    Cheng, H.S.; Wulff, W.; Mallen, A.N.; Lekach, S.V.; Stritar, A.; Cerbone, R.J.

    1985-01-01

    Advanced technology for high-speed interactive nuclear power plant simulations is of great value for timely resolution of safety issues, for plant monitoring, and for computer-aided emergency responses to an accident. Presented is the methodology employed at BNL to develop a BWR plant analyzer capable of simulating severe plant transients at much faster than real-time process speeds. Five modeling principles are established and a criterion is given for selecting numerical procedures and efficient computers to achieve the very high simulation speeds. Typical results are shown to demonstrate the modeling fidelity of the BWR plant analyzer.

  5. 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.

  6. DOE - Office of Legacy Management -- General Electric Co - San Jose - CA 13

    Office of Legacy Management (LM)

    San Jose - CA 13 FUSRAP Considered Sites Site: General Electric Co. - San Jose (CA.13 ) Eliminated from consideration under FUSRAP Designated Name: Not Designated Alternate Name: None Location: San Jose , California CA.13-1 Evaluation Year: 1995 CA.13-2 Site Operations: Fabricated uranium metal. CA.13-1 Site Disposition: Eliminated - No Authority - NRC licensed CA.13-2 Radioactive Materials Handled: Yes Primary Radioactive Materials Handled: Uranium CA.13-1 Radiological Survey(s): No Site

  7. INEL BWR severe accidnet ATWS study

    SciTech Connect (OSTI)

    Jouse, W.C.

    1983-01-01

    The subject of this study is a postulate Anticipated Transient Without Scram (ATWS) at unit one of the Browns Ferry nuclear plant, a boiling water reactor (BWR). The development work is being conducted at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC). It has long been recognized that the dominant ATWS transient in BWRs is the main steamline isolation valve (MSIV) closure pressurization type of event. The analytic tool used in this study is RELAP5/MOD1.6. This version of RELAP5 has the capability to simulate BWR plants in that several special process models, such as a jet pump momentum mixer model, have been installed.

  8. Steam Line Break and Station Blackout Transients for Proliferation Resistant Hexagonal Tight Lattice BWR

    SciTech Connect (OSTI)

    Upendra Rohatgi; Jae Jo; Bub Dong Chung; Hiroshi Takahashi [Brookhaven National Laboratory, Energy Sciences and Technology Department, Upton, New York 11973 (United States); Downar, T.J. [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47906-1290 (United States)

    2002-07-01

    Safety analyses of a proliferation resistant, economically competitive, high conversion, boiling water reactor (HCBWR) fueled with fissile plutonium and fertile thorium oxide fuel elements, and with passive safety systems are presented here. The HCBWR developed here is characterized by a very tight lattice with a relatively small water volume fraction in the core which therefore operates with a fast reactor neutron spectrum, and a considerably improved neutron economy compared to the current generation of Light Water Reactors. A tight lattice BWR core has very narrow flow channels with a hydraulic diameter less than half of the regular BWR core. The tight lattice core presented a special challenge to core cooling, because of reduced water inventory and high friction in the core. The primary safety concern when reducing the moderator to fuel ratio and when using a tightly packed lattice arrangement is to maintain adequate cooling of the core during both normal operation and accident scenarios. In the preliminary HCBWR design, the core has been placed in a vessel with a large chimney section, and the vessel is connected with an Isolation Cooling System (ICS). The vessel is placed in a containment with a Gravity Driven Cooling System (GDCS) and a Passive Containment Cooling System (PCCS) in a configuration similar to General Electric's (GE) Simplified Boiling Water Reactor (SBWR). The safety systems are similar to the SBWR; the ICS and PCCS are scaled with power. An internal recirculation pump was placed in the downcomer to augment the buoyancy head provided by the chimney. The buoyancy provided by the chimney alone could not generate sufficient recirculation in the vessel since the tight lattice configuration resulted in much larger friction in the core than the SBWR. A modified RELAP5 Code was used to simulate and analyze two of the most limiting events for a tight pitch lattice core: the Station Blackout and the Main Steam Line Break events. The constitutive

  9. Generalized Bohms criterion and negative anode voltage fall in electric discharges

    SciTech Connect (OSTI)

    Londer, Ya. I.; Ulyanov, K. N.

    2013-10-15

    The value of the voltage fall across the anode sheath is found as a function of the current density. Analytic solutions are obtained in a wide range of the ratio of the directed velocity of plasma electrons v{sub 0} to their thermal velocity v{sub T}. It is shown that the voltage fall in a one-dimensional collisionless anode sheath is always negative. At the small values of v{sub 0}/v{sub T}, the obtained expression asymptotically transforms into the Langmuir formula. Generalized Bohms criterion for an electric discharge with allowance for the space charge density ?(0), electric field E(0), ion velocity v{sub i}(0), and ratio v{sub 0}/v{sub T} at the plasma-sheath interface is formulated. It is shown that the minimum value of the ion velocity v{sub i}{sup *}(0) corresponds to the vanishing of the electric field at one point inside the sheath. The dependence of v{sub i}{sup *} (0) on ?(0), E(0), and v{sub 0}/v{sub T} determines the boundary of the existence domain of stationary solutions in the sheath. Using this criterion, the maximum possible degree of contraction of the electron current at the anode is determined for a short high-current vacuum arc discharge.

  10. ELECTRIC

    Office of Legacy Management (LM)

    ELECTRIC cdrtrokArJclaeT 3 I+ &i, y$ \I &OF I*- j< t j,fci..- ir )(yiT !E-li, ( \-,v? Cl -p/4.4 RESEARCH LABORATORIES EAST PITTSBURGH, PA. 8ay 22, 1947 Mr. J. Carrel Vrilson General ?!!mager Atomic Qxzgy Commission 1901 Constitution Avenue Kashington, D. C. Dear Sir: In the course of OUT nuclenr research we are planning to study the enc:ri;y threshold anti cross section for fission. For thib program we require a s<>piAroted sample of metallic Uranium 258 of high purity. A

  11. LBB application in Swedish BWR design

    SciTech Connect (OSTI)

    Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P.

    1997-04-01

    The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as these installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions.

  12. Evaluation of BWR emergency procedure guidelines for BWR ATWS using RAMONA-3B code

    SciTech Connect (OSTI)

    Neymotin, L.; Slovik, G.; Cazzoli, E.; Saha, P.

    1985-01-01

    An MSIV Closure ATWS calculation for a typical BWR/4 (Browns Ferry, Unit 1) was performed using the RAMONA-3B code which is a BWR systems transient code combining three-dimensional neutronic core representation with multi-channel one-dimensional thermal hydraulics. The main objective of the study was to perform a best-estimate evaluation of the recently proposed Emergency Procedure Guidelines for Anticipated Transients Without Scram (ATWS). Emphasis was placed on evaluating the effects of lowering the downcomer water level to the Top of Active Fuel (TAF) and vessel depressurization. The calculation was run up to approximately 1200 seconds. Both actions, namely, lowering the water level and vessel depressurization, lowered the reactor power to some extent. However, the pressure suppression pool water temperature still reached approximately 90/sup 0/C (potential High Pressure Coolant Injection (HPCI) pump seal failure temperature) in twenty minutes. Thus, other actions such as boron injection and/or manual control rod insertion are necessary to mitigate a BWR/4 Main Steam Isolation Valve (MSIV) closure ATWS. 19 refs., 14 figs., 3 tabs.

  13. Evaluation of ATWS core damage frequency for a BWR/4

    SciTech Connect (OSTI)

    Shiu, K.; Ilberg, D.; Hanan, N.

    1985-01-01

    This paper reports a study performed to evaluate the core damage frequency contribution from Anticipated Transient Without Scram (ATWS) in a BWR/4 plant. Discussions on improvements in the design and operation of BWR plants to reduce the likelihood of occurrence and core damage frequency of ATWS have continued for years. In November 1981, subsequent to the issuance of three alternate proposed ATWS rules, the Nuclear Regulatory Commission invited comments on these rules. In June 1984, a final rule on the reduction of risk from ATWS events was issued. In the study, it is assumed that the BWR/4 reactor is of an earlier vintage. However, only two of the modifications have been implemented in accordance with the final rule: a diverse scram system and automatic recirculation pump trip. It is further assumed that the setpoint for Main Steam Isolation Valves (MSIVs) closure is at reactor pressure vessel (RPV) water level 1 and that the BWR emergency procedure guidelines are implemented.

  14. Reassessment of the BWR scram failure probability

    SciTech Connect (OSTI)

    Burns, E.T.

    1989-01-01

    As part of the Severe Accident Policy Statement implementation, the probabilistic quantification of accident sequence frequencies that may lead to core damage is a key element in demonstrating a plant's safety status relative to US Nuclear Regulatory Commission (NRC) staff goals. One of the key quantitative inputs in a boiling water reactor (BWR) probabilistic risk assessment is the probability of a failure to scram. The assessment of this failure probability has been the subject of a long and continuing debate over the adequacy of available data and analytic modeling. This report provides a summary of the status of this debate, including the latest data, and provides a revision to the characterization of the failure probability originally published in NUREG 0460 and the Utility Group on Anticipated Transient Without Scram (ATWS) Petition.

  15. Uncertainty in BWR power during ATWS events

    SciTech Connect (OSTI)

    Diamond, D.J.

    1986-01-01

    A study was undertaken to improve our understanding of BWR conditions following the closure of main steam isolation valves and the failure of reactor trip. Of particular interest was the power during the period when the core had reached a quasi-equilibrium condition with a natural circulation flow rate determined by the water level in the downcomer. Insights into the uncertainity in the calculation of this power with sophisticated computer codes were quantified using a simple model which relates power to the principal thermal-hydraulic variables and reactivity coefficients; the latter representing the link between the thermal-hydraulics and the neutronics. Assumptions regarding the uncertainty in these variables and coefficients were then used to determine the uncertainty in power.

  16. BWR/4 loss of feedwater transient analysis

    SciTech Connect (OSTI)

    Lu, M.S.; Levine, M.M.; Shier, W.G.

    1983-01-01

    This paper presents an analysis of a series of loss of feedwater (LOF) transients for a typical BWR/4 reactor. These calculations were prompted by the events that occurred during the TMI incident and hence include various assumed failures in the safety/relief valve system and the assumed inoperability of various safety systems. This analysis provides transient results necessary to evaluate the potential for core uncovery and excessive average fuel temperatures which can then be used in the evaluation of the adequacy of the engineered safety features and the plant operating procedures. The RELAP5/MOD1 computer code was used for this analysis. The version of the code is designated as Cycle 13 with additional modifications provided by INEL. The modifications affect the jet pump model, interphase drag model and separator model.

  17. BWR Core Heat Transfer Code System.

    Energy Science and Technology Software Center (OSTI)

    1999-04-27

    Version 00 MOXY is used for the thermal analysis of a planar section of a boiling water reactor (BWR) fuel element during a loss-of-coolant accident (LOCA). The code emplyoys models that describe heat transfer by conduction, convection, and thermal radiation, and heat generation by metal-water reaction and fission product decay. Models are included for considering fuel-rod swelling and rupture, energy transport across the fuel-to-cladding gap, and the thermal response of the canister. MOXY requires thatmore » time-dependent data during the blowdown process for the power normalized to the steady-state power, for the heat-transfer coefficient, and for the fluid temperature be provided as input. Internal models provide these parameters during the heatup and emergency cooling phases.« less

  18. Radioactive waste shipments to Hanford Retrievable Storage from the General Electric Vallecitos Nuclear Center, Pleasanton, California

    SciTech Connect (OSTI)

    Vejvoda, E.J.; Pottmeyer, J.A.; DeLorenzo, D.S.; Weyns-Rollosson, M.I.; Duncan, D.R.

    1993-10-01

    During the next two decades the transuranic (TRU) wastes now stored in the burial trenches and storage facilities at the Hanford Site are to be retrieved, processed at the Waste Receiving and Processing Facility, and shipped to the Waste Isolation Pilot Plant near Carlsbad, New Mexico for final disposal. Approximately 3.8% of the TRU waste to be retrieved for shipment to WIPP was generated at the General Electric (GE) Vallecitos Nuclear Center (VNC) in Pleasanton, California and shipped to the Hanford Site for storage. The purpose of this report is to characterize these radioactive solid wastes using process knowledge, existing records, and oral history interviews. The waste was generated almost exclusively from the activities, of the Plutonium Fuels Development Laboratory and the Plutonium Analytical Laboratory. Section 2.0 provides further details of the VNC physical plant, facility operations, facility history, and current status. The solid radioactive wastes were associated with two US Atomic Energy Commission/US Department of Energy reactor programs -- the Fast Ceramic Reactor (FCR) program, and the Fast Flux Test Reactor (FFTR) program. These programs involved the fabrication and testing of fuel assemblies that utilized plutonium in an oxide form. The types and estimated quantities of waste resulting from these programs are discussed in detail in Section 3.0. A detailed discussion of the packaging and handling procedures used for the VNC radioactive wastes shipped to the Hanford Site is provided in Section 4.0. Section 5.0 provides an in-depth look at this waste including the following: weight and volume of the waste, container types and numbers, physical description of the waste, radiological components, hazardous constituents, and current storage/disposal locations.

  19. BWR Source Term Generation and Evaluation

    SciTech Connect (OSTI)

    J.C. Ryman

    2003-07-31

    This calculation is a revision of a previous calculation (Ref. 7.5) that bears the same title and has the document identifier BBAC00000-01717-0210-00006 REV 01. The purpose of this revision is to remove TBV (to-be-verified) -41 10 associated with the output files of the previous version (Ref. 7.30). The purpose of this and the previous calculation is to generate source terms for a representative boiling water reactor (BWR) spent nuclear fuel (SNF) assembly for the first one million years after the SNF is discharged from the reactors. This calculation includes an examination of several ways to represent BWR assemblies and operating conditions in SAS2H in order to quantify the effects these representations may have on source terms. These source terms provide information characterizing the neutron and gamma spectra in particles per second, the decay heat in watts, and radionuclide inventories in curies. Source terms are generated for a range of burnups and enrichments (see Table 2) that are representative of the waste stream and stainless steel (SS) clad assemblies. During this revision, it was determined that the burnups used for the computer runs of the previous revision were actually about 1.7% less than the stated, or nominal, burnups. See Section 6.6 for a discussion of how to account for this effect before using any source terms from this calculation. The source term due to the activation of corrosion products deposited on the surfaces of the assembly from the coolant is also calculated. The results of this calculation support many areas of the Monitored Geologic Repository (MGR), which include thermal evaluation, radiation dose determination, radiological safety analyses, surface and subsurface facility designs, and total system performance assessment. This includes MGR items classified as Quality Level 1, for example, the Uncanistered Spent Nuclear Fuel Disposal Container (Ref. 7.27, page 7). Therefore, this calculation is subject to the requirements of the

  20. RELAP-7: Demonstrating the integration of two-phase flow components for an ideal BWR loop

    SciTech Connect (OSTI)

    Hongbin Zhang; Haihua Zhao; Ling Zou; David Andrs; John Peterson; Ray Berry; Richard Martineua

    2013-06-01

    This is DOE Level 3 milestone report documenting RELAP-7's capability to simulate an ideal BWR loop.

  1. Analysis of the MSIV closure ATWS in a BWR/6

    SciTech Connect (OSTI)

    Pan, Chin; Chen, Genshun; Hsiue, J.K.; Liaw, T.J.

    1988-01-01

    Anticipated transient without scram (ATWS) has received much attention since the beginning of the last decade. It has been recognized as a dominant accident sequence for possible core melt and containment damage in a boiling water reactor (BWR) power plant. In the literature, a great deal of study has been reported on the investigation of BWR behavior during an ATWS, especially for BWR/2 and BWR/4. The objective of this study is to assess reactor behavior during a main steam isolation valve (MSIV) closure ATWS for the Kuosheng nuclear power plant of Taiwan Power Company, which has two units of BWR/6 with Mark-III containment. The analyses were performed using RETRAN-02/MOD3, which solves a one-dimensional homogeneous equilibrium model of two-phase mixture. The relative motion between two phases is treated by the dynamic slip model. Due to the unavailability of appropriate three-dimensional neutronic codes, such as SIMULATE-E, in this country, the option of point kinetics has been adopted for calculations. This approach is not expected to give the exact power shape variation during the transient, but is believed to be appropriate for providing the accurate overall power level, which is the major concern for core and containment integrity.

  2. BWR ASSEMBLY SOURCE TERMS FOR WASTE PACKAGE DESIGN

    SciTech Connect (OSTI)

    T.L. Lotz

    1997-02-15

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide boiling water reactor (BWR) assembly radiation source term data for use during Waste Package (WP) design. The BWR assembly radiation source terms are to be used for evaluation of radiolysis effects at the WP surface, and for personnel shielding requirements during assembly or WP handling operations. The objectives of this evaluation are to generate BWR assembly radiation source terms that bound selected groupings of BWR assemblies, with regard to assembly average burnup and cooling time, which comprise the anticipated MGDS BWR commercial spent nuclear fuel (SNF) waste stream. The source term data is to be provided in a form which can easily be utilized in subsequent shielding/radiation dose calculations. Since these calculations may also be used for Total System Performance Assessment (TSPA), with appropriate justification provided by TSPA, or radionuclide release rate analysis, the grams of each element and additional cooling times out to 25 years will also be calculated and the data included in the output files.

  3. Preliminary Risk Assessment Associated with IGSCC of BWR Vessel Internals

    SciTech Connect (OSTI)

    A. Ware; K. Morton; M. Nitzel; N. Chokshi; T-Y. Chang

    1999-08-01

    BWR core shrouds and other reactor internals important to safety are experiencing intergranular stress corrosion cracking (IGSCC). The United States Nuclear Regulatory Commission (NRC) has followed the problem, and as part of its investigations, contracted with the Idaho National Engineering and Environmental Laboratory (INEEL) to conduct a risk assessment. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, with specific consideration given to potential cascading and common mode effects. The paper presents an overview of the program, discusses the results of a preliminary qualitative assessment, and summarizes a simplified risk assessment that was conducted on sequences resulting from failures of jet pump components of a BWR/4 plant.

  4. Trace Assessment for BWR ATWS Analysis

    SciTech Connect (OSTI)

    Cheng, L.Y.; Diamond, D.; Arantxa Cuadra, Gilad Raitses, Arnold Aronson

    2010-04-22

    A TRACE/PARCS input model has been developed in order to be able to analyze anticipated transients without scram (ATWS) in a boiling water reactor. The model is based on one developed previously for the Browns Ferry reactor for doing loss-of-coolant accident analysis. This model was updated by adding the control systems needed for ATWS and a core model using PARCS. The control systems were based on models previously developed for the TRAC-B code. The PARCS model is based on information (e.g., exposure and moderator density (void) history distributions) obtained from General Electric Hitachi and cross sections for GE14 fuel obtained from an independent source. The model is able to calculate an ATWS, initiated by the closure of main steam isolation valves, with recirculation pump trip, water level control, injection of borated water from the standby liquid control system and actuation of the automatic depres-surization system. The model is not considered complete and recommendations are made on how it should be improved.

  5. STEAM LINE BREAK AND STATION BLACKOUT TRANSIENTS FOR PROLIFERATION RESISTANT HEXAGONAL TIGHT LATTICE BWR.

    SciTech Connect (OSTI)

    ROHATGI,U.S.; JO,J.; CHUNG,B.D.; TAKAHASHI,H.

    2002-06-09

    Safety analyses of a proliferation resistant, economically competitive, high conversion, boiling water reactor (HCBWR) fueled with fissile plutonium and fertile thorium oxide fuel elements, and with passive safety systems are presented here. The HCBWR developed here is characterized by a very tight lattice with a relatively small water volume fraction in the core which therefore operates with a fast reactor neutron spectrum, and a considerably improved neutron economy compared to the current generation of Light Water Reactors. The tight lattice core has a very narrow flow channels with a hydraulic diameter less than half of the regular BWR core and, thus, presents a special challenge to core cooling, because of reduced water inventory and high friction in the core. The primary safety concern when reducing the moderator to fuel ratio and when using a tightly packed lattice arrangement is to maintain adequate cooling of the core during both normal operation and accident scenarios. In the preliminary HCBWR design, the core has been placed in a vessel with a large chimney section, and the vessel is connected with Isolation Condenser System (ICs). The vessel is placed in containment with Gravity Driven Cooling System (GDCS) and Passive Containment Cooling System (PCCS) in a configuration similar to General Electric's Simplified Boiling Water Reactor (SBWR). The safety systems are similar to SBWR; ICs and PCCS are scaled with power. An internal recirculation pump was placed in the downcomer to augment the buoyancy head provided by the chimney, since the buoyancy provided by the chimney alone could not generate sufficient recirculation in the vessel as the tight lattice configuration resulted in much larger friction in the core than the SBWR. The constitutive relationships for RELAP5 were assessed for narrow channels, and as a result the heat transfer package was modified. The modified RELAP5 was used to simulate and analyze two of the most limiting events for a tight

  6. The Advanced BWR Nuclear Plant: Safe, economic nuclear energy

    SciTech Connect (OSTI)

    Redding, J.R.

    1994-12-31

    The safety and economics of Advanced BWR Nuclear Power Plants are outlined. The topics discussed include: ABWR Programs: status in US and Japan; ABWR competitiveness: safety and economics; SBWR status; combining ABWR and SBWR: the passive ABWR; and Korean/GE partnership.

  7. Simulations of Cyclic Voltammetry for Electric Double Layers in Asymmetric Electrolytes: A Generalized Modified Poisson-Nernst-Planck Model

    SciTech Connect (OSTI)

    Wang, Hainan; Thiele, Alexander; Pilon, Laurent

    2013-11-15

    This paper presents a generalized modified Poisson–Nernst–Planck (MPNP) model derived from first principles based on excess chemical potential and Langmuir activity coefficient to simulate electric double-layer dynamics in asymmetric electrolytes. The model accounts simultaneously for (1) asymmetric electrolytes with (2) multiple ion species, (3) finite ion sizes, and (4) Stern and diffuse layers along with Ohmic potential drop in the electrode. It was used to simulate cyclic voltammetry (CV) measurements for binary asymmetric electrolytes. The results demonstrated that the current density increased significantly with decreasing ion diameter and/or increasing valency |zi| of either ion species. By contrast, the ion diffusion coefficients affected the CV curves and capacitance only at large scan rates. Dimensional analysis was also performed, and 11 dimensionless numbers were identified to govern the CV measurements of the electric double layer in binary asymmetric electrolytes between two identical planar electrodes of finite thickness. A self-similar behavior was identified for the electric double-layer integral capacitance estimated from CV measurement simulations. Two regimes were identified by comparing the half cycle period τCV and the “RC time scale” τRC corresponding to the characteristic time of ions’ electrodiffusion. For τRC ← τCV, quasi-equilibrium conditions prevailed and the capacitance was diffusion-independent while for τRC → τCV, the capacitance was diffusion-limited. The effect of the electrode was captured by the dimensionless electrode electrical conductivity representing the ratio of characteristic times associated with charge transport in the electrolyte and that in the electrode. The model developed here will be useful for simulating and designing various practical electrochemical, colloidal, and biological systems for a wide range of applications.

  8. EA-1869: Supplement to General Motors Corp., Electric Vehicle/Battery Manufacturing Application, White Marsh, Maryland, and Wixom, Michigan (DOE/EA-1723-S1)

    Broader source: Energy.gov [DOE]

    Based on the analysis in the Environmental Assessment DOE determined that its proposed action, to award a federal grant to General Motors to establish an electric motor components manufacturing and electric drive assembly facility would result in no significant adverse impacts.

  9. P.C. 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; BWR...

    Office of Scientific and Technical Information (OSTI)

    Erosioncorrosion-induced pipe wall thinning in US Nuclear Power Plants Wu, P.C. 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; BWR TYPE REACTORS; PIPES; CORROSION; EROSION;...

  10. 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.

  11. Calcium/calcium chromate thermal battery and thermal battery assignment at the General Electric Neutron Devices Department

    SciTech Connect (OSTI)

    Neale, J.B.; Walton, R.D.

    1980-10-10

    A nontechnical overview of thermal battery design and fabrication methods is given, along with a description of the role of the General Electric Neutron Devices Department (GEND) in the Department of Energy's battery program. A thermal battery is a primary, reserve electrochemical power source; that is, it can be used only once and then for a relatively short period, measured in minutes. To energize the battery, an external electrical signal ignites a heat source in the battery to melt the electrolyte and initiate an electrochemical reaction. The battery is made up of several series-connected cells, each with an anode, a cathode, and a current collector. A cell's anode is calcium; its cathode is hexavalent chromium. The electrochemical reaction takes place when the electrolyte is melted by heat supplied from ignition of an iron-potassium perchlorate disk. Since no reaction occurs while the electrolyte is in the solid state, the battery does not deteriorate with time and has a shelf life exceeding 20 years. Presented are such critical battery operating characteristics as temperature, rise time, active life, current capacity, etc. Design factors described include size and shape, pellet density, ignition methods, anode construction, etc. These batteries are designed by Sandia National Laboratories, Albuquerque. GEND acts as a procurement agency and provides engineering support to suppliers. 18 figures.

  12. Stress corrosion cracking of Alloys 600 and 182 in BWR environments

    SciTech Connect (OSTI)

    Ljungberg, L.G.; Hofling, C.G.; Sahlberg, A.; Moeller, J. )

    1992-05-01

    Wrought Alloy 600 and weldments of Alloy 182 are being tested for initiation and propagation of intergranular stress corrosion cracking (IGSCC). Crack initiation is tested on compact tension (CT) specimens with U-notches of various radii under enhanced crevice conditions, in a test loop in a Swedish BWR. After one year exposure there was initiation of IGSCC in a large portion of the Alloy 182 specimens, but nearly no initiation in Alloy 600. Crack propagation was measured in a laboratory loop on CT specimens under constant or cyclic load. Low carbon Alloy 600, or Alloy 182 high in titanium and niobium versus carbon, cracked at lower rates than material with high carbon activity. Materials with low concentrations of phosphorus and sulfur cracked slower than those high in these elements in clean environment, but no such effect was found in environment with sulfate. Alloy 182 weld metal generally cracked at higher rates than Alloy 600.

  13. Stress corrosion cracking of Alloys 600 and 182 in BWR environments. Interim report

    SciTech Connect (OSTI)

    Ljungberg, L.G.; Hofling, C.G.; Sahlberg, A.; Moeller, J.

    1992-05-01

    Wrought Alloy 600 and weldments of Alloy 182 are being tested for initiation and propagation of intergranular stress corrosion cracking (IGSCC). Crack initiation is tested on compact tension (CT) specimens with U-notches of various radii under enhanced crevice conditions, in a test loop in a Swedish BWR. After one year exposure there was initiation of IGSCC in a large portion of the Alloy 182 specimens, but nearly no initiation in Alloy 600. Crack propagation was measured in a laboratory loop on CT specimens under constant or cyclic load. Low carbon Alloy 600, or Alloy 182 high in titanium and niobium versus carbon, cracked at lower rates than material with high carbon activity. Materials with low concentrations of phosphorus and sulfur cracked slower than those high in these elements in clean environment, but no such effect was found in environment with sulfate. Alloy 182 weld metal generally cracked at higher rates than Alloy 600.

  14. A counter-charge layer in generalized solvents framework for electrical double layers in neat and hybrid ionic liquid electrolytes

    SciTech Connect (OSTI)

    Huang, Jingsong; Feng, Guang; Sumpter, Bobby G; Qiao, Rui; Meunier, Vincent

    2011-01-01

    Room-temperature ionic liquids (RTILs) have received significant attention as electrolytes due to a number of attractive properties such as their wide electrochemical windows. Since electrical double layers (EDLs) are the cornerstone for the applications of RTILs in electrochemical systems such as supercapacitors, it is important to develop an understanding of the structure capacitance relationships for these systems. Here we present a theoretical framework termed counter-charge layer in generalized solvents (CGS) for describing the structure and capacitance of the EDLs in neat RTILs and in RTILs mixed with different mass fractions of organic solvents. Within this framework, an EDL is made up of a counter-charge layer exactly balancing the electrode charge, and of polarized generalized solvents (in the form of layers of ion pairs, each of which has a zero net charge but has a dipole moment the ion pairs thus can be considered as a generalized solvent) consisting of all RTILs inside the system except the counter-ions in the counter-charge layer, together with solvent molecules if present. Several key features of the EDLs that originate from the strong ion ion correlation in RTILs, e.g., overscreening of electrode charge and alternating layering of counter-ions and co-ions, are explicitly incorporated into this framework. We show that the dielectric screening in EDLs is governed predominately by the polarization of generalized solvents (or ion pairs) in the EDL, and the capacitance of an EDL can be related to its microstructure with few a priori assumptions or simplifications. We use this framework to understand two interesting phenomena observed in molecular dynamics simulations of EDLs in a neat IL of 1-butyl-3- methylimidazolium tetrafluoroborate ([BMIM][BF4]) and in a mixture of [BMIM][BF4] and acetonitrile (ACN): (1) the capacitance of the EDLs in the [BMIM][BF4]/ACN mixture increases only slightly when the mass fraction of ACN in the mixture increases from zero

  15. Possible Methods to Estimate Core Location in a Beyond-Design-Basis Accident at a GE BWR with a Mark I Containment Stucture

    SciTech Connect (OSTI)

    Walston, S; Rowland, M; Campbell, K

    2011-07-27

    It is difficult to track to the location of a melted core in a GE BWR with Mark I containment during a beyond-design-basis accident. The Cooper Nuclear Station provided a baseline of normal material distributions and shielding configurations for the GE BWR with Mark I containment. Starting with source terms for a design-basis accident, methods and remote observation points were investigated to allow tracking of a melted core during a beyond-design-basis accident. The design of the GE BWR with Mark-I containment highlights an amazing poverty of expectations regarding a common mode failure of all reactor core cooling systems resulting in a beyond-design-basis accident from the simple loss of electric power. This design is shown in Figure 1. The station blackout accident scenario has been consistently identified as the leading contributor to calculated probabilities for core damage. While NRC-approved models and calculations provide guidance for indirect methods to assess core damage during a beyond-design-basis loss-of-coolant accident (LOCA), there appears to be no established method to track the location of the core directly should the LOCA include a degree of fuel melt. We came to the conclusion that - starting with detailed calculations which estimate the release and movement of gaseous and soluble fission products from the fuel - selected dose readings in specific rooms of the reactor building should allow the location of the core to be verified.

  16. BWR Spent Nuclear Fuel Interfacial Bonding Efficiency Study

    SciTech Connect (OSTI)

    Wang, Jy-An John; Jiang, Hao

    2015-04-30

    The objective of this project is to perform a systematic study of spent nuclear fuel (SNF, also known as “used nuclear fuel” [UNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. Additional CIRFT testing was conducted on three HBR rods; two specimens failed, and one specimen was tested to over 2.23 × 107 cycles without failing. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of the SNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, eleven SNF rod segments from the Limerick BWR were tested using the ORNL CIRFT equipment; one test under static conditions and ten tests under dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at a maximum curvature of 4.0 m-1. The specimen did not show any sign of failure during three repeated loading cycles to a similar maximum curvature. Ten cyclic tests were conducted with amplitudes varying from 15.2 to 7.1 N·m. Failure was observed in nine of

  17. Electricity Monthly Update

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

    Methodology and Documentation General The Electricity Monthly Update is prepared by the Electric Power Operations Team, Office of Electricity, Renewables and Uranium Statistics,...

  18. EA-1723: General Motors LLC Electric Drive Vehicle Battery and Component Manufacturing Initiative Application White Marsh, Maryland and Wixom, Michigan

    Broader source: Energy.gov [DOE]

    DOE’s Proposed Action is to provide GM with $105,387,000 in financial assistance in a cost sharing arrangement to facilitate construction and operation of a manufacturing facility to produce electric motor components and assemble an electric drive unit. This Proposed Action through the Vehicle Technologies Program will accelerate the development and production of electric-drive vehicle systems and reduce the United States’ consumption of petroleum. This Proposed Action will also meaningfully assist in the nation’s economic recovery by creating manufacturing jobs in the United States in accordance with the objectives of the Recovery Act.

  19. 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

  20. The Modeling of Advanced BWR Fuel Designs with the NRC Fuel Depletion Codes PARCS/PATHS

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

    Ward, Andrew; Downar, Thomas J.; Xu, Y.; March-Leuba, Jose A; Thurston, Carl; Hudson, Nathanael H.; Ireland, A.; Wysocki, A.

    2015-04-22

    The PATHS (PARCS Advanced Thermal Hydraulic Solver) code was developed at the University of Michigan in support of U.S. Nuclear Regulatory Commission research to solve the steady-state, two-phase, thermal-hydraulic equations for a boiling water reactor (BWR) and to provide thermal-hydraulic feedback for BWR depletion calculations with the neutronics code PARCS (Purdue Advanced Reactor Core Simulator). The simplified solution methodology, including a three-equation drift flux formulation and an optimized iteration scheme, yields very fast run times in comparison to conventional thermal-hydraulic systems codes used in the industry, while still retaining sufficient accuracy for applications such as BWR depletion calculations. Lastly, themore » capability to model advanced BWR fuel designs with part-length fuel rods and heterogeneous axial channel flow geometry has been implemented in PATHS, and the code has been validated against previously benchmarked advanced core simulators as well as BWR plant and experimental data. We describe the modifications to the codes and the results of the validation in this paper.« less

  1. The Modeling of Advanced BWR Fuel Designs with the NRC Fuel Depletion Codes PARCS/PATHS

    SciTech Connect (OSTI)

    Ward, Andrew; Downar, Thomas J.; Xu, Y.; March-Leuba, Jose A; Thurston, Carl; Hudson, Nathanael H.; Ireland, A.; Wysocki, A.

    2015-04-22

    The PATHS (PARCS Advanced Thermal Hydraulic Solver) code was developed at the University of Michigan in support of U.S. Nuclear Regulatory Commission research to solve the steady-state, two-phase, thermal-hydraulic equations for a boiling water reactor (BWR) and to provide thermal-hydraulic feedback for BWR depletion calculations with the neutronics code PARCS (Purdue Advanced Reactor Core Simulator). The simplified solution methodology, including a three-equation drift flux formulation and an optimized iteration scheme, yields very fast run times in comparison to conventional thermal-hydraulic systems codes used in the industry, while still retaining sufficient accuracy for applications such as BWR depletion calculations. Lastly, the capability to model advanced BWR fuel designs with part-length fuel rods and heterogeneous axial channel flow geometry has been implemented in PATHS, and the code has been validated against previously benchmarked advanced core simulators as well as BWR plant and experimental data. We describe the modifications to the codes and the results of the validation in this paper.

  2. Calculation of a BWR partial ATWS using RAMONA-3B

    SciTech Connect (OSTI)

    Garber, D.I.; Diamond, D.J.; Cheng, H.S.

    1982-01-01

    The RAMONA-3B code has been used to simulate a boiling water reactor (BWR) transient initiated by the closure of the main steam line isolation valves in which all the control rods in one-half the core fail to scram after reactor trip. The modeling of the nuclear steam supply system included three-dimensional neutron kinetics and parallel hydraulic channels (including a bypass channel). The transient is characterized by an initial pressure spike and then by oscillations in the pressure due to the opening and closing of relief valves. These oscillations in turn affect all thermohydraulic properties in the vessel. The simulation was continued for 7 minutes of reactor time at which point boron began to accumulate in the core. The calculation demonstrates the importance of using three-dimensional neutron kinetics in conjunction with the modeling of the nuclear steam supply system for this type of transient. RAMONA-3B is unique in its ability to do this type of calculation.

  3. BWR Anticipated Transients Without Scram Leading to Instability

    SciTech Connect (OSTI)

    Cheng L. Y.; Baek J.; Cuadra, A.; Aronson, A.; Diamond, D.; Yarsky, P.

    2013-11-10

    Anticipated transients without scram (ATWS) in aboiling water reactor (BWR) were simulated in order to understand reactor response and determine the effectiveness of automatic and operator actions to mitigate this beyond-design-basis accident. The events of interest herein are initiated by a turbine trip when the reactor is operating in the expanded operating domainMELLLA+ [maximum extended load line limit plus]. In these events the reactor may initially be at up to 120% of the original licensed thermal power (OLTP) and at flow rates as low as 80% of rated.For these (and similar) ATWS events the concern isthat when the reactor power decreases in response to a dual recirculation pump trip, the core will become unstable and large amplitude oscillations will begin. The occurrence of these power oscillations, if left unmitigated, may result in fuel damage, and the amplitude of the poweroscillations may hamper the effectiveness of the injection of dissolved neutron absorber through the standby liquid control system (SLCS).

  4. GOTHIC MODEL OF BWR SECONDARY CONTAINMENT DRAWDOWN ANALYSES

    SciTech Connect (OSTI)

    Hansen, P.N.

    2004-10-06

    This article introduces a GOTHIC version 7.1 model of the Secondary Containment Reactor Building Post LOCA drawdown analysis for a BWR. GOTHIC is an EPRI sponsored thermal hydraulic code. This analysis is required by the Utility to demonstrate an ability to restore and maintain the Secondary Containment Reactor Building negative pressure condition. The technical and regulatory issues associated with this modeling are presented. The analysis includes the affect of wind, elevation and thermal impacts on pressure conditions. The model includes a multiple volume representation which includes the spent fuel pool. In addition, heat sources and sinks are modeled as one dimensional heat conductors. The leakage into the building is modeled to include both laminar as well as turbulent behavior as established by actual plant test data. The GOTHIC code provides components to model heat exchangers used to provide fuel pool cooling as well as area cooling via air coolers. The results of the evaluation are used to demonstrate the time that the Reactor Building is at a pressure that exceeds external conditions. This time period is established with the GOTHIC model based on the worst case pressure conditions on the building. For this time period the Utility must assume the primary containment leakage goes directly to the environment. Once the building pressure is restored below outside conditions the release to the environment can be credited as a filtered release.

  5. A Program for Risk Assessment Associated with IGSCC of BWR Vessel Internals

    SciTech Connect (OSTI)

    A. G. Ware; D. K. Morton; J. D. Page; M. E. Nitzel; S. A. Eide; T. -Y. Chang

    1999-08-01

    A program is being carried out for the US Nuclear Regulatory Commission (NRC) by the Idaho National Engineering and Environmental Laboratory (INEEL), to conduct an independent risk assessment of the consequences of failures initiated by intergranular stress corrosion cracking (IGSCC) of the reactor vessel internals of boiling water reactor (BWR) plants. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, both singly and in combination with the failures of others, with specific consideration given to potential cascading and common mode effects on system performance. This paper presents a description of the overall program that is underway to modify an existing probabilistic risk assessment (PRA) of the BWR/4 plant to include IGSCC-initiated failures, subsequently to complete a quantitative PRA.

  6. Code System for Best-Estimate Analysis of LOCA in BWR.

    Energy Science and Technology Software Center (OSTI)

    2001-07-23

    Version 00 TRAC-BD1 performs best estimate analyses of loss-of-coolant accidents (LOCA) and other transients in boiling water reactors (BWRs). The program provides LOCA analysis capability for BWRs and for many BWR-related thermal-hydraulic experimental facilities. The program features a three-dimensional treatment of the BWR pressure vessel, a detailed model of a BWR fuel bundle including multi-rod, multi-bundle, radiation heat transfer, and leakage path modeling capability; flow-regime-dependent constitutive equation treatment; reflood tracking capability both for falling filmsmore » and bottom flood quench fronts; and consistent treatment of the entire accident sequence. Dump/restart capabilities are also provided.« less

  7. Key Parameters for Operator Diagnosis of BWR Plant Condition during a Severe Accident

    SciTech Connect (OSTI)

    Clayton, Dwight A.; Poore, III, Willis P.

    2015-01-01

    The objective of this research is to examine the key information needed from nuclear power plant instrumentation to guide severe accident management and mitigation for boiling water reactor (BWR) designs (specifically, a BWR/4-Mark I), estimate environmental conditions that the instrumentation will experience during a severe accident, and identify potential gaps in existing instrumentation that may require further research and development. This report notes the key parameters that instrumentation needs to measure to help operators respond to severe accidents. A follow-up report will assess severe accident environmental conditions as estimated by severe accident simulation model analysis for a specific US BWR/4-Mark I plant for those instrumentation systems considered most important for accident management purposes.

  8. Electricity Monthly Update

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

    of electricity. End-use data is the first "data page" based on the assumption that information about retail electricity service is of greatest interest to a general...

  9. The response of BWR Mark II containments to station blackout severe accident sequences

    SciTech Connect (OSTI)

    Greene, S.R.; Hodge, S.A.; Hyman, C.R.; Tobias, M.L. (Oak Ridge National Lab., TN (USA))

    1991-05-01

    This report describes the results of a series of calculations conducted to investigate the response of BWR Mark 2 containments to short-term and long-term station blackout severe accident sequences. The BWR-LTAS, BWRSAR, and MELCOR codes were employed to conduct quantitative accident sequence progression and containment response analyses for several station blackout scenarios. The accident mitigation effectiveness of automatic depressurization system actuation, drywell flooding via containment spray operation, and debris quenching in Mark 2 suppression pools is assessed. 27 refs., 16 figs., 21 tabs.

  10. Photoelectrochemical protection of stainless alloys from the stress-corrosion cracking in BWR primary coolant environment

    SciTech Connect (OSTI)

    Akashi, Masatsune; Iso-o, Hiroyuki; Kubota, Nobuhiko; Fukuda, Takanori; Ayabe, Muneo; Hirano, Kenji

    1995-12-31

    The feasibility of counteracting or preventing the stress-corrosion cracking in the BWR core internals by the photoelectrochemical method has been examined. For the purpose TiO{sub 2} semiconductor is noted for its capability of photo electrochemically inducing the water-oxidizing anodic reaction in low enough potential domain if supplied with a light of a wavelength shorter than 410 nm. This paper offers an empirical proof by showing that Type 304 stainless steel and Alloy 600 stainless alloy that have been plasma-spray coated with TiO{sub 2} film will do quite well in environments of BWR primary coolant.

  11. Analysis of BWR high burnup fuel in LOCA conditions

    SciTech Connect (OSTI)

    Garcia Sedano, Pablo; Dey Navarro, Jose Manuel; Gallego Cabezon, Ines; Orive Moreno, Raul

    2004-07-01

    High Burnup Fuel Behaviour has been growing in importance since middle 80's when pellet microstructure changes (rim effect) and cladding oxidation rates increase were observed. Later on, Cadarache reactivity tests revealed cladding integrity failures below safety limits. These phenomena, occurred at high burnup, stressed the necessity of having a wide experimental data base that would allow to dispose non-extrapolated data of material properties submitted to higher burnups than 40000 MWd/TM and data of new materials at the same time. One of the objectives of the EPRI's Fuel Reliability Program is to establish the bases for the licensing of nuclear fuel to burnup levels beyond the current licensed value of 62 GWd/MTU rod average burnup. The technical bases to support those high burnup levels are being developed. One of the licensing points of concern is the behaviour of the high burnup fuel in LOCA conditions. To respond to this concern a series of LOCA experiments are being performed at Argonne National Laboratory using fuel rods from Limerick NPP at 57 GWd/TM and H.B. Robinson at 67 GWd/MTU. When the ANL tests have been finished, a conservative Peak Cladding Temperature/ Equivalent Cladding Reacted (PCT/ECR) limit will be determine from the residual ductility tests to be applied to the high burnup fuel. This makes necessary to determine the behaviour of the high burnup fuel in LOCA conditions and to determine the available safety margin. In licensing LOCA calculations, corresponding to present core designs and future core designs, the calculated PCT and ECR values as a function of the fuel burnup could be used to determine the relative severity of LOCA for the high burnup fuel. This report presents the LOCA analyses performed by IBERDROLA (Spanish utility), using results from the Cofrentes NPP (BWR-6) LOCA evaluations. (authors)

  12. Minor Actinides Loading Optimization for Proliferation Resistant Fuel Design - BWR

    SciTech Connect (OSTI)

    G. S. Chang; Hongbin Zhang

    2009-09-01

    One approach to address the United States Nuclear Power (NP) 2010 program for the advanced light water reactor (LWR) (Gen-III+) intermediate-term spent fuel disposal need is to reduce spent fuel storage volume while enhancing proliferation resistance. One proposed solution includes increasing burnup of the discharged spent fuel and mixing minor actinide (MA) transuranic nuclides (237Np and 241Am) in the high burnup fuel. Thus, we can reduce the spent fuel volume while increasing the proliferation resistance by increasing the isotopic ratio of 238Pu/Pu. For future advanced nuclear systems, MAs are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. A typical boiling water reactor (BWR) fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of adding MAs (237Np and/or 241Am) to enhance proliferation resistance and improve fuel cycle performance for the intermediate-term goal of future nuclear energy systems. However, adding MAs will increase plutonium production in the discharged spent fuel. In this work, the Monte-Carlo coupling with ORIGEN-2.2 (MCWO) method was used to optimize the MA loading in the UO2 fuel such that the discharged spent fuel demonstrates enhanced proliferation resistance, while minimizing plutonium production. The axial averaged MA transmutation characteristics at different burnup were compared and their impact on neutronics criticality and the ratio of 238Pu/Pu discussed.

  13. Risk perception & strategic decision making :general insights, a framework, and specific application to electricity generation using nuclear energy.

    SciTech Connect (OSTI)

    Brewer, Jeffrey D.

    2005-11-01

    The objective of this report is to promote increased understanding of decision making processes and hopefully to enable improved decision making regarding high-consequence, highly sophisticated technological systems. This report brings together insights regarding risk perception and decision making across domains ranging from nuclear power technology safety, cognitive psychology, economics, science education, public policy, and neural science (to name a few). It forms them into a unique, coherent, concise framework, and list of strategies to aid in decision making. It is suggested that all decision makers, whether ordinary citizens, academics, or political leaders, ought to cultivate their abilities to separate the wheat from the chaff in these types of decision making instances. The wheat includes proper data sources and helpful human decision making heuristics; these should be sought. The chaff includes ''unhelpful biases'' that hinder proper interpretation of available data and lead people unwittingly toward inappropriate decision making ''strategies''; obviously, these should be avoided. It is further proposed that successfully accomplishing the wheat vs. chaff separation is very difficult, yet tenable. This report hopes to expose and facilitate navigation away from decision-making traps which often ensnare the unwary. Furthermore, it is emphasized that one's personal decision making biases can be examined, and tools can be provided allowing better means to generate, evaluate, and select among decision options. Many examples in this report are tailored to the energy domain (esp. nuclear power for electricity generation). The decision making framework and approach presented here are applicable to any high-consequence, highly sophisticated technological system.

  14. THERMAL EVALUATION OF THE USE OF BWR MOX SNF IN THE WASTE PACKAGE DESIGN (SCPB: N/A)

    SciTech Connect (OSTI)

    H. Wang

    1997-01-23

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) as specified in the Waste Package Implementation Plan (pp. 4-8,4-11,4-24, 5-1, and 5-13; Ref. 5.10) and Waste Package Plan (pp. 3-15,3-17, and 3-24; Ref. 5.9). The design data request addressed herein is: (1) Characterize the conceptual 40 BWR and 24 BWR Multi-Purpose Canister (MPC) Waste Package (WP) design to show that the design is feasible for use in the MGDS environment when loaded with BWR MOX SNF. (2) Characterize the conceptual 44 BWR and 24 BWR Uncanistered Fuel (UCF) Waste Package (WP) design to show that the design is feasible for use in the MGDS environment when loaded with BWR MOX SNF. The purpose of this analysis is to respond to a concern that the long-term disposal thermal issues for the WP Design, if used with SNF designed for a MOX fuel cycle, do not preclude WP compatibility with the MGDS. The objective of this analysis is to provide thermal parameter information for the conceptual WP design with disposal container which is loaded with BWR MOX SNF under nominal MGDS repository conditions. The results are intended to show that the design has a reasonable chance to meet the MGDS design requirements for normal MGDS operation, and to provide the required guidance to determining the major design issues for future design efforts, and to show that the BWR MOX SNF loaded WP performance is similar to an WP loaded with commercial BWR SNF.

  15. Conditioning of BWR Control - Elements Using the New MOSAIK 80T/SWR-SE Cask - Concept

    SciTech Connect (OSTI)

    Oldiges, O.; Blenski, H.-J.; Engelage, H.; Behrens, W.; Majunke, J.; Schwarz, W.; Hallfarth, Dr.

    2002-02-27

    During the operation of Boiling Water Reactors, Control - Elements are used to control the neutron flux inside the reactor vessel. After the end of the lifetime, the Control - Elements are usually stored in the fuel - elements - pool of the reactor. Up to now, in Germany no conditioning of Control - Elements has been done in a BWR under operation.

  16. Electrochemical Potential (ECP) of Clean Heated Fuel Cladding Material and Structural SS under BWR Operating Conditions

    SciTech Connect (OSTI)

    Pop, Mike G.; Bell, Merl; Kilian, Renate; Dorsch, Thomas; Christian, Mueller

    2007-07-01

    To preliminarily monitor the relative effect of advanced water chemistry measures on SS structural material and fuel cladding in BWR environments a number of experiments were performed using laboratory equipment (recirculation loop, autoclave with heated electrodes, reference electrodes, etc.). The simulation of the plant condition was done without impurities or crud deposit contribution (clean surfaces). Subsequent testing, performed during 2007 and not yet cleared for release, is considering the effect of combined complex BWR chemistries and crud deposition. The heated Zircaloy fuel cladding tubing was prepared to simulate heat transfer by internal heating at levels existing in BWR (70 W/cm{sup 2}). For comparison purposes additional type SS347 electrode and unheated zirconium was used. A platinum electrode was used to measure the redox potential of the electrolyte. A high temperature Ag/AgCl electrode was used as a reference electrode. The assembly was installed in a recirculation 1 liter autoclave. Present report presents corrosion potential measurements performed under the following BWR water chemistry conditions (at 288 deg. C fluid exit temperature, 86 bar with surface temperature of Zirconium hot finger at 296 deg. C) - normal (inert) water conditions, - hydrogen injection in three steps from 0.68 ppm to 1.6 ppm, - oxygen injection in three steps from 2.4 ppm to 10 ppm - -methanol 2 ppm and oxygen 2 ppm in a close loop (without methanol refreshing) (authors)

  17. Straight and chopped dc performance data for a General Electric 5BT 2366C10 motor and an EV-1 controller. Final report

    SciTech Connect (OSTI)

    Edie, P.C.

    1981-01-01

    This report is intended to supply the electric vehicle manufacturer with performance data on the General Electric 5BT 2366C10 series wound dc motor and EV-1 chopper controller. Data are provided for both straight and chopped dc input to the motor, at 2 motor temperature levels. Testing was done at 6 voltage increments to the motor, and 2 voltage increments to the controller. Data results are presented in both tabular and graphical forms. Tabular information includes motor voltage and current input data, motor speed and torque output data, power data and temperature data. Graphical information includes torque-speed, motor power output-speed, torque-current, and efficiency-speed plots under the various operating conditions. The data resulting from this testing shows the speed-torque plots to have the most variance with operating temperature. The maximum motor efficiency is between 86% and 87%, regardless of temperature or mode of operation. When the chopper is utilized, maximum motor efficiency occurs when the chopper duty cycle approaches 100%. At low duty cycles the motor efficiency may be considerably less than the efficiency for straight dc. Chopper efficiency may be assummed to be 95% under all operating conditions. For equal speeds at a given voltage level, the motor operated in the chopped mode develops slightly more torque than it does in the straight dc mode. System block diagrams are included, along with test setup and procedure information.

  18. Electric vehicles

    SciTech Connect (OSTI)

    Not Available

    1990-03-01

    Quiet, clean, and efficient, electric vehicles (EVs) may someday become a practical mode of transportation for the general public. Electric vehicles can provide many advantages for the nation's environment and energy supply because they run on electricity, which can be produced from many sources of energy such as coal, natural gas, uranium, and hydropower. These vehicles offer fuel versatility to the transportation sector, which depends almost solely on oil for its energy needs. Electric vehicles are any mode of transportation operated by a motor that receives electricity from a battery or fuel cell. EVs come in all shapes and sizes and may be used for different tasks. Some EVs are small and simple, such as golf carts and electric wheel chairs. Others are larger and more complex, such as automobile and vans. Some EVs, such as fork lifts, are used in industries. In this fact sheet, we will discuss mostly automobiles and vans. There are also variations on electric vehicles, such as hybrid vehicles and solar-powered vehicles. Hybrid vehicles use electricity as their primary source of energy, however, they also use a backup source of energy, such as gasoline, methanol or ethanol. Solar-powered vehicles are electric vehicles that use photovoltaic cells (cells that convert solar energy to electricity) rather than utility-supplied electricity to recharge the batteries. This paper discusses these concepts.

  19. Environmental consequences of postulated plutonium releases from General Electric Company Vallecitos Nuclear Center, Vallecitos, California, as a result of severe natural phenomena

    SciTech Connect (OSTI)

    Jamison, J.D.; Watson, E.C.

    1980-11-01

    Potential environmental consequences in terms of radiation dose to people are presented for postulated plutonium releases caused by severe natural phenomena at the General Electric Company Vallecitos Nuclear Center, Vallecitos, California. The severe natural phenomena considered are earthquakes, tornadoes, and high straight-line winds. Maximum plutonium deposition values are given for significant locations around the site. All important potential exposure pathways are examined. The most likely 50-year committed dose equivalents are given for the maximum-exposed individual and the population within a 50-mile radius of the plant. The maximum plutonium deposition values likely to occur offsite are also given. The most likely calculated 50-year collective committed dose equivalents are all much lower than the collective dose equivalent expected from 50 years of exposure to natural background radiation and medical x-rays. The most likely maximum residual plutonium contamination estimated to be deposited offsite following the earthquakes, and the 180-mph and 230-mph tornadoes are above the Environmental Protection Agency's (EPA) proposed guideline for plutonium in the general environment of 0.2 ..mu..Ci/m/sup 2/. The deposition values following the 135-mph tornado are below the EPA proposed guidelines.

  20. Enhancing BWR Proliferation Resistance Fuel with Minor Actinides

    SciTech Connect (OSTI)

    Gray S. Chang

    2008-07-01

    reactivity control of the systems into which they are incorporated. In the study, a typical boiling water reactor (BWR) fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate term goal for future nuclear energy systems. To account for the water coolant density variation from the bottom (0.76 g/cm3) to the top (0.35 g/cm3) of the core, the axial coolant channel and fuel pin were divided to 24 nodes. The MA transmutation characteristics at different elevations were compared and their impact on neutronics criticality discussed. We concluded that the concept of MARA, which involves the use of transuranic nuclides (237Np and/or 241Am), significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate term of nuclear energy rennaissance.

  1. Coolant Density and Control Blade History Effects in Extended BWR Burnup Credit

    SciTech Connect (OSTI)

    Ade, Brian J; Marshall, William BJ J; Bowman, Stephen M; Gauld, Ian C; Ilas, Germina; Martinez-Gonzalez, Jesus S

    2015-01-01

    Oak Ridge National Laboratory and the US Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (keff) calculations and predicted spent fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date investigating some aspects of extended BUC. (The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper.) Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC: (1) the effect of axial void profile and (2) the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of a modern operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. Although a single cycle does not provide complete data, the data obtained are sufficient to determine the primary effects and to identify conservative modeling approaches. These data were used in a study of the effect of axial void profile. The first stage of the study was determination of the necessary moderator density temporal fidelity in depletion modeling. After the required temporal fidelity was established, multiple void profiles were used to examine the effect on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied was control blade history. Control blades are inserted in

  2. Preliminary design study of small long life boiling water reactor (BWR) with tight lattice thorium nitride fuel

    SciTech Connect (OSTI)

    Trianti, Nuri E-mail: szaki@fi.itba.c.id; Su'ud, Zaki E-mail: szaki@fi.itba.c.id; Arif, Idam E-mail: szaki@fi.itba.c.id; Riyana, EkaSapta

    2014-09-30

    Neutronic performance of small long-life boiling water reactors (BWR) with thorium nitride based fuel has been performed. A recent study conducted on BWR in tight lattice environments (with a lower moderator percentage) produces small power reactor which has some specifications, i.e. 10 years operation time, power density of 19.1 watt/cc and maximum excess reactivity of about 4%. This excess reactivity value is smaller than standard reactivity of conventional BWR. The use of hexagonal geometry on the fuel cell of BWR provides a substantial effect on the criticality of the reactor to obtain a longer operating time. Supported by a tight concept lattice where the volume fraction of the fuel is greater than the moderator and fuel, Thorium Nitride give good results for fuel cell design on small long life BWR. The excess reactivity of the reactor can be reduced with the addition of gadolinium as burnable poisons. Therefore the hexagonal tight lattice fuel cell design of small long life BWR that has a criticality more than 20 years of operating time has been obtained.

  3. Using Electricity",,,"Electricity Consumption",,,"Electricity...

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

    . Total Electricity Consumption and Expenditures, 2003" ,"All Buildings* Using Electricity",,,"Electricity Consumption",,,"Electricity Expenditures" ,"Number of Buildings...

  4. Roles of electricity: Electric steelmaking

    SciTech Connect (OSTI)

    Burwell, C.C.

    1986-07-01

    Electric steel production from scrap metal continues to grow both in total quantity and in market share. The economics of electric-steel production in general, and of electric minimills in particular, seem clearly established. The trend towards electric steelmaking provides significant economic and competitive advantages for producers and important overall economic, environmental, and energy advantages for the United States at large. Conversion to electric steelmaking offers up to a 4-to-1 advantage in terms of the overall energy used to produce a ton of steel, and s similar savings in energy cost for the producer. The amount of old scrap used to produce a ton of steel has doubled since 1967 because of the use of electric furnaces.

  5. Thermometry in the multirod burst test program. [PWR; BWR

    SciTech Connect (OSTI)

    Anderson, R.L.; Carr, K.R.; Kollie, T.G.

    1982-03-01

    A temperature measurement error analysis was performed for the Type S (0.25-mm-diam, bare-wire) and Type K (0.71-mm-diam, sheathed) thermocouple circuits used to measure the temperature of the Zircaloy-clad, electrically heated fuel-rod simulators in the Multirod Burst Test program (MRBT) at Oak Ridge National Laboratory (ORNL). An important objective of the MRBT program is to improve the understanding of the behavior of the Zircaloy cladding of nuclear fuel rods under conditions postulated for a large-break, loss-of-coolant accident. Eight categories of error sources were studied both analytically and experimentally: thermal shunting; electrical shuntng and leakage; thermocouple calibration; thermocouple decalibration in service; thermoelectric properties of thermocouple extension wire, plugs, and jacks; thermocouple reference junction; data acquisition system; and electrical noise.

  6. Assessment of TRAC-BD1 amd RAMONA-3B codes fpr BWR ATWS application

    SciTech Connect (OSTI)

    Neymotin, L.; Hsu, C.J.; Saha, P.

    1984-01-01

    Based on comparisons between the TRAC-BD1 power imposed calculation and the RAMONA-3B results, it can be said that the thermal-hydraulic models of both RAMONA-3B and TRAC-BD1 provide adequate representation of an ATWS event in a BWR. However, for the reactor power calculation, RAMONA-3B with space-time neutron kinetics is a superior and preferable tool to the TRAC-BD1 with point kinetics for ATWS type events where the spatial core power distribution varies with time. Also, the computer running time for RAMONA-3B (with 115 hydraulic cells and 192 neutronic cells has been found to be about four times lower than TRAC-BD1 (with 63 hydraulic cells and point kinetics). Therefore, it is recommended that RAMONA-3B be further used for best-estimate analysis of BWR ATWS-type events.

  7. Recent SCDAP/RELAP5 improvements for BWR severe accident simulations

    SciTech Connect (OSTI)

    Griffin, F.P.

    1995-12-31

    A new model for the SCDAP/RELAP5 severe accident analysis code that represents the control blade and channel box structures in a boiling water reactor (BWR) has been under development since 1991. This model accounts for oxidation, melting, and relocation of these structures, including the effects of material interactions between B{sub 4}C, stainless steel, and Zircaloy. This paper describes improvements that have been made to the BWR control blade/channel box model during 1994 and 1995. These improvements include new capabilities that represent the relocation of molten material in a more realistic manner and modifications that improve the usability of the code by reducing the frequency of code failures. This paper also describes a SCDAP/RELAP5 assessment calculation for the Browns Ferry Nuclear Plant design based upon a short-term station blackout accident sequence.

  8. Evaluation of ATWS core damage frequency for an earlier vintage BWR/4

    SciTech Connect (OSTI)

    Shiu, K.K.; Ilberg, D.; Hanan, N.

    1986-01-01

    In summary, this study evaluated the total core damage contribution due to ATWS events for an earlier vintage BWR/4. A realistic calculation was performed for a plant with a particular ATWS prevention and mitigation configuration and with some of the ATWS rule modifications implemented. Results are compared with the ATWS task force findings; three areas have been identified which could potentially have significant impact upon the ATWS core damage frequency contribution. 4 refs.

  9. High-Burnup BWR Fuel Behavior Under Simulated Reactivity-Initiated Accident Conditions

    SciTech Connect (OSTI)

    Nakamura, Takehiko; Kusagaya, Kazuyuki; Fuketa, Toyoshi; Uetsuka, Hiroshi

    2002-06-15

    Boiling water reactor (BWR) fuel at 56 to 61 GWd/tonne U was pulse irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity-initiated accident conditions. Current Japanese 8 x 8 type Step II BWR fuel from Fukushima Daini Unit 2 was refabricated to short segments, and thermal energy from 272 to 586 J/g (65 to 140 cal/g) was promptly inserted to the test rods. Cladding deformation of the BWR fuel by the pulse irradiation was smaller than that of pressurized water reactor (PWR) fuels. However, cladding failure occurred in tests with fuel at burnup of 61 GWd/tonne U at fuel enthalpies of 260 to 360 J/g (62 to 86 cal/g) during the early stages of transients, while the cladding remained cool. The failure was comparable to the one observed in high-burnup PWR fuel tests, in which embrittled cladding with dense hydride precipitation near the outer surface was fractured due to pellet cladding mechanical interaction. Transient fission gas release by the pulse irradiation was {approx}9.6 to 17% depending on the peak fuel enthalpy.

  10. Using Electricity",,,"Electricity Consumption",,,"Electricity...

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

    A. Total Electricity Consumption and Expenditures for All Buildings, 2003" ,"All Buildings Using Electricity",,,"Electricity Consumption",,,"Electricity Expenditures" ,"Number of...

  11. Electricity",,,"Electricity Consumption",,,"Electricity Expenditures...

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

    C9. Total Electricity Consumption and Expenditures, 1999" ,"All Buildings Using Electricity",,,"Electricity Consumption",,,"Electricity Expenditures" ,"Number of Buildings...

  12. Electricity",,,"Electricity Consumption",,,"Electricity Expenditures...

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

    DIV. Total Electricity Consumption and Expenditures by Census Division, 1999" ,"All Buildings Using Electricity",,,"Electricity Consumption",,,"Electricity Expenditures" ,"Number...

  13. Electricity Monthly Update

    Gasoline and Diesel Fuel Update (EIA)

    Methodology and Documentation General The Electricity Monthly Update is prepared by the Electric Power Operations Team, Office of Electricity, Renewables and Uranium Statistics, U.S. Energy Information Administration (EIA), U.S. Department of Energy. Data published in the Electricity Monthly Update are compiled from the following sources: U.S. Energy Information Administration, Form EIA-826,"Monthly Electric Utility Sales and Revenues with State Distributions Report," U.S. Energy

  14. 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

  15. TRACE Model for Simulation of Anticipated Transients Without Scram in a BWR

    SciTech Connect (OSTI)

    Cheng L. Y.; Baek J.; Cuadra,A.; Aronson, A.; Diamond, D.; Yarsky, P.

    2013-11-10

    A TRACE model has been developed for using theTRACE/PARCS computational package [1, 2] to simulate anticipated transients without scram (ATWS) events in a boiling water reactor (BWR). The model represents a BWR/5 housed in a Mark II containment. The reactor and the balance of plant systems are modeled in sufficient detail to enable the evaluation of plant responses and theeffectiveness of automatic and operator actions tomitigate this beyond design basis accident.The TRACE model implements features thatfacilitate the simulation of ATWS events initiated by turbine trip and closure of the main steam isolation valves (MSIV). It also incorporates control logic to initiate actions to mitigate the ATWS events, such as water levelcontrol, emergency depressurization, and injection of boron via the standby liquid control system (SLCS). Two different approaches have been used to model boron mixing in the lower plenum of the reactor vessel: modulate coolant flow in the lower plenum by a flow valve, and use control logic to modular.

  16. Thermal Response of the 44-BWR Waste Package to a Hypothetical Fire Accident

    SciTech Connect (OSTI)

    J.R. Smotrel; H. Marr; M.J. Anderson

    2001-04-05

    The purpose of this calculation is to determine the thermal response of the 44-boiling water reactor (BWR) waste package (WP) to the hypothetical regulatory fire accident. The objective is to calculate the temperature response of the waste package materials to the hypothetical short-term fire defined in 10 CFR 7 1, Section 73(c)(4), Reference 1. The scope of the calculation includes evaluation of the accident with the waste package above ground, at the Yucca Mountain surface facility. 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 is that for the potential design of the type of WP considered in this calculation. In addition to the nominal design configuration thermal load case, the effects of varying the BWR thermal load are determined. The associated activity is the development of engineering evaluations to support the Licensing Application (LA) design activities.

  17. Design study of Thorium-232 and Protactinium-231 based fuel for long life BWR

    SciTech Connect (OSTI)

    Trianti, N.; Su'ud, Z.; Riyana, E. S.

    2012-06-06

    A preliminary design study for the utilization of thorium added with {sup 231}Pa based fuel on BWR type reactor has been performed. In the previous research utilization of fuel based Thorium-232 and Uranium-233 show 10 years operation time with maximum excess-reactivity about 4.075% dk/k. To increase reactor operation time and reduce excess-reactivity below 1% dk/k, Protactinium (Pa-231) is used as Burnable Poison. Protactinium-231 has very interesting neutronic properties, which enable the core to reduce initial excess-reactivity and simultaneously increase production of {sup 233}U to {sup 231}Pa in burn-up process. Optimizations of the content of {sup 231}Pa in the core enables the BWR core to sustain long period of operation time with reasonable burn-up reactivity swing. Based on the optimization of fuel element composition (Th and Pa) in various moderation ratio we can get reactor core with longer operation time, 20 {approx} 30 years operation without fuel shuffling or refuelling, with average power densities maximum of about 35 watt/cc, and maximum excess-reactivity 0.56% dk/k.

  18. Development of a new methodology for stability analysis in BWR NPP

    SciTech Connect (OSTI)

    Garcia-Fenoll, M.; Abarca, A.; Barrachina, T.; Miro, R.; Verdu, G.

    2012-07-01

    In this work, a new methodology to reproduce power oscillations in BWR NPP is presented. This methodology comprises the modal analysis techniques, the signal analysis techniques and the simulation with the coupled code RELAP5/PARCSv2.7. Macroscopic cross sections are obtained by using the SIMTAB methodology, which is fed up with CASMO-4/SIMULATE-3 data. The input files for the neutronic and thermohydraulic codes are obtained automatically and the thermalhydraulic-to-neutronic representation (mapping) used is based on the fundamental, first and second harmonics shapes of the reactor power, calculated with the VALKIN code (developed in UPV). This mapping was chosen in order not to condition the oscillation pattern. To introduce power oscillations in the simulation a new capability in the coupled code, for generate density perturbations (both for the whole core and for chosen axial levels) according with the power modes shapes, has been implemented. The purpose of the methodology is to reproduce the driving mechanism of the out of phase oscillations appeared in BWR type reactors. In this work, the methodology is applied to the Record 9 point, collected in the NEA benchmark of Ringhals 1 NPP. A set of different perturbations are induced in the first active axial level and the LPRM signals resulting are analyzed. (authors)

  19. 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.

  20. Workplace Charging Challenge Partner: General Motors | Department...

    Office of Environmental Management (EM)

    ... General Electric Google Nissan San Diego Gas & Electric Siemens Tesla Verizon Behind the workplace charging goal is the EV Everywhere Challenge, which is dedicated to accelerating ...

  1. Modeling of BWR core meltdown accidents - for application in the MELRPI. MOD2 computer code

    SciTech Connect (OSTI)

    Koh, B R; Kim, S H; Taleyarkhan, R P; Podowski, M Z; Lahey, Jr, R T

    1985-04-01

    This report summarizes improvements and modifications made in the MELRPI computer code. A major difference between this new, updated version of the code, called MELRPI.MOD2, and the one reported previously, concerns the inclusion of a model for the BWR emergency core cooling systems (ECCS). This model and its computer implementation, the ECCRPI subroutine, account for various emergency injection modes, for both intact and rubblized geometries. Other changes to MELRPI deal with an improved model for canister wall oxidation, rubble bed modeling, and numerical integration of system equations. A complete documentation of the entire MELRPI.MOD2 code is also given, including an input guide, list of subroutines, sample input/output and program listing.

  2. An efficient modeling method for thermal stratification simulation in a BWR suppression pool

    SciTech Connect (OSTI)

    Haihua Zhao; Ling Zou; Hongbin Zhang; Hua Li; Walter Villanueva; Pavel Kudinov

    2012-09-01

    The suppression pool in a BWR plant not only is the major heat sink within the containment system, but also provides major emergency cooling water for the reactor core. In several accident scenarios, such as LOCA and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; and the pool temperature distribution also affects the NPSHa (Available Net Positive Suction Head) and therefore the performance of the pump which draws cooling water back to the core. Current safety analysis codes use 0-D lumped parameter methods to calculate the energy and mass balance in the pool and therefore have large uncertainty in prediction of scenarios in which stratification and mixing are important. While 3-D CFD methods can be used to analyze realistic 3D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, therefore long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by 1-D transient partial differential equations and substructures such as free or wall jets are modeled with 1-D integral models. This allows very large reductions in computational effort compared to 3-D CFD modeling. The POOLEX experiments at Finland, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, are used for validation. GOTHIC lumped parameter models are used to obtain boundary conditions for BMIX++ code and CFD simulations. Comparison between the BMIX++, GOTHIC, and CFD calculations against the POOLEX experimental data is discussed in detail.

  3. Simulation of Thermal Stratification in BWR Suppression Pools with One Dimensional Modeling Method

    SciTech Connect (OSTI)

    Haihua Zhao; Ling Zou; Hongbin Zhang

    2014-01-01

    The suppression pool in a boiling water reactor (BWR) plant not only is the major heat sink within the containment system, but also provides the major emergency cooling water for the reactor core. In several accident scenarios, such as a loss-of-coolant accident and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; the pool temperature distribution also affects the NPSHa (available net positive suction head) and therefore the performance of the Emergency Core Cooling System and Reactor Core Isolation Cooling System pumps that draw cooling water back to the core. Current safety analysis codes use zero dimensional (0-D) lumped parameter models to calculate the energy and mass balance in the pool; therefore, they have large uncertainties in the prediction of scenarios in which stratification and mixing are important. While three-dimensional (3-D) computational fluid dynamics (CFD) methods can be used to analyze realistic 3-D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, resulting in a long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code (Berkeley mechanistic MIXing code in C++) has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by one-dimensional (1-D) transient partial differential equations and substructures (such as free or wall jets) are modeled with 1-D integral models. This allows very large reductions in computational effort compared to multi-dimensional CFD modeling. One heat-up experiment performed at the Finland POOLEX facility, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, is used for

  4. Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments

    SciTech Connect (OSTI)

    Chen, Y.; Chopra, O. K.; Gruber, Eugene E.; Shack, William J.

    2010-06-01

    The internal components of light water reactors are exposed to high-energy neutron irradiation and high-temperature reactor coolant. The exposure to neutron irradiation increases the susceptibility of austenitic stainless steels (SSs) to stress corrosion cracking (SCC) because of the elevated corrosion potential of the reactor coolant and the introduction of new embrittlement mechanisms through radiation damage. Various nonsensitized SSs and nickel alloys have been found to be prone to intergranular cracking after extended neutron exposure. Such cracks have been seen in a number of internal components in boiling water reactors (BWRs). The elevated susceptibility to SCC in irradiated materials, commonly referred to as irradiation-assisted stress corrosion cracking (IASCC), is a complex phenomenon that involves simultaneous actions of irradiation, stress, and corrosion. In recent years, as nuclear power plants have aged and irradiation dose increased, IASCC has become an increasingly important issue. Post-irradiation crack growth rate and fracture toughness tests have been performed to provide data and technical support for the NRC to address various issues related to aging degradation of reactor-core internal structures and components. This report summarizes the results of the last group of tests on compact tension specimens from the Halden-II irradiation. The IASCC susceptibility of austenitic SSs and heat-affected-zone (HAZ) materials sectioned from submerged arc and shielded metal arc welds was evaluated by conducting crack growth rate and fracture toughness tests in a simulated BWR environment. The fracture and cracking behavior of HAZ materials, thermally sensitized SSs and grain-boundary engineered SSs was investigated at several doses (?3 dpa). These latest results were combined with previous results from Halden-I and II irradiations to analyze the effects of neutron dose, water chemistry, alloy compositions, and welding and processing conditions on IASCC

  5. BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

    SciTech Connect (OSTI)

    McKinnon, M.A.; Doman, J.W.; Tanner, J.E.; Guenther, R.J.; Creer, J.M.; King, C.E.

    1986-02-01

    This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior.

  6. Analysis of the OECD/NRC BWR Turbine Trip Benchmark by the Coupled-Code System ATHLET-QUABOX/CUBBOX

    SciTech Connect (OSTI)

    Langenbuch, S.; Schmidt, K.-D.; Velkov, K.

    2004-10-15

    The OECD/NRC boiling water reactor (BWR) turbine trip benchmark has been calculated by the coupled thermal-hydraulic neutronics system code ATHLET-QUABOX/CUBBOX developed by Gesellschaft fuer Anlagen- und Reaktorsicherheit. The results obtained for all three exercises and for the additional four hypothetical cases are presented. The physical phenomena determining the BWR pressure transient are studied and explained. The sensitivity of results to variations of the initial steady-state conditions and of parameters of the two-phase flow model is discussed. A comparison is also performed for exercise 2 between the reactor core model with 33 thermal-hydraulic channels (THCs) as specified and a reactor core model with 764 THCs using a 1:1 mapping scheme.

  7. Electricity Fuel Basics | Department of Energy

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

    Vehicles & Fuels » Fuels » Electricity Fuel Basics Electricity Fuel Basics August 19, 2013 - 5:44pm Addthis Electricity used to power vehicles is generally provided by the electricity grid and stored in the vehicle's batteries. Vehicles that run on electricity have no tailpipe emissions. Emissions that can be attributed to electric vehicles are generated during electricity production at the power plant. Charging plug-in electric vehicles at home is as simple as plugging them into an

  8. Distributed charging of electrical assets

    DOE Patents [OSTI]

    Ghosh, Soumyadip; Phan, Dung; Sharma, Mayank; Wu, Chai Wah; Xiong, Jinjun

    2016-02-16

    The present disclosure relates generally to the field of distributed charging of electrical assets. In various examples, distributed charging of electrical assets may be implemented in the form of systems, methods and/or algorithms.

  9. Electric Vehicles

    Broader source: Energy.gov [DOE]

    This album contains a variety of all-electric, plug-in hybrid electric and fuel cell electric vehicles. For a full list of all electric vehicles visit the EV Everywhere website.

  10. Impact of modeling Choices on Inventory and In-Cask Criticality Calculations for Forsmark 3 BWR Spent Fuel

    SciTech Connect (OSTI)

    Martinez-Gonzalez, Jesus S.; Ade, Brian J.; Bowman, Stephen M.; Gauld, Ian C.; Ilas, Germina; Marshall, William BJ J.

    2015-01-01

    Simulation of boiling water reactor (BWR) fuel depletion poses a challenge for nuclide inventory validation and nuclear criticality safety analyses. This challenge is due to the complex operating conditions and assembly design heterogeneities that characterize these nuclear systems. Fuel depletion simulations and in-cask criticality calculations are affected by (1) completeness of design information, (2) variability of operating conditions needed for modeling purposes, and (3) possible modeling choices. These effects must be identified, quantified, and ranked according to their significance. This paper presents an investigation of BWR fuel depletion using a complete set of actual design specifications and detailed operational data available for five operating cycles of the Swedish BWR Forsmark 3 reactor. The data includes detailed axial profiles of power, burnup, and void fraction in a very fine temporal mesh for a GE14 (10×10) fuel assembly. The specifications of this case can be used to assess the impacts of different modeling choices on inventory prediction and in-cask criticality, specifically regarding the key parameters that drive inventory and reactivity throughout fuel burnup. This study focused on the effects of the fidelity with which power history and void fraction distributions are modeled. The corresponding sensitivity of the reactivity in storage configurations is assessed, and the impacts of modeling choices on decay heat and inventory are addressed.

  11. Stress-corrosion crack initiation process for Alloy 182 weld metal in simulated BWR environments

    SciTech Connect (OSTI)

    Nakayama, Guen; Akashi, Masatsune

    1995-09-01

    For preventing SCC from occurring in the internal structure of materials of the BWR plant, the injection of hydrogen into the core-water so as to reduce the free corrosion potential of the materials were proposed. Because of the lack of basic data of stress-corrosion cracking susceptibility in BWR environment on Ni-based alloys in comparison with stainless steels, the slow strain-rate tensile (SSRT) tests and the creviced bent-beam (CBB) test were conducted for a sensitized Alloy 182 weld metal in high-purity water environments containing dissolved oxygen (DO) and hydrogen (DH) to varied concentrations at 288 C, and the SCC initiation process were examined. The susceptibility of a material to SCC was discussed in terms of the electrode potential effect, and the effects of impurities of the testing water were examined by adding slightly Na{sub 2}, SO{sub 4}. In high purity waters and in the electrode potential region higher than {minus} 0.2 V vs. SHE, the interdendritic stress-corrosion cracks were observed both in the slow strain-rate test and the creviced bent-beam test. SEM observations of sub-cracks at the specimen surfaces revealed that stress-corrosion cracks were initiated when the oxide film had cracked to under-hundred {micro}m wide, that no such individual cracks could grow per se, but that those micro-cracks which happened to be formed in each other`s vicinity would coalesce into large cracks, one of which made propagated as stress-corrosion cracking, and that the stress-corrosion cracking sensitivity became more acute on addition of impurity. In the electrode potential region lower than 0 V, on the other hand, the stress-corrosion cracks were observed to be initiated at bottoms of corrosion pits formed on the specimen surfaces in the former, whereas both type of stress-corrosion cracks were observed between 0 to {minus}0.2V. No stress-corrosion crack was observed even though much the same corrosion pits in the CBB test at {minus}0.4 V.

  12. Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    SciTech Connect (OSTI)

    Marshall, William BJ J; Ade, Brian J; Bowman, Stephen M; Gauld, Ian C; Ilas, Germina; Mertyurek, Ugur; Radulescu, Georgeta

    2015-01-01

    Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (keff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup

  13. TRACE/PARCS Core Modeling of a BWR/5 for Accident Analysis of ATWS Events

    SciTech Connect (OSTI)

    Cuadra A.; Baek J.; Cheng, L.; Aronson, A.; Diamond, D.; Yarsky, P.

    2013-11-10

    The TRACE/PARCS computational package [1, 2] isdesigned to be applicable to the analysis of light water reactor operational transients and accidents where the coupling between the neutron kinetics (PARCS) and the thermal-hydraulics and thermal-mechanics (TRACE) is important. TRACE/PARCS has been assessed for itsapplicability to anticipated transients without scram(ATWS) [3]. The challenge, addressed in this study, is to develop a sufficiently rigorous input model that would be acceptable for use in ATWS analysis. Two types of ATWS events were of interest, a turbine trip and a closure of main steam isolation valves (MSIVs). In the first type, initiated by turbine trip, the concern is that the core will become unstable and large power oscillations will occur. In the second type,initiated by MSIV closure,, the concern is the amount of energy being placed into containment and the resulting emergency depressurization. Two separate TRACE/PARCS models of a BWR/5 were developed to analyze these ATWS events at MELLLA+ (maximum extended load line limit plus)operating conditions. One model [4] was used for analysis of ATWS events leading to instability (ATWS-I);the other [5] for ATWS events leading to emergency depressurization (ATWS-ED). Both models included a large portion of the nuclear steam supply system and controls, and a detailed core model, presented henceforth.

  14. Containment venting as a mitigation technique for BWR Mark I plant ATWS

    SciTech Connect (OSTI)

    Harrington, R.M.

    1986-01-01

    Containment venting is studied as a mitigation strategy for preventing or delaying severe fuel damage following hypothetical BWR Anticipated Transient Without SCRAM (ATWS) accidents initiated by MSIV-closure, and compounded by failure of the Standby Liquid Control (SLC) system injection of sodium pentaborate solution and by the failure of manually initiated control rod insertion. The venting of primary containment after reaching 75 psia (0.52 MPa) is found to result in the release of the vented steam inside the reactor building, and to result in inadequate Net Positive Suction Head (NPSH) for any system pumping from the pressure suppression pool. CONTAIN code calculations show that personnel access to large portions of the reactor building would be lost soon after the initiation of venting and that the temperatures reached would be likely to result in independent equipment failures. It is concluded that containment venting would be more likely to cause or to hasten the onset of severe fuel damage than to prevent or to delay it.

  15. Generic BWR-4 degraded core in-vessel study. Status report

    SciTech Connect (OSTI)

    Not Available

    1984-11-01

    Original intent of this project was to produce a phenomenological study of the in-vessel degradation which occurs during the TQUX and TQUV sequences for a generic BWR-4 from the initiation of the FSAR Chapter 15 operational transient through core debris bed formation to the failure of the primary pressure boundary. Bounding calculations were to be performed for the two high pressure and low pressure non-LOCA scenarios to assess the uncertainties in the current state of knowledge regarding the source terms for containment integrity studies. Source terms as such were defined in terms of hydrogen generation, unreacted metal, and coolant inventroy, and in terms of the form, sequencing and mode of dispersal through the primary vessel boundary. Fission product release was not to be considered as part of this study. Premature termination of the project, however, led to the dicontinuation of work on an as is basis. Work on the in-core phase from the point of scram to core debris bed formation was largely completed. A preliminary scoping calculation on the debris bed phase had been initiated. This report documents the status of the study at termination.

  16. Application of the IBERDROLA RETRAN Licensing Methodology to the Confrentes BWR-6 110% Extended Power Uprate

    SciTech Connect (OSTI)

    Fuente, Rafael de la; Iglesias, Javier; Sedano, Pablo G.; Mata, Pedro

    2003-04-15

    IBERDROLA (Spanish utility) and IBERDROLA INGENIERIA (engineering branch) have been developing during the last 2 yr the 110% Extended Power Uprate Project for Cofrentes BWR-6. IBERDROLA has available an in-house design and licensing reload methodology that has been approved in advance by the Spanish Nuclear Regulatory Authority. This methodology has been applied to perform the nuclear design and the reload licensing analysis for Cofrentes cycles 12 and 13 and to develop a significant number of safety analyses of the Cofrentes Extended Power.Because the scope of the licensing process of the Cofrentes Extended Power Uprate exceeds the range of analysis included in the Cofrentes generic reload licensing process, it has been required to extend the applicability of the Cofrentes RETRAN model to the analysis of new transients. This is the case of the total loss of feedwater (TLFW) transient.The content of this paper shows the benefits of having an in-house design and licensing methodology and describes the process to extend the applicability of the Cofrentes RETRAN model to the analysis of new transients, particularly in this paper the TLFW transient.

  17. 100% MOX BWR experimental program design using multi-parameter representative

    SciTech Connect (OSTI)

    Blaise, P.; Fougeras, P.; Cathalau, S.

    2012-07-01

    A new multiparameter representative approach for the design of Advanced full MOX BWR core physics experimental programs is developed. The approach is based on sensitivity analysis of integral parameters to nuclear data, and correlations among different integral parameters. The representativeness method is here used to extract a quantitative relationship between a particular integral response of an experimental mock-up and the same response in a reference project to be designed. The study is applied to the design of the 100% MOX BASALA ABWR experimental program in the EOLE facility. The adopted scheme proposes an original approach to the problem, going from the initial 'microscopic' pin-cells integral parameters to the whole 'macroscopic' assembly integral parameters. This approach enables to collect complementary information necessary to optimize the initial design and to meet target accuracy on the integral parameters to be measured. The study has demonstrated the necessity of new fuel pins fabrication, fulfilling minimal costs requirements, to meet acceptable representativeness on local power distribution. (authors)

  18. Analysis of fission product revaporization in a BWR reactor cooling system during a station blackout accident

    SciTech Connect (OSTI)

    Yang, J.W.; Schmidt, E.; Cazzoli, E.; Khatib-Rahbar, M.

    1988-01-01

    This report presents a preliminary analysis of fission product revaporization in the Reactor Cooling System (RCS) after the vessel failure. The station blackout transient for BWR Mark I Power Plant is considered. The TRAPMELT3 models of evaporization, chemisorption, and the decay heating of RCS structures and gases are adopted in the analysis. The RCS flow models based on the density-difference between the RCS and containment pedestal region are developed to estimate the RCS outflow which carries the revaporized fission product to the containment. A computer code called REVAP is developed for the analysis. The REVAP is incorporated with the MARCH, TRAPMELT3 and NAUA codes of the Source Term Code Pack Package (STCP). The NAUA code is used to estimate the impact of revaporization on environmental release. The results show that the thermal-hydraulic conditions between the RCS and the pedestal region are important factors determining the magnitude of revaporization and subsequent release of the volatile fission product. 8 figs., 1 tab.

  19. Electric power monthly, March 1995

    SciTech Connect (OSTI)

    1995-03-20

    This report for March 1995, presents monthly electricity statistics for a wide audience including Congress, Federal and State agencies, the electric utility industry, and the general public. The purpose of this publication is to provide energy decisionmakers with accurate and timely information that may be used in forming various perspectives on electric issues that lie ahead.

  20. Electric power monthly

    SciTech Connect (OSTI)

    1995-08-01

    The Energy Information Administration (EIA) prepares the Electric Power Monthly (EPM) for a wide audience including Congress, Federal and State agencies, the electric utility industry, and the general public. This publication provides monthly statistics for net generation, fossil fuel consumption and stocks, quantity and quality of fossil fuels, cost of fossil fuels, electricity sales, revenue, and average revenue per kilowatthour of electricity sold. Data on net generation, fuel consumption, fuel stocks, quantity and cost of fossil fuels are also displayed for the North American Electric Reliability Council (NERC) regions. The EIA publishes statistics in the EPM on net generation by energy source, consumption, stocks, quantity, quality, and cost of fossil fuels; and capability of new generating units by company and plant. The purpose of this publication is to provide energy decisionmakers with accurate and timely information that may be used in forming various perspectives on electric issues that lie ahead.

  1. General Information

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

    information General Information JLF Contacts Request a Tour

  2. Instability due to a two recirculation pump trip in a BWR using RAMONA-4B computer code with 3D neutron kinetics

    SciTech Connect (OSTI)

    Cheng, H.S.; Rohatgi, U.S.

    1993-06-01

    An investigation was made of the potential for thermal-hydraulic instabilities coupled to neutronic feedback in a BWR due to a two recirculation pump trip event using the RAMONA-4B computer code with 3D neutron kinetics. It is concluded that a high-power (100%) and low-flow (75%) initial condition would most likely lead to in-phase density wave oscillations after the tripping of both recirculation pumps, and that RAMONA-4B is capable of predicting such thermal-hydraulic instabilities coupled to neutronic feedback in BWR and in SBWR.

  3. Workplace Charging Challenge Partner: Portland General Electric...

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

    The U.S. Department of Energy EV Project provided the funding for most of these charging stations, and manufacturers provided several others to PGE for demonstration purposes. Many ...

  4. Portland General Electric Co | Open Energy Information

    Open Energy Info (EERE)

    1,336,537.655 812,088 2008-04 69,252.201 698,275.983 710,362 49,100.6 587,547.939 96,607 12,663.109 193,242.595 1,116 36.174 444.97 23 131,052.084 1,479,511.487 808,108 2008-03...

  5. Portland General Electric Co | Open Energy Information

    Open Energy Info (EERE)

    Landing Page: www.portlandgeneral.comd Green Button Reference Page: energy.govarticlesgreen References: EIA Form EIA-861 Final Data File for 2010 - File1a1 Energy...

  6. QER- Comment of Portland General Electric

    Broader source: Energy.gov [DOE]

    Thanks, Karen. As per our email exchange earlier this week, we're attaching prepared comments from PGE to support the statement Jim Piro offered at the QER public meeting in Portland this past summer. We appreciate your flexibility on the submission deadline! Please let me know if you need anything else from us.

  7. Inventors Behind General Electric | GE Global Research

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

    studies, Bernardo puts his biochemistry background to work as a home brewer. He enjoys sports and watching movies. Published Work By Bernardo Cinelli Granular starch hydrolysis of...

  8. Inventors Behind General Electric | GE Global Research

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

    Ismail Sezal Ismail Sezal Research Engineer Turbomachinery Aero Click to email this to a friend (Opens in new window) Share on Facebook (Opens in new window) Click to share (Opens in new window) Click to share on LinkedIn (Opens in new window) Click to share on Tumblr (Opens in new window) "The technology landscape is rapidly changing. We must always be learning and adapting to changes to stay ahead." -Ismail Sezal Ismail loves a challenge. Whether he is coming up with the next big

  9. Inventors Behind General Electric | GE Global Research

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

    Sultan Shair Sultan Shair Research Engineer Composites Manufacturing Click to email this to a friend (Opens in new window) Share on Facebook (Opens in new window) Click to share (Opens in new window) Click to share on LinkedIn (Opens in new window) Click to share on Tumblr (Opens in new window) "I feel really energized coming to work because work is my playground." -Sultan Shair When Sultan Shair gets up in the morning, he doesn't think about going to work, he thinks about going to

  10. Inventors Behind General Electric | GE Global Research

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

    Adnan Bohori Adnan Bohori Manager Electromagnetic Systems Click to email this to a friend (Opens in new window) Share on Facebook (Opens in new window) Click to share (Opens in new window) Click to share on LinkedIn (Opens in new window) Click to share on Tumblr (Opens in new window) "Every day is new, filled with unknowns and different challenges." -Adnan Bohori Adnan Bohori likes to think out of the box. Whether it's working with his team in Bangalore to build a new technology space

  11. Inventors Behind General Electric | GE Global Research

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

    Amol Kolwalkar Amol Kolwalkar Senior Engineer Control & Optimization Systems Click to email this to a friend (Opens in new window) Share on Facebook (Opens in new window) Click to share (Opens in new window) Click to share on LinkedIn (Opens in new window) Click to share on Tumblr (Opens in new window) "What motivates me is the thrill and anticipation of finding something that was not previously identified." -Amol Kolwalkar Amol Kolwalkar epitomizes the GE Belief "deliver

  12. Inventors Behind General Electric | GE Global Research

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

    Leao Bruno Leao Lead Engineer Smart Systems CoE Click to email this to a friend (Opens in new window) Share on Facebook (Opens in new window) Click to share (Opens in new window) Click to share on LinkedIn (Opens in new window) Click to share on Tumblr (Opens in new window) "The resources are available to those who have the creativity. We all have the ability to turn knowledge into something amazing." -Bruno Leao Bruno began his academic career in control and automation engineering. He

  13. Inventors Behind General Electric | GE Global Research

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

    Dragan Filipovic Dragan Filipovic Senior Engineer Manufacturing Chemical & Materials Technologies Click to email this to a friend (Opens in new window) Share on Facebook (Opens in new window) Click to share (Opens in new window) Click to share on LinkedIn (Opens in new window) Click to share on Tumblr (Opens in new window) "Believing that "the sky is the limit" is what sets apart an ordinary researcher from a great one." -Dragan Filipovic Dragan is a believer. He believes

  14. Inventors Behind General Electric | GE Global Research

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

    Fabio Fonseca Fabio Fonseca Lead Engineer Software and Productivity Analytics CoE Click to email this to a friend (Opens in new window) Share on Facebook (Opens in new window) Click to share (Opens in new window) Click to share on LinkedIn (Opens in new window) Click to share on Tumblr (Opens in new window) "I try to keep the big picture in mind, especially the positive impact I'm creating with my research." -Fabio Fonseca Fabio started with GE in 2012, where he focused on middleware

  15. Inventors Behind General Electric | GE Global Research

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

    Hari NS Hari NS Manager Electrochemistry, Corrosion & Tribology Click to email this to a friend (Opens in new window) Share on Facebook (Opens in new window) Click to share (Opens in new window) Click to share on LinkedIn (Opens in new window) Click to share on Tumblr (Opens in new window) "Working on large programs over multiple years has taught me to develop the patience and resilience required to pursue the ideas in which I believe." -Hari NS Hari NS has many accomplishments of

  16. Inventors Behind General Electric | GE Global Research

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

    Lucas Malta Lucas Malta Program Manager Smart Systems CoE Click to email this to a friend (Opens in new window) Share on Facebook (Opens in new window) Click to share (Opens in new window) Click to share on LinkedIn (Opens in new window) Click to share on Tumblr (Opens in new window) "My role gives me the opportunity to create something from scratch every day. " -Lucas Malta Before Lucas was leading a program in the Smart Systems CoE, he was paving the way in data analytics for the

  17. Inventors Behind General Electric | GE Global Research

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

    Prasanth Kumar Prasanth Kumar Senior Scientist Ceramics Synthesis & Processing Click to email this to a friend (Opens in new window) Share on Facebook (Opens in new window) Click to share (Opens in new window) Click to share on LinkedIn (Opens in new window) Click to share on Tumblr (Opens in new window) "I am constantly discovering and rediscovering myself in different application spaces." -Prasanth Kumar Prasanth Kumar's passion - both for his work and his personal life - is

  18. Inventors Behind General Electric | GE Global Research

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

    Ricardo Hernandez Pereira Ricardo Hernandez Pereira Combustion Engineer - Turbines & Reciprocating Engines Bioenergy Systems Click to email this to a friend (Opens in new window) Share on Facebook (Opens in new window) Click to share (Opens in new window) Click to share on LinkedIn (Opens in new window) Click to share on Tumblr (Opens in new window) "Contemporary machines perform in ways early designers would consider unfeasible. 50 years into the future, people will have a similar

  19. Inventors Behind General Electric | GE Global Research

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

    Sundeep Kumar Sundeep Kumar Senior Scientist Ceramics Synthesis & Processing Click to email this to a friend (Opens in new window) Share on Facebook (Opens in new window) Click to share (Opens in new window) Click to share on LinkedIn (Opens in new window) Click to share on Tumblr (Opens in new window) "I leverage the experience and network I have built over the years working with some of the finest minds at GE to help me solve tough technical problems for GE." -Sundeep Kumar

  20. Parametric Analysis of a Turbine Trip Event in a BWR Using a 3D Nodal Code

    SciTech Connect (OSTI)

    Gorzel, A.

    2006-07-01

    Two essential thermal hydraulics safety criteria concerning the reactor core are that even during operational transients there is no fuel melting and not-permissible cladding temperatures are avoided. A common concept for boiling water reactors is to establish a minimum critical power ratio (MCPR) for steady state operation. For this MCPR it is shown that only a very small number of fuel rods suffers a short-term dryout during the transient. It is known from experience that the limiting transient for the determination of the MCPR is the turbine trip with blocked bypass system. This fast transient was simulated for a German BWR by use of the three-dimensional reactor analysis transient code SIMULATE-3K. The transient behaviour of the hot channels was used as input for the dryout calculation with the transient thermal hydraulics code FRANCESCA. By this way the maximum reduction of the CPR during the transient could be calculated. The fast increase in reactor power due to the pressure increase and to an increased core inlet flow is limited mainly by the Doppler effect, but automatically triggered operational measures also can contribute to the mitigation of the turbine trip. One very important method is the short-term fast reduction of the recirculation pump speed which is initiated e. g. by a pressure increase in front of the turbine. The large impacts of the starting time and of the rate of the pump speed reduction on the power progression and hence on the deterioration of CPR is presented. Another important procedure to limit the effects of the transient is the fast shutdown of the reactor that is caused when the reactor power reaches the limit value. It is shown that the SCRAM is not fast enough to reduce the first power maximum, but is able to prevent the appearance of a second - much smaller - maximum that would occur around one second after the first one in the absence of a SCRAM. (author)

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

    SciTech Connect (OSTI)

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

    2014-06-01

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

  2. 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.

  3. 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 MOX fuel is generally obtained from reprocessed irradiated nuclear fuel, whereas weapons-grade plutonium is obtained from decommissioned nuclear weapons material and thus has a different plutonium (and other actinides) concentration. Using MOX fuel instead of UOX fuel has potential impacts on the neutronic performance of the nuclear fuel and the design of the nuclear fuel must take these differences into account. Each of the plutonium sources (RG and WG) has different implications on the neutronic behavior of the fuel because each contains a different blend of plutonium nuclides. The amount of heat and the number of neutrons produced from fission of plutonium nuclides is different from fission of {sup 235}U. These differences in UOX and MOX do not end at discharge of the fuel from the reactor core - the short- and long-term storage of MOX fuel may have different requirements than UOX fuel because of the different discharged fuel decay heat characteristics. The research documented in this report compares MOX and UOX fuel during storage and disposal of the fuel by comparing decay heat rates for typical pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies with and without weapons-grade (WG) and reactor-grade (RG) MOX fuel.

  4. Analysis of high-pressure boiloff situation during an MSIV closure ATWS in a typical BWR/4

    SciTech Connect (OSTI)

    Neymotin, L.Y.; Slovik, G.C.; Saha, P.

    1986-01-01

    An anticipated transient without scram (ATWS) is recognized as one of the boiling water reactor (BWR) accident sequences potentially leading to core damage. Of all the various ATWS initiating events, the main steam isolation valve (MSIV) closure ATWS is the most severe, because of its relatively high frequency of occurrence and its challenge to the residual heat removal and containment integrity systems. Although under investigation for quite a long period of time, different aspects of this type of transient are still being analyzed. The final outcome of these studies should be a well-defined set of recommendations for the plant operator to mitigate an ATWS accident. The objective of this paper is to provide a best estimate analysis of the MSIV closure ATWS in the Browns Ferry Unit 1 BWR with Mark-1 containment. The calculations have been performed using the RAMONA-3B code which as a three-dimensional neutron kinetics model coupled with one-dimensional four-equation, nonhomogeneous, nonequilibrium thermal hydraulics. The code also allows for one-dimensional neutronic core representation. The one-dimensional capability of the code has been employed in this calculation since a thorough sensitivity study showed that for a full ATWS, a one-dimensional neutron kinetics adequately describes the core behavior. The calculation described in the paper was started from a steady-state fuel condition corresponding to the end of cycle 5 of the Browns Ferry reactor.

  5. Efficiency and accuracy of the perturbation response coefficient generation method for whole core comet calculations in BWR and CANDU configurations

    SciTech Connect (OSTI)

    Zhang, D.; Rahnema, F.

    2013-07-01

    The coarse mesh transport method (COMET) is a highly accurate and efficient computational tool which predicts whole-core neutronics behaviors for heterogeneous reactor cores via a pre-computed eigenvalue-dependent response coefficient (function) library. Recently, a high order perturbation method was developed to significantly improve the efficiency of the library generation method. In that work, the method's accuracy and efficiency was tested in a small PWR benchmark problem. This paper extends the application of the perturbation method to include problems typical of the other water reactor cores such as BWR and CANDU bundles. It is found that the response coefficients predicted by the perturbation method for typical BWR bundles agree very well with those directly computed by the Monte Carlo method. The average and maximum relative errors in the surface-to-surface response coefficients are 0.02%-0.05% and 0.06%-0.25%, respectively. For CANDU bundles, the corresponding quantities are 0.01%-0.05% and 0.04% -0.15%. It is concluded that the perturbation method is highly accurate and efficient with a wide range of applicability. (authors)

  6. Computational fluid dynamics modeling of two-phase flow in a BWR fuel assembly. Final CRADA Report.

    SciTech Connect (OSTI)

    Tentner, A.; Nuclear Engineering Division

    2009-10-13

    A direct numerical simulation capability for two-phase flows with heat transfer in complex geometries can considerably reduce the hardware development cycle, facilitate the optimization and reduce the costs of testing of various industrial facilities, such as nuclear power plants, steam generators, steam condensers, liquid cooling systems, heat exchangers, distillers, and boilers. Specifically, the phenomena occurring in a two-phase coolant flow in a BWR (Boiling Water Reactor) fuel assembly include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. Traditionally, the best analysis tools for this purpose of two-phase flow phenomena inside the BWR fuel assembly have been the sub-channel codes. However, the resolution of these codes is too coarse for analyzing the detailed intra-assembly flow patterns, such as flow around a spacer element. Advanced CFD (Computational Fluid Dynamics) codes provide a potential for detailed 3D simulations of coolant flow inside a fuel assembly, including flow around a spacer element using more fundamental physical models of flow regimes and phase interactions than sub-channel codes. Such models can extend the code applicability to a wider range of situations, which is highly important for increasing the efficiency and to prevent accidents.

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

    SciTech Connect (OSTI)

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

    1983-01-01

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

  8. An assessment of BWR (boiling water reactor) Mark-II containment challenges, failure modes, and potential improvements in performance

    SciTech Connect (OSTI)

    Kelly, D.L.; Jones, K.R.; Dallman, R.J. ); Wagner, K.C. )

    1990-07-01

    This report assesses challenges to BWR Mark II containment integrity that could potentially arise from severe accidents. Also assessed are some potential improvements that could prevent core damage or containment failure, or could mitigate the consequences of such failure by reducing the release of fission products to the environment. These challenges and improvements are analyzed via a limited quantitative risk/benefit analysis of a generic BWR/4 reactor with Mark II containment. Point estimate frequencies of the dominant core damage sequences are obtained and simple containment event trees are constructed to evaluate the response of the containment to these severe accident sequences. The resulting containment release modes are then binned into source term release categories, which provide inputs to the consequence analysis. The output of the consequences analysis is used to construct an overall base case risk profile. Potential improvements and sensitivities are evaluated by modifying the event tree spilt fractions, thus generating a revised risk profile. Several important sensitivity cases are examined to evaluate the impact of phenomenological uncertainties on the final results. 75 refs., 25 figs., 65 tabs.

  9. Electric power annual 1992

    SciTech Connect (OSTI)

    Not Available

    1994-01-06

    The Electric Power Annual presents a summary of electric utility statistics at national, regional and State levels. The objective of the publication is to provide industry decisionmakers, government policymakers, analysts and the general public with historical data that may be used in understanding US electricity markets. The Electric Power Annual is prepared by the Survey Management Division; Office of Coal, Nuclear, Electric and Alternate Fuels; Energy Information Administration (EIA); US Department of Energy. ``The US Electric Power Industry at a Glance`` section presents a profile of the electric power industry ownership and performance, and a review of key statistics for the year. Subsequent sections present data on generating capability, including proposed capability additions; net generation; fossil-fuel statistics; retail sales; revenue; financial statistics; environmental statistics; electric power transactions; demand-side management; and nonutility power producers. In addition, the appendices provide supplemental data on major disturbances and unusual occurrences in US electricity power systems. Each section contains related text and tables and refers the reader to the appropriate publication that contains more detailed data on the subject matter. Monetary values in this publication are expressed in nominal terms.

  10. Risk evaluation of the alternate-3A modification to the ATWS prevention/mitigation system in a BWR-4, Mark-II power plant

    SciTech Connect (OSTI)

    Papazoglou, I.A.; Karol, R.; Shiu, K.; Bari, R.A.

    1983-01-01

    Purpose of this paper is to present a risk evaluation of the ATWS Alternate 3A modification (ATWS-3A) proposed by NRC staff in NUREG-0460 to the ATWS prevention/mitigation system in a BWR nuclear power plant. The evaluation is done relative to three risk indices: the frequency of core damage, the expected early fatalities, and the expected latent fatalities.

  11. Mixed-oxide fuel decay heat analysis for BWR LOCA safety evaluation

    SciTech Connect (OSTI)

    Chiang, R. T.

    2013-07-01

    The mixed-oxide (MOX) fuel decay heat behavior is analyzed for Boiling Water Reactor (BWR) Loss of Coolant Accident (LOCA) safety evaluation. The physical reasoning on why the decay heat power fractions of MOX fuel fission product (FP) are significantly lower than the corresponding decay heat power fractions of uranium-oxide (UOX) fuel FP is illustrated. This is primarily due to the following physical phenomena. -The recoverable energies per fission of plutonium (Pu)-239 and Pu-241 are significantly higher than those of uranium (U)-235 and U-238. Consequently, the fission rate required to produce the same amount of power in MOX fuel is significantly lower than that in UOX fuel, which leads to lower subsequent FP generation rate and associated decay heat power in MOX fuel than those in UOX fuel. - The effective FP decay energy per fission of Pu-239 is significantly lower than the corresponding effective FP decay energy per fission of U-235, e.g., Pu-239's 10.63 Mega-electron-Volt (MeV) vs. U-235's 12.81 MeV at the cooling time 0.2 second. This also leads to lower decay heat power in MOX fuel than that in UOX fuel. The FP decay heat is shown to account for more than 90% of the total decay heat immediately after shutdown. The FP decay heat results based on the American National Standard Institute (ANSI)/American Nuclear Society (ANS)-5.1-1979 standard method are shown very close to the corresponding FP decay heat results based on the ANSI/ANS-5.1-2005 standard method. The FP decay heat results based on the ANSI/ANS-5.1-1979 simplified method are shown very close to but mostly slightly lower than the corresponding FP decay heat results based on the ANSI/ANS-5.1-1971 method. The FP decay heat results based on the ANSI/ANS-5.1-1979 simplified method or the ANSI/ANS-5.1-1971 method are shown significantly larger than the corresponding FP decay heat results based on the ANSI/ANS-5.1-1979 standard method or the ANSI/ANS-5.1-2005 standard method. (authors)

  12. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    SciTech Connect (OSTI)

    Bevard, Bruce Balkcom; Mertyurek, Ugur; Belles, Randy; Scaglione, John M.

    2015-10-01

    utilized or referenced, justification has been provided as to why the data can be utilized for BWR fuel.

  13. Mexican Electric Research Institute IIE | Open Energy Information

    Open Energy Info (EERE)

    Mexican Electric Research Institute IIE Jump to: navigation, search Name: Mexican Electric Research Institute (IIE) Place: Mexico Sector: Services Product: General Financial &...

  14. An Introduction to Electric Power Transmission | Open Energy...

    Open Energy Info (EERE)

    An Introduction to Electric Power Transmission Jump to: navigation, search OpenEI Reference LibraryAdd to library General: An Introduction to Electric Power Transmission Abstract...

  15. Healthcare Energy: Spotlight on Lighting and Other Electric Loads...

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

    See below for a few highlights from monitoring lighting and other electric loads. Lighting and Other Electric Loads at the Gray Building For the Massachusetts General Hospital Gray ...

  16. National Electrical Manufacturers Association

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

    July 24, 2014 VIA EMAIL TO: Regulatory.Review@hq.doe.gov Steven Croley, General Counsel Office of the General Counsel U.S. Department of Energy 1000 Independence Avenue SW., Washington, DC 20585 NEMA Comments on DOE Reducing Regulatory Burden RFI 79 Fed.Reg. 28518 (July 3, 2014) Dear Mr. Croley, The National Electrical Manufacturers Association (NEMA) thanks you for the opportunity to provide comments on the Department of Energy's efforts to make its regulatory program more effective and less

  17. Application to Export Electric Energy OE Docket No. EA-97-D Portland

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

    General Electric Company | Department of Energy Company Application to Export Electric Energy OE Docket No. EA-97-D Portland General Electric Company Application from PGE to export electric energy to Canada. EA-97-D PGE (CN).pdf (2.49 MB) More Documents & Publications EA-97-D Portland General Electric Company Application to Export Electric Energy OE Docket No. EA-97-D Portland General Electric Company: Federal Register Notice, Volume 79, No. 103 - May 29, 2014 Application to Export

  18. Safety and licensing issues that are being addressed by the Power Burst Facility test programs. [PWR; BWR

    SciTech Connect (OSTI)

    McCardell, R.K.; MacDonald, P.E.

    1980-01-01

    This paper presents an overview of the results of the experimental program being conducted in the Power Burst Facility and the relationship of these results to certain safety and licensing issues. The safety issues that were addressed by the Power-Cooling-Mismatch, Reactivity Initiated Accident, and Loss of Coolant Accident tests, which comprised the original test program in the Power Burst Facility, are discussed. The resolution of these safety issues based on the results of the thirty-six tests performed to date, is presented. The future resolution of safety issues identified in the new Power Burst Facility test program which consists of tests which simulate BWR and PWR operational transients, anticipated transients without scram, and severe fuel damage accidents, is described.

  19. Review results of a BWR standard plant PRA and an assessment of potential benefits from design modifications

    SciTech Connect (OSTI)

    Shiu, K.; Hanan, N.; Rubin, M.

    1985-01-01

    Brookhaven National Laboratory (BNL) has participated in the review of the GESSAR II Standard Boiling Water Reactor (BWR) Plant probabilistic risk assessment (PRA). One major objective of this review was to utilize the PRA as a tool for investigation of the relative benefits available for incorporation of various proposed modifications to the baseline design. This paper presents the findings of the BNL review and assessment of the impact upon core damage frequency from two suggested design modifications. This work was restricted to consideration of interal events only. Review results indicated that the point estimate core damage frequency of the GESSAR II plant is equal to 2.2 x 10/sup -5//reactor-year for a plant site located within the Mid-Atlantic Area Council Grid (MAAC) and 3.8 x 10/sup -5//reactor-year if the national average loss of offsite power initiator frequency is used.

  20. Analysis of high pressure boil-off situation during MSIV closure ATWS in a typical BWR/4

    SciTech Connect (OSTI)

    Neymotin, L.Y.; Slovik, G.C.; Saha, P.

    1986-01-01

    The objective of this paper is to provide a best-estimate analysis of the MSIV Closure ATWS in the Browns Ferry Unit 1 BWR with Mark 1 containment. The calculations have been performed using the RAMONA-3B code which has a three-dimensional neutron kinetics model coupled with one-dimensional (multi-channel core representation), four-equation, nonhomogeneous, nonequilibrium thermal hydraulics. The code also allows for one-dimensional neutronic core representation. The 1-D capability of the code has been employed in this calculation since a thorough sensitivity study showed that for a full ATWS, a one-dimensional (axial) neutron kinetics adequately describes the core behavior. (Note that the core steady-state symmetry in this case was preserved throughout the transient so that radial effects could be neglected.) The calculation described in the paper was started from a steady-state fuel condition corresponding to the end of Cycle 5 of the Browns Ferry reactor.

  1. Field test and evaluation of the IAEA coincidence collar for the measurement of unirradiated BWR fuel assemblies

    SciTech Connect (OSTI)

    Menlove, H.O.; Keddar, A.

    1982-12-01

    The neutron coincidence counter has been field tested and evaluated for the measurement of boiling-water-reactor (BWR) fuel assemblies at the ASEA-ATOM Fuel Fabrication Facility. The system measures the /sup 235/U content per unit length of full fuel assemblies using neutron interrogation and coincidence counting. The /sup 238/U content is measured in the passive mode without the AmLi neutron interrogatioin source. The field tests included both standard production movable fuel rods to investigate enrichment and absorber variations. Results gave a response standard deviation of 0.9% for the active case and 2.1% for the passive case in 1000-s measurement times. 10 figures, 2 tables.

  2. Celebrating Electric Vehicles | Department of Energy

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

    Electric Vehicles Celebrating Electric Vehicles September 29, 2015 - 4:01pm Addthis The United States has the largest electric vehicle fleet in the world, which includes cars like the Chevrolet Volt. | Photo courtesy of General Motors The United States has the largest electric vehicle fleet in the world, which includes cars like the Chevrolet Volt. | Photo courtesy of General Motors Paul Lester Paul Lester Digital Content Specialist, Office of Public Affairs KEY FACTS More than 1 million plug-in

  3. EA-376 Societe Generale Energy Corp | Department of Energy

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

    6 Societe Generale Energy Corp EA-376 Societe Generale Energy Corp Order authorizing Societe Generale Energy Corp to export electric energy to Canada EA-376 Societe Generale Energy Corp (3.53 MB) More Documents & Publications Application to Export Electric Energy OE Docket No. EA-376 Societe Generale Energy Corp. EA-171-D Powerex Corp. EA-145-E Powerex Corp.

  4. Electric Power monthly, November 1996

    SciTech Connect (OSTI)

    1996-11-01

    This publication presents monthly electricity statistics for a wide audience including Congress, Federal and state agencies, the electric utility industry, and the general public. Purpose is to provide energy decisionmakers with accurate and timely information that may be used in forming various perspectives on electric issues that lie ahead. EIA collected the information in this report to fulfill its data collection and dissemination responsibilities as specified in the Federal Energy Administration Act of 1974 (Public Law 93-275) as amended.

  5. Electric power monthly, May 1996

    SciTech Connect (OSTI)

    1996-05-01

    This publication presents monthly electricity statistics for a wide audience including Congress, Federal and Stage agencies, the electric utility industry, and the general public. Purpose is to provide energy decisionmakers with accurate and timely information that may be used in forming various perspectives on electric issues that lie ahead. EIA collected the information to fulfill its data collection and dissemination responsibilities in Federal Energy Administration Act of 1974 (Public Law 93-275) as amended.

  6. Electric Vehicles

    ScienceCinema (OSTI)

    Ozpineci, Burak

    2014-07-23

    Burak Ozpineci sees a future where electric vehicles charge while we drive them down the road, thanks in part to research under way at ORNL.

  7. Electrical Engineer

    Broader source: Energy.gov [DOE]

    Transmission Field Services is responsible for field switching operation and maintenance of Bonneville Power Administration's high-voltage electrical transmission system to provide safe, reliable,...

  8. Electrical Safety

    Office of Environmental Management (EM)

    Handbook that was originally issued in 1998, and revised in 2004. DOE handbooks are ... the National Fire Protection Association (NFPA) 70, the National Electrical Code (NEC), ...

  9. Electric Vehicles

    SciTech Connect (OSTI)

    Ozpineci, Burak

    2014-05-02

    Burak Ozpineci sees a future where electric vehicles charge while we drive them down the road, thanks in part to research under way at ORNL.

  10. General Engineers

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

    General Engineers The U.S. Energy Information Administration (EIA) within the Department of Energy has forged a world-class information program that stresses quality, teamwork, and employee growth. In support of our program, we offer a variety of profes- sional positions, including the General Engineer, whose work is associated with analytical studies and evaluation projects pertaining to the operations of the energy industry. Responsibilities: General Engineers perform or participate in one or

  11. Registration Contact List: Electricity Transmission System Workshop

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

    ... Gordon H. Matthews General Engineer Bonneville Power Administration PO Box 3621 Portland OR 97208 United States 503-230-3275 Registration Contact List: Electricity Transmission ...

  12. Electric avenues

    SciTech Connect (OSTI)

    Stone, P.; Chang, A.

    1994-12-31

    Highly efficient electric drive technology developed originally for defense applications is being applied to the development of all electric shuttle buses for the San Jose International Airport. An innovative opportunity charging system using induction chargers will be incorporated to extend operation hours. The project, if successful, is expected to reduce pollution at the airport and generate jobs for displaced defense workers.

  13. Electric machine

    DOE Patents [OSTI]

    El-Refaie, Ayman Mohamed Fawzi; Reddy, Patel Bhageerath

    2012-07-17

    An interior permanent magnet electric machine is disclosed. The interior permanent magnet electric machine comprises a rotor comprising a plurality of radially placed magnets each having a proximal end and a distal end, wherein each magnet comprises a plurality of magnetic segments and at least one magnetic segment towards the distal end comprises a high resistivity magnetic material.

  14. Containment failure time and mode for a low-pressure short-term station blackout in a BWR-4 with Mark-I containment

    SciTech Connect (OSTI)

    Carbajo, J.J.; Greene, S.R. (Oak Ridge National Lab., TN (United States))

    1993-01-01

    This study investigates containment failure time and mode for a low-pressure, short-term station blackout severe accident sequence in a boiling water reactor (BWR-4) with a Mark-I containment. The severe accident analysis code MELCOR, version 1.8.1, was used in these calculations. Other results using the MELCOR/CORBH package and the BWRSAR and CONTAIN codes are also presented and compared to the MELCOR results. The plant analyzed is the Peach Bottom atomic station, a BWR-4 with a Mark-I containment. The automatic depressurization system was used to depressurize the vessel in accordance with the Emergency Procedure Guidelines. Two different variations of the station blackout were studied: one with a dry cavity and the other with a flooded cavity. For the flooded cavity, it is assumed that a control rod drive (CRD) pump becomes operational after vessel failure, and it is used to pump water into the cavity.

  15. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - models and correlations

    SciTech Connect (OSTI)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

    1998-03-01

    This document describes the major modifications and improvements made to the modeling of the RAMONA-3B/MOD0 code since 1981, when the code description and assessment report was completed. The new version of the code is RAMONA-4B. RAMONA-4B is a systems transient code for application to different versions of Boiling Water Reactors (BWR) such as the current BWR, the Advanced Boiling Water Reactor (ABWR), and the Simplified Boiling Water Reactor (SBWR). This code uses a three-dimensional neutron kinetics model coupled with a multichannel, non-equilibrium, drift-flux, two-phase flow formulation of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients and instability issues. Chapter 1 is an overview of the code`s capabilities and limitations; Chapter 2 discusses the neutron kinetics modeling and the implementation of reactivity edits. Chapter 3 is an overview of the heat conduction calculations. Chapter 4 presents modifications to the thermal-hydraulics model of the vessel, recirculation loop, steam separators, boron transport, and SBWR specific components. Chapter 5 describes modeling of the plant control and safety systems. Chapter 6 presents and modeling of Balance of Plant (BOP). Chapter 7 describes the mechanistic containment model in the code. The content of this report is complementary to the RAMONA-3B code description and assessment document. 53 refs., 81 figs., 13 tabs.

  16. Electric power monthly, August 1994

    SciTech Connect (OSTI)

    Not Available

    1994-08-24

    The Electric Power Monthly (EPM) presents monthly electricity statistics. The purpose of this publication is to provide energy decisionmakers with accurate and timely information that may be used in forming various perspectives on electric issues that lie ahead. Data in this report are presented for a wide audience including Congress, Federal and State agencies, the electric utility industry, and the general public. The EIA collected the information in this report to fulfill its data collection and dissemination responsibilities as specified in the Federal Energy Administration Act of 1974 (Public Law 93-275) as amended.

  17. Electric power monthly, June 1994

    SciTech Connect (OSTI)

    Not Available

    1994-06-01

    The Electric Power Monthly (EPM) presents monthly electricity statistics. The purpose of this publication is to provide energy decisionmakers with accurate and timely information that may be used in forming various perspectives on electric issues that lie ahead. Data in this report are presented for a wide audience including Congress, Federal and State agencies, the electric utility industry, and the general public. The EIA collected the information in this report to fulfill its data collection and dissemination responsibilities as specified in the Federal Energy Administration Act of 1974 (Public Law 93-275) as amended.

  18. Electric power monthly, July 1993

    SciTech Connect (OSTI)

    Not Available

    1993-07-29

    The Electric Power Monthly (EPM) presents monthly electricity statistics. The purpose of this publication is to provide energy decisionmakers with accurate and timely information that may be used in forming various perspectives on electric issues that lie ahead. Data in this report are presented for a wide audience including Congress, Federal and State agencies, the electric utility industry, and the general public. The EIA collected the information in this report to fulfill its data collection and dissemination responsibilities as specified in the Federal Energy Administration Act of 1974 (Public Law 93-275) as amended.

  19. Electrical connector

    DOE Patents [OSTI]

    Dilliner, Jennifer L.; Baker, Thomas M.; Akasam, Sivaprasad; Hoff, Brian D.

    2006-11-21

    An electrical connector includes a female component having one or more receptacles, a first test receptacle, and a second test receptacle. The electrical connector also includes a male component having one or more terminals configured to engage the one or more receptacles, a first test pin configured to engage the first test receptacle, and a second test pin configured to engage the second test receptacle. The first test receptacle is electrically connected to the second test receptacle, and at least one of the first test pin and the second test pin is shorter in length than the one or more terminals.

  20. Electrical Safety

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

    ... Fig. 1-1. Flow down of Electrical AHJ and worker responsibility. 3 DOE-HDBK-1092-2013 2.0 ... When equipment contains storage batteries, workers should be protected from the various ...

  1. General Engineer

    Broader source: Energy.gov [DOE]

    This position is located in Office of Standard Contract Management, within the Office of the General Counsel (GC). The purpose of the position is to conduct technical and engineering reviews of the...

  2. Electric generator

    DOE Patents [OSTI]

    Foster, Jr., John S.; Wilson, James R.; McDonald, Jr., Charles A.

    1983-01-01

    1. In an electrical energy generator, the combination comprising a first elongated annular electrical current conductor having at least one bare surface extending longitudinally and facing radially inwards therein, a second elongated annular electrical current conductor disposed coaxially within said first conductor and having an outer bare surface area extending longitudinally and facing said bare surface of said first conductor, the contiguous coaxial areas of said first and second conductors defining an inductive element, means for applying an electrical current to at least one of said conductors for generating a magnetic field encompassing said inductive element, and explosive charge means disposed concentrically with respect to said conductors including at least the area of said inductive element, said explosive charge means including means disposed to initiate an explosive wave front in said explosive advancing longitudinally along said inductive element, said wave front being effective to progressively deform at least one of said conductors to bring said bare surfaces thereof into electrically conductive contact to progressively reduce the inductance of the inductive element defined by said conductors and transferring explosive energy to said magnetic field effective to generate an electrical potential between undeformed portions of said conductors ahead of said explosive wave front.

  3. Electric power monthly, May 1994

    SciTech Connect (OSTI)

    Not Available

    1994-05-01

    The Electric Power Monthly (EPM) presents monthly electricity statistics. The purpose of this publication is to provide energy decisionmakers with accurate and timely information that may be used in forming various perspectives on electric issues that lie ahead. Data in this report are presented for a wide audience including Congress, Federal and State agencies, the electric utility industry, and the general public. This publication provides monthly statistics for net generation, fossil fuel consumption and stocks, quantity and quality of fossil fuels, cost of fossil fuels, electricity sales, revenue, and average revenue per kilowatthour of electricity sold. Statistics by company and plant are published on the capability of new generating units, net generation, fuel consumption, fuel stocks, quantity and quality of fuel, and cost of fossil fuels.

  4. Simple Electric Vehicle Simulation

    Energy Science and Technology Software Center (OSTI)

    1993-07-29

    SIMPLEV2.0 is an electric vehicle simulation code which can be used with any IBM compatible personal computer. This general purpose simulation program is useful for performing parametric studies of electric and series hybrid electric vehicle performance on user input driving cycles.. The program is run interactively and guides the user through all of the necessary inputs. Driveline components and the traction battery are described and defined by ASCII files which may be customized by themore » user. Scaling of these components is also possible. Detailed simulation results are plotted on the PC monitor and may also be printed on a printer attached to the PC.« less

  5. Electric power annual 1997. Volume 1

    SciTech Connect (OSTI)

    1998-07-01

    The Electric Power Annual presents a summary of electric power industry statistics at national, regional, and State levels. The objective of the publication is to provide industry decisionmakers, government policy-makers, analysts, and the general public with data that may be used in understanding US electricity markets. The Electric Power Annual is prepared by the Electric Power Division; Office of Coal, Nuclear, Electric and Alternate Fuels; Energy Information Administration (EIA); US Department of Energy. Volume 1 -- with a focus on US electric utilities -- contains final 1997 data on net generation and fossil fuel consumption, stocks, receipts, and cost; preliminary 1997 data on generating unit capability, and retail sales of electricity, associated revenue, and the average revenue per kilowatthour of electricity sold (based on a monthly sample: Form EIA-826, ``Monthly Electric Utility Sales and Revenue Report with State Distributions``). Additionally, information on net generation from renewable energy sources and on the associated generating capability is included in Volume 1 of the EPA.

  6. 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 BWR Boiling Water Reactor. Source: Form EIA-860, "Annual Electric ...

  7. General Publications

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

    General Publications General Publications Print ALS Strategic Plan 2015-19 cover image An updated version of the ALS Strategic Plan, covering the five-year period from 2015 to 2019. As in the 2014-18 version, Section I gives a brief synopsis on beamline and endstation projects. The science drivers behind these projects are explained in greater detail in Section II, and a very brief description of emerging plans for a ALS-U are in Section III. Soft X-ray Science Opportunities Using

  8. Typical BWR/4 MSIV closure ATWS analysis using RAMONA-3B code with space-time neutron kinetics

    SciTech Connect (OSTI)

    Neymotin, L.; Saha, P.

    1984-01-01

    A best-estimate analysis of a typical BWR/4 MSIV closure ATWS has been performed using the RAMONA-3B code with three-dimensional neutron kinetics. All safety features, namely, the safety and relief valves, recirculation pump trip, high pressure safety injections and the standby liquid control system (boron injection), were assumed to work as designed. No other operator action was assumed. The results show a strong spatial dependence of reactor power during the transient. After the initial peak of pressure and reactor power, the reactor vessel pressure oscillated between the relief valve set points, and the reactor power oscillated between 20 to 50% of the steady state power until the hot shutdown condition was reached at approximately 1400 seconds. The suppression pool bulk water temperature at this time was predicted to be approx. 96/sup 0/C (205/sup 0/F). In view of code performance and reasonable computer running time, the RAMONA-3B code is recommended for further best-estimate analyses of ATWS-type events in BWRs.

  9. Development of Probabilistic Risk Assessment Model for BWR Shutdown Modes 4 and 5 Integrated in SPAR Model

    SciTech Connect (OSTI)

    S. T. Khericha; S. Sancakter; J. Mitman; J. Wood

    2010-06-01

    Nuclear plant operating experience and several studies show that the risk from shutdown operation during modes 4, 5, and 6 can be significant This paper describes development of the standard template risk evaluation models for shutdown modes 4, and 5 for commercial boiling water nuclear power plants (BWR). The shutdown probabilistic risk assessment model uses full power Nuclear Regulatory Commissions (NRCs) Standardized Plant Analysis Risk (SPAR) model as the starting point for development. The shutdown PRA models are integrated with their respective internal events at-power SPAR model. This is accomplished by combining the modified system fault trees from SPAR full power model with shutdown event tree logic. For human reliability analysis (HRA), the SPAR HRA (SPAR-H) method is used which requires the analysts to complete relatively straight forward worksheet, including the performance shaping factors (PSFs). The results are then used to estimate HEP of interest. The preliminary results indicate the risk is dominated by the operators ability to diagnose the events and provide long term cooling.

  10. Analyzing simulation-based PRA data through traditional and topological clustering: A BWR station blackout case study

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

    Maljovec, D.; Liu, S.; Wang, B.; Mandelli, D.; Bremer, P. -T.; Pascucci, V.; Smith, C.

    2015-07-14

    Here, dynamic probabilistic risk assessment (DPRA) methodologies couple system simulator codes (e.g., RELAP and MELCOR) with simulation controller codes (e.g., RAVEN and ADAPT). Whereas system simulator codes model system dynamics deterministically, simulation controller codes introduce both deterministic (e.g., system control logic and operating procedures) and stochastic (e.g., component failures and parameter uncertainties) elements into the simulation. Typically, a DPRA is performed by sampling values of a set of parameters and simulating the system behavior for that specific set of parameter values. For complex systems, a major challenge in using DPRA methodologies is to analyze the large number of scenarios generated,more » where clustering techniques are typically employed to better organize and interpret the data. In this paper, we focus on the analysis of two nuclear simulation datasets that are part of the risk-informed safety margin characterization (RISMC) boiling water reactor (BWR) station blackout (SBO) case study. We provide the domain experts a software tool that encodes traditional and topological clustering techniques within an interactive analysis and visualization environment, for understanding the structures of such high-dimensional nuclear simulation datasets. We demonstrate through our case study that both types of clustering techniques complement each other for enhanced structural understanding of the data.« less

  11. Analysis of a typical BWR/4 MSIV closure ATWS using RAMONA-3B and TRAC-BD1 codes

    SciTech Connect (OSTI)

    Hsu, C.J.; Neymotin, L.; Saha, P.

    1984-01-01

    Analysis of a typical BWR/4 Anticipated Transient Without Scram (ATWS) has been performed using two advanced, best-estimate computer codes, namely, RAMONA-3B and TRAC-BD1. The transient was initiated by an inadvertant closure of all Main Steam Isolation Valves (MSIVs) with subsequent failure to scram the reactor. However, all other safety features namely, the safety and relief valves, recirculation pump trip, high pressure coolant injection and the standby liquid (boron) control system were assumed to work as designed. No other operator action was assumed. It has been found that both RAMONA-3B (with three-dimensional neutron kinetics) and TRAC-BD1 (with point kinetics) yielded similar results for the global parameters such as reactor power, system pressure and the suppression pool temperature. Both calculations showed that the reactor can be brought to hot shutdown in approximately twenty to twenty-five minutes with borated water mass flow rate of 2.78 kg/s (43 gpm) with 23800 ppM of boron. The suppression pool water temperature (assuming no pool cooling) at this time could be in the range of 170 to 205/sup 0/F. An additional TRAC-BD1 calculation with RAMONA-3B reactor power indicates that the thermal-hydraulic models in RAMONA-3B, although simpler than those in TRAC-BD1, can adequately represent the system behavior during the ATWS-type transient.

  12. Analysis of burnup and isotopic compositions of BWR 9 x 9 UO{sub 2} fuel assemblies

    SciTech Connect (OSTI)

    Suzuki, M.; Yamamoto, T.; Ando, Y.; Nakajima, T.

    2012-07-01

    In order to extend isotopic composition data focusing on fission product nuclides, measurements are progressing using facilities of JAEA for five samples taken from high burnup BWR 9 x 9 UO{sub 2} fuel assemblies. Neutronics analysis with an infinite assembly model was applied to the preliminary measurement data using a continuous-energy Monte Carlo burnup calculation code MVP-BURN with nuclear libraries based on JENDL-3.3 and JENDL-4.0. The burnups of the samples were determined to be 28.0, 39.3, 56.6, 68.1, and 64.0 GWd/t by the Nd-148 method. They were compared with those calculated using node-average irradiation histories of power and in-channel void fractions which were taken from the plant data. The comparison results showed that the deviations of the calculated burnups from the measurements were -4 to 3%. It was confirmed that adopting the nuclear data library based on JENDL-4.0 reduced the deviations of the calculated isotopic compositions from the measurements for {sup 238}Pu, {sup 144}Nd, {sup 145}Nd, {sup 146}Nd, {sup 148}Nd, {sup 134}Cs, {sup 154}Eu, {sup 152}Sm, {sup 154}Gd, and {sup 157}Gd. On the other hand, the effect of the revision in the nuclear. data library on the neutronics analysis was not significant for major U and Pu isotopes. (authors)

  13. Application Of The Iberdrola Licensing Methodology To The Cofrentes BWR-6 110% Extended Power Up-rate

    SciTech Connect (OSTI)

    Mata, Pedro; Fuente, Rafael de la; Iglesias, Javier; Sedano, Pablo G.

    2002-07-01

    Iberdrola (spanish utility) and Iberdrola Ingenieria (engineering branch) have been developing during the last two years the 110% Extended Power Up-rate Project (EPU 110%) for Cofrentes BWR-6. IBERDROLA has available an in-house design and licensing reload methodology that has been approved by the Spanish Nuclear Regulatory Authority. This methodology has been already used to perform the nuclear design and the reload licensing analysis for Cofrentes cycles 12 to 14. The methodology has been also applied to develop a significant number of safety analysis of the Cofrentes Extended Power Up-rate including: Reactor Heat Balance, Core and Fuel performance, Thermal Hydraulic Stability, ECCS LOCA Evaluation, Transient Analysis, Anticipated Transient Without Scram (ATWS) and Station Blackout (SBO) Since the scope of the licensing process of the Cofrentes Extended Power Up-rate exceeds the range of analysis included in the Cofrentes generic reload licensing process, it has been required to extend the applicability of the Cofrentes licensing methodology to the analysis of new transients. This is the case of the TLFW transient. The content of this paper shows the benefits of having an in-house design and licensing methodology, and describes the process to extend the applicability of the methodology to the analysis of new transients. The case of analysis of Total Loss of Feedwater with the Cofrentes Retran Model is included as an example of this process. (authors)

  14. Analysis of the FeCrAl Accident Tolerant Fuel Concept Benefits during BWR Station Blackout Accidents

    SciTech Connect (OSTI)

    Robb, Kevin R

    2015-01-01

    Iron-chromium-aluminum (FeCrAl) alloys are being considered for fuel concepts with enhanced accident tolerance. FeCrAl alloys have very slow oxidation kinetics and good strength at high temperatures. FeCrAl could be used for fuel cladding in light water reactors and/or as channel box material in boiling water reactors (BWRs). To estimate the potential safety gains afforded by the FeCrAl concept, the MELCOR code was used to analyze a range of postulated station blackout severe accident scenarios in a BWR/4 reactor employing FeCrAl. The simulations utilize the most recently known thermophysical properties and oxidation kinetics for FeCrAl. Overall, when compared to the traditional Zircaloy-based cladding and channel box, the FeCrAl concept provides a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. Finally, due to the slower oxidation kinetics, substantially less hydrogen is generated, and the generation is delayed in time. This decreases the amount of non-condensable gases in containment and the potential for deflagrations to inhibit the accident response.

  15. Analyzing simulation-based PRA data through traditional and topological clustering: A BWR station blackout case study

    SciTech Connect (OSTI)

    Maljovec, D.; Liu, S.; Wang, B.; Mandelli, D.; Bremer, P. -T.; Pascucci, V.; Smith, C.

    2015-07-14

    Here, dynamic probabilistic risk assessment (DPRA) methodologies couple system simulator codes (e.g., RELAP and MELCOR) with simulation controller codes (e.g., RAVEN and ADAPT). Whereas system simulator codes model system dynamics deterministically, simulation controller codes introduce both deterministic (e.g., system control logic and operating procedures) and stochastic (e.g., component failures and parameter uncertainties) elements into the simulation. Typically, a DPRA is performed by sampling values of a set of parameters and simulating the system behavior for that specific set of parameter values. For complex systems, a major challenge in using DPRA methodologies is to analyze the large number of scenarios generated, where clustering techniques are typically employed to better organize and interpret the data. In this paper, we focus on the analysis of two nuclear simulation datasets that are part of the risk-informed safety margin characterization (RISMC) boiling water reactor (BWR) station blackout (SBO) case study. We provide the domain experts a software tool that encodes traditional and topological clustering techniques within an interactive analysis and visualization environment, for understanding the structures of such high-dimensional nuclear simulation datasets. We demonstrate through our case study that both types of clustering techniques complement each other for enhanced structural understanding of the data.

  16. Commercial Miscellaneous Electric Loads Report: Energy Consumption

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

    Characterization and Savings Potential in 2008 by Building Type | Department of Energy Commercial Miscellaneous Electric Loads Report: Energy Consumption Characterization and Savings Potential in 2008 by Building Type Commercial Miscellaneous Electric Loads Report: Energy Consumption Characterization and Savings Potential in 2008 by Building Type Commercial miscellaneous electric loads (MELs) are generally defined as all electric loads except those related to main systems for heating,

  17. Electrically powered hand tool

    DOE Patents [OSTI]

    Myers, Kurt S.; Reed, Teddy R.

    2007-01-16

    An electrically powered hand tool is described and which includes a three phase electrical motor having a plurality of poles; an electrical motor drive electrically coupled with the three phase electrical motor; and a source of electrical power which is converted to greater than about 208 volts three-phase and which is electrically coupled with the electrical motor drive.

  18. Experiment data report for Multirod Burst Test (MRBT) bundle B-6. [PWR; BWR

    SciTech Connect (OSTI)

    Chapman, R H; Longest, A W; Crowley, J L

    1984-07-01

    A reference source of MRBT bundle B-6 test data is presented with minimum interpretation. The primary objective of this 8 x 8 multirod burst test was to investigate cladding deformation in the alpha-plus-beta-Zircaloy temperature range under simulated light-water-reactor (LWR) loss-of-coolant accident (LOCA) conditions. B-6 test conditions simulated the adiabatic heatup (reheat) phase of an LOCA and produced very uniform temperature distributions. The fuel pin simulators were electrically heated (average linear power generation of 1.42 kW/m) and were slightly cooled with a very low flow (Re approx. 140) of low-pressure superheated steam. The cladding temperature increased from the initial temperature (330/sup 0/C) to the burst temperature at a rate of 3.5/sup 0/C/s. The simulators burst in a very narrow temperature range, with an average of 930/sup 0/C. Cladding burst strain ranged from 21 to 56%, with an average of 31%. Volumetric expansion over the heated length of the cladding ranged from 16 to 32%, with an average of 23%. 23 references.

  19. Boundary effects on Zircaloy-4 cladding deformation in LOCA simulation tests. [PWR; BWR

    SciTech Connect (OSTI)

    Longest, A.W.; Chapman, R.H.; Crowley, J.L.

    1982-01-01

    Deformation behavior of Zircaloy-4 cladding under simulated loss-of-coolant accident (LOCA) conditions is being investigated in the Multirod Burst Test (MRBT) program in single rod and multirod tests. In these tests, internally-pressurized unirradiated Zircaloy-4 tubes containing internal electrical heaters are heated to failure in a low-pressure, superheated-steam environment (200 < Re < 800). The results provide a data base for evaluating deformation and blockage models employed with design-basis accident sequences to assess LWR core coolability for licensing purposes. Results of a recent 8 X 8 test indicate that models derived from smaller test arrays may not be representative of the behavior in large arrays, particularly for those temperature ranges in which large deformation can be expected. Two MRBT LOCA simulation tests conducted under the same nominal conditions (approx. 10 K/s heating rate from approx. 340/sup 0/C to failure at approx. 770/sup 0/C) were examined to determine the effects of array size and boundary conditions on deformation.

  20. Electric vehicle climate control

    SciTech Connect (OSTI)

    Dauvergne, J.

    1994-04-01

    EVs have insufficient energy sources for a climatic comfort system. The heat rejection of the drivetrain is dispersed in the vehicle (electric motor, batteries, electronic unit for power control). Its level is generally low (no more than 2-kW peaks) and variable according to the trip profile, with no heat rejection at rest and a maximum during regenerative braking. Nevertheless, it must be used for heating. It is not realistic to have the A/C compressor driven by the electric traction motor: the motor does not operate when the vehicle is at rest, precisely when maximum cooling power is required. The same is true for hybrid vehicles during electric operation. It is necessary to develop solutions that use stored onboard energy either from the traction batteries or specific storage source. In either case, it is necessary to design the climate control system to use the energy efficiently to maximize range and save weight. Heat loss through passenger compartment seals and the walls of the passenger compartment must be limited. Plastic body panes help to reduce heat transfer, and heat gain is minimized with insulating glazing. This article describes technical solutions to solve the problem of passenger thermal comfort. However, the heating and A/C systems of electrically operated vehicles may have marginal performance at extreme outside temperatures.

  1. Application to Export Electric Energy OE Docket No. EA-376 Societe...

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

    Application from Societe Generale Energy Corp to export electric energy to Canada ... Application to Export Electric Energy OE Docket No. EA-297-B SESCO Enterprises Canada

  2. Electric power monthly, June 1997 with data for March 1997

    SciTech Connect (OSTI)

    1997-06-01

    The Electric Power Monthly (EPM) presents monthly electricity statistics for a wide audience including Congress, Federal and State agencies, the electric utility industry, and the general public. The purpose of this publication is to provide energy decisionmakers with accurate and timely information that may be used in forming various perspectives on electric issues that lie ahead. 63 tabs.

  3. Electric power monthly, July 1997 with data for April 1997

    SciTech Connect (OSTI)

    1997-07-01

    The Electric Power Monthly (EPM) presents monthly electricity statistics for a wide audience including Congress, Federal and State agencies, the electric utility industry, and the general public. The purpose of this publication is to provide energy decisionmakers with accurate and timely information that may be used in forming various perspectives on electric issues that lie ahead. 57 tabs.

  4. Integrating The Non-Electrical Worker Into The Electrical Safety Program

    SciTech Connect (OSTI)

    Mills, T. David; McAlhaney, John H.

    2012-08-17

    The intent of this paper is to demonstrate an electrical safety program that incorporates all workers into the program, not just the electrical workers. It is largely in response to a paper presented at the 2012 ESW by Lanny Floyd entitled "Facilitating Application of Electrical Safety Best Practices to "Other" Workers" which requested all attendees to review their electrical safety program to assure that non-electrical workers were protected as well as electrical workers. The referenced paper indicated that roughly 50% of electrical incidents involve workers whose primary function is not electrical in nature. It also encouraged all to "address electrical safety for all workers and not just workers whose job responsibilities involve working on or near energized electrical circuits." In this paper, a program which includes specific briefings to non-electrical workers as well as to workers who may need to perform their normal activities in proximity to energized electrical conductors is presented. The program uses a targeted approach to specific areas such as welding, excavating, rigging, chart reading, switching, cord and plug equipment and several other general areas to point out hazards that may exist and how to avoid them. NFPA 70E-2004 was incorporated into the program several years ago and with it the need to include the "other" workers became apparent. The site experience over the years supports the assertion that about half of the electrical incidents involve non-electrical workers and this prompted us to develop specific briefings to enhance the knowledge of the non-electrical worker regarding safe electrical practices. The promotion of "May is Electrical Safety Month" and the development of informative presentations which are delivered to the general site population as well as electrical workers have greatly improved the hazards awareness status of the general worker on site.

  5. Electricity Monthly Update

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

    Contact Information and Staff The Electricity Monthly Update is prepared by the Electric Power Operations Team, Office of Electricity, Renewables and Uranium Statistics, U.S....

  6. Effects of thermocouple installation and location on fuel rod temperature measurements. [PWR; BWR

    SciTech Connect (OSTI)

    McCormick, R.D.

    1983-01-01

    This paper describes the results of analyses of nuclear fuel rod cladding temperature data obtained during in-reactor experiments under steady state and transient (simulated loss-of-coolant accident) operating conditions. The objective of the analyses was to determine the effect of thermocouple attachment method and location on measured thermal response. The use of external thermocouples increased the time to critical heat flux (CHF), reduced the blowdown peak temperature, and enhanced rod quench. A comparison of laser welded and resistance welded external thermocouple responses showed that the laser welding technique reduced the indicated cladding steady state temperatures and provided shorter time-to-CHF. A comparison of internal welded and embedded thermocouples indicated that the welded technique gave generally unsatisfactory cladding temperature measurements. The embedded thermocouple gave good, consistent results, but was possibly more fragile than the welded thermocouples. Detailed descriptions of the thermocouple designs, attachment methods and locations, and test conditions are provided.

  7. ORPS General Analysis Desk Guide | Department of Energy

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

    General Analysis Desk Guide ORPS General Analysis Desk Guide November 2015 The ORPS Desk Guide serves as a quick reference for ORPS account holders to navigate ORPS, narrow search fields, and produce various types of reports. (November 2015) ORPS General Analysis Desk Guide (2.39 MB) More Documents & Publications Monthly Analysis of Electrical Safety Occurrences - May 2012 Monthly Analysis of Electrical Safety Occurrences - June 2013 Monthly Analysis of Electrical Safety Occurrences -

  8. Electrical receptacle

    DOE Patents [OSTI]

    Leong, Robert

    1993-01-01

    The invention is a receptacle for a three prong electrical plug which has either a tubular or U-shaped grounding prong. The inventive receptacle has a grounding prong socket which is sufficiently spacious to prevent the socket from significantly stretching when a larger, U-shaped grounding prong is inserted into the socket, and having two ridges to allow a snug fit when a smaller tubular shape grounding prong is inserted into the socket. The two ridges are made to prevent the socket from expanding when either the U-shaped grounding prong or the tubular grounding prong is inserted.

  9. Electrical receptacle

    DOE Patents [OSTI]

    Leong, R.

    1993-06-22

    The invention is a receptacle for a three prong electrical plug which has either a tubular or U-shaped grounding prong. The inventive receptacle has a grounding prong socket which is sufficiently spacious to prevent the socket from significantly stretching when a larger, U-shaped grounding prong is inserted into the socket, and having two ridges to allow a snug fit when a smaller tubular shape grounding prong is inserted into the socket. The two ridges are made to prevent the socket from expanding when either the U-shaped grounding prong or the tubular grounding prong is inserted.

  10. Electrical machine

    DOE Patents [OSTI]

    De Bock, Hendrik Pieter Jacobus; Alexander, James Pellegrino; El-Refaie, Ayman Mohamed Fawzi; Gerstler, William Dwight; Shah, Manoj Ramprasad; Shen, Xiaochun

    2016-06-21

    An apparatus, such as an electrical machine, is provided. The apparatus can include a rotor defining a rotor bore and a conduit disposed in and extending axially along the rotor bore. The conduit can have an annular conduit body defining a plurality of orifices disposed axially along the conduit and extending through the conduit body. The rotor can have an inner wall that at least partially defines the rotor bore. The orifices can extend through the conduit body along respective orifice directions, and the rotor and conduit can be configured to provide a line of sight along the orifice direction from the respective orifices to the inner wall.

  11. Ellen C. Ginsberg VICE PRESIDENT, GENERAL COUNSEL & SECRETARY

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

    Ellen C. Ginsberg VICE PRESIDENT, GENERAL COUNSEL & SECRETARY March 20, 2012 Mr. ... Lopatto (Westinghouse Electric Company), and Bud Piland (The Babcock and Wilcox Company). ...

  12. Miniature electrical connector

    DOE Patents [OSTI]

    Casper, Robert F.

    1976-01-01

    A miniature coaxial cable electrical connector includes an annular compressible gasket in a receptacle member, the gasket having a generally triangular cross section resiliently engaging and encircling a conically tapered outer surface of a plug member to create an elongated current leakage path at their interface; means for preventing rotation of the plug relative to the receptacle; a metal sleeve forming a portion of the receptacle and encircling the plug member when interconnected; and a split ring in the plug having outwardly and rearwardly projecting fingers spaced from and encircling a portion of a coaxial cable and engageable with the metal sleeve to interlock the receptacle and plug.

  13. Electric power annual 1995. Volume I

    SciTech Connect (OSTI)

    1996-07-01

    The Electric Power Annual presents a summary of electric power industry statistics at national, regional, and State levels. The objective of the publication is to provide industry decisionmakers, government policymakers, analysts, and the general public with data that may be used in understanding U.S. electricity markets. The Electric Power Annual is prepared by the Coal and Electric Data and Renewables Division; Office of Coal, Nuclear, Electric and Alternate Fuels; Energy Information Administration (EIA); U.S. Department of Energy. In the private sector, the majority of the users of the Electric Power Annual are researchers and analysts and, ultimately, individuals with policy- and decisionmaking responsibilities in electric utility companies. Financial and investment institutions, economic development organizations interested in new power plant construction, special interest groups, lobbyists, electric power associations, and the news media will find data in the Electric Power Annual useful. In the public sector, users include analysts, researchers, statisticians, and other professionals with regulatory, policy, and program responsibilities for Federal, State, and local governments. The Congress and other legislative bodies may also be interested in general trends related to electricity at State and national levels. Much of the data in these reports can be used in analytic studies to evaluate new legislation. Public service commissions and other special government groups share an interest in State-level statistics. These groups can also compare the statistics for their States with those of other jurisdictions.

  14. Electric Motors

    Broader source: Energy.gov [DOE]

    Section 313 of the Energy Independence and Security Act (EISA) of 2007 raised Federal minimum efficiency standards for general-purpose, single-speed, polyphase induction motors of 1 to 500 horsepower (hp). This new standard took effect in December 2010. The new minimum efficiency levels match FEMP's performance requirement for these motors.

  15. Electric power monthly, October 1993

    SciTech Connect (OSTI)

    Not Available

    1993-10-20

    The Electric Power Monthly (EPM) presents monthly electricity statistics. The purpose of this publication is to provide energy decisionmakers with accurate and timely information that may be used in forming various perspectives on electric issues that lie ahead. Data in this report are presented for a wide audience including Congress, Federal and State agencies, the electric utility industry, and the general public. The EIA collected the information in this report to fulfill its data collection and dissemination responsibilities as specified in the Federal Energy Administration Act of 1974 (Public Law 93-275) as amended. This publication provides monthly statistics at the US, Census division, and State levels for net generation, fossil fuel consumption and stocks, quantity and quality of fossil fuels, cost of fossil fuels, electricity sales, revenue, and average revenue per kilowatthour of electricity sold. Data on net generation, fuel consumption, fuel stocks, quantity and cost of fossil fuels are also displayed for the North American Electric Reliability Council (NERC) regions. Statistics by company and plant are published in the EPM on the capability of new generating units, net generation, fuel consumption, fuel stocks, quantity and quality of fuel, and cost of fossil fuels.

  16. Electric power monthly, January 1994

    SciTech Connect (OSTI)

    Not Available

    1994-01-26

    The Electric Power Monthly (EPM) presents monthly electricity statistics. The purpose of this publication is to provide energy decisionmakers with accurate and timely information that may be used in forming various perspectives on electric issues that lie ahead. Data in this report are presented for a wide audience including Congress, Federal and State agencies, the electric utility industry, and the general public. The EIA collected the information in this report to fulfill its data collection and dissemination responsibilities as specified in the Federal Energy Administration Act of 1974 (Public Law 93-275) as amended. This publication provides monthly statistics at the US Census division, and State levels for net generation, fossil fuel consumption and stocks, quantity and quality of fossil fuels, cost of fossil fuels, electricity sales, revenue, and average revenue per kilowatthour of electricity sold. Data on net generation, fuel consumption, fuel stocks, quantity and cost of fossil fuels are also displayed for the North American Electric Reliability Council (NERC) regions. Statistics by company and plant are published in the EPM on the capability of new generating units, net generation, fuel consumption, fuel stocks, quantity and quality of fuel, and cost of fossil fuels.

  17. Electric power monthly, February 1994

    SciTech Connect (OSTI)

    Not Available

    1994-02-16

    The Electric Power Monthly (EMP) presents monthly electricity statistics. The purpose of this publication is to provide energy decisionmakers with accurate and timely information that may be used in forming various perspectives on electric issues that lie ahead. Data in this report are presented for a wide audience including Congress, Federal and State agencies, the electric utility industry, and the general public. The EIA collected the information in this report to fulfill its data collection and dissemination responsibilities as specified in the Federal Energy Administration Act of 1974 (Public Law 93-275) as amended. This publication provides monthly statistics at the US, Census division, and State levels for net generation, fossil fuel consumption and stocks, quantity and quality of fossil fuels, cost of electricity sales, revenue, and average revenue per kilowatthour of electricity sold. Data on net generation, fuel consumption, fuel stocks, quantity and cost of fossil fuels are also displayed for the North American Electric Reliability Council (NERC) regions. Statistics by company and plant are published in the EPM on the capability of new generating units, net generation, fuel consumption, fuel stocks, quantity and quality of fuel, and cost of fossil fuels.

  18. Application to Export Electric Energy OE Docket No. EA-97-D Portland...

    Office of Environmental Management (EM)

    Company Application to Export Electric Energy OE Docket No. EA-97-D Portland General Electric Company Application from PGE to export electric energy to Canada. EA-97-D PGE (CN).pdf...

  19. Aging Management Guideline for commercial nuclear power plants: Electrical switchgear. Final report

    SciTech Connect (OSTI)

    Toman, G.; Gazdzinski, R.; Schuler, K.

    1993-07-01

    This Aging Management Guideline (AMG) provides recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant electrical switchgear important to license renewal. The latent of this AMG to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner which allows personnel responsible for performance analysis and maintenance, to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.

  20. Electric power annual 1997. Volume 2

    SciTech Connect (OSTI)

    1998-10-01

    The Electric Power Annual 1997, Volume 2 contains annual summary statistics at national, regional, and state levels for the electric power industry, including information on both electric utilities and nonutility power producers. Included are data for electric utility retail sales of electricity, associated revenue, and average revenue per kilowatthour of electricity sold; financial statistics; environmental statistics; power transactions; and demand-side management. Also included are data for US nonutility power producers on installed capacity; gross generation; emissions; and supply and disposition of energy. The objective of the publication is to provide industry decisionmakers, government policymakers, analysts, and the general public with historical data that may be used in understanding US electricity markets. 15 figs., 62 tabs.

  1. Texas - PUC Substantive Rule 25.5 - Electric Service Providers...

    Open Energy Info (EERE)

    5 - Electric Service Providers-General Provisions Jump to: navigation, search OpenEI Reference LibraryAdd to library Legal Document- RegulationRegulation: Texas - PUC Substantive...

  2. Simulations of Cyclic Voltammetry for Electric Double Layers...

    Office of Scientific and Technical Information (OSTI)

    Simulations of Cyclic Voltammetry for Electric Double Layers in Asymmetric Electrolytes: A Generalized Modified PoissonNernstPlanck Model Citation Details In-Document Search...

  3. Electric power monthly, July 1994

    SciTech Connect (OSTI)

    Not Available

    1994-07-01

    The Electric Power Monthly (EPM) presents monthly electricity statistics. The purpose of this publication is to provide energy decisionmakers with accurate and timely information that may be used in forming various perspectives on electric issues that lie ahead. Data in this report are presented for a wide audience including Congress, Federal and State agencies, the electric utility industry, and the general public. The EIA collected the information in this report to fulfill its data collection and dissemination responsibilities as specified in the Federal Energy Administration Act of 1974 (Public Law 93-275) as amended. The EPM is prepared by the Survey Management Division; Office of Coal, Nuclear, Electric and Alternate Fuels, Energy Information Administration (EIA), Department of Energy. This publication provides monthly statistics at the US, Census division, and State levels for net generation, fossil fuel consumption and stocks, quantity and quality of fossil fuels, cost of fossil fuels, electricity sales, revenue, and average revenue per kilowatthour of electricity sold. Data on net generation, fuel consumption, fuel stocks, quantity and cost of fossil fuels are also displayed for the North American Electric Reliability Council (NERC) regions. Statistics by company and plant are published in the EPM on the capability of new generating units, net generation, fuel consumption, fuel stocks, quantity and quality of fuel, and cost of fossil fuels. Data on quantity, quality, and cost of fossil fuels lag data on net generation, fuel consumption, fuel stocks, electricity sales, and average revenue per kilowatthour by 1 month. This difference in reporting appears in the US, Census division, and State level tables. However, for purposes of comparison, plant-level data are presented for the earlier month.

  4. Keeping Nuclear as a Viable Option for Electric Power Generation in the Brazilian Matrix

    SciTech Connect (OSTI)

    Henning, F.

    2004-10-06

    This paper discusses all alternatives that are part of the general solution for the electric energy problem in Brazil.

  5. Projecting Electricity Demand in 2050

    SciTech Connect (OSTI)

    Hostick, Donna J.; Belzer, David B.; Hadley, Stanton W.; Markel, Tony; Marnay, Chris; Kintner-Meyer, Michael C. W.

    2014-07-01

    This paper describes the development of end-use electricity projections and load curves that were developed for the Renewable Electricity (RE) Futures Study (hereafter RE Futures), which explored the prospect of higher percentages (30% - 90%) of total electricity generation that could be supplied by renewable sources in the United States. As input to RE Futures, two projections of electricity demand were produced representing reasonable upper and lower bounds of electricity demand out to 2050. The electric sector models used in RE Futures required underlying load profiles, so RE Futures also produced load profile data in two formats: 8760 hourly data for the year 2050 for the GridView model, and in 2-year increments for 17 time slices as input to the Regional Energy Deployment System (ReEDS) model. The process for developing demand projections and load profiles involved three steps: discussion regarding the scenario approach and general assumptions, literature reviews to determine readily available data, and development of the demand curves and load profiles.

  6. Effects of Cr and Nb contents on the susceptibility of Alloy 600 type Ni-base alloys to stress-corrosion cracking in a simulated BWR environment

    SciTech Connect (OSTI)

    Akashi, Masatsune

    1995-09-01

    In order to discuss the effects of chromium and niobium contents on the susceptibility of Alloy 600 type nickel-base alloys to stress-corrosion cracking in the BWR primary coolant environment, a series of creviced bent-beam (CBB) tests were conducted in a high-temperature, high-purity water environment. Chromium, niobium, and titanium as alloying elements improved the resistivity to stress-corrosion cracking, whereas carbon enhanced the susceptibility to it. Alloy-chemistry-based correlations have been defined to predict the relative resistances of alloys to stress-corrosion cracking. A strong correlation was found, for several heats of alloys, between grain-boundary chromium depletion and the susceptibility to stress-corrosion cracking.

  7. Analysis of the OECD/NRC BWR Turbine Trip Transient Benchmark with the Coupled Thermal-Hydraulics and Neutronics Code TRAC-M/PARCS

    SciTech Connect (OSTI)

    Lee, Deokjung; Downar, Thomas J.; Ulses, Anthony; Akdeniz, Bedirhan; Ivanov, Kostadin N.

    2004-10-15

    An analysis of the Peach Bottom Unit 2 Turbine Trip 2 (TT2) experiment has been performed using the U.S. Nuclear Regulatory Commission coupled thermal-hydraulics and neutronics code TRAC-M/PARCS. The objective of the analysis was to assess the performance of TRAC-M/PARCS on a BWR transient with significance in two-phase flow and spatial variations of the neutron flux. TRAC-M/PARCS results are found to be in good agreement with measured plant data for both steady-state and transient phases of the benchmark. Additional analyses of four fictitious extreme scenarios are performed to provide a basis for code-to-code comparisons and comprehensive testing of the thermal-hydraulics/neutronics coupling. The obtained results of sensitivity studies on the effect of direct moderator heating on transient simulation indicate the importance of this modeling aspect.

  8. Estimating boiling water reactor decommissioning costs: A user`s manual for the BWR Cost Estimating Computer Program (CECP) software. Final report

    SciTech Connect (OSTI)

    Bierschbach, M.C.

    1996-06-01

    Nuclear power plant licensees are required to submit to the US Nuclear Regulatory Commission (NRC) for review their decommissioning cost estimates. This user`s manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personal computer, provides estimates for the cost of decommissioning boiling water reactor (BWR) power stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning.

  9. Estimating boiling water reactor decommissioning costs. A user`s manual for the BWR Cost Estimating Computer Program (CECP) software: Draft report for comment

    SciTech Connect (OSTI)

    Bierschbach, M.C.

    1994-12-01

    With the issuance of the Decommissioning Rule (July 27, 1988), nuclear power plant licensees are required to submit to the U.S. Regulatory Commission (NRC) for review, decommissioning plans and cost estimates. This user`s manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personal computer, provides estimates for the cost of decommissioning BWR power stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning.

  10. Electric power annual 1996. Volume 1

    SciTech Connect (OSTI)

    1997-08-01

    The Electric Power Annual presents a summary of electric power industry statistics at national, regional, and State levels. The objective of the publication is to provide industry decisionmakers, government policy-makers, analysts, and the general public with data that may be used in understanding US electricity markets. The Electric Power Annual is prepared by the Coal and Electric Data and Renewables Division; Office of Coal, Nuclear, Electric and Alternate Fuels; Energy Information Administration (EIA); US Department of Energy. Volume 1--with a focus on US electric utilities--contains final 1996 data on net generation and fossil fuel consumption, stocks, receipts, and cost; preliminary 1996 data on generating unit capability, and retail sales of electricity, associated revenue, and the average revenue per kilowatthour of electricity sold. Additionally, information on net generation from renewable energy sources and on the associated generating capability is included in Volume 1 of the EPA. Data published in the Electric Power Annual Volume 1 are compiled from three statistical forms filed monthly and two forms filed annually by electric utilities. These forms are described in detail in the Technical Notes. 5 figs., 30 tabs.