Sample records for niagara mohawk pwr

  1. Customer Response to RTP in Competitive Markets: A Study of Niagara Mohawk's Standard Offer Tariff

    E-Print Network [OSTI]

    Boisvert, Richard N.; Cappers, Peter; Goldman, Charles; Neenan, Bernie; Hopper, Nicole

    2006-01-01T23:59:59.000Z

    Niagara Mohawk’s Standard Offer Tariff * Richard N. BoisvertThis default-service commodity tariff (“SC-3A Option One”)electricity usage data, tariff history, basic customer

  2. Review of integrated resource bidding at Niagara Mohawk

    SciTech Connect (OSTI)

    Goldman, C.A.; Busch, J.F.; Kahn, E.P.; Stoft, S.S.; Cohen, S.

    1992-05-01T23:59:59.000Z

    In June 1988, the New York Public Service Commission (PSC) ordered the state's investor-owned utilities to develop competitive bidding programs that included both supply and demandside resource options. The New York State Energy Research Development Authority (NYSERDA), the New York Department of Public Service, and the Department of Energy's Integrated Resource Planning program asked Lawrence Berkeley Laboratory (LBL) to review the integrated bidding processes of two New York utilities, Niagara Mohawk and Consolidated Edison. This interim report focuses on Niagara Mohawk (NMPC). In terms of overall approach, our analysis is intended as a critical review of a large-scale experiment in competitive resource acquisition implemented by New York utilities at the direction of their state regulators. The study is not a formal impact or process evaluation. Based on priorities established jointly with project sponsors, the report focuses on selected topics: analysis of the two-stage scoring system used by NMPC, ways that the scoring system can be improved, an in-depth review of the DSM bidding component of the solicitation including surveys of DSM bidders, relationship between DSM bidding and other utility-sponsored DSM programs, and major policy issues that arise in the design and implementation of competitive resource procurements. The major findings of this report are: NMPC's solicitation elicited an impressive response from private power developers and energy service companies. In the initial ranking of bids, DSM projects were awarded significantly more points on price and environmental factors compared to supply-side bids. NMPC's scoring system gave approximately twice as much weight nominally to price as to non-price factors (850 vs. 460 points).

  3. Review of integrated resource bidding at Niagara Mohawk

    SciTech Connect (OSTI)

    Goldman, C.A.; Busch, J.F.; Kahn, E.P.; Stoft, S.S.; Cohen, S.

    1992-05-01T23:59:59.000Z

    In June 1988, the New York Public Service Commission (PSC) ordered the state`s investor-owned utilities to develop competitive bidding programs that included both supply and demandside resource options. The New York State Energy Research & Development Authority (NYSERDA), the New York Department of Public Service, and the Department of Energy`s Integrated Resource Planning program asked Lawrence Berkeley Laboratory (LBL) to review the integrated bidding processes of two New York utilities, Niagara Mohawk and Consolidated Edison. This interim report focuses on Niagara Mohawk (NMPC). In terms of overall approach, our analysis is intended as a critical review of a large-scale experiment in competitive resource acquisition implemented by New York utilities at the direction of their state regulators. The study is not a formal impact or process evaluation. Based on priorities established jointly with project sponsors, the report focuses on selected topics: analysis of the two-stage scoring system used by NMPC, ways that the scoring system can be improved, an in-depth review of the DSM bidding component of the solicitation including surveys of DSM bidders, relationship between DSM bidding and other utility-sponsored DSM programs, and major policy issues that arise in the design and implementation of competitive resource procurements. The major findings of this report are: NMPC`s solicitation elicited an impressive response from private power developers and energy service companies. In the initial ranking of bids, DSM projects were awarded significantly more points on price and environmental factors compared to supply-side bids. NMPC`s scoring system gave approximately twice as much weight nominally to price as to non-price factors (850 vs. 460 points).

  4. PP-190 Niagara Mohawk Power Corporation | Department of Energy

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

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  5. Radiological Survey Results for the Niagara Mohawk Right-of-Way, Tonawanda, New York (TNY004)

    SciTech Connect (OSTI)

    McKenzie, S.P.; Uziel, M.S.

    1998-11-01T23:59:59.000Z

    At the request of the U.S. Department of Energy (DOE), a team from Oak Ridge National Laboratory conducted a radiological survey of a small portion of the Niagara Mohawk Power Corporation right-of-way in Tonawanda, New York. The purpose of the survey was to determine if radioactive residuals had migrated from or been redistributed onto the Niagara Mohawk right-of-way from the former Linde property to the west. The Linde Air Products Division of Union Carbide Corporation, Tonawanda New York, had used radioactive materials at that location for work performed under government contract from 1942 through 1948. The survey was performed in May 1996 in response to Formerly Utilized Sites Remedial Action Program (FUSRAP) requirements. These requirements dictate that the radiological status of certain vicinity properties shall be assessed and documented according to prescribed procedures prior to certification of the property for release for unrestricted use. Such release can only be granted if the property is found to be within current applicable authorized limits. The survey included a gamma scan of accessible areas and the collection and radionuclide analysis of soil samples from the portion of right-of-way located east of the former Linde plant site and north of the railway spur entrance gate. Results of the survey indicate that radioactive material probably originating from the Linde plant is located on the Niagara Mohawk right-of-way in the area surveyed. Surface gamma exposure rates were elevated above typical background levels. Four scattered surface soil samples exceeded DOE guideline values for {sup 238}U, and 8 of 13 surface soil samples exceeded DOE guideline values for {sup 226}Ra. The radionuclide distribution in these samples was similar to that found in materials resulting from former processing activities at the Linde site. It is recommended that the property be designated for remedial action by DOE.

  6. Niagrara Mohawk Power Corporation`s energy awareness program

    SciTech Connect (OSTI)

    Gentile, P.R.; Oughterson, M.

    1995-12-01T23:59:59.000Z

    The Energy Awareness pilot program introduced the notion of energy-efficient office equipment to Niagara Mohawk and some selected customer sites. The program was designed to acquaint customers with the Energy Star Program and with energy-efficient office technologies and with promotional activities at the national, state, and regional levels. A major element is customer education to stimulate use of the information available. Another was to work with customer and vendor personnel to develop standards, procedures, etc., that would be useful for other participants in the program. The program was named the {open_quotes}Energy Watchdog Program.{close_quotes} Niagara Mohawk is working to establish itself as the preferred energy provider in the region, and in keeping with that objective the program goals are to: (1) Raise the level of customer satisfaction by providing service of value over and above that of electricity alone. (2) Measure energy savings or efficiency improvements achieved through the program to help meet DSM goals. (3) Contribute to the current education and information program. (4) Apply principles of energy-efficiency to in-house equipment and practices to develop expertise and experience. (5) Build awareness among utility and customer employees of energy-efficient office technologies and practices so that they can use the information at home as well as on the job. (6) Nurture trade ally relationships that will last after rebates are phased out.

  7. PP-190 Niagara Mohawk Power Corporation | Department of Energy

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742Energy ChinaofSchaeferAprilOverviewEfficiencyofHSSPIA - I-ManageMisuse of Position1 1 1

  8. SOXAL{trademark} pilot plant demonstration at Niagara Mohawk`s Dunkirk Station

    SciTech Connect (OSTI)

    Strangway, P.K. [Niagara Mohawk Power Corp., Syracuse, NY (United States)

    1995-12-31T23:59:59.000Z

    The Clean Air Act Amendments of 1990 made it necessary to accelerate the development of scrubber systems for use by some utilities burning sulfur-containing fuels, primarily coal. While many types of Flue Gas Desulfurization (FGD) systems operate based on lime and limestone scrubbing, these systems have drawbacks when considered for incorporation into long-term emissions control plans. Although the costs associated with disposal of large amounts of scrubber sludge may be manageable today, the trend is toward increased disposal costs. Many new SO{sub 2} control technologies are being pursued in the hope of developing an economical regenerable FGD system did recovers the SO{sub 2} as a saleable commercial product, thus minimizing the formation of disposal waste. Some new technologies include the use of exotic chemical absorbents which are alien to the utility industry and utilities` waste treatment facilities. These systems present utilities with new environmental issues. The SOXAL{trademark} process has been developed so as to eliminate such issues.

  9. Customer Response to RTP in Competitive Markets: A Study of Niagara Mohawk's Standard Offer Tariff

    E-Print Network [OSTI]

    Boisvert, Richard N.; Cappers, Peter; Goldman, Charles; Neenan, Bernie; Hopper, Nicole

    2006-01-01T23:59:59.000Z

    of Residential Time-of-Use Pricing Experiments”, Journal ofResidential Response in Time of Use Pricing Experiments. ”Across Time-of-Use Electricity Pricing Experiments. ”

  10. Customer Response to RTP in Competitive Markets: A Study of Niagara Mohawk's Standard Offer Tariff

    E-Print Network [OSTI]

    Boisvert, Richard N.; Cappers, Peter; Goldman, Charles; Neenan, Bernie; Hopper, Nicole

    2006-01-01T23:59:59.000Z

    Operator (NYISO) and New York State Energy Research andin New York. ” Final Report prepared for California Energy

  11. Mohawk Municipal Comm | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directedAnnual SiteofEvaluatingGroup |JilinLu anMicrogreen PolymersModular Energy Devices IncMohawk Municipal

  12. Final Independent External Peer Review Report Mohawk Dam Major Rehabilitation Report

    E-Print Network [OSTI]

    US Army Corps of Engineers

    and to minimize the potential for catastrophic failure of the dam during such events. Several alternatives wereFinal Independent External Peer Review Report Mohawk Dam Major Rehabilitation Report Warsaw, Ohio Report Mohawk Dam Major Rehabilitation Report Warsaw, Ohio by Battelle 505 King Avenue Columbus, OH 43201

  13. ,"Niagara Falls, NY Natural Gas Pipeline Imports From Canada...

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

    Imports From Canada (MMcf)" ,"Click worksheet name or tab at bottom for data" ,"Worksheet Name","Description"," Of Series","Frequency","Latest Data for" ,"Data 1","Niagara Falls,...

  14. Wellton-Mohawk Irr & Drain Dist | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directedAnnualProperty Edit withTianlin BaxinUmweltVillageGraphWellton-Mohawk Irr & Drain Dist Jump to:

  15. ,"Niagara Falls, NY Natural Gas Pipeline Exports to Canada (MMcf...

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

    Exports to Canada (MMcf)" ,"Click worksheet name or tab at bottom for data" ,"Worksheet Name","Description"," Of Series","Frequency","Latest Data for" ,"Data 1","Niagara Falls, NY...

  16. Volume 29, Number 1 October 2006 The McMaster University Faculty of Engineering and the Mohawk

    E-Print Network [OSTI]

    Thompson, Michael

    directly from high school. Programs will be offered in automotive and vehicle technology, and biotechnology. A 4-year process automation program is already underway from the Mohawk College campus. Both the two

  17. DOE - Office of Legacy Management -- Niagara Falls Vicinity Properties NY -

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA group currentBradleyTableSelling CorpNew MexicoUtah MexicanNiagara Falls

  18. Niagara Falls Storage Site Vicinity Properties in Lewiston, New York,

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA groupTuba City, Arizona, DisposalFourthN V O 1 8 7 +New York, New York,Niagara

  19. Assessing environmental exposure to PCBs among Mohawks at Akwesasne through the use of geostatistical methods

    SciTech Connect (OSTI)

    Hwang, S.; Fitzgerald, E.F.; Cayo, M.; Yang, B.Z. [New York State Dept. of Health, Albany, NY (United States). Bureau of Environmental and Occupational Epidemiology] [New York State Dept. of Health, Albany, NY (United States). Bureau of Environmental and Occupational Epidemiology; Tarbell, A.; Jacobs, A. [Mohawk Nation at Akwesasne (United States). Akwesasne Task Force on the Environment] [Mohawk Nation at Akwesasne (United States). Akwesasne Task Force on the Environment

    1999-02-01T23:59:59.000Z

    The Mohawk Nation at Akwesasne is a Native American community located along the St. Lawrence River in New York State, Ontario, and Quebec. One component of a multiphase human health study was to assess the impact of different pathways of human exposure resulting from the off-site migration of polychlorinated biphenyl (PCB) contamination in this area. This paper illustrates how mapped residential information and environmental sampling data can be united to assist in exposure assessment for epidemiologic studies using geographic information system (GIS) technology and statistical methods. A proportional sampling scheme was developed to collect 119 surface soils. Using a method of cross validation, the average estimated error can be computed and the best estimator can be selected. Seven spatial methods were examined to estimate surface soil PCB concentrations; the lowest relative mean error was 0.42% for Inverse 3 nearest neighbor weighted according to the inverse distance, and the highest relative mean error was 4.4% for Voronoi polygons. Residual plots indicated that all methods performed well except near some of the sampling points that formed the outer boundaries of the sampling distribution.

  20. Local fish consumption and serum PCB concentrations among Mohawk men at Akwesasne

    SciTech Connect (OSTI)

    Fitzgerald, E.F.; Deres, D.A.; Hwang, S.A.; Bush, B.; Yang, B. [New York State Dept. of Health, Albany, NY (United States)] [New York State Dept. of Health, Albany, NY (United States); Tarbell, A.; Jacobs, A. [Mohawk Nation at Akwesasne (United States). Akwesasne Task Force on the Environment] [Mohawk Nation at Akwesasne (United States). Akwesasne Task Force on the Environment

    1999-02-01T23:59:59.000Z

    A study was conducted to assess local fish consumption patterns and their relationship to concentrations of total polychlorinated biphenyls (PCBs) in the serum of Mohawk men residing near three hazardous waste sites. From 1992 to 1995, 139 men were interviewed and donated a 20-ml venous blood sample. The results indicated that the men ate a mean of 21.2 local fish meals during the past year, compared with annual means of 27.7 meals 1--2 years before and 88.6 meals more than 2 years before. This change is probably a consequence of advisories issued against the consumption of local fish, since 97% of the mean were aware of the advisories and two-third had changed their behavior as a result. Multiple regression analysis revealed that serum PCB levels increased with age and local fish consumption. The data suggest that local fish consumption has contributed to body burdens in this population and that the advisories have been effective in modifying local fish consumption habits.

  1. Niagara Falls Storage Site annual environmental report for calendar year 1991, Lewiston, New York. [Niagara Falls Storage Site

    SciTech Connect (OSTI)

    Not Available

    1992-09-01T23:59:59.000Z

    This document describes the environmental monitoring program at the Niagara Falls Storage Site (NFSS) and surrounding area, implementation of the program, and monitoring results for 1991. Environmental monitoring at NFSS began in 1981. The site is owned by the US Department of Energy (DOE) and is assigned to the DOE Formerly Utilized Sites Remedial Action Program (FUSRAP). FUSRAP is a program to decontaminate or otherwise control sites where residual radioactive materials remain from the early years of the nation's atomic energy program or from commercial operations causing conditions that Congress has authorized DOE to remedy. The environmental monitoring program at NFSS includes sampling networks for radon concentrations in air; external gamma radiation exposure; and total uranium and radium-226 concentrations in surface water, sediments, and groundwater. Additionally, several nonradiological parameters including seven metals are routinely measured in groundwater. Monitoring results are compared with applicable Environmental Protection Agency (EPA) standards, DOE derived concentration guides (DCGs), dose limits, and other requirements in DOE orders. Environmental standards are established to protect public health and the environment.

  2. American Ref-Fuel of Niagara Biomass Facility | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directedAnnualProperty EditCalifornia: Energy Resources Jump to:Almo, Idaho:ReligiousHempstead BiomassNiagara

  3. DOE - Office of Legacy Management -- Niagara Falls Storage Site NY - NY 17

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA group currentBradleyTableSelling CorpNew MexicoUtah MexicanNiagara Falls Storage

  4. BWR Fuel Assembly BWR Fuel Assembly PWR Fuel Assembly

    National Nuclear Security Administration (NNSA)

    BWR Fuel Assembly BWR Fuel Assembly PWR Fuel Assembly PWR Fuel Assembly The PWR 17x17 assembly is approximately 160 inches long (13.3 feet), 8 inches across, and weighs 1,500 lbs....

  5. EIS-0109: Long-Term Management of the Existing Radioactive Wastes and Residues at the Niagara Falls Storage Site

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy developed this statement to evaluate the environmental impacts of several alternatives for management and control of the radioactive wastes and residues at the Niagara Falls Storage Site, including a no action alternative, an alternative to manage wastes on-site, and two off-site management alternatives.

  6. Niagara Falls Storage Site environmental report for calendar year 1989, Lewiston, New York

    SciTech Connect (OSTI)

    Not Available

    1990-05-01T23:59:59.000Z

    The environmental monitoring program, which began in 1981, was continued during 1989 at the Niagara Falls Storage Site (NFSS), a United States Department of Energy (DOE) surplus facility located in Niagara County, New York, that is currently used for interim storage of radioactive residues, contaminated soils, and rubble. The monitoring program is being conducted by Bechtel National, Inc. The monitoring program at NFSS measures radon concentrations in air; external gamma radiation levels; and uranium and radium concentrations in surface water, groundwater, and sediment. Additionally, several nonradiological parameters are measured in groundwater. To verify that the site is in compliance with the DOE radiation protection standard and to assess its potential effect on public health, the radiation dose was calculated for a hypothetical maximally exposed individual. Based on the conservative scenario described in this report, this hypothetical individual receives an annual external exposure equivalent to approximately 2 percent of the DOE radiation protection standard of 100 mrem/yr. This exposure is less than a person receives during a one-way flight from New York to Los Angeles (because of the greater amounts of cosmic radiation at higher altitudes). The cumulative dose to the population within an 80-km (50-mi) radius of NFSS that results from radioactive materials present at the site is indistinguishable from the dose that the same population receives from naturally occurring radioactive sources. Results of the 1989 monitoring show that NFSS is in compliance with applicable DOE radiation protection standards. 18 refs., 26 figs., 18 tabs.

  7. Results of radiological measurements taken in the Niagara Falls, New York, area (NF002)

    SciTech Connect (OSTI)

    Williams, J.K.; Berven, B.A.

    1986-11-01T23:59:59.000Z

    The results of a radiological survey of 100 elevated gamma radiation anomalies in the Niagara Falls, New York, area are presented. These radiation anomalies were identified by a mobile gamma scanning survey during the period October 3-16, 1984, and were recommended for an onsite survey to determine if the elevated levels of radiation may be related to the transportation of radioactive waste material to the Lake Ontario Ordnance Works for storage. In this survey, radiological measurements included outdoor gamma exposure rates at 1 m above the surface; outdoor gamma exposure rates at the surface, range of gamma exposure rates during scan; and uranium, radium, and thorium concentrations in biased surface soil samples. The results show 38 anomalies (35 located along Pletcher Road and 3 associated with other unreleated locations) were found to exceed Formerly Utilized Sites Remedial Action Program (FUSRAP) remedial action guidelines and were recommended for formal characterization surveys. (Since the time of this survey, remedial actions have been conducted on the 38 anomalies identified as exceeding FUSRAP guidelines, and the radioactive material above guidelines has been removed.) The remaining 62 anomalies are associated with asphalt driveways and parking lots, which used a phosphate slag material (previously identified as cyclowollastonite, synthetic CaSiO/sub 3/). This rocky-slag waste material was used for bedding under asphalt surfaces and in general gravel applications. Most of the contaminated soil and rock samples collected at the latter anomalies had approximately equal concentrations of /sup 226/Ra and /sup 238/U and, therefore, are not related to materials connected with the Niagara Falls Storage Site (NFSS), including material that was transported to the NFSS. 13 refs., 7 figs., 14 tabs.

  8. Preliminary study on direct recycling of spent PWR fuel in PWR system

    SciTech Connect (OSTI)

    Waris, Abdul; Nuha; Novitriana; Kurniadi, Rizal; Su'ud, Zaki [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jalan Ganesa 10 Bandung 40132 (Indonesia)

    2012-06-06T23:59:59.000Z

    Preliminary study on direct recycling of PWR spent fuel to support SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario has been conducted. Several spent PWR fuel compositions in loaded PWR fuel has been evaluated to obtain the criticality of reactor. The reactor can achieve it criticality for U-235 enrichment in the loaded fresh fuel is at least 4.0 a% with the minimum fraction of the spent fuel in the core is 15.0 %. The neutron spectra become harder with the escalating of U-235 enrichment in the loaded fresh fuel as well as the amount of the spent fuel in the core.

  9. Engineering evaluation of alternatives for the disposition of Niagara Falls Storage Site, its residues and wastes

    SciTech Connect (OSTI)

    Not Available

    1984-01-01T23:59:59.000Z

    The final disposition scenarios selected by DOE for assessment in this document are consistent with those stated in the Notice of Intent to prepare an Environmental Impact Statement (EIS) for the Niagara Falls Storage Site (NFSS) (DOE, 1983d) and the modifications to the alternatives resulting from the public scoping process. The scenarios are: take no action beyond interim remedial measures other than maintenance and surveillance of the NFSS; retain and manage the NFSS as a long-term waste management facility for the wastes and residues on the site; decontaminate, certify, and release the NFSS for other use, with long-term management of the wastes and residues at other DOE sites; and partially decontaminate the NFSS by removal and transport off site of only the more radioactive residues, and upgrade containment of the remaining wastes and residues on site. The objective of this document is to present to DOE the conceptual engineering, occupational radiation exposure, construction schedule, maintenance and surveillance requirements, and cost information relevant to design and implementation of each of the four scenarios. The specific alternatives within each scenario used as the basis for discussion in this document were evaluated on the bases of engineering considerations, technical feasibility, and regulatory requirements. Selected alternatives determined to be acceptable for each of the four final disposition scenarios for the NFSS were approved by DOE to be assessed and costed in this document. These alternatives are also the subject of the EIS for the NFSS currently being prepared by Argonne National Laboratory (ANL). 40 figures, 38 tables.

  10. Niagara Falls storage site annual environmental report for calendar year 1990, Lewiston, New York

    SciTech Connect (OSTI)

    Not Available

    1991-08-01T23:59:59.000Z

    Environmental monitoring of the US DOE Niagara Falls Storage Site (NFSS) and surrounding area began in 1981. NFSS is part of a DOE program to decontaminate or otherwise control sites where residual radioactive materials remain from the early years of the nation's atomic energy program or from commercial, operations causing conditions the Congress has authorized DOE to remedy. Environmental monitoring systems at NFSS include sampling networks for radon concentrations in air; external gamma radiation exposure; and total uranium and radium-226 concentrations in surface water sediments, and groundwater. Additionally, several nonradiological parameters are routinely measured in groundwater. During 1990, the average ambient air radon concentration (including background) at NFSS ranged from 0.3 to 0.7 pCi/L (0.01 to 0.03 Bq/L); the maximum at any location for any quarter was 1.6 pCi/L (0.06 Bq/L). The average on-site external gamma radiation exposure level was 69 mR/yr; the average at the property line was 68 mR/yr (including background). The average background radiation level in the area was 66 mR/yr. Average annual concentrations of radium-226 and total uranium in surface water ranged from 0.4E-9 to 0.9E-9 {mu}Ci/m1 (0.02 to 0.03 Bq/L) and from 5E-9 to 9E-9 {mu}Ci/m1 (0.2 to 0.3 Bq/L), respectively. Routine analyses of groundwater samples from NFSS included the indicator parameters total organic carbon, total organic halides, pH, and specific conductivity.

  11. The graphs at right show overall variability distribution estimated for the Pentium D 800 series (near) and the T1 Niagara (far) using the FPGA data

    E-Print Network [OSTI]

    Renau, Jose

    RESULTS The graphs at right show overall variability distribution estimated for the Pentium D 800 where a core no longer works properly. In the Sun T1 Niagara cores this is done with a built-in- self processors we record the temperature at which the failure occurred and adjust to the frequencies

  12. Niagara Falls Storage Site, Annual site environmental report, Lewiston, New York, Calendar year 1986: Surplus Facilities Management Program (SFMP)

    SciTech Connect (OSTI)

    Not Available

    1987-06-01T23:59:59.000Z

    During 1986, the environmental monitoring program was continued at the Niagara Falls Storage Site (NFSS), a US Department of Energy (DOE) surplus facility located in Niagara County, New York, presently used for the interim storage of radioactive residues and contaminated soils and rubble. The monitoring program is being conducted by Bechtel National, Inc. The monitoring program at the NFSS measures radon gas concentrations in air; external gamma radiation levels; and uranium and radium concentrations in surface water, groundwater, and sediment. To verify that the site is in compliance with the DOE radiation protection standard and to assess its potential effect on public health, the radiation dose was calculated for the maximally exposed individual. Based on the conservative scenario described in the report, this individual would receive an annual external exposure approximately equivalent to 6% of the DOE radiation protection standard of 100 mrem/yr. By comparison, the incremental dose received from living in a brick house versus a wooden house is 10 mrem/yr above background. The cumulative dose to the population within an 80-km (50-mi) radius of the NFSS that would result from radioactive materials present at the site would be indistinguishable from the dose that the same population would receive from naturally occurring radioactive sources. Results of the 1986 monitoring show that the NFSS is in compliance with the DOE radiation protection standard. 14 refs., 11 figs., 14 tabs.

  13. applications pwr bwr: Topics by E-print Network

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

    reactivity and power distributions ... Inoue, Yuichiro, 1969- 2004-01-01 5 Ris9-R-609(EN) Simulation ofa PWR Power Plant Multidisciplinary Databases and Resources Websites Summary:...

  14. Western Minnesota Mun Pwr Agny | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directedAnnualProperty Edit withTianlin BaxinUmweltVillageGraphWellton-MohawkWestern Iowa Power Coop

  15. Ris9-R-609(EN) Simulation ofa PWR Power Plant

    E-Print Network [OSTI]

    Ris9-R-609(EN) Simulation ofa PWR Power Plant for Process Control and Diagnosis Finn Ravnsbjerg Nielsen Risø National Laboratory, Roskilde, Denmark December 1991 #12;Simulation of a PWR Power Plant *^R a compute simulation of a simplified pressurized nuclear power plant model directed towards process control

  16. Zebra: An advanced PWR lattice code

    SciTech Connect (OSTI)

    Cao, L.; Wu, H.; Zheng, Y. [School of Nuclear Science and Technology, Xi'an Jiaotong Univ., No. 28, Xianning West Road, Xi'an, ShannXi, 710049 (China)

    2012-07-01T23:59:59.000Z

    This paper presents an overview of an advanced PWR lattice code ZEBRA developed at NECP laboratory in Xi'an Jiaotong Univ.. The multi-group cross-section library is generated from the ENDF/B-VII library by NJOY and the 361-group SHEM structure is employed. The resonance calculation module is developed based on sub-group method. The transport solver is Auto-MOC code, which is a self-developed code based on the Method of Characteristic and the customization of AutoCAD software. The whole code is well organized in a modular software structure. Some numerical results during the validation of the code demonstrate that this code has a good precision and a high efficiency. (authors)

  17. Fuel cycle optimization of thorium and uranium fueled PWR systems

    E-Print Network [OSTI]

    Garel, Keith Courtnay

    1977-01-01T23:59:59.000Z

    The burnup neutronics of uniform PWR lattices are examined with respect to reduction of uranium ore requirements with an emphasis on variation of the fuel-to-moderator ratio

  18. Niagara falls storage site: Annual site environmental report, Lewiston, New York, Calendar Year 1988: Surplus Facilities Management Program (SFMP)

    SciTech Connect (OSTI)

    Not Available

    1989-04-01T23:59:59.000Z

    The monitoring program at the Niagara Falls Storage Site (NFSS) measures radon concentrations in air; external gamma radiation levels; and uranium and radium concentrations in surface water, groundwater, and sediment. To verify that the site is in compliance with the DOE radiation protection standard and to assess its potential effect on public health, the radiation dose was calculated for a hypothetical maximally exposed individual. Based on the conservative scenario described in this report, this hypothetical individual receives an annual external exposure approximately equivalent to 6 percent of the DOE radiation protection standard of 100 mrem/yr. This exposure is less than a person receives during two round-trip flights from New York to Los Angeles (because of the greater amounts of cosmic radiation at higher altitudes). The cumulative dose to the population within an 80-km (50-mi) radius of the NFSS that results from radioactive materials present at the site is indistinguishable from the dose that the same population receives from naturally occurring radioactive sources. Results of the 1988 monitoring show that the NFSS is in compliance with applicable DOE radiation protection standards. 17 refs., 31 figs., 20 tabs.

  19. Fort Drum integrated resource assessment. Volume 3, Resource assessment

    SciTech Connect (OSTI)

    Dixon, D.R.; Armstrong, P.R.; Daellenbach, K.K.; Dagle, J.E.; Di Massa, F.V.; Elliott, D.B.; Keller, J.M.; Richman, E.E.; Shankle, S.A.; Sullivan, G.P.; Wahlstrom, R.R.

    1992-12-01T23:59:59.000Z

    The US Army Forces Command (FORSCOM) has tasked Pacific Northwest Laboratory (PNL) as the lead laboratory supporting the US Department of Energy (DOE) Federal Energy Management Program`s (FEMP) mission to identify, evaluate, and assist in acquiring all cost-effective energy projects at Fort Drum. This is a model program PNL is designing for federal customers served by the Niagara Mohawk Power Company (Niagara Mohawk). It will (1) identify and evaluate all electric and fossil fuel cost-effective energy projects; (2) develop a schedule at each installation for project acquisition considering project type, size, timing, capital requirements, as well as energy and dollar savings; and (3) secure 100% of the financing required to implement electric energy efficiency projects from Niagara Mohawk and have Niagara Mohawk procure the necessary contractors to perform detailed audits and install the technologies. This report provides the results of the fossil fuel and electric energy resource opportunity (ERO) assessments performed by PNL at one of Niagara Mohawk`s primary federal facilities, the FORSCOM Fort Drum facility located near Watertown, New York. It is a companion report to Volume 1, the Executive Summary, and Volume 2, the Baseline Detail.

  20. RIS-M-2264 CONSTRUCTION OF PWR NUCLEAR CROSS SECTIONS FOR TRANSIENT

    E-Print Network [OSTI]

    RISØ-M-2264 CONSTRUCTION OF PWR NUCLEAR CROSS SECTIONS FOR TRANSIENT CALCULATIONS. TEST OF THE ANTI recent Westinghouse designs, representing two different PWR reactor cores, are calculated as functions oi; COMPUTER CALCULATIONS; COUPLING CONSTANTS; CROSS SECTIONS; POWER DISTRIBUTION; PWR TYPE REACTORS

  1. A Comparison Between Model Reduction and Controller Reduction: Application to a PWR Nuclear Planty

    E-Print Network [OSTI]

    Gevers, Michel

    A Comparison Between Model Reduction and Controller Reduction: Application to a PWR Nuclear Planty model reduction with controller reduction for the same PWR system. We show that closed-loop techniques to the design of a low-order con- troller for a realistic model of order 42 of a Pressurized Water Reactor (PWR

  2. PWR fuel performance and future trend in Japan

    SciTech Connect (OSTI)

    Kondo, Y.

    1988-01-01T23:59:59.000Z

    Since the first PWR power plant Mihama Unit 1 initiated its commercial operation in 1970, Japanese utilities and manufacturers have expended much of their resources and efforts to improve PWR technology. The results are already seen in significantly improved performance of 16 PWR plants now in operation. Mitsubishi Heavy Industries Ltd. (MHI) has been supplying them with nuclear fuel assemblies, which are over 5700. As the reliability of the current design fuel has been achieved, the direction of RandD on nuclear fuel has changed to make nuclear power more competitive to the other power generation methods. The most important RandD targets are the burnup extension, Gd contained fuel, utilization and the load follow capability.

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

    SciTech Connect (OSTI)

    Faidy, C. [EDF SEPTEN, Villeurbanne (France)

    1997-04-01T23:59:59.000Z

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

  4. Analysis of PWR RCS Injection Strategy During Severe Accident

    SciTech Connect (OSTI)

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

    2004-05-15T23:59:59.000Z

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

  5. Effects of Burnable Absorbers on PWR Spent Nuclear Fuel

    SciTech Connect (OSTI)

    P.M. O'Leary; Dr. M.L. Pitts

    2000-08-21T23:59:59.000Z

    Burnup credit is an ongoing issue in designing and licensing transportation and storage casks for spent nuclear fuel (SNF). To address this issue, in July 1999, the U.S. Nuclear Regulatory Commission (NRC), Spent Fuel Project Office, issued Interim Staff Guidance-8 (ISG-8), Revision 1 allowing limited burnup credit for pressurized water reactor (PWR) spent nuclear fuel (SNF) to be used in transport and storage casks. However, one of the key limitations for a licensing basis analysis as stipulated in ISG-8, Revision 1 is that ''burnup credit is restricted to intact fuel assemblies that have not used burnable absorbers''. Because many PWR fuel designs have incorporated burnable-absorber rods for more than twenty years, this restriction places an unnecessary burden on the commercial nuclear power industry. This paper summarizes the effects of in-reactor irradiation on the isotopic inventory of PWR fuels containing different types of integral burnable absorbers (BAs). The work presented is illustrative and intended to represent typical magnitudes of the reactivity effects from depleting PWR fuel with different types of burnable absorbers.

  6. Fort Drum integrated resource assessment

    SciTech Connect (OSTI)

    Dixon, D.R.; Armstrong, P.R.; Daellenbach, K.K.; Dagle, J.E.; Di Massa, F.V.; Elliott, D.B.; Keller, J.M.; Richman, E.E.; Shankle, S.A.; Sullivan, G.P.; Wahlstrom, R.R.

    1992-12-01T23:59:59.000Z

    The US Army Forces Command (FORSCOM) has tasked Pacific Northwest Laboratory (PNL) as the lead laboratory supporting the US Department of Energy (DOE) Federal Energy Management Program's (FEMP) mission to identify, evaluate, and assist in acquiring all cost-effective energy projects at Fort Drum. This is a model program PNL is designing for federal customers served by the Niagara Mohawk Power Company (Niagara Mohawk). It will (1) identify and evaluate all electric and fossil fuel cost-effective energy projects; (2) develop a schedule at each installation for project acquisition considering project type, size, timing, capital requirements, as well as energy and dollar savings; and (3) secure 100% of the financing required to implement electric energy efficiency projects from Niagara Mohawk and have Niagara Mohawk procure the necessary contractors to perform detailed audits and install the technologies. This report provides the results of the fossil fuel and electric energy resource opportunity (ERO) assessments performed by PNL at one of Niagara Mohawk's primary federal facilities, the FORSCOM Fort Drum facility located near Watertown, New York. It is a companion report to Volume 1, the Executive Summary, and Volume 2, the Baseline Detail.

  7. Condensate polishing guidelines for PWR and BWR plants

    SciTech Connect (OSTI)

    Robbins, P.; Crinigan, P.; Graham, B.; Kohlmann, R.; Crosby, C.; Seager, J.; Bosold, R.; Gillen, J.; Kristensen, J.; McKeen, A.; Jones, V.; Sawochka, S.; Siegwarth, D.; Keeling, D.; Polidoroff, T.; Morgan, D.; Rickertsen, D.; Dyson, A.; Mills, W.; Coleman, L.

    1993-03-01T23:59:59.000Z

    Under EPRI sponsorship, an industry committee, similar in form and operation to other guideline committees, was created to develop Condensate Polishing Guidelines for both PWR and BWR systems. The committee reviewed the available utility and water treatment industry experience on system design and performance and incorporated operational and state-of-the-art information into document. These guidelines help utilities to optimize present condensate polisher designs as well as be a resource for retrofits or new construction. These guidelines present information that has not previously been presented in any consensus industry document. The committee generated guidelines that cover both deep bed and powdered resin systems as an integral part of the chemistry of PWR and BWR plants. The guidelines are separated into sections that deal with the basis for condensate polishing, system design, resin design and application, data management and performance and management responsibilities.

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

    SciTech Connect (OSTI)

    Subkhi, Moh. Nurul; Su'ud, Zaki; Waris, Abdul [Physics.Dept., Bandung Institute of Technology.Ganesha 10, Bandung (Indonesia)

    2012-06-06T23:59:59.000Z

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

  9. NIAGARA FALLS STORAGE SITE

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA groupTuba City, Arizona, DisposalFourthN V O 1 8 7 + PROJECTpi/L +3(YJ 4:i"

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

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directedAnnual Siteof EnergyInnovation inOpenadd: China DatangCentral ElCentral Montana E Pwr Coop Inc

  11. Northwest Rural Pub Pwr Dist | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directedAnnual SiteofEvaluatingGroup |JilinLuOpenNorth AmericaNorthwest Rural Pub Pwr Dist Jump to:

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

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directedAnnual SiteofEvaluatingGroup |JilinLuOpenNorthOlympiaAnalysis)Pearl River Valley El Pwr Assn Jump to:

  13. Arizona Electric Pwr Coop Inc | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directedAnnualProperty EditCalifornia: Energy Resources JumpAnaconda,AnzaArcade,theElectric Pwr Coop Inc Jump

  14. Cuming County Public Pwr Dist | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are beingZealand JumpConceptual Model, clickInformationNew|CoreCpWing County,ElectricCuming County Public Pwr Dist

  15. Niagara Falls Storage Site, Lewiston, New York: Annual site environmental report, Calendar year 1987: Formerly Utilized Sites Remedial Action Program (FUSRAP)

    SciTech Connect (OSTI)

    Not Available

    1988-04-01T23:59:59.000Z

    The monitoring program at the Niagara Falls Storage Site (NFSS) measures radon gas concentrations in air; external gamma radiation levels; and uranium and radium concentrations in surface water, groundwater, and sediment. To verify that the site is in compliance with the DOE radiation protection standard and to assess its potential effect on public health, the radiation dose was calculated for the maximally exposed individual. Based on the conservative scenario described in the report, this individual would receive an annual external exposure approximately equivalent to 6 percent of the DOE radiation protection standard of 100 mrem/yr. By comparison, the incremental dose received from living in a brick house versus a wooden house is 10 mrem/yr above background. The cumulative dose to the population within an 80-km (50-mi) radius of the NFSS that would result from radioactive materials present at the site would be indistinguishable from the dose that the same population would receive from naturally occurring radioactive sources. Results of the 1987 monitoring show that the NFSS is in compliance with the DOE radiation protection standard. 13 refs., 10 figs., 20 tabs.

  16. Niagara Falls Storage Site environmental report for calendar year 1992, 1397 Pletcher Road, Lewiston, New York. Formerly Utilized Sites Remedial Action Program (FUSRAP)

    SciTech Connect (OSTI)

    Not Available

    1993-05-01T23:59:59.000Z

    This report describes the environmental surveillance program at the Niagara Falls Storage Site (NFSS) and provides the results for 1992. From 1944 to the present, the primary use of NFSS has been storage of radioactive residues produced as a by-product of uranium production. All onsite areas of residual radioactivity above guidelines have been remediated. Materials generated during remediation are stored onsite in the 4-ha (10-acre) waste containment structure (WCS). The WCS is a clay-lined, clay-capped, and grass-covered storage pile. The environmental surveillance program at NFSS includes sampling networks for radon concentrations in air; external gamma radiation exposure; and total uranium and radium-226 concentrations in surface water, sediments, and groundwater. Several chemical parameters, including seven metals, are also routinely measured in groundwater. This surveillance program assists in fulfilling the DOE policy of measuring and monitoring effluents from DOE activities and calculating hypothetical doses. Monitoring results are compared with applicable Environmental Protection Agency (EPA) and New York State Department of Environmental Conservation (NYSDEC) standards, DOE derived concentration guides (DCGs), dose limits, and other DOE requirements. Results of environmental monitoring during 1992 indicate that levels of the parameters measured were in compliance with all but one requirement: Concentrations of iron and manganese in groundwater were above NYSDEC groundwater quality standards. However, these elements occur naturally in the soils and groundwater associated with this region. In 1992 there were no environmental occurrences or reportable quantity releases.

  17. Recent advances in analysis of PWR containment bypass accidents

    SciTech Connect (OSTI)

    Warman, E.A.; Metcalf, J.E.; Donahue, M.L. (Stone and Webster Engineering Corp., Boston, MA (United States))

    1991-01-01T23:59:59.000Z

    The Reactor Safety Study identified and quantified the contribution to off-site radiological risks of accident sequences at pressurized water reactors (PWRs) in which the release of fission products may be released by bypassing the containment building. These so-called bypass accidents were also referred to as interfacing systems loss-of-coolant accidents (LOCAs) or Event 5 sequences due to the postulated failure of valves separating the high-pressure reactor coolant system (RCS) from low-pressure piping located outside containment. Containment bypass sequence risks constitute a large fraction of the total pressurized water reactor (PWR) in NUREG-1150 in large part because estimates of competing risks from early containment failures have been greatly reduced since WASH-1400. Rigorous analyses of both SGTR and V sequence bypass sequences result in reductions in fission product release to such an extent that in-containment sequences are expected to dominate PWR risks at levels substantially lower than reported in NUREG-1150. It is important that these findings be confirmed by other investigators, particularly in light of the NRC's ongoing study of the frequency of occurrence of interfacing systems. LOCAs based on extensive investigations at operating plants. Progress in this latter effort should be matched by progress in the knowledge and understanding of the progression of bypass sequences, once initiated.

  18. Gamma and Neutron Radiolysis in the 21-PWR Waste Package

    SciTech Connect (OSTI)

    J.S. Tang

    2001-05-03T23:59:59.000Z

    The objective of this calculation is to compute gamma and neutron dose rates in order to determine the maximum radiolytic production of nitric acid and other chemical species inside the 21-PWR (pressurized-water reactor) waste package (WP). The scope of this calculation is limited to the time period between 5,000 and 100,000 years after emplacement. The information provided by the sketches attached to this calculation is that of the potential design for the type of WP considered in this calculation. The results of this calculation will be used to evaluate nitric acid corrosion of fuel cladding from radiolysis in the 21-PWR WP. This calculation was performed in accordance with the Technical Work Plan for: Waste Package Design Description for LA (Civilian Radioactive Waste Management System (CRWMS) Management and Operating Contractor (M&O) 2000a). AP-3.124, Calculations, is used to perform the calculation and develop the document. This calculation is associated with the total system performance assessment (TSPA) of which the spent fuel cladding integrity is to be evaluated.

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

    SciTech Connect (OSTI)

    Tanaka, T. [Kansai Electric Power Company, Osaka (Japan); Shimizu, S.; Ogata, Y. [Mitsubishi Heavy Industries, Ltd., Kobe (Japan)

    1997-04-01T23:59:59.000Z

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

  20. SENSITIVITY STUDIES FOR THE PWR ROD EJECTION ACCIDENT

    SciTech Connect (OSTI)

    DIAMOND,D.J.; YANG,C.Y.; ARONSON,A.L.

    1999-11-14T23:59:59.000Z

    The objective of this study was to understand the uncertainty in fuel enthalpy calculated for the rod ejection accident (REA) in a pressurized water reactor (PWR). This is to help the US Nuclear Regulatory Commission in making judgments about acceptance criteria for the REA when high burnup fuel is used and for assessing the validity of licensee methods for calculating the REA. The approach is twofold. Sensitivity studies were first done to determine the effect on calculated fuel enthalpy of uncertainties in the important parameters which determine the outcome of the REA. The second step, which will be carried out at a later date, is to use the sensitivity to estimate the random error in the fuel enthalpy due to random errors in these key parameters once the variance of these parameters is determined.

  1. Griffiss AFB integrated resource assessment. Volume 2, Electric baseline detail

    SciTech Connect (OSTI)

    Dixon, D.R.; Armstrong, P.R.; Keller, J.M.

    1993-02-01T23:59:59.000Z

    The US Air Force Air Combat Command has tasked the Pacific Northwest Laboratory (PNL) as the lead laboratory supporting the US Department of Energy (DOE) Federal Energy Management Program`s (FEMP) mission to identify, evaluate, and assist in acquiring all cost-effective energy projects at Griffiss Air Force Base (AFB). This is a model program PNL is designing for federal customers served by the Niagara Mohawk Power Company (Niagara Mohawk). It will (1) identify and evaluate all electric cost-effective energy projects; (2) develop a schedule at each installation for project acquisition considering project type, size, timing, and capital requirements, as well as energy and dollar savings; and (3) secure 100% of the financing required to implement electric energy efficiency projects from Niagara Mohawk and have Niagara Mohawk procure the necessary contractors to perform detailed audits and install the technologies. This report documents the assessment of baseline energy use at one of Niagara Mohawk`s primary federal facilities, Griffiss AFB, an Air Combat Command facility located near Rome, New York. It is a companion report to Volume 1, the Executive Summary, and Volume 3, the Electric Resource Assessment. The analysis examines the characteristics of electric, gas, oil, propane, coal, and purchased thermal capacity use for fiscal year (FY) 1990. The results include energy-use intensities for the facilities at Griffiss AFB by building type and electric energy end use. A complete electric energy consumption reconciliation is presented that accounts for the distribution of all major electric energy uses and losses among buildings, utilities, and central systems.

  2. WAPD-SC-545 HYDROGEN FLAMMABILITY DATA AND APPLICATION TO PWR

    Office of Scientific and Technical Information (OSTI)

    WAPD-SC-545 HYDROGEN FLAMMABILITY DATA AND APPLICATION TO PWR LOSS-OF-COOLANT ACCIDENT CONTRACT A T - I M - G E N - H BETTIS PLANT PITTSBURGH, PENNSYLVANIA Operated for the U.S....

  3. A comparison of fuzzy logic-PID control strategies for PWR pressurizer control

    SciTech Connect (OSTI)

    Kavaklioglu, K.; Ikonomopoulos, A. (Univ. of Tennessee, Knoxville (United States))

    1993-01-01T23:59:59.000Z

    This paper describes the results obtained from a comparison performed between classical proportional-integral-derivative (PID) and fuzzy logic (FL) controlling the pressure in a pressurized water reactor (PWR). The two methodologies have been tested under various transient scenarios, and their performances are evaluated with respect to robustness and on-time response to external stimuli. One of the main concerns in the safe operation of PWR is the pressure control in the primary side of the system. In order to maintain the pressure in a PWR at the desired level, the pressurizer component equipped with sprayers, heaters, and safety relief valves is used. The control strategy in a Westinghouse PWR is implemented with a PID controller that initiates either the electric heaters or the sprayers, depending on the direction of the coolant pressure deviation from the setpoint.

  4. Evaluation of Final Radiological Conditions at Areas of the Niagara Falls Storage Site Remediated under the Formerly Utilized Sites Remedial Action Program -12184

    SciTech Connect (OSTI)

    Clayton, Christopher [U.S Department of Energy Office of Legacy Management, Washington, DC; Kothari, Vijendra [U.S Department of Energy Office of Legacy Management, Morgantown, West Virginia; Starr, Ken [U.S Department of Energy Office of Legacy Management, Westminster, Colorado; Widdop, Michael; Gillespie, Joey [SM Stoller Corporation, Grand Junction, Colorado

    2012-02-26T23:59:59.000Z

    The U. S. Department of Energy (DOE) methods and protocols allow evaluation of remediation and final site conditions to determine if remediated sites remain protective. Two case studies are presented that involve the Niagara Falls Storage Site (NFSS) and associated vicinity properties (VPs), which are being remediated under the Formerly Utilized Sites Remedial Action Program (FUSRAP). These properties are a part of the former Lake Ontario Ordnance Works (LOOW). In response to stakeholders concerns about whether certain remediated NFSS VPs were putting them at risk, DOE met with stakeholders and agreed to evaluate protectiveness. Documentation in the DOE records collection adequately described assessed and final radiological conditions at the completed VPs. All FUSRAP wastes at the completed sites were cleaned up to meet DOE guidelines for unrestricted use. DOE compiled the results of the investigation in a report that was released for public comment. In conducting the review of site conditions, DOE found that stakeholders were also concerned about waste from the Separations Process Research Unit (SPRU) at the Knolls Atomic Power Laboratory (KAPL) that was handled at LOOW. DOE agreed to determine if SPRU waste remained at that needed to be remediated. DOE reviewed records of waste characterization, historical handling locations and methods, and assessment and remediation data. DOE concluded that the SPRU waste was remediated on the LOOW to levels that pose no unacceptable risk and allow unrestricted use and unlimited exposure. This work confirms the following points as tenets of an effective long-term surveillance and maintenance (LTS&M) program: ? Stakeholder interaction must be open and transparent, and DOE must respond promptly to stakeholder concerns. ? DOE, as the long-term custodian, must collect and preserve site records in order to demonstrate that remediated sites pose no unacceptable risk. ? DOE must continue to maintain constructive relationships with the U.S. Army Corps of Engineers and state and federal regulators.

  5. Bias identification in PWR pressurizer instrumentation using the generalized liklihood-ratio technique

    SciTech Connect (OSTI)

    Tylee, J.L.

    1981-01-01T23:59:59.000Z

    A method for detecting and identifying biases in the pressure and level sensors of a pressurized water reactor (PWR) pressurizer is described. The generalized likelihood ratio (GLR) technique performs statistical tests on the innovations sequence of a Kalman filter state estimator and is capable of determining when a bias appears, in what sensor the bias exists, and estimating the bias magnitude. Simulation results using a second-order linear, discrete PWR pressurizer model demonstrate the capabilities of the GLR method.

  6. Effect of component aging on PWR control rod drive systems

    SciTech Connect (OSTI)

    Grove, E.; Gunther, W.; Sullivan, K.

    1992-01-01T23:59:59.000Z

    An aging assessment of PWR control rod drive (CRD) systems has been completed as part of the US NRC Nuclear Plant Aging Research (NPAR) Program. The design, construction, maintenance, and operation of the Babcock Wilcox (B W), Combustion Engineering (CE), and Westinghouse (W) systems were evaluated to determine the potential for degradation as each system ages. Operating experience data were evaluated to identify the predominant failure modes, causes, and effects. This, coupled with an assessment of the materials of construction and operating environment, demonstrate that each design is subject to degradation, which if left unchecked, could affect its safety function as the plant ages. An industry survey, conducted with the assistance of EPRI and NUMARC, identified current CRD system maintenance and inspection practices. The results of this survey indicate that some plants have performed system modifications, replaced components, or augmented existing preventive maintenance practices in response to system aging. The survey results also supported the operating experience data, which concluded that the timely replacement of degraded components, prior to failure, was not always possible using existing condition monitoring techniques. The recommendations presented in this study also include a discussion of more advanced monitoring techniques, which provide trendable results capable of detecting aging.

  7. Effect of component aging on PWR control rod drive systems

    SciTech Connect (OSTI)

    Grove, E.; Gunther, W.; Sullivan, K.

    1992-06-01T23:59:59.000Z

    An aging assessment of PWR control rod drive (CRD) systems has been completed as part of the US NRC Nuclear Plant Aging Research (NPAR) Program. The design, construction, maintenance, and operation of the Babcock & Wilcox (B & W), Combustion Engineering (CE), and Westinghouse (W) systems were evaluated to determine the potential for degradation as each system ages. Operating experience data were evaluated to identify the predominant failure modes, causes, and effects. This, coupled with an assessment of the materials of construction and operating environment, demonstrate that each design is subject to degradation, which if left unchecked, could affect its safety function as the plant ages. An industry survey, conducted with the assistance of EPRI and NUMARC, identified current CRD system maintenance and inspection practices. The results of this survey indicate that some plants have performed system modifications, replaced components, or augmented existing preventive maintenance practices in response to system aging. The survey results also supported the operating experience data, which concluded that the timely replacement of degraded components, prior to failure, was not always possible using existing condition monitoring techniques. The recommendations presented in this study also include a discussion of more advanced monitoring techniques, which provide trendable results capable of detecting aging.

  8. Automatic load follow control system for PWR plants

    SciTech Connect (OSTI)

    Nakakura, H.; Ishiguro, A.

    1987-01-01T23:59:59.000Z

    In Japan, load follow operation (daily load follow, automatic frequency control (AFC) operation, and governor free (GF) operation) of nuclear plants will be required in the near future to control grid frequency, as the ratio of nuclear plant electrical production to total grid production will increase. The AFC operation regulated power by demand from the central load dispatcher to control mainly the fringe component of the grid frequency fluctuation, and GF operation regulates power by turbine revolution or grid frequency to control mainly the cyclic component of grid frequency fluctuation. This paper deals with the automatic power distribution control system, which is important to load follow operation and possibly will be applied to pressurized water reactor (PWR) nuclear plants. The reactor control systems noted below are conventional design with some improvements for AFC/GF operation, so that the reactor operates the turbine as before: (1) rod control system (reactor power control); (2) pressurizer pressure control system; (3) pressurizer level control system; and (4) steam generator level control system.

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

    SciTech Connect (OSTI)

    Waris, A.; Permana, S.; Kurniadi, R.; Su'ud, Z. [Bosscha Laboratory, Department of Physics, Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Sekimoto, H. [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology (Japan)

    2010-06-22T23:59:59.000Z

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

  10. Griffiss AFB integrated resource assessment

    SciTech Connect (OSTI)

    Dixon, D.R.; Armstrong, P.R.; Keller, J.M.

    1993-02-01T23:59:59.000Z

    The US Air Force Air Combat Command has tasked the Pacific Northwest Laboratory (PNL) as the lead laboratory supporting the US Department of Energy (DOE) Federal Energy Management Program's (FEMP) mission to identify, evaluate, and assist in acquiring all cost-effective energy projects at Griffiss Air Force Base (AFB). This is a model program PNL is designing for federal customers served by the Niagara Mohawk Power Company (Niagara Mohawk). It will (1) identify and evaluate all electric cost-effective energy projects; (2) develop a schedule at each installation for project acquisition considering project type, size, timing, and capital requirements, as well as energy and dollar savings; and (3) secure 100% of the financing required to implement electric energy efficiency projects from Niagara Mohawk and have Niagara Mohawk procure the necessary contractors to perform detailed audits and install the technologies. This report documents the assessment of baseline energy use at one of Niagara Mohawk's primary federal facilities, Griffiss AFB, an Air Combat Command facility located near Rome, New York. It is a companion report to Volume 1, the Executive Summary, and Volume 3, the Electric Resource Assessment. The analysis examines the characteristics of electric, gas, oil, propane, coal, and purchased thermal capacity use for fiscal year (FY) 1990. The results include energy-use intensities for the facilities at Griffiss AFB by building type and electric energy end use. A complete electric energy consumption reconciliation is presented that accounts for the distribution of all major electric energy uses and losses among buildings, utilities, and central systems.

  11. WG-MOX Fuel Zr-tube Neutron Spectrum Comparison in ATR and PWR

    SciTech Connect (OSTI)

    Gray S. Chang

    2005-02-01T23:59:59.000Z

    An experiment containing WG-MOX fuel has been designed and irradiated from 1998 to 2004 in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). Important neutronics parameters were computed using novel Monte Carlo methods. The purpose of this summary is to compare the Weapons-Grade Mixed Oxide fuel (WG-MOX) Zr-tube’s neutron spectrum in ATR and PWR. The results indicate that the Zrtube’s neutron spectrum in ATR are softer than in PWR.

  12. Categorization of PWR accident sequences and guidelines for fault trees: seismic initiators

    SciTech Connect (OSTI)

    Kimura, C.Y.

    1984-09-01T23:59:59.000Z

    This study developed a set of dominant accident sequences that could be applied generically to domestic commercial PWRs as a standardized basis for a probabilistic seismic risk assessment. This was accomplished by ranking the Zion 1 accident sequences. The pertinent PWR safety systems were compared on a plant-by-plant basis to determine the applicability of the dominant accident sequences of Zion 1 to other PWR plants. The functional event trees were developed to describe the system functions that must work or not work in order for a certain accident sequence to happen, one for pipe breaks and one for transients.

  13. Ris-M-2209 THE THREE-DIMENSIONAL PWR TRANSIENT CODE

    E-Print Network [OSTI]

    , REACTOR KINETICS, ROD DROP ACCIDENTS, THREE- DIMENSIONAL CALCULATIONS, TRANSIENTS. UDC 621 more or less by change. The calculation is there- fore not representative of any existing reactorRisø-M-2209 THE THREE-DIMENSIONAL PWR TRANSIENT CODE ANTI; ROD EJECTION TEST CALCULATION A

  14. Westinghouse Approach and Experience on Operating VVER (PWR)1000 I and C Modernization

    SciTech Connect (OSTI)

    Mahlab Moshe [Kozloduy Project Director, Westinghouse Electric Company (Bulgaria); Naydenov, Nayden [Kozloduy NPP Modernization Manager (Bulgaria); Sechensky, Boyan [Chief Engineer, Westinghouse Energy Systems Bulgaria (Bulgaria)

    2004-07-01T23:59:59.000Z

    The paper will describe the background, current implementation approach and experience on the largest ever modernization program on operating units VVER 1000 (PWR) at Kozloduy Nuclear Power Plant in Bulgaria. The Modernization Program itself includes more than 212 measures. Westinghouse is modernizing the major I and C Systems at VVER 1000. (authors)

  15. Reactor analysis support package (RASP). Volume 7. PWR set-point methodology. Final report

    SciTech Connect (OSTI)

    Temple, S.M.; Robbins, T.R.

    1986-09-01T23:59:59.000Z

    This report provides an overview of the basis and methodology requirements for determining Pressurized Water Reactor (PWR) technical specifications related setpoints and focuses on development of the methodology for a reload core. Additionally, the report documents the implementation and typical methods of analysis used by PWR vendors during the 1970's to develop Protection System Trip Limits (or Limiting Safety System Settings) and Limiting Conditions for Operation. The descriptions of the typical setpoint methodologies are provided for Nuclear Steam Supply Systems as designed and supplied by Babcock and Wilcox, Combustion Engineering, and Westinghouse. The description of the methods of analysis includes the discussion of the computer codes used in the setpoint methodology. Next, the report addresses the treatment of calculational and measurement uncertainties based on the extent to which such information was available for each of the three types of PWR. Finally, the major features of the setpoint methodologies are compared, and the principal effects of each particular methodology on plant operation are summarized for each of the three types of PWR.

  16. Crack growth rates of nickel alloy welds in a PWR environment.

    SciTech Connect (OSTI)

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

    2006-05-31T23:59:59.000Z

    In light water reactors (LWRs), vessel internal components made of nickel-base alloys are susceptible to environmentally assisted cracking. A better understanding of the causes and mechanisms of this cracking may permit less conservative estimates of damage accumulation and requirements on inspection intervals. A program is being conducted at Argonne National Laboratory to evaluate the resistance of Ni alloys and their welds to environmentally assisted cracking in simulated LWR coolant environments. This report presents crack growth rate (CGR) results for Alloy 182 shielded-metal-arc weld metal in a simulated pressurized water reactor (PWR) environment at 320 C. Crack growth tests were conducted on 1-T compact tension specimens with different weld orientations from both double-J and deep-groove welds. The results indicate little or no environmental enhancement of fatigue CGRs of Alloy 182 weld metal in the PWR environment. The CGRs of Alloy 182 in the PWR environment are a factor of {approx}5 higher than those of Alloy 600 in air under the same loading conditions. The stress corrosion cracking for the Alloy 182 weld is close to the average behavior of Alloy 600 in the PWR environment. The weld orientation was found to have a profound effect on the magnitude of crack growth: cracking was found to propagate faster along the dendrites than across them. The existing CGR data for Ni-alloy weld metals have been compiled and evaluated to establish the effects of key material, loading, and environmental parameters on CGRs in PWR environments. The results from the present study are compared with the existing CGR data for Ni-alloy welds to determine the relative susceptibility of the specific Ni-alloy weld to environmentally enhanced cracking.

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

    SciTech Connect (OSTI)

    Fei, T.; Hoffman, E.A.; Kim, T.K.; Taiwo, T.A. [Nuclear Engineering Division Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL (United States)

    2013-07-01T23:59:59.000Z

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

  18. TITAN code development for application to a PWR steam line break accident : final report 1983-1984

    E-Print Network [OSTI]

    Tsai, Chon-Kwo

    1984-01-01T23:59:59.000Z

    Modification of the TITAN computer code which enables it to be applied to a PWR steam line break accident has been accomplished. The code now has the capability of simulating an asymmetric inlet coolant temperature transient ...

  19. Transient thermal analysis of PWR’s by a single-pass procedure using a simplified nodal layout

    E-Print Network [OSTI]

    Liu, Jack S. H.

    1979-01-01T23:59:59.000Z

    PWR accident conditions and analysis methods have been reviewed. Limitations of the simplified method with respect to analysis of these accident conditions are drawn and two transients ( loss of coolant flow, seized rotor) ...

  20. Fort Drum integrated resource assessment. Volume 2, Baseline detail

    SciTech Connect (OSTI)

    Dixon, D.R.; Armstrong, P.R.; Brodrick, J.R.; Daellenbach, K.K.; Di Massa, F.V.; Keller, J.M.; Richman, E.E.; Sullivan, G.P.; Wahlstrom, R.R.

    1992-12-01T23:59:59.000Z

    The US Army Forces Command (FORSCOM) has tasked the Pacific Northwest Laboratory (PNL) as the lead laboratory supporting the US Department of Energy (DOE) Federal Energy Management Program`s mission to identify, evaluate, and assist in acquiring all cost-effective energy projects at Fort Drum. This is a model program PNL is designing for federal customers served by the Niagara Mohawk Power Company. It will identify and evaluate all electric and fossil fuel cost-effective energy projects; develop a schedule at each installation for project acquisition considering project type, size, timing, and capital requirements, as well as energy and dollar savings; and secure 100% of the financing required to implement electric energy efficiency projects from Niagara Mohawk and have Niagara Mohawk procure the necessary contractors to perform detailed audits and install the technologies. This report documents the assessment of baseline energy use at one of Niagara Mohawk`s primary federal facilities, the FORSCOM Fort Drum facility located near Watertown, New York. It is a companion report to Volume 1, the Executive Summary, and Volume 3, the Resource Assessment. This analysis examines the characteristics of electric, gas, oil, propane, coal, and purchased thermal capacity use for fiscal year (FY) 1990. It records energy-use intensities for the facilities at Fort Drum by building type and energy end use. It also breaks down building energy consumption by fuel type, energy end use, and building type. A complete energy consumption reconciliation is presented that includes the accounting of all energy use among buildings, utilities, central systems, and applicable losses.

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

    SciTech Connect (OSTI)

    Subkhi, M. Nurul [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology (Ganesha 10 Bandung, Indonesia) and Physics Dept., Faculty of Science and Technology, State Islamic University of Sunan Gunung (Indonesia); Su'ud, Zaki; Waris, Abdul; Permana, Sidik [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology (Ganesha 10 Bandung) (Indonesia)

    2014-09-30T23:59:59.000Z

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

  2. Safety analysis of B and W Standard PWR using thorium-based fuels

    SciTech Connect (OSTI)

    Uotinen, V.O.; Carroll, W.P.; Jones, H.M.; Toops, E.C.

    1980-06-01T23:59:59.000Z

    A study was performed to assess the safety and licenseability of the Babcock and Wilcox standard 205-fuel assembly PWR when it is fueled with three types of thoria-based fuels denatured (/sup 233/U//sup 238/U-Th)O/sub 2/, denatured (/sup 235//U/sup 238/U-Th)O/sub 2/, and (Th-Pu)O/sub 2/. Selected transients were analyzed using typical PWR safety analysis calculational methods. The results support the conclusion that it is feasible from a safety standpoint to utilize either of the denatured urania-thoria fuels in the standard B and W plant. In addition, it appears that the use of thoria-plutonia fuels would probably also be feasible. These tentative conclusions depend on a data that is more limited than that available for UO/sub 2/ fuels.

  3. An Empirical Approach to Bounding the Axial Reactivity Effects of PWR Spent Nuclear Fuel

    SciTech Connect (OSTI)

    P. M. O'Leary; J. M. Scaglione

    2001-04-04T23:59:59.000Z

    One of the significant issues yet to be resolved for using burnup credit (BUC) for spent nuclear fuel (SNF) is establishing a set of depletion parameters that produce an adequately conservative representation of the fuel's isotopic inventory. Depletion parameters (such as local power, fuel temperature, moderator temperature, burnable poison rod history, and soluble boron concentration) affect the isotopic inventory of fuel that is depleted in a pressurized water reactor (PWR). However, obtaining the detailed operating histories needed to model all PWR fuel assemblies to which BUC would be applied is an onerous and costly task. Simplifications therefore have been suggested that could lead to using ''bounding'' depletion parameters that could be broadly applied to different fuel assemblies. This paper presents a method for determining a set of bounding depletion parameters for use in criticality analyses for SNF.

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

    SciTech Connect (OSTI)

    Cardoni, Jeffrey

    2010-11-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Swamy, S.A.; Bhowmick, D.C. [Westinghouse Nuclear Technology Division, Pittsburgh, PA (United States)

    1997-04-01T23:59:59.000Z

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

  6. Nuclear data uncertainties by the PWR MOX/UO{sub 2} core rod ejection benchmark

    SciTech Connect (OSTI)

    Pasichnyk, I.; Klein, M.; Velkov, K.; Zwermann, W.; Pautz, A. [Boltzmannstr. 14, D-85748 Garching b. Muenchen (Germany)

    2012-07-01T23:59:59.000Z

    Rod ejection transient of the OECD/NEA and U.S. NRC PWR MOX/UO{sub 2} core benchmark is considered under the influence of nuclear data uncertainties. Using the GRS uncertainty and sensitivity software package XSUSA the propagation of the uncertainties in nuclear data up to the transient calculations are considered. A statistically representative set of transient calculations is analyzed and both integral as well as local output quantities are compared with the benchmark results of different participants. It is shown that the uncertainties in nuclear data play a crucial role in the interpretation of the results of the simulation. (authors)

  7. MELCOR analyses of severe accident scenarios in Oconee, a B W PWR plant

    SciTech Connect (OSTI)

    Madni, I.K.; Nimnual, S. (Brookhaven National Lab., Upton, NY (United States)); Foulds, R. (Nuclear Regulatory Commission, Washington, DC (United States))

    1993-01-01T23:59:59.000Z

    This paper presents the results and insights gained from MELCOR analyses of two severe accident scenarios, a Loss of Coolant Accident (LOCA) and a Station Blackout (TMLB) in Oconee, a Babcock Wilcox (B W) designed PWR with a large dry containment, and comparisons with Source Term Code Package (STCP) calculations of the same sequences. Results include predicted timing of key events, thermal-hydraulic response in the reactor coolant system and containment, and environmental releases of fission products. The paper also explores the impact of varying concrete type, vessel failure temperature, and break location on the accident progression, containment pressurization, and environmental releases of radionuclides.

  8. MELCOR analyses of severe accident scenarios in Oconee, a B&W PWR plant

    SciTech Connect (OSTI)

    Madni, I.K.; Nimnual, S. [Brookhaven National Lab., Upton, NY (United States); Foulds, R. [Nuclear Regulatory Commission, Washington, DC (United States)

    1993-03-01T23:59:59.000Z

    This paper presents the results and insights gained from MELCOR analyses of two severe accident scenarios, a Loss of Coolant Accident (LOCA) and a Station Blackout (TMLB) in Oconee, a Babcock & Wilcox (B&W) designed PWR with a large dry containment, and comparisons with Source Term Code Package (STCP) calculations of the same sequences. Results include predicted timing of key events, thermal-hydraulic response in the reactor coolant system and containment, and environmental releases of fission products. The paper also explores the impact of varying concrete type, vessel failure temperature, and break location on the accident progression, containment pressurization, and environmental releases of radionuclides.

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

    SciTech Connect (OSTI)

    Shaver, Mark W.; Lanning, Donald D.

    2010-02-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2009-05-01T23:59:59.000Z

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

  11. Development of a new lattice physics code robin for PWR application

    SciTech Connect (OSTI)

    Zhang, S.; Chen, G. [Shanghai NuStar Nuclear Power Technology Co., Ltd., 81 South Qinzhou Road, Shanghai 200235 (China)

    2013-07-01T23:59:59.000Z

    This paper presents a description of methodologies and preliminary verification results of a new lattice physics code ROBIN, being developed for PWR application at Shanghai NuStar Nuclear Power Technology Co., Ltd. The methods used in ROBIN to fulfill various tasks of lattice physics analysis are an integration of historical methods and new methods that came into being very recently. Not only these methods like equivalence theory for resonance treatment and method of characteristics for neutron transport calculation are adopted, as they are applied in many of today's production-level LWR lattice codes, but also very useful new methods like the enhanced neutron current method for Dancoff correction in large and complicated geometry and the log linear rate constant power depletion method for Gd-bearing fuel are implemented in the code. A small sample of verification results are provided to illustrate the type of accuracy achievable using ROBIN. It is demonstrated that ROBIN is capable of satisfying most of the needs for PWR lattice analysis and has the potential to become a production quality code in the future. (authors)

  12. Bi-content Gadolinia as Burnable Absorber in PWR to Improve the Reactor Core Behaviour

    SciTech Connect (OSTI)

    Zheng, S. [AREVA, AREVA NP Fuel Sector, 10, Rue Juliette Recamier 69456 Lyon cedex (France)

    2007-07-01T23:59:59.000Z

    The gadolinia product is one of the standard burnable absorbers used in the PWR long and low leakage fuel cycle in order to control the radial power distribution and to hold down the initial core reactivity. This product presents a large number of advantages such as the high efficiency with only a small number of gadolinia-bearing rods, the easy adjustment between the number and the content of the gadolinia-bearing rods according to the cycle length need and the initial reactivity hold-down, no increasing of boron concentration versus cycle depletion, no additional increasing of internal pressure in poisoned rods, very low additional manufacture cost. On the other hand, some unfavourable phenomena are also observed during the utilization of the gadolinia: amplification of the asymmetrical power distribution and more negative axial offset. Based on the correlation between the gadolinia burnout and its content, the use of gadolinia bi-content will improve the parameters indicated here above. The gadolinia bi-content have been used in BWR for more than 20 years. In this paper, the comparison of the main reactor core physical parameters in PWR, calculated with the AREVA NP standard neutronic code package SCIENCE, is made by using the mono- and bi-content of the gadolinia products in the same fuel assembly. The results show that the asymmetrical axial and azimuthal power distribution can be improved in the case of the bi-content gadolinia product. (authors)

  13. Investigation of optimal reactor control for a load-following PWR

    SciTech Connect (OSTI)

    Yim, M.S.

    1987-01-01T23:59:59.000Z

    Characteristics of optimal load-follow control of PWR plants are investigated in this study. A simple system model that describes main features of physical processes in the system was developed. The system model includes core neutronics with all the spatial dependent feedback effects, Xe-I dynamics, core thermal balances, primary-loop thermal balances, and steam-generator dynamic responses to turbine load changes. An optimal control problem that describes power-level control and power-distribution control problem together and considers all the important system operation limits as hard inquality constraints was formulated. The full-length control rod bank positions, part-length control rod positions, and boron concentration changes were modeled as control variables and turbine load variations were used as the forcing variable. Because modern PWR operating policy is to leave the part-length rods uninserted, the part-length rods were not used as a control variable in the optimal control calculations. The optimal control problem was converted to unconstrained nonlinear optimization problem by using the discretization approximation and the penalty function technique. The converted problem was solved by the nonlinear Gauss-Newton method which showed superior performance over all of the other tested optimization methods.

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

    SciTech Connect (OSTI)

    Santamarina, A.; Bernard, D.; Blaise, P.; Leconte, P.; Palau, J. M.; Roque, B.; Vaglio, C.; Vidal, J. F. [Commissariat a l'Energie Atomique et aux Energies Alternatives, CEA, DEN, DER, SPRC, Cadarache, F-13108 Saint-Paul-Lez-Durance (France)

    2013-07-01T23:59:59.000Z

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

  15. Non-Linear Dynamics Analysis of a PWR with Up-to-date Fuel Design

    SciTech Connect (OSTI)

    Riverola Gurruchaga, Javier [ENUSA Industrias Avanzadas S.A., Santiago Rusinol 12, 28040 Madrid (Spain)

    2007-07-01T23:59:59.000Z

    The Lyapunov stability theorems are applied to a simplified system of non-linear differential equations representative of a current 3 loop /12 feet contemporary PWR (Generation II) with up-to-date 17x17 lattice fuel design. The one-speed non-linear point kinetics model with six delayed neutron groups and lumped parameter heat transfer equations in the fuel rod and coolant along with a reactivity function with Doppler and moderator feedback effects is considered. First, local asymptotic stability is demonstrated at a variety of equilibrium state-points ranging from start-up to 150% nominal power. Then, a Lyapunov V function is found with the mathematical condition for sign definiteness and the stability region of attraction around the equilibrium HFP state is obtained. This study is complemented with the application of the Welton criterion for non linear kinetics and linear feedback in the frequency domain. As expected and consistently with Reactor Physics theory and experience, the strong asymptotic stable trend of a PWR is confirmed again for all analyzed conditions. This method is general and adaptable to other fuel assembly designs and reactor types. (authors)

  16. Fort Drum integrated resource assessment

    SciTech Connect (OSTI)

    Dixon, D.R.; Armstrong, P.R.; Brodrick, J.R.; Daellenbach, K.K.; Di Massa, F.V.; Keller, J.M.; Richman, E.E.; Sullivan, G.P.; Wahlstrom, R.R.

    1992-12-01T23:59:59.000Z

    The US Army Forces Command (FORSCOM) has tasked the Pacific Northwest Laboratory (PNL) as the lead laboratory supporting the US Department of Energy (DOE) Federal Energy Management Program's mission to identify, evaluate, and assist in acquiring all cost-effective energy projects at Fort Drum. This is a model program PNL is designing for federal customers served by the Niagara Mohawk Power Company. It will identify and evaluate all electric and fossil fuel cost-effective energy projects; develop a schedule at each installation for project acquisition considering project type, size, timing, and capital requirements, as well as energy and dollar savings; and secure 100% of the financing required to implement electric energy efficiency projects from Niagara Mohawk and have Niagara Mohawk procure the necessary contractors to perform detailed audits and install the technologies. This report documents the assessment of baseline energy use at one of Niagara Mohawk's primary federal facilities, the FORSCOM Fort Drum facility located near Watertown, New York. It is a companion report to Volume 1, the Executive Summary, and Volume 3, the Resource Assessment. This analysis examines the characteristics of electric, gas, oil, propane, coal, and purchased thermal capacity use for fiscal year (FY) 1990. It records energy-use intensities for the facilities at Fort Drum by building type and energy end use. It also breaks down building energy consumption by fuel type, energy end use, and building type. A complete energy consumption reconciliation is presented that includes the accounting of all energy use among buildings, utilities, central systems, and applicable losses.

  17. Calculation of releases of radioactive materials in gaseous and liquid effluents from pressurized water reactors (PWR-GALE Code). Revision 1

    SciTech Connect (OSTI)

    Chandrasekaran, T.; Lee, J.Y.; Willis, C.A.

    1985-04-01T23:59:59.000Z

    This report revises the original issuance of NUREG-0017, ''Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE-Code)'' (April 1976), to incorporate more recent operating data now available as well as the results of a number of in-plant measurement programs at operating pressurized water reactors. The PWR-GALE Code is a computerized mathematical model for calculating the releases of radioactive material in gaseous and liquid effluents (i.e., the gaseous and liquid source terms). The US Nuclear Regulatory Commission uses the PWR-GALE Code to determine conformance with the requirements of Appendix I to 10 CFR Part 50.

  18. Seismic qualification of equipment by means of probabilistic risk assessment. [PWR

    SciTech Connect (OSTI)

    Azarm, M.A.; Farahzad, P.; Boccio, J.L.

    1982-01-01T23:59:59.000Z

    Upon the sponsorship of the Equipment Qualification Branch (EQB) of NRC, Brookhaven National Laboratory (BNL) has utilized a risk-based approach for identifying, in a generic fashion, seismically risk-sensitive equipment. It is anticipated that the conclusions drawn therefrom and the methodology employed will, in part, reconcile some of the concerns dealing with the seismic qualification of equipment in operating plants. The approach taken augments an existing sensitivity analysis, based upon the WASH-1400 Reactor Safety Study (RSS), by accounting for seismicity and component fragility with the Kennedy model and by essentially including the requisite seismic data presented in the Zion Probabilistic Safety Study (ZPSS). Parametrically adjusting the seismic-related variables and ascertaining their effects on overall plant risk, core-melt probability, accident sequence probability, etc., allows one to identify those seismically risk-sensitive systems and equipment. This paper describes the approach taken and highlights the results obtained thus far for a hypothetical pressurized water reactor (PWR).

  19. Risk analysis of highly combustible gas storage, supply, and distribution systems in PWR plants

    SciTech Connect (OSTI)

    Simion, G.P. [Science Applications International Corp., Albuquerque, NM (United States); VanHorn, R.L.; Smith, C.L.; Bickel, J.H.; Sattison, M.B. [EG and G Idaho, Inc., Idaho Falls, ID (United States); Bulmahn, K.D. [SCIENTECH, Inc., Idaho Falls, ID (United States)

    1993-06-01T23:59:59.000Z

    This report presents the evaluation of the potential safety concerns for pressurized water reactors (PWRs) identified in Generic Safety Issue 106, Piping and the Use of Highly Combustible Gases in Vital Areas. A Westinghouse four-loop PWR plant was analyzed for the risk due to the use of combustible gases (predominantly hydrogen) within the plant. The analysis evaluated an actual hydrogen distribution configuration and conducted several sensitivity studies to determine the potential variability among PWRs. The sensitivity studies were based on hydrogen and safety-related equipment configurations observed at other PWRs within the United States. Several options for improving the hydrogen distribution system design were identified and evaluated for their effect on risk and core damage frequency. A cost/benefit analysis was performed to determine whether alternatives considered were justifiable based on the safety improvement and economics of each possible improvement.

  20. The key to superior water chemistry at a PWR nuclear station

    SciTech Connect (OSTI)

    Dolan, R.; Miller, L.K.; Olejar, L.L.; Salem, E.

    1983-01-01T23:59:59.000Z

    This paper demonstrates how a condensate polishing unit can be successfully used to treat the feedwater for circulating-type pressurized water reactors (PWRs). Water chemistry at the Salem Generating Station, a two-unit, four-loop Westinghouse PWR located in New Jersey, is discussed. Topics considered include a plant description and the history of early operation, the role of constant surveillance, makeup water quality, the effect of freezing on gel-type anion exchange resin, a total organic carbon (TOC) survey, steam generator chemistry, steam generator inspection, condensate polisher operation, and management philosophy. The SEPREX condensate polishing process, in which the complete separation of the anion exchange resin from the cation exchange resin is achieved by flotation separation, is examined. It is concluded that the utilization of a condensate polishing process such as SEPREX provides the operating personnel at the plant with the necessary means to maintain the minimum desired level of contaminants within the steam generator.

  1. Hydrazine usage for corrosion control in PWR plants with powdered-resin-condensate polishers. Final report

    SciTech Connect (OSTI)

    Barkich, J.L.; Battaglia, P.J.

    1983-03-01T23:59:59.000Z

    The objective of this project was to obtain the data necessary to determine the optimum injection point and amount of hydrazine to be used for oxygen control in PWR units with condensate polishing demineralizers. An additional objective was to demonstrate that the condensate polisher can be used as a means to reduce the oxygen concentration in the condensate when a sufficient excess concentration of hydrazine (over the oxygen) concentration is carried over with the main steam into the condensers. Testing was performed at North Anna Unit 2 which employs powdered resin-type condensate polishing equipment. Testing was also scheduled to be performed at Ginna and Sequoyah Unit 1 plants; however, because of an unscheduled shutdown at Ginna and inability to control hydrazine dosage at Sequoyah, these tests were not performed.

  2. Quantitative uncertainty and sensitivity analysis of a PWR control rod ejection accident

    SciTech Connect (OSTI)

    Pasichnyk, I.; Perin, Y.; Velkov, K. [Gesellschaft flier Anlagen- und Reaktorsicherheit - GRS mbH, Boltzmannstasse 14, 85748 Garching bei Muenchen (Germany)

    2013-07-01T23:59:59.000Z

    The paper describes the results of the quantitative Uncertainty and Sensitivity (U/S) Analysis of a Rod Ejection Accident (REA) which is simulated by the coupled system code ATHLET-QUABOX/CUBBOX applying the GRS tool for U/S analysis SUSA/XSUSA. For the present study, a UOX/MOX mixed core loading based on a generic PWR is modeled. A control rod ejection is calculated for two reactor states: Hot Zero Power (HZP) and 30% of nominal power. The worst cases for the rod ejection are determined by steady-state neutronic simulations taking into account the maximum reactivity insertion in the system and the power peaking factor. For the U/S analysis 378 uncertain parameters are identified and quantified (thermal-hydraulic initial and boundary conditions, input parameters and variations of the two-group cross sections). Results for uncertainty and sensitivity analysis are presented for safety important global and local parameters. (authors)

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

    SciTech Connect (OSTI)

    Gilles Youinou; Andrea Alfonsi

    2012-03-01T23:59:59.000Z

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

  4. Effect of aging on the PWR Chemical and Volume Control System

    SciTech Connect (OSTI)

    Grove, E.J.; Travis, R.J.; Aggarwal, S.K. [Brookhaven National Lab., Upton, NY (United States)

    1995-06-01T23:59:59.000Z

    The PWR Chemical and Volume Control System (CVCS) is designed to provide both safety and non-safety related functions. During normal plant operation it is used to control reactor coolant chemistry, and letdown and charging flow. In many plants, the charging pumps also provide high pressure injection, emergency boration, and RCP seal injection in emergency situations. This study examines the design, materials, maintenance, operation and actual degradation experiences of the system and main sub-components to assess the potential for age degradation. A detailed review of the Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Report (LER) databases for the 1988--1991 time period, together with a review of industry and NRC experience and research, indicate that age-related degradations and failures have occurred. These failures had significant effects on plant operation, including reactivity excursions, and pressurizer level transients. The majority of these component failures resulted in leakage of reactor coolant outside the containment. A representative plant of each PWR design (W, CE, and B and W) was visited to obtain specific information on system inspection, surveillance, monitoring, and inspection practices. The results of these visits indicate that adequate system maintenance and inspection is being performed. In some instances, the frequencies of inspection were increase in response to repeated failure events. A parametric study was performed to assess the effect of system aging on Core Damage Frequency (CDF). This study showed that as motor-operated valve (MOV) operating failures increased, the contribution of the High Pressure Injection to CDF also increased.

  5. Initiation stress threshold irradiation assisted stress corrosion cracking criterion assessment for core internals in PWR environment

    SciTech Connect (OSTI)

    Tanguy, Benoit; Stern, Anthony; Bossis, Philippe [CEA, DEN-DMN, Gif-sur-Yvette, (France); Pokor, Cedric [EDF les Renardieres, Moret-sur-Loing, (France)

    2012-07-01T23:59:59.000Z

    Irradiation assisted stress corrosion cracking (IASCC) is a problem of growing importance in pressurized water reactors (PWR). An understanding of the mechanism(s) of IASCC is required in order to provide guidance for the development of mitigation strategies. One of the principal reasons why the IASCC mechanism(s) has been so difficult to understand is the inseparability of the different IASCC potential contributors evolutions due to neutron irradiation. The potential contributors to IASCC in PWR primary water are: (i) radiation induced segregation (RIS) at grain boundaries, (ii) radiation induced microstructure (formation and growth of dislocations loops, voids, bubbles, phases), (iii) localized deformation under loading, (iv) irradiation creep and transmutations. While the development of some of the contributors (RIS, microstructure) with increasing doses are at least qualitatively well understood, the role of these changes on IASCC remains unclear. Parallel to fundamental understanding developments relative to IASCC, well controlled laboratory tests on neutron irradiated stainless steels are needed to assess the main mechanisms and also to establish an engineering criterion relative to the initiation of fracture due to IASCC. First part of this study describes the methodology carried out at CEA in order to provide more experimental data from constant load tests dedicated to the study of initiation of SCC on neutron irradiated stainless steel. A description of the autoclave recirculation loop dedicated to SCC tests on neutron irradiated materials is then given. This autoclave recirculation loop has been started on July 2010 with the first SCC test on an irradiated stainless steel (grade 316) performed at CEA. The main steps of the interrupted SCC tests are then described. Second part of this paper reports the partial results of the first test performed on a highly neutron irradiated material. (authors)

  6. Sensitivity of risk parameters to component unavailability in reactor safety study (PSAP/PSAB computer codes). [PWR; BWR

    SciTech Connect (OSTI)

    Azarm, M.; Farahzad, P.; Tingle, A.

    1982-06-01T23:59:59.000Z

    The Probabilistic Sensitivity Analysis for Pressurized and Boiling Water Reactors (PSAP/PSAB) codes have been developed to update the WASH-1400 conclusions. The initial effort, reported in NUREG/CR-1879 Sensitivity of Risk Parameters to Human Errors in Reactor Safety Study for a PWR, concentrated on developing a code for system sensitivity to human errors based on an expanded version of the PWR fault trees from the Reactor Safety Study (RSS). The success of that effort and the insights gained from the code's use initiated the development of the PSAP/PSAB codes. The two codes allow the user to evaluate the impact of new data and system models on the conclusions drawn in WASH-1400. They are designed to be fast running and modular so that detailed sensitivity studies can be run efficiently.

  7. akwesasne mohawk youth: Topics by E-print Network

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

    Dissertations Summary: ??This thesis explores the subjectivities available to young people experiencing homelessness in contemporary modern societies such as Australia. Youth...

  8. akwesasne mohawk young: Topics by E-print Network

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

    Consumer Vulnerability in the Web Physics Websites Summary: Vulnerability in the Web 2.0 Society Abstract The young consumers constitute one of the fastest growing Internet:...

  9. MHK Projects/Mohawk MHK Project | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are beingZealand Jump to: navigation, searchOf KilaueaInformationCygnet <| OpenMarisol Peru

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

    SciTech Connect (OSTI)

    J.W. Davis

    1996-08-29T23:59:59.000Z

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

  11. Development and Application of Laser Peening System for PWR Power Plants

    SciTech Connect (OSTI)

    Masaki Yoda; Itaru Chida; Satoshi Okada; Makoto Ochiai; Yuji Sano; Naruhiko Mukai; Gaku Komotori; Ryoichi Saeki [Toshiba Corporation (Japan); Toshimitsu Takagi; Masanori Sugihara; Hirokata Yoriki [Shikoku Electric Co., Inc. (Japan)

    2006-07-01T23:59:59.000Z

    Laser peening is a process to improve residual stress from tensile to compressive in surface layer of materials by irradiating high-power laser pulses on the material in water. Toshiba has developed a laser peening system composed of Q-switched Nd:YAG laser oscillators, laser delivery equipment and underwater remote handling equipment. We have applied the system for Japanese operating BWR power plants as a preventive maintenance measure for stress corrosion cracking (SCC) on reactor internals like core shrouds or control rod drive (CRD) penetrations since 1999. As for PWRs, alloy 600 or 182 can be susceptible to primary water stress corrosion cracking (PWSCC), and some cracks or leakages caused by the PWSCC have been discovered on penetrations of reactor vessel heads (RVHs), reactor bottom-mounted instrumentation (BMI) nozzles, and others. Taking measures to meet the unconformity of the RVH penetrations, RVHs themselves have been replaced in many PWRs. On the other hand, it's too time-consuming and expensive to replace BMI nozzles, therefore, any other convenient and less expensive measures are required instead of the replacement. In Toshiba, we carried out various tests for laser-peened nickel base alloys and confirmed the effectiveness of laser peening as a preventive maintenance measure for PWSCC. We have developed a laser peening system for PWRs as well after the one for BWRs, and applied it for BMI nozzles, core deluge line nozzles and primary water inlet nozzles of Ikata Unit 1 and 2 of Shikoku Electric Power Company since 2004, which are Japanese operating PWR power plants. In this system, laser oscillators and control devices were packed into two containers placed on the operating floor inside the reactor containment vessel. Laser pulses were delivered through twin optical fibers and irradiated on two portions in parallel to reduce operation time. For BMI nozzles, we developed a tiny irradiation head for small tubes and we peened the inner surface around J-groove welds after laser ultrasonic testing (LUT) as the remote inspection, and we peened the outer surface and the weld for Ikata Unit 2 supplementary. For core deluge line nozzles and primary water inlet nozzles, we peened the inner surface of the dissimilar metal welding, which is of nickel base alloy, joining a safe end and a low alloy metal nozzle. In this paper, the development and the actual application of the laser peening system for PWR power plants will be described. (authors)

  12. Development of a coupling code for PWR reactor cavity radiation streaming calculation

    SciTech Connect (OSTI)

    Zheng, Z.; Wu, H.; Cao, L.; Zheng, Y.; Zhang, H.; Wang, M. [NECP Laboratory, School of Nuclear Science and Technology, Xi'an Jiaotong Univ., Xi'an Shaanxi 710049 (China)

    2012-07-01T23:59:59.000Z

    PWR reactor cavity radiation streaming is important for the safe of the personnel and equipment, thus calculation has to be performed to evaluate the neutron flux distribution around the reactor. For this calculation, the deterministic codes have difficulties in fine geometrical modeling and need huge computer resource; and the Monte Carlo codes require very long sampling time to obtain results with acceptable precision. Therefore, a coupling method has been developed to eliminate the two problems mentioned above in each code. In this study, we develop a coupling code named DORT2MCNP to link the Sn code DORT and Monte Carlo code MCNP. DORT2MCNP is used to produce a combined surface source containing top, bottom and side surface simultaneously. Because SDEF card is unsuitable for the combined surface source, we modify the SOURCE subroutine of MCNP and compile MCNP for this application. Numerical results demonstrate the correctness of the coupling code DORT2MCNP and show reasonable agreement between the coupling method and the other two codes (DORT and MCNP). (authors)

  13. Aging mechanisms in the Westinghouse PWR (Pressurized Water Reactor) Control Rod Drive system

    SciTech Connect (OSTI)

    Gunther, W.; Sullivan, K.

    1991-01-01T23:59:59.000Z

    An aging assessment of the Westinghouse Pressurized Water Reactor (PWR) Control Rod System (CRD) has been completed as part of the US NRC's Nuclear Plant Aging Research, (NPAR) Program. This study examined the design, construction, maintenance, and operation of the system to determine its potential for degradation as the plant ages. Selected results from this study are presented in this paper. The operating experience data were evaluated to identify the predominant failure modes, causes, and effects. From our evaluation of the data, coupled with an assessment of the materials of construction and the operating environment, we conclude that the Westinghouse CRD system is subject to degradation which, if unchecked, could affect its safety function as a plant ages. Ways to detect and mitigate the effects of aging are included in this paper. The current maintenance for the control rod drive system at fifteen Westinghouse PWRs was obtained through a survey conducted in cooperation with EPRI and NUMARC. The results of the survey indicate that some plants have modified the system, replaced components, or expanded preventive maintenance. Several of these activities have effectively addressed the aging issue. 2 refs., 2 figs., 2 tabs.

  14. Modern Fuel Cladding in Demanding Operation - ZIRLO in Full Life High Lithium PWR Coolant

    SciTech Connect (OSTI)

    Kargol, Kenneth [Pacific Gas and Electric Company, Diablo Canyon Power Plant, Avila Beach, California (United States); Stevens, Jim [TXU Power, Comanche Peak Steam Electric Station, Glen Rose, Texas (United States); Bosma, John [Westinghouse Electric Company, Dallas, Texas (United States); Iyer, Jayashri; Wikmark, Gunnar [Westinghouse Electric Company, Columbia, South Carolina (United States)

    2007-07-01T23:59:59.000Z

    There is an increasing demand to optimize the PWR water chemistry in order to minimize activity build-up in the plants and to avoid CIPS and other fuel related issues. Operation with a constant pH between 7.2 and 7.4 is generally considered an important part in achieving the optimized water chemistry. The extended long cycles currently used in most of the U.S. PWRs implies that the lithium concentration at BOC will be outside the general operating experience with such a coolant chemistry regime. With the purpose to extend the experience of high lithium coolant operation, such water chemistry has been used in a few PWRs, i.e. CPSES Unit 2 and Diablo Canyon Units 1 and 2, all with ZIRLO{sup TM} cladding. Operation with a lithium concentration up to 4.2 ppm does not show any impact of the elevated lithium, while operation with up to 6 ppm possibly produce some limited corrosion acceleration in the region of sub-nucleate boiling but has no detrimental impact under the conditions limited by current operating experience. (authors)

  15. Assessment of molten debris freezing in a severe RIA in-pile test. [PWR; BWR

    SciTech Connect (OSTI)

    El-Genk, M.S.; Moore, R.L.

    1980-01-01T23:59:59.000Z

    An understanding of the freezing of molten debris on cold core structures following a hypothetical core meltdown accident in a light water reactor (LWR) is of importance to reactor safety analysis. The purpose of the present investigation was to analyze the transient freezing of the molten debris produced in a severe reactivity initiated accident (RIA) scoping test, designated RIA-ST-4, which was performed in the Power Burst Facility and simulated a BWR control rod drop accident. In the RIA-ST-4 experiment, a single, unirradiated, 20 wt % enriched, UO/sub 2/ fuel rod contained within a Zircaloy flow shroud was subjected to a single power burst which deposited a total energy of about 700 cal/g UO/sub 2/. This energy deposition is well above what is possible in a commercial LWR during a hypothetical control rod drop (BWR) or ejection (PWR) accident. However, the performance of such an in-pile test has provided important information regarding molten debris movement, relocation, and freezing on cold walls.

  16. Head Loss Evaluation in a PWR Reactor Vessel Using CFD Analysis

    SciTech Connect (OSTI)

    Ji Hwan Jeong; Jong Pil Park [School of Mechanical Engineering, Pusan National University, Enesys Jangjeon-dong, Geumjeong-gu, Busan (Korea, Republic of); Byoung-Sub Han [Jangdae-dong, Yusong-gu, Daejeon (Korea, Republic of)

    2006-07-01T23:59:59.000Z

    Nuclear vendors and utilities perform lots of simulations and analyses in order to ensure the safe operation of nuclear power plants (NPPs). In general, the simulations are carried out using vendor-specific design codes and best-estimate system analysis codes and most of them were developed based on 1-dimensional lumped parameter models. During the past decade, however, computers, parallel computation methods, and 3-dimensional computational fluid dynamics (CFD) codes have been dramatically enhanced. It is believed to be beneficial to take advantage of advanced commercial CFD codes in safety analysis and design of NPPs. The present work aims to analyze the flow distribution in downcomer and lower plenum of Korean standard nuclear power plants (KSNPs) using STAR-CD. The lower plenum geometry of a PWR is very complicated since there are so many reactor internals, which hinders in CFD analysis for real reactor geometry up to now. The present work takes advantage of 3D CAD model so that real geometry of lower plenum is used. The results give a clear figure about flow fields in the reactor vessel, which is one of major safety concerns. The calculated pressure drop across downcomer and lower plenum appears to be in good agreement with the data in engineering calculation note. A algorithm which can evaluate head loss coefficient which is necessary for thermal-hydraulic system code running was suggested based on this CFD analysis results. (authors)

  17. ASTM standards associated with PWR and BWR power plant licensing, operation and surveillance

    SciTech Connect (OSTI)

    McElroy, W.N. [Consultants and Technology Services, Richland, WA (United States); McElroy, R.J. [AEA Reactor Services, Harwell (United Kingdom); Gold, R. [Metrology Control Corp., Richland, WA (United States); Lippincott, E.P. [Westinghouse Electric Corp., Pittsburgh, PA (United States); Lowe, A.L. Jr. [BW Nuclear Technologies, Lynchburg, VA (United States)

    1994-12-31T23:59:59.000Z

    This paper considers ASTM Standards that are available, under revision, and are being considered in support of Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) Nuclear Power Plant (NPP) licensing, regulation, operation, surveillance and life attainment. The current activities of ASTM Committee E10 and its Subcommittees E10.02 and current activities of ASTM Committee E10 and its Subcommittees E10.02 and E10.05 and their Task Groups (TG) are described. A very important aspect of these efforts is the preparation, revision, and balloting of standards identified in the ASTM E706 Standard on Master Matrix for Light Water Reactor (LWR) Pressure Vessel (PV) Surveillance Standards. The current version (E706-87) of the Master Matrix identifies 21 ASTM LWR physics-dosimetry-metallurgy standards for Reactor Pressure Vessel (RPV) and Support Structure (SS) surveillance programs, whereas, for the next revision 34 standards are identified. The need for national and international coordination of Standards Technology Development, Transfer and Training (STDTT) is considered in this and other Symposium papers that address specific standards related physics-dosimetry-metallurgy issues. 69 refs.

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

    SciTech Connect (OSTI)

    Swamy, S.A.; Bhowmick, D.C.; Prager, D.E. [Westinghouse Nuclear Technology Division, Pittsburgh, PA (United States)

    1997-04-01T23:59:59.000Z

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

  19. Analysis of a rod withdrawal in a PWR core with the neutronic- thermalhydraulic coupled code RELAP/PARCS and RELAP/VALKIN

    SciTech Connect (OSTI)

    Miro, R.; Maggini, F.; Barrachina, T.; Verdu, G. [Departamento de Ingenieria Quimica y Nuclear, Universidad Politecnica de Valencia, Camino de Vera, 14, 46022, Valencia (Spain); Gomez, A.; Ortego, A. [IBERINCO, Avenida de Burgos, Madrid (Spain); Murillo, J. C. [CNAT, Av. Manoteras, Madrid (Spain)

    2006-07-01T23:59:59.000Z

    The Reactor Ejection Accident (REA) belongs to the Reactor Initiated Accidents (RIA) category of accidents and it is part of the licensing basis accident analyses required for pressure water reactors (PWR). The REA at hot zero power (HZP) is characterized by a single rod ejection from a core position with a very low power level. The evolution consists basically of a continuous reactivity insertion. The main feature limiting the consequences of the accident in a PWR is the Doppler Effect. To check the performance of the coupled code RELAP5/PARCS2.5 and RELAP5/VALKIN a REA in Trillo NPP is simulated. These analyses will allow knowing more accurately the PWR real plant phenomenology in the RIA most limiting conditions. (authors)

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

    SciTech Connect (OSTI)

    Travis, R.J.; Davis, R.E.; Grove, E.J.; Azarm, M.A. [Brookhaven National Lab., Upton, NY (United States)

    1997-08-01T23:59:59.000Z

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

  1. Testing and analyses of the TN-24P PWR spent-fuel dry storage cask loaded with consolidated fuel

    SciTech Connect (OSTI)

    McKinnon, M A; Michener, T E; Jensen, M F; Rodman, G R

    1989-02-01T23:59:59.000Z

    A performance test of a Transnuclear, Inc. TN-24P storage cask configured for pressurized water reactor (PWR) spent fuel was performed. The work was performed by the Pacific Northwest Laboratory (PNL) and Idaho National Engineering Laboratory (INEL) for the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) and the Electric Power Research Institute. The performance test consisted of loading the TN-24P cask with 24 canisters of consolidated PWR spent fuel from Virginia Power's Surry and Florida Power and Light's Turkey Point reactors. Cask surface and fuel canister guide tube temperatures were measured, as were cask surface gamma and neutron dose rates. Testing was performed with vacuum, nitrogen, and helium backfill environments in both vertical and horizontal cask orientations. Transnuclear, Inc., arranged to have a partially insulated run added to the end of the test to simulate impact limiters. Limited spent fuel integrity data were also obtained. From both heat transfer and shielding perspectives, the TN-24P cask with minor refinements can be effectively implemented at reactor sites and central storage facilities for safe storage of unconsolidated and consolidated spent fuel. 35 refs., 93 figs., 17 tabs.

  2. TOPICAL REPORT ON ACTINIDE-ONLY BURNUP CREDIT FOR PWR SPENT NUCLEAR FUEL PACKAGES

    SciTech Connect (OSTI)

    DOE

    1997-04-01T23:59:59.000Z

    A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package loading criteria and confirm proper assembly selection prior to loading. A measurement of the average assembly burnup is required and that measurement must be within 10% of the utility burnup record for the assembly to be accepted. The measurement device must be accurate to within 10%. Each step is described in detail for use with any computer code system and is then demonstrated with the SCALE 4.2 computer code package using 27BURNUPLIB cross sections.

  3. Thomas, J.R. and Clem, A.W, 1991, PWR moderator temperature coefficient via noise analysis: time series methods, Proceedings of SMORNVI, Gatlinburg, 34.01

    E-Print Network [OSTI]

    Pázsit, Imre

    and Testing Symposium, Knoxville, Tennes­ see 52.01 Uhrig, R.E., 1990, Use of artificial intelligence/Computer Interactions: Nuclear and Beyond, Nash­ ville, Tennessee, 210 Uhrig R.E., 1991, Potential application of neural­ 36 ­ Thomas, J.R. and Clem, A.W, 1991, PWR moderator temperature coefficient via noise analysis

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

    SciTech Connect (OSTI)

    Bi, G.; Liu, C.; Si, S. [Shanghai Nuclear Engineering Research and Design Inst., No. 29, Hongcao Road, Shanghai, 200233 (China)

    2012-07-01T23:59:59.000Z

    This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade {sup 233}U-Thorium (U{sub 3}ThOX) FAs on the basis of reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade {sup 233}U extracted from burnt PuThOX fuel was used to fabrication of U{sub 3}ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U{sub 3}ThOX mixed core, the well designed U{sub 3}ThOX FAs with 1.94 w/o fissile uranium (mainly {sup 233}U) were located on the periphery of core as a blanket region. U{sub 3}ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U{sub 3}ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U{sub 3}ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U{sub 3}ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U{sub 3}ThOX loading on the periphery of core has no visible impacts on neutronic characteristics compared with reference full UOX core. The fuel cycle analysis has shown that {sup 233}U mono-recycling with U{sub 3}ThOX fuel could save 13% of natural uranium resource compared with UOX once through fuel cycle, slightly more than that of Plutonium single-recycling with MOX fuel. If {sup 233}U multi-recycling with U{sub 3}ThOX fuel is implemented, more natural uranium resource would be saved. (authors)

  5. 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.

    1980-01-01T23:59:59.000Z

    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.

  6. Development code for sensitivity and uncertainty analysis of input on the MCNPX for neutronic calculation in PWR core

    SciTech Connect (OSTI)

    Hartini, Entin, E-mail: entin@batan.go.id; Andiwijayakusuma, Dinan, E-mail: entin@batan.go.id [Center for Development of Nuclear Informatics - National Nuclear Energy Agency, PUSPIPTEK, Serpong, Tangerang, Banten (Indonesia)

    2014-09-30T23:59:59.000Z

    This research was carried out on the development of code for uncertainty analysis is based on a statistical approach for assessing the uncertainty input parameters. In the butn-up calculation of fuel, uncertainty analysis performed for input parameters fuel density, coolant density and fuel temperature. This calculation is performed during irradiation using Monte Carlo N-Particle Transport. The Uncertainty method based on the probabilities density function. Development code is made in python script to do coupling with MCNPX for criticality and burn-up calculations. Simulation is done by modeling the geometry of PWR terrace, with MCNPX on the power 54 MW with fuel type UO2 pellets. The calculation is done by using the data library continuous energy cross-sections ENDF / B-VI. MCNPX requires nuclear data in ACE format. Development of interfaces for obtaining nuclear data in the form of ACE format of ENDF through special process NJOY calculation to temperature changes in a certain range.

  7. The toughness of irradiated pressure water reactor (PWR) vessel shell rings and the effect of segregation zones

    SciTech Connect (OSTI)

    Bethmont, M.; Frund, J.M. [Electricite de France, Moret-sur-Loing (France); Housin, B. [Framatome, Paris La Defense (France). Materials and Technology Dept.; Soulat, P. [Commissariat a l`Energie Atomique, Gif-sur-Yvette (France)

    1996-12-31T23:59:59.000Z

    To establish the integrity of pressure water reactor (PWR) vessels it is necessary to determine the toughness of A508Cl.3 steel at the end of its life, that is after thermal aging and irradiation embrittlement. In safety analyses the toughness can be deduced from a reference curve set forth in the code (ASME or RCC-M). The validity of the reference curve has been verified for several years for unirradiated French reactor vessels. Tests were performed on specimens taken from materials having heterogeneities in chemical composition. For most of the test results the reference curve is a lower bound. To solve te problem of determining the toughness of SA 508 Cl.3 steel after irradiation and in the presence of possible heterogeneities, the toughness results were gathered. The synthesis shows that the RCC-M code curve is conservative.

  8. Ten year RPV inspections experiences in a PWR in Spain: Improvements in the inner-radius inspection techniques

    SciTech Connect (OSTI)

    Gonzalez, E.; Willke, A. [Tecnatom, S.A., Madrid (Spain)

    1994-12-31T23:59:59.000Z

    The in-service inspection of an RPV, performed in accordance with the scope and requirements of Section 11 of the ASME Code at the end of the ten year interval, is one of the most complicated ISI activities carried out. Special resources and tools are required for successful performance of this type of inspection: (1) preparation and planning; (2) mechanical scanner; (3) data acquisition and analysis system; and (4) ultrasonic techniques. This paper describes the most relevant issues relating to RPV inspection, along with the experience obtained during the inspection of the RPV of a 930 MW Spanish PWR plant in 1992. Special attention is paid to the improvements achieved with respect to inspection of the inner-radius areas of the primary nozzles.

  9. Chemical System Decontamination at PWR Power Stations Biblis A and B by Advanced System Decontamination by Oxidizing Chemistry (ASDOC-D) Process Technology - 13081

    SciTech Connect (OSTI)

    Loeb, Andreas; Runge, Hartmut; Stanke, Dieter [NIS Ingenieurgesellschaft mbH, Industriestrasse 13, 63755 Alzenau (Germany)] [NIS Ingenieurgesellschaft mbH, Industriestrasse 13, 63755 Alzenau (Germany); Bertholdt, Horst-Otto [NCT Consulting, Leonhardstrasse 16-18, 90443 Nuernberg (Germany)] [NCT Consulting, Leonhardstrasse 16-18, 90443 Nuernberg (Germany); Adams, Andreas; Impertro, Michael; Roesch, Josef [RWE Power, 68643 Biblis (Germany)] [RWE Power, 68643 Biblis (Germany)

    2013-07-01T23:59:59.000Z

    For chemical decontamination of PWR primary systems the so called ASDOC-D process has been developed and qualified at the German PWR power station Biblis. In comparison to other chemical decontamination processes ASDOC-D offers a number of advantages: - ASDOC-D does not require separate process equipment but is completely operated and controlled by the nuclear site installations. Feeding of chemical concentrates into the primary system is done by means of the site's dosing systems. Process control is performed by standard site instrumentation and analytics. - ASDOC-D safely prevents any formation and precipitation of insoluble constituents - Since ASDOC-D is operated without external equipment there is no need for installation of such equipment in high radioactive radiation surrounding. The radioactive exposure rate during process implementation and process performance may therefore be neglected in comparison to other chemical decontamination processes. - ASDOC-D does not require auxiliary hose connections which usually bear high leakage risk. The above mentioned technical advantages of ASDOC-D together with its cost-effectiveness gave rise to Biblis Power station to agree on testing ASDOC-D at the volume control system of PWR Biblis unit A. By involving the licensing authorities as well as expert examiners into this test ASDOC-D received the official qualification for primary system decontamination in German PWR. As a main outcome of the achieved results NIS received contracts for full primary system decontamination of both units Biblis A and B (each 1.200 MW) by end of 2012. (authors)

  10. Performance-based ratemaking for electric utilities: Review of plans and analysis of economic and resource-planning issues. Volume 2, Appendices

    SciTech Connect (OSTI)

    Comnes, G.A.; Stoft, S.; Greene, N. [Lawrence Berkeley Lab., CA (United States); Hill, L.J. [Oak Ridge National Lab., TN (United States)

    1995-11-01T23:59:59.000Z

    This document contains summaries of the electric utilities performance-based rate plans for the following companies: Alabama Power Company; Central Maine Power Company; Consolidated Edison of New York; Mississippi Power Company; New York State Electric and Gas Corporation; Niagara Mohawk Power Corporation; PacifiCorp; Pacific Gas and Electric; Southern California Edison; San Diego Gas & Electric; and Tucson Electric Power. In addition, this document also contains information about LBNL`s Power Index and Incentive Properties of a Hybrid Cap and Long-Run Demand Elasticity.

  11. BIOMASS REBURNING - MODELING/ENGINEERING STUDIES

    SciTech Connect (OSTI)

    Vladimir Zamansky; Chris Lindsey; Vitali Lissianski

    2000-01-28T23:59:59.000Z

    This project is designed to develop engineering and modeling tools for a family of NO{sub x} control technologies utilizing biomass as a reburning fuel. During the ninth reporting period (September 27--December 31, 1999), EER prepared a paper Kinetic Model of Biomass Reburning and submitted it for publication and presentation at the 28th Symposium (International) on Combustion, University of Edinburgh, Scotland, July 30--August 4, 2000. Antares Group Inc, under contract to Niagara Mohawk Power Corporation, evaluated the economic feasibility of biomass reburning options for Dunkirk Station. A preliminary report is included in this quarterly report.

  12. Electric and gas utility marketing of residential energy conservation case studies

    SciTech Connect (OSTI)

    None

    1980-05-01T23:59:59.000Z

    The objective of this research was to obtain information about utility conservation marketing techniques from companies actively engaged in performing residential conservation services. Many utilities currently are offering comprehensive services (audits, listing of contractors and lenders, post-installation inspection, advertising, and performing consumer research). Activities are reported for the following utilities: Niagara Mohawk Power Corporation; Tampa Electric Company; Memphis Light, Gas, and Water Division; Northern States Power-Wisconsin; Public Service Company of Colorado; Arizona Public Service Company; Pacific Gas and Electric Company; Sacramento Municipal Utility District; and Pacific Power and Light Company.

  13. Scoping design analyses for optimized shipping casks containing 1-, 2-, 3-, 5-, 7-, or 10-year-old PWR spent fuel

    SciTech Connect (OSTI)

    Bucholz, J.A.

    1983-01-01T23:59:59.000Z

    This report details many of the interrelated considerations involved in optimizing large Pb, Fe, or U-metal spent fuel shipping casks containing 1, 2, 3, 5, 7, or 10-year-old PWR fuel assemblies. Scoping analyses based on criticality, shielding, and heat transfer considerations indicate that some casks may be able to hold as many as 18 to 21 ten-year-old PWR fuel assemblies. In the criticality section, a new type of inherently subcritical fuel assembly separator is described which uses hollow, borated stainless-steel tubes in the wall-forming structure between the assemblies. In another section, details of many n/..gamma.. shielding optimization studies are presented, including the optimal n/..gamma.. design points and the actual shielding requirements for each type of cask as a function of the age of the spent fuel and the number of assemblies in the cask. Multigroup source terms based on ORIGEN2 calculations at these and other decay times are also included. Lastly, the numerical methods and experimental correlations used in the steady-state and transient heat transfer analyses are fully documented, as are pertinent aspects of the SCOPE code for Shipping Cask Optimization and Parametric Evaluation. (While only casks for square, intact PWR fuel assemblies were considered in this study, the SCOPE code may also be used to design and analyze casks containing canistered spent fuel or other waste material. An abbreviated input data guide is included as an appendix).

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

    SciTech Connect (OSTI)

    Lau, C. W.; Demaziere, C. [Dept. of Applied Physics, Div. of Nuclear Engineering, Chalmers Univ. of Technology, 412 96 Gothenburg (Sweden); Nylen, H.; Sandberg, U. [Ringhals AB, 432 85 Vaeroebacka (Sweden)

    2012-07-01T23:59:59.000Z

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

  15. Preliminary results of the PWR low power and shutdown accident frequencies program: Coarse screening analysis for Surry

    SciTech Connect (OSTI)

    Chu, T.L.; Musicki, Z.; Luckas, W.; Wong, S.M.; Neymotin, L.; Diamond, D.J.; Hsu, C.J.; Bozoki, G.; Kohut, P.; Fitzpatrick, R. [Brookhaven National Lab., Upton, NY (United States); Siu, N. [Massachusetts Inst. of Tech., Cambridge, MA (United States)

    1991-12-31T23:59:59.000Z

    This document presents the preliminary internal events Level 1 results (including fire and flood) obtained as a result of a coarse screening analysis on the low power and shutdown accident frequencies of the Surry Nuclear Power Plant. The work was performed by Brookhaven National Laboratory (BNL) for the Nuclear Regulatory Commission Office of Nuclear Regulatory Research (RES). This coarse screening analysis was performed in support of the NRC staff`s follow-up actions subsequent to the March 20, 1990 Vogtle incident with the objective of providing high-level qualitative insights within a relatively short time frame. It is the first phase of a major study that will ultimately produce estimates on the core damage frequency of a pressurized water reactor (PWR) during low power and shutdown conditions. Phase 2 of the study will be guided by the Phase 1 results in order to concentrate the effort on the various plant operational states, the dominant accident sequences, and pertinent data items according to their importance to core damage frequency and risk.

  16. Preliminary results of the PWR low power and shutdown accident frequencies program: Coarse screening analysis for Surry

    SciTech Connect (OSTI)

    Chu, T.L.; Musicki, Z.; Luckas, W.; Wong, S.M.; Neymotin, L.; Diamond, D.J.; Hsu, C.J.; Bozoki, G.; Kohut, P.; Fitzpatrick, R. (Brookhaven National Lab., Upton, NY (United States)); Siu, N. (Massachusetts Inst. of Tech., Cambridge, MA (United States))

    1991-01-01T23:59:59.000Z

    This document presents the preliminary internal events Level 1 results (including fire and flood) obtained as a result of a coarse screening analysis on the low power and shutdown accident frequencies of the Surry Nuclear Power Plant. The work was performed by Brookhaven National Laboratory (BNL) for the Nuclear Regulatory Commission Office of Nuclear Regulatory Research (RES). This coarse screening analysis was performed in support of the NRC staff's follow-up actions subsequent to the March 20, 1990 Vogtle incident with the objective of providing high-level qualitative insights within a relatively short time frame. It is the first phase of a major study that will ultimately produce estimates on the core damage frequency of a pressurized water reactor (PWR) during low power and shutdown conditions. Phase 2 of the study will be guided by the Phase 1 results in order to concentrate the effort on the various plant operational states, the dominant accident sequences, and pertinent data items according to their importance to core damage frequency and risk.

  17. Level 1 probabilistic risk assessment of low power and shutdown operations at a PWR: Phase 2 results

    SciTech Connect (OSTI)

    Chu, T.L.; Bozoki, G.; Kohut, P.; Musicki, Z.; Wong, S.M.; Yang, J.; Hsu, C.J.; Diamond, D.J.; Su, R.F. (Brookhaven National Lab., Upton, NY (United States)); Holmes, B. (AEA Technology, London (United Kingdom)); Siu, N. (Massachusetts Inst. of Tech., Cambridge, MA (United States)); Bley, D.; Lin, J. (Pickard, Lowe and Garrick, Inc., Newport Beach, CA (United States))

    1992-01-01T23:59:59.000Z

    As a result of the Chernobyl accident and other precursor events (e.g., Diablo Canyon), the US Nuclear Regulatory Commission's (NRC's) Office of Nuclear Regulatory Research (RES) initiated an extensive project during 1989 to carefully examine the potential risks during Low Power and Shutdown (LP S) operations. Shortly after the program began, an event occurred at the Vogtle plant during shutdown, which further intensified the effort of the LP S program. In the LP S program, one pressurized water reactor (PWR), Surry, and one boiling water reactor (BWR), Grand Gulf, were selected, mainly because they were previously analyzed in the NUREG-1150 Study. The Level-1 Program is being performed in two phases. Phase 1 was dedicated to performing a coarse screening level-1 analysis including internal fire and flood. A draft report was completed in November, 1991. In the phase 2 study, mid-loop operations at the Surry plant were analyzed in detail. The objective of this paper is to present the approach of the phase 2 study and the preliminary results and insights.

  18. Level 1 probabilistic risk assessment of low power and shutdown operations at a PWR: Phase 2 results

    SciTech Connect (OSTI)

    Chu, T.L.; Bozoki, G.; Kohut, P.; Musicki, Z.; Wong, S.M.; Yang, J.; Hsu, C.J.; Diamond, D.J.; Su, R.F. [Brookhaven National Lab., Upton, NY (United States); Holmes, B. [AEA Technology, London (United Kingdom); Siu, N. [Massachusetts Inst. of Tech., Cambridge, MA (United States); Bley, D.; Lin, J. [Pickard, Lowe and Garrick, Inc., Newport Beach, CA (United States)

    1992-12-31T23:59:59.000Z

    As a result of the Chernobyl accident and other precursor events (e.g., Diablo Canyon), the US Nuclear Regulatory Commission`s (NRC`s) Office of Nuclear Regulatory Research (RES) initiated an extensive project during 1989 to carefully examine the potential risks during Low Power and Shutdown (LP&S) operations. Shortly after the program began, an event occurred at the Vogtle plant during shutdown, which further intensified the effort of the LP&S program. In the LP&S program, one pressurized water reactor (PWR), Surry, and one boiling water reactor (BWR), Grand Gulf, were selected, mainly because they were previously analyzed in the NUREG-1150 Study. The Level-1 Program is being performed in two phases. Phase 1 was dedicated to performing a coarse screening level-1 analysis including internal fire and flood. A draft report was completed in November, 1991. In the phase 2 study, mid-loop operations at the Surry plant were analyzed in detail. The objective of this paper is to present the approach of the phase 2 study and the preliminary results and insights.

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

    SciTech Connect (OSTI)

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

    2012-05-01T23:59:59.000Z

    The document contains the simulation results of a steady state model PWR problem with the RELAP-7 code. The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on INL's modern scientific software development framework - MOOSE (Multi-Physics Object-Oriented Simulation Environment). This report summarizes the initial results of simulating a model steady-state single phase PWR problem using the current version of the RELAP-7 code. The major purpose of this demonstration simulation is to show that RELAP-7 code can be rapidly developed to simulate single-phase reactor problems. RELAP-7 is a new project started on October 1st, 2011. It will become the main reactor systems simulation toolkit for RISMC (Risk Informed Safety Margin Characterization) and the next generation tool in the RELAP reactor safety/systems analysis application series (the replacement for RELAP5). The key to the success of RELAP-7 is the simultaneous advancement of physical models, numerical methods, and software design while maintaining a solid user perspective. Physical models include both PDEs (Partial Differential Equations) and ODEs (Ordinary Differential Equations) and experimental based closure models. RELAP-7 will eventually utilize well posed governing equations for multiphase flow, which can be strictly verified. Closure models used in RELAP5 and newly developed models will be reviewed and selected to reflect the progress made during the past three decades. RELAP-7 uses modern numerical methods, which allow implicit time integration, higher order schemes in both time and space, and strongly coupled multi-physics simulations. RELAP-7 is written with object oriented programming language C++. Its development follows modern software design paradigms. The code is easy to read, develop, maintain, and couple with other codes. Most importantly, the modern software design allows the RELAP-7 code to evolve with time. RELAP-7 is a MOOSE-based application. MOOSE (Multiphysics Object-Oriented Simulation Environment) is a framework for solving computational engineering problems in a well-planned, managed, and coordinated way. By leveraging millions of lines of open source software packages, such as PETSC (a nonlinear solver developed at Argonne National Laboratory) and LibMesh (a Finite Element Analysis package developed at University of Texas), MOOSE significantly reduces the expense and time required to develop new applications. Numerical integration methods and mesh management for parallel computation are provided by MOOSE. Therefore RELAP-7 code developers only need to focus on physics and user experiences. By using the MOOSE development environment, RELAP-7 code is developed by following the same modern software design paradigms used for other MOOSE development efforts. There are currently over 20 different MOOSE based applications ranging from 3-D transient neutron transport, detailed 3-D transient fuel performance analysis, to long-term material aging. Multi-physics and multiple dimensional analyses capabilities can be obtained by coupling RELAP-7 and other MOOSE based applications and by leveraging with capabilities developed by other DOE programs. This allows restricting the focus of RELAP-7 to systems analysis-type simulations and gives priority to retain and significantly extend RELAP5's capabilities.

  20. Reactivity loss validation of high burn-up PWR fuels with pile-oscillation experiments in MINERVE

    SciTech Connect (OSTI)

    Leconte, P.; Vaglio-Gaudard, C.; Eschbach, R.; Di-Salvo, J.; Antony, M.; Pepino, A. [CEA, DEN, DER, Cadarache, F-13108 Saint-Paul-Lez-Durance (France)

    2012-07-01T23:59:59.000Z

    The ALIX experimental program relies on the experimental validation of the spent fuel inventory, by chemical analysis of samples irradiated in a PWR between 5 and 7 cycles, and also on the experimental validation of the spent fuel reactivity loss with bum-up, obtained by pile-oscillation measurements in the MINERVE reactor. These latter experiments provide an overall validation of both the fuel inventory and of the nuclear data responsible for the reactivity loss. This program offers also unique experimental data for fuels with a burn-up reaching 85 GWd/t, as spent fuels in French PWRs never exceeds 70 GWd/t up to now. The analysis of these experiments is done in two steps with the APOLLO2/SHEM-MOC/CEA2005v4 package. In the first one, the fuel inventory of each sample is obtained by assembly calculations. The calculation route consists in the self-shielding of cross sections on the 281 energy group SHEM mesh, followed by the flux calculation by the Method Of Characteristics in a 2D-exact heterogeneous geometry of the assembly, and finally a depletion calculation by an iterative resolution of the Bateman equations. In the second step, the fuel inventory is used in the analysis of pile-oscillation experiments in which the reactivity of the ALIX spent fuel samples is compared to the reactivity of fresh fuel samples. The comparison between Experiment and Calculation shows satisfactory results with the JEFF3.1.1 library which predicts the reactivity loss within 2% for burn-up of {approx}75 GWd/t and within 4% for burn-up of {approx}85 GWd/t. (authors)

  1. Evaluation of the thermal-hydraulic response and fuel rod thermal and mechanical deformation behavior during the power burst facility test LOC-3. [PWR

    SciTech Connect (OSTI)

    Yackle, T.R.; MacDonald, P.E.; Broughton, J.M.

    1980-01-01T23:59:59.000Z

    An evaluation of the results from the LOC-3 nuclear blowdown test conducted in the Power Burst Facility is presented. The test objective was to examine fuel and cladding behavior during a postulated cold leg break accident in a pressurized water reactor (PWR). Separate effects of rod internal pressure and the degree of irradiation were investigated in the four-rod test. Extensive cladding deformation (ballooning) and failure occurred during blowdown. The deformation of the low and high pressure rods was similar; however, the previously irradiated test rod deformed to a greater extent than a similar fresh rod exposed to identical system conditions.

  2. Please cite this article in press as: Malen, J.A., et al., Thermal hydraulic design of a hydride-fueled inverted PWR core. Nucl. Eng. Des. (2009), doi:10.1016/j.nucengdes.2009.02.026

    E-Print Network [OSTI]

    Malen, Jonathan A.

    2009-01-01T23:59:59.000Z

    Please cite this article in press as: Malen, J.A., et al., Thermal hydraulic design of a hydride hydraulic design of a hydride-fueled inverted PWR core J.A. Malena, , N.E. Todreasb , P. Hejzlarb , P and its thermal hydraulic performance is compared to that of a standard rod bundle core design also fueled

  3. A Critical Review of Practice of Equating the Reactivity of Spent Fuel to Fresh Fuel in Burnup Credit Criticality Safety Analyses for PWR Spent Fuel Pool Storage

    SciTech Connect (OSTI)

    Wagner, J.C.; Parks, C.V.

    2000-09-01T23:59:59.000Z

    This research examines the practice of equating the reactivity of spent fuel to that of fresh fuel for the purpose of performing burnup credit criticality safety analyses for PWR spent fuel pool (SFP) storage conditions. The investigation consists of comparing k{sub inf} estimates based on reactivity equivalent fresh fuel enrichment (REFFE) to k{sub inf} estimates using the actual spent fuel isotopics. Analyses of selected storage configurations common in PWR SFPs show that this practice yields nonconservative results (on the order of a few tenths of a percent) in configurations in which the spent fuel is adjacent to higher-reactivity assemblies (e.g., fresh or lower-burned assemblies) and yields conservative results in configurations in which spent fuel is adjacent to lower-reactivity assemblies (e.g., higher-burned fuel or empty cells). When the REFFE is determined based on unborated water moderation, analyses for storage conditions with soluble boron present reveal significant nonconservative results associated with the use of the REFFE. This observation is considered to be important, especially considering the recent allowance of credit for soluble boron up to 5% in reactivity. Finally, it is shown that the practice of equating the reactivity of spent fuel to fresh fuel is acceptable, provided the conditions for which the REFFE was determined remain unchanged. Determination of the REFFE for a reference configuration and subsequent use of the REFFE for different configurations violates the basis used for the determination of the REFFE and, thus, may lead to inaccurate, and possibly, nonconservative estimates of reactivity. A significant concentration ({approximately}2000 ppm) of soluble boron is typically (but not necessarily required to be) present in PWR SFPs, of which only a portion ({le} 500 ppm) may be credited in safety analyses. Thus, a large subcritical margin currently exists that more than accounts for errors or uncertainties associated with the use of the REFFE. Consequently, the findings presented here do not represent a significant safety concern unless/until the subcritical margin associated with the soluble boron (that is not currently explicitly credited) is offset by the uncertainties associated with burnup credit and/or the expanded allowance of credit for the soluble boron.

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

    SciTech Connect (OSTI)

    Ade, Brian J [ORNL; Gauld, Ian C [ORNL

    2011-10-01T23:59:59.000Z

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

  5. PWR FLECHT SEASET 21-rod bundle flow blockage task data and analysis report. NRC/EPRI/Westinghouse Report No. 11. Appendices K-P

    SciTech Connect (OSTI)

    Loftus, M.J.; Hochreiter, L.E.; Lee, N.; McGuire, M.F.; Wenzel, A.H.; Valkovic, M.M.

    1982-09-01T23:59:59.000Z

    This report presents data and limited analysis from the 21-Rod Bundle Flow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Separate Effects and Systems Effects Test Program (FLECHT SEASET). The tests consisted of forced and gravity reflooding tests utilizing electrical heater rods with a cosine axial power profile to simulate PWR nuclear core fuel rod arrays. Steam cooling and hydraulic characteristics tests were also conducted. These tests were utilized to determine effects of various flow blockage configurations (shapes and distributions) on reflooding behavior, to aid in development/assessment of computational models in predicting reflooding behavior of flow blockage configurations, and to screen flow blockage configurations for future 163-rod flow blockage bundle tests.

  6. Crack growth rates and metallographic examinations of Alloy 600 and Alloy 82/182 from field components and laboratory materials tested in PWR environments.

    SciTech Connect (OSTI)

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

    2008-05-05T23:59:59.000Z

    In light water reactors, components made of nickel-base alloys are susceptible to environmentally assisted cracking. This report summarizes the crack growth rate results and related metallography for field and laboratory-procured Alloy 600 and its weld alloys tested in pressurized water reactor (PWR) environments. The report also presents crack growth rate (CGR) results for a shielded-metal-arc weld of Alloy 182 in a simulated PWR environment as a function of temperature between 290 C and 350 C. These data were used to determine the activation energy for crack growth in Alloy 182 welds. The tests were performed by measuring the changes in the stress corrosion CGR as the temperatures were varied during the test. The difference in electrochemical potential between the specimen and the Ni/NiO line was maintained constant at each temperature by adjusting the hydrogen overpressure on the water supply tank. The CGR data as a function of temperature yielded activation energies of 252 kJ/mol for a double-J weld and 189 kJ/mol for a deep-groove weld. These values are in good agreement with the data reported in the literature. The data reported here and those in the literature suggest that the average activation energy for Alloy 182 welds is on the order of 220-230 kJ/mol, higher than the 130 kJ/mol commonly used for Alloy 600. The consequences of using a larger value of activation energy for SCC CGR data analysis are discussed.

  7. MHK Projects/Niagara Community 2 | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are beingZealand Jump to: navigation, searchOf KilaueaInformationCygnet <| OpenMarisolNJBPUVicksberg, MS

  8. MHK Projects/Niagara Community | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are beingZealand Jump to: navigation, searchOf KilaueaInformationCygnet <| OpenMarisolNJBPUVicksberg,

  9. Niagara Falls, New York: Energy Resources | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directedAnnual SiteofEvaluatingGroup |JilinLuOpen EnergyNelsoniX LtdNew EnergyCity DataNextEra

  10. Niagara County, New York: Energy Resources | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov YouKizildere I Geothermal Pwer PlantMunhall,Missouri: Energy Resources Jump to: navigation,NextEraCounty, New York: Energy

  11. Niagara Falls, New York: Energy Resources | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov YouKizildere I Geothermal Pwer PlantMunhall,Missouri: Energy Resources Jump to: navigation,NextEraCounty, New York:

  12. DOE - Office of Legacy Management -- Niagara VP_FUSRAP

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA group currentBradleyTableSelling Corp - CT 0-01Naturita36New York, New

  13. Niagara Falls, NY Natural Gas Exports to Canada

    Gasoline and Diesel Fuel Update (EIA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40 Buildingto ChinaThousand CubicSeparation 29 0Year Jan0 0 0 0 03a

  14. Niagara Falls, NY Natural Gas Imports by Pipeline from Canada

    Gasoline and Diesel Fuel Update (EIA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40 Buildingto ChinaThousand CubicSeparation 29 0Year Jan0 0 0 0 03a188,525

  15. Mr. Frank Archer President Niagara Cold Drawn Steel Corporation

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA groupTuba City, Arizona, DisposalFourth Five-Year38Report3Department of Energy

  16. ADDENDUM TO ACTION DESCRIPTION MEMORANDUM NIAGARA FALLS STORAGE SITE

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA groupTuba City, Arizona, DisposalFourthN V4100U.S.Indian10WayneAnnualLinited[

  17. St. Regis Mohawk Tribe Paves the Way to a Sustainable Future...

    Energy Savers [EERE]

    efforts, including the expansion of AHA's Sunrise Acres in 2011 to include 20 new ultra-efficient homes for low-income seniors. For the strategic energy planning portion of...

  18. A review of "Mohawk Saint: Catherine Tekakwitha and the Jesuits." by Allan Greer

    E-Print Network [OSTI]

    Br. Benet Exton, O.S.B.

    2005-01-01T23:59:59.000Z

    on major writers in early modern England. With its insistence that inwardness matters as much as the social forces that regulate identity, the book represents an important contribution to theories of Renaissance subjec- tivity and identity. Allan Greer... Tekakawitha. She died in 1680, and progress of her cause for sainthood has taken a long time. She has not been canonized a saint although the elusive miracle needed has reportedly occurred, and so it is possible that Pope Benedict XVI will canonize her. Allan...

  19. St. Regis Mohawk Tribe Paves the Way to a Sustainable Future; Kicks Off

    Energy Savers [EERE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directed offOCHCO Overview OCHCO OverviewRepositoryManagement |SolarSpecial Report:SpectrumEnergy

  20. ORNL rod-bundle heat-transfer test data. Volume 3. Thermal-hydraulic test facility experimental data report for test 3. 06. 6B - transient film boiling in upflow. [PWR

    SciTech Connect (OSTI)

    Mullins, C.B.; Felde, D.K.; Sutton, A.G.; Gould, S.S.; Morris, D.G.; Robinson, J.J.

    1982-05-01T23:59:59.000Z

    Reduced instrument responses are presented for Thermal-Hyraulic Test Facility (THTF) Test 3.06.6B. This test was conducted by members of the Oak Ridge National Laboratory Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on August 29, 1980. The objective of the program was to investigate heat transfer phenomena believed to occur in PWR's during accidents, including small and large break loss-of-coolant accidents. Test 3.06.6B was conducted to obtain transient film boiling data in rod bundle geometry under reactor accident-type conditions. The primary purpose of this report is to make the reduced instrument responses for THTF Test 3.06.6B available. Included in the report are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers.

  1. Experimental Investigation on the Effects of Coolant Concentration on Sub-Cooled Boiling and Crud Deposition on Reactor Cladding at Prototypical PWR Operating Conditions

    SciTech Connect (OSTI)

    Schultis, J., Kenneth; Fenton, Donald, L.

    2006-10-20T23:59:59.000Z

    Increasing demand for energy necessitates nuclear power units to increase power limits. This implies significant changes in the design of the core of the nuclear power units, therefore providing better performance and safety in operations. A major hindrance to the increase of nuclear reactor performance especially in Pressurized Deionized water Reactors (PWR) is Axial Offset Anomaly (AOA)--the unexpected change in the core axial power distribution during operation from the predicted distribution. This problem is thought to be occur because of precipitation and deposition of lithiated compounds like boric acid (H{sub 2}BO{sub 3}) and lithium metaborate (LiBO{sub 2}) on the fuel rod cladding. Deposited boron absorbs neutrons thereby affecting the total power distribution inside the reactor. AOA is thought to occur when there is sufficient build-up of crud deposits on the cladding during subcooled nucleate boiling. Predicting AOA is difficult as there is very little information regarding the heat and mass transfer during subcooled nucleate boiling. An experimental investigation was conducted to study the heat transfer characteristics during subcooled nucleate boiling at prototypical PWR conditions. Pool boiling tests were conducted with varying concentrations of lithium metaborate (LiBO{sub 2}) and boric acid (H{sub 2}BO{sub 3}) solutions in deionized water. The experimental data collected includes the effect of coolant concentration, subcooling, system pressure and heat flux on pool the boiling heat transfer coefficient. The analysis of particulate deposits formed on the fuel cladding surface during subcooled nucleate boiling was also performed. The results indicate that the pool boiling heat transfer coefficient degrades in the presence of boric acid and lithium metaborate compared to pure deionized water due to lesser nucleation. The pool boiling heat transfer coefficients decreased by about 24% for 5000 ppm concentrated boric acid solution and by 27% for 5000 ppm lithium metaborate solution respectively at the saturation temperature for 1000 psi (68.9 bar) coolant pressure. Boiling tests also revealed the formation of fine deposits of boron and lithium on the cladding surface which degraded the heat transfer rates. The boron and lithium metaborate precipitates after a 5 day test at 5000 ppm concentration and 1000 psi (68.9 bar) operating pressure reduced the heat transfer rate 21% and 30%, respectively for the two solutions.

  2. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 5. Probabilistic fracture mechanics analysis. Load Combination Program Project I final report

    SciTech Connect (OSTI)

    Harris, D.O.; Lim, E.Y.; Dedhia, D.D.

    1981-06-01T23:59:59.000Z

    The primary purpose of the Load Combination Program covered in this report is to estimate the probability of a seismic induced LOCA in the primary piping of a commercial pressurized water reactor (PWR). Best estimates, rather than upper bound results are desired. This was accomplished by use of a fracture mechanics model that employs a random distribution of initial cracks in the piping welds. Estimates of the probability of cracks of various sizes initially existing in the welds are combined with fracture mechanics calculations of how these cracks would grow during service. This then leads to direct estimates of the probability of failure as a function of time and location within the piping system. The influence of varying the stress history to which the piping is subjected is easily determined. Seismic events enter into the analysis through the stresses they impose on the pipes. Hence, the influence of various seismic events on the piping failure probability can be determined, thereby providing the desired information.

  3. Spent fuel dry storage technology development: fuel temperature measurements under imposed dry storage conditions (I kW PWR spent fuel assembly)

    SciTech Connect (OSTI)

    Unterzuber, R.; Wright, J.B.

    1980-09-01T23:59:59.000Z

    A spent fuel assembly temperature test under imposed dry storage conditions was conducted at the Engine Maintenance Assembly and Disassembly (E-MAD) facility on the Nevada Test Site in support of spent fuel dry storage technology development. This document presents the test data and results obtained from an approximately 1.0 kW decay heat level PWR spent fuel assembly. A spent fuel test apparatus was designed to utilize a representative stainless steel spent fuel canister, a canister lid containing internal temperature instrumentation to measure fuel cladding temperatures, and a carbon steel liner that encloses the canister and lid. Electrical heaters along the liner length, on the lid, and below the canister are used to impose dry storage canister temperature profiles. Temperature instrumentation is provided on the liner and canister. The liner and canister are supported by a test stand in one of the large hot cells (West Process Cell) inside E-MAD. Fuel temperature measurements have been performed using imposed canister temperature profiles from the electrically heated and spent fuel drywell tests being conducted at E-MAD as well as for four constant canister temperature profiles, each with a vacuum, helium and air backfill. Computer models have been utilized in conjunction with the test to predict the thermal response of the fuel cladding. Computer predictions are presented, and they show good agreement with the test data.

  4. Neutronic Analysis of the Burning of Transuranics in Fully Ceramic Micro-Encapsulated Tri-Isotropic Particle-Fuel in a PWR

    SciTech Connect (OSTI)

    Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag; Gilles Youinou; Brian Boer

    2012-11-01T23:59:59.000Z

    Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU) – only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO2 and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO2 and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior is dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.

  5. The impact of wind turbines on birds in upstate New York

    SciTech Connect (OSTI)

    Cooper, B.A. [ABR, Inc., Forest Grove, OR (United States); Johnson, C.B. [ABR, Inc., Fairbanks, AK (United States)

    1995-12-31T23:59:59.000Z

    During spring and fall 1995, ABR, Inc., an environmental research firm, used radar and visual techniques to study bird migration near proposed and existing wind-turbine sites in upstate New York for Niagara Mohawk Power Corporation. The primary goal of the study was to evaluate the possible impacts of wind turbines and meteorological towers on local and migratory birds during the spring and fall migration periods. Here we primarily report on data collected from the existing wind-turbine site at Copenhagen. In addition to visual observations of diurnal movements of birds, two radars were used for observations of migrating birds at night. The surveillance radar provided information on nocturnal migration rates, flight directions, and flight behavior. The vertical radar provided information on flight altitudes.

  6. Death of a carbonate basin: The Niagara-Salina transition in the Michigan basin

    SciTech Connect (OSTI)

    Leibold, A.W.; Howell, P.D. (Univ. of Michigan, Ann Arbor (United States))

    1991-03-01T23:59:59.000Z

    The A-O Carbonate in the Michigan basin comprises a sequence of laminated calcite/anhydrite layers intercalated with bedded halite at the transition between normal marine Niagaran carbonates and lower Salina Group evaporites. The carbonate/anhydrite interbeds represent freshing events during initial evaporative concentration of the Michigan basin. Recent drilling in the Michigan basin delineates two distinct regions of A-O Carbonate development: a 5 to 10 m thick sequence of six 'laminites' found throughout most of the western and northern basin and a 10 to 25 m thick sequence in the southeastern basin containing both thicker 'laminates' and thicker salt interbeds. Additionally, potash deposits of the overlying A-1 evaporite unit are restricted to the northern and western basin regions. The distribution of evaporite facies in these two regions is adequately explained by a source of basin recharge in the southeast-perhaps the 'Clinton Inlet' of earlier workers. This situation suggest either that: (1) the source of basin recharge is alternately supplying preconcentrated brine and more normal marine water, or (2) that the basin received at least two distinct sources of water during A-O deposition.

  7. Niagara Falls, NY Natural Gas Pipeline Exports to Canada (Million Cubic

    Gasoline and Diesel Fuel Update (EIA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for On-Highway4,1,50022,3,,,,6,1,9,1,50022,3,,,,6,1,Decade1 Source: Office(BillionYear Jan Feb Mar Apr May JunNew

  8. Niagara Falls, NY Natural Gas Pipeline Imports From Canada (Million Cubic

    Gasoline and Diesel Fuel Update (EIA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for On-Highway4,1,50022,3,,,,6,1,9,1,50022,3,,,,6,1,Decade1 Source: Office(BillionYear Jan Feb Mar Apr May JunNewFeet) Decade

  9. Evaluation of Final Radiological Conditions at Areas of the Niagara Falls

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic Plan Departmentof1-SCORECARD-09-21-11 Page5-03 Evaluation

  10. Evaluation of Final Radiological Conditions at Areas of the Niagara Falls

    Office of Environmental Management (EM)

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  11. PRELIMINARY SURVEY OF THE UNION CARBIDE CORPORATION METALS DIVISION PLANT, NIAGARA FALLS, NEW YORK

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA group currentBradleyTableSelling7 AugustAFRICAN .METALS~TEXAS CITY CHEMICALS,e -

  12. Niagara Falls, NY Natural Gas Pipeline Exports to Canada (Million Cubic

    Gasoline and Diesel Fuel Update (EIA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40 Buildingto ChinaThousand CubicSeparation 29 0Year Jan0 0 0 0Thousand

  13. Niagara Falls, NY Natural Gas Pipeline Imports From Canada (Million Cubic

    Gasoline and Diesel Fuel Update (EIA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122 40 Buildingto ChinaThousand CubicSeparation 29 0Year Jan0 0 0Feet) Year Jan

  14. Niagara Falls, NY Natural Gas Pipeline Exports to Canada (Dollars per

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5 Tables July 1996 Energy Information Administration Office ofthrough 1996) inThousandWithdrawals (MillionNine Mile Pointper0 0Thousand

  15. Niagara Falls, NY Natural Gas Pipeline Exports to Canada (Dollars per

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5 Tables July 1996 Energy Information Administration Office ofthrough 1996) inThousandWithdrawals (MillionNine Mile Pointper0

  16. Niagara Falls, NY Natural Gas Pipeline Exports to Canada (Million Cubic

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5 Tables July 1996 Energy Information Administration Office ofthrough 1996) inThousandWithdrawals (MillionNine Mile

  17. Niagara Falls, NY Natural Gas Pipeline Imports From Canada (Dollars per

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5 Tables July 1996 Energy Information Administration Office ofthrough 1996) inThousandWithdrawals (MillionNine MileThousand Cubic

  18. Niagara Falls, NY Natural Gas Pipeline Imports From Canada (Dollars per

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5 Tables July 1996 Energy Information Administration Office ofthrough 1996) inThousandWithdrawals (MillionNine MileThousand

  19. Niagara Falls, NY Natural Gas Pipeline Imports From Canada (Million Cubic

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5 Tables July 1996 Energy Information Administration Office ofthrough 1996) inThousandWithdrawals (MillionNine

  20. GROUND LEVEL INVESTIGATION OF ANOMALOUS RADIATION LEVELS IN NIAGARA FALLS, NEW YORK

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA group currentBradleyTableSelling CorpNewCF INDUSTRIES,L? . -. .- *' * ..+GROUND

  1. COT"IPREITENS IVE RADIOLOGICAL SURVEY OFF-SITE PROPERTY P NIAGARA FALIS STORAGE SITE

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA groupTuba City, Arizona, DisposalFourthNrr-osams ADMIN RCDBaseline t-)

  2. PWR Fuel Shipping Limits & RNP Core Design

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

    operating reserve at all nuclear units Spent fuel shipping program to reduce inventories at Brunswick and Robinson Maximize use of Harris spent fuel pools 5 Transportation...

  3. Filtered-vented containment systems. [PWR; BWR

    SciTech Connect (OSTI)

    Benjamin, A.S.; Walling, H.C.; Cybulskis, P.; DiSalvo, R.

    1980-01-01T23:59:59.000Z

    The potential benefits of filtered-vented containment systems as a means for mitigating the effects of severe accidents are analyzed. Studies so far have focused upon two operating reactor plants in the United States, a large-containment pressurized water reactor and a Mark I containment boiling water reactor. Design options that could be retrofitted to these plants are described including single-component once-through venting systems, multiple-component systems with vent and recirculation capabilities, and totally contained venting systems. A variety of venting strategies are also described which include simple low-volume containment pressure relief strategies and more complicated, high-volume venting strategies that require anticipatory actions. The latter type of strategy is intended for accidents that produce containment-threatening pressure spikes.

  4. Evaluation of condensate polishers. Final report. [PWR

    SciTech Connect (OSTI)

    Lurie, S.W.

    1983-06-01T23:59:59.000Z

    The potential for steam generator corrosion caused by the release of resins or soluble impurity chemicals from full-flow condensate polishing was evaluated in a series of high temperature tests. The tests were designed to operate within the then prevailing NSSS steam generator chemistry specifications, consistent with realistic release of these impurities to steam generators. Each potential corrodent was tested separately in the absence of other corrosive conditions.

  5. Harquahala Valley Pwr District | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are8COaBulkTransmissionSitingProcess.pdfGetec AG| OpenInformation Handbook forHansungHarney

  6. Overhearing in 802.11 mesh networks

    E-Print Network [OSTI]

    Afanasyev, Mikhail

    2009-01-01T23:59:59.000Z

    through a port on an HP 2626-PWR switch. There are sevenuplink, but no neighbors in the mesh. ) CDF of nodes Pwr5 Pwr 10 Pwr 15 Pwr 20 Pwr 30 Pwr 40 Pwr 50 Pwr 60

  7. Combining water budgets and IFIM results for analyzing operation alternatives at peaking projects

    SciTech Connect (OSTI)

    Conners, M.E.; Homa, J. Jr. [Ichthyological Associates, Inc., Lansing, NY (United States); Carrington, G. [Northrup, Devine, and Tarbell, Inc., Vancouver, WA (United States)

    1995-12-31T23:59:59.000Z

    Licensing of hydropower projects often involves evaluating and comparing several different alternatives for project operation. Projects with peaking capabilities, in particular, are frequently required to compare peaking operation with substantially different alternatives, such as continuous run-of-the-river flows. Instream flow studies are used to assess the environmental impacts of hydropower operation by modeling the amount of aquatic habitat available at various flows. It can be difficult, however, to apply instream flow models downstream of peaking operations, or to present habitat model results in a way that clearly compares operation alternatives. This paper presents a two-stage analysis that was used in the successful negotiation of a licensing settlement for Niagara Mohawk Power Corporation`s Salmon River Project in upstate New York. A water budget model based on project configuration was used to compile flow-duration curves for the project under several alternative operating rules. A spreadsheet model was developed that combines the results of instream flow habitat models with flow-duration statistics. This approach provides a clear, quantitative comparison of the effect of alternative project operations on downstream aquatic habitat.

  8. Report on discussions with utility engineers about superconducting generators

    SciTech Connect (OSTI)

    none,

    1996-03-01T23:59:59.000Z

    This report relates to a series of discussions with electric utility engineers concerning the integration of high-temperature superconducting (HTS) generators into the present electric power system. The current and future interest of the utilities in the purchase and use of HTS generators is assessed. Various performance and economic factors are also considered as part of this inspection of the utility prospects for HTS generators. Integration of HTS generators into the electric utility sector is one goal of the Superconductivity Partnership Initiative (SPI). The SPI, a major part of the Department of Energy (DOE) Superconductivity Program for Electric Systems, features vertical teaming of a major industrial power apparatus manufacturers, a producer of HTS wire, and an end-user with assistance and technical support for the national laboratories. The SPI effort on HTS generators is headed by a General Electric Corporation internal team comprised of the Corporate Research Laboratories, Power Generation Engineering, and Power Systems Group. Intermagnetics General corporation, which assisted in the development of the superconducting coils, is the HTS wire and tape manufacturer. Additional technical support is provided by the national laboratories: Argonne, Los Alamos, and Oak Ridge, and the New York State Institute on Superconductivity. The end-user is represented by Niagara-Mohawk and the Electric Power Research Institute.

  9. Not All Large Customers are Made Alike: Disaggregating Response toDefault-Service Day-Ahead Market Pricing

    SciTech Connect (OSTI)

    Hopper, Nicole; Goldman, Charles; Neenan, Bernie

    2006-05-12T23:59:59.000Z

    For decades, policymakers and program designers have gone onthe assumption that large customers, particularly industrial facilities,are the best candidates for realtime pricing (RTP). This assumption isbased partly on practical considerations (large customers can providepotentially large load reductions) but also on the premise thatbusinesses focused on production cost minimization are most likely toparticipate and respond to opportunities for bill savings. Yet fewstudies have examined the actual price response of large industrial andcommercial customers in a disaggregated fashion, nor have factors such asthe impacts of demand response (DR) enabling technologies, simultaneousemergency DR program participation and price response barriers been fullyelucidated. This second-phase case study of Niagara Mohawk PowerCorporation (NMPC)'s large customer RTP tariff addresses theseinformation needs. The results demonstrate the extreme diversity of largecustomers' response to hourly varying prices. While two-thirdsexhibitsome price response, about 20 percent of customers provide 75-80 percentof the aggregate load reductions. Manufacturing customers are mostprice-responsive as a group, followed by government/education customers,while other sectors are largely unresponsive. However, individualcustomer response varies widely. Currently, enabling technologies do notappear to enhance hourly price response; customers report using them forother purposes. The New York Independent System Operator (NYISO)'semergency DR programs enhance price response, in part by signaling tocustomers that day-ahead prices are high. In sum, large customers docurrently provide moderate price response, but there is significant roomfor improvement through targeted programs that help customers develop andimplement automated load-response strategies.

  10. Biomass Reburning: Modeling/Engineering Studies

    SciTech Connect (OSTI)

    Vladimir M. Zamansky

    1998-01-20T23:59:59.000Z

    Reburning is a mature fuel staging NO{sub x} control technology which has been successfully demonstrated at full scale by Energy and Environmental Research Corporation (EER) and others on numerous occasions. Based on chemical kinetic modeling and experimental combustion studies, EER is currently developing novel concepts to improve the efficiency of the basic gas reburning process and to utilize various renewable and waste fuels for NO{sub x} control. This project is designed to develop engineering and modeling tools for a family of NO{sub x} control technologies utilizing biomass as a reburning fuel. Basic and advanced biomass reburning have the potential to achieve 60-90+% NO{sub x} control in coal fired boilers at a significantly lower cost than SCR. The scope of work includes modeling studies (kinetic, CFD, and physical modeling), experimental evaluation of slagging and fouling associated with biomass reburning, and economic study of biomass handling requirements. Project participants include: EER, FETC R and D group, Niagara Mohawk Power Corporation and Antares, Inc. Most of the combustion experiments on development of biomass reburning technologies are being conducted in the scope of coordinated SBIR program funded by USDA. The first reporting period (October 1--December 31, 1997) included preparation of project management plan and organization of project kick-off meeting at DOE FETC. The quarterly report briefly describes the management plan and presents basic information about the kick-off meeting.

  11. Cycling operation of fossil plants

    SciTech Connect (OSTI)

    Devendorf, D.; Kulczycky, T.G. (Niagara Mohawk Power Corp., Syracuse, NY (USA))

    1991-05-01T23:59:59.000Z

    A necessity for many utilities today is the cycling of their fossil units. Fossil plants with their higher fuel costs are being converted to cycling operation to accommodate daily load swings and to decrease the overall system fuel costs. For a large oil-fired unit, such as Oswego Steam Station Unit 5, millions of dollars can be saved annually in fuel costs if the unit operates in a two-shift mode. However, there are also penalties attributable to cycling operation which are associated with availability and thermal performance. The objectives of Niagara Mohawk Power Corporation were to minimize the losses in availability and performance, and the degradation in the life of the equipment by incorporating certain cycling modifications into the unit. The objective of this project was to evaluate the effectiveness of three of these cycling modifications: (1) the superheater and turbine bypass (Hot Restart System), (2) the use of variable pressure operation, and (3) the full-flow condensate polishing system. To meet this objective, Unit 5 was tested using the cycling modifications, and a dynamic mathematical model of this unit was developed using the Modular Modeling System (MMS) Code from EPRI. This model was used to evaluate various operating modes and to assist in the assessment of operating procedures. 15 refs., 41 figs., 22 tabs.

  12. U.S. Army Corps of Engineers Buffalo District Office 1776 Niagara Street, Buffalo, New York, 14207

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA groupTuba City, Arizona, DisposalFourthN V O'1repository designd ' . ,$!

  13. DOE/OR/20722-133 POST-REMEDIAL ACTION REPORT FOR THE NIAGARA FALLS STORAGE SITE

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA groupTuba City, Arizona, DisposalFourthNrr-osams ADMIN RCDBaseline0

  14. RESULTS OF RADIOLOGICAL I'IEASUREMENTS HIGHT{AYS 18 AI.ID IO4 IN NIAGARA

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA groupTuba City, Arizona,Site Operations Guide Doc.5 R A D I O L O G I C A L9s'

  15. RESULTS OF RADIOLOGICAL MEASUREMENTS TAKEN NEAR JUNCTION OF HIGHWAY 3I AND MILITARY ROAD IN NIAGARA FALLSI NEI{ YOR

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA groupTuba City, Arizona,Site Operations Guide Doc.5 R A D I O L O G I C A L9s'7At

  16. Real Time Pricing and the Real Live Firm

    SciTech Connect (OSTI)

    Moezzi, Mithra; Goldman, Charles; Sezgen, Osman; Bharvirkar, Ranjit; Hopper, Nicole

    2004-05-26T23:59:59.000Z

    Energy economists have long argued the benefits of real time pricing (RTP) of electricity. Their basis for modeling customers response to short-term fluctuations in electricity prices are based on theories of rational firm behavior, where management strives to minimize operating costs and optimize profit, and labor, capital and energy are potential substitutes in the firm's production function. How well do private firms and public sector institutions operating conditions, knowledge structures, decision-making practices, and external relationships comport with these assumptions and how might this impact price response? We discuss these issues on the basis of interviews with 29 large (over 2 MW) industrial, commercial, and institutional customers in the Niagara Mohawk Power Corporation service territory that have faced day-ahead electricity market prices since 1998. We look at stories interviewees told about why and how they respond to RTP, why some customers report that they can't, and why even if they can, they don't. Some firms respond as theorized, and we describe their load curtailment strategies. About half of our interviewees reported that they were unable to either shift or forego electricity consumption even when prices are high ($0.50/kWh). Reasons customers gave for why they weren't price-responsive include implicit value placed on reliability, pricing structures, lack of flexibility in adjusting production inputs, just-in-time practices, perceived barriers to onsite generation, and insufficient time. We draw these observations into a framework that could help refine economic theory of dynamic pricing by providing real-world descriptions of how firms behave and why.

  17. Planning guidance for nuclear-power-plant decontamination. [PWR; BWR

    SciTech Connect (OSTI)

    Munson, L.F.; Divine, J.R.; Martin, J.B.

    1983-06-01T23:59:59.000Z

    Direct and indirect costs of decontamination are considered in the benefit-cost analysis. A generic form of the benefit-cost ratio is evaluated in monetary and nonmonetary terms, and values of dollar per man-rem are cited. Federal and state agencies that may have jurisiction over various aspects of decontamination and waste disposal activities are identified. Methods of decontamination, their general effectiveness, and the advantages and disadvantages of each are outlined. Dilute or concentrated chemical solutions are usually used in-situ to dissolve the contamination layer and a thin layer of the underlying substrate. Electrochemical techniques are generally limited to components but show high decontamination effectiveness with uniform corrosion. Mechanical agents are particularly appropriate for certain out-of-system surfaces and disassembled parts. These processes are catagorized and specific concerns are discussed. The treatment, storage, and disposal or discharge or discharge of liquid, gaseous, and solid wastes generated during the decontamination process are discussed. Radioactive and other hazardous chemical wastes are considered. The monitoring, treatment, and control of radioactive and nonradioactive effluents, from both routine operations and possible accidents, are discussed. Protecting the health and safety of personnel onsite during decontamination is of prime importance and should be considered in each facet of the decontamination process. The radiation protection philosophy of reducing exposure to levels as low as reasonably achievable should be stressed. These issues are discussed.

  18. Fuel performance annual report for 1981. [PWR; BWR

    SciTech Connect (OSTI)

    Bailey, W.J.; Tokar, M.

    1982-12-01T23:59:59.000Z

    This annual report, the fourth in a series, provides a brief description of fuel performance during 1981 in commercial nuclear power plants. Brief summaries of fuel operating experience, fuel problems, fuel design changes and fuel surveillance programs, and high-burnup fuel experience are provided. References to additional, more detailed information and related NRC evaluations are included.

  19. Critical discharge of initially subcooled water through slits. [PWR; BWR

    SciTech Connect (OSTI)

    Amos, C N; Schrock, V E

    1983-09-01T23:59:59.000Z

    This report describes an experimental investigation into the critical flow of initially subcooled water through rectangular slits. The study of such flows is relevant to the prediction of leak flow rates from cracks in piping, or pressure vessels, which contain sufficient enthalpy that vaporization will occur if they are allowed to expand to the ambient pressure. Two new analytical models, which allow for the generation of a metastable liquid phase, are developed. Experimental results are compared with the predictions of both these new models and with a Fanno Homogeneous Equilibrium Model.

  20. Integrated TRAC/MELPROG analyses of a PWR station blackout

    SciTech Connect (OSTI)

    Henninger, R.; Dearing, J.F.

    1987-01-01T23:59:59.000Z

    The first complete, coupled, and largely mechanistic analysis of the entire reactor-coolant system during a station blackout (TMLB') core-meltdown accident has been made with MELPROG/TRAC. The calculation was initiated at the start of the transient and ended with a late recovery of cooling. Additional cooling provided by water from the primary system delayed events relative to a standalone MELPROG calculation. Natural circulation within the vessel was established and primary-relief-valve action did little to disturb this flow. In addition, it was calculated directly that the hot leg reached a failure temperature long before vessel failure. Beyond relocation of the core, we have calculated the boiloff of the water in the lower head and have estimated the time of vessel failure to be at about 14,700 s into the transient. For ''nominal'' corium-water heat transfer, the boiloff process (steam-production rate) is slow enough that the relief valves prevent pressurization beyond 17.5 MPa. Parametric cases with increased corium-water heat transfer resulted in steaming rates beyond the capability of the relief valves, leading to pressures in excess of 19.2 MPa. Natural convection flow around the loop, if started by removing the water in the loop seal, was blocked by a relatively less-dense hydrogen/steam mixture that flowed to the top of the steam generator. Emergency core-cooling system activation late in the transient (after core slump) resulted in rapid cooling of the periphery of the debris region but slower cooling in the interior regions because of poor water penetration.

  1. Component failures that lead to reactor scrams. [PWR; BWR

    SciTech Connect (OSTI)

    Burns, E. T.; Wilson, R. J.; Lim, E. Y.

    1980-04-01T23:59:59.000Z

    This report summarizes the operating experience scram data compiled from 35 operating US light water reactors (LWRs) to identify the principal components/systems related to reactor scrams. The data base utilized to identify the scram causes is developed from a EPRI-utility sponsored survey conducted by SAI coupled with recent data from the USNRC Gray Books. The reactor population considered in this evaluation is limited to 23 PWRs and 12 BWRs because of the limited scope of the program. The population includes all the US NSSS vendors. It is judged that this population accurately characterizes the component-related scrams in LWRs over the first 10 years of plant operation.

  2. Hydrogen production during fragmented debris/concrete interactions. [PWR; BWR

    SciTech Connect (OSTI)

    Tarbell, W.W.; Blose, R.E.

    1982-01-01T23:59:59.000Z

    In the unlikely event that molten core debris escapes the reactor pressure vessel, the interactions of the debris with concrete and structural materials become the driving forces for severe accident phenomena. The Ex-vessel Core Debris Interactions Program at Sandia Laboratories is a research effort to characterize the nature of these interactions and the magnitude of safety-related phenomena such as hydrogen generation, aerosol production, and fission product release that arise because of the melt/concrete interactions.

  3. analysis program pwr: Topics by E-print Network

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

    refinement and modification is needed to extend it to thread creation Mller-Olm, Markus 219 Precise FixpointBased Analysis of Programs with ThreadCreation and...

  4. Support arrangements for core modules of nuclear reactors. [PWR

    DOE Patents [OSTI]

    Bollinger, L.R.

    1983-11-03T23:59:59.000Z

    A support arrangement is provided for the core modules of a nuclear reactor which provides support access through the control drive mechanisms of the reactor. This arrangement provides axial support of individual reactor core modules from the pressure vessel head in a manner which permits attachment and detachment of the modules from the head to be accomplished through the control drive mechanisms after their leadscrews have been removed. The arrangement includes a module support nut which is suspended from the pressure vessel head and screw threaded to the shroud housing for the module. A spline lock prevents loosening of the screw connection. An installation tool assembly, including a cell lifting and preloading tool and a torquing tool, fits through the control drive mechanism and provides lifting of the shroud housing while disconnecting the spline lock, as well as application of torque to the module support nut.

  5. Nuclear reactor cooling system decontamination reagent regeneration. [PWR; BWR

    DOE Patents [OSTI]

    Anstine, L.D.; James, D.B.; Melaika, E.A.; Peterson, J.P. Jr.

    1980-06-06T23:59:59.000Z

    An improved method for decontaminating the coolant system of water-cooled nuclear power reactors and for regenerating the decontamination solution is described. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution.

  6. POWER LEVEL EFFECT IN A PWR ROD EJECTION ACCIDENT.

    SciTech Connect (OSTI)

    DIAMOND,D.J.; BROMLEY,B.P.; ARONSON,A.L.

    2002-10-07T23:59:59.000Z

    The purpose of this study is to determine the effect of the initial power level during a rod ejection accident (REA) on the ejected rod worth and the resulting energy deposition in the fuel. The model used is for the hot zero power (HZP) conditions at the end of a typical fuel cycle for the Three Mile Island Unit 1 pressurized water reactor. PARCS, a transient, three-dimensional, two-group neutron nodal diffusion code, coupled with its own thermal-hydraulics model, is used to perform both steady-state and transient simulations. The worth of an ejected control rod is affected by both power level, and the positions of control banks. As the power level is increased, the worth of a single central control rod tends to drop due to thermal-hydraulic feedback and control bank removal, both of which flatten the radial neutron flux and power distributions. Although the peak fuel pellet enthalpy rise during an REA will be greater for a given ejected rod worth at elevated initial power levels, it is more likely the HZP condition will cause a greater net energy deposition because an ejected rod will have the highest worth at HZP. Thus, the HZP condition can be considered the most conservative in a safety evaluation.

  7. INTERCOMPARISON OF RESULTS FOR A PWR ROD EJECTION ACCIDENT

    SciTech Connect (OSTI)

    DIAMOND,D.J.; ARONSON,A.; JO,J.; AVVAKUMOV,A.; MALOFEEV,V.; SIDOROV,V.; FERRARESI,P.; GOUIN,C.; ANIEL,S.; ROYER,M.E.

    1999-10-01T23:59:59.000Z

    This study is part of an overall program to understand the uncertainty in best-estimate calculations of the local fuel enthalpy during the rod ejection accident. Local fuel enthalpy is used as the acceptance criterion for this design-basis event and can also be used to estimate fuel damage for the purpose of determining radiological consequences. The study used results from neutron kinetics models in PARCS, BARS, and CRONOS2, codes developed in the US, the Russian Federation, and France, respectively. Since BARS uses a heterogeneous representation of the fuel assembly as opposed to the homogeneous representations in PARCS and CRONOS, the effect of the intercomparison was primarily to compare different intra-assembly models. Quantitative comparisons for core power, reactivity, assembly fuel enthalpy and pin power were carried out. In general the agreement between methods was very good providing additional confidence in the codes and providing a starting point for a quantitative assessment of the uncertainty in calculated fuel enthalpy using best-estimate methods.

  8. accident consequences pwr: Topics by E-print Network

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

    Denmark December 1991 12;Abstract. A computer model of a simplified pressurized nuclear power plant a compute simulation of a simplified pressurized nuclear power plant model...

  9. accidents pwr bwr: Topics by E-print Network

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

    Denmark December 1991 12;Abstract. A computer model of a simplified pressurized nuclear power plant a compute simulation of a simplified pressurized nuclear power plant model...

  10. air primer pwr: Topics by E-print Network

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

    Denmark December 1991 12;Abstract. A computer model of a simplified pressurized nuclear power plant a compute simulation of a simplified pressurized nuclear power plant model...

  11. advanced pwr core: Topics by E-print Network

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

    Denmark December 1991 12;Abstract. A computer model of a simplified pressurized nuclear power plant a compute simulation of a simplified pressurized nuclear power plant model...

  12. advanced pwr fuel: Topics by E-print Network

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

    Denmark December 1991 12;Abstract. A computer model of a simplified pressurized nuclear power plant a compute simulation of a simplified pressurized nuclear power plant model...

  13. analysis pwr bwr: Topics by E-print Network

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

    Denmark December 1991 12;Abstract. A computer model of a simplified pressurized nuclear power plant a compute simulation of a simplified pressurized nuclear power plant model...

  14. An expert system for PWR core operation management

    SciTech Connect (OSTI)

    Ida, Toshio; Masuda, Masahiro; Nishioka, Hiromasa

    1988-01-01T23:59:59.000Z

    Planning for restartup after planned or unplanned reactor shutdown and load-follow operations is an important task in the core operation management of pressurized water reactors (PWRs). These planning problems have been solved by planning experts using their expertise and the computational prediction of core behavior. Therefore, the quality of the plan and the time consumed in the planning depend heavily on the skillfulness of the planning experts. A knowledge engineering approach has been recently considered as a promising means to solve such complicated planning problems. Many knowledge-based systems have been developed so far, and some of them have already been applied because of their effectiveness. The expert system REPLEX has been developed to aid core management engineers in making a successful plan for the restartup or the load-follow operation of PWRs within a shorter time. It can maintain planning tasks at a high-quality level independent of the skillfulness of core management engineers and enhance the efficiency of management. REPLEX has an explanation function that helps user understanding of plans. It could be a useful took, therefore, for the training of core management engineers.

  15. PWR core monitoring and simulation during load follow operation

    SciTech Connect (OSTI)

    Beard, C. (Westinghouse Electric Corp., Pittsburgh, PA (USA). Commercial Nuclear Fuel Div.); Winter, M.; Niederer, R. (Commonwealth Edison Co., Zion, IL (USA))

    1989-01-01T23:59:59.000Z

    This paper presents a new operation core support system developed for pressurized water reactors. This system provides an enhanced understanding of the operating core with significant benefits in operational flexibility. It also permits evaluation of alternatives and specific situations that allows for enhanced operation of the unit, which provides benefits in power capability and minimizes potential operational issues.

  16. Reactor physics assessment of thick silicon carbide clad PWR fuels

    E-Print Network [OSTI]

    Bloore, David A. (David Allan)

    2013-01-01T23:59:59.000Z

    High temperature tolerance, chemical stability and low neutron affinity make silicon carbide (SiC) a potential fuel cladding material that may improve the economics and safety of light water reactors (LWRs). "Thick" SiC ...

  17. CONTAIN: recent highlights in code testing and validation. [PWR; BWR

    SciTech Connect (OSTI)

    Murata, K.K.; Bergeron, K.D.; Rexroth, P.E.; Clauser, M.J.; Tills, J.L.; Sciacca, F.W.; Senglaub, M.E.; Trebilcock, W.; Williams, D.C.

    1983-01-01T23:59:59.000Z

    The CONTAIN code is currently being developed for LWR applications. The status of the LWR features and the code testing program will be briefly reviewed. New CONTAIN blind predictions for code validation experiments in the ABCOVE program and the HDR program will be presented, along with CONTAIN comparisons to an NSPP aerosol experiment in a steam environment and to a CSE spray experiment.

  18. Steam generator operating experience update, 1982-1983. [PWR

    SciTech Connect (OSTI)

    Frank, L.

    1984-06-01T23:59:59.000Z

    This report is a continuation of earlier reports by the staff addressing pressurized water reactor steam generator operating experience. NUREG-0886, Steam Generator Tube Experience, published in February 1982 summarized experience in domestic and foreign plants through December 1981. This report summarizes steam generator operating experience in domestic plants for the years 1982 and 1983. Included are new problems encountered with secondary-side loose parts, sulfur-induced stress-assisted corrosion cracking, and flow-induced vibrational wear in the new preheater design steam generators. The status of Unresolved Safety Issues A3, A4, and A5 is also discussed.

  19. Wire wrapped fuel pin hexagonal arrays for PWR service

    E-Print Network [OSTI]

    Diller, Peter Ray

    2005-01-01T23:59:59.000Z

    This work contributes to the Hydride Fuels Project, a collaborative effort between UC Berkeley and MIT aimed at investigating the potential benefits of hydride fuel use in light water reactors (LWRs). Core design is ...

  20. Corrosion-product transport in PWR secondary systems

    SciTech Connect (OSTI)

    Sawochka, S.G.; Copley, S.E.; Pearl, W.L.

    1981-12-01T23:59:59.000Z

    Corrosion product concentration data for the secondary systems of ten pressurized water reactors were obtained employing integrating sampling techniques. Based on these data, mass transport relations were developed for iron, copper, and nickel and employed to evaluate the effect of parameters such as blowdown, condensate polishing, operating pH and oxygen ingress on corrosion product transport and the buildup of sludge in steam generators. Several generic system designs were considered to illustrate application to plant design for the purpose of minimizing corrosion product input to generating equipment.

  1. Nuclear power plant fire protection: philosophy and analysis. [PWR; BWR

    SciTech Connect (OSTI)

    Berry, D. L.

    1980-05-01T23:59:59.000Z

    This report combines a fire severity analysis technique with a fault tree methodology for assessing the importance to nuclear power plant safety of certain combinations of components and systems. Characteristics unique to fire, such as propagation induced by the failure of barriers, have been incorporated into the methodology. By applying the resulting fire analysis technique to actual conditions found in a representative nuclear power plant, it is found that some safety and nonsafety areas are both highly vulnerable to fire spread and impotant to overall safety, while other areas prove to be of marginal importance. Suggestions are made for further experimental and analytical work to supplement the fire analysis method.

  2. Minor Actinides Transmutation Scenario Studies in PWR with Innovative Fuels

    SciTech Connect (OSTI)

    Grouiller, J. P.; Boucher, L.; Golfier, H.; Dolci, F.; Vasile, A.; Youinou, G.

    2003-02-26T23:59:59.000Z

    With the innovative fuels (CORAIL, APA, MIX, MOX-UE) in current PWRs, it is theoretically possible to obtain different plutonium and minor actinides transmutation scenarios, in homogeneous mode, with a significant reduction of the waste radio-toxicity inventory and of the thermal output of the high level waste. Regarding each minor actinide element transmutation in PWRs, conclusions are : neptunium : a solution exists but the gain on the waste radio-toxicity inventory is not significant, americium : a solution exists but it is necessary to transmute americium with curium to obtain a significant gain, curium: Cm244 has a large impact on radiation and residual power in the fuel cycle; a solution remains to be found, maybe separating it and keeping it in interim storage for decay into Pu240 able to be transmuted in reactor.

  3. East Mississippi Elec Pwr Assn | Open Energy Information

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    Open Energy Info (EERE)

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  14. Crawfordsville Elec, Lgt & Pwr | Open Energy Information

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  15. Prentiss County Elec Pwr Assn | Open Energy Information

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    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov YouKizildere I Geothermal PwerPerkins County, Nebraska:Precourt Institute for Energy Efficiency JumpPrenova Inc

  16. Central Electric Pwr Coop, Inc | Open Energy Information

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    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directedAnnual Siteof EnergyInnovation inOpenadd: China DatangCentral El trica Anhanguera

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

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are8COaBulkTransmissionSitingProcess.pdfGetec AG Contracting JumpGove County,Texas: Energy ResourcesGrand Valley Rrl

  18. Northeast Missouri El Pwr Coop | Open Energy Information

    Open Energy Info (EERE)

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  19. North Central Public Pwr Dist | Open Energy Information

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  20. Michigan South Central Pwr Agy | Open Energy Information

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  1. South Mississippi El Pwr Assn | Open Energy Information

    Open Energy Info (EERE)

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  2. Stanton County Public Pwr Dist | Open Energy Information

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  3. Keosauqua Municipal Light & Pwr | Open Energy Information

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  7. Cuming County Public Pwr Dist | Open Energy Information

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  10. Vermont Yankee Nucl Pwr Corp | Open Energy Information

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  12. Wolverine Pwr Supply Coop, Inc | Open Energy Information

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  14. SUBMILLIMETER OPTICAL PROPERTIES OF HEXAGONAL BORON NITRIDE A. J. Gatesman, R. H. Giles and J. Waldman

    E-Print Network [OSTI]

    Massachusetts at Lowell, University of

    boron nitride was obtained in four grades (A, HP, M, M26) from The Carborundum Co. in Niagara Fall, NY

  15. ORNL/RASA-85/1 RESULTS OF THE II4OBILE GAMMA SCANNING ACTIVITIES IN NIAGARA FALLS, NEvl YORK AREA

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA groupTuba City, Arizona,Site Operations Guide Doc. No.GS05:or _^r a ' 'LJ4e?Ll

  16. Nordisk kernesikkerhedsforskning Norrnar kjarnryggisrannsknir

    E-Print Network [OSTI]

    ..................................................................................................................6 2.2. PWR TS

  17. Customer Strategies for Responding to Day-Ahead Market HourlyElectricity Pricing

    SciTech Connect (OSTI)

    Goldman, Chuck; Hopper, Nicole; Bharvirkar, Ranjit; Neenan,Bernie; Boisvert, Dick; Cappers, Peter; Pratt, Donna; Butkins, Kim

    2005-08-25T23:59:59.000Z

    Real-time pricing (RTP) has been advocated as an economically efficient means to send price signals to customers to promote demand response (DR) (Borenstein 2002, Borenstein 2005, Ruff 2002). However, limited information exists that can be used to judge how effectively RTP actually induces DR, particularly in the context of restructured electricity markets. This report describes the second phase of a study of how large, non-residential customers' adapted to default-service day-ahead hourly pricing. The customers are located in upstate New York and served under Niagara Mohawk, A National Grid Company (NMPC)'s SC-3A rate class. The SC-3A tariff is a type of RTP that provides firm, day-ahead notice of hourly varying prices indexed to New York Independent System Operator (NYISO) day-ahead market prices. The study was funded by the California Energy Commission (CEC)'s PIER program through the Demand Response Research Center (DRRC). NMPC's is the first and longest-running default-service RTP tariff implemented in the context of retail competition. The mix of NMPC's large customers exposed to day-ahead hourly prices is roughly 30% industrial, 25% commercial and 45% institutional. They have faced periods of high prices during the study period (2000-2004), thereby providing an opportunity to assess their response to volatile hourly prices. The nature of the SC-3A default service attracted competitive retailers offering a wide array of pricing and hedging options, and customers could also participate in demand response programs implemented by NYISO. The first phase of this study examined SC-3A customers' satisfaction, hedging choices and price response through in-depth customer market research and a Constant Elasticity of Substitution (CES) demand model (Goldman et al. 2004). This second phase was undertaken to answer questions that remained unresolved and to quantify price response to a higher level of granularity. We accomplished these objectives with a second customer survey and interview effort, which resulted in a higher, 76% response rate, and the adoption of the more flexible Generalized Leontief (GL) demand model, which allows us to analyze customer response under a range of conditions (e.g. at different nominal prices) and to determine the distribution of individual customers' response.

  18. articulate brachiopods argyrotheca: Topics by E-print Network

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

    characteristic cardinalia with inner hinge plates separated by fissure, Lockport, Niagara County, New York, 11420, X34 (new). 8 The University of Kansas Paleontological...

  19. articulate brachiopod terebratal: Topics by E-print Network

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

    characteristic cardinalia with inner hinge plates separated by fissure, Lockport, Niagara County, New York, 11420, X34 (new). 8 The University of Kansas Paleontological...

  20. QUANTITATIVE STUDIES OF THERMAL SHOCK IN CERAMICS BASED ON A NOVEL TEST TECHNIQUE

    E-Print Network [OSTI]

    Faber, K.T.

    2013-01-01T23:59:59.000Z

    Silicon Carbide Division The Carborundum Co. Niagara Falls,of this research provided by the Carborundum Co. (K.T.F. and

  1. Modeling Coupled Processes in Clay Formations for Radioactive Waste Disposal

    E-Print Network [OSTI]

    Liu, Hui-Hai

    2010-01-01T23:59:59.000Z

    Pressurized Water Reactor (PWR) used nuclear fuel. The firstrepository tunnels, the PWR type of used fuel is typicallyby the length of individual PWR fuel elements and the number

  2. From rational numbers to algebra: Separable contributions of decimal magnitude and relational understanding of fractions

    E-Print Network [OSTI]

    DeWolf, M; Bassok, M; Holyoak, KJ

    2015-01-01T23:59:59.000Z

    disseminated broadly. OTO fraction PWR NOTO fraction Decimala part- to-whole ratio (PWR) is the relation between theof relationships. The PWR is a conventional relationship for

  3. T? tunable porous silicon iron oxide nanocomposites for magnetic resonance imaging guided drug delivey

    E-Print Network [OSTI]

    Ananda Yogendran, Shalini

    2012-01-01T23:59:59.000Z

    > y) & (0 z)) if (sqrt(pwr(x-xpos[sphere],2) + pwr (y-ypos[sphere],2)+pwr( z-zpos[sphere], 2)) <=sphere_radius) { int

  4. Reactive Transport and Coupled THM Processes in Engineering Barrier Systems (EBS)

    E-Print Network [OSTI]

    Steefel, Carl

    2010-01-01T23:59:59.000Z

    Pressurized Water Reactor (PWR) used nuclear fuel. The firstrepository tunnels, the PWR type of used fuel is typicallyby the length of individual PWR fuel elements and the number

  5. A vascular access system (VAS) for preclinical models

    E-Print Network [OSTI]

    Berry-Pusey, Brittany Nan

    2012-01-01T23:59:59.000Z

    control system for the VAS include SIGNAL PWR GND SIGNALK7 K7 PWR GNDSIGNAL PWR GND SIGNAL ´ SHLD FWD REV ENC ENC GND DIR DIR

  6. A Technical Review of Non-Destructive Assay Research for the Characterization of Spent Nuclear Fuel Assemblies Being Conducted Under the US DOE NGSI - 11544

    E-Print Network [OSTI]

    Croft, S.

    2012-01-01T23:59:59.000Z

    Determining Fissile Content in PWR Spent Fuel Assembliesalong the length of several PWR fuel rods (including somebeen studied for a wide range of PWR assembly cases and two

  7. Pwr fuel assembly optimization using adaptive simulated annealing coupled with translat

    E-Print Network [OSTI]

    Rogers, Timothy James

    2009-05-15T23:59:59.000Z

    assembly to be used in an operating nuclear power reactor. The two main cases of optimization are: one that finds the optimal 235U enrichment layout of the fuel pins in the assembly and another that finds both the optimal 235U enrichments where gadolinium...

  8. Effects of upper-plenum steam condensation phenomena on heat transfer in a rod bundle. [PWR

    SciTech Connect (OSTI)

    Chon, W.Y.; Addabbo, C.; Liao, N.S.

    1980-02-01T23:59:59.000Z

    System performance and thermohydraulic response to simultaneous bottom and top water injection were investigated in a 3 x 3 rod bundle Reflood Test Facility. An extensive series of tests, encompassing both simple bottom and combined injection reflooding, were carried out. A number of phenomenological events governing the thermodynamic coupling between the bottom reflood updraft and the top deluge were identified. Due to the countercurrent motion of the upflowing steam and water injected in the upper plenum counter current flow limiting phenomena hindered the penetration of water from inventory in the upper plenum into the bundle section. Consequently, condensation phenomena in the upper plenum and in the venting pipework characterized the thermohydraulic response of the bundle to simultaneous bottom and top water injection.

  9. Monitoring system for a liquid-cooled nuclear fission reactor. [PWR

    DOE Patents [OSTI]

    DeVolpi, A.

    1984-07-20T23:59:59.000Z

    The invention provides improved means for detecting the water levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting the density of the water in these regions. The invention utilizes a plurality of exterior gamma radiation detectors and a collimator technique operable to sense separate regions of the reactor vessel to give respectively, unique signals for these regions, whereby comparative analysis of these signals can be used to advise of the presence and density of cooling water in the vessel.

  10. Dynamic system characterization of an integral test facility of an advanced PWR

    E-Print Network [OSTI]

    Smith, Simon Gregory

    1995-01-01T23:59:59.000Z

    gives: P = pph&+p gh + p RT Differentiating with respect to time leads to, dp dp/ dhf dp dh dp gh. + p g ? + ? gh + p g ? s+ ? sRT+ p R? dt dt t / dt d? s dt dt t dt For a fixed tank with area A, -dhf/dt can be substituted for dhs/dt, and (H - hf...) for hs, dp dp/ dh/ dp dh& dp dT gh + pg ? + ? sg(H ? h) ? p g ? + ? RT+ p R? dt dt / /g dt dt / s dt dt & dt (] 2) 16 Since pt is approximately constant, or changes very slowly compared to other dynamic changes in the system: dpf Substituting...

  11. CFD Analysis of Nuclear Fuel Bundles and Spacer Grids for PWR Reactors

    E-Print Network [OSTI]

    Capone, Luigi

    2012-10-19T23:59:59.000Z

    The analysis of the turbulent flows in nuclear fuel bundles is a very interesting task to optimize the efficiency of modern nuclear power plants. The proposed study utilizes Computational Fluid Dynamics (CFD) to characterize the flow pattern...

  12. Design and fuel management of PWR cores to optimize the once-through fuel cycle

    E-Print Network [OSTI]

    Fujita, Edward Kei

    The once-through fuel cycle has been analyzed to see if there are substantial prospects for improved uranium ore utilization in current

  13. Dose reduction and optimization studies (ALARA) at nuclear power facilities. [PWR; BWR

    SciTech Connect (OSTI)

    Baum, J.W.; Meinhold, C.B.

    1983-01-01T23:59:59.000Z

    Brookhaven National Laboratory (BNL) has been commissioned by the Nuclear Regulatory Commission (NRC) to study dose-reduction techniques and effectiveness of as low as reasonably achievable (ALARA) planning at LWR plants. These studies have the following objectives: identify high-dose maintenance tasks; identify dose-reduction techniques; examine incentives for dose reduction; evaluate cost-effectiveness and optimization of dose-reduction techniques; and compile an ALARA handbook on data, engineering modifications, cost-effectiveness calculations, and other information of interest to ALARA practioners.

  14. THE CALCULATION OF BURNABLE POISON CORRECTION FACTORS FOR PWR FRESH FUEL ACTIVE COLLAR MEASUREMENTS

    SciTech Connect (OSTI)

    Croft, Stephen [Los Alamos National Laboratory; Favalli, Andrea [Los Alamos National Laboratory; Swinhoe, Martyn T. [Los Alamos National Laboratory

    2012-06-19T23:59:59.000Z

    Verification of commercial low enriched uranium light water reactor fuel takes place at the fuel fabrication facility as part of the overall international nuclear safeguards solution to the civilian use of nuclear technology. The fissile mass per unit length is determined nondestructively by active neutron coincidence counting using a neutron collar. A collar comprises four slabs of high density polyethylene that surround the assembly. Three of the slabs contain {sup 3}He filled proportional counters to detect time correlated fission neutrons induced by an AmLi source placed in the fourth slab. Historically, the response of a particular collar design to a particular fuel assembly type has been established by careful cross-calibration to experimental absolute calibrations. Traceability exists to sources and materials held at Los Alamos National Laboratory for over 35 years. This simple yet powerful approach has ensured consistency of application. Since the 1980's there has been a steady improvement in fuel performance. The trend has been to higher burn up. This requires the use of both higher initial enrichment and greater concentrations of burnable poisons. The original analytical relationships to correct for varying fuel composition are consequently being challenged because the experimental basis for them made use of fuels of lower enrichment and lower poison content than is in use today and is envisioned for use in the near term. Thus a reassessment of the correction factors is needed. Experimental reassessment is expensive and time consuming given the great variation between fuel assemblies in circulation. Fortunately current modeling methods enable relative response functions to be calculated with high accuracy. Hence modeling provides a more convenient and cost effective means to derive correction factors which are fit for purpose with confidence. In this work we use the Monte Carlo code MCNPX with neutron coincidence tallies to calculate the influence of Gd{sub 2}O{sub 3} burnable poison on the measurement of fresh pressurized water reactor fuel. To empirically determine the response function over the range of historical and future use we have considered enrichments up to 5 wt% {sup 235}U/{sup tot}U and Gd weight fractions of up to 10 % Gd/UO{sub 2}. Parameterized correction factors are presented.

  15. Behavior of triplex silicon carbide fuel cladding designs tested under simulated PWR conditions

    E-Print Network [OSTI]

    Stempien, John D. (John Dennis)

    2011-01-01T23:59:59.000Z

    A silicon carbide (SiC) fuel cladding for LWRs may allow a number of advances, including: increased safety margins under transients and accident scenarios, such as loss of coolant accidents; improved resource utilization ...

  16. Attack of fragmented-core debris on concrete in the presence of water. [PWR; BWR

    SciTech Connect (OSTI)

    Tarbell, W.W.; Bradley, D.R.

    1982-01-01T23:59:59.000Z

    In the unlikely event that core debris escapes the reactor pressure vessel, the interactions of the debris with concrete, structural materials, and coolant become the driving force for severe accident phenomena. The Ex-Vessel Core Debris Interactions Program at Sandia National Laboratories is an experimental research effort to characterize these interactions and the magnitude of safety-related phenomena such as flammable gas generation, aerosol production, fission product release, and concrete attack. Major areas of study within the program include molten core simultants in contact with concrete, high pressure melt streaming into scaled reactor cavities, the addition of coolant to high-temperature melt/concrete interactions, and the attack of hot, solid core debris on concrete. This paper describes results from the last of these efforts, i.e., hot, but not molten debris attacking concrete.

  17. Aerosol source term in high-pressure-melt ejection. [PWR; BWR

    SciTech Connect (OSTI)

    Brockmann, J.E.; Tarbell, W.W.

    1983-01-01T23:59:59.000Z

    Pressurized ejection of melt from a reactor pressure vessel has been identified as an important element of a severe reactor accident. Copious aerosol production is observed when thermitically generated melts pressurized with nitrogen or carbon dioxide to 1.3 to 17 MPa are ejected into an air atmosphere. Aerosol particle size distributions measured in the tests have modes of about 0.5, 5, and > 10..mu..m. Mechanisms leading to formation of these multimodal size distributions are suggested. This aerosol is a potentially important fission product source term which has not been considered in previous severe accident analyses.

  18. Aging considerations for PWR (pressurized water reactor) control rod drive mechanisms and reactor internals

    SciTech Connect (OSTI)

    Ware, A.G.

    1988-01-01T23:59:59.000Z

    This paper describes age-related degradation mechanisms affecting life extension of pressurized water reactor control rod drive mechanisms and reactor internals. The major sources of age-related degradation for control rod drive mechanisms are thermal transients such as plant heatups and cooldowns, latchings and unlatchings, long-term aging effects on electrical insulation, and the high temperature corrosive environment. Flow induced loads, the high-temperature corrosive environment, radiation exposure, and high tensile stresses in bolts all contribute to aging related degradation of reactor internals. Another problem has been wear and fretting of instrument guide tubes. The paper also discusses age-related failures that have occurred to date in pressurized water reactors.

  19. Comparison of nuclear data uncertainty propagation methodologies for PWR burn-up simulations

    E-Print Network [OSTI]

    Diez, Carlos Javier; Hoefer, Axel; Porsch, Dieter; Cabellos, Oscar

    2014-01-01T23:59:59.000Z

    Several methodologies using different levels of approximations have been developed for propagating nuclear data uncertainties in nuclear burn-up simulations. Most methods fall into the two broad classes of Monte Carlo approaches, which are exact apart from statistical uncertainties but require additional computation time, and first order perturbation theory approaches, which are efficient for not too large numbers of considered response functions but only applicable for sufficiently small nuclear data uncertainties. Some methods neglect isotopic composition uncertainties induced by the depletion steps of the simulations, others neglect neutron flux uncertainties, and the accuracy of a given approximation is often very hard to quantify. In order to get a better sense of the impact of different approximations, this work aims to compare results obtained based on different approximate methodologies with an exact method, namely the NUDUNA Monte Carlo based approach developed by AREVA GmbH. In addition, the impact ...

  20. Modeling the performance of high burnup thoria and urania PWR fuel

    E-Print Network [OSTI]

    Long, Yun, 1972-

    2002-01-01T23:59:59.000Z

    Fuel performance models have been developed to assess the performance of ThO?-UO? fuels that can be operated to a high burnup up to 80-100MWd/kgHM in current and future Light Water Reactors (LWRs). Among the various issues ...

  1. Observation of the Isotopic Evolution of PWR Fuel Using an Antineutrino Detector

    E-Print Network [OSTI]

    Bowden, N S; Dazeley, S; Svoboda, R; Misner, A; Palmer, T

    2008-01-01T23:59:59.000Z

    By operating an antineutrino detector of simple design during several fuel cycles, we have observed long term changes in antineutrino flux that result from the isotopic evolution of a commercial pressurized water reactor. Measurements made with simple antineutrino detectors of this kind offer an alternative means for verifying fissile inventories at reactors, as part of IAEA and other reactor safeguards regimes.

  2. Liquid level, void fraction, and superheated steam sensor for nuclear-reactor cores. [PWR; BWR

    DOE Patents [OSTI]

    Tokarz, R.D.

    1981-10-27T23:59:59.000Z

    This disclosure relates to an apparatus for monitoring the presence of coolant in liquid or mixed liquid and vapor, and superheated gaseous phases at one or more locations within an operating nuclear reactor core, such as pressurized water reactor or a boiling water reactor.

  3. Status of verification and validation of AREVA's ARCADIA{sup R} code system for PWR applications

    SciTech Connect (OSTI)

    Porsch, D. [AREVA, AREVA NP GmbH (Germany); P.O.Box 1109, 91001 Erlangen (Germany); Leberig, M.; Kuch, S. [AREVA, AREVA NP GmbH (Germany); Magat, P. [AREVA, AREVA NP SAS, Paris (France); Segard, K. [AREVA, AREVA NP Inc., Lynchburg (United States)

    2012-07-01T23:59:59.000Z

    In March 2010 the submittal of Topical Reports for ARCADIA{sup R} and COBRA-FLX, the thermal-hydraulic module of ARCADIA{sup R}, to the U.S. Nuclear Regulatory Commission (NRC) concluded a major step in the development of AREVA's new code system for core design and safety analyses. This submittal was dedicated to the application of the code system to uranium fuel in pressurized water reactors. The submitted information comprised results for plants operated in the US (France)) and Germany and provided uncertainties for in-core measuring systems with traveling in-core detectors and for the aero-ball system of the EPR. A reduction of the uncertainties in the prediction of F{sub AH} and F{sub Q} of > 1 % (absolute) was derived compared to the current code systems. This paper extents the verification and validation base for uranium based fuel and demonstrates the basic capabilities of ARCADIA{sup R} of describing MOX. The achieved status of verification and validation is described in detail. All applications followed the same standard without any specific calibration. The paper gives also insight in the new capability of 3D full core steady-state and transient pin-by-pin/sub-channel-by-sub-channel calculations and the opportunities offered by this feature. The gain of margins with increasing detail of the representation is outlined. Currently, the strategies for worldwide implementation of ARCADIA{sup R} are developed. (authors)

  4. Effect of lithium hydroxide on zircaloy corrosion in the Ringhals-3 PWR plant

    SciTech Connect (OSTI)

    Polley, M.V.; Evans, H.E. (Nuclear Electric plc, Berkeley (United Kingdom). Berkeley Nuclear Labs.); Anderson, P.O.; Larson, J. (Statens Vattenfallsverk, Stockholm (Sweden))

    1992-03-01T23:59:59.000Z

    Zircaloy oxide thicknesses were predicted for several fuel rods irradiated in Ringhals 3 over cycles 1b-4. During most of this period the fuel cladding was exposed to a high pH primary coolant chemistry regime in which lithium was present up to a concentration of 3.5 ppm. Comparison of prediction with measurement showed that the presence of lithium had produced no enhancement in oxidation within the limits of experimental and computational error.

  5. Impact of PWR spent fuel variations on TRU-fueled VHTRS

    E-Print Network [OSTI]

    Alajo, Ayodeji Babatunde

    2009-05-15T23:59:59.000Z

    Several alternative strategies are being considered as spent nuclear fuel (SNF) management options. Transuranic nuclides (TRU) are responsible for the SNF long-term radiotoxicity beyond the first 500 years. One of the most viable approaches suggests...

  6. Technical specification action requirements for AFW system failures: Method development and application to four PWR plants

    SciTech Connect (OSTI)

    Mankamo, T. [Avaplan Oy, Espoo (Finland); Kim, I.S.; Yang, Ji Wu; Samanta, P.K. [Brookhaven National Lab., Upton, NY (United States)

    1996-09-01T23:59:59.000Z

    Failures in the auxiliary feedwater (AFW) system of pressurized water reactors (PWRs) are considered to involve substantial risk whether a decision is made to either continue power operation while repair is being done, or to shut down the plant to undertake repairs. Technical specification action requirements usually require immediate plant shutdown in the case of multiple failures in the system (in some cases, immediate repair of one train is required when all AFW trains fail). This paper presents a probabilistic risk assessment-based method to quantitatively evaluate and compare both the risks of continued power operation and of shutting the plant down, given known failures in the system. The method is applied to the AFW system for four different PWRs. Results show that the risk of continued power operation and plant shutdown both are substantial, but the latter is larger than the former over the usual repair time. This was proven for four plants with different designs: two operating Westinghouse plants, one operating Asea-Brown Boveri Combustion Engineering Plant, and one of evolutionary design. The method can be used to analyze individual plant design and to improve AFW action requirements using risk-informed evaluations.

  7. Institutional implications of establishing safety goals for nuclear power plants. [PWR; BWR

    SciTech Connect (OSTI)

    Morris, F.A.; Hooper, R.L.

    1983-07-01T23:59:59.000Z

    The purpose of this project is to anticipate and address institutional problems that may arise from the adoption of NRC's proposed Policy Statement on Safety Goals for Nuclear Power Plants. The report emphasizes one particular category of institutional problems: the possible use of safety goals as a basis for legal challenges to NRC actions, and the resolution of such challenges by the courts. Three types of legal issues are identified and analyzed. These are, first, general legal issues such as access to the legal system, burden of proof, and standard of proof. Second is the particular formulation of goals. Involved here are such questions as sustainable rationale, definitions, avoided issues, vagueness of time and space details, and degree of conservatism. Implementation brings up the third set of issues which include interpretation and application, linkage to probabilistic risk assessment, consequences as compared to events, and the use of results.

  8. Structure/piping sensitivity studies for the US NRC Seismic Safety Margins Research Program. [PWR; BWR

    SciTech Connect (OSTI)

    Shieh, L.C.; O'Connell, W.J.; Johnson, J.J.

    1983-01-01T23:59:59.000Z

    The Seismic Safety Margins Research Program (SSMRP) is a NRC-funded, multi-year program conducted by Lawrence Livermore National Laboratory (LLNL). One of the goals of the program is to develop a complete, fully coupled analysis procedure (including methods and computer codes) for estimating the risk of an earthquake-caused radioactive release from commercial nuclear power plant. The analysis procedure is based upon a state-of-the-art evaluation of the current seismic analysis design process and explicitly includes the uncertainties inherent in such a process. The results will be used to improve seismic licensing requirements for nuclear power plants. In Phase I, a probabilistic computational procedure was developed for the seismic safety assessment. In Phase II, sensitivity studies were performed, codes and models were improved, and analysis of the Zion plant was completed.

  9. TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

    SciTech Connect (OSTI)

    Joy L. Rempe; Darrell L. Knudson

    2014-05-01T23:59:59.000Z

    The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation process that could be implemented in upcoming Fukushima Daiichi recovery efforts.

  10. TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

    SciTech Connect (OSTI)

    Joy L. Rempe; Darrell L. Knudson

    2013-03-01T23:59:59.000Z

    The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation process that could be implemented in upcoming Fukushima Daiichi recovery efforts.

  11. Multi-Recycling of Transuranic Elements in a Modified PWR Fuel Assembly 

    E-Print Network [OSTI]

    Chambers, Alex

    2012-10-19T23:59:59.000Z

    The nuclear waste currently generated in the United States is stored in spent fuel pools and dry casks throughout the country awaiting a permanent disposal solution. One efficient solution would be to remove the actinides from the waste...

  12. CFD Analysis of Nuclear Fuel Bundles and Spacer Grids for PWR Reactors 

    E-Print Network [OSTI]

    Capone, Luigi

    2012-10-19T23:59:59.000Z

    The analysis of the turbulent flows in nuclear fuel bundles is a very interesting task to optimize the efficiency of modern nuclear power plants. The proposed study utilizes Computational Fluid Dynamics (CFD) to characterize the flow pattern...

  13. Reactor physics considerations for implementing silicon carbide cladding into a PWR environment

    E-Print Network [OSTI]

    Dobisesky, Jacob P. (Jacob Paul), 1987-

    2011-01-01T23:59:59.000Z

    Silicon carbide (SiC) offers several advantages over zirconium (Zr)-based alloys as a potential cladding material for Pressurized Water Reactors: very slow corrosion rate, ability to withstand much higher temperature with ...

  14. Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation - Vandellos II Reactor

    SciTech Connect (OSTI)

    Ilas, Germina [ORNL; Gauld, Ian C [ORNL

    2011-01-01T23:59:59.000Z

    This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.

  15. The design of a compact integral medium size PWR : the CIRIS

    E-Print Network [OSTI]

    Shirvan, Koroush

    2010-01-01T23:59:59.000Z

    The International Reactor Innovative and Secure (IRIS) is an advanced medium size, modular integral light water reactor design, rated currently at 1000 MWt. IRIS design has been under development by over 20 organizations ...

  16. A QUALITATIVE APPROACH TO UNCERTAINTY ANALYSIS FOR THE PWR ROD EJECTION ACCIDENT

    SciTech Connect (OSTI)

    DIAMOND,D.J.; ARONSON,A.; YANG,C.

    2000-06-19T23:59:59.000Z

    In order to understand best-estimate calculations of the peak local fuel enthalpy during a rod ejection accident, an assessment of the uncertainty has been completed. The analysis took into account point kinetics parameters which would be available from a three-dimensional core model and engineering judgment as to the uncertainty in those parameters. Sensitivity studies to those parameters were carried out using the best-estimate code PARCS. The results showed that the uncertainty (corresponding to one standard deviation) in local fuel enthalpy would be determined primarily by the uncertainty in ejected rod worth and delayed neutron fraction. For an uncertainty in the former of 8% and the latter of 5%, the uncertainty in fuel enthalpy varied from 51% to 69% for control rod worth varying from $1.2 to $1.0. Also considered in the uncertainty were the errors introduced by uncertainties in the Doppler reactivity coefficient, the fuel pellet specific heat, and assembly and fuel pin peaking factors.

  17. Prediction of departure from nucleate boiling in PWR fast power transients

    E-Print Network [OSTI]

    Lenci, Giancarlo

    2013-01-01T23:59:59.000Z

    An assessment is conducted of the differences in predicted results between use of steady state versus transient Departure from Nucleate Boiling (DNB) models, for fast power transients under forced convective heat exchange ...

  18. Analysis of molten debris freezing and wall erosion during a severe RIA test. [PWR; BWR

    SciTech Connect (OSTI)

    El-Genk, M.S.; Moore, R.L.

    1980-01-01T23:59:59.000Z

    A one-dimensional physical model was developed to study the transient freezing of the molten debris layer (a mixture of UO/sub 2/ fuel and zircaloy cladding) produced in a severe reactivity initiated accident in-pile test and deposited on the inner surface of the test shroud wall. The wall had a finite thickness and was cooled along its outer surface by coolant bypass flow. Analyzed are the effects of debris temperature, radiation cooling at the debris layer surface, zircaloy volume ratio within the debris, and initial wall temperature on the transient freezing of the debris layer and the potential melting of the wall. The governing equations of this two-component, simultaneous freezing and melting problem in a finite geometry were solved using a one-dimensional finite element code based on the method of weighted residuals.

  19. annular-dispersed flow pwr: Topics by E-print Network

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

    Denmark December 1991 12;Abstract. A computer model of a simplified pressurized nuclear power plant a compute simulation of a simplified pressurized nuclear power plant model...

  20. alarm-p1 pwr thermohydraulics: Topics by E-print Network

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

    Denmark December 1991 12;Abstract. A computer model of a simplified pressurized nuclear power plant a compute simulation of a simplified pressurized nuclear power plant model...

  1. Extension of load follow capability of a PWR reactor by optimal control

    SciTech Connect (OSTI)

    Winokur, M.; Tepper, L.

    1984-04-01T23:59:59.000Z

    The problem of extending that part of the fuel life cycle during which a reactor is capable of sustaining load-follow operation is formulated as an optimal control problem. A two-node model representation of pressurized water reactor dynamics is used, leading to a set of non-linear ordinary differential equations. Differential Dynamic Programming is used to solve directly the resulting nonlinear optimization problem and obtain the trajectories of soluble boron concentration and control rod insertion. Results of computations performed for a reference reactor are presented, showing how the optimal control policy stretches the capability of the reactor to follow an average daily load curve towards the end of the fuel life cycle.

  2. Analysis of conventional and plutonium recycle unit-assemblies for the Yankee (Rowe) PWR

    E-Print Network [OSTI]

    Mertens, Paul Gustaaf

    1971-01-01T23:59:59.000Z

    An analysis and comparison of Unit Conventional UO2 Fuel-Assemblies and proposed Plutonium Recycle Fuel Assemblies for the Yankee (Rowe) Reactor has been made. The influence of spectral effects, at the watergaps -and ...

  3. ANALYSIS OF PWR SBO CAUSED BY EXTERNAL FLOODING USING THE RISMC TOOLKIT

    SciTech Connect (OSTI)

    Mandelli, Diego; Smith, Curtis; Prescott, Steven; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua; Kinoshita, Robert

    2014-08-01T23:59:59.000Z

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impacts of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization project aims to provide insights to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This paper focuses on the impacts of power uprate on the safety margin of a boiling water reactor for a flooding induced station black-out event. Analysis is performed by using a combination of thermal-hydraulic codes and a stochastic analysis tool currently under development at the Idaho National Laboratory, i.e. RAVEN. We employed both classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. Results obtained give a detailed investigation of the issues associated with a plant power uprate including the effects of station black-out accident scenarios. We were able to quantify how the timing of specific events was impacted by a higher nominal reactor core power. Such safety insights can provide useful information to the decision makers to perform risk informed margins management.

  4. Identification and Resolution of Safety Issues for the Advanced Integral Type PWR

    SciTech Connect (OSTI)

    Kim, Woong Sik; Jo, Jong Chull; Yune, Young Gill; Kim, Hho Jung [Korea Institute of Nuclear Safety, 19 Kusung-dong, Yusung-ku, Taejon, 305-338 (Korea, Republic of)

    2004-07-01T23:59:59.000Z

    This paper presents the interim results of a study on the identification and resolution of safety issues for the AIPWR licensing. The safety issues discussed in this paper include (1) policy issues for which decision-makings are needed for the procedural requirements of licensing system in the regulatory policy point of view, (2) technical issues for which either development of new requirements or amendment of some existing requirements is needed, or (3) other technical issues for which safety verifications are required. The study covers (a) the assessment of applicability of the issues identified from the previous studies to the case of the AIPWR, (b) identification of safety issues through analysis of the international experiences in the design and licensing of advanced reactors, and technical review of the AIPWR design, and (c) development of the resolutions of safety issues, and application of the resolutions to the amendment of regulatory requirements and the licensing review of the AIPWR. As the results of this study, a total of twenty eight safety issues was identified: fourteen issues from the previous studies, including the establishment of design safety goals; four issues from the foreign practices and experiences, including the risk-informed licensing; and ten issues by the AIPWR design review, including reliability of passive safety systems. Ten issues of them have been already resolved and the succeeding study is under way to resolve the remaining ones. (authors)

  5. Reactor Safety Research Programs. Quarterly report, July-September 1984. Volume 3. [PWR; BWR

    SciTech Connect (OSTI)

    Edler, S.K. (ed.)

    1985-02-01T23:59:59.000Z

    This document summarizes work performed by Pacific Northwest Laboratory from July 1 through September 30, 1984, for the Division of Accident Evaluation and the Division of Engineering Technology, US Nuclear Regulatory Commission. Results from an instrumented fuel assembly irradiation program being performed at Halden, Norway, are reported. Accelerated pellet-cladding interaction modeling is being conducted to predict the probability of fuel rod failure under normal operating conditions. Experimental data and analytical models are being provided to aid in decision making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Fuel assemblies and analytical support are being provided for experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory, Idaho Falls, Idaho. High-temperature materials property tests are being conducted to provide data on severe core damage fuel behavior. Thermal-hydraulic models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Severe fuel damage accident tests are being conducted in the NRU Reactor, Chalk River, Canada.

  6. Reactor safety research programs. Quarterly report, January-March 1984. Vol. 1. [PWR; BWR

    SciTech Connect (OSTI)

    Edler, S.K. (ed.)

    1984-06-01T23:59:59.000Z

    This document summarizes work performed by Pacific Northwest Laboratory from January 1 through March 31, 1984, for the Division of Accident Evaluation and the Division of Engineering Technology, US Nuclear Regulatory Commission. Results from an instrumented fuel assembly irradiation program being performed at Halden, Norway, are reported. Accelerated pellet-cladding interaction modeling is being conducted to predict the probability of fuel rod failure under normal operating conditions. Experimental data on analytical models are being provided to aid in decision making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Fuel assemblies and analytical support are being provided for experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory, Idaho Falls, Idaho. High-temperature materials property tests are being conducted to provide data on severe core damage fuel behavior. Thermal-hydraulic models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Severe fuel damage accident tests are being conducted at the NRU reactor, Chalk River, Canada.

  7. Dosimetry assessments for the reactor pressure vessel and core barrel in UK PWR plant

    SciTech Connect (OSTI)

    Thornton, D.A.; Allen, D.A.; Huggon, A.P.; Picton, D.J.; Robinson, A.T.; Steadman, R.J. [Serco, Rutherford House, Quedgeley, Gloucester, Gl2 4NF (United Kingdom); Seren, T.; Lipponen, M.; Kekki, T. [VTT, Technical Research Centre of Finland, Otakaari 3 K, P.O. BOX 1000, Espoo, FI-02044 (Finland)

    2011-07-01T23:59:59.000Z

    Specimens for the Sizewell B reactor pressure vessel (RPV) inservice steels surveillance program are irradiated inside eight capsules located within the reactor pressure vessel and loaded prior to commissioning. The periodic removal of these capsules and testing of their contents provides material properties data at intervals during the lifetime of the plant. Neutron activation measurements and radiation transport calculations play an essential role in assessing the neutron exposure of the specimens and RPV. Following the most recent withdrawal, seven capsules have now been removed covering nine cycles of reactor operation. This paper summarizes the dosimetry results of the Sizewell B surveillance program obtained to date. In addition to an overview of the calculational methodology it includes a review of the measurements. Finally, it describes an extension of the methodology to provide dosimetry recommendations for the core barrel and briefly discusses the results that were obtained. (authors)

  8. The analysis of normative requirements to materials of PWR components, basing on LBB concepts

    SciTech Connect (OSTI)

    Anikovsky, V.V.; Karzov, G.P.; Timofeev, B.T. [CRISM Prometey, St. Petersburg (Russian Federation)

    1997-04-01T23:59:59.000Z

    The paper discusses the advisability of the correction of Norms to solve in terms of material science the Problem: how the normative requirements to materials must be changed in terms of the concept {open_quotes}leak before break{close_quotes} (LBB).

  9. Impact of PWR spent fuel variations on TRU-fueled VHTRS 

    E-Print Network [OSTI]

    Alajo, Ayodeji Babatunde

    2009-05-15T23:59:59.000Z

    The VHTR is a graphite moderated helium-cooled reactor that supplies heat with core outlet temperatures equal to or greater than 850 degree Celsius. Its basic technology has been well established in High Temperature Gas Reactors (HTGR..., representing an attractive spent fuel form. The prismatic core configuration features annular fuel compacts composed of TRISO coated fuel particles that are embedded in graphite matrix. The fuel compacts are stacked together forming the fuel element...

  10. Comparison of nuclear data uncertainty propagation methodologies for PWR burn-up simulations

    E-Print Network [OSTI]

    Carlos Javier Diez; Oliver Buss; Axel Hoefer; Dieter Porsch; Oscar Cabellos

    2014-11-04T23:59:59.000Z

    Several methodologies using different levels of approximations have been developed for propagating nuclear data uncertainties in nuclear burn-up simulations. Most methods fall into the two broad classes of Monte Carlo approaches, which are exact apart from statistical uncertainties but require additional computation time, and first order perturbation theory approaches, which are efficient for not too large numbers of considered response functions but only applicable for sufficiently small nuclear data uncertainties. Some methods neglect isotopic composition uncertainties induced by the depletion steps of the simulations, others neglect neutron flux uncertainties, and the accuracy of a given approximation is often very hard to quantify. In order to get a better sense of the impact of different approximations, this work aims to compare results obtained based on different approximate methodologies with an exact method, namely the NUDUNA Monte Carlo based approach developed by AREVA GmbH. In addition, the impact of different covariance data is studied by comparing two of the presently most complete nuclear data covariance libraries (ENDF/B-VII.1 and SCALE 6.0), which reveals a high dependency of the uncertainty estimates on the source of covariance data. The burn-up benchmark Exercise I-1b proposed by the OECD expert group "Benchmarks for Uncertainty Analysis in Modeling (UAM) for the Design, Operation and Safety Analysis of LWRs" is studied as an example application. The burn-up simulations are performed with the SCALE 6.0 tool suite.

  11. Gas bubbling-enhanced film boiling of Freon-11 on liquid metal pools. [PWR; BWR

    SciTech Connect (OSTI)

    Greene, G.A.

    1985-01-01T23:59:59.000Z

    In the analysis of severe core damage accidents in LWRs, a major driving force which must be considered in evaluating containment loading and fission product transport is the ex-vessel interaction between molten core debris and structural concrete. Two computer codes have been developed for this purpose, the CORCON-MOD2 model of ex-vessel, core concrete interactions and the VANESA model for aerosol generation and fission product release as a result of molten core-concrete interactions. Under a wide spectrum of reactor designs and accident sequences, it is possible for water to come into contact with the molten core debris and form a coolant pool overlying the core debris which is attacking the concrete. As the concrete decomposes, noncondensable gases are released, which bubble through the melt and across the boiling interface, affecting the liquid-liquid boiling process. Currently, the CORCON code includes the classical Berenson model for film boiling over a horizontal flat plate for this phenomenon. The objectives of this activity are to investigate the influence of transverse noncondensable gas flux on the magnitude of the stable liquid-liquid film boiling heat flux and develop a gas flux-enhanced, liquid-liquid film boiling model for incorporation into the CORCON-MOD2 computer code to replace or modify the Berenson model.

  12. Aerosol release and transport program. Quarterly progress report, October-December 1981. [LMFBR; PWR; BWR

    SciTech Connect (OSTI)

    Adams, R. E.; Tobias, M. L.

    1982-05-01T23:59:59.000Z

    This report summarizes progress for the Aerosol Release and Transport Program sponsored by the Nuclear Regulatory Commission's Office of Nuclear Regulatory Research, Division of Accident Evaluation, for the period October-December 1981. Topics discussed include (1) under-sodium tests in the Fuel Aerosol Simulant Test (FAST) Facility, (2) U/sub 3/O/sub 8/ and Fe/sub 2/O/sub 3/ in steam (light-water reactor accident) aerosol experiments in the Nuclear Safety Pilot Plant, (3) generation and characterization of cadmium and CdO aerosols in the basic aerosol experimental program, (4) core-melt tests of Zircaloy-clad fuel capsules, (5) initial results of a piston-model bubble oscillation code allowing liquid bypass, and (6) calculations with the UVABUBL code to compare with underwater and under-sodium period measurements in FAST experiments.

  13. Three dimensional effects in analysis of PWR steam line break accident

    E-Print Network [OSTI]

    Tsai, Chon-Kwo

    A steam line break accident is one of the possible severe abnormal transients in a pressurized water reactor. It is required to present an analysis of a steam line break accident in the Final Safety Analysis Report (FSAR) ...

  14. Uranium resource utilization improvements in the once-through PWR fuel cycle

    SciTech Connect (OSTI)

    Matzie, R A [ed.

    1980-04-01T23:59:59.000Z

    In support of the Nonproliferation Alternative Systems Assessment Program (NASAP), Combustion Engineering, Inc. performed a comprehensive analytical study of potential uranium utilization improvement options that can be backfit into existing PWRs operating on the once-through uranium fuel cycle. A large number of potential improvement options were examined as part of a preliminary survey of candidate options. The most attractive of these, from the standpoint of uranium utilization improvement, economic viability, and ease of implementation, were then selected for detailed analysis and were included in a single composite improvement case. This composite case represents an estimate of the total savings in U/sub 3/O/sub 8/ consumption that can be achieved in current-design PWRs by implementing improvements which can be developed and demonstrated in the near term. The improvement options which were evaluated in detail and included in the composite case were a new five-batch, extended-burnup fuel management scheme, low-leakage fuel management, modified lattice designs, axial blankets, reinsertion of initial core batches, and end-of-cycle stretchout.

  15. The comparison of available data on PWR assembly thermal behavior with analytical predictions

    E-Print Network [OSTI]

    Liu, Jack S. H.

    The comparison of available data with analytical predictions has been illustrated in this report. Since few data on the cross flow are available, a study of parameters in the transverse momentum equation were performed to ...

  16. State-of-the-art evaluation of condensate polisher performance. Final report. [PWR

    SciTech Connect (OSTI)

    Elmiger, S.J.; Potterton, S.J.

    1983-04-01T23:59:59.000Z

    The operating environments in power plant cycles require the use of high-purity feedwater combined with careful chemical control to minimize deposition and corrosion problems. Condensate polishing system provide plant operators with a tool to control the transport of soluble impurities and corrosion products during normal operation, start-ups, shutdowns, and other plant transients. The objective of this program was to provide an understanding of the factors that influence condensate polisher performance. In order to achieve a large data base for evaluation, a lengthy and detailed plant survey was chosen as the means by which to accumulate the data. Response to the survey was excellent. Of the 268 surveys distributed, 147 were completed and returned with 135 of these included in a computerized condensate polishing data base. The data from the effort have been summarized in this report to show the environment, operation problems as performance of both deep bed and powdered condensate polisher system.

  17. MODELING THE PERFORMANCE OF HIGH BURNUP THORIA AND URANIA PWR FUEL

    E-Print Network [OSTI]

    Long, Y.

    Fuel performance models have been developed to assess the performance of ThO[subscript 2]-UO[subscript 2]

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

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

    current approach of long-term storage at its nuclear power plants and independent spent fuel storage installation, and deferred transportation of used nuclear fuel (UNF), along...

  19. Final assessment of MOX fuel performance experiment with Japanese PWR specification fuel in the HBWR

    SciTech Connect (OSTI)

    Fujii, Hajime; Teshima, Hideyuki; Kanasugi, Katsumasa [Mitsubishi Heavy Industries, Ltd., 1-1, Wadasaki-cho 1-chome, Hyogo-ku, Kobe 652-8585 (Japan); Kosaka, Yuji [Nuclear Development Corporation, 622-12 Funaishikawa, Tokai-mura, Ibaraki 319-1111 (Japan); Arakawa, Yasushi [The Kansai Electric Power Co., Inc., 8 Yokota, 13 Goichi, Mihama-cho, Mikata-gun, Fukui, 919-1141 (Japan)

    2007-07-01T23:59:59.000Z

    In order to obtain high burn-up MOX fuel irradiation performance data, SBR and MIMAS MOX fuel rods with Pu-fissile enrichment of about 6 wt% had been irradiated in the HBWR from 1995 to 2006. The peak burn-up of MOX pellet achieved 72 GWd/tM. In this test, fuel centerline temperature, rod internal pressure, stack length and cladding length were measured for MOX fuel and UO{sub 2} fuel as reference. MOX fuel temperature is confirmed to have no significant difference in comparison with UO{sub 2}, taking into account of adequate thermal conductivity degradation due to PuO{sub 2} addition and burn-up development. And the measured fuel temperature agrees well with FINE code calculation up to high burn-up region. Fission gas release of MOX is possibly greater than UO{sub 2} based on temperature and pressure assessment. No significant difference is confirmed between SBR and MIMAS MOX on FGR behavior. MOX fuel swelling rate agrees well with solid swelling rate in the literature. Cladding elongation data shows onset of PCMI in high power region. (authors)

  20. SYNTHESE EN FRANAIS TITRE: NEUTRONIC STUDY OF THE MONO-RECYCLING OF AMERICIUM IN PWR

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    retraitement standard consiste à séparer le plutonium afin de fabriquer un combustible MOX sur base d'uranium appauvri. La concentration du Pu dans le MOX est déterminée pour atteindre un taux d'irradiation du MOX de refroidissement de 30 ans demande à augmenter la teneur en Pu dans le MOX. L'241Am, avec une durée de vie de 432

  1. Los Alamos PWR feed-and-bleed studies summary results and conclusions

    SciTech Connect (OSTI)

    Boyack, B.E.; Henninger, R.J.; Lime, J.F.

    1985-01-01T23:59:59.000Z

    The adequacy of shutdown decay heat removal in pressurized water reactors (PWRs) is currently under investigation by the Nuclear Regulatory Commission. One part of this effort is review of feed-and-bleed procedures that could be used if the normal cooling mode through the steam generators was unavailable. Feed-and-bleed cooling is effected by manually activating the high-pressure injection (HPI) system and opening the power-operated relief valves (PORVs) to release the core decay energy. The feasibility of the feed-and-bleed concept as a diverse mode of heat removal has been evaluated at the Los Alamos National Laboratory. The TRAC-PF1 code has been used to predict the expected performances of the Oconee-1 and Calvert Cliffs-1 reactors of Babcock and Wilcox and Combustion Engineering, respectively, and the Zion-1 and H.B. Robinson-2 plants of Westinghouse. Feed and bleed was successfully applied in each of the four plants studied, provided it was initiated no later than the time of loss-of-secondary heat sink.

  2. Advanced design concepts for PWR and BWR high-performance annular fuel assemblies

    E-Print Network [OSTI]

    Ellis, Tyler Shawn

    2006-01-01T23:59:59.000Z

    Sobering electricity supply and demand projections, coupled with the current volatility of energy prices, have underscored the seriousness of the challenges which lay ahead for the utility industry. This research addresses ...

  3. Feasibility and economics of existing PWR transition to a higher power core using annular fuel

    E-Print Network [OSTI]

    Beccherle, Julien

    2007-01-01T23:59:59.000Z

    The internally and externally cooled annular fuel is a new type of fuel for PWRs that enables an increase in core power density by 50% within the same or better safety margins as the traditional solid fuel. Each annular ...

  4. Multi-Recycling of Transuranic Elements in a Modified PWR Fuel Assembly

    E-Print Network [OSTI]

    Chambers, Alex

    2012-10-19T23:59:59.000Z

    with improved spent fuel management technologies; • Enhance energy security by extracting energy recoverable in spent fuel and depleted Uranium ensuring Uranium does not become a limiting resource for nuclear energy; • Improve fuel cycle management, while...

  5. P(R) P(W|R) P(S|R, W) (1) LR (PGLR) [8

    E-Print Network [OSTI]

    Shirai, Kiyoaki

    , , , , , , , , , , , , , , , , , , , 20 Cv v p p pi Pd PI (13) 1 2 P(pi|PI) 4 EDR [10] 20 v p (p, v) (p , Cv) (3.3 ) 1 1: m 1 2 3 4 235591(p|Cv, 1) = O(p, Cv) + p( O(p, Cv) + ) (14) = 1 m = 2, 3 3.2 F m = 2, 3 987 1311 EDR 1000 4 8.16 118 4­4, 1996. [8] V. Sornlertlamvanich, , , . LR . 3 , 1997. [9] , . : . , (NLC­96), 1996. [10] . EDR

  6. Effects of Multiple Drying Cycles on HBU PWR Cladding Alloys | Department

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page onYouTube YouTube Note: Since the YouTube|6721 Federal Register / Vol.6: RecordJune-YearEffect ofof Energy Effects of

  7. Fuel Assembly Shaker Test for Determining Loads on a PWR Assembly under

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn't YourTransport inEnergy0.pdf Flash2010-60.pdf2JessiNicholasRE:EnergyEngine OilsSurrogate

  8. Embrittlement and DBTT of High-Burnup PWR Fuel Cladding Alloys | Department

    Energy Savers [EERE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directed offOCHCO2: Final EnvironmentalCounties, Idaho || Department:June 5,regarding Data Privacy |of

  9. WAPD-SC-545 HYDROGEN FLAMMABILITY DATA AND APPLICATION TO PWR

    Office of Scientific and Technical Information (OSTI)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOnItem Not Found Item Not Found The item youThe DiscoveryFuelsOfficeWAPD-SC-545 HYDROGEN

  10. Effects of Multiple Drying Cycles on High-Burnup PWR Cladding

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742Energy Chinaof EnergyImpactOn July 2, 2014 in theGroup Report |ofM

  11. Microsoft PowerPoint - MISO-SPP Market Impacts HydPwrConf 2014

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOEThe Bonneville PowerCherries 82981-1cnHighand RetrievalsFinalModule8.ppt MicrosoftDOE'sR.G. VanInChanging Energy

  12. Understanding Hydraulic Processes Primary Investigator: Frank H. Quinn

    E-Print Network [OSTI]

    Understanding Hydraulic Processes Primary Investigator: Frank H. Quinn Overview The hydraulic and connecting channel hydraulics models for use in Great Lakes water resource studies. 2000 Plans Niagara River Hydraulic Studies: Detailed analysis of the impact of hydraulic regime changes in the Niagara River

  13. TECHNISCHE UNIVERSITEIT EINDHOVEN Tentamen 2IC08: ComputerSystemen 2

    E-Print Network [OSTI]

    Franssen, Michael

    die is aangesloten via een H- brug op de PWR0 en PWR1 uitgangen van de practicumprocessor. De uitgangen PWR2 t/m PWR7 worden niet gebruikt en mogen elke w

  14. On the Disposition of Graphite Containing TRISO Particles and the Aqueous Transport of Radionuclides via Heterogeneous Geological Formations

    E-Print Network [OSTI]

    van den Akker, Bret Patrick

    2012-01-01T23:59:59.000Z

    element) 0.225 (compact only) 5.144 Graphite CSNF 21-PWR12-PWR 44-BWR 24-BWR UO2 21 PWR fuel assemblies 12 PWR fuel assemblies, 44

  15. Scratch, Click & Vote: E2E voting over the Internet Miroslaw Kutylowski and Filip Zagrski

    E-Print Network [OSTI]

    Institute of Mathematics and Computer Science Wroclaw University of Technology mirekk@im.pwr.wroc.pl filipz@im.pwr

  16. On Some Distributed Disorder Detection Krzysztof Szajowski

    E-Print Network [OSTI]

    ; e-mail: Krzysztof.Szajowski@pwr.edu.pl http://im.pwr.edu.pl/~szajow AMS Subject Classification(2010

  17. IEEE Computer Society Board of Governors kasahara@waseda.jp

    E-Print Network [OSTI]

    Kasahara, Hironori

    Compile Option: (*1) Sequential: -O3 ­qarch=pwr6, XLF: -O3 ­qarch=pwr6 ­qsmp=auto, OSCAR: -O3 ­qarch=pwr6 ­qsmp=noauto (*2) Sequential: -O5 -q64 ­qarch=pwr6, XLF: -O5 ­q64 ­qarch=pwr6 ­qsmp=auto, OSCAR: -O5 ­q64 ­qarch=pwr6 ­qsmp=noauto (Others) Sequential: -O5 ­qarch=pwr6, XLF: -O5 ­qarch=pwr6 ­qsmp

  18. Microsoft Word - Cover Page - Exhibit 7

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

    Steady Easement Appalachian Trail Tract 164-03 Jahoda CE SILVIO O. CONTE NATIONAL FISH AND WILDLIFE Appalachian Trail Tract 164-05 Mohawk Div. of Silvio O Conte NFWR...

  19. 1 Objective The aim of the work was to install the Trilinos software library on the HPCx

    E-Print Network [OSTI]

    Silvester, David J.

    ='xlc_r -q64 -O3 -qarch=pwr5 -qtune=pwr5' \\ CXX='xlC_r -q64 -qrtti=all -O3 -qarch=pwr5 -qtune=pwr5' \\ F77='xlf_r -q64 -O3 -qarch=pwr5 -qtune=pwr5' \\ --prefix=$HOME/trilinos7-mpi-libs \\ --enable-mpi \\ --with

  20. URL: http://www.kasahara.cs.waseda.ac.jp/ , ,,TV, DVD

    E-Print Network [OSTI]

    Kasahara, Hironori

    -core SMP Server Compile Option: (*1) Sequential: -O3 ­qarch=pwr6, XLF: -O3 ­qarch=pwr6 ­qsmp=auto, OSCAR: -O3 ­qarch=pwr6 ­qsmp=noauto (*2) Sequential: -O5 -q64 ­qarch=pwr6, XLF: -O5 ­q64 ­qarch=pwr6 ­qsmp=auto, OSCAR: -O5 ­q64 ­qarch=pwr6 ­qsmp=noauto (Others) Sequential: -O5 ­qarch=pwr6, XLF: -O5

  1. In-Vessel Retention Technology Development and Use for Advanced PWR Designs in the USA and Korea

    SciTech Connect (OSTI)

    T.G. Theofanous; S.J. Oh; J.H. Scobel

    2004-05-18T23:59:59.000Z

    In-Vessel Retention (IVR) of molten core debris by means of external reactor vessel flooding is a cornerstone of severe accident management for Westinghouse's AP600 (advanced passive light water reactor) design. The case for its effectiveness (made in previous work by the PI) has been thoroughly documented, reviewed as part of the licensing certification, and accepted by the US Nuclear Regulatory Commission. A successful IVR would terminate a severe accident, passively, with the core in a stable, coolable configuration (within the lower head), thus avoiding the largely uncertain accident evolution with the molten debris on the containment floor. This passive plant design has been upgraded by Westinghouse to the AP1000, a 1000 MWe plant very similar to the AP600. The severe accident management approach is very similar too, including In-Vessel Retention as the cornerstone feature, and initial evaluations indicated that this would be feasible at the higher power as well. A similar strategy is adopted in Korea for the APR1400 plant. The overall goal of this project is to provide experimental data and develop the necessary basic understanding so as to allow the robust extension of the AP600 In-Vessel Retention strategy for severe accident management to higher power reactors, and in particular, to the AP1000 advanced passive design.

  2. Role of ex-vessel interactions in determining the severe reactor-accident source term for fission products. [PWR; BWR

    SciTech Connect (OSTI)

    Powers, D.A.; Brockmann, J.E.; Bradley, D.R.; Tarbell, W.W.

    1983-01-01T23:59:59.000Z

    The role fission-product release and aerosol generation outside the primary system can have in determining the severe reactor-accident source term is reviewed. Recent analytical and experimental studies of major causes of ex-vessel fission product release and aerosol generation are described. The ejection of molten-core debris from a pressurized-reactor vessel is shown to be a potentially large source of aerosols that has not been recognized in past severe-accident evaluations. A mechanistic model of fission-product release during core-debris interactions with concrete is discussed. Calculations with this model are compared to correlations of experimental data and previous estimates of ex-vessel fission-product release. Predictions with the mechanistic model agree quite well with the data correlations but do not agree at all well with estimates made in the past.

  3. Results and insights of a level-1 internal event PRA of a PWR during mid-loop operations

    SciTech Connect (OSTI)

    Chu, T.L.; Musicki, Z.; Kohut, P.

    1993-12-31T23:59:59.000Z

    Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analysis that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. The objective of this paper is to present the approach utilized in the level-1 PRA for the Surry plant, and discuss the results obtained. A comparison of the results with those of other shutdown studies is provided. Relevant safety issues such as plant and hardware configurations, operator training, and instrumentation and control is discussed.

  4. Enhanced Control of PWR Primary Coolant Water Chemistry Using Selective Separation Systems for Recovery and Recycle of Enriched Boric Acid

    SciTech Connect (OSTI)

    Ken Czerwinski; Charels Yeamans; Don Olander; Kenneth Raymond; Norman Schroeder; Thomas Robison; Bryan Carlson; Barbara Smit; Pat Robinson

    2006-02-28T23:59:59.000Z

    The objective of this project is to develop systems that will allow for increased nuclear energy production through the use of enriched fuels. The developed systems will allow for the efficient and selective recover of selected isotopes that are additives to power water reactors' primary coolant chemistry for suppression of corrosion attack on reactor materials.

  5. Life Estimation of PWR Steam Generator U-Tubes Subjected to Foreign Object-Induced Fretting Wear

    SciTech Connect (OSTI)

    Jo, Jong Chull; Jhung, Myung Jo; Kim, Woong Sik; Kim, Hho Jung [Korea Institute of Nuclear Safety (Korea, Republic of)

    2005-10-15T23:59:59.000Z

    This paper presents an approach to the remaining life prediction of steam generator (SG) U-tubes, which are intact initially, subjected to fretting-wear degradation due to the interaction between a vibrating tube and a foreign object in operating nuclear power plants. The operating SG shell-side flow field conditions are obtained from a three-dimensional SG flow calculation using the ATHOS3 code. Modal analyses are performed for the finite element models of U-tubes to get the natural frequency, corresponding mode shape, and participation factor. The wear rate of a U-tube caused by a foreign object is calculated using the Archard formula, and the remaining life of the tube is predicted. Also discussed in this study are the effects of the tube modal characteristics, external flow velocity, and tube internal pressure on the estimated results of the remaining life of the tube.

  6. Organizational analysis and safety for utilities with nuclear power plants: perspectives for organizational assessment. Volume 2. [PWR; BWR

    SciTech Connect (OSTI)

    Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Nadel, M.V.; Scott, W.G.; Connor, P.E.; Kerwin, N.; Kennedy, J.K. Jr.

    1983-08-01T23:59:59.000Z

    This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. Volume 1 of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety. The six chapters of this volume discuss the major elements in our general approach to safety in the nuclear industry. The chapters include information on organizational design and safety; organizational governance; utility environment and safety related outcomes; assessments by selected federal agencies; review of data sources in the nuclear power industry; and existing safety indicators.

  7. Technical considerations related to interim source-term assumptions for emergency planning and equipment qualification. [PWR; BWR

    SciTech Connect (OSTI)

    Niemczyk, S.J.; McDowell-Boyer, L.M.

    1982-09-01T23:59:59.000Z

    The source terms recommended in the current regulatory guidance for many considerations of light water reactor (LWR) accidents were developed a number of years ago when understandings of many of the phenomena pertinent to source term estimation were relatively primitive. The purpose of the work presented here was to develop more realistic source term assumptions which could be used for interim regulatory purposes for two specific considerations, namely, equipment qualification and emergency planning. The overall approach taken was to adopt assumptions and models previously proposed for various aspects of source term estimation and to modify those assumptions and models to reflect recently gained insights into, and data describing, the release and transport of radionuclides during and after LWR accidents. To obtain illustrative estimates of the magnitudes of the source terms, the results of previous calculations employing the adopted assumptions and models were utilized and were modified to account for the effects of the recent insights and data.

  8. Influence of steam on the behavior of U/sub 3/O/sub 8/ aerosols. [PWR; BWR

    SciTech Connect (OSTI)

    Adams, R.E.; Tobias, M.L.; Kress, T.S.

    1982-01-01T23:59:59.000Z

    A project is being conducted in the Nuclear Safety Pilot Plant (NSPP), located at the Oak Ridge National Laboratory (ORNL), to study the behavior of aerosols assumed to be generated during LWR reactor accident sequences and released into containment. This project, which is part of the ORNL Aerosol Release and Transport (ART) Program, is sponsored by the Nuclear Regulatory Commission and its purpose is to provide experimental qualification for LWR aerosol behavioral codes being developed independently by other NRC-sponsored programs. The program plan for the NSPP aerosol project provides for the study of the behavior of LWR accident aerosols emanating from fuel, reactor core structural materials, and from concrete-molten metal reactions. The behavior of each of these aerosols is being studied individually to establish their characteristics; future experiments will involve mixtures of these aerosols to establish their interaction and collective behavior within containment. The purpose of this paper is to document observations illustrating the influence that steam has on the behavior of U/sub 3/O/sub 8/ aerosols within the NSPP vessel.

  9. FEMAXI-V benchmarking study on peak temperature and fission gas release prediction of PWR rod fuel

    SciTech Connect (OSTI)

    Suwardi; Dewayatna, W.; Briyatmoko, B. [Center for Nuclear Fuel Technology - National Nuclear Energy Agency, Puspiptek Tangerang - 15310 (Indonesia)

    2012-06-06T23:59:59.000Z

    The present paper reports a study of FEMAXI-V code and related report on code benchmarking. Capabilities of the FEMAXI-V code to predict the thermal and fission gas release have been tested on MOX fuels in LWRs which has been done in SCK{center_dot}CEN and Belgonucleaire by using PRIMO MOX rod BD8 irradiation experiment after V Sobolev as reported O. J. Ott. Base irradiation in the BR3 reactor, the BD8 rod was transported to CEA-Saclay for irradiation in the OSIRIS reactor (ramp power excursion). The irradiation device used for the PRIMO ramps was the ISABELLE 1 loop, installed on a movable structure of the core periphery. The power variations were obtained by inwards/backwards movements of the loop in the core water. The preconditioning phase for rod BD8 occurred at a peak power level of 189 W/cm with a hold time of 27 hours. The subsequent power excursion rate amounted to 77 W/ (cm.min), reaching a terminal peak power level of 395 W/cm that lasted for 20 hours.

  10. Structure and high-temperature stability of compositionally graded CVD mullite coatings

    E-Print Network [OSTI]

    Basu, Soumendra N.

    C (Carborundum, Niagara Falls, NY). The coatings were deposited using the AlCl3± SiCl4±CO2±H2 system in a hot

  11. Lining Over Refractory - Conserve Energy and Capital

    E-Print Network [OSTI]

    Jost, M. L.; Barrows, G. L.

    1980-01-01T23:59:59.000Z

    .~. LINING OVER REFRACTORY - CONSERVE ENERGY & CAPITAL by Mark L. Jost Gerald L. Barrows The Carborundum Company Niagara Falls, New York INTRODUCTION Companies operating industrial heating equip Advantages ment find themselves coming under...

  12. JOINT PROGRAM IN TRANSPORTATION UNIVERSITY OF TORONTO

    E-Print Network [OSTI]

    Toronto, University of

    JOINT PROGRAM IN TRANSPORTATION UNIVERSITY OF TORONTO 2001 TRANSPORTATION TOMORROW SURVEY of Transportation, Ontario Additions in 1996 Regional Municipalities of Niagara, Waterloo Counties of Peterborough not to participate) #12;JOINT PROGRAM IN TRANSPORTATION UNIVERSITY OF TORONTO 2001 TRANSPORTATION TOMORROW SURVEY

  13. Effective: Saturday, November 15 (MD Football vs Michigan State) For more information visit transportation.umd.edu or call (301) 314-2255

    E-Print Network [OSTI]

    Hill, Wendell T.

    12:00AM midnight. 122 Green Suspended Service will be suspended until 12:00AM midnight. Nite Ride Lackawanna St Lackawanna St Laguna Rd Muskogee St Muskogee St Laguna Rd Edgewood Rd Niagara Rd Locust Hill Dr

  14. Superior Energy Performance?: Recognizing Excellence in Energy...

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

    Niagara Falls, NY: http:industrial-energy.lbl.govfiles industrial-energyactive0LBNL-6349E.pdf 2 Find more information on the Better Buildings, Better Plants Program at:...

  15. A Domain-Specific Safety Analysis for Digital Nuclear Plant Protection Systems

    E-Print Network [OSTI]

    Specification method - NuSCR(3) 9 [k_VAR_OVER_PWR_Trip_Dly, k_VAR_OVER_PWR_Trip_Dly] th_VAR_OVER_PWR_Trip_Logic := k_TRUE #12;NUFTA 10 #12;Overview of Nu

  16. 2.1E Supplement

    E-Print Network [OSTI]

    Winkelmann, F.C.

    2010-01-01T23:59:59.000Z

    version added HERM-CENT-COND-PWR HERM-CENT-COND-TYPE P -QUAD V 4 , 3 HERM-REC-COND-PWR P - PLANT-PARAMETERS V 2 . .FT DEFROST-CAP-FT DEFROST-PWR-FT HPDefrst HPDefrst HPDefrst

  17. Determining Plutonium Mass in Spent Fuel with Nondestructive Assay Techniques NGSI Research Overview and Update on NDA Techniques

    E-Print Network [OSTI]

    A., V. Mozin, S.J. Tobin, L.W. Cambell, J.R. Cheatham, C.R. Freeman, C.J. Gesh,

    2012-01-01T23:59:59.000Z

    considered one of the 17x17 PWR assemblies from the NGSIplutonium signal because in a PWR spent fuel its content isspectra for a single PWR fuel pin with fresh and spent UO 2

  18. Geological Problems in Radioactive Waste Isolation: Second Worldwide Review

    E-Print Network [OSTI]

    2010-01-01T23:59:59.000Z

    pressurized water reactors (PWR) with a combined capacity ofelements from the Loviisa PWRs assemblies as well. The emptyBWR/4 BWR/4 BWR/6 BWR/6 PWR PWR ABWR (scheduled) operating

  19. In-situ Surface Enhanced Raman Spectroscopy Investigation of the Surface Films on Alloy 600 and Alloy 690 in Pressurized Water Reactor-Primary Water

    E-Print Network [OSTI]

    Wang, Feng

    2012-01-01T23:59:59.000Z

    oxidation of Alloy 600 in PWR Primary Water. The layered-oxidation of Alloy 690 in PWR Primary Water. The film ofwith oxidation of Alloy 600 in PWR Primary Water. The film

  20. Full Core, Heterogeneous, Time Dependent Neutron Transport Calculations with the 3D Code DeCART

    E-Print Network [OSTI]

    Hursin, Mathieu

    2010-01-01T23:59:59.000Z

    core layout for the 1/8th PWR core model ________________ 76and future of the NEACRP PWR core transient benchmark. 199449. Hursin, M. (2008). PWR Control Rod Ejection Analysis

  1. PRELIMINARY THERMAL AND THERMOMECHANICAL MODELING FOR THE NEAR SURFACE TEST FACILITY HEATER EXPERIMENTS AT HANFORD

    E-Print Network [OSTI]

    chan, T.

    2011-01-01T23:59:59.000Z

    to the power generated by a PWR (Pressurized Water Reactor)or 1 kW, corresponding to a PWR spent fuel assembly 2.5 andcanister, the heat output of a PWR of spent fuel assembly

  2. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    of conventional LWR systems (PWR & BWRs), partly due to thethe margin to boiling in a PWR is ?15 ? C, while the coolantprimary heat exhangers of a PWR, in which borated water is

  3. Development of Superconducting High-Resolution Gamma-Ray Spectrometers for Nuclear Safeguards

    E-Print Network [OSTI]

    Dreyer, Jonathan

    2012-01-01T23:59:59.000Z

    Production of plutonium in PWR fuel as function ofNDA NEP PNCC PR PSD PTR PWR SQUID TES TIMS UHV wt% Adiabaticcomposition of plutonium in a PWR as a function of burn up

  4. Spin-On for the Renaissance? The Current State of China's Nuclear Industry

    E-Print Network [OSTI]

    Yuan, Jing-dong

    2010-01-01T23:59:59.000Z

    are primarily based on two PWR designs: the CPR-1000 andpressurized water reactor (PWR) in Qinshan in the mid-1980s,civilian purposes. Fabrication of PWR fuel is undertaken at

  5. A Coupled Model for Natural Convection and Condensation in Heated Subsurface Enclosures Embedded in Fractured Rock

    E-Print Network [OSTI]

    Halecky, N.; Birkholzer, J.T.; Webb, S.W.; Peterson, P.F.; Bodvarsson, G.S.

    2006-01-01T23:59:59.000Z

    packages such as the “21 PWR” or the “44 BWR” (Figure 3).drift length (for the “21 PWR”). For comparison: the initialdrift) "5 HLW Long" "21 PWR AP" "44 BWR AP" "5 HLW SHORT"

  6. Optimal Partial Feedback Design for MIMO Block Fading Channels with Causal Noiseless Feedback

    E-Print Network [OSTI]

    Liu, Youjian "Eugene"

    T . . . . . . . . y1 ynR (a) No CSIT. Encoder Encoder Encoder Pwr Pwr Pwr Eigen- beamforming N=min{nT,nR} X1 X2 XNT X1

  7. LIPs on Venus Vicki L. Hansen

    E-Print Network [OSTI]

    Hansen, Vicki

    -O, Quetzalpetlatl, Atahensik), crustal plateaus, and `plains with wrinkle ridges', unit pwr. Unit pwr, widely by massive partial mantle melting caused by large bolide impact on thin lithosphere. The status of unit pwr

  8. Drift Natural Convection and Seepage at the Yucca Mountain Repository

    E-Print Network [OSTI]

    Halecky, Nicholaus Eugene

    2010-01-01T23:59:59.000Z

    between to hot 21-PWR waste packages. . . . . . . . . . .difference between a hot 21-PWR waste canister (having anis placed next to a hot 21-PWR waste canister and see what

  9. A STUDY OF REGIONAL TEMPERATURE AND THERMOHYDROLOGICAL EFFECTS OF AN UNDERGROUND REPOSITORY FOR NUCLEAR WASTES IN HARD ROCK

    E-Print Network [OSTI]

    Wang, J.S.Y.

    2010-01-01T23:59:59.000Z

    to a BWR Effects of different PWR fuel cycles from the sameFig. 14 Effects of different PWR cycles from the same amountof wastes from different PWR fuel cycles normalized to 10 W/

  10. GEOTECHNICAL ASSESSMENT AND INSTRUMENTATION NEEDS FOR NUCLEAR WASTE ISOLATION IN CRYSTALLINE AND ARGILLACEOUS ROCKS SYMPOSIUM

    E-Print Network [OSTI]

    Authors, Various

    2011-01-01T23:59:59.000Z

    recycling of plutonium in PWR. KBS Report 111. Lakatos, T. (hogaktivt avfall fr%n·en PWR beraknade med ORIGEN ("Emissionand high-level waste from a PWR, calculated using ORIGEW'),

  11. Alternative Energy Development and China's Energy Future

    E-Print Network [OSTI]

    Zheng, Nina

    2012-01-01T23:59:59.000Z

    the earliest one- and two-loop PWR design and the CNP-1000as the standard three-loop PWR design with a high burn-upCPR-1000 and 1000+ Generation II PWR designs and the AP1000

  12. China Energy Databook -- User Guide and Documentation, Version 7.0

    E-Print Network [OSTI]

    Fridley, Ed., David

    2008-01-01T23:59:59.000Z

    in Guangdong has two 900 MW PWR units. Source: EB, Chinain Guangdong has two 900 MW PWR units. Source: EB, ChinaUnits (MW x no units) 900x2 PWR nuclear 300x2 600x2 [1

  13. Uncertainties in the Anti-neutrino Production at Nuclear Reactors

    E-Print Network [OSTI]

    Djurcic, Zelimir

    2009-01-01T23:59:59.000Z

    Isotopic Analysis of High-Burnup PWR Spent Fuel Samples FromIsotopic Predictions for PWR Spent Fuel”, ORNL/TM-13317,Analysis for San Onofre PWR MOX Fuel”, ORNL/TM-1999/326,

  14. Small-Scale Readout Systems Prototype for the STAR PIXEL Detector

    E-Print Network [OSTI]

    Szelezniak, Michal A.

    2008-01-01T23:59:59.000Z

    PIXEL detector. PIXEL COLUMN CIRCUITRY VREF1 PWR_ON VREF2VDD RESET PWR_ONREAD PWR_ON CALIB RESET VR1 VR2 Q READ MOSCAP SOURCE

  15. Boric Acid Causes ER Stress and Activates the eIF2alpha/ATF4 and ATF6 Branches of the Unfolded Protein Response in Prostate Cancer Cells and Using Toxicology in the Public Interest

    E-Print Network [OSTI]

    Kobylewski, Sarah Ellen

    2012-01-01T23:59:59.000Z

    isoforms in DU-145, LNCaP, and PWR-1E cells. Biochem Biophysisoforms in DU-145, LNCaP, and PWR-1E cells. Biochem Biophysin prostate DU-145, LNCaP, and PWR-1E cells. Biochemical and

  16. IEEE Computer Society Board of Governors kasahara@waseda.jp

    E-Print Network [OSTI]

    Kasahara, Hironori

    on IBM p6 595 Power6 (4.2GHz) based 32-core SMP Server Compile Option: (*1) Sequential: -O3 ­qarch=pwr6, XLF: -O3 ­qarch=pwr6 ­qsmp=auto, OSCAR: -O3 ­qarch=pwr6 ­qsmp=noauto (*2) Sequential: -O5 -q64 ­qarch=pwr6, XLF: -O5 ­q64 ­qarch=pwr6 ­qsmp=auto, OSCAR: -O5 ­q64 ­qarch=pwr6 ­qsmp=noauto (Others

  17. TECHNISCHE UNIVERSITEIT EINDHOVEN Tentamen 2IC08: ComputerSystemen 2

    E-Print Network [OSTI]

    Franssen, Michael

    de uitgangen PWR0 t/m PWR3. Daarbij is telkens één uitgang hoog en de overige uitgangen zijn laag. Door de hoge uitgang één poort modulo 4 naar boven op te schuiven (bijvoorbeeld van PWR1=1 naar PWR2=1 of van PWR3=1 naar PWR0=1) zet de motor een stap naar links. Schuift de hoge uitgang naar de andere kant

  18. NUCLEAR ENERGY RENAISSANCE:NUCLEAR ENERGY RENAISSANCE: ADDRESSING THE CHALLENGES OF CLIMATE CHANGE AND SUSTAINABILITYADDRESSING THE CHALLENGES OF CLIMATE CHANGE AND SUSTAINABILITY

    E-Print Network [OSTI]

    ­­LWR, BWR, PWR, CANDU, VVER, RBMKLWR, BWR, PWR, CANDU, VVER, RBMK 19951995 ­­ 2010 GEN III2010 GEN III and1 and--2; Fukushima2; Fukushima--1;1; GariglianoGarigliano PWR:PWR: HB Robinson; Palisades; Early:BWR: OskarshamnOskarshamn--2; La Salle; Fuku2; La Salle; Fuku--2;2; TokaiTokai--2, Leibstatd2, Leibstatd PWR:PWR

  19. The chemistry of OH and HO2 radicals in the boundary layer over the tropical Atlantic Ocean

    E-Print Network [OSTI]

    2010-01-01T23:59:59.000Z

    CO and NO), respectively. Pwr is the laser power enterings). Using the above values and Pwr = 9 mW, LODs of 1.1×10 6

  20. Automatic aligning free space communication platform

    E-Print Network [OSTI]

    Andrews, John Michael

    2008-01-01T23:59:59.000Z

    addchannel(d,2,'Power'); %% PWR else addchannel(d,0,'Y addchannel(d,2,'Power'); %% PWR end %%% Set up sampling

  1. Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors

    E-Print Network [OSTI]

    Terrani, Kurt Amir

    2010-01-01T23:59:59.000Z

    1.2.1 PWRs . . . . . . . . . . . . . . . . . . . . 1.2.2Actinides Multi-Recycling in PWR Using Hydride Fuels. InRecycling in Hydride Fueled PWR Cores. Nuclear Engineering

  2. Half-Scale Model Tests on the Three Quarter Wave R.F. System for the 184-inch Frequency Modulated Cyclotron

    E-Print Network [OSTI]

    Anderson, Robert L.

    2010-01-01T23:59:59.000Z

    ~ o DC PLATE VOLTAGE o PWR INPUT WATTS TO TRIODES ~--o--,110-I DC PLATE' VOLTAGE' PWR INPUT WATTS TO TRIODES

  3. Power Efficiency and the Top500 John Shalf, Shoaib Kamil, Erich Strohmaier, David Bailey

    E-Print Network [OSTI]

    SystemPower(kW) max pwr avg pwr Growth in Power Consumption (Top50) Excluding Cooling 0.00 100.00 200

  4. DESIGN OF A MOBILE LABORATORY FOR VENTILATION STUDIES AND INDOOR AIR POLLUTION MONITORING

    E-Print Network [OSTI]

    Berk, James V.

    2011-01-01T23:59:59.000Z

    210 GA 2/0 GA I I I I C I 20A Instrument regulated pwr I 15AInstrument non-regulated pwr 20A I • I I 30 A twistA twist lock tool box Pump pwr Zero gas generated pwr

  5. DATA ACQUISITION, HANDLING, AND DISPLAY FOR THE HEATER EXPERIMENTS AT STRIPA

    E-Print Network [OSTI]

    McEvoy, M.B.

    2011-01-01T23:59:59.000Z

    8M. INTLKS. COIfIPL. t..O HZ. PWR. ON X EMERt:'£NCY LlGHT! ;~ ~"cf? ;;:~~ (}t; ( 50NZ PWR. ,wHlPWR. I 77M>PWR. HA ISH . (jb"" IN~li1L 50

  6. DESIGN OF A MOBILE LABORATORY FOR VENTILATION STUDIES AND INDOOR AIR POLLUTION MONITORING

    E-Print Network [OSTI]

    Berk, James V.

    2011-01-01T23:59:59.000Z

    i co i 2KVA sou reg. X Instrument regulated pwr 20A 20AInstrument non-regulated pwr 30A twist lock tool box 30 Alock ( __J~ v x-k \\^_J~ Pump pwr 30A Zero gas generated pwr

  7. Mechanisms of Small RNA Degradation and Characterization of THO Complex Mutants in Arabidopsis

    E-Print Network [OSTI]

    ZHAO, YUANYUAN

    2013-01-01T23:59:59.000Z

    LIL HSP20/ LIL Ler WT-Ler_r2 CGATGT Ctrl Ler pwr-1_r2 TGACCAWT-Ler_r2 PWR Ler top1a_r2 ACAGTG WT-Ler_r2 TOP1A ColGCCAAT 9-7-2_r2 TAF6 Col pwr-2_r2 GTCCGC Col_r2 PWR Col

  8. Rhodopsin Reconstituted into a Planar-Supported Lipid Bilayer Retains Photoactivity after Cross-Linking Polymerization of Lipid Monomers

    E-Print Network [OSTI]

    Brown, Michael F.

    -waveguide resonance (PWR) spectroscopy10 to characterize Rho in PSLBs. PWR is highly sensitive to the optical the aqueous volume of the PWR cell has been described.7,11 Rho12 was reconstituted into the PSLB by introducing small aliquots of octylglucoside-solubi- lized receptor into the PWR cell, which contained 10 m

  9. JOURNAL DE PHYSIQUE Colloque C2, supplment au n03, Tome 47, mars 1986 page c2-191

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    reactors (PWR), respectively. The Zircaloys contain tin and small amounts of iron and chromium. Zircaloy-2 is not the same in PWR and BWR environments. In water (PWR) "general corrosion" dominates. This type of corrosion in a PWR, "nodular corrosion" is the life determining mechanism. In this case, after some time of general

  10. Molecular Cell, Volume 45 Supplemental Information

    E-Print Network [OSTI]

    van Oudenaarden, Alexander

    imaging of (A) FLO11 and PWR1, (A) FLO11 and ICR1, and (C) ICR1 and PWR1 transcripts in individual cells) and PWR1 (Cy5; red dots) transcripts in fields of intact individual WT cells. DAPI staining (blue) shows the locations of nuclei. (C) Merged fluorescence and DIC microscopy images show ICR1 (TMR; green dots) and PWR

  11. MENTORING FROM THE TOP: STORIES OF SUCCESS & LESSONS LEARNED

    E-Print Network [OSTI]

    Nelson, Tim

    The SUNY Registrar's Association ­ building future leaders & other stories JudyTatum, Senior Director MohawkValley Community College, Utica branch The SUNY Registrar's Association Building future leadersValley Community College Located in Utica, NY- population 60,600 The first NYS Community College (1946) Utica

  12. axon reflex test: Topics by E-print Network

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

    axon reflex test First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Reflexives in Mohawk University of...

  13. Spatial patterns of flow and their modification within and around a giant kelp forest Brian Gaylord1

    E-Print Network [OSTI]

    California at Santa Cruz, University of

    Spatial patterns of flow and their modification within and around a giant kelp forest Brian Gaylord and over the full extent of a giant kelp (Macrocystis pyrifera) forest located at Mohawk Reef, Santa reported for larger (kilometer-scale) kelp beds, suggesting that alongshore currents may play a greater

  14. CX-000033: Categorical Exclusion Determination

    Broader source: Energy.gov [DOE]

    Saint Regis Mohawk Tribe Energy Efficiency and Conservation Programs for Buildings and FacilitiesCX(s) Applied: B5.1, A9Date: 11/02/2009Location(s): New YorkOffice(s): Energy Efficiency and Renewable Energy

  15. Light-water reactors: preliminary safety and environmental information document. Volume I

    SciTech Connect (OSTI)

    Not Available

    1980-01-01T23:59:59.000Z

    Information is presented concerning the reference PWR reactor system; once-through, low-enrichment uranium-235 fuel, 30 MWD per kilogram (PWR LEU(5)-OT); once-through, low-enrichment, high-burnup uranium fuel (PWR LEU(5)-Mod OT); self-generated plutonium spiked recycle (PWR LEU(5)-Pu-Spiked Recycle); denatured uranium-233/thorium cycle (PWR DU(3)-Th Recycle DU(3)); and plutonium/thorium cycle (Pu/ThO/sub 2/ Burner).

  16. Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods; Revision 1

    SciTech Connect (OSTI)

    Johnson, G.L.

    1991-11-01T23:59:59.000Z

    Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. One such package would store tightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97{degrees}C and whether the cladding of the stored spent fuel ever exceeds 350{degrees}C. Limiting the borehole to temperatures of 97{degrees}C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350{degrees}C cladding limit minimizes the possibility of creep- related failure in the spent fuel rod cladding. For a series of packages stored in a 8 {times} 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97{degrees}C for the full 10000-yr analysis period. For the 4.74-kW load, the peak cladding temperature rises to just below the 350{degrees}C limit about 4 years after emplacement. If the packages are stored using the spacing specified in the Site Characterization Plan (15 ft {times} 126 ft), a maximum of 4.1 kW per container may be stored. If the 0.05-m-thick void between the container and the borehole wall is filled with loosely packed bentonite, the peak cladding temperature rises more than 40{degrees}C above the allowed cladding limit. In all cases the dominant heat transfer mode between container components is thermal radiation.

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

    SciTech Connect (OSTI)

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

    1982-03-01T23:59:59.000Z

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

  18. IBM United States Withdrawal Announcement 908-175, dated August 05, 2008

    E-Print Network [OSTI]

    M LC/LC FIBER CABLE 2053 All 5601 PWR CORD, 250V 2.5A, ARGEN 2053 All 9231 PWR CORD, 250V 2.5A, AUSTR 2053 All 9232 PWR CORD, 250V 2.5A, EU 2053 All 9233 PWR CORD, 125V 3A, US 2053 All 9234 PWR CORD, 250V 2.5A, UK 2053 All 9235 PWR CORD, 250V 2.5A, KOREA 2053 All 9236 Machine type 2054 8-PORT IP

  19. Mining Large Graphs And Streams Using Matrix And Tensor Tools

    E-Print Network [OSTI]

    Kolda, Tamara G.

    RESULTS The graphs at right show overall variability distribution estimated for the Pentium D 800 where a core no longer works properly. In the Sun T1 Niagara cores this is done with a built-in- self processors we record the temperature at which the failure occurred and adjust to the frequencies

  20. Water Qual. Res. J. Canada, 2004 Volume 39, No. 3, 213222 Copyright 2004, CAWQ

    E-Print Network [OSTI]

    Mazumder, Asit

    produced by algae and microbial processes in raw or stored water. Examining and understanding the principal and odour compounds in water supplies. The most investigated taste and odour forming com- pounds produced Ontario, the St. Lawrence River, and treatment plants that draw water from Lake Erie and from the Niagara

  1. ASM Dinner MeetingASM Dinner MeetingASM Dinner MeetingASM Dinner Meeting ASM Ottawa Valley ChapterASM Ottawa Valley ChapterASM Ottawa Valley ChapterASM Ottawa Valley Chapter

    E-Print Network [OSTI]

    Ellis, Randy

    for the Class 2 nuclear facility was granted by the Canadian Nuclear Safety Commission in October 2011. The RMTL Chalk River Laboratory Chalk River, ON K0J 1J0 Canada Institute for S T I National Research Council Longueuil, QC J4G 1T5 MetLab Corporation PO Box 1075 Niagara Falls, NY 14302 Institute for Aerospace

  2. It was a hot summer night in New York in 1894, and the reporter had decided that it was time to meet the Wizard.

    E-Print Network [OSTI]

    Rowley, Clarence W.

    - ventor, Nikola Tesla. His name was on everyone's lips: "Every scientist knows his work and every foolish without prior written permission of the publisher. #12;Dinner at Delmonico's © 3 Nikola Tesla is almost to gener- ate electricity at the new plant under construction at Niagara Falls, but Tesla had taken 250

  3. Wireless networks and mobile Ivan Stojmenovic

    E-Print Network [OSTI]

    Stojmenovic, Ivan

    medium #12;Nikola Tesla 1856-1943 · The Serbian-American inventor, electrical engineer, and scientist plant at Niagara Falls 1895 #12;Nikola Tesla's inventions · Radio/wireless transmission · US Supreme Court awarded patent to Tesla in 1945, taking it from Marconi #12;Lost inventions · When Nikola Tesla

  4. Analysis of a small sample geometry for concurrent identification and quantification of mixed-nuclide samples

    E-Print Network [OSTI]

    Krieger, Kenneth Vincent

    1999-01-01T23:59:59.000Z

    to each channel of the MCA is used to display a histogram, or spectrum, which is representative of the photon energy deposited in the crystal. * Aptec Instruments Inc. , 908 Niagara Falls Blvd, Ste 524, N. Tonawanda, NY. 14120. HPGe Detector Pre...

  5. ITP Industrial Distributed Energy: Cooling, Heating, and Power...

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

    60. Southwestern Electric Pwr 61. Texas Utilities Electric Co 62. Toledo Edison Co 63. Tucson Electric Power Co 64. Union Electric Co 65. Virginia Electric & Pwr Co 66. West Penn...

  6. A universal low-noise analog receiver baseband in 65-nm CMOS

    E-Print Network [OSTI]

    Tekin, Ahmet; Elwan, Hassan; Pedrotti, Kenneth

    2010-01-01T23:59:59.000Z

    ð ð BW Á C tot Þ=N Þ Á ð ð Pwr Þ= ð N Á BW Þ Þ where DR isamount of capacitance used, Pwr is the power consumption and

  7. Nordisk kernesikkerhedsforskning Norrnar kjarnryggisrannsknir

    E-Print Network [OSTI]

    , PWR, neutron detector, keff, IACIP: NKS_R_2008_61 NKS-211 ISBN 978-87-7893-280-8 Electronic report Keywords CYGNUS, VNEM, Ringhals, unit 3, PWR, neutron detector, keff, IACIP: NKS_R_2008_61 Comparison

  8. Hypothetical Reactor Accident Study

    E-Print Network [OSTI]

    POPULATIONS; IODINE 131; MELTDOWN; METEOROLOGY; NUCLEAR POWER PLANTS; P CODES; PWR TYPE REACTORS; RADIATION in a Typical BWR and in a typical PWR. Comparison with WASH-1400 by C F . Højerup 202 APPENDIX 3. Calculation

  9. Thermodynamic Investigations of Aqueous Ternary Complexes for Am/Cm Separation

    E-Print Network [OSTI]

    Leggett, Christina Joy

    2012-01-01T23:59:59.000Z

    discharge (uranium-fueled PWR). Figure 1.2 Production of Cf.and pressurized water reactors (PWR). The majority of thepressurized water reactors. In PWRs, the coolant (water),

  10. Advanced phase modulation techniques for stimulated brillouin scattering suppression in fiber optic parametric amplifiers

    E-Print Network [OSTI]

    Coles, James

    2009-01-01T23:59:59.000Z

    s-1. —Measurement Results Integ Pwr: -Markers B 190.3398 THzMeasurement Results Integ Pwr: 0.375 dBrn MeanWL: 1575.03832

  11. 3.3.3AC Sweep AC . AC

    E-Print Network [OSTI]

    ­. DC ­ AC ­) .( . ­ ,, '­Spice .Spice . : 0 0 E1 PWR(V(%IN+, %IN . )2.3( Etable " . . : 00 V1 0Vdc E2 Pwr(V(%IN+, %IN-),2) ETABLE TABLE = (5

  12. Nordisk kernesikkerhedsforskning Norrnar kjarnryggisrannsknir

    E-Print Network [OSTI]

    follows vessel gen- erations ­ first to third generation; and reactor types ­ PWR and LMC technol- ogy generation; and reactor types ­ PWR and LMC technology. Most of the available information is related

  13. CRICKET V2.0 NETWORKS AND MOBILE SYSTEMS GROUP

    E-Print Network [OSTI]

    _AMP_VREF US _AMP_PWR US_AMP_OUT US_AMP_VREF US_AMP_PWR PDATA POT_CS POT_SCK US_IN_ENA US_DETECT VCC VCC R45 R

  14. China Energy Primer

    E-Print Network [OSTI]

    Ni, Chun Chun

    2010-01-01T23:59:59.000Z

    pressurized water reactors (PWRs) the principal but not theJiangsu Capacity (MW) Reactor Type PWR:CNP-300 Operator CNNC1991 Qinshan Phase II Unit 1 PWR:CNP-600 CNNC 2-Jun-1996 15-

  15. Study of a low Mach nuclear core model for two-phase ows with phase

    E-Print Network [OSTI]

    Figure 1 for schematic pictures of PWR and BWR reactors). A natural approach is to represent Paris, France yohan.penel@cerema.fr 1PWR is the acronym for Pressurized Water Reactor. 2BWR

  16. RIS-M-2302 LIST OF SELECTED PUBLICATIONS 1980

    E-Print Network [OSTI]

    -Dimensional PWR Transient Code ANTI. Riso-M-2256 (1980) 106 pp. Friis Jensen, J. and I. Misfeldt, User Manual Neutron Dosemeter. Risø-M- 2247 (1980) 14 pp. Hvidtfeldt Larsen, A. M., The Three-Dimensional PWR

  17. A. E. K. Ris Ris-M-l1 Title and author^*)

    E-Print Network [OSTI]

    Simulator ANDYOAP 30 6.2 The PWR Power Plant Simulator 31 6.3 A l-Dime^sional BWR Plant Dynamic Model. PWR Reports 51 #12;- 2 - 1. General Introduction The work of the Department of Reactor Technology

  18. RIS-M-2357 MULTILEVEL FLOW MODELLING OF PROCESS

    E-Print Network [OSTI]

    of complex systems. A model of a nuclear power plant (PWR) is presented in the paper for illustration. Due SPECIFICATIONS 19 A MULTILEVEL FLOW MODEL OF A PWR 22 APPLICATIONS OF MULTILEVEL FLOW MODELS 24 ACKNOWLEDGEMENTS

  19. Connecticut Nuclear Profile - Millstone

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

    vnd.ms-excel" 3,"1,233","9,336",86.4,"PWR","applicationvnd.ms-excel","applicationvnd.ms-excel" ,"2,103","16,750",90.9 "Data for 2010" "PWR Pressurized Light Water Reactor."...

  20. Georgia Nuclear Profile - Vogtle

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

    vnd.ms-excel" 2,"1,152","9,363",92.8,"PWR","applicationvnd.ms-excel","applicationvnd.ms-excel" ,"2,302","19,610",97.2 "Data for 2010" "PWR Pressurized Light Water Reactor."...

  1. New York Nuclear Profile - Indian Point

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

    vnd.ms-excel" 3,"1,040","8,995",98.7,"PWR","applicationvnd.ms-excel","applicationvnd.ms-excel" ,"2,063","16,321",90.3 "Data for 2010" "PWR Pressurized Light Water Reactor."...

  2. South Carolina Nuclear Profile - Oconee

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

    vnd.ms-excel" 3,846,"6,779",91.5,"PWR","applicationvnd.ms-excel","applicationvnd.ms-excel" ,"2,538","20,943",94.2 "Data for 2010" "PWR Pressurized Light Water Reactor."...

  3. Transportation Issues and Resolutions Compilation of Laboratory...

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

    Phase I Ring Compression Testing of High Burnup Cladding Embrittlement and DBTT of High-Burnup PWR Fuel Cladding Alloys Effects of Multiple Drying Cycles on HBU PWR Cladding Alloys...

  4. Energy and Security in Northeast Asia: Proposals for Nuclear Cooperation

    E-Print Network [OSTI]

    Kaneko, Kumao; Suzuki, Atsuyuki; Choi, Jor-Shan; Fei, Edward

    1998-01-01T23:59:59.000Z

    PWR based on ABB-Combustion Engineering's System 80 design;modified the ABB-Combustion Engineering's System 80 PWR, andElectric, ABB-Combustion engineering and many smaller

  5. Energy efficient data centers

    E-Print Network [OSTI]

    Tschudi, William; Xu, Tengfang; Sartor, Dale; Koomey, Jon; Nordman, Bruce; Sezgen, Osman

    2004-01-01T23:59:59.000Z

    contribution to demand growth Units Msf W/sf GW TWh Totalcontribution to demand growth Units Msf W/sf GW TWh TotalPower Density (W/SF) Avg Pwr Demand (KW) Peak Pwr Demand (

  6. Supplementary Data to Global risk of radioactive fallout after nuclear reactor accidents

    E-Print Network [OSTI]

    Meskhidze, Nicholas

    Afrika Koeberg 1 PWR operational 900 944 04.04.1984 (14.08.2024) 127.092 Koeberg 2 PWR operational 900 under construction since 2010 610 650 ­ ­ ­ Daya Bay (Guangdong) 1 PWR operational 944 984 31.08.1993 (06.05.2034) 93.096 Daya Bay (Guangdong) 2 PWR operational 944 984 07.02.1994 (31.01.2034) 91

  7. ICLU RFP December 2011 Page 1 of 12

    E-Print Network [OSTI]

    December 2011 Page 3 of 12 Attachments 1 to 7 of Schedule A Cisco 1841 Nortel 4550T-PWR x2 67 Users Single mode AA1419002 24 x Multi Mode AA1419001 10 servers Parktown Campus Gables End Nortel 4548GT-PWR 15 Users Knockando (RES) Nortel 4526GT PWR x2 Nortel BPS2000 x1 Nortel 4548GT PWR x1 360 Users

  8. January 2005 Sun Mon Tue Wed Thu Fri Sat

    E-Print Network [OSTI]

    Wechsler, Risa H.

    signoff Intermediate No Chopper bulk pwr supplies on Regen HER arc 11 Some RF Processing RF processing RFCAV HER septum flow pwr dip actually on pipe PBL pwr tap NDR RF, T-gun 4-3 NIRP 4-3 NIRP Mini-ROD LER 4-4 NIRP PPS, SDR klys Bldg 685 pwr out T3018K59 RF 12-5, 12-6 flow swtch SDR septum 18 beam losses

  9. 0.5 -1.0 GHZ-3 1.25-1.75GHZ

    E-Print Network [OSTI]

    . NOTES ALFA MONITOR ALFA 6A ALFA "7A" PWR SPLITTER FIBER RECEIVER RACK 5 Ortel 10450 .01-6 GHz 1,2,...11d -20 -20 -20 -20 GPIB READ- OUT -10 HP4412 PWR HEAD -10 HP4412 PWR HEAD FROM POL B IF AMP (RACK 6) DUAL CHANNEL POWER METER #1 HP E4419A RACK 5 LEFT HAND PWR METER TRANSFER (REVERSING) SWITCH: EXAMPLE: if2 "if2

  10. CMAD IV 11/14/96 Information Security

    E-Print Network [OSTI]

    California at Davis, University of

    utilities, power pools, vendors etc.. #12;CMAD IV 11/14/96 #12; #12; GridCo LineCo PoolCo Energy Merchant INFO INFO INFO $ $ $ PWR PWR PWR #12;CMAD IV 11/14/96 "Future" Is At Hand · Federal Energy Regulatory protection and audit practices inadequate. · Internal priorities limiting attention to security concerns

  11. Ranger: CircumstancesRanger: Circumstances, Events, Legacy, g y

    E-Print Network [OSTI]

    RA 3: Mirror image m/c, missed Moon · RA-4: Main pwr. short at Agena separation · RA 5: Main pwr lost; 10 32 screw overheat· RA-5: Main pwr. lost; 10-32 screw overheat · RA-6: Plasma short circuit

  12. Nuclear Resonance Fluorescence to Measure Plutonium Mass in Spent Nuclear Fuel

    E-Print Network [OSTI]

    Ludewigt, Bernhard A

    2011-01-01T23:59:59.000Z

    NJOY NRES NRFXSSI NRF PDF PWR QM RIPL SNF TD TTB wt. % XCOMrotation angle for typical 17x17 PWR assembly. The source ofux. Assembly IE Type 17x17 PWR 3.1 9x9 BWR 1.91 VVER440 M U

  13. Modeling and Optimization of PEMFC Systems and its Application to Direct Hydrogen Fuel Cell Vehicles

    E-Print Network [OSTI]

    Zhao, Hengbing; Burke, Andy

    2008-01-01T23:59:59.000Z

    s Comp . Speed as_exp _sh_pwr Goto 1 Torque = deltaPower / w= Inertia * dw/dt Add as_comp _sh_pwr Goto as_net _sh_pwr Goto 2 Figure 19 Compressor speed calculation

  14. Class Name: OS101 W07 Class Key: I25425H764 You will need

    E-Print Network [OSTI]

    Kudela, Raphael M.

    by pressing the PWR/JOIN button. Join ­ the response pads automatically search for a class roster to join to join, turn on the response pad and press the PWR/JOIN button. Manually Join ­ to manually join a class, turn on the response pad and press the PWR/JOIN button twice. Join: appears on the LCD screen. Type

  15. ESAIM: PROCEEDINGS, Vol. ?, 2012, 1-10 Editors: Will be set by the publisher

    E-Print Network [OSTI]

    Boyer, Edmond

    variables (like temperature) within the reactor. Let us rst present the normal functioning of a PWR (Pressurized Water Reactor) see Fig. 1. In a PWR, the primary coolant (water) is pumped under high pressure circuit of a PWR. where steam is generated and ows to a turbine which, in turn, spins an electric

  16. Compiling Esterel Better Circuits

    E-Print Network [OSTI]

    _state = STANDBY_PWR_DN; end else if (valid_diag_window | ibuf_full | jmp_e) begin next_state = cur_state; end else STANDBY_PWR_DN: begin if(!pcsu_powerdown | jmp_e ) begin next_state = IDLE; end else next_state = STANDBY_PWR

  17. An Assessment of 238Puand239+240

    E-Print Network [OSTI]

    An Assessment of 238Puand239+240 Puinthe Primary Cooling Waterofa PWR Q. Chen, S.P. Nielsen and S+240puinthe Primary Cooling Water of a PWR Q. Chen1 , S.P. Nielsen1 and S. Duniec2 1 Risø National Laboratory of transuranics in these devices. Unit 2 (PWR) of the Ringhals power plant was investigated in this study which

  18. Multi-band high efficiency power amplifier

    E-Print Network [OSTI]

    Besprozvanny, Randy-Alexander Randolph

    2011-01-01T23:59:59.000Z

    $FPRJ p3: Freq = 0.75 GHz Pwr = 28 dBm p1: Freq = 1.25 GHzTime_Output p2 p1 p3 p1: Freq = 2 GHz Pwr = 30 dBm Time (ns)p2: Freq = 2 GHz Pwr = 30 dBm Current Waveform (mA) Voltage

  19. THERMAL IMPACT OF WASTE EMPLACEMENT AND SURFACE COOLING ASSOCIATED WITH GEOLOGIC DISPOSAL OF NUCLEAR WASTE

    E-Print Network [OSTI]

    Wang, J.S.Y.

    2010-01-01T23:59:59.000Z

    f o r 1 0 - y e a r - o l d PWR SF. I n i t i a l Heat G e nt nuclear fuel cycles for a PWR. Decay heat power for d i fKBS LWR MOX NRC NWTS ONWI OWI PWR RH-TRU WIPP b o i l i n g

  20. A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01T23:59:59.000Z

    W NA C C OK/W NA C C OK/W PWR secondary-to- B. Metal—waterfrom core binding OK(WDB) 3. PWR steam a. b. c. Pump Ap S(C Post-CHF heat transfer PWR reflood heat transfer OK (CUP)/

  1. Laboratoire des Solides Irradis, UMR 7642 Laboratoire des Solides Irradis Tl. : 33 1 69 33 44 80

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43 2.3 PWR Water Radiolysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 62 3 hal-00841142,version1-19Aug2013 #12;CONTENTS 3 Corrosion issues of 316L under Primary PWR Conditions 69 3.1 The Oxide on 316L Formed under Primary PWR Water . . . . . . . . . . . . . . . . . 71 3.1.A

  2. U238 (1) (2)

    E-Print Network [OSTI]

    Hong, Deog Ki

    , . . 6.1. 6. 1 . 3.3% (PWR) 33,000MWd/MT 150 . . 50MWe PWR 33,000MWd/MT, 32.5% 0.8 3.3% 27,271 . 26,011 #12; 0.83%. 6. 1 (PWR

  3. NuSCR Quick Checker Reformation of Quick Checker for verifying NuSCR

    E-Print Network [OSTI]

    FOD, FSM, TTS, SDT . 1 KNICS RPS(Reactor Protection System) BP (Bistable Process) g_VAR_OVER_PWR . . , . , . , . . 1 g_VAR_OVER_PWR FOD FSM history variable node . FSM . NuSCR . FSM . . . Timed history variable node TTS . TTS FSM . . 2 Waiting Trip k_VAR_OVER_PWR

  4. SmartCast - Novel Textile Sensors for Embedded Pressure Sensing of Orthopedic Casts

    E-Print Network [OSTI]

    Danilovic, Andrew

    2013-01-01T23:59:59.000Z

    set_sleep_mode(SLEEP_MODE_PWR_DOWN); //enter Power-down Mode4 #define SEL2_PB0 8 #define SEL1_PD7 7 #define PWR_CTRL_PINPD5 #define PWR_CTRL_SD_CARD 6 //SmartCast bitFields for

  5. CERTS Microgrid Laboratory Test Bed - PIER Final Project Report

    E-Print Network [OSTI]

    Eto, Joseph H.

    2008-01-01T23:59:59.000Z

    N1 Relay 3 (also Sheet 10) Ia PWR Ib Ic In F to CB32, andalso Sheet 10) Ia Ib Ic In PWR F to CB42 and Microsource A2DAS_24dcPOS Vc com Vs E Relay PWR Vsn Ia A52 B52 C52 (5A) (

  6. Energy Efficient Computing with the Low Power, Energy Aware Processing (LEAP) Architecture

    E-Print Network [OSTI]

    McIntire, Dustin Hale

    2012-01-01T23:59:59.000Z

    and communications module PWR RT618+RT620 Power module andwith a power supply module (PWR module), and a high fidelityAUX_BUS_CHAIN_OUT EMAP2 +MAG_PWR V5BUS V33BUS V9BUS -V9BUS

  7. ANALYSIS AND APPLICATION OF INDUCTANCE IN CLOCK DISTRIBUTION NETWORKS

    E-Print Network [OSTI]

    Hu, Xuchu

    2012-01-01T23:59:59.000Z

    17ps on average. Non-resonant CDN Sink MA S.Cap MA mm 2 Pwr.1 Pwr. 2 Resonant CDN LA Skew MA P wr. 2 P wr.mm 2 mW mW ps m Avg. 1431 Pwr. 1 : Switched capacitance CDN

  8. 2.1E BDL Summary

    E-Print Network [OSTI]

    Winkelmann, F.C.

    2010-01-01T23:59:59.000Z

    TOWERAIR) ' OPEN-CENT-COND-PWR(0.3;0.0 to 1.0 Btu/Btu) •changed i n 2.1E OPEN-REC-COND-PWR(0.03;0.0 to 1.0Btu/Btu) HERM-CENT-COND-PWR(0.3;0.0 to 1.0 Btu/Btu) •

  9. Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay

    E-Print Network [OSTI]

    Quiter, Brian

    2012-01-01T23:59:59.000Z

    the mass of 239 Pu in a 17x17 PWR fuel assembly with 45 GWd/center of 40 GWd/MTU burn-up PWR fuel assembly with coolingrate for the 11 y cooled PWR fuel was used as a source term

  10. Rapid Mixing and Security of Chaum's Visual Electronic Voting

    E-Print Network [OSTI]

    Rivest, Ronald L.

    . Wybrzeze Wyspia´nskiego 27 50-370 Wroclaw, Poland gomulkie@im.pwr.wroc.pl klonowsk@ulam.im.pwr.wroc.pl mirekk@im.pwr.wroc.pl 2 CC Signet Abstract. Recently, David Chaum proposed an electronic voting scheme

  11. MO. RIV. MO. LEAM Reliability Numerical Analysis PAGE 1 OF 6

    E-Print Network [OSTI]

    Rathbun, Julie A.

    Components Derating IC's Digital 40 to 60% of fan out Transistors 1 to 10% of rated PWR Capacitors 1 to 5O% of rated voltage Resistors 1 to 35% of rated PWR Diodes 1 to 20% of rated PWR #12;NO. ltiV. NO. ATM 1030

  12. CALIFORNIA SOLAR DATA MANUAL

    E-Print Network [OSTI]

    Berdahl, P.

    2010-01-01T23:59:59.000Z

    o .1Q O. ~3 :l.62 SAUGUS PWR PL N~ I SCIHJA SEARSv XL LEN37H N3423 N3721 SAUGUS PWR PL NO I SURSVILlE LAKE SIERRAV!INA. FAA AP SALT SPRINGS PWR HOUSE SAN BERNARDINO eo HOSP

  13. w Ris Report No. 318 J-Danish Atomic Energy Commission

    E-Print Network [OSTI]

    equipment and computing time for simulation of a control system for the power plant. A diagram showing Description of the Real Time PWR Power Plant Model PWR-PLASIM by P. la Cour Christensen November 1974 M n Risø Department of Reactor Technology Abstract A description is given of a PWR power plant model

  14. The Energy Technologist

    E-Print Network [OSTI]

    Tumber, A. J.; Molczan, T. J.

    1981-01-01T23:59:59.000Z

    THE ENERGY TECHNOLOGIST Adrian J. Tumber Mohawk College Hamilton, Ontario, Canada ABSTRACT The special need for personnel to assist with energy management in the industrial and commercial/institutional sectors has resulted in a new three.... Balancing our energy budget, if that is ever to "'l be achieved, will be as a result of i~1 increasing our energy supplies, and by reducing consumption, that is, by conservation. Ted J. Molczan Canada Packers Inc. Toronto, Ontario, Canada...

  15. National Coexistence is Our Bull Durham: Revisiting "The Indian Today"

    E-Print Network [OSTI]

    Valandra, Edward C.

    2006-03-01T23:59:59.000Z

    ); and the development of the St. Lawrence Seaway Project (Mohawk Nation), and their activism has been well documented.5 The Haudenosaunee Confederacy's proactive opposition to these colonial assaults caught the attention of Native Country, indicating that other... social movement to characterize the Indian movement is a deliberate choice. In their study of four different social or protest movements—unemployed workers, industrial workers, civil rights, and welfare rights—Frances F. Piven and Richard A. Cloward...

  16. Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

    SciTech Connect (OSTI)

    Woo, H.H.; Lu, S.C.

    1981-09-15T23:59:59.000Z

    Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design.

  17. http://www.kasahara.cs.waseda.ac.jp kasahara@waseda.jp

    E-Print Network [OSTI]

    Kasahara, Hironori

    Servers IBM Power4,5,5+,6, HPCS 10PFLOP(Power7) S Ni (S T1 T2) R kNEDO Sun Niagara(SparcT1,T2), Rock M #12; /NEDO (2005.7 2008.3)** OSCAR: Optimally Scheduled Advanced Multiprocessor DVD swim su2cor hydro2d mgrid applu turb3d apsi fpppp wave5 wupwise swim mgrid applu sixtrack apsi SPEC CFP

  18. Compiling Esterel Better Circuits

    E-Print Network [OSTI]

    _diag_window) begin next_state = STANDBY_PWR_DN; end else if (valid_diag_window | ibuf_full | jmp_e) begin next_state ; end STANDBY_PWR_DN: begin if(!pcsu_powerdown | jmp_e ) begin next_state = IDLE; end else next_state = STANDBY_PWR_DN; end default: next_state = 7'bx; endcase loop await case [icu_miss and not cacheable] do

  19. High Level Synthesis from Synchronous Language Es

    E-Print Network [OSTI]

    parallel_case IDLE: begin if (pcsu_powerdown & !jmp_e & !valid_diag_window) begin next_state = STANDBY_PWR_WD: begin if(normal_ack| error_ack) begin next_state = IDLE; end else next_state = cur_state ; end STANDBY_PWR_DN: begin if(!pcsu_powerdown | jmp_e ) begin next_state = IDLE; end else next_state = STANDBY_PWR_DN; end

  20. High Level Synthesis from the Synchronous Language Esterel 1068.003

    E-Print Network [OSTI]

    parallel_case IDLE: begin if (pcsu_powerdown & !jmp_e & !valid_diag_window) begin next_state = STANDBY_PWR_WD: begin if(normal_ack| error_ack) begin next_state = IDLE; end else next_state = cur_state ; end STANDBY_PWR_DN: begin if(!pcsu_powerdown | jmp_e ) begin next_state = IDLE; end else next_state = STANDBY_PWR_DN; end

  1. High Level Synthesis from the Synchronous Language Esterel 1068.003

    E-Print Network [OSTI]

    _case IDLE: begin if (pcsu_powerdown & !jmp_e & !valid_diag_window) begin next_state = STANDBY_PWR_DN; end_WD: begin if(normal_ack| error_ack) begin next_state = IDLE; end else next_state = cur_state ; end STANDBY_PWR_DN: begin if(!pcsu_powerdown | jmp_e ) begin next_state = IDLE; end else next_state = STANDBY_PWR_DN; end

  2. Compiling Esterel Better Circuits

    E-Print Network [OSTI]

    _powerdown & !jmp_e & !valid_diag_window) begin next_state = STANDBY_PWR_DN; end else if (valid_diag_window | ibuf_state = IDLE; end else next_state = cur_state ; end STANDBY_PWR_DN: begin if(!pcsu_powerdown | jmp_e ) begin next_state = IDLE; end else next_state = STANDBY_PWR_DN; end default: next_state = 7'bx; endcase loop

  3. High Level Synthesis from the Synchronous Language Esterel 1068.00

    E-Print Network [OSTI]

    parallel_case IDLE: begin if (pcsu_powerdown & !jmp_e & !valid_diag_window) begin next_state = STANDBY_PWR_WD: begin if(normal_ack| error_ack) begin next_state = IDLE; end else next_state = cur_state ; end STANDBY_PWR_DN: begin if(!pcsu_powerdown | jmp_e ) begin next_state = IDLE; end else next_state = STANDBY_PWR_DN; end

  4. 1.25-1.75GHZ 1.5 -2.0 GHZ

    E-Print Network [OSTI]

    . SHEET 4) FP FP -10 -10 RP RP 8-T0-1 SWITCH, POL A (SIGNAL SELECTOR FOR PWR METER 2a) 8-T0-1 SWITCH, POL B (SIGNAL SELECTOR FOR PWR METER 2b) FROM AMPS 1-8 POL. B FROM AMPS 1-8 POL. A COUGAR AC582C AMPS 20 RACK 5 COMPUTER READ OUT NOT USED RACK 5 RIGHT HAND PWR METER DUAL-CHANNEL POWER METER #2 HP E4419B

  5. THE ROLE OF BUFFER GASES IN OPTOAOOUSTIC SPECTROSCOPY

    E-Print Network [OSTI]

    Thomas III, L.J.

    2011-01-01T23:59:59.000Z

    at 160 Torr Temp. 22.5° C e Pwr. abs. -1.6mW o c: N I I enTorr c ~O""'"' Temp.21.2°C Pwr. abs. -1.6 mW a c: en N 0 p4 at 760 Torr Temp. 22.4ac Pwr. abs. · 2.6mW I I I I I I I o

  6. Crosstalk-Aware Anycast Routing and Wavelength Assignment in Optical WDM Networks Balagangadhar G. Bathula1

    E-Print Network [OSTI]

    Bathula, Balagangadhar G

    | while A do for h PATH(d i ) do PWR(h, i) PWR(h - 1, i) - LOSS(h, i) ASE(h, i) ASE(h - 1, i) + ASE.SW(i) XT(h, i) XT(h, i) + XT.SW(i) end OSNR(d i , i) = PWR(d i ,i) (ASE(d i ,i)+XT(d i ,i)) if OSNR(d i

  7. V >K S l O O ' f g -f RisO-M-2640

    E-Print Network [OSTI]

    V �>K S l O O ' f g - f o se ii cø S � «T"? RisO-M-2640 Simulation Model of a PWR Power Plant Niels PLANT Niels Larsen Abstract. A simulation model of a hypothetical PWR power plant is described. A large - 1. INTRODUCTION A simulation model of a hypothetical PWR power plant has been constructed. The model

  8. Data:D6eb72d7-15ef-4a25-9afa-66996380b15f | Open Energy Information

    Open Energy Info (EERE)

    Pwr Dist Effective date: 20120101 End date if known: Rate name: Urban Residential - Gas Water Heater Sector: Residential Description: Source or reference: ISU Documentation...

  9. Sandia National Laboratories: Concentrating Solar Power

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

    by PWR in cooperation with the Department ... Excellence Award in the 2012 Facilities Environmental, Safety and Health Go Green Initiative On December 19, 2012, in...

  10. CASL Symposium: Celebrating the Past - Visualizing the Future

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

    on helping to address key industry challenges related to pressurized water reactor (PWR) reactor core performance in normal and accident conditions. VERA has been deployed to...

  11. Sandia Energy - EC Publications

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

    -Finite element analysis for generic salt repository (waste package size up to 32-PWR) -"Open" disposal concept development: shale unbackfilled, sedimentary backfilled, and...

  12. Sandia Energy - EC Publications

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

    - Finite element analysis for generic salt repository (waste package size up to 32-PWR) - "Open" disposal concept development: shale unbackfilled, sedimentary backfilled, and...

  13. OutageMapURL Phases Energy Services

    Open Energy Info (EERE)

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  14. Product Description (<100) Total

    E-Print Network [OSTI]

    Berns, Hans-Gerd

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  15. affecting reactor accident: Topics by E-print Network

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

    METEOROLOGY; NUCLEAR POWER PLANTS; P CODES; PWR TYPE REACTORS; RADIATION 2 Does Daylight Savings Time Affect Traffic Accidents? Texas A&M University - TxSpace Summary: This...

  16. Data:34dab3ad-9128-402a-8a7e-5b31d61e0d79 | Open Energy Information

    Open Energy Info (EERE)

    Information Utility name: Singing River Elec Pwr Assn (Mississippi) Effective date: 20091204 End date if known: Rate name: Security Lighting HPS 400 W w Pole Sector: Lighting...

  17. Data:C6cc60b8-c6f5-467c-82bc-be54db91f9fa | Open Energy Information

    Open Energy Info (EERE)

    >> Basic Information Utility name: Cuming County Public Pwr Dist Effective date: 20111214 End date if known: Rate name: Large Power Rate Sector: Industrial Description: Source...

  18. Data:05070b10-3c15-4871-8021-7ea15a4bea96 | Open Energy Information

    Open Energy Info (EERE)

    >> Basic Information Utility name: Cuming County Public Pwr Dist Effective date: 20111214 End date if known: Rate name: High Voltage Delivery Rate Sector: Industrial...

  19. Data:1b0f0ce3-ceb3-4295-93e4-fa2b77c76cb4 | Open Energy Information

    Open Energy Info (EERE)

    >> Basic Information Utility name: Cuming County Public Pwr Dist Effective date: 20111214 End date if known: Rate name: Security Lighting Unmetered Lights 250W MV Sector:...

  20. Data:16337ee5-241c-435d-8a06-04cff7fb9a92 | Open Energy Information

    Open Energy Info (EERE)

    >> Basic Information Utility name: Cuming County Public Pwr Dist Effective date: 20111214 End date if known: Rate name: Security Lighting Unmetered Lights 100W HPS Sector:...

  1. Data:18e92d7d-ab66-4e91-9e74-9ff84d6e479e | Open Energy Information

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    Next >> Basic Information Utility name: Cuming County Public Pwr Dist Effective date: 20121214 End date if known: Rate name: *Municipal Commercial Demand Rate Single Phase V2...

  2. Data:39a41a1e-9265-4f35-872c-780c287a6458 | Open Energy Information

    Open Energy Info (EERE)

    Information Utility name: Singing River Elec Pwr Assn (Mississippi) Effective date: 20091204 End date if known: Rate name: Security Lighting HPS 150 W Flood w Pole Sector:...

  3. Data:2a5c2d50-a8ca-41c8-a1ec-4cca9b37f532 | Open Energy Information

    Open Energy Info (EERE)

    >> Basic Information Utility name: Cuming County Public Pwr Dist Effective date: 20111214 End date if known: Rate name: Municipal SpaceWater Heating Rate Single Phase Sector:...

  4. Data:188ef1da-71c2-489f-ac19-5b3da3149f9c | Open Energy Information

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    >> Basic Information Utility name: Cuming County Public Pwr Dist Effective date: 20111214 End date if known: Rate name: Municipal Commercial Rate Single Phase Sector:...

  5. Data:5b44a54d-2284-43db-acbf-981f07d169e3 | Open Energy Information

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    Information Utility name: Singing River Elec Pwr Assn (Mississippi) Effective date: 20091204 End date if known: Rate name: Security Lighting MH 1000 W Sector: Lighting...

  6. V

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    Livennore National Laboratory (LLNL) for the USHRC (U) which employs the Monte Carlo simulation, has been modified to simulate th. . . ' ' history of PWR feedwater lines,...

  7. "The whole is more than the sum of the parts" Aristotle

    E-Print Network [OSTI]

    Sóbester, András

    vessel ·Operation like nuclear battery ·Shielded inaccessible container ·Low cost compared to PWR ·Zero

  8. Novel Laser-Based Manufacturing of nano-LiFePO4-Based Materials for High Power Li Ion Batteries

    E-Print Network [OSTI]

    Horne, Craig R.; Jaiswal, Abhishek; Chang, On; Crane, S.; Doeff, Marca M.; Wang, Emile

    2006-01-01T23:59:59.000Z

    NanoParticle Manufacturing (NPM™), has been used tomaterials synthesized by the NPM™ process (branded as nPWR™)phosphoric acid into an NPM™ reactor. The powder collected

  9. Data:996ecf92-0300-4f93-be1c-1845621f5f1b | Open Energy Information

    Open Energy Info (EERE)

    Information 2. Demand 3. Energy << Previous 1 2 3 Next >> Basic Information Utility name: Puerto Rico Electric Pwr Authority Effective date: End date if known: Rate name: 312 -...

  10. DOE/ID-Number

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    PW Procedure Writers PWR Pressurized Water Reactor RHR Reactor Heat Removal RCS Reactor Cooling System RMS Radiation Monitoring System RO Reactor Operator RSF Remote Shutdown...

  11. Data:0af43158-f79b-42f3-85f1-5bd7dce06f4e | Open Energy Information

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    Pwr Dist Effective date: 20130101 End date if known: Rate name: RATE 76,77 Municipal Pumping Service Sector: Commercial Description: AVAILABLE: In the lease towns of Clearwater,...

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    Public Pwr Dist Effective date: 20111214 End date if known: Rate name: Municipal Pumping Rate Sector: Commercial Description: Source or reference: Illinois State University...

  13. Data:F56e2b50-2f4d-4b6d-81d3-31f03cd7d67b | Open Energy Information

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    Rural Pub Pwr Dist Effective date: 20120101 End date if known: Rate name: Municipal Pumping Service Sector: Residential Description: Source or reference: ISU Documentation...

  14. Data:Deb1c116-8c09-4685-98b7-2aa21a790238 | Open Energy Information

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    Public Pwr Dist Effective date: 20130101 End date if known: Rate name: 36- Single Phase Farm Residenital - PV DG (solar) Sector: Residential Description: Source or reference:...

  15. Mohegan Tribal Utility Auth | Open Energy Information

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    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directedAnnual SiteofEvaluatingGroup |JilinLu anMicrogreen PolymersModular Energy Devices IncMohawk

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  19. Wenchuan Huaming Power Development Co Ltd | Open Energy Information

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  20. Wenchuan Longfa Electric Power Co Ltd | Open Energy Information

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