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

    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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power.pdf11-161-LNG |September 15, 2010Energy SeriesPOSTKATRINAMEMOATTACHMENTS.pdfPOlicy-23-18190

  3. 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.

  4. 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk Municipal Comm Jump to: navigation,

  5. E-Print Network 3.0 - akwesasne mohawk young Sample Search Results

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

    haplotype, ancient DNA, migra- tion, admixture, Arikara, Mohawk, Sioux, Chippewa, Cherokee, Ojibwa, Pawnee... the Mohawk, Arikara, and Sioux populations than between any of...

  6. 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. ”

  7. 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

  8. ,"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,...

  9. DOE - Office of Legacy Management -- Niagara Falls Storage Site...

    Office of Legacy Management (LM)

    of the Niagara Falls, New York, Storage Site NY.17-1 - AEC Memorandum; Malone to Smith; Subject: Monthly Progress Report for April; April 24, 1951. Attachment: Tonawanda...

  10. EIS-0153: Niagara Import Point Project

    Broader source: Energy.gov [DOE]

    The Federal Energy Regulatory Commission prepared this statement to assess the environmental impacts of the proposed Niagara Import Point project that would construct an interstate natural gas pipeline to transport gas from Canada and domestic sources to the Northeastern United States market. The U.S. Department of Energy's Office of Fossil Energy was a cooperating agency during statement development and adopted this statement on 6/15/1990.

  11. ,"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...

  12. 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

  13. Niagara Falls Storage Site environmental monitoring report. Calendar year 1983

    SciTech Connect (OSTI)

    Not Available

    1984-07-01T23:59:59.000Z

    During 1983, an environmental monitoring program was continued at the Niagara Falls Storage Site, a United States Department of Energy (DOE) surplus facility located in Niagara County, New York presently used for the storage of radioactive residues, contaminated soils and rubble. The monitoring program at NFSS measures radon concentrations in air, uranium and radium concentrations in surface water, groundwater, and sediments, and external gamma exposure rates. Radiation doses to the public are also calculated. Environmental samples collected are analyzed to determine compliance with applicable standards. Comparison of 1983 monitoring results with 1982 results shows a significant decrease in radon levels at almost every monitoring location. External gamma exposure rates also showed a general decrease. 9 references, 10 figures, 11 tables

  14. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartment ofDepartment ofof EnergyYou$0.C. 20545*.MSE Cores" _ ,' ,:.' :r-2NewNiagara

  15. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartment ofDepartment ofof EnergyYou areDowntown SiteOhioMissouriMaywoodNY 17 Niagara

  16. 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.

  17. PWR AXIAL BURNUP PROFILE ANALYSIS

    SciTech Connect (OSTI)

    J.M. Acaglione

    2003-09-17T23:59:59.000Z

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

  18. 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....

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. Niagara Falls Storage Site environmental surveillance report for calendar year 1993

    SciTech Connect (OSTI)

    Not Available

    1994-06-01T23:59:59.000Z

    This report summarizes the results of environmental surveillance activities conducted at the Niagara Falls Storage Site (NFSS) during calendar year 1993. It includes an overview of site operations, the basis for radiological and nonradiological monitoring, a summary of the results, and the estimated dose to the offsite population. Environmental surveillance activities were conducted in accordance with the site environmental monitoring plan, which describes the rationale and design criteria for the surveillance program, the frequency of sampling and analysis, specific sampling and analysis procedures, and quality assurance requirements. NFSS is in compliance with National Emission Standards for Hazardous Air Pollutants (NESHAPs) Subpart H of the Clean Air Act as well as the requirements of the National Pollutant Discharge Elimination System (NPDES) under the Clean Water Act. Located in northwestern New York, the site covers 191 acres. From 1944 to the present, the primary use of NFSS has been storage of radioactive residues that were by-products of uranium production. Most onsite areas of residual radioactivity above regulatory guidelines were remediated during the early 1980s. Additional isolated areas of onsite contamination were remediated in 1989, and the materials were consolidated into the waste containment structure in 1991. Remediation of the site has now been completed.

  4. Improving the economics of PWR cores

    SciTech Connect (OSTI)

    Ober, T.G. [Entergy Operations, Jackson, MS (United States)

    1996-08-01T23:59:59.000Z

    Economic fuel cycles have become of paramount importance to the nuclear power industry due to the increasing impact of deregulation and competition. This paper describes the PWR core design techniques being employed at Entergy in the quest to meet the ever-decreasing fuel cost targets for these units.

  5. 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.

  6. 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.

  7. 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.

  8. 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

  9. 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:...

  10. 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

  11. 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)

  12. Radiation embrittlement of PWR vessel supports

    SciTech Connect (OSTI)

    Cheverton, R.D.; Robinson, G.C.; Pennell, W.E.; Nanstad, R.K.

    1989-01-01T23:59:59.000Z

    Several studies pertaining to radiation damage of PWR vessel supports were conducted between 1978 and 1987. During this period, apparently there was no reason to believe that low-temperature (<100/degree/C) MTR embrittlement data were not appropriate for evaluating embrittlement of PWR vessel supports. However, late in 1986, data from the High Flux Isotope Reactor (HFIR) vessel surveillance program indicated that the embrittlement rates of the several HFIR vessel materials (A212-B, A350-LF3, A105-II) were substantially greater than anticipated on the basis of MTR data. Further evaluation of the HFIR data suggested that a fluence-rate effect was responsible for the apparent discrepancy, and shortly thereafter it became apparent that this rate effect was applicable to the evaluation of LWR vessel supports. As a result, the Nuclear Regulatory Commission (NRC) requested that the Oak Ridge National Laboratory (ORNL) evaluate the impact of the apparent embrittlement rate effect on the integrity of light-water-reactor (LWR) vessel supports. The purpose of the study was to provide an indication of whether the integrity of reactor vessel supports is likely to be challenged by radiation-induced embrittlement. The scope of the evaluation included correlation of the HFIR data for application to the evaluation of LWR vessel supports; a survey and cursory evaluation of all US LWR vessel support designs, selection of two plants for specific-plant evaluation, and a specific-plant evaluation of both plants to determine critical flaw sizes for their vessel supports. 19 refs., 8 figs., 2 tabs.

  13. 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

  14. 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.

  15. 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.

  16. Timing analysis of PWR fuel pin failures

    SciTech Connect (OSTI)

    Jones, K.R.; Wade, N.L.; Katsma, K.R.; Siefken, L.J. (EG and G Idaho, Inc., Idaho Falls, ID (United States)); Straka, M. (Halliburton NUS, Idaho Falls, ID (United States))

    1992-09-01T23:59:59.000Z

    This report discusses research conducted to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B W) design (Oconee) and a Westinghouse (W) four-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin burnup, axial peaking factor, break size, emergency core cooling system availability, and main coolant pump trip on these times. The analysis was performed using the following codes: FRAPCON-2, for the calculation of steady-state fuel behavior; SCDAP/RELAP5/MOD3 and TRACPF1/MOD1, for the calculation of the transient thermal-hydraulic conditions in the reactor system; and FRAP-T6, for the calculation of transient fuel behavior. In addition to the calculation of fuel pin failure timing, this analysis provides a comparison of the predicted results of SCDAP/RELAP5/MOD3 and TRAC-PF1/MOD1 for large-break LOCA analysis. Using SCDAP/RELAP5/MOD3 thermal-hydraulic data, the shortest time intervals calculated between initiation of containment isolation and fuel pin failure are 10.4 seconds and 19.1 seconds for the B W and W plants, respectively. Using data generated by TRAC-PF1/MOD1, the shortest intervals are 10.3 seconds and 29.1 seconds for the B W and W plants, respectively. These intervals are for a double-ended, offset-shear, cold leg break, using the technical specification maximum peaking factor and applied to fuel with maximum design burnup. Using peaking factors commensurate with actual burnups would result in longer intervals for both reactor designs. This document provides appendices K and L of this report which provide plots for the timing analysis of PWR fuel pin failures for Oconee and Seabrook respectively.

  17. 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.

  18. Examination of dissimilar metal welds in BWR and PWR piping

    SciTech Connect (OSTI)

    MacDonald, D.E. [Electric Power Research Inst., Charlotte, NC (United States). NDE Center

    1994-12-31T23:59:59.000Z

    This paper addresses dissimilar metal weld examinations at PWRS. Surveys were conducted to document the dissimilar metal weld configurations at PWR plants and to update the information known about dissimilar metal weld configurations at BWR plants. The experiences which BWR utilities have had with dissimilar metal weld examinations are documented and include: correct identification of IGSCC, indications thought to be IGSCC but were actually fabrication flaws, and difficulties encountered with the examination of dissimilar metal welds after stress improvement. An experimental program was conducted which verified that the longitudinal wave procedures developed for BWRs are also applicable to PWR designs.

  19. 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.

  20. Design report for the interim waste containment facility at the Niagara Falls Storage Site. [Surplus Facilities Management Program

    SciTech Connect (OSTI)

    Not Available

    1986-05-01T23:59:59.000Z

    Low-level radioactive residues from pitchblende processing and thorium- and radium-contaminated sand, soil, and building rubble are presently stored at the Niagara Falls Storage Site (NFSS) in Lewiston, New York. These residues and wastes derive from past NFSS operations and from similar operations at other sites in the United States conducted during the 1940s by the Manhattan Engineer District (MED) and subsequently by the Atomic Energy Commission (AEC). The US Department of Energy (DOE), successor to MED/AEC, is conducting remedial action at the NFSS under two programs: on-site work under the Surplus Facilities Managemnt Program and off-site cleanup of vicinity properties under the Formerly Utilized Sites Remedial Action Program. On-site remedial action consists of consolidating the residues and wastes within a designated waste containment area and constructing a waste containment facility to prevent contaminant migration. The service life of the system is 25 to 50 years. Near-term remedial action construction activities will not jeopardize or preclude implementation of any other remedial action alternative at a later date. Should DOE decide to extend the service life of the system, the waste containment area would be upgraded to provide a minimum service life of 200 years. This report describes the design for the containment system. Pertinent information on site geology and hydrology and on regional seismicity and meteorology is also provided. Engineering calculations and validated computer modeling studies based on site-specific and conservative parameters confirm the adequacy of the design for its intended purposes of waste containment and environmental protection.

  1. 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 on Google Bookmark EERE: Alternative Fuels Data CenterFranconia, Virginia: Energy Resources Jump to: navigation,Ohio: Energy ResourcesGordon, Alabama:5812144° Loading map...Pwr Line, Inc

  2. Twin County Electric 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating SolarElectric Coop, IncTipmontInformationKentucky)Bank,Turkmenistan:Pwr Assn Jump

  3. 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.

  4. 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.

  5. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartment ofDepartment ofof EnergyYou$0.C. 20545*.MSE Cores" _ ,' ,:.' :

  6. 21-PWR WASTE PACKAGE WITH ABSORBER PLATES LOADING CURVE EVALUATION

    SciTech Connect (OSTI)

    J.M. Scaglione

    2004-12-17T23:59:59.000Z

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

  7. 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocusOski Energy LLCPascoag Utility DistrictPea RiverAwarenessPwr

  8. Polk County 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocusOskiPhilips Color KineticsGrowth JumpPub Pwr Dist Jump to:

  9. North Central 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 on Google Bookmark EERE: Alternative Fuels Data CenterFranconia, Virginia: Energy ResourcesLoading map...(Utility Company) Jump to:City)Norristown,Braddock isStateCentral Public Pwr Dist

  10. Northeast Missouri El Pwr Coop | 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus AreaDataBusPFAN) | OpenIncNobleNorrisElecEnergy JumpEl Pwr

  11. Stanton 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 on Google Bookmark EERE: Alternative Fuels Data Center Home5b9fcbce19 No revisionEnvReviewNonInvasiveExplorationUT-g GrantAtlas (PACA RegionSpringview IISt. Mary'sStanislausPwr Dist

  12. Virgin Islands Wtr&Pwr Auth | 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating SolarElectric Coop,Save EnergyGlouster,Winside, Nebraska (Utility Company)Wtr&Pwr

  13. Elkhorn Rural 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Power Basics (The followingDirectLow CarbonOpen1Model |Rural Public Pwr Dist

  14. Central 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of Inspector GeneralDepartmentAUDIT REPORTOpenWendeGuoCatalyst Renewables JumpViewCentral Electric Coop,Pwr

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. Solving chemical and mechanical problems of PWR steam generators

    SciTech Connect (OSTI)

    Green, S.J.

    1987-07-01T23:59:59.000Z

    Steam generators in power plants, based on pressurized water reactors (PWRs), transfer heat from a primary coolant system (pressurized water) to a secondary coolant system. Primary coolant water is heated in the core and passes through the steam generator that transfers heat to the secondary coolant water to make steam. The steam then drives a turbine that turns an electric generator. Steam is condensed and returned to the steam generator as feedwater. Two types of PWR steam generators are in use: recirculating steam generators (RSGs) and once-through steam generators (OTSGs). Since most of the units are vertical, only vertical units are discussed in this article. Some vertical units have operated with a minimum of problems, while others have experienced a variety of corrosion and mechanically-induced problems that have caused unscheduled outages and expensive repairs.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

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

    SciTech Connect (OSTI)

    Putney, J.M.; Preece, R.J. [National Power, Leatherhead (GB). Technology and Environment Centre

    1993-06-01T23:59:59.000Z

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

  7. 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.

  8. 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.

  9. 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@csam.ucl.ac.be, Gevers@csam.ucl.ac.be 2 Electricite de France, Direction des Etudes et Recherches, 6 Quai Watier, F-78041 of a controller for the secondary circuit of a nu- clear Pressurized Water Reactor, leading to the conclu- sions

  10. 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

  11. 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.

  12. Assessment of void swelling in austenitic stainless steel PWR core internals.

    SciTech Connect (OSTI)

    Chung, H. M.; Energy Technology

    2006-01-31T23:59:59.000Z

    As many pressurized water reactors (PWRs) age and life extension of the aged plants is considered, void swelling behavior of austenitic stainless steel (SS) core internals has become the subject of increasing attention. In this report, the available database on void swelling and density change of austenitic SSs was critically reviewed. Irradiation conditions, test procedures, and microstructural characteristics were carefully examined, and key factors that are important to determine the relevance of the database to PWR conditions were evaluated. Most swelling data were obtained from steels irradiated in fast breeder reactors at temperatures >385 C and at dose rates that are orders of magnitude higher than PWR dose rates. Even for a given irradiation temperature and given steel, the integral effects of dose and dose rate on void swelling should not be separated. It is incorrect to extrapolate swelling data on the basis of 'progressive compounded multiplication' of separate effects of factors such as dose, dose rate, temperature, steel composition, and fabrication procedure. Therefore, the fast reactor data should not be extrapolated to determine credible void swelling behavior for PWR end-of-life (EOL) or life-extension conditions. Although the void swelling data extracted from fast reactor studies is extensive and conclusive, only limited amounts of swelling data and information have been obtained on microstructural characteristics from discharged PWR internals or steels irradiated at temperatures and at dose rates comparable to those of a PWR. Based on this relatively small amount of information, swelling in thin-walled tubes and baffle bolts in a PWR is not considered a concern. As additional data and relevant research becomes available, the newer results should be integrated with existing data, and the worthiness of this conclusion should continue to be scrutinized. PWR baffle reentrant corners are the most likely location to experience high swelling rates, and hence, high swelling at EOL, especially in internal regions of small volume where irradiation temperature is high. However, it is considered unlikely that void swelling in a reentrant corner will exceed the threshold level of {approx}4% beyond which the swelling rate reaches the steady state rate of 1%/dpa. However, this estimation is only preliminary, and a more accurate quantification of maximum temperature of reentrant corners at EOL and life-extension situations would be useful.

  13. 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.

  14. 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)

  15. 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 ...

  16. 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) ...

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

    E-Print Network [OSTI]

    Shirai, Kiyoaki

    (W|R) W P(S|R, W) S P(W|R) 1 P(W|R) = li P(wi|li) � ni P(ni|Ni) j D(ni|Ni[pj : vj]) � (pj1···pjm) P(pj1 · · · pjm|Pj1 · · · Pjm[vj]) (2) 1(2) (2) 1 wi( ) li wi R (2) 2 ni D(ni|Ni[pj : vj]) ni vj pj ni ni D(ni|N[pj : vj]) = P(ni|N[pj : vj]) P(ni|N) (3) ni vj pj ( ) 1 ( ) 1 D 1 [4, 6] ni (2) 3 pi vj m pj1, · · · , pjm

  18. 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.

  19. Cameras, PwrMeter,and MultiMeter rfi August 27, 1999 5 FIGURE 3: Power meter birdie with door closed.

    E-Print Network [OSTI]

    Cameras, PwrMeter,and MultiMeter rfi August 27, 1999 5 FIGURE 3: Power meter birdie with door closed. Fri Aug 27 23:17:58 1999 phil #12; Cameras, PwrMeter,and MultiMeter rfi August 27, 1999 4 FIGURE 2. Power meter, multimeter birdies with door closed. Fri Aug 27 23:14:06 1999 phil #12; Cameras, PwrMeter

  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. Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR plants

    SciTech Connect (OSTI)

    Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

    1989-01-01T23:59:59.000Z

    Recent data from the HFIR vessel surveillance program indicate a substantial radiation embrittlement rate effect at low irradiation temperatures (/approximately/120/degree/F) for A212-B, A350-LF3, A105-II, and corresponding welds. PWR vessel supports are fabricated of similar materials and are subjected to the same low temperatures and fast neutron fluxes (10/sup 8/ to 10/sup 9/ neutrons/cm/sup 2//center dot/s, E > 1.0 MeV) as those in the HFIR vessel. Thus, the embrittlement rate of these structures may be greater than previously anticipated. A study sponsored by the NRC is under way at ORNL to determine the impact of the rate effect on PWR vessel-support life expectancy. The scope includes the interpretation and application of the HFIR data, a survey of all light-water-reactor vessel support designs, and a structural and fracture-mechanics analysis of the supports for two specific PWR plants of particular interest with regard to a potential for support failure as a result of propagation of flaws. Calculations performed thus far indicate best-estimate critical flaw sizes, corresponding to 32 EFPY, of /approximately/0.2 in. for one plant and /approximately/0.4 in. for the other. These flaw sizes are small enough to be of concern. However, it appears that low-cycle fatigue is not a viable mechanism for creation of flaws of this size, and thus, presumably, such flaws would have to exist at the time of fabrication. 59 refs., 128 figs., 49 tabs.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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)

  7. 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.

  8. 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)

  9. 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.

  10. 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.

  11. San Onofre PWR Data for Code Validation of MOX Fuel Depletion Analyses

    SciTech Connect (OSTI)

    Hermann, O.W.

    1999-09-01T23:59:59.000Z

    The isotopic composition of mixed-oxide fuel (fabricated with both uranium and plutonium isotope) discharged from reactors is of interest to the Fissile Material Disposition Program. The validation of depletion codes used to predict isotopic compositions of MOX fuel, similar to studies concerning uranium-only fueled reactors, thus, is very important. The EEI-Westinghouse Plutonium Recycle Demonstration Program was conducted to examine the use of MOX fuel in the San Onofre PWR, Unit I, during cycles 2 and 3. The data usually required as input to depletion codes, either one-dimensional or lattice codes, were taken from various sources and compiled into this report. Where data were either lacking or determined inadequate, the appropriate data were supplied from other references. The scope of the reactor operations and design data, in addition to the isotopic analyses, were considered to be of sufficient quality for depletion code validation.

  12. 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.

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

    SciTech Connect (OSTI)

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

    1985-02-01T23:59:59.000Z

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

  14. Evaluation of steam-generator fluid mixing during layup. Final report. [PWR

    SciTech Connect (OSTI)

    MacArthur, A.D.

    1983-05-01T23:59:59.000Z

    The objective of this project was to develop practical methods of achieving an adequately mixed chemical environment on the secondary side of PWR steam generators during periods of shutdown, cold shutdown (layup), and startup. Layup chemicals introduced into the steam generator could then be evenly dispersed to minimize corrosion processes which may occur if the chemical environment was not properly maintained. Systems for chemical feed, mixing, sampling, and removal of contaminant chemicals in the steam generator secondary side were also evaluated and recommendations have been made. Test results from a plexiglass model indicated that forced circulation and turbulent mixing were the most effective methods of achieving a rapid, homogeneous chemical environment. Natural convection and diffusion, on the other hand, were found to be less effective in achieving a thorough mixing.

  15. 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.

  16. 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.

  17. 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)

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

    SciTech Connect (OSTI)

    Yang, Yong Sik; Bang, Je Geon; Kim, Sun Ki; Song, Kun Woo [LWR Fuel Development Division, Korea Atomic Energy Research Institute, 1045, Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of); Park, Su Ki [HANARO Utilization Technology Development Division, Korea Atomic Energy Research Institute, 1045, Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of); Seo, Chul Gyo [HANARO Management Division, Korea Atomic Energy Research Institute, 1045, Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

    2007-07-01T23:59:59.000Z

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

  19. Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR (pressurized-water-reactor) plants

    SciTech Connect (OSTI)

    Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

    1988-01-01T23:59:59.000Z

    Recent pressure-vessel surveillance data from the High Flux Isotope Reactor (HFIR) indicate an embrittlement fluence-rate effect that is applicable to the evaluation of the integrity of light-water reactor (LWR) pressure vessel supports. A preliminary evaluation using the HFIR data indicated increases in the nil ductility transition temperature at 32 effective full-power years (EFPY) of 100 to 130/degree/C for pressurized-water-reactor (PWR) vessel supports located in the cavity at midheight of the core. This result indicated a potential problem with regard to life expectancy. However, an accurate assessment required a detailed, specific-plant, fracture-mechanics analysis. After a survey and cursory evaluation of all LWR plants, two PWR plants that appeared to have a potential problem were selected. Results of the analyses indicate minimum critical flaw sizes small enough to be of concern before 32 EFPY. 24 refs., 16 figs., 7 tabs.

  20. Subchannel Thermal-Hydraulic Experimental Program (STEP). Volume 1. Mixing in a pressurized water reactor (PWR) rod bundle. Final report

    SciTech Connect (OSTI)

    Barber, A.R.; Zielke, L.A.

    1980-08-01T23:59:59.000Z

    This volume describes an experiment that was performed to determine the mixing characteristics of a pressurized water reactor (PWR) rod bundle. The objective of this project was to improve the subchannel computer code models of the reactor core. The experimental technique was isokinetic subchannel withdrawal of the entire flow from two sample subchannels. Once withdrawn, the sample fluid was condensed and its enthalpy was measured by regenerative heat exchange calorimetry. The test bundle was a 4 x 6 electrically heated array with a 50% power upset. The COBRA IIIC code was used to model the experiment and to determine the value of the thermal mixing coefficient, ..beta.., that was necessary to predict the measured results. Both single- and two-phase data were obtained over a range of PWR operating conditions. The results indicate that both single- and two-phase mixing is small. The COBRA model predicts the enthalpy data using a turbulent mixing coefficient, ..beta.. approx. = 0.002.

  1. 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.

  2. Impact of an apparent radiation embrittlement rate on the life expectancy of PWR (pressurized-water-reactor) vessel supports

    SciTech Connect (OSTI)

    Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

    1989-01-01T23:59:59.000Z

    Recent data from the HFIR vessel surveillance program indicate a substantial radiation embrittlement rate effect at low irradiation temperatures (/approximately/120/degree/F) for A212-B, A350-LF3, A105-II and corresponding welds. PWR vessel supports are fabricated of similar materials and are subjected to the same low temperatures and fast neutron fluxes (10/sup 8/ /minus/ 10/sup 9/ n/cm/sup 2//center dot/s, E > 1.0 MeV) as those in the HFIR vessel. Thus, the embrittlement rate of these structures may be greater than previously anticipated. A study sponsored by the NRC is under way at ORNL to determine the impact of the rate effect on PWR vessel support life expectancy. The scope includes the interpretation and application of the HFIR data, a survey of all LWR vessel support designs, and a structural and fracture-mechanics analysis of the supports for two specific PWR plants of particular interest with regard to a potential for support failure as a result of propagation of flaws. Calculations performed for one of the plants indicate best-estimate critical flaw size corresponding to 32 EFPY, of /approximately/0.4 in. It appears that low-cycle fatigue is not a viable mechanism for creation of flaws of this size. Thus, presumably such flaws would have to exist at the time of fabrication. 19 refs., 8 figs., 3 tabs.

  3. 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)

  4. 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...

  5. 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 on Google Bookmark EERE: Alternative Fuels Data Center Home5b9fcbce19 No revision hasInformation Earth's Heat JumpIncMAK Technologies JumpLuangwa Zambia5.1719°,

  6. 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating SolarElectric Coop,Save EnergyGlouster,Winside,WarrenWells Rural Electric

  7. 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.

  8. 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)

  9. 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.

  10. 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)

  11. Remote Gamma Scanning System for Characterization of BWR and PWR Fuel Rod Sections

    SciTech Connect (OSTI)

    Crowell, Shannon L.; Alzheimer, James M.

    2011-08-08T23:59:59.000Z

    Sometimes challenges with the design and deployment of automated equipment in remote environments deals more with the constraints imposed by the remote environment than it does with the details of the automation. This paper discusses the development of a scanning system used to provide gamma radiation profiles of irradiated fuel rod segments. The system needed the capability to provide axial scans of cut segments of BWR and PWR fuel rods. The scanning location is A-Cell at the Radiochemical Processing Laboratory (RPL) at the Hanford site in Washington State. The criteria for the scanning equipment included axial scanning increments of a tenth of an inch or less, ability to scan fuel rods with diameters ranging from 3/8 inch to 5/8 inch in diameter, and fuel rod segments up to seven feet in length. Constraints imposed by the environment included having the gamma detector and operator controls on the outside of the hot cell and the scanning hardware on the inside of the hot cell. This entailed getting a narrow, collimated beam of radiation from the fuel rod to the detector on the outside of the hot cell while minimizing the radiation exposure caused by openings for the wires and cables traversing the hot cell walls. Setup and operation of all of the in-cell hardware needed to accommodate limited access ports and use of hot cell manipulators. The radiation levels inside the cell also imposed constraints on the materials used.

  12. 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.

  13. 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)

  14. 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.

  15. Design, Feasibility, and Testing of Instrumented Rod Bundles to Improve Heat Transfer Knowledge in PWR Fuel Assemblies

    SciTech Connect (OSTI)

    Bergeron, A. [CEA, Saclay (France); Chataing, T.; Garnier, J. [CEA, Genoble (France); Decossin, E.; Peturaud, P. [EDF/R and D, Chatou (France); Yagnik, S.K. [Electric Power Research Institute - EPRI (United States)

    2007-07-01T23:59:59.000Z

    Two 5 x 5 test rod bundles mimicking the PWR fuel assembly have been adapted into two suitable test loop facilities, respectively, to carry out sufficiently detailed hydraulic and thermal measurements in identical geometric configuration. The objective is to investigate heat transfer phenomena in single-phase as well as with onset of nucleate boiling (ONB). The accuracy and reproducibility of the temperature measurements using the sliding-traversing thermocouple device under typical PWR conditions has been demonstrated in the thermal test facility. In the hydraulic loop, a Laser Doppler Velocimetry (LDV) system to precisely scan the local axial velocity component in each sub-channel has been implemented. The approach is to utilize mean sub-channel axial velocity distributions and pressure drop data from the hydraulic loop and the global boundary conditions (Pressure, Temperature, flow rate) from the thermal loop to simulate sub-channels in appropriate T/H codes. This permits computation of sub-channel averaged fluid temperatures (as well as mass velocity) in various subchannels within the test bundle. Subsequently, in conjunction with the wall temperatures and applied heat flux values from the thermal loop, it is possible to develop a complete map of heat transfer coefficients along the 9 instrumented central heater rods. Locations downstream of spacer grids would be of special interest. Depending on pressure, mass velocity and heat flux conditions of a given test, the inlet temperature will be a parameter to be varied so that the ONB boundary can be observed within the bundle. Detailed designs of the test section, required loop modifications, and adaptation of specialized instrumentation and data acquisition systems have been accomplished in both test loops. Further we have established that based on such detailed rod surface temperature and sub-channel axial velocity measurements, it is possible to achieve sufficient accuracy in the temperature measurements to meet the objective of improving the heat transfer correlations applicable to PWR cores. (authors)

  16. Investigation of the behaviour of high burn-up PWR fuel under RIA conditions in the CABRI test reactor

    SciTech Connect (OSTI)

    Schmitz, F.; Papin, J.; Haessler, M.; Nervi, J.C. [Institut de Protection et de Surete Nucleaire (France); Permezel, P. [Electricite de France, Septen (France)

    1994-10-01T23:59:59.000Z

    Performance, reliability and economics are the goal criteria for fuel pin design and development. For steady state behaviour and operational transients, the demonstration is made worldwide that burn-up of more than 60 GWd/t can be reached reliably with improved PWR fuel. It has however not been demonstrated yet that safety criteria, related to design basis accident scenarios, are still respected at these high burn-up levels. In particular, for the reactivity initiated accident (RIA), resulting from a postulated, rapid removal of control rod elements, the amount of energy injection must be limited by design such that no severe damage to the core and its structures might occur.

  17. 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.

  18. Analysis of a Defected Dissimilar Metal Weld in a PWR Power Plant

    SciTech Connect (OSTI)

    Efsing, P. [Barseback Kraft AB, P.O. Box 524, Loddekopinge SE-246 25 (Sweden); Lagerstrom, J. [Vattenfall AB, Ringhals, 430 22 Vaeroebacka (Sweden)

    2002-07-01T23:59:59.000Z

    During the refueling outage 2000, inspections of the RC-loops of one of the Ringhals PWR-units, Ringhals 4, indicated surface breaking defects in the axial direction of the piping in a dissimilar weld between the Low alloy steel nozzle and the stainless safe end in the hot leg. In addition some indications were found that there were embedded defects in the weld material. These defects were judged as being insignificant to the structural integrity. The welds were inspected in 1993 with the result that no significant indications were found. The weld it self is a double U weld, where the thickness of the material is ideally 79,5 mm. Its is constructed by Inconel 182 weld material. At the nozzle a buttering was applied, also by Inconel 182. The In-service inspection, ISI, of the object indicated four axial defects, 9-16 mm deep. During fabrication, the areas where the defects are found were repaired at least three times, onto a maximum depth of 32 mm. To evaluate the defects, 6 boat samples from the four axial defects were cut from the perimeter and shipped to the hot-cell laboratory for further examination. This examination revealed that the two deep defects had been under sized by the ISI outside the requirement set by the inspection tolerances, while the two shallow defects were over sized, but within the tolerances of the detection system. When studying the safety case it became evident that there were several missing elements in the way this problems is handled with respect to the Swedish safety evaluation code. Among these the most notable at the beginning was the absence of reliable fracture mechanical data such as crack growth laws and fracture toughness at elevated temperature. Both these questions were handled by the project. The fracture mechanical evaluation has focused on a fit for service principal. Thus defects both in the unaffected zones and the disturbed zones, boat sample cutouts, of the weld have been analyzed. With reference to the Swedish safety evaluation system in accordance to the regulatory demands, a safety evaluation was performed using the R6-method. The failure assessment diagram is modified by the addition of the ASME XI safety factors both for limit load analysis and fracture assessment. This results in a very high conservatism since the secondary stresses such as residual stresses are high in the area. In order to quantify this effect an analysis in accordance to ASME IWB-3640, App. C was performed. This analysis provides the decision-makers with a sensitivity study; important to have to value the real risk of any missed defects in the area. (authors)

  19. 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.

  20. 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)

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. Development of a Safeguards Verification Method and Instrument to Detect Pin Diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies

    SciTech Connect (OSTI)

    Ham, Y S; Maldonado, G I; Burdo, J; He, T

    2006-10-10T23:59:59.000Z

    A technical safeguards challenge has remained for decades for the IAEA to identify possible diversion of nuclear fuel pins from Light Water Reactor (LWR) spent fuel assemblies. In fact, as modern nuclear power plants are pushed to higher power levels and longer fuel cycles, fuel failures (i.e., ''leakers'') as well as the corresponding fuel assembly repairs (i.e., ''reconstitutions'') are commonplace occurrences within the industry. Fuel vendors have performed hundreds of reconstitutions in the past two decades, thus, an evolved know-how and sophisticated tools exist to disassemble irradiated fuel assemblies and replace damaged pins with dummy stainless steel or other type rods. Various attempts have been made in the past two decades to develop a technology to identify a possible diversion of pin(s) and to determine whether some pins are missing or replaced with dummy or fresh fuel pins. However, to date, there are no safeguards instruments that can detect a possible pin diversion scenario to the requirements of the IAEA. The FORK detector system [1-2] can characterize spent fuel assemblies using operator declared data, but it is not sensitive enough to detect missing pins from spent fuel assemblies. Likewise, an emission computed tomography system [3] has been used to try to detect missing pins from a spent fuel assembly, which has shown some potential for identifying possible missing pins but this capability has not yet been fully demonstrated. The use of such a device in the future would not be envisaged, especially in an inexpensive, easy to handle setting for field applications. In this article, we describe a concept and ongoing research to help develop a new safeguards instrument for the detection of pin diversions in a PWR spent fuel assembly. The proposed instrument is based on one or more very thin radiation detectors that could be inserted within the guide tubes of a Pressurized Water Reactor (PWR) assembly. Ultimately, this work could lead to the development of a detector cluster and corresponding high-precision driving system to collect radiation signatures inside PWR spent fuel assemblies. The data obtained would provide the spatial distribution of the neutron and gamma flux fields within the spent fuel assembly, while the data analysis would be used to help identify missing or replaced pins. Monte Carlo simulations have been performed to help validate this concept using a realistic 17 x 17 PWR spent fuel assembly [4-5]. The initial results of this study show that neutron profile in the guide tubes, when obtained in the presence of missing pins, can be identifiably different from the profiles obtained without missing pins, Our latest simulations have focused upon a specific type of fission chamber that could be tested for this application.

  6. PWR blowdown heat transfer separate-effects program - Thermal-Hydraulic Test Facility experimental data report for test 177. [Contains microfiche data

    SciTech Connect (OSTI)

    Clemons, V.D.; Flanders, R.M.; Craddick, W.G.

    1980-08-01T23:59:59.000Z

    Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 177, which is part of the ORNL Pressurized-Water Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. Objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant accident in a PWR system. Test 177 was conducted at the request of Idaho National Engineering Laboratory ''for use in the independent assessment of RELAP4/MOD6.'' Primary purpose of this report is to make the reduced instrument responses during test 177 available. The responses are presented in graphical form in engineering units and have been analyzed only to the extent necessary to assure reasonableness and consistency. The data are presented in microfiche form.

  7. Evaluation of storing Shippingport Core II spent blanket fuel assemblies in the T Plant PWR Core II fuel pool without active cooling

    SciTech Connect (OSTI)

    Gilbert, E.R.; Lanning, D.D. [Pacific Northwest Lab., Richland, WA (United States); Dana, C.M.; Hedengren, D.C. [Westinghouse Hanford Co., Richland, WA (United States)

    1994-10-01T23:59:59.000Z

    PWR Core II fuel pool chiller-off test was conducted because it appeared possible that acceptable pool-water temperatures could be maintained without operating the chillers, thus saving hundreds of thousands of dollars in maintenance and replacement costs. Test results showed that the water-cooling capability is no longer needed to maintain pool temperature below 38{degrees}C (100{degrees}F).

  8. 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).

  9. Benchmark of SCALE (SAS2H) isotopic predictions of depletion analyses for San Onofre PWR MOX fuel

    SciTech Connect (OSTI)

    Hermann, O.W.

    2000-02-01T23:59:59.000Z

    The isotopic composition of mixed-oxide (MOX) fuel, fabricated with both uranium and plutonium, after discharge from reactors is of significant interest to the Fissile Materials Disposition Program. The validation of the SCALE (SAS2H) depletion code for use in the prediction of isotopic compositions of MOX fuel, similar to previous validation studies on uranium-only fueled reactors, has corresponding significance. The EEI-Westinghouse Plutonium Recycle Demonstration Program examined the use of MOX fuel in the San Onofre PWR, Unit 1, during cycles 2 and 3. Isotopic analyses of the MOX spent fuel were conducted on 13 actinides and {sup 148}Nd by either mass or alpha spectrometry. Six fuel pellet samples were taken from four different fuel pins of an irradiated MOX assembly. The measured actinide inventories from those samples has been used to benchmark SAS2H for MOX fuel applications. The average percentage differences in the code results compared with the measurement were {minus}0.9% for {sup 235}U and 5.2% for {sup 239}Pu. The differences for most of the isotopes were significantly larger than in the cases for uranium-only fueled reactors. In general, comparisons of code results with alpha spectrometer data had extreme differences, although the differences in the calculations compared with mass spectrometer analyses were not extremely larger than that of uranium-only fueled reactors. This benchmark study should be useful in estimating uncertainties of inventory, criticality and dose calculations of MOX spent fuel.

  10. Linking Grain Boundary Microstructure to Stress Corrosion Cracking of Cold Rolled Alloy 690 in PWR Primary Water

    SciTech Connect (OSTI)

    Bruemmer, Stephen M.; Olszta, Matthew J.; Toloczko, Mychailo B.; Thomas, Larry E.

    2012-10-01T23:59:59.000Z

    Grain boundary microstructures and microchemistries are examined in cold-rolled alloy 690 tubing and plate materials and comparisons are made to intergranular stress corrosion cracking (IGSCC) behavior in PWR primary water. Chromium carbide precipitation is found to be a key aspect for materials in both the mill annealed and thermally treated conditions. Cold rolling to high levels of reduction was discovered to produce small IG voids and cracked carbides in alloys with a high density of grain boundary carbides. The degree of permanent grain boundary damage from cold rolling was found to depend directly on the initial IG carbide distribution. For the same degree of cold rolling, alloys with few IG precipitates exhibited much less permanent damage. Although this difference in grain boundary damage appears to correlate with measured SCC growth rates, crack tip examinations reveal that cracked carbides appeared to blunt propagation of IGSCC cracks in many cases. Preliminary results suggest that the localized grain boundary strains and stresses produced during cold rolling promote IGSCC susceptibility and not the cracked carbides and voids.

  11. 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)

  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. 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.

  14. 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.

  15. 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)

  16. An assessment of the core degradation frequency in a typical large LMFBR design for internal accident initiators-a comparison with PWR predictions

    SciTech Connect (OSTI)

    Tzanos, C.P.; Adamantiades, A.G.; Hanan, N.A.

    1983-12-01T23:59:59.000Z

    A comparative assessment of the core degradation frequency due to internal accident initiators between a typical large liquid-metal fast breeder reactor (LMFBR) design and pressurized water reactors (PWRs) has been performed. For the PWR system, existing analyses have been utilized. For the reference LMFBR, an extensive analysis has been performed for two accident initiators, i.e., loss of off-site power and loss of main feedwater. Based on this analysis an estimate of about1 X 10/sup -6//reactor X yr has been obtained for the core degradation frequency of the reference LMFBR. This estimate is significantly smaller than the PWR core degradation frequency ( about 6 X 10/sup -5//yr). A sensitivity analysis shows that the parameters having the largest impact on the unavailability of decay heat removal are (a) for the ''loss of off-site power'' initiator: human error and failure to restore off-site power, and (b) for the ''loss of main feedwater'' initiator: the leakage rates of the passive decay heat removal system and the adoption of the policy to repair the Na-NaK heat exchanger only during normal shutdowns. The results indicate that the LMFBR system has the potential of higher resistance than the PWR system to the accident initiators considered. The lower core degradation frequency estimated for the LMFBR system is due to the presence of two redundant and diverse reactor shutdown systems, with a self-actuated feature included in one of them, the incorporation of a passive decay heat removal system, and the significantly lower sensitivity of the reference LMFBR to primary system pipe breaks.

  17. PWR FLECHT SEASET 163-Rod Bundle Flow Blockage Task data report. NRC/EPRI/Westinghouse report No. 13, August-October 1982

    SciTech Connect (OSTI)

    Loftus, M J; Hochreiter, L E; McGuire, M F; Valkovic, M M

    1983-10-01T23:59:59.000Z

    This report presents data from the 163-Rod Bundle Blow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Systems Effects and Separate Effects Test Program (FLECHT SEASET). The task 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. These tests were designed to determine effects of flow blockage and flow bypass on reflooding behavior and to aid in the assessment of computational models in predicting the reflooding behavior of flow blockage in rod bundle arrays.

  18. 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.

  19. Fuel cycle cost, reactor physics and fuel manufacturing considerations for Erbia-bearing PWR fuel with > 5 wt% U-235 content

    SciTech Connect (OSTI)

    Franceschini, F.; Lahoda, E. J.; Kucukboyaci, V. N. [Westinghouse Electric Co. LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01T23:59:59.000Z

    The efforts to reduce fuel cycle cost have driven LWR fuel close to the licensed limit in fuel fissile content, 5.0 wt% U-235 enrichment, and the acceptable duty on current Zr-based cladding. An increase in the fuel enrichment beyond the 5 wt% limit, while certainly possible, entails costly investment in infrastructure and licensing. As a possible way to offset some of these costs, the addition of small amounts of Erbia to the UO{sub 2} powder with >5 wt% U-235 has been proposed, so that its initial reactivity is reduced to that of licensed fuel and most modifications to the existing facilities and equipment could be avoided. This paper discusses the potentialities of such a fuel on the US market from a vendor's perspective. An analysis of the in-core behavior and fuel cycle performance of a typical 4-loop PWR with 18 and 24-month operating cycles has been conducted, with the aim of quantifying the potential economic advantage and other operational benefits of this concept. Subsequently, the implications on fuel manufacturing and storage are discussed. While this concept has certainly good potential, a compelling case for its short-term introduction as PWR fuel for the US market could not be determined. (authors)

  20. 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

  1. 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.

  2. Sandia National Laboratories: PWR

    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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 -theErik Spoerke SSLSMolten-SaltReliability PVracks HelioVoltPV-TechPV_LIB

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

    SciTech Connect (OSTI)

    Ade, Brian J [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.

  4. 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.

  5. Development of a Safeguards Verification Method and Instrument to Detect Pin Diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies Phase I Study

    SciTech Connect (OSTI)

    Ham, Y S; Sitaraman, S

    2008-12-24T23:59:59.000Z

    A novel methodology to detect diversion of spent fuel from Pressurized Water Reactors (PWR) has been developed in order to address a long unsolved safeguards verification problem for international safeguards community such as International Atomic Energy Agency (IAEA) or European Atomic Energy Community (EURATOM). The concept involves inserting tiny neutron and gamma detectors into the guide tubes of a spent fuel assembly and measuring the signals. The guide tubes form a quadrant symmetric pattern in the various PWR fuel product lines and the neutron and gamma signals from these various locations are processed to obtain a unique signature for an undisturbed fuel assembly. Signatures based on the neutron and gamma signals individually or in a combination can be developed. Removal of fuel pins from the assembly will cause the signatures to be visibly perturbed thus enabling the detection of diversion. All of the required signal processing to obtain signatures can be performed on standard laptop computers. Monte Carlo simulation studies and a set of controlled experiments with actual commercial PWR spent fuel assemblies were performed and validated this novel methodology. Based on the simulation studies and benchmarking measurements, the methodology developed promises to be a powerful and practical way to detect partial defects that constitute 10% or more of the total active fuel pins. This far exceeds the detection threshold of 50% missing pins from a spent fuel assembly, a threshold defined by the IAEA Safeguards Criteria. The methodology does not rely on any operator provided data like burnup or cooling time and does not require movement of the fuel assembly from the storage rack in the spent fuel pool. A concept was developed to build a practical field device, Partial Defect Detector (PDET), which will be completely portable and will use standard radiation measuring devices already in use at the IAEA. The use of the device will not require any information provided by the operator or any prior knowledge of the spent fuel assembly. The device can also be operated without any movement of the spent fuel from its storage position. Based on parametric studies conducted via computer simulation, the device should be able to detect diversion of as low as ten percent of the missing or replaced fuel from an assembly regardless of the location of the missing fuel within the assembly, of the cooling time, initial fuel enrichment or burnup levels. Conditions in the spent fuel pool such as clarity of the water or boron content are also not issues for this device. The shape of the base signature is principally dependent on the layout of the guide tubes in the various types of PWR fuel assemblies and perturbations in the form of replaced fuel pins will distort this signature. These features of PDET are all unique and overcome limitation and disadvantages presented by currently used devices such as the Fork detector or the Cerenkov Viewing Device. Thus, this device when developed and tested could fill an important need in the safeguards area for partial defect detection, a technology that the IAEA has been seeking for the past few decades.

  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. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartment ofDepartment ofofOxford SiteToledo SiteTonawanda North Site Unit3.1 03/13[ {

  8. 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 on Google Bookmark EERE: Alternative Fuels Data CenterFranconia, Virginia: Energy ResourcesLoading map...(Utility Company) Jump to:City) Jump to:Newmarket,3034505°NextGen

  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 on Google Bookmark EERE: Alternative Fuels Data CenterFranconia, Virginia: Energy ResourcesLoading map...(Utility Company) Jump to:City) Jump

  10. 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 on Google Bookmark EERE: Alternative Fuels Data Center Home5b9fcbce19 No revision hasInformation Earth's Heat JumpIncMAK Technologies JumpLuangwaoldid=676597"NenanaNewVicksberg,

  11. 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 on Google Bookmark EERE: Alternative Fuels Data Center Home5b9fcbce19 No revision hasInformation Earth's Heat JumpIncMAK Technologies

  12. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartment ofDepartment ofof EnergyYou$0.C. 20545*.MSE Cores TubaySite, Ohio, Site

  13. 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus AreaDataBusPFAN) | OpenInc JumpNew YorkNewNextEra

  14. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartment ofDepartment ofof EnergyYou are herePA 3003A AECMexicoEngland

  15. Niagara Falls, NY Natural Gas Exports to Canada

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelinesProved ReservesFeet) Year Jan FebFeet)SalesYear Jan Feb Mar0 0 0 0 03a

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

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelinesProved ReservesFeet) Year Jan FebFeet)SalesYear Jan Feb Mar0 0 0 0 03a188,525

  17. 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.

  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. The French codes RCC-M and RSE-M -- Design, construction and in-service inspection rules for the mechanical components of PWR nuclear islands: An overview and a comparison to the ASME codes

    SciTech Connect (OSTI)

    Journet, J.; Masson, S.H.; Morel, A.; Remond, A.; Grandemange, J.M.

    1995-12-01T23:59:59.000Z

    The RCC-M, ``Regles de Conception et de Construction des Materiels Mecaniques des Ilots Nucleaires REP`` or, in English, ``Design and Construction Rules for the Mechanical Components of PWR Nuclear Islands`` and the RSE-M, ``Regles de Surveillance en Exploitation des Materiels Mecaniques des Ilots Nucleaires REP`` or, in English, ``In-Service Inspection Rules for the Mechanical Components of PWR Nuclear Islands`` gather all design, construction and operating practices relating to the mechanical components of French PWR nuclear islands. This paper is a presentation of these two codes. Throughout this presentation the specific aspects of the French approach will be underlined and will be compared to that of the ASME codes--mainly Section 3 and Section 11. The broad general technical scopes of the French codes are similar to those of the ASME codes. However, in some important areas of design, material specifications, procurement and manufacturing, the provisions of the RCC-M and RSE-M deviate from those of a strict mechanical Code and are more self-sustaining than those of ASME.

  20. 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.

  1. 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.

  2. Degraded voltage resulting in non-safety UPS failure at Nine Mile Point Unit 2 August 13, 1991

    SciTech Connect (OSTI)

    Julka, A.K. (Niagara Mohawk Power Corp. (United States))

    1992-08-01T23:59:59.000Z

    At approximately 5:48 a.m. on august 13, 1991, phase B of the main stepup transformer of Niagara Mohawk's Nine Mile Point Unit 2 (NMP2) Nuclear Power Plant experienced a failure resulting in degraded voltage in phase B of the electrical AC distribution system. The duration of the degraded voltage lasted 12 cycles, the time required to clear the fault and to fast transfer the house loads to alternate offsite sources. The protective relaying schemes accomplished this without any abnormalities. This paper reports that due to the nature of the fault and its protection, the turbine tripped, resulting in an automatic reactor scram. However, during the fast transfer under degraded voltage conditions, five non-0safety related Uninterruptible Power Supplies (UPS) tripped; these UPS's were supplied by Exide Electronics. The tripping of these UPS's resulted in the loss of plant process computers, Control Room annunciation, and a significant portion of non-safety related instrumentation and control circuits.

  3. 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.

  4. TRAC methods and models. [PWR

    SciTech Connect (OSTI)

    Mahaffy, J.H.; Liles, D.R.; Bott, T.F.

    1981-01-01T23:59:59.000Z

    The numerical methods and physical models used in the Transient Reactor Analysis Code (TRAC) versions PD2 and PF1 are discussed. Particular emphasis is placed on TRAC-PF1, the version specifically designed to analyze small-break loss-of-coolant accidents (LOCAs).

  5. PWR FLECHT SEASET 21-rod-bundle flow-blockage task: data and analysis report. NRC/EPRI/Westinghouse report No. 11, main report and appendices A-J

    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. Review of consolidated Edison`s integrated resource bidding program

    SciTech Connect (OSTI)

    Goldman, C.A.; Busch, J.F.; Kahn, E.P.; Baldick, R.; Milne, A.

    1993-07-01T23:59:59.000Z

    Competitive bidding has emerged as the dominant method for procuring new resources by US utilities. In New York, the Public Service Commission (NYPSC) ordered the state`s seven investor-owned utilities to develop bidding programs to acquire supply and DSM resource options. Utilities were allowed significant discretion in program design in order to encourage experimentation. Competitive bidding programs pose formidable policy, design, and management challenges for utilities and their regulators. Yet, there have been few detailed case studies of bidding programs, particularly of those utilities that take on the additional challenge of having supply and DSM resources compete head-to-head for a designated block of capacity. To address that need, the New York State Energy Research and 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 bidding programs of two utilities that tested the integrated ``all-sources`` approach. This study focuses primarily on Consolidated Edison Company of New York`s (Con Edison) bidding program; an earlier report discusses our review of Niagara Mohawk`s program (Goldman et al 1992). We reviewed relevant Commission decisions, utility filings and signed contracts, interviewed utility and regulatory staff, surveyed DSM bidders and a selected sample of DSM non-bidders, and analyzed the bid evaluation system used in ranking bids based on detailed scoring information on individual bids provided by Con Edison.

  7. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartment ofDepartment ofofOxford SiteToledo SiteTonawanda North SiteD&D D&DKEY

  8. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec 2007 10,998 9,933 10,998 10,643 10,998through 1996) inDecade Year-0 Year-1 Year-2Thousand NetThousand Cubic

  9. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec 2007 10,998 9,933 10,998 10,643 10,998through 1996) inDecade Year-0 Year-1 Year-2Thousand NetThousand

  10. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec 2007 10,998 9,933 10,998 10,643 10,998through 1996) inDecade Year-0 Year-1 Year-2Thousand

  11. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec 2007 10,998 9,933 10,998 10,643 10,998through 1996) inDecade Year-0 Year-1 Year-2ThousandFeet) Year Jan

  12. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec 2007 10,998 9,933 10,998 10,643 10,998through 1996) inDecade Year-0 Year-1 Year-2ThousandFeet) Year

  13. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec 2007 10,998 9,933 10,998 10,643 10,998through 1996) inDecade Year-0 Year-1 Year-2ThousandFeet)

  14. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec 2007 10,998 9,933 10,998 10,643 10,998through 1996) inDecade Year-0 Year-1 Year-2ThousandFeet)Feet)

  15. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec 2007 10,998 9,933 10,998 10,643 10,998through 1996) inDecade Year-0 Year-1

  16. 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating SolarElectricEnergy

  17. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartment ofDepartment ofof EnergyYou areDowntown SiteOhioMissouriMaywood

  18. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartment ofDepartment ofof EnergyYou$ EGcG ENERGYELIkNATION REPORTFairfield,? . -

  19. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power SystemsResources DOEElectricalonJustice EnvironmentalDISTRIBUTIO192-01 EvaluationStorage

  20. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power SystemsResources DOEElectricalonJustice EnvironmentalDISTRIBUTIO192-01

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

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelinesProved ReservesFeet) Year Jan FebFeet)SalesYear Jan Feb Mar0 0 0 0Thousand

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

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelinesProved ReservesFeet) Year Jan FebFeet)SalesYear Jan Feb Mar0 0 0Feet) Year Jan

  3. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartment ofDepartment ofof EnergyYou$0.C. 20545 OCT 28 1% - :NEW;ORAU/ I_ROHM ANDTEXASe -

  4. PWR cores with silicon carbide cladding

    SciTech Connect (OSTI)

    Dobisesky, J. P.; Carpenter, D.; Pilat, E.; Kazimi, M. S. [Center for Advanced Nuclear Energy Systems, Dept. of Nuclear Science and Engineering, Massachusetts Inst. of Technology, 77 Massachusetts Avenue 24-215, Cambridge, MA 02139-4307 (United States)

    2012-07-01T23:59:59.000Z

    The feasibility of using silicon carbide rather than Zircaloy cladding, to reach higher power levels and higher discharge burnups in PWRs has been evaluated. A preliminary fuel design using fuel rods with the same dimensions as in the Westinghouse Robust Fuel Assembly but with fuel pellets having 10 vol% central void has been adopted to mitigate the higher fuel temperatures that occur due to the lower thermal conductivity of the silicon carbide and to the persistence of the open clad-pellet gap over most of the fuel life. With this modified fuel design, it is possible to achieve 18 month cycles that meet present-day operating constraints on peaking factor, boron concentration, reactivity coefficients and shutdown margin, while allowing batch average discharge burnups up to 80 MWD/kgU and peak rod burnups up to 100 MWD/kgU. Power uprates of 10% and possibly 20% also appear feasible. For non-uprated cores, the silicon carbide-clad fuel has a clear advantage that increases with increasing discharge burnup. Even for comparable discharge burnups, there is a savings in enriched uranium. Control rod configuration modifications may be required to meet the shutdown margin criterion for the 20% up-rate. Silicon carbide's ability to sustain higher burnups than Zircaloy also allows the design of a licensable two year cycle with only 96 fresh assemblies, avoiding the enriched uranium penalty incurred with use of larger batch sizes due to their excessive leakage. (authors)

  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 on Google Bookmark EERE: Alternative Fuels Data CenterFranconia, Virginia: Energy Resources Jump to: navigation,Ohio:Greer County is8584°,Hardy County, West Virginia:HarmonJumpHarquahala

  6. PWR Fuel Shipping Limits & RNP Core Design

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

    Excellent Operational Record No radioactive spills or releases No radiation exposure to train personnel or general public 8 Safety - Success Factors Robust package design Nuclear...

  7. 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.

  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. E-Print Network 3.0 - african buffalo suggested Sample Search...

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

    of St Andrews Collection: Environmental Sciences and Ecology 91 BUFFALO-NIAGARA INTERNATIONAL AIRPORT Summary: BUFFALO-NIAGARA INTERNATIONAL AIRPORT MICHIGANST. GOODELL...

  11. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartment ofDepartment ofofOxford SiteToledo SiteTonawanda North SiteD&Dir^0 0 0 0

  12. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartment ofDepartment ofofOxford SiteToledo SiteTonawandaUniversity21Prepared by,..~ i;9s'

  13. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartment ofDepartment ofofOxford SiteToledo SiteTonawandaUniversity21Prepared by,..~ i;9s'7At

  14. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartment ofDepartment ofof EnergyYou$0.C.Green River,The Secretary ofd ' . ,$! liSite~

  15. Current Positions: Director Oil, Gas, & Mining, Vecron Lordstock Group (VLG)

    E-Print Network [OSTI]

    Boolchand, Punit

    : Chief Scientist and Director Carborundum Technology Center, The Carborundum Corporation (Niagara Falls The Carborundum Company #12;

  16. 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.

  17. 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.

  18. 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...

  19. 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...

  20. 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...

  1. 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...

  2. 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...

  3. 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.

  4. 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.

  5. Stress-corrosion cracking in BWR and PWR piping

    SciTech Connect (OSTI)

    Weeks, R.W.

    1983-07-01T23:59:59.000Z

    Intergranular stress-corrosion cracking of weld-sensitized wrought stainless steel piping has been an increasingly ubiquitous and expensive problem in boiling-water reactors over the last decade. In recent months, numerous cracks have been found, even in large-diameter lines. A number of potential remedies have been developed. These are directed at providing more resistant materials, reducing weld-induced stresses, or improving the water chemistry. The potential remedies are discussed, along with the capabilities of ultrasonic testing to find and size the cracks and related safety issues. The problem has been much less severe to date in pressurized-water reactors, reflecting the use of different materials and much lower coolant oxygen levels.

  6. 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.

  7. Annular burnout data from rod-bundle experiments. [PWR

    SciTech Connect (OSTI)

    Yoder, G.L.; Morris, D.G.; Mullins, C.B.

    1983-01-01T23:59:59.000Z

    Burnout data for annular flow in a rod bundle are presented for both transient and steady-state conditions. Tests were performed at the Oak Ridge National Laboratory in the Thermal Hydraulic Test Facility (THTF), a pressurized-water loop containing an electrically heated 64-rod bundle. The bundle configuration is typical of later generation pressurized-water reactors with 17 x 17 fuel arrays. Both axial and radial power profiles are flat. All experiments were carried out in upflow with subcooled inlet conditions, insuring accurate flow measurement. Conditions within the bundle were typical of those which could be encountered during a nuclear reactor loss-of-coolant accident.

  8. 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 ...

  9. 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...

  10. 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 ...

  11. 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.

  12. Steam generator sludge pile model boiler testing: sludge characterization. [PWR

    SciTech Connect (OSTI)

    Becker, L.F. Jr.; Esposito, J.N.

    1981-09-01T23:59:59.000Z

    As part of a program to understand the thermal and hydraulic transport process that can lead to chemical concentration in sludge piles on the tubesheet in a steam generator, the chemical composition and physical properties of eight sludges and several simulants were determined. Analyses performed by emission and x-ray fluorescence spectroscopy indicated that most of the sludges were mainly composed of iron oxides, copper, and other elements at trace levels. X-ray diffraction measurements identified iron to exist in the form of magnetite and copper to exist in the form of a metal. The densities, porosity, particle size, surface area, pore size distribution, and hydrodynamic permeabilities were determined on all plant sludges and selected simulants. Wide variations were observed in the physical measurements of the different plant sludges.

  13. 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.

  14. Puerto Rico Electric Pwr Authority | 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar PowerstoriesNrelPartnerType Jump to:Co Jump to: navigation, search Name:Rico

  15. Red River Valley Coop 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar PowerstoriesNrelPartnerType Jump to:Co JumpRETScreenJam Home NameRed LakeAssn

  16. Renville-Sibley Coop 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar PowerstoriesNrelPartnerType Jump to:Co

  17. Seward County Rrl 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar PowerstoriesNrelPartnerTypePonsa, Mallorca:up DataBus as a systemSevin

  18. South Central 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar PowerstoriesNrelPartnerTypePonsa,HomeIndiana: Energy ResourcesSouth

  19. South Mississippi 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar PowerstoriesNrelPartnerTypePonsa,HomeIndiana:Rhode Island References:Korea:El

  20. Stanton 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating SolarElectric Coop, Inc Place: Missouri References:InformationStanford-Dist Jump

  1. 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 on Google Bookmark EERE: Alternative Fuels Data Center Home5b9fcbce19 NoPublic Utilities Address: 160Benin:EnergyWisconsin:2003)Crowley County, Colorado:CumberlandCumbria WindCuming

  2. 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.

  3. Keosauqua Municipal Light & Pwr | 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories on climateJuno Beach, Florida:Kenyon Municipal Utilities Jump

  4. South Mississippi 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 on Google Bookmark EERE: Alternative Fuels Data Center Home5b9fcbce19 No revisionEnvReviewNonInvasiveExplorationUT-g GrantAtlas (PACA Region -Sonelgaz JumpSouthDakota‎ |

  5. 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentratingRenewable Solutions LLC JumpCrow Lake Wind Jump to:Roadmap Meeting Home >Rose

  6. Singing River Elec Pwr Assn (Mississippi) | 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 on Google Bookmark EERE: Alternative Fuels Data CenterFranconia, Virginia:FAQ < RAPID Jump to:Seadov Pty Ltd Jump to: navigation,Pvt LtdShrubSimpsonville, SouthSingaporeSinging River

  7. Michigan South Central Pwr Agy | 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 on Google Bookmark EERE: Alternative Fuels Data CenterFranconia, Virginia: Energy Resources Jump to:46 -Energieprojekte GmbH Jump to: navigation,MetalysisMiMichigan Public Power AgencyAgy

  8. 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating SolarElectricEnergyCT Biomass FacilityArdica Technologies JumpArizona

  9. Property:EnvReviewPwrPlantSiting | 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 on Google Bookmark EERE: Alternative Fuels Data Center Home5b9fcbce19 No revisionEnvReviewNonInvasiveExploration Jump to: navigation, search Property Name

  10. Rushmore 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 on Google Bookmark EERE: Alternative Fuels Data CenterFranconia, Virginia:FAQ < RAPID Jump to: navigation, searchVirginiaRoosevelt GardensUK-based supplier and

  11. Sam Rayburn Municipal 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 on Google Bookmark EERE: Alternative Fuels Data CenterFranconia, Virginia:FAQ < RAPID Jump to: navigation, searchVirginiaRooseveltVI Solaris a city in Utah607793°,SaltonSaltonElecSam

  12. Vermont Public Pwr Supply Auth | 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating SolarElectric Coop,Save Energy NowNew HampshireValeroTrans Co Inc Jump to:

  13. Vermont Yankee Nucl Pwr Corp | 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating SolarElectric Coop,Save Energy NowNew HampshireValeroTrans Co Inc Jump to:LLC

  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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating SolarElectric Coop,Save EnergyGlouster,Winside,WarrenWellsLoadingREMC Jump to:Elec

  15. Wolverine Pwr Supply 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating SolarElectric Coop,SaveWhiskey Flats Geothermal Areaarticle isWisconsinWiseSupply

  16. Crawfordsville Elec, Lgt & Pwr | 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Power Basics (The following text isRica NREL CooperationCraighead

  17. East Mississippi Elec 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Power Basics (The followingDirectLow CarbonOpen1 June, 2013 -Mississippi Elec

  18. East River Elec 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Power Basics (The followingDirectLow CarbonOpen1 June, 2013 -Mississippi

  19. 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of Inspector GeneralDepartmentAUDIT REPORTOpenWendeGuoCatalyst Renewables JumpViewCentralCentral LincolnCentralE

  20. 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.

  1. Interphase transport in horizontal stratified cocurrent flow. [PWR; BWR

    SciTech Connect (OSTI)

    Jensen, R.J.; Yuen, M.C.

    1982-05-01T23:59:59.000Z

    The problem of interfacial transport in cocurrent, horizontal stratified gas-liquid systems is considered. Local condensation heat transfer coefficients and interface shear stress were obtained from mass and force balances. These balances were based on gas phase pitot traverses at various streamwise locations. Laser anemometer measurements of liquid mean and rms fluctuation velocities were made at similar locations. The laser anemometer data supported the value for the interface shear velocity obtained by the gas phase force balance. Based on cocurrent stratified air-water flow data the noncondensing interface shear stress was found to be a function of the relative velocity between the phases and the liquid fraction. Incorporated into Linehan's relation for condensing flow shear stress, the correlation was found to estimate the shear velocity for the condensation data considered. Local condensation heat transfer coefficients and gas absorption mass transfer coefficients were found to be directly proportional to the shear velocity.

  2. 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.

  3. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartment ofDepartment ofofOxford SiteToledo SiteTonawandaUniversity21 theB 4q-utSPJ\Nf7 n-q

  4. 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

  5. E-Print Network 3.0 - ancient mtdna control Sample Search Results

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

    haplotype, ancient DNA, migra- tion, admixture, Arikara, Mohawk, Sioux, Chippewa, Cherokee, Ojibwa, ... Source: Kemp, Brian M. - Department of Anthropology & School of...

  6. 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.

  7. 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.

  8. 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 ...

  9. Advances in steam turbine technology for the power generation industry. PWR-Volume 26

    SciTech Connect (OSTI)

    Moore, W.G. [ed.

    1994-12-31T23:59:59.000Z

    This is a collection of the papers on advances in steam turbine technology for the power generation industry presented at the 1994 International Joint Power Generation Conference. The topics include advances in steam turbine design, application of computational fluid dynamics to turbine aerodynamic design, life extension of fossil and nuclear powered steam turbine generators, solid particle erosion control technologies, and artificial intelligence, monitoring and diagnostics.

  10. 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...

  11. 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...

  12. 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.

  13. 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.

  14. 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.

  15. 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 ...

  16. 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)

  17. 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...

  18. 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.

  19. 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.

  20. Evaluation of flaw characteristics and their influence on inservice inspection reliability. [PWR; BWR

    SciTech Connect (OSTI)

    Becker, F.L.

    1980-01-01T23:59:59.000Z

    This report describes the results of the first year's effort of a five year program which is being conducted by Battelle, Pacific Northwest Laboratories, on behalf of the US Nuclear Regulatory Commission. This initial effort was directed toward identification and quantification of inspection uncertainties, which are likely to occur during inservice inspection of LWR primary piping systems, and their influence on inspection reliability. These experiments were conducted on 304 stainless steel samples, however, the results are equally applicable to other materials. Later portions of the program will extend these measurements and evaluations to other materials and conditions.

  1. Physical protection solutions for security problems at nuclear power plants. [PWR; BWR

    SciTech Connect (OSTI)

    Darby, J.L.; Jacobs, J.

    1980-09-01T23:59:59.000Z

    Under Department of Energy sponsorship, Sandia National Laboratories has developed a broad technological base of components and integrated systems to address security concerns at facilities of importance, including nuclear reactors. The primary security concern at a light water reactor is radiological sabotage, a deliberate set of actions at a plant which could expose the public to a significant amount of radiation (on the order of 10 CFR 100 limits). (Also of importance to plant operators are acts of industrial sabotage that could prevent a plant from producing electrical power).

  2. Gamma-thermometer-based reactor-core liquid-level detector. [PWR

    SciTech Connect (OSTI)

    Burns, T.J.

    1981-06-16T23:59:59.000Z

    A system is provided which employs a modified gamma thermometer for determining the liquid coolant level within a nuclear reactor core. The gamma thermometer which normally is employed to monitor local core heat generation rate (reactor power), is modified by thermocouple junctions and leads to obtain an unambiguous indication of the presence or absence of coolant liquid at the gamma thermometer location. A signal processor generates a signal based on the thermometer surface heat transfer coefficient by comparing the signals from the thermocouples at the thermometer location. The generated signal is a direct indication of loss of coolant due to the change in surface heat transfer when coolant liquid drops below the thermometer location. The loss of coolant indication is independent of reactor power at the thermometer location. Further, the same thermometer may still be used for the normal power monitoring function.

  3. 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)

  4. Aggressive fuel designs minimize fuel costs for the ANO-1 PWR

    SciTech Connect (OSTI)

    Ober, T.G.; Megehee, K.B.; Bencheikh, A.; Thompson, R.A. (Entergy Operations, Jackson, MS (United States))

    1993-01-01T23:59:59.000Z

    Fuel cycle design objectives are influenced by the desire of utilities to attain top performer status in the industry and to become more cost competitive. At Energy, we are seeking aggressive fuel designs and core management schemes that reduce costs without compromising operating margins. Recent efforts at the Arkansas Nuclear One (ANO-1) plant demonstrated the effectiveness of this approach and led to important benefits for both the utility and the fuel vendor, Babcock Wilcox. With our acquisition of the CASMO-3/SIMULATE-3 advanced physics code, we initiated a proactive approach to the design of cycle 12 of ANO-1. The primary goal was to explore the use of advanced designs to reduce front-end fuel cycle costs for cycle 12. A secondary goal was to incorporate those features into cycle 12 that could lead to further cost or margin improvements in later cycles.

  5. Analysis of the FLECHT SEASET unblocked bundle steam-cooling and boiloff tests. [PWR

    SciTech Connect (OSTI)

    Wong, S.; Hochreiter, L.E.

    1981-05-01T23:59:59.000Z

    A series of forced convection steam cooling tests at low Reynolds numbers and bundle boiloff tests were conducted in the unblocked bundle task of the FLECHT SEASET program. The COBRA-IV-I computer code was utilized to simulate the steam cooling tests, so that the effects of the housing, disconnected heater rods in the bundle, and subchannel mixing were accurately accounted for. After careful data screening, a steady-state forced convection steam cooling heat transfer correlation was developed using the measured heater rod power, heater rod surface temperatures calculated from the measured cladding inner surface temperature by an inverse conduction code, and the vapor temperatures at various subchannels calculated by the COBRA-IV-I code The new correlation was found to give higher heat transfer than the conventional Dittus-Boelter correlation in the low Reynolds number region. At higher Reynolds numbers, the data begin to merge with the Dittus-Boelter correlation.

  6. Evaluation of nonequilibrium effects in bundle dispersed-flow film boiling. [PWR; BWR

    SciTech Connect (OSTI)

    Morris, D.G.; Mullins, C.B.; Yoder, G.L.

    1983-01-01T23:59:59.000Z

    The effects of thermodynamic nonequilibrium in dispersed flow film boiling heat transfer are examined. Steady-state and transient rod-bundle data are used to evaluate several empirical heat-transfer models commonly employed to predict post-CHF behavior. The models that account for thermodynamic nonequilibrium perform adequately, while those that ignore nonequilibrium effects incur errors in wall superheat as high as 190/sup 0/K. Nonequilibrium effects can also be treated by explicitly modeling the phenomena. The thermal-hydraulic code COBRA-TF employs this approach. Using bundle data, the models in the code are evaluated. Analysis suggests that the interfacial heat transfer is overpredicted.

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

    E-Print Network [OSTI]

    Chambers, Alex

    2012-10-19T23:59:59.000Z

    production/destruction, and radiotoxicity reduction as compared to a UOX and MOX assembly. It is found that the most beneficial recycling strategy is the one where all of the transuranics are recycled. The inclusion of Cm reduces the required U-235...

  8. Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations.

    SciTech Connect (OSTI)

    Blomquist, R.N.; Finck, P.J.; Jammes, C.; Stenberg, C.G.

    1999-02-17T23:59:59.000Z

    Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k{sub eff} of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data.

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

    Office of Scientific and Technical Information (OSTI)

    and results of a major loss-of-coolant accident in RVR (the Shippingport Atomic Power Station). There it is shown that a zirconium-water (or steam) reaction is a possible...

  10. 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 ...

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. Full-scale turbine-missile-casing tests. Final report. [PWR; BWR

    SciTech Connect (OSTI)

    Yoshimura, H.R.; Schamaun, J.T.

    1983-01-01T23:59:59.000Z

    Results are presented of two full-scale tests simulating the impact of turbine disk fragments on simple ring and shell structures that represent the internal stator blade ring and the outer housing of an 1800-rpm steam turbine casing. The objective was to provide benchmark data on both the energy-absorbing mechanisms of the impact process and, if breakthrough occured, the exit conditions of the turbine missile. A rocket sled was used to accelerate a 1527-kg (3366-lb) segment of a turbine disk, which impacted a steel ring 12.7 cm (5 in.) thick and a steel shell 3.2 cm (1.25 in.) thick. The impact velocity of about 150 m/s (492 ft/s) gave a missile kinetic energy corresponding to the energy of a fragment from a postulated failure at the design overspeed (120% of operating speed). Depending on the orientation of the missile at impact, the steel test structure either slowed the missile to 60% of its initial translational velocity or brought it almost to rest (an energy reduction of 65 and 100%, respectively). The report includes structural and finite element analysis and data interpretation, estimates of energy during impact, missile displacement and velocity histories, and selected strain gage data.

  16. Comparison of BALON2 with cladding ballooning strain tables in NUREG-0630. [PWR; BWR

    SciTech Connect (OSTI)

    Resch, S.C.; Laats, E.T.

    1982-01-01T23:59:59.000Z

    For this comparison study, the two computer models used for calculating fuel rod cladding failure and the resulting permanent strains were compared against experiment data. The two models considered were the mechanistic BALON2 model and the empirical model described in the NUREG-0630 report. The purpose for making this comparison was simply to gain insight into the relative strengths and weaknesses of each model. The experiment data sample consisted of data from both single and bundle tests conducted sometimes in in-pile facilities, but mostly in out-of-pile facilities. Comparisons between models indicated that the empirical NUREG-0630 model more accurately calculated the local cladding temperature and pressure conditions at rupture, but the mechanistic BALON2 model more accurately calculated the resulting cladding permanent strain at the rupture location.

  17. 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).

  18. 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 ...

  19. 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]

  20. 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...

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

    E-Print Network [OSTI]

    Alajo, Ayodeji Babatunde

    2009-05-15T23:59:59.000Z

    of legacy spent fuel that is given in the Yucca Mountain Safety Assessment Report. Chapter III discusses the details of the study. The concluding study was accomplished through the use of ORNL SCALE5.1 code system. The detailed whole-core 3-D models were...

  3. 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

  4. 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

    This work characterizes the dynamic behavior for the modified Large Scale Test Facility (LSTF), which has been selected by the U.S. Nuclear Regulatory Commission for confirmatory testing of the Westinghouse AP600 design. The LSTF is performing a...

  5. 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.

  6. Parametric calculations of fatigue-crack growth in piping. [PWR; BWR

    SciTech Connect (OSTI)

    Simonen, F.A.; Goodrich, C.W.

    1983-03-01T23:59:59.000Z

    A major objective of this program is to provide data that can be used to formulate recommended revisions to ASME Section XI and regulatory requirements for inservice inspection of piping and pressure vessels. This study presents calculations of the growth of piping flaws produced by fatigue. Flaw growth was predicted as a function of the initial flaw size, the level and number of stress cycles, the piping material, and environmental factors.

  7. Experimental facility for containment sump reliability studies (Generic Task A-43). [PWR; BWR

    SciTech Connect (OSTI)

    Durgin, W. W.; Padmanabhan, M.; Janik, C. R.

    1980-12-01T23:59:59.000Z

    On July 3, 1979, Sandia National Laboratories (Sandia) contracted the Alden Research Laboratory (ARL) to conduct tests on unresolved safety issues associated with containment sump performance during the recirculation mode (Generic Task A-43). This report describes the test facility constructed and completed under Phase I, Task III of the contract. Sump performance is determined through the observation of vortex formation in the main tank and the measurement of swirl, pressure gradient, and entrained air in the suction pipes. The use of electrically operated valves and a sophisticated data acquisition system, with computer interface, allows the test flow parameters to be set and test data to be taken (with the exception of vortex observations) from a single central office.

  8. 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)

  9. Analysis of operating data related to power and flow distribution in a PWR

    E-Print Network [OSTI]

    Herbin, Henry Christophe

    1974-01-01T23:59:59.000Z

    The analysis of the effects of the uncertainties associated with temperature and power measurements in the Connecticut Yankee Reactor leads to the evaluation of the uncertainty associated with the effective flow factor. ...

  10. Thermal and hydraulic code verification: ATHOS2 and Model Boiler No. 2 data. Final report. [PWR

    SciTech Connect (OSTI)

    Hopkins, G.W.; Lee, A.Y.; Mendler, O.J.

    1983-02-01T23:59:59.000Z

    As part of the EPRI/Westinghouse Project S168-1, Westinghouse was contracted to conduct steady-state and transient tests on the Westinghouse Model Boiler No. 2 (MB-2) steam genertor test model at the Engineering Test Facility in Tampa, Florida, and to use the data obtained in these tests for the verification of the ATHOS2 (an updated version of URSULA2) code developed for EPRI by CHAM of North America, Inc. This document presents a description of: (1) the model boiler and the associated test facility; (2) the ATHOS2 code analytical model of MB-2; (3) the tests performed for the code verification program; (4) the comparisons of the test data with ATHOS2 calculations; and (5) recommendations for improving the ATHOS2 code.

  11. Thermal and hydraulic code verification: ATHOS2 and Model Boiler No. 2 data. Final report. [PWR

    SciTech Connect (OSTI)

    Hopkins, G.W.; Lee, A.Y.; Mendler, O.J.

    1983-02-01T23:59:59.000Z

    As part of the EPRI/Westinghouse Project S168-1, Westinghouse was contracted to conduct steady-state and transient tests on the Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, and to use the data obtained in these tests for the verification of the ATHOS2 (an updated version of URSULA2) code developed for EPRI by CHAM of North America, Inc. This document presents a description of (1) the model boiler and the associated test facility, (2) the ATHOS2 code analytical model of MB-2, (3) the tests performed for the code verification program, (4) the comparisons of the test data with ATHOS2 calculations, and (5) recommendations for improving the ARHOS2 code.

  12. Effect of calcium hydroxide and carbonates on IGA and SCC of Alloy 600. Final report. [PWR

    SciTech Connect (OSTI)

    Balavage, J.R.

    1983-05-01T23:59:59.000Z

    A series of five tests was conducted at the Westinghouse Forest Hills Single Tube Model Boiler (STMB) Facility under this project. The objective of the project was to determine if alkaline earth carbonates and/or hydroxides are key ingreidents in causing intergranular attack (IGA) or stress corrosion cracking (SCC) of mill-annealed Alloy 600. Also as part of this program a report was written and issued detailing the earlier related STMB work Westinghouse conducted that led to reproducible SCC. The work reported here was an extension of that work.

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

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    AND OF THE CORE CONVERSION IN MNSR USING THE MURE CODE R�SUM� Le code MURE est basé sur le couplage d'un code questions : le mono- recyclage de l'Am dans les réacteurs français de type REP et la conversion du coeur du MNSR (Miniature Neutron Source Reactor) au Ghana d'un combustible à uranium hautement enrichi (HEU

  14. 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.

  15. 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) ...

  16. 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 ...

  17. 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 ...

  18. 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.

  19. E-Print Network 3.0 - advanced pwr fuel Sample Search Results

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

    Advanced nuclear reactor theory... equations, prompt jump approximation; subcritical reactor kinetics, circulating fuel reactor dynamics 5... Short-term Reactivity...

  20. 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 Data Center Home Page on Delicious Rank EERE:YearRound-UpHeatMulti-Dimensional ElectricalEnergy Frozen Telescope Looks to Ends

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

    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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear SecurityTensile Strain Switched Ferromagnetism inS-4500IIVasudhaSurface. |VolunteeringMap2-5: EastW.W7WAPD-SC-545

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

    E-Print Network [OSTI]

    Capone, Luigi

    2012-10-19T23:59:59.000Z

    4.3 The Numerical Approach ......................................................... 95 viii 4.4 Implementation in CFD Codes ................................................. 99 4... Total standard error refinement 3 ................................................................... 56 xii Figure 62 Experimental results plane B1 axial velocity (a), KER 138M mesh constant inlet (b), SST138M mesh periodic inlet (c...

  3. 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 ...

  4. 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.

  5. 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 ...

  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 on Delicious Rank EERE:Year in Review: TopEnergyIDIQ Contract ESPC IDIQEnergy EdisonNon-Road Engines, Report 1

  7. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Career Scientists'Montana.Program - LibbyofThisStatement Tuesday,Department ofNon-Road Engines, Report

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

    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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Career Scientists'Montana.Program - LibbyofThisStatementNOTElectricityof Energy776 I Street,linesof

  9. 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: VegetationEquipment Surfaces andMapping Richland OperationsU.S. CommercialIn this paper, weSchool

  10. Reactivity initiated accident test series Test RIA 1-4 fuel behavior report. [PWR; BWR

    SciTech Connect (OSTI)

    Cook, B.A.; Martinson, Z.R.

    1984-09-01T23:59:59.000Z

    This report presents and discusses results from the final test in the Reactivity Initiated Accident (RIA) Test Series, Test RIA 1-4, conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. Nine preirradiated fuel rods in a 3 x 3 bundle configuration were subjected to a power burst while at boiling water reactor hot-startup system conditions. The test resulted in estimated axial peak, radial average fuel enthalpies of 234 cal/g UO/sub 2/ on the center rod, 255 cal/g UO/sub 2/ on the side rods, and 277 cal/g UO/sub 2/ on the corner rods. Test RIA 1-4 was conducted to investigate fuel coolability and channel blockage within a bundle of preirradiated rods near the present enthalpy limit of 280 cal/g UO/sub 2/ established by the US Nuclear Regulatory Commission. The test design and conduct are described, and the bundle and individual rod thermal and mechanical responses are evaluated. Conclusions from this final test and the entire PBF RIA Test Series are presented.

  11. 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.

  12. Approach to fluid-mechanics calculations on serial and parallel computer architectures. [PWR; BWR; TRAC code

    SciTech Connect (OSTI)

    Liles, D.R.; Mahaffy, J.H.; Giguere, P.T.

    1983-01-01T23:59:59.000Z

    The Transient Reactor Analysis Code (TRAC) is a large FORTRAN thermal-hydraulics program designed to solve problems involving internal flows in nuclear reactors. The current versions have been designed for a CDC 7600 and CRAY 1 but benchmarks have been run in parallel simulations. This paper will discuss the methods in use, the reason that these techniques are effective, and their extension to parallel machines.

  13. 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.

  14. Parametric Analysis of PWR Spent Fuel Depletion Parameters for Long-Term-Disposal Criticality Safety

    SciTech Connect (OSTI)

    DeHart, M.D.

    1999-08-01T23:59:59.000Z

    Utilization of burnup credit in criticality safety analysis for long-term disposal of spent nuclear fuel allows improved design efficiency and reduced cost due to the large mass of fissile material that will be present in the repository. Burnup-credit calculations are based on depletion calculations that provide a conservative estimate of spent fuel contents (in terms of criticality potential), followed by criticality calculations to assess the value of the effective neutron multiplication factor (k(sub)eff) for the a spent fuel cask or a fuel configuration under a variety of probabilistically derived events. In order to ensure that the depletion calculation is conservative, it is necessary to both qualify and quantify assumptions that can be made in depletion models.

  15. Electrical impedance string probes for two-phase void and velocity measurements. [PWR

    SciTech Connect (OSTI)

    Hardy, J E; Hylton, J O

    1982-05-01T23:59:59.000Z

    An instrumentation scheme has been developed to measure two-phase flow velocity and void fraction during the refill/reflood stages of a loss-of-coolant accident in experimental test facilities. The instrumentation's principle of operation was based on measurement of the electrical impedance of two-phase mixtures. Two-phase velocity is estimated by time-of-flight analysis of signals from two spatially separate sensors. A relative capacitive technique was employed to measure void fraction. The impedance sensor consists of a pair of stainless steel wires strung back and forth across a stainless steel frame. This sensor was dubbed string probe for this reason. The string probe was designed to withstand temperatures of 350/sup 0/C, thermal transients of approx. 300/sup 0/C/s, and severe fluid- and condensation-induced shocks.

  16. The Honorable Anthony Mosillo'

    Office of Legacy Management (LM)

    DC 20585 The Honorable Anthony Mosillo' ', ' 6500 Niagara Square. 'Buffalo, New York 14202 Dear Mayor Mosillo: , ' ", Secretary of Energy Hazel, O'Leary has announced...

  17. E-Print Network 3.0 - artificial canal network Sample Search...

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

    (opened to Lake Ontario in 1819) Hudson River Erie Canal 12... ;Gained access to Lake Erie through Welland Canal around Niagara Falls (completed 1829), but not noted... ...

  18. E-Print Network 3.0 - actin-lined canals controls Sample Search...

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

    (opened to Lake Ontario in 1819) Hudson River Erie Canal 12... ;Gained access to Lake Erie through Welland Canal around Niagara Falls (completed 1829), but not noted... ...

  19. ,"Plant","Primary Energy Source","Operating Company","Net Summer...

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

    York" ,"Plant","Primary Energy Source","Operating Company","Net Summer Capacity (MW)" 1,"Robert Moses Niagara","Hydroelectric","New York Power Authority",2353.2...

  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. "To Feel the Drumming Earth Come Upward": Indigenizing the American Studies Discipline, Field, Movement

    E-Print Network [OSTI]

    Clark, D. Anthony Tyeeme; Yetman, Norman R.

    2006-03-01T23:59:59.000Z

    two of whom (Rob ert K. Thomas and Shirley Hill Witt) identified as American Indians (Oklahoma Cherokee and Akwesasne Mohawk, Wolf Clan, respectively).17 Finally, the issue's contemporary focus contrasted with the emphasis on the past in most...

  2. 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...

  3. Protecting Elm Trees from Elm Bark Beetle on the Texas High Plains

    E-Print Network [OSTI]

    Porter, Patrick; Baugh, Brant A.; Siders, Kerry; Riley, Cherinell; Young, Stanley

    2001-06-27T23:59:59.000Z

    is to grow a Texas native. Several new USDA selections and hybrids look promising. Pecan (native) The state tree of Texas does well in most locations in the state. In the South Plains, ?Pawnee,? Carya illinoinensis ?Caddo,? ?Shoshoni,? ?Maramec,? ?Mohawk...

  4. CX-004898: Categorical Exclusion Determination

    Broader source: Energy.gov [DOE]

    Gila-Wellton-Mohawk (Structure Maintenance)CX(s) Applied: B1.3Date: 11/05/2010Location(s): Yuma County, ArizonaOffice(s): Western Area Power Administration-Desert Southwest Region

  5. A review of "The Poetics of Melancholy in Early Modern England." by Douglas Trevor

    E-Print Network [OSTI]

    Thomas P. Anderson

    2005-01-01T23:59:59.000Z

    as the social forces that regulate identity, the book represents an important contribution to theories of Renaissance subjec- tivity and identity. Allan Greer. Mohawk Saint: Catherine Tekakwitha and the Jesuits. Oxford: Oxford University Press, 2005. xiv...

  6. 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

  7. Effect of temperature on the fatty acid profile of pecan oil

    E-Print Network [OSTI]

    Bahri, Sami

    1992-01-01T23:59:59.000Z

    accumulated 12 weeks prior to shuck split were studied. Growing area affected the fatty acid profile for a l l cultivars. 'Cheyenne' and 'Mohawk' showed a positive correlation between heat units and oleic/linoleic acid ratios (r = 0.905 and r = 0.... The posit ive response of 'Cheyenne' and 'Mohawk' to temperature indicates that the oleic, and linoleic composition of these cultivars might be controlled by the selection of growing location. SUMMARY AND CONCLUSIONS The primary intent of this study...

  8. 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.

  9. Assessment of Biasi and Columbia University CHF correlations with GE 3x3 rod bundle experiment. [PWR; BWR

    SciTech Connect (OSTI)

    Chen, B.C.J.; Chien, T.H.; Sha, W.T.; Kim, J.H.

    1984-01-01T23:59:59.000Z

    The critical heat flux (CHF), at which a sudden degradation of heat transfer occurs without corresponding decrease in heat generation, is one of the limiting parameters for safe operation of nuclear reactors. Reactor operation beyond the CHF causes a rapid rise in fuel cladding temperature and thus should be avoided to maintain the fuel element integrity. Reactor power limits are therefore set so that a prescribed safety margin below the CHF is maintained. Two CHF correlations are evaluated for reactor core thermal hydraulic analysis: the Biasi correlation and the Columbia University correlation. The BODYFIT-2PE computer code is used for this assessment. The CHF predicted by the BODYFIT-2PE using the two correlations is compared with GE 3x3 rod bundle CHF experiment.

  10. Critical heat flux predictions based on the BODYFIT-2PE computer code and Columbia University CHF correlation. [PWR; BWR

    SciTech Connect (OSTI)

    Chen, B.C.J.; Chien, T.H.; Sha, W.T.; Kim, J.H.

    1983-01-01T23:59:59.000Z

    The BODYFIT-2PE (Boundary-Fitted Coordinate, 2-Phase Flow with Partially Elliptic) computer code has been developed and was employed to simulate the critical heat flux experiment in a General Electric 3 x 3 rod bundle by using a CHF correlation recently developed at Columbia University under EPRI sponsorship. CHF predictions are important in analyzing rod bundle performance in nuclear reactor operation. The results of the BODYFIT calculations compared favorably with the experimental measurements.

  11. Seismic fragility testing of naturally-aged, safety-related, class 1E battery cells. [PWR; BWR

    SciTech Connect (OSTI)

    Bonzon, L.L.; Hente, D.B.; Kukreti, B.M.; Schendel, J.S.; Black, D.A.; Paulsen, G.D.; Tulk, J.D.; Janis, W.J.; Aucoin, B.D.

    1984-01-01T23:59:59.000Z

    The concern over seismic susceptibility of naturally-aged lead-acid batteries used for safety-related emergency power in nuclear power stations was brought about by battery problems that periodically had been reported in Licensee Event Reports (LERs). The Turkey Point Station had reported cracked and buckled plates in several cells in October 1974 (LER 75-5). The Fitzpatrick Station had reported cracked battery cell cases in October 1977 (LER 77-55) and again in September 1979 (LER 79-59). The Browns Ferry Station had reported a cracked cell leaking a small quantity of electrolyte in July 1981 (LER 81-42). The Indian Point Station had reported cracked and leaking cells in both February (LER 82-7) and April 1982 (LER 82-16); both of these LERs indicated the cracked cells were due to expansion (i.e., growth) of the positive plates.

  12. 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.

  13. 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.

  14. 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.

  15. Calculation of the neutron source distribution in the venus PWR engineering mock-up and comparison with experimental measurements

    SciTech Connect (OSTI)

    Morakinyo, P.; Williams, M.L.; Kam, F.B.K.

    1984-01-01T23:59:59.000Z

    The VENUS experiment is sponsored by the USNRC in conjunction with CEN/SCK in Mol, Belgium. The VENUS configuration consists of a central water hole, surrounded by a 2.888 cm thick inner sheet baffle. The inner core zone in the immediate vicinity of the inner baffle contains 752 3.3% /sup 235/U, zircalloy fuel cells, with 48 pyrex rods interspersed among them. The outer core zone contains 1800 4.0% /sup 235/U, steel clad fuel cells. The core itself is surrounded by a 2.858 cm thick outer steel baffle, a water reflector, a 4.972 cm thick steel core barrel, a water gap, a neutron pad, and the pressure vessel. The primary aim of this study is to calculate the VENUS neutron source distribution, as part of the USNRC's overall program goal of benchmarking RPV fluence calculations. Of particular concern is the accuracy of the source calculation near the core-baffle interface, which is the important region for contributing to RPV fluence. The accuracy of the calculations is evaluated by comparison with gamma scan measurements performed by CEN/SCK.

  16. Prediction of annular liquid-gas flow with entrainment: cocurrent vertical pipe flow with gravity. [PWR; BWR

    SciTech Connect (OSTI)

    Levy, S.; Healzer, J.M.

    1980-09-01T23:59:59.000Z

    A simplified semi-empirical model is developed for annular two-phase (gas-liquid) flow with liquid entrainment in a vertical pipe. Gravity effects are included. Model predictions are compared to test data obtained with air-water, air-trichloroethane, and steam-water mixtures. The agreement is generally good between model and test results for pressure drop, liquid film thickness and wavyness, and liquid entrainment.

  17. Experimental investigation of a flow monitoring instrument in an upper plenum of an air-water reflood test facility. [PWR

    SciTech Connect (OSTI)

    Combs, S.K.; Hardy, J.E.

    1980-01-01T23:59:59.000Z

    Instrumentation was developed for measuring fluid phenomena in the upper plenum of pressurized water reactor reflood facilities. In particular, the instrumentation measured two-phase flow velocity and void fraction. The principle of operation of the instrumentation scheme was based on the measurement of electrical impedance. The technique of analysis of random signals from two spatially separated impedance sensors was employed to measure two-phase flow velocity. A relative admittance technique was used to determine void fraction. The performance of the instrumentaton was studied in an air-water test facility.

  18. Catalysinganenergyrevolution Nuclear Failures

    E-Print Network [OSTI]

    Laughlin, Robert B.

    . First and foremost responsible for overseeing development of the electricity supply across France, today CSA Soulaines Bure Pressurised Water Reactor (PWR) PWR loaded with MOX (or licensed to be) PWR under construction Fast Breeder Reactor (FBR) GCR reactor Heavy

  19. - United States Government

    Office of Legacy Management (LM)

    .' 41 G I? SUBJECT: Elimination of the T itanium Alloy Manufacturing Co., Niagara Falls, New York TO: The F ile I have reviewed the attached site. summary and elimination...

  20. 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:...

  1. Amtrak and its Host Railroads Paul Vilter, Amtrak AVP Host Railroads

    E-Print Network [OSTI]

    Bustamante, Fabián E.

    Brunswick FDOT VTR NECR Pan Am MBTA Metro- North SCRRA Metra Operated by VIA Rail SCRRA / BNSF / SDN Richmond Springfield Rutland Albany Pontiac Port Huron Niagara Falls St. Albans Brunswick MBTA (Boston

  2. 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

  3. 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

  4. 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

  5. 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...

  6. E-Print Network 3.0 - americium Sample Search Results

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

    thermal response: removal of plutonium, americium... , cesium, and strontium from spent PWR fuel,americium, cesium, and strontium from spent PWR fuel,...

  7. E-Print Network 3.0 - americium 241 Sample Search Results

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

    thermal response: removal of plutonium, americium... , cesium, and strontium from spent PWR fuel,americium, cesium, and strontium from spent PWR fuel, recycling...

  8. Comparison of different methods of aggregation of model ensemble outcomes in the validation and reconstruction of real power plant signals

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    signals measured at a Finnish nuclear Pressurized Water Reactor (PWR) located in Loviisa. References

  9. 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

  10. 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...

  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. 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

  13. The identity and construction of Wreck Baker: a War of 1812 era Royal Navy frigate

    E-Print Network [OSTI]

    Walker, Daniel Robert

    2009-06-02T23:59:59.000Z

    for this new command.28 Chauncey immediately undertook to send 170 sailors and marines, 140 ship carpenters, as well as more than 100 cannon and other supplies to 17 Sackets Harbor. The supplies and men traveled up the Hudson River to the Mohawk and from...

  14. CX-009805: Categorical Exclusion Determination

    Broader source: Energy.gov [DOE]

    Maintenance and Wood Pole Replacement along the Gila Wellton Mohawk 161 Kilovolt Transmission Line CX(s) Applied: B1.3 Date: 01/03/2013 Location(s): Arizona Offices(s): Western Area Power Administration-Desert Southwest Region

  15. Variation in ecogeographical traits of pecan cultivars and provenances

    E-Print Network [OSTI]

    Sagaram, Madhulika

    2009-05-15T23:59:59.000Z

    and provenances (i.e., the area of origin of seed). An assessment of leaf anatomical traits of pecan cultivars (Pawnee, Mohawk and Starking Hardy Giant) collected from three locations (Tifton, GA., Chetopa, KS., and Stillwater, OK.) was conducted to provide...

  16. Multivariable analysis of the effects of Li, H{sub 2}, and pH on PWR primary water stress corrosion cracking. Final report

    SciTech Connect (OSTI)

    Eason, E.D.; Merton, A.A.; Wright, J.E.

    1996-05-01T23:59:59.000Z

    The effects of Li, pH and H, on primary water stress corrosion cracking (PWSCC) of Alloy 600 were investigated for temperatures between 320 and 330{degrees}C. Specimens included in the study were reverse U-bends (RUBs) made from several different heats of Alloy 600. The characteristic life, {eta}, which represents the time until 63.2% of the population initiates PWSCC, was computed using a modified Weibull statistical analysis algorithm and was analyzed for effects of the water chemistry variables previously mentioned. It was determined that the water chemistry variables are less sensitive than the metallurgical characteristics defined by the heat, heat treatment and initial stress state of the specimen (diameter and style of RUB); the maximum impact of chemistry effects was 0.13 to 0.59 standard deviations compared to a range of three (3) standard deviations for all variables. A first-order model was generated to estimate the effect of changes in pH, Li and H, concentrations on the characteristic life. The characteristic time to initiate cracks, {eta}, is not sensitive to Li and H{sub 2} concentrations in excess of 3.5 ppm and 25 ml/kg, respectively. Below these values, (1) {eta} decreases by {approximately}20% when [Li] is increased from 0.7 to 3.5 ppm; (2) {eta} decreases by {approximately}9% when [H{sub 2}] is increased from 13.1 to 25.0 ml/kg; and (3) {eta} decreases by {approximately}14% when pH is increased from 7.0 to 7.4, in each case holding the other two variables constant.

  17. Experimental investigations of uncovered-bundle heat transfer and two-phase mixture-level swell under high-pressure low heat-flux conditions. [PWR

    SciTech Connect (OSTI)

    Anklam, T. M.; Miller, R. J.; White, M. D.

    1982-03-01T23:59:59.000Z

    Results are reported from a series of uncovered-bundle heat transfer and mixture-level swell tests. Experimental testing was performed at Oak Ridge National Laboratory in the Thermal Hydraulic Test Facility (THTF). The THTF is an electrically heated bundle test loop configured to produce conditions similar to those in a small-break loss-of-coolant accident. The objective of heat transfer testing was to acquire heat transfer coefficients and fluid conditions in a partially uncovered bundle. Testing was performed in a quasi-steady-state mode with the heated core 30 to 40% uncovered. Linear heat rates varied from 0.32 to 2.22 kW/m.rod (0.1 to 0.68 kW/ft.rod). Under these conditions peak clad temperatures in excess of 1050 K (1430/sup 0/F) were observed, and total heat transfer coefficients ranged from 0.0045 to 0.037 W/cm/sup 2/.K (8 to 65 Btu/h.ft/sup 2/./sup 0/F). Spacer grids were observed to enhance heat transfer at, and downstream of, the grid. Radiation heat transfer was calculated to account for as much as 65% of total heat transfer in low-flow tests.

  18. 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.

  19. Design concept and testing of an in-bundle gamma densitometer for subchannel void fraction measurements in the THTF electrically heated rod bundle. [PWR

    SciTech Connect (OSTI)

    Felde, D. K.

    1982-04-01T23:59:59.000Z

    A design concept is presented for an in-bundle gamma densitometer system for measurement of subchannel average fluid density and void fraction in rod or tube bundles. This report describes (1) the application of the design concept to the Thermal-Hydraulic Test Facility (THTF) electrically heated rod bundle; and (2) results from tests conducted in the THTF.

  20. Development of a generalized correlation for phase-velocity measurements obtained from impedance-probe pairs in two-phase flow systems. [PWR

    SciTech Connect (OSTI)

    Hsu, C.T.; Keshock, E.G.; McGill, R.N.

    1983-01-01T23:59:59.000Z

    A flag type electrical impedance probe has been developed at the Oak Ridge National Lab (ORNL) to measure liquid- and vapor-phase velocities in steam-water mixtures flowing through rod bundles. Measurements are made by utilizing the probes in pairs, installed in line, parallel to the flow direction, and extending out into the flow channel. The present study addresses performance difficulties by examining from a fundamental point of view the two-phase flow system which the impedance probes typically operate in. Specifically, the governing equations (continuity, momentum, energy) were formulated for both air-water and steam-water systems, and then subjected to a scaling analysis. The scaling analysis yielded the appropriate dimensionless parameters of significance in both kinds of systems. Additionally, with the aid of experimental data obtained at ORNL, those parameters of significant magnitude were established. As a result, a generalized correlation was developed for liquid and vapor phase velocities that makes it possible to employ the impedance probe velocity measurement technique in a wide variety of test configurations and fluid combinations.

  1. Measurement of two-phase flow at the core/upper plenum interface for a PWR geometry under simulated reflood conditions

    SciTech Connect (OSTI)

    Thomas, D.G.; Combs, S.K.

    1983-08-01T23:59:59.000Z

    The Instrument Development Loop (IDL) Program is part of the International 2D/3D Refill and Reflood Experimental and Analytical Research Program. Among the objectives of the International Program are: the study of the steam binding effect during reflood flow distribution (chimney effect) in a heated core; and the study of flow hydrodynamics in the core, downcomer and upper plenum during refill and reflood. Three experimental facilities were used in these studies: a one-bundle air/water loop, a three-bundle air/water loop, and a one-bundle steam/water loop. The loops represent full-scale vertical sections of the UPTF, extending from spray nozzles to the top of the upper plenum and including a short length of dummy fuel rods, upper end boxes, core support plate and control rod guide tubes. Three flow regimes were identified and studied: (1) all liquid down; (2) countercurrent flow in which gas (or vapor) goes up and liquid goes both up and down; and (3) cocurrent flow in which both gas (or vapor) and liquid go up.

  2. 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."...

  3. 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."...

  4. 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."...

  5. 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."...

  6. Robust nuclear signal reconstruction by a novel ensemble model aggregation procedure P. Baraldi1

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    at the Finnish Pressurized Water Reactor (PWR) situated in Loviisa. 1. Introduction Sensors are placed at various Pressurized Water Reactor (PWR) situated in Loviisa. The novel procedure is applied and tested on both

  7. April 2008 Nuclear News

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

    Constellation 2. South Texas-1 96.88 1250.6 PWR STPNOC 3. FitzPatrick 96.69 816 BWR Entergy 4. Ginna 96.55 585 PWR Constellation 5. Braidwood-2 96.45 1155 PWR Exelon 6. Comanche...

  8. 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

  9. Recent Changes in Greenland & Antarctica

    E-Print Network [OSTI]

    Bitz, Cecilia

    lake drainage her height Ã?5 = old water level Lake used to be 8m (24 ft) deep Flow rate Niagara Falls depth: Volume of water in lake over time Field project: If meltwater reaches the bed, does it affect ice the ability to make accurate projections (IPCC, 2007). From ice sheets /glaciers: 0.04 to 0.23 meters #12;IPCC

  10. 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 for long distance transfers · Polyphase motors to use the current · Built the world's first hydroelectric plant at Niagara Falls 1895 #12;Nikola Tesla's inventions · Radio/wireless transmission · US Supreme

  11. Great Lakes Geologic Mapping Coalition -Annual Science Meeting April 15-17, 2014

    E-Print Network [OSTI]

    Polly, David

    :40 Innovative 3D digital map demonstration Abigail Burt, MNDMF 9:00 The passive-seismic experiment (HVSR of Calhoun County Al Kehew, MIGS 3:50 Rationale and methods for regional 3D geological mapping Harvey Thorleifson, MNGS 4:10 The Niagara Peninsula in 3D and a cool mini-project Abigail Burt, MNDMF 4

  12. Lexical Substitutability 1. Too Much and Too Little Evidence

    E-Print Network [OSTI]

    Church, Kenneth W.

    . A lexicographer is like a person standing underneath Niagara Falls holding a rainwater gauge, while the evidence. Some seek to capture particularly fine and unusual droplets of spray: they collect citations for rare for a citation collection might not be the most appropriate sampling procedure, especially if one wants

  13. Water Qual. Res. J. Canada, 2004 Volume 39, No. 3, 213222 Copyright 2004, CAWQ

    E-Print Network [OSTI]

    Mazumder, Asit

    by algae are geosmin and MIB (2- methylisoborneol). In Canada, geosmin and MIB have been detected in Lake Ontario, the St. Lawrence River, and treatment plants that draw water from Lake Erie and from the Niagara for the analysis of lake and reservoir water samples from several coastal British Columbia drinking water sources

  14. 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

  15. 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

  16. Experience an Ivy League institution on Exchange for a semester or a year!

    E-Print Network [OSTI]

    Keinan, Alon

    Tuh-GAN-ick), is three stories taller than Niagara Falls Along with the falls, gorges and beautiful lakefront Transportation Ithaca Airport: Direct flights to Detroit, Philadelphia and NYC Good public bus system (TCAT Engineering Hotel Administration Human Ecology Industrial & Labor Relations Second largest college

  17. 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...

  18. Mohegan Tribal Utility Auth | 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk Municipal Comm Jump to:

  19. Mojave, California: 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk Municipal Comm Jump to:Mojave,

  20. Moldova National Inventory Report - Lessons Learned | Open Energy

    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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk Municipal Comm Jump

  1. Mongolia-GTZ Development of RE 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk Municipal Comm Jump(EC-LEDS) |

  2. Monitoring and Assessment of Greenhouse Gas Emissions and Mitigation

    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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk Municipal Comm Jump(EC-LEDS)

  3. Monitoring, Verification and Reporting: Improving Compliance Within Energy

    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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk Municipal Comm

  4. Monroe County Elec 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk Municipal CommMonongahela Power Co

  5. Montana-Dakota Utilities Co (South Dakota) | 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk Municipal CommMonongahela Power

  6. Montana: 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk Municipal CommMonongahela

  7. Montenegro: 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk Municipal CommMonongahelaMontenegro:

  8. Montpelier, Vermont: 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk Municipal

  9. Montvale, New Jersey: 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk MunicipalMontvale, New Jersey: Energy

  10. Moon Lake Electric Assn 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk MunicipalMontvale, New Jersey:

  11. Mora-San Miguel Elec 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk MunicipalMontvale, New

  12. Morocco-Enhancing Low-carbon Development by Greening the Economy: Policy

    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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk MunicipalMontvale,

  13. Morocco-Low Carbon Development Planning in the Power Sector | Open Energy

    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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk MunicipalMontvale,GTZ Partner Ministry

  14. Morocco-UNEP Green Economy Advisory Services | 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk MunicipalMontvale,GTZ Partner

  15. Morocco-UNEP Risoe Technology Needs Assessment Program | Open Energy

    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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk MunicipalMontvale,GTZ

  16. Motor Vehicle Emission Simulator (MOVES) | 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk MunicipalMontvale,GTZVehicle Emission

  17. Mountain Electric Coop, Inc (North Carolina) | 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk MunicipalMontvale,GTZVehicle

  18. Mountain Electric 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk MunicipalMontvale,GTZVehicleMountain

  19. Mountain View Elec Assn, 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk MunicipalMontvale,GTZVehicleMountainAssn,

  20. Mountain View, California: 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk

  1. 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.

  2. 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

  3. V

    Office of Scientific and Technical Information (OSTI)

    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,...

  4. E-Print Network 3.0 - ac metal-enclosed switchgear Sample Search...

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

    Collection: Mathematics 70 jiangy@iastate.edu Zhong Zhang Summary: , "Reliability centered maintenance applied to electrical switchgear," Proc. IEEE Pwr Engr Society Meeting,...

  5. Object Management Systems

    E-Print Network [OSTI]

    Gollu, Aleks Ohannes

    1995-01-01T23:59:59.000Z

    Object-Oriented Database Management Systems for EngineeringR. Cassel. Distribution Management Systems: Functions and8-PWR 1988. Network Management Systems 52 Subodh Bapat.

  6. "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

  7. TRACE Code Validation for Natural Circulation During Small Break LOCA in EPR-Type Reactor.

    E-Print Network [OSTI]

    Bertran Morancho, Joan

    2011-01-01T23:59:59.000Z

    ?? The PWR PACTEL test facility was built in Lappeenranta (Finland) to gain experience in thermal-hydraulics behavior of vertical steam generators used by EPR (European… (more)

  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. E-Print Network 3.0 - automatic reactor control Sample Search...

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

    Control: PWR v. BWR, and systems (e.g., pressurizer & steam... generator) control 9 Power Plant Modeling: reactor ... Source: Zhang, Junshan - Department of Electrical...

  10. March 2014 Most Viewed Documents for Fission And Nuclear Technologies...

    Office of Scientific and Technical Information (OSTI)

    J. (1978) 30 > Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Executive summary: main report. PWR and BWR Not Available (1975)...

  11. Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors

    E-Print Network [OSTI]

    Terrani, Kurt Amir

    2010-01-01T23:59:59.000Z

    of hydride fueled BWRs. Nuclear Engineering and Design, 239:Fueled PWR Cores. Nuclear Engineering and Design, 239:1489–Hydride Fueled LWRs. Nuclear Engineering and Design, 239:

  12. 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

  13. Energy Systems Technology - A Development in Experiential Learning

    E-Print Network [OSTI]

    Tumber, A. J.

    1980-01-01T23:59:59.000Z

    outside I make a smooth transition away from the Hamilton but Mohawk is the only college in fossil fuels on which we have become so Ontario, or in Canada, to offer this pro dependent. The theme of this conference is gram. conservation through energy... and shorter certificate programs. of the Steel Company of Canada and Dominion These colleges, unlike many in other parts Foundries and Steel, the two largest pro of Canada and in the United States, do not ducers of steel in Canada. Firestone, function...

  14. Warped Space Issue 29/30

    E-Print Network [OSTI]

    Multiple Contributors

    1977-01-01T23:59:59.000Z

    and the Bandersnatchi Press have moved to 2100 N. Halsted, Third Floor, Chicago, IL, 60614. Steve has been in the process of moving and hopes to catch up real soon now on answering all orders and inquiries. The album (Leslie Fish's) and filksong booklet should be out.... But it's done with flair, it hangs to gether. It is for enjoyment. "To the Water and the Wild" — Spock and an ersatz unicorn? Amazing how a Vulcan who goes into Pon Farr as often as he does can remain a virgin ... —Leslie Fish, 1906 N. Mohawk, rear house...

  15. Moldova-Enhancing Capacity for Low Emission Development Strategies

    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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk Municipal Comm Jump(EC-LEDS) | Open

  16. Mongolia-GTZ Energy Efficiency within the Grid-Connected Energy Supply |

    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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk Municipal Comm Jump(EC-LEDS) |Open Energy

  17. Monitoring Climate Finance and ODA | 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk Municipal Comm Jump(EC-LEDS) |Open

  18. Monitoring and Assessment of Greenhouse Gas Emissions and Mitigation

    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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk Municipal Comm Jump(EC-LEDS)Potential in

  19. Monitoring and Tracking Long-Term Finance to Support Climate Action | Open

    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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk Municipal Comm Jump(EC-LEDS)Potential

  20. Monongahela Power Co (West Virginia) | 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk Municipal CommMonongahela Power Co (West

  1. Mont Alto Borough | 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk Municipal CommMonongahela Power CoBorough

  2. Montana-Dakota Utilities Co (Wyoming) | 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk Municipal CommMonongahela PowerWyoming

  3. Moon Lake Electric Assn Inc (Utah) | 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk MunicipalMontvale, New Jersey: EnergyInc

  4. Moose Lake Water & Light 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk MunicipalMontvale, New Jersey:Water &

  5. Mor-Gran-Sou Electric 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk MunicipalMontvale, New Jersey:Water

  6. Moreau-Grand Electric 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk MunicipalMontvale, NewMoreau-Grand

  7. Morocco-GTZ Promotion of EERE | 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 on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of InspectorConcentrating Solar Powerstories onFocus Area EnergyMohawk MunicipalMontvale,GTZ Partner Ministry of

  8. Assessment of uranium-free nitride fuels for spent fuel transmutation in fast reactor systems

    E-Print Network [OSTI]

    Szakaly, Frank Joseph

    2004-09-30T23:59:59.000Z

    . ....................................................................................... 18 Fig. 4. Standard PWR ¼ core model with fresh, once- and twice-burned fuel, and the location of MOX fuel assemblies with respect to original layout, 32% MOX loading................................................................................................................ 21 Fig. 5. Control rod locations......................................................................................... 21 Fig. 6. Net change of U, Pu and Am for PWR and 1/3 MOX fueled whole cores, 360 day burn...

  9. Open Archive TOULOUSE Archive Ouverte (OATAO) OATAO is an open access repository that collects the work of Toulouse researchers and

    E-Print Network [OSTI]

    Boyer, Edmond

    and Feuillebois, François and Simonin, Olivier Eulerian Simulation of Interacting PWR Sprays Including Droplet. 133-143" #12;KEYWORDS: containment, spray, collision EULERIAN SIMULATION OF INTERACTING PWR SPRAYS d Institut de Mécanique des Fluides de Toulouse, Toulouse, France A numerical simulation

  10. The long-term development of a watershed: spatial patterns, streamflow, and sustainability

    E-Print Network [OSTI]

    DeFee, Buren Brooks, II

    2005-02-17T23:59:59.000Z

    for determining developed patches....................................................... 68 18 Maps of the study area showing development patch configuration for roads (Rds) alone and parcels with roads (PWR) for selected years... index for roads and parcels with roads. ..................... 78 27 Landscape division for roads and parcels with roads by year. ....... 79 28 Lacunarity curves for parcels with roads for five time points. ...... 80 29 PWR lacunarity over time...

  11. Document (501k)

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

    DebtCapital DebtInterest DebtN DeconYr DeconYrs DeconPmt ELN EquityN FTax GenPer GPP GridPWR GridElec1 GridElec2 GridElecPrice GrossPWR HeatFlow HeatPrice HeatPriceLB...

  12. RIS-M-2256 INPUT DESCRIPTION FOR THE THREE-DIMENSIONAL

    E-Print Network [OSTI]

    , PWR TYPE REACTORS, REACTOR KINETICS, THREE-DIMENSIONAL CALCULATIONS, TRANSIENTS. UDC 621.039.514 : 621 calculations for the PWR core. It combines a nodal theory neutron kinetics calculation with transient sub calculations. The present report describes the input for ANTI. INIS descriptors; A CODES, BURNUP, HYDRAULICS

  13. Control and Spring Emergence of the Cotton Flea Hopper.

    E-Print Network [OSTI]

    Reinhard, H. J. (Henry Jonathan)

    1927-01-01T23:59:59.000Z

    in experiments on control of the cotton flea hopper by the use of insecticides ap- plied as dusts and as sprays under field conditions. Superfine dusting sulphur, flowers of sulphur, Niagara sulphur-naphtha- lene, and mixtures of sulphur-tobacco dust resulted... in an aver- age daily control ranging from 68 to 75 per cent. When applied at the rate of 20 pounds per acre it was found that these dusts remained effective in preventing multiplication of the insects for a period of six or seven days under favorable...

  14. Kylteknik ("KYL")Kylteknik ("KYL") RefrigerationRefrigerationRefrigerationRefrigeration

    E-Print Network [OSTI]

    Zevenhoven, Ron

    + moisture a_Falls.jpg For the range -10 ~ +50°C, dry air can be treated as an ideal gas with c i 1 005 kJ/(kg· K) _Mist-Niagara cp,air 1.005 kJ/(kg K) The saturation pressure of water at 50°C is 12.3 k.3 kJ/kg c t 1.82 kJ/(kg· K) Picture:http:// 8.11.2012 �bo Akademi Univ - Thermal and Flow

  15. http://www.kasahara.cs.waseda.ac.jp/ . Everywhere

    E-Print Network [OSTI]

    Kasahara, Hironori

    Power4,5,5+,6, HPCS 10PFLOP(Power7)NEDO OSCAR , , , , ( ) Sun Niagara(SparcT1,T2), Rock , 20023, 5120.7 2008.3)** OSCAR: Optimally Scheduled Advanced Multiprocessor DVD TV CMP m ( m) ( PC0 ( 0) PC1 ( 1.0) (1.0) (1.8) 00 2.0 Fortran Ver.8.13.5 10.7 0.0 tomcatv swim su2cor hydro2d mgrid applu turb3d apsi

  16. E-Print Network 3.0 - actinides fuel research Sample Search Results

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

    Summary: in safety, proliferation resistance, and can be designed to breed fuel or burn heavy actinides. One... . The number of fuel pins in a fuel assembly of a PWR core is...

  17. E-Print Network 3.0 - actinide burning fuel Sample Search Results

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

    Summary: in safety, proliferation resistance, and can be designed to breed fuel or burn heavy actinides. One... . The number of fuel pins in a fuel assembly of a PWR core is...

  18. An assessment of silicon carbide as a cladding material for light water reactors

    E-Print Network [OSTI]

    Carpenter, David Michael

    2011-01-01T23:59:59.000Z

    An investigation into the properties and performance of a novel silicon carbide-based fuel rod cladding under PWR conditions was conducted. The novel design is a triplex, with the inner and outermost layers consisting of ...

  19. CRUD resistant fuel cladding materials

    E-Print Network [OSTI]

    Paramonova, Ekaterina (Ekaterina D.)

    2013-01-01T23:59:59.000Z

    CRUD is a term commonly used to describe deposited corrosion products that form on the surface of fuel cladding rods during the operation of Pressurized Water Reactors (PWR). CRUD has deleterious effects on reactor operation ...

  20. E-Print Network 3.0 - abwr feedwater pump Sample Search Results

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

    Gases, Summary: 610 2011 China, Taiwan (2) Lungmen 1 ABWR 1,300 A Lungmen 2 ABWR 1,300 A Finland (1) Olkiluoto 3 PWR 1... Agency PRIS database http:www.iaea.orgprogrammesa2...

  1. Interim Report on Thermal Cycling Model Development for Representative Unisolable Piping Configurations (MRP-81)

    SciTech Connect (OSTI)

    J. Keller, A. Bilanin

    2002-11-30T23:59:59.000Z

    Thermal fatigue can lead to cracking in dead-ended branch lines attached to PWR primary coolant piping. This interim report describes the results of on-going research to provide an improved screening tool for identification of susceptible piping.

  2. Flooding Experiments with Steam and Water in a Large Diameter Vertical Tube

    E-Print Network [OSTI]

    Williams, Susan Nicole

    2010-10-12T23:59:59.000Z

    An experimental study on flooding with steam and water in a large diameter vertical tube was conducted. This research has been performed to provide a better prediction of flooding in a pressurized water reactor (PWR) pressurizer surge line...

  3. Nuclear power plants: structure and function

    SciTech Connect (OSTI)

    Hendrie, J.M.

    1983-01-01T23:59:59.000Z

    Topics discussed include: steam electric plants; BWR type reactors; PWR type reactors; thermal efficiency of light water reactors; other types of nuclear power plants; the fission process and nuclear fuel; fission products and reactor afterheat; and reactor safety.

  4. Rl*-M-2413 LIST OF SELECTED PUBLICATIONS 1982

    E-Print Network [OSTI]

    Cour, A Model of the Ringhals 3 PWR Power Plant. In: 10th IMACS World Congress on System Simulation. Lagerstrom, B. Lehnert, B. P. Peregud, A. Sillesen, and A. A. Semenov, Use of Thermal Imaging in Experiments

  5. CASL Plan of Record 2 (1/11-6/11)

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

    IC CASL Industry Council LWR Light Water Reactor M&S Modeling and Simulation NPP Nuclear Power Plant ORNL Oak Ridge National Laboratory PWR Pressurized Water Reactor R&D Research...

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

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

    PWR Fuel CRUD," Proceedings of the TMS 2013 142nd Annual Meeting and Exhibition, March 3-7, 2013, San Antonio, TX, 2013. Tryggvason, G., S. Dabiri, B. Aboulhasanzadeh, J. Lu.,...

  7. A core reload pattern and composition optimization methodology for pressurized water reactors

    E-Print Network [OSTI]

    Sauer, Ildo Luis

    1985-01-01T23:59:59.000Z

    The primary objective of this research was the development of a comprehensive, rapid and conceptually simple methodology for PWR core reload pattern and fuel composition optimization, capable of systematic incorporation ...

  8. Design strategies for optimizing high burnup fuel in pressurized water reactors

    E-Print Network [OSTI]

    Xu, Zhiwen, 1975-

    2003-01-01T23:59:59.000Z

    This work is focused on the strategy for utilizing high-burnup fuel in pressurized water reactors (PWR) with special emphasis on the full array of neutronic considerations. The historical increase in batch-averaged discharge ...

  9. ESAIM: PROCEEDINGS, Vol. ?, 2012, 1-10 Editors: Will be set by the publisher

    E-Print Network [OSTI]

    Boyer, Edmond

    cela, nous utilisons un modèle bas-Mach monophasique (modèle Lmnc) pour une loi d'état de type gaz circuit of a PWR. where steam is generated and ows to a turbine which, in turn, spins an electric

  10. Linear Extrusion 400 Tons/Day Dry Solids Pump

    SciTech Connect (OSTI)

    Kenneth Sprouse; David Matthews

    2008-04-30T23:59:59.000Z

    Pratt & Whitney Rocketdyne (PWR) has developed an innovative gasifier concept that uses rocket engine experience to significantly improve gasifier performance, life, and cost compared to current state-of-the-art systems. The PWR gasifier concept uses a compact and highly efficient (>50%) dry solids pump that has excellent availability (>99.5%). PWR is currently developing this dry solids pump under a U.S. Department of Energy (DOE) cooperative agreement. The conceptual design on two dry solids pumps were completed under this agreement and one pump concept was selected for preliminary design. A preliminary design review (PDR) of the selected pump was presented on September 20, 2007 to PWR management and numerous technical specialists. Feedback from the PDR review team has been factored into the design and a Delta-PDR was held on April 9, 2008.

  11. Most Viewed Documents for Fission And Nuclear Technologies: September...

    Office of Scientific and Technical Information (OSTI)

    PWR and BWR Not Available (1975) 27 Sloshing analysis of viscous liquid storage tanks Uras, R.Z. (1995) 22 Flow-induced vibration of circular cylindrical structures Chen,...

  12. Three-Dimensional Velocity Measurement Reconstruction for a Rod Bundle Array using Matched Refractive Index Particle Tracking Velocimetry

    E-Print Network [OSTI]

    Reyes, Denny L

    2013-08-09T23:59:59.000Z

    In a pressurized water reactor (PWR), pressurized water flows over fuel rods containing radioactive uranium. Potential failure of these nuclear fuel rods is a primary concern, as fuel rod failure typically results in power generation losses...

  13. Initial Modeling of a Pressurized Water Reactor Completed Using RELAP-7

    Broader source: Energy.gov [DOE]

    RELAP-7 is a nuclear reactor system safety analysis code where initial capabilities were demonstrated by simulating a steady-state single-phase pressurized water reactor (PWR) with two parallel loops and multiple reactor core flow channels.

  14. Optimization of the axial power shape in pressurized water reactors

    E-Print Network [OSTI]

    Melik, M. A.

    1981-01-01T23:59:59.000Z

    Analytical and numerical methods have been applied to find the optimum axial power profile in a PWR with respect to uranium utilization. The preferred shape was found to have a large central region of uniform power density, ...

  15. Proceedings: 2003 EPRI Workshop on Condensate Polishing

    SciTech Connect (OSTI)

    None

    2004-02-01T23:59:59.000Z

    Successful condensate polishing operations maintain control of ionic and particulate impurity transport to the pressurized water reactor (PWR) steam generator and the boiling water reactor (BWR) reactor and recirculation system, thus allowing the units to operate more reliably. This report contains the work presented at EPRI's 2003 Workshop on Condensate Polishing, where 30 papers were presented on current issues, research, and utility experiences involving polishing issues at both PWR and BWR units.

  16. Multirecycling of Plutonium from LMFBR Blanket in Standard PWRs Loaded with MOX Fuel

    SciTech Connect (OSTI)

    Sonat Sen; Gilles Youinou

    2013-02-01T23:59:59.000Z

    It is now well-known that, from a physics standpoint, Pu, or even TRU (i.e. Pu+M.A.), originating from LEU fuel irradiated in PWRs can be multirecycled also in PWRs using MOX fuel. However, the degradation of the isotopic composition during irradiation necessitates using enriched U in conjunction with the MOX fuel either homogeneously or heterogeneously to maintain the Pu (or TRU) content at a level allowing safe operation of the reactor, i.e. below about 10%. The study is related to another possible utilization of the excess Pu produced in the blanket of a LMFBR, namely in a PWR(MOX). In this case the more Pu is bred in the LMFBR, the more PWR(MOX) it can sustain. The important difference between the Pu coming from the blanket of a LMFBR and that coming from a PWR(LEU) is its isotopic composition. The first one contains about 95% of fissile isotopes whereas the second one contains only about 65% of fissile isotopes. As it will be shown later, this difference allows the PWR fed by Pu from the LMFBR blanket to operate with natural U instead of enriched U when it is fed by Pu from PWR(LEU)

  17. Assessment for advanced fuel cycle options in CANDU

    SciTech Connect (OSTI)

    Morreale, A.C.; Luxat, J.C. [McMaster University, 1280 Main St. W. Hamilton, Ontario, L8S 4L7 (Canada); Friedlander, Y. [AMEC-NSS Ltd., 700 University Ave. 4th Floor, Toronto, Ontario, M5G 1X6 (Canada)

    2013-07-01T23:59:59.000Z

    The possible options for advanced fuel cycles in CANDU reactors including actinide burning options and thorium cycles were explored and are feasible options to increase the efficiency of uranium utilization and help close the fuel cycle. The actinide burning TRUMOX approach uses a mixed oxide fuel of reprocessed transuranic actinides from PWR spent fuel blended with natural uranium in the CANDU-900 reactor. This system reduced actinide content by 35% and decreased natural uranium consumption by 24% over a PWR once through cycle. The thorium cycles evaluated used two CANDU-900 units, a generator and a burner unit along with a driver fuel feedstock. The driver fuels included plutonium reprocessed from PWR, from CANDU and low enriched uranium (LEU). All three cycles were effective options and reduced natural uranium consumption over a PWR once through cycle. The LEU driven system saw the largest reduction with a 94% savings while the plutonium driven cycles achieved 75% savings for PWR and 87% for CANDU. The high neutron economy, online fuelling and flexible compact fuel make the CANDU system an ideal reactor platform for many advanced fuel cycles.

  18. A framework for the assessment of severe accident management strategies

    SciTech Connect (OSTI)

    Kastenberg, W.E. [ed.; Apostolakis, G.; Dhir, V.K. [California Univ., Los Angeles, CA (United States). Dept. of Mechanical, Aerospace and Nuclear Engineering] [and others

    1993-09-01T23:59:59.000Z

    Severe accident management can be defined as the use of existing and/or altemative resources, systems and actors to prevent or mitigate a core-melt accident. For each accident sequence and each combination of severe accident management strategies, there may be several options available to the operator, and each involves phenomenological and operational considerations regarding uncertainty. Operational uncertainties include operator, system and instrumentation behavior during an accident. A framework based on decision trees and influence diagrams has been developed which incorporates such criteria as feasibility, effectiveness, and adverse effects, for evaluating potential severe accident management strategies. The framework is also capable of propagating both data and model uncertainty. It is applied to several potential strategies including PWR cavity flooding, BWR drywell flooding, PWR depressurization and PWR feed and bleed.

  19. Analysis of high-burnup fuel performance during load-follow operation

    SciTech Connect (OSTI)

    Matsui, T.; Fukuya, K.; Kinoshita, M.

    1987-01-01T23:59:59.000Z

    In Japan, an objective of the burnup extension of nuclear fuel is to raise the licensing limit of burnup from 39 to 48 GWd/t for pressurized water reactors (PWRs) in the near future. Because of an increasing ratio of nuclear power generation, the necessity of the load-follow operation, which responds flexibly to changing power demands, is more apparent. To evaluate accurately the mechanical integrity of PWR fuel at high burnup during a load-follow operation, the FEMAXI-III code, originally developed for analyses of fuel experiments, was modified, improving submodels to evaluate PWR fuel; the new code was named IRON. The results of verification work on the code using data on PWR fuel covering wide ranges of burnup and linear heat rate show that it has good predictability and, therefore, that the improvement was confirmed as effective.

  20. Seismic risk assessment of a BWR: status report

    SciTech Connect (OSTI)

    Chuang, T.Y.; Bernreuter, D.L.; Wells, J.E.; Johnson, J.J.

    1985-02-01T23:59:59.000Z

    The seismic risk methodology developed in the US NRC Seismic Safety Margins Research Program (SSMRP) was demonstrated by its application to the Zion nuclear power plant, a pressurized water reactor (PWR). A detailed model of Zion, including systems analysis models (initiating events, event trees, and fault trees), SSI and structure models, and piping models was developed and analyzed. The SSMRP methodology can equally be applied to a boiling water reactor (BWR). To demonstrate its applicability, to identify fundamental differences in seismic risk between a PWR and a BWR, and to provide a basis of comparison of seismic risk between a PWR and a BWR when analyzed with comparable methodology and assumptions, a seismic risk analysis is being performed on the LaSalle County Station nuclear power plant.

  1. Natural convection phenomena in a nuclear power plant during a postulated TMLB' accident

    SciTech Connect (OSTI)

    Domanus, H.M.; Schmitt, R.C.; Sha, W.T.; Shah, V.L.; Han, J.T.

    1987-01-01T23:59:59.000Z

    After the TMI (Three Mile Island) accident, there has been significant interest in analyzing and understanding the phenomena that may occur in a PWR (Pressurized Water Reactor) accident which may lead to partial or total core meltdown and degradation. Natural convection is one of the important phenomena. In the present paper the results of two numerical simulations of (1) four-loop PWR and (2) three-loop PWR are presented. The simulations were performed with the COMMIX(2) computer code. Our analysis shows that in severe accident scenarios, natural convection phenomena does occur and that it helps to delay core degradation by transferring decay heat from the reactor core to other internal structures of the reactor system. The amount of heat transfer and delay in core degradation depends on the geometry and internal structures of the system and on the events of an accident.

  2. NPP operation and maintenance with French-built NPPs

    SciTech Connect (OSTI)

    Mira, J.J.; Charbonneau, S.

    1988-01-01T23:59:59.000Z

    In France, 80% of the electricity production will be nuclear in 1990. More than to-day, PWR units will be operated by Electricite de France on a load-follow basis. Every effort is made to reduce planned and forced outages. Maintenance is shared between EdF and Framatome, the latter being in charge of high technology operations. All these actions are eased by the standardization of units, within each power class, the resulting build-up of experience being available to all PWR operators in the world, more particularly those of Framatome-built units.

  3. Knowledge and abilities catalog for nuclear power plant operators: Pressurized water reactors. Revision 1

    SciTech Connect (OSTI)

    NONE

    1995-08-01T23:59:59.000Z

    This document provides the basis for the development of content-valid licensing examinations for reactor operators and senior reactor operators. The examinations developed using the PWR catalog will cover those topics listed under Title 10, (ode of Federal Regulations Part 55. The PWR catalog contains approximately 5100 knowledge and ability (K/A) statements for reactor operators and senior reactor operators. The catalog is organized into six major sections: Catalog Organization; Generic Knowledge and Abilities; Plant Systems; Emergency and Abnormal Plant Evolutions; Components and Theory.

  4. Dry Coal Feed System and Multi-Element Injector Test Plan

    SciTech Connect (OSTI)

    Ken Sprouse; Fred Widman; Alan Darby

    2006-08-30T23:59:59.000Z

    Pratt & Whitney Rocketdyne (PWR) has developed an innovative gasifier concept that uses rocket engine technology to significantly improve gasifier performance, life, and cost compared to current state-of-the-art systems. One key feature of the PWR concept is the use of an ultra-dense phase feed system to provide dry coal to the multi-element injector. This report describes the layout, test procedures, instrumentation and data acquisition requirements for an ultradense phase multi-element injector and feed system to be operated at the University of North Dakota Energy and Environmental Research Center (UNDEERC).

  5. Use of Solid Hydride Fuel for Improved long-Life LWR Core Designs

    SciTech Connect (OSTI)

    Greenspan, E

    2006-04-30T23:59:59.000Z

    The primary objective of this project was to assess the feasibility of improving the performance of PWR and BWR cores by using solid hydride fuels instead of the commonly used oxide fuel. The primary measure of performance considered is the bus-bar cost of electricity (COE). Additional performance measures considered are safety, fuel bundle design simplicity – in particular for BWR’s, and plutonium incineration capability. It was found that hydride fuel can safely operate in PWR’s and BWR’s without restricting the linear heat generation rate of these reactors relative to that attainable with oxide fuel. A couple of promising applications of hydride fuel in PWR’s and BWR’s were identified: (1) Eliminating dedicated water moderator volumes in BWR cores thus enabling to significantly increase the cooled fuel rods surface area as well as the coolant flow cross section area in a given volume fuel bundle while significantly reducing the heterogeneity of BWR fuel bundles thus achieving flatter pin-by-pin power distribution. The net result is a possibility to significantly increase the core power density – on the order of 30% and, possibly, more, while greatly simplifying the fuel bundle design. Implementation of the above modifications is, though, not straightforward; it requires a design of completely different control system that could probably be implemented only in newly designed plants. It also requires increasing the coolant pressure drop across the core. (2) Recycling plutonium in PWR’s more effectively than is possible with oxide fuel by virtue of a couple of unique features of hydride fuel – reduced inventory of U-238 and increased inventory of hydrogen. As a result, the hydride fuelled core achieves nearly double the average discharge burnup and the fraction of the loaded Pu it incinerates in one pass is double that of the MOX fuel. The fissile fraction of the Pu in the discharged hydride fuel is only ~2/3 that of the MOX fuel and the discharged hydride fuel is more proliferation resistant. Preliminary feasibility assessment indicates that by replacing some of the ZrH1.6 by ThH2 it will be possible to further improve the plutonium incineration capability of PWR’s. Other possibly promising applications of hydride fuel were identified but not evaluated in this work. A number of promising oxide fueled PWR core designs were also found as spin-offs of this study: (1) The optimal oxide fueled PWR core design features smaller fuel rod diameter of D=6.5 mm and a larger pitch-to-diameter ratio of P/D=1.39 than presently practiced by industry – 9.5mm and 1.326. This optimal design can provide a 30% increase in the power density and a 24% reduction in the cost of electricity (COE) provided the PWR could be designed to have the coolant pressure drop across the core increased from the reference 29 psia to 60 psia. (2) Using wire wrapped oxide fuel rods in hexagonal fuel assemblies it is possible to design PWR cores to operate at 54% higher power density than the reference PWR design that uses grid spacers and a square lattice, provided 60 psia coolant pressure drop across the core could be accommodated. Uprating existing PWR’s to use such cores could result in 40% reduction in the COE. The optimal lattice geometry is D = 8.08 mm and P/D = 1.41. The most notable advantages of wire wraps over grid spacers are their significant lower pressure drop, higher critical heat flux and improved vibrations characteristics.

  6. A.E. K.Ris Ris -M -C509 Title and author(s)

    E-Print Network [OSTI]

    Station Model 4 1.3. Hybrid Computer 7 1.4. Analysis of Power Plant Control Tasks 9 1.5. Man Machine - CONTENTS Page 1. Systems Techniques 2 1.1. Off-line Graphics Terminal 2 1.2. Development of a PWR Power chapters will describe: graphics terminal, reactor model, kidney simulation, analysis of control tasks

  7. Spent Nuclear Fuel Self-Induced XRF to Predict Pu to U Content

    E-Print Network [OSTI]

    Stafford, Alissa Sarah

    2010-10-12T23:59:59.000Z

    ) in July 2008 and January 2009. These measurements successfully showed that it is possible to measure the Pu x-ray peak at 103.7 keV in PWR spent fuel (~1 percent Pu) using a planar HPGe detector. Prior to these measurement campaigns, the Pu peak has only...

  8. Deployment Scenario of Heavy Water Cooled Thorium Breeder Reactor

    SciTech Connect (OSTI)

    Mardiansah, Deby; Takaki, Naoyuki [Course of Applied Science, School of Engineering, Tokai University (Japan)

    2010-06-22T23:59:59.000Z

    Deployment scenario of heavy water cooled thorium breeder reactor has been studied. We have assumed to use plutonium and thorium oxide fuel in water cooled reactor to produce {sup 233}U which will be used in thorium breeder reactor. The objective is to analysis the potential of water cooled Th-Pu reactor for replacing all of current LWRs especially in Japan. In this paper, the standard Pressurize Water Reactor (PWR) has been designed to produce 3423 MWt; (i) Th-Pu PWR, (ii) Th-Pu HWR (MFR = 1.0) and (iii) Th-Pu HWR (MFR 1.2). The properties and performance of the core were investigated by using cell and core calculation code. Th-Pu PWR or HWR produces {sup 233}U to introduce thorium breeder reactor. The result showed that to replace all (60 GWe) LWR by thorium breeder reactor within a period of one century, Th-Pu oxide fueled PWR has insufficient capability to produce necessary amount of {sup 233}U and Th-Pu oxide fueled HWR has almost enough potential to produce {sup 233}U but shows positive void reactivity coefficient.

  9. Water Reactor Safety Research Division quarterly progress report, January 1-March 31, 1980

    SciTech Connect (OSTI)

    Romano, A.J. (comp.)

    1980-06-01T23:59:59.000Z

    The Water Reactor Safety Research Programs Quarterly Report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: LWR Thermal Hydraulic Development, Advanced Code Evaluation, TRAC Code Assessment, and Stress Corrosion Cracking of PWR Steam Generator Tubing.

  10. Water Reactor Safety Research Division. Quarterly progress report, April 1-June 30, 1980

    SciTech Connect (OSTI)

    Abuaf, N.; Levine, M.M.; Saha, P.; van Rooyen, D.

    1980-08-01T23:59:59.000Z

    The Water Reactor Safety Research Programs quarterly report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: LWR Thermal Hydraulic Development, Advanced Code Evlauation, TRAC Code Assessment, and Stress Corrosion Cracking of PWR Steam Generator Tubing.

  11. Water Reactor Safety Research Division. Quarterly progress report, October 1-December 31, 1980

    SciTech Connect (OSTI)

    Cerbone, R.J.; Saha, P.; van Rooyen, D.

    1981-02-01T23:59:59.000Z

    The Water Reactor Safety Research Programs Quarterly Report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: Stress Corrosion Cracking of PWR Steam Generator Tubing, Advanced Code Evaluation, Simulator Improvement Program, and LWR Assessment and Application.

  12. Wireless Sensor Networks Lecture 1: Introduction

    E-Print Network [OSTI]

    Maróti, Miklós

    Ptr, Vanderbilt #12;2009 ­© Matt Welsh Harvard University 10 Solar panels for charging car battery (used by Free Harvard University 4 Wireless Technologies Comparison Data rate Complexity/power/cost CC1000 Bluetooth 802 University 5 Wireless Technologies Comparison Type Data rate Transmit pwr Range (approx) Cost 802.11b 11

  13. Elucidating and Mapping Heat Tolerance in Wild Tetraploid Wheat (Triticum turgidum L.)

    E-Print Network [OSTI]

    Ali, Mohamed Badry Mohamed

    2012-02-14T23:59:59.000Z

    Component PWR Pedigree Wide Regression LD Linkage Disequilibrium AA Association Analysis viii TABLE OF CONTENTS Page ABSTRACT... ....................................... 9 Metabolic adaptation to heat stress ............................................... 11 Quantitative trait loci (QTL) and their importance for breeding .... 14 Simple sequence repeats (SSR) and their importance for breeding 15 Linkage map...

  14. Multi-barrier borehole canister designs for a tuff repository

    SciTech Connect (OSTI)

    James, D.E.; Skaggs, R.L.; Mohansingh, S.

    1994-05-01T23:59:59.000Z

    Initial dimensions are presented for proposed multi-barrier spent fuel borehole canisters using coated shells combined with sacrificial anodes and alkaline, oxide barriers to adjust potential and pH of the exterior shell into thermodynamically passive or immune regions of the Pourbaix diagram. Configuration of the 3 PWR canister is similar to the 1983 Site Characterization Project (SCP) borehole design. Canister dimensions were determined by using material performance data to calculate wall thickness, criticality, and sacrificial anode life. For the 3-PWR canister. Incoloy 825 is the preferred exterior canister shell material; copper-nickel alloy CDA 715 is the preferred interior canister shell material. High-lime concrete or alumina is preferred for the alkaline filler. Magnesium alloy is the preferred sacrificial anode material. Coating the canister exterior would be necessary to reduce corrosion current density to the point where a 10,000 year design life is possible. A 1 PWR canister has lower mass, thinner walls and lower criticality than the 3 PWR design. Equilibrium calculations for the historical average composition of J-13 water using the aquatic chemical speciation program WQ4F show positive saturation indices for several minerals, indicating potential for deposition on the canister exterior over long time periods. Uniform deposition could reduce corrosion rate by hindering transport of corrosion products from the canister surface. If deposition is non-uniform, local corrosion could increase through development of differential oxygen concentration cells.

  15. REACTOR OPERATIONS AND CONTROL

    E-Print Network [OSTI]

    Pázsit, Imre

    REACTOR OPERATIONS AND CONTROL KEYWORDS: core calculations, neural networks, control rod elevation of a control rod, or a group of control rods, is an important parameter from the viewpoint of reactor control DETERMINATION OF PWR CONTROL ROD POSITION BY CORE PHYSICS AND NEURAL NETWORK METHODS NINOS S. GARIS* and IMRE

  16. Journal of ASTM International, May 2005, Vol. 2, No. 5 Paper ID JAI12375

    E-Print Network [OSTI]

    Motta, Arthur T.

    on three Zr-based alloys with varied corrosion behavior were studied with micro-beam synchrotron radiation alloys, corrosion, synchrotron radiation, X-ray diffraction, oxide microstructure Introduction The extended fuel burnup, longer fuel cycles, power uprates, higher temperatures, and increased lithium in PWR

  17. Research Collaboration with local Centers of

    E-Print Network [OSTI]

    Tennessee, University of

    Faculty 2014 Enrollment 2013 Graduates Brief History For more information, see our Annual Report at www.engr.utk.edu/nuclear, and Control Laboratory · PWR Simulator (hardware and software) · Radiochemistry and Nuclear Forensics in the past 56 years · 85 MW High Flux Isotope Reactor (HFIR) · Spallation Neutron Source (SNS) · Nuclear

  18. To link to this article: DOI:10.1016/j.corsci.2011.12.007 URL: http://dx.doi.org/10.1016/j.corsci.2011.12.007

    E-Print Network [OSTI]

    Mailhes, Corinne

    ] reported, for 316 stainless steel un- der PWR nominal solution, a duplex structure layer: a compact in- ner- Christine and Andrieu, Eric Effect of surface preparation on the corrosion of austenitic stainless steel 304-oatao@listes.diff.inp-toulouse.fr #12;Effect of surface preparation on the corrosion of austenitic stainless steel 304L in high

  19. Department of Energy's team's analyses of Soviet designed VVERs (water-cooled water-moderated atomic energy reactors)

    SciTech Connect (OSTI)

    Not Available

    1989-09-01T23:59:59.000Z

    This document contains apprendices A through P of this report. Topics discussed are: a cronyms and technical terms, accident analyses reactivity control; Soviet safety regulations; radionuclide inventory; decay heat; operations and maintenance; steam supply system; concrete and concrete structures; seismicity; site information; neutronic parameters; loss of electric power; diesel generator reliability; Soviet codes and standards; and comparisons of PWR and VVER features. (FI)

  20. Genetic algorithm-based wrapper approach for grouping condition monitoring signals of

    E-Print Network [OSTI]

    Boyer, Edmond

    ) of a Pressurized Water Reactor (PWR). The optimization results are evaluated with respect to the accuracy monitoring of nuclear power plants requires to optimally group the usually very large number of signals- proach to optimally group the signals. We use a Genetic Algorithm (GA) for the optimization of the groups

  1. Proceedings: 2000 EPRI Workshop on PWSCC of Alloy 600 in PWRs (PWRMRP-27)

    SciTech Connect (OSTI)

    None

    2000-11-01T23:59:59.000Z

    Primary water stress corrosion cracking (PWSCC) can lead to increased costs for operation, maintenance, assessment, repair, and replacement of PWR components. This EPRI workshop emphasized issues related to PWSCC of reactor vessel (RV) closure head nozzles, including control rod drive mechanism (CRDM) nozzles, and other primary system alloy 600 penetrations.

  2. Proceedings: EPRI MRP Alloy 600 Industry Workshop: June 12-14, 2001, Atlanta

    SciTech Connect (OSTI)

    None

    2001-09-01T23:59:59.000Z

    Recent instances of significant cracking in Alloy 82/182 welds have occurred in commercial pressurized water reactor (PWR) nuclear power plants, including (1) cracking found in the hot leg nozzle weld at the V. C. Summer Nuclear Station and at a foreign plant, and (2) circumferential cracking found in the control rod drive mechanism (CRDM) nozzles at Oconee Nuclear Station.

  3. Proceedings: 2002 Workshop on Condensate Polishing

    SciTech Connect (OSTI)

    None

    2002-06-01T23:59:59.000Z

    Condensate polishing aims to control impurities in a nuclear power plant, thus allowing the unit to operate more reliably. This report contains the work presented at EPRI's 2002 Workshop on Condensate Polishing, where 36 papers were presented on current issues, research, and utility experiences involving polishing issues at both pressurized water reactor (PWR) and boiling water reactor (BWR) units.

  4. Thermal Shock Effects Modeling on a Globe Valve Body-Bonnet Bolted Flange Joint

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    and are sources of issues to suppliers and users. This is also the case when it comes to valve reliability. In French nuclear industry, EDF used to work closely with valve suppliers to ensure the reliability. Thus of qualification was created in the beginning of the French nuclear Pressurized Water Reactors (PWR) industry

  5. Improvement of LWR thermal margins by introducing thorium Cheuk Wah Lau a,*, Christophe Demazire a

    E-Print Network [OSTI]

    Demazière, Christophe

    Improvement of LWR thermal margins by introducing thorium Cheuk Wah Lau a,*, Christophe Demazière Keywords: Thorium PWR Thermal margins Transport calculations a b s t r a c t The use of thorium pins contain a mixture of uranium and thorium oxides, while a few fuel pins contain a mixture between

  6. Progress in Nuclear Energy 61 (2012) 48 56 Contents lists available at SciVerse ScienceDirect

    E-Print Network [OSTI]

    Demazière, Christophe

    : Thorium PWR Thermal margins Transport calculations The use of thorium in pressurized water reactor fuel is a traditional 17 x 17 pressurized water reactor fuel design. The majority of the fuel pins contain a mixture. The calculation were performed by two dimensional transport calculations with the Studsvik Scandpower CASMO 4E

  7. Senior Design Projects 2013 Project Title 1 : Monte Carlo Simulations Using a Benchmark Full-Core Pressured Water Rector Model

    E-Print Network [OSTI]

    Danon, Yaron

    that it would be 2019 before such a full reactor core calculation could be accomplished in 1 hour. In this case a full-core PWR reactor model for parallel MCNP calculations on the CCNI system 4. Code optimization in inhomogeneous, 3D media such as a nuclear reactor assembly. However, the large computation time that is required

  8. Ris Report No. 341 Ris National Laboratory

    E-Print Network [OSTI]

    OPTIMIZATION POWER DISTRIBUTION PWR TYPE REACTORS THREE-DIMENSIONAL CALCULATIONS TWO-DIMENSIONAL CALCULATIONS Synthetization of the Optimization Problem 25 4.1. Method of Calculation 25 4.2. The Reactor Model 25 4 9. Three-Dimensional Calculations 49 9.1. Control Strategy for a Three-Dimensional Reactor 4 8 9

  9. Non-Intrusive Experiemental Investigation of Multi-Scale Flow Behavior in Rod Bundle with Spacer-Grids

    E-Print Network [OSTI]

    Dominguez Ontiveros, Elvis Efren

    2011-08-08T23:59:59.000Z

    5 x 5 rod bundle with spacer-grids. Measurements were performed using two different grid designs. One typical of Boiling Water Reactors (BWR) with swirl type mixing vanes and the other typical of Pressurized Water Reactors (PWR) with split type...

  10. Fourier Transform-Plasmon Waveguide Spectroscopy: A Nondestructive Multifrequency Method for Simultaneously Determining Polymer Thickness and Apparent Index of Refraction

    SciTech Connect (OSTI)

    Bobbitt, Jonathan M [Ames Laboratory; Weibel, Stephen C [GWC Technologies Inc; Elshobaki, Moneim [Iowa State University; Chaudhary, Sumit [Iowa State University; Smith, Emily A.

    2014-12-16T23:59:59.000Z

    Fourier transform (FT)-plasmon waveguide resonance (PWR) spectroscopy measures light reflectivity at a waveguide interface as the incident frequency and angle are scanned. Under conditions of total internal reflection, the reflected light intensity is attenuated when the incident frequency and angle satisfy conditions for exciting surface plasmon modes in the metal as well as guided modes within the waveguide. Expanding upon the concept of two-frequency surface plasmon resonance developed by Peterlinz and Georgiadis [ Opt. Commun. 1996, 130, 260], the apparent index of refraction and the thickness of a waveguide can be measured precisely and simultaneously by FT-PWR with an average percent relative error of 0.4%. Measuring reflectivity for a range of frequencies extends the analysis to a wide variety of sample compositions and thicknesses since frequencies with the maximum attenuation can be selected to optimize the analysis. Additionally, the ability to measure reflectivity curves with both p- and s-polarized light provides anisotropic indices of refraction. FT-PWR is demonstrated using polystyrene waveguides of varying thickness, and the validity of FT-PWR measurements are verified by comparing the results to data from profilometry and atomic force microscopy (AFM).

  11. Nuclear Regulatory Commission issuances, January 1995. Volume 41, Number 1

    SciTech Connect (OSTI)

    NONE

    1995-01-01T23:59:59.000Z

    This book contains issuances of the Atomic Safety and Licensing Boards for January 1995. The issuances include Babcock and Wilcox Company materials license; Hydro Resources, Inc. application for uranium mining; low-level waste storage in Utah; communication of emerging and existing generic, technical issues with PWR owners groups; and radioactive waste management by Sierra Nuclear Corporation.

  12. Threshold sensor for high-doses of radiation I. Augustyniak, P. Knapkiewicz, J. Dziuban

    E-Print Network [OSTI]

    Boyer, Edmond

    of spent nuclear fuel, nuclear waste disposal site as well as after nuclear accidents. High radiation doses.augustyniak@pwr.wroc.pl M. Olszacki National Centre for Nuclear Research, Otwock, Poland michal.olszacki@ncbj.gov.pl A membrane I. INTRODUCTION High doses of radiation (>10 kGy) can be found in nuclear power plants, storage

  13. Recent results on the RIA test in IGR reactor

    SciTech Connect (OSTI)

    Asmolov, V.; Yegorova, L. [Nuclear Safety Institute, Moscow (Russian Federation)

    1997-01-01T23:59:59.000Z

    At the 23d WRSM meeting the data base characterizing results of VVER high burnup fuel rods tests under reactivity-initiated accident (RIA) conditions was presented. Comparison of PWR and VVER failure thresholds was given also. Additional analysis of the obtained results was being carried out during 1996. The results of analysis show that the two different failure mechanisms were observed for PWR and VVER fuel rods. Some factors which can be as the possible reasons of these differences are presented. First of them is the state of preirradiated cladding. Published test data for PWR high burnup fuel rods demonstrated that the PWR high burnup fuel rods failed at the RIA test are characterized by very high level of oxidation and hydriding for the claddings. Corresponding researches were performed at Institute of Atomic Reactors (RLAR, Dimitrovgrad, Russia) for large set of VVER high burnup fuel rods. Results of these investigations show that preirradiated commercial Zr-1%Nb claddings practically keep their initial levels of oxidation and H{sub 2} concentration. Consequently the VVER preirradiated cladding must keep the high level of mechanical properties. The second reason leading to differences between failure mechanisms for two types of high burnup fuel rods can be the test conditions. Now such kind of analysis have been performed by two methods.

  14. Co-combustion feasibility study. Final report

    SciTech Connect (OSTI)

    Handcock, D.J. [Clough, Harbour and Associates, Albany, NY (United States)

    1995-01-01T23:59:59.000Z

    This report investigates the technical and economic feasibility of co-combusting municipal sewage sludge produced by the Saratoga County Sewer District No. 1 with paper mill sludge produced by the Cottrell Paper Company, Encore Paper Company, International Paper Company, Mohawk Paper Mills, and TAGSONS Papers at the Saratoga County Sewer District No. 1`s secondary wastewater treatment plant and recovering any available energy products. The co-combustion facility would consist of sludge and wood chip storage and conveying systems, belt filter presses, screw presses, fluidized-bed incinerators, venturi scrubbers and tray cooling systems, ash dewatering facilities, heat recovery steam generators, gas-fired steam superheaters, and a back-pressure steam turbine system. Clean waste wood chips would be used as an auxiliary fuel in the fluidized-bed incinerators. It is recommended that the ash produced by the proposed facility be beneficially used, potentially as a raw material in the manufacture of cement and/or as an interim barrier layer in landfills.

  15. 94-A13 Native American Initiative Short Course Management Plan

    SciTech Connect (OSTI)

    Carroll, Herbert B.; Johnson, William I.; Kokesh, Judith H.

    1999-04-27T23:59:59.000Z

    A training program conducted in Bartlesville by BDM-Oklahoma technical staff, which included geologists, geophysicists, exploration and drilling specialists, and environmental policy experts. The proposed training schedule offered four courses per year and included those coursed identified by the tribes in the survey. The training program was outlined for members of Native American Tribes whose lands have oil and gas resources. The proposed program contributed to meeting the goals of the U.S. Department of Energy's (DOE) Domestic Oil and Gas Initiative to help Native American tribes become more self-sufficient in developing and managing their resources through training in cost-effective, improved technologies for hydrocarbon production that will meet environmental regulations. The training program outlined was for adult tribal representatives who are responsible for managing tribal mineral holdings or setting policy, or who work in the oil and gas industry. The course content is in response to a survey that was developed by BDM-Oklahoma and sent in the Spring of 1995 to 26 tribal agencies identified through previous contact with DOE. Tribes were asked to indicate course content needs, levels, preferred time of year, and location. Six tribes responded with specific recommendations and needs. These tribes, were the Creek, Pueblo, Cherokee, St. Regis Mohawk, Northern Arapho, and Ute Mountain Ute.

  16. Consumption of freshwater fish in Kahnawake: Risks and benefits

    SciTech Connect (OSTI)

    Chan, H.M.; Trifonopoulos, M.; Ing, A.; Receveur, O. [McGill Univ., Ste-Anne-de-Bellevue, Quebec (Canada)] [McGill Univ., Ste-Anne-de-Bellevue, Quebec (Canada); Johnson, E. [Kahnawake Environment, Quebec (Canada)] [Kahnawake Environment, Quebec (Canada)

    1999-02-01T23:59:59.000Z

    Kahnawake is a Mohawk community located on the south shore of the Saint Lawrence River near Montreal. A comprehensive study was conducted in 1996--1997 to address the local concern regarding health risks of contaminant exposure associated with freshwater fish consumption. Forty-two participants, including most of the identified active fishermen were interviewed. Walleye, perch, bullhead, and smallmouth bass were the species most consumed. Average daily intake of locally caught fish was 23 g/day. Nutrient and contaminant levels of locally collected fish were analyzed. Fish were good sources of protein, polyunsaturated fatty acids, calcium, zinc, and iron. Levels of cadmium, lead, arsenic, polychlorinated biphenyls (PCBs), and other chlorinated pesticides were at least 10 times lower than the guideline levels. Mercury levels of some predatory fish exceeded the guideline of 0.5 {micro}g/g. Average daily intakes of all contaminants were below the guideline levels by a factor of 10 except for mercury. Average mercury intake rate was about one-third that of the guideline level. Contrary to residents` perception, Kahnawake fish were not particularly contaminated. In view of the nutritional as well as cultural benefits, fishing and fish consumption may be promoted.

  17. DEVELOPMENT OF ALTERNATE METHODS OF DETERMINING INTEGRATED SMR SOURCE TERMS

    SciTech Connect (OSTI)

    Barry, Kenneth

    2014-06-10T23:59:59.000Z

    The Nuclear Energy Institute (NEI) Small Modular Reactor (SMR) Licensing Task Force (TF) has been evaluating licensing issues unique and important to iPWRs, ranking these issues, and developing NEI position papers for submittal to the U.S. Nuclear Regulatory Commission (NRC) during the past three years. Papers have been developed and submitted to the NRC in a range of areas including: Price- Anderson Act, NRC annual fees, security, modularity, and staffing. In December, 2012, NEI completed a draft position paper on SMR source terms and participated in an NRC public meeting presenting a summary of this paper, which was subsequently submitted to the NRC. One important conclusion of the source term paper was the evaluation and selection of high importance areas where additional research would have a significant impact on source terms. The highest ranked research area was iPWR containment aerosol natural deposition. The NRC accepts the use of existing aerosol deposition correlations in Regulatory Guide 1.183, but these were developed for large light water reactor (LWR) containments. Application of these correlations to an iPWR design has resulted in greater than a ten-fold reduction of containment airborne aerosol inventory as compared to large LWRs. Development and experimental justification of containment aerosol natural deposition correlations specifically for the unique iPWR containments is expected to result in a large reduction of design basis and beyond-design-basis accident source terms with concomitantly smaller dose to workers and the public. Therefore, NRC acceptance of iPWR containment aerosol natural deposition correlations will directly support the industry’s goal of reducing the Emergency Planning Zone (EPZ) for SMRs. Based on the results in this work, it is clear that thermophoresis is relatively unimportant for iPWRs. Gravitational settling is well understood, and may be the dominant process for a dry environment. Diffusiophoresis and enhanced settling by particle growth are the dominant processes for determining DFs for expected conditions in an iPWR containment. These processes are dependent on the areato- volume (A/V) ratio, which should benefit iPWR designs because these reactors have higher A/Vs compared to existing LWRs.

  18. Phase I (Year 1) Summary of Research--Establishing the Relationship between Fracture-Related Dolomite and Primary Rock Fabric on the Distribution of Reservoirs in the Michigan Basin

    SciTech Connect (OSTI)

    G. Michael Grammer

    2005-11-09T23:59:59.000Z

    This topical report covers the first 12 months of the subject 3-year grant, evaluating the relationship between fracture-related dolomite and dolomite constrained by primary rock fabric in the 3 most prolific reservoir intervals in the Michigan Basin (Ordovician Trenton-Black River Formations; Silurian Niagara Group; and the Devonian Dundee Formation). Phase I tasks, including Developing a Reservoir Catalog for selected dolomite reservoirs in the Michigan Basin, Characterization of Dolomite Reservoirs in Representative Fields and Technology Transfer have all been initiated and progress is consistent with our original scheduling. The development of a reservoir catalog for the 3 subject formations in the Michigan Basin has been a primary focus of our efforts during Phase I. As part of this effort, we currently have scanned some 13,000 wireline logs, and compiled in excess of 940 key references and 275 reprints that cover reservoir aspects of the 3 intervals in the Michigan Basin. A summary evaluation of the data in these publications is currently ongoing, with the Silurian Niagara Group being handled as a first priority. In addition, full production and reservoir parameter data bases obtained from available data sources have been developed for the 3 intervals in Excel and Microsoft Access data bases. We currently have an excess of 25 million cells of data for wells in the Basin. All Task 2 objectives are on time and on target for Phase I per our original proposal. Our mapping efforts to date, which have focused in large part on the Devonian Dundee Formation, have important implications for both new exploration plays and improved enhanced recovery methods in the Dundee ''play'' in Michigan--i.e. the interpreted fracture-related dolomitization control on the distribution of hydrocarbon reservoirs. In an exploration context, high-resolution structure mapping using quality-controlled well data should provide leads to convergence zones of fault/fracture trends that are not necessarily related to structural elevation. Further work in Phase II will be focused on delineating the relative contribution to fracture-only dolomitization to that which occurs in conjunction with primary facies and/or sequence stratigraphic framework.

  19. Neurocontrol of Pressurized Water Reactors in Load-Follow Operations

    SciTech Connect (OSTI)

    Lin Chaung; Shen Chihming

    2000-12-15T23:59:59.000Z

    The neurocontrol technique was applied to control a pressurized water reactor (PWR) in load-follow operations. Generalized learning or direct inverse control architecture was adopted in which the neural network was trained off-line to learn the inverse model of the PWR. Two neural network controllers were designed: One provided control rod position, which controlled the axial power distribution, and the other provided the change in boron concentration, which adjusted core total power. An additional feedback controller was designed so that power tracking capability was improved. The time duration between control actions was 15 min; thus, the xenon effect is limited and can be neglected. Therefore, the xenon concentration was not considered as a controller input variable, which simplified controller design. Center target strategy and minimum boron strategy were used to operate the reactor, and the simulation results demonstrated the effectiveness and performance of the proposed controller.

  20. Assessment of RELAP5/MOD2 code using loss of offsite power transient data of KNU (Korea Nuclear Unit) No. 1 Plant

    SciTech Connect (OSTI)

    Chung, Bud-Dong; Kim, Hho-Jung (Korea Advanced Energy Research Inst., Daeduk-Danji (Republic of Korea). Korea Nuclear Safety Center); Lee, Young-Jin (Seoul National Univ. (Republic of Korea))

    1990-04-01T23:59:59.000Z

    This report presents a code assessment study based on a real plant transient that occurred on June 9, 1981 at the KNU {number sign}1 (Korea Nuclear Unit Number 1). KNU {number sign}1 is a two-loop Westinghouse PWR plant of 587 Mwe. The loss of offsite power transient occurred at the 77.5% reactor power with 0.5%/hr power ramp. The real plant data were collected from available on-line plant records and computer diagnostics. The transient was simulated by RELAP5/MOD2/36.05 and the results were compared with the plant data to assess the code weaknesses and strengths. Some nodalization studies were performed to contribute to developing a guideline for PWR nodalization for the transient analysis. 5 refs., 18 figs., 3 tabs.