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Note: This page contains sample records for the topic "niagara mohawk pwr" from the National Library of EnergyBeta (NLEBeta).
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1

PP-190 Niagara Mohawk Power Corporation | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Niagara Mohawk Power Corporation More Documents & Publications PP-230-2 International Transmission Company EIS-0183: Notice of Availability of the Revised Record of Decision...

2

STATEMENT OF CONSIDERATIONS REQUEST BY NIAGARA MOHAWK POWER CORPORATION FOR AN  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

NIAGARA MOHAWK POWER CORPORATION FOR AN NIAGARA MOHAWK POWER CORPORATION FOR AN ADVANCE WAIVER OF DOMESTIC AND FOREIGN PATENT RIGHTS UNDER. CONTRACT NO. DE-FC36-96GO10132, W(A)-96-018, CH-0910 The Niagara Mohawk Power Corporation (hereafter NMPC), a large business, has petitioned for an advance waiver of patent rights under DOE Contract No. DE-FC36- 96GO10132. NMPC is the lead company of a group of organizations call the Salix Consortium which expects to demonstrate that a process called the "Swedish Willow Biomass System" is suitable for electrical power production in the Northeastern United States. NMPC has requested a waiver of domestic and foreign patent rights for all subject inventions of its employees, including those of its wholly owned or controlled subsidiaries or affiliates, as well as those of its subcontractors, at any tier, other than those of domestic small

3

Does RTP Deliver Demand Response?: Case Studies of Niagara Mohawk RTP and  

E-Print Network (OSTI)

/ educational 40% 46% Average monthly maximum demand 3.0 MW 3.4 MW Option 2 9% 18% The survey response rateDoes RTP Deliver Demand Response?: Case Studies of Niagara Mohawk RTP and ~43 Voluntary Utility RTP Programs Charles Goldman Lawrence Berkeley National Laboratory Mid-Atlantic Demand Response Initiative

4

Cycling Operation of Fossil Plants: Volume 1: Cycling Considerations for Niagara Mohawk's Oswego Unit 5  

Science Conference Proceedings (OSTI)

Fossil plants are being converted to cycling operation to accommodate daily load swings and to decrease the overall system fuel costs. This report summarizes the methods and results of an engineering study of three two-shift cycling approaches considered for Niagara Mohawk's Oswego unit 5: superheater/turbine bypass, variable pressure operations, and full-flow condensate polishing.

1991-05-01T23:59:59.000Z

5

Results of the radiological survey at the Niagara-Mohawk property, Railroad Avenue, Colonie, New York (AL218)  

SciTech Connect

A number of properties in the Albany/Colonie area have been identified as being potentially contaminated with uranium originating from the former National Lead Company's uranium forming plant in Colonie, New York. The Niagara-Mohawk property on Railroad Avenue in Colonie, New York, was the subject of a radiological investigation initiated June 11, 1987. This commercial property is an irregularly shaped lot partially occupied by an electric power substation and associated transmission lines. Portions of the property that were swampy and heavily vegetated were inaccessible to the survey team. There are no buildings on the property. A diagram showing the approximate boundaries and the 15-m grid network established for measurements on the property is shown. The lot included in the radiological survey was /approximately/45 m wide by 246 m deep. Two views of the property are shown. 13 refs., 6 figs., 4 tabs.

Marley, J.L.; Carrier, R.F.

1987-12-01T23:59:59.000Z

6

Niagara Limestone  

NLE Websites -- All DOE Office Websites (Extended Search)

Niagara Limestone Niagara Limestone Nature Bulletin No. 282-A November 11, 1967 Forest Preserve District of Cook County Richard B. Ogilvie, President Roland F. Eisenbeis, Supt. of Conservation NIAGARA LIMESTONE Chicago stands at the crossroads of America -- the heart of the Middle West -- and one of the most important natural resources upon which it depends is the Niagara limestone beneath it. The bedrock in this region consists of layer upon layer of limestones, shales and sandstones stacked almost a half mile thick on top of the ancient granite, once molten, that formed the original surface of the earth before oceans formed and life appeared. The Niagara limestone is the uppermost layer here but few of us are aware of it because it is covered with soil and ground up rock -- glacial drift -- ranging from a few feet to a hundred or more feet in depth.

7

niagaraVP  

Office of Legacy Management (LM)

Niagara Falls Vicinity Properties, New York. Niagara Falls Vicinity Properties, New York. This site is managed by the U.S. Department of Energy Office of Legacy Management. Site Description and History The Niagara Falls Vicinity Properties, New York (formerly the Niagara Falls Storage Site) is located in Lewiston, New York, approximately 10 miles north of the city of Niagara Falls, New York. The site is a remnant of the U.S. Army's 7,500-acre Lake Ontario

8

STATEMENT OF CONSIDERATIONS REQUEST BY NIAGARA MOHAWK POWER CORPORATIO...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Willow Biomass System to be demonstrated under this agreement is a willow energy crop system utilizing a biomass fuel source (i.e., willow trees) which can be used in...

9

The Fossils of Niagara Limestone  

NLE Websites -- All DOE Office Websites (Extended Search)

sights -- a mile wide and 300 feet deep. Niagara Limestone shows no trace of fishes or other backboned animals because these had not yet appeared on the earth at the time...

10

NIAGARA FALLS STORAGE SITE  

Office of Legacy Management (LM)

:i" :i" _,, ' _~" ORISE 95/C-70 :E : i:; :' l,J : i.: RADIOLOGICAL SURVEY Op BUILDINGS 401, ' 403, AND ' m HITTMAN BUILDING $ <,' 2:. NIAGARA FALLS STORAGE SITE I .~~ ; " LEWISTON, ' NEW YORK : f? j:,:i I ,.J- ;b f" /: Li _e.*. ~,, I ,,~, ,:,,;:, Prepared by T. .I. Vitkus i,c Environmental Survey and Site Assessment Program Energy/Environment Systems Division ;>::; Oak Ridge Institute for Science and Education .,:, "Oak Ridge, Temressee 37831-0117 .F P ., ? :_ &,d ,,,, ;<:x,, Prepared for the 3 I. Office of Environmental Restoration I, U.S. Department of Energy i gy i. ~: ,,, "! ? ' :' : "' ,//, FINAL REPORT ".$ :,a ,,, MARCH 1995 ; m L ,, ,, ,,,. ., ,,. ' 1 jq ,Ij:,., .,~ _,I_ 1 This report is based on work performed under contract number DE-AC05-760R00033 with the

11

The Effects of Electricity Tariff Structure on Distributed Generation Adoption in New York State  

E-Print Network (OSTI)

FLT TOU RTP Consolidated Edison Niagara Mohawk Orange andRTP FLT TOU Consolidated Edison Niagara Mohawk FLT TOU RTPFLT TOU RTP Consolidated Edison Niagara Mohawk Orange and

Firestone, Ryan; Marnay, Chris

2005-01-01T23:59:59.000Z

12

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

E-Print Network (OSTI)

customers are to their maximum demand. We proceed with anSubstitution by Account and Maximum Demand N=119 Number offrom 60% to 70% of maximum demand) results in much smaller

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

2006-01-01T23:59:59.000Z

13

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

E-Print Network (OSTI)

day. This view of daily electricity demand is supported bypeak Summer THE ELECTRICITY DEMAND MODEL For the empiricalwe adopt a model for electricity demand consistent with that

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

2006-01-01T23:59:59.000Z

14

Customer Response to RTP in Competitive Markets: A Study ofNiagara Mohawk's Standard Offer Tariff  

Science Conference Proceedings (OSTI)

Utilizing load, price, and survey data for 119 largecustomers that paid competitively determined hourly electricity pricesannounced the previous day between 2000 and 2004, this study providesinsight into the factors that determine the intensity of price response.Peak and off-peak electricity can be: perfect complements, substitutes,or substitutes where high peak prices cause temporary disconnection fromthe grid, as for some firms with on-site generation. The averageelasticity of substitution is 0.11. Thirty percent of the customers usepeak and off-peak electricity in fixed proportions. The 18 percent withelasticities greater than 0.10 provide 75 percent of the aggregate priceresponse. In contrast to Industrial customers, Commercial/Retail andGovernment/Education customers are more price responsive on hot days andwhen the ratio of peak to off-peak prices is high. Price responsivenessis not substantially reduced when customers operate near peak usage.Diversity of customer circumstances and price response suggest dynamicpricing is suited for some, but not all customers.

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

2006-06-01T23:59:59.000Z

15

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

E-Print Network (OSTI)

2002). “Industrial Response to Electricity Real-Time Prices:Industrial Response To Real Time Electricity Prices. ”price elasticity of large commercial and industrial customers served under a day-ahead RTP rate to identify patterns based on whether electricity

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

2006-01-01T23:59:59.000Z

16

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

E-Print Network (OSTI)

two MW. NMPC indexes RTP energy usage rates to the NYISO’sof peak and off-peak energy usage, and E cannot be observedsubstantial proportion of energy usage—Government/Education

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

2006-01-01T23:59:59.000Z

17

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

E-Print Network (OSTI)

Response to Electricity Real-Time Prices: Short Run and LongResponse To Real Time Electricity Prices. ” December.$500/MWh or the real-time market price for load curtailments

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

2006-01-01T23:59:59.000Z

18

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

Gasoline and Diesel Fuel Update (EIA)

View History: Monthly Annual Download Data (XLS File) Niagara Falls, NY Natural Gas Pipeline Exports to Canada (Million Cubic Feet) Niagara Falls, NY Natural Gas Pipeline Exports...

19

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

Gasoline and Diesel Fuel Update (EIA)

View History: Monthly Annual Download Data (XLS File) Niagara Falls, NY Natural Gas Pipeline Imports From Canada (Million Cubic Feet) Niagara Falls, NY Natural Gas Pipeline...

20

Mohawk Municipal Comm | Open Energy Information  

Open Energy Info (EERE)

Municipal Comm Municipal Comm Jump to: navigation, search Name Mohawk Municipal Comm Place New York Utility Id 12759 Utility Location Yes Ownership M NERC Location NPCC Activity Distribution Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png Industrial Rate Industrial Large Commercial Commercial Public Street Lighting Lighting Security Lighting 150 w lamp Lighting Security Lighting 175 w lamp Lighting Security Lighting 250 w lamp Lighting Security Lighting 400 w lamp Lighting Single-Phase Residential Residential Small Commercial Business Commercial Average Rates Residential: $0.0366/kWh

Note: This page contains sample records for the topic "niagara mohawk pwr" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

Wellton-Mohawk Irr & Drain Dist | Open Energy Information  

Open Energy Info (EERE)

Wellton-Mohawk Irr & Drain Dist Wellton-Mohawk Irr & Drain Dist Jump to: navigation, search Name Wellton-Mohawk Irr & Drain Dist Place Arizona Utility Id 25060 Utility Location Yes Ownership P NERC Location WECC NERC WECC Yes Activity Distribution Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png Agricultural Pumping Service Commercial General Service Commercial Large General Service Commercial Residential Service Residential Average Rates Residential: $0.1000/kWh Commercial: $0.0842/kWh Industrial: $0.0956/kWh References ↑ "EIA Form EIA-861 Final Data File for 2010 - File1_a"

22

DOE - Office of Legacy Management -- Niagara VP_FUSRAP  

Office of Legacy Management (LM)

Niagara Falls Vicinity Properties, New York, Site Niagara Falls Vicinity Properties, New York, Site FUSRAP Site Niagara Falls Vicinity Properties Map Background-The Niagara Falls Vicinity Properties Site was remediated under the Formerly Utilized Sites Remedial Action Program (FUSRAP). FUSRAP was established in 1974 to remediate sites where radioactive contamination remained from Manhattan Project and early U.S. Atomic Energy Commission operations. History-Niagara Falls Storage Site Vicinity Properties, located near Lewiston, New York, consists of 26 properties sold to private owners; the properties were formerly part of the Lake Ontario Ordnance Works. Another portion of the former ordnance works was transferred to the U.S. Atomic Energy Commission and became the Niagara Falls Storage Site. Beginning in 1944, the Manhattan Engineer District stored uranium

23

DOE - Office of Legacy Management -- Niagara VP_FUSRAP  

NLE Websites -- All DOE Office Websites (Extended Search)

was established in 1974 to remediate sites where radioactive contamination remained from Manhattan Project and early U.S. Atomic Energy Commission operations. History-Niagara...

24

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

U.S. Energy Information Administration (EIA)

Niagara Falls, NY Natural Gas Pipeline Imports From Canada (Million Cubic Feet) Year Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec; 2011: 9,497: 6,894: 4,421: 2,459 ...

25

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

Office of Legacy Management (LM)

Niagara Falls Vicinity Properties Niagara Falls Vicinity Properties NY - NY 17 FUSRAP Considered Sites Niagara Falls Vicinity Properties, NY Alternate Name(s): Lake Ontario Ordnance Works (LOOW) Niagara Falls Storage Site (NFSS) DOE-Niagara Falls Storage Site NY.17-1 NY.17-3 Location: Lewiston , New York NY.17-5 Historical Operations: Stored, shipped, and buried radioactive equipment and waste for MED and AEC containing uranium, radium, and thorium. Portions of the former site are privately owned, creating a "site" for the vicinity properties. NY.17-1 NY.17-2 NY.17-14 Eligibility Determination: Eligible NY.17-4 Radiological Survey(s): Assessment Surveys, Verification Surveys NY.17-3 NY.17-5 NY.17-6 NY.17-7 NY.17-8 NY.17-9 NY.17-10 NY.17-11 NY.17-12 NY.17-14 Site Status: Certification Basis, including Federal Register Notice for 23 properties. Cleanup in progress for additional 3 VPs. NY.17-13

26

Review of Integrated Resource Bidding at Niagara Mohawk C.A. Goldman, J.F. Busch, E.P. Kahn, S.S. Stoft, and S. Cohen  

E-Print Network (OSTI)

Secretary for Conservation and Renewable Energy, Office of Utility Technologies, Office ofEnergy Management

27

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

Open Energy Info (EERE)

Niagara Falls, NY) Niagara Falls, NY) Jump to: navigation, search Equivalent URI DBpedia Coordinates 43.0944999°, -79.0567111° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":43.0944999,"lon":-79.0567111,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

28

MHK Projects/Niagara Community 2 | Open Energy Information  

Open Energy Info (EERE)

Niagara Community 2 Niagara Community 2 < MHK Projects Jump to: navigation, search << Return to the MHK database homepage Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":5,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"500px","height":"350px","centre":false,"title":"","label":"","icon":"File:Aquamarine-marker.png","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":43.081,"lon":-79.0111,"alt":0,"address":"","icon":"http:\/\/prod-http-80-800498448.us-east-1.elb.amazonaws.com\/w\/images\/7\/74\/Aquamarine-marker.png","group":"","inlineLabel":"","visitedicon":""}]}

29

Mr. Frank Archer President Niagara Cold Drawn Steel Corporation  

Office of Legacy Management (LM)

Department of Energy Department of Energy Washington, DC 20585 FEB 2 1 1991 ' i-. 1,; ' -, f ' + \ 1 : , .J p- * c - Mr. Frank Archer President Niagara Cold Drawn Steel Corporation 110 Hopkins Street P.O. Box 399 Buffalo, NY 14240 Dear Mr. Archer: I have executed the consent forms for the performance of a radiological survey of the Niagara Cold Drawn Steel Corporation's property under the Formerly Utilized Sites Remedial Action Program (FUSRAP) of the U.S. Department of Energy (DOE). I enclose a copy of the consent for your company's records. Oak Ridge Associated Universities (ORAU) will perform the survey of your company's property. MS. Michelle Landis (615-576-2908) is the project manager for ORAU. We have worked with Mr. Shells of your staff on this project and

30

Customer reponse to day-ahead wholesale market electricity prices: Case study of RTP program experience in New York  

E-Print Network (OSTI)

of Niagara Mohawk’s Large Customers (Peak Demand >large customers with peak demand in excess of two megawattsmonthly peak and off- peak demand blocks (at 100% load

2004-01-01T23:59:59.000Z

31

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

Open Energy Info (EERE)

Niagara Biomass Facility Niagara Biomass Facility Jump to: navigation, search Name American Ref-Fuel of Niagara Biomass Facility Facility American Ref-Fuel of Niagara Sector Biomass Facility Type Municipal Solid Waste Location Niagara County, New York Coordinates 43.3119496°, -78.7476208° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":43.3119496,"lon":-78.7476208,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

32

ADDENDUM TO ACTION DESCRIPTION MEMORANDUM NIAGARA FALLS STORAGE SITE  

Office of Legacy Management (LM)

ADDENDUM TO ADDENDUM TO ACTION DESCRIPTION MEMORANDUM NIAGARA FALLS STORAGE SITE PROPOSED INTERIM REMEDIAL ACTIONS FOR FY 1983-85 ACCELERATED PROGRAM (1984 VICINITY PROPERTIES CLEANUP) Prepared by Environmental Research Division Argonne National Laboratory Argonne, Illinois July 1984 Prepared for U.S. Department of Energy Oak Ridge Operations Technical Services Division Oak Ridge, Tennessee CONTENTS Page SUMMARY OF PROPOSED ACTION AND RELATED ACTIVITIES ........... 1 HISTORY AND ENVIRONMENTAL SETTING ........................ 4 RADIOLOGICAL CONTAMINATION AND NEED FOR PROPOSED ACTION ........ 4 Property A .. . . . . . . . . . . . . . . . .. . . . . . . 6 Property C' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 Property H ...... ............. ... 7 Property H' . . . . . . . . . . . . . . . . . . .. . . . . . . . . 7 Property L ..... ...... .

33

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

Office of Legacy Management (LM)

Niagara Falls Storage Site Vicinity Niagara Falls Storage Site Vicinity Properties in Lewiston, New York, from 7983 through 7986 Depatfment of Energy Former Sites Restoration Division Oak Ridge Field Office July 7 992 I I I I I I I I I I I I I I I I I I I CONTENTS Figures .......................... Tables .......................... Abbreviations ....................... Acronyms ......................... 1.0 Introduction ..................... 2.0 Site History ..................... 3.0 Property Descriptions ................ 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 3.10 3.11 3.12 3.13 3.14 3.15 3.16 3.17 3.18 3.19 3.20 3.21 3.22 3.23 3.24 3.25 3.26 3.27 Property A ............ Property B ............ PropertyC' ........... Property D ............ Property F ........ .' ... PropertyH' ...........

34

Option Value of Electricity Demand Response  

E-Print Network (OSTI)

Duke Power d FirstEnergy Georgia Power Niagara Mohawk Pacific Gas & Electric Pennsylvania Power & Light Progress Energy

Sezgen, Osman; Goldman, Charles; Krishnarao, P.

2005-01-01T23:59:59.000Z

35

Potential of breccia pipes in the Mohawk Canyon Area, Hualapai Indian Reservation, Arizona  

Science Conference Proceedings (OSTI)

The Hualapai Indian Reservation is on the southwestern corner of the Colorado Plateau in northern Arizona. Hundreds of solution-collapse breccia pipes crop out in the canyons and on the plateaus of northern Arizona. The pipes originated in the Mississippian Redwall Limestone and stoped their way upward through the upper Paleozoic strata, locally extending into the Triassic Moenkopi and Chinle Formations. The occurrence of high-grade U ore, associated with potentially economic concentrations of Cu, Ag, Pb, Zn, V, Co, and Ni in some of these pipes, has stimulated mining activity in northern Arizona despite the depressed market for most of these metals. Two breccia pipes, 241, and 242, have significant mineralized rock exposed on the Esplanade erosion surface; unfortunately, their economic potential is questionable because of their inaccessibility at the bottom of Mohawk Canyon. All warrant further exploration.

Wenrich, K.J.; Billingsley, G.H.; Van Gosen, B.S.

1990-09-21T23:59:59.000Z

36

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

Office of Legacy Management (LM)

Niagara Falls Storage Site NY - NY Niagara Falls Storage Site NY - NY 17 FUSRAP Considered Sites Niagara Falls Storage Site, NY Alternate Name(s): Lake Ontario Ordnance Works (LOOW) Niagara Falls Storage Site (NFSS) DOE-Niagara Falls Storage Site NY.17-1 NY.17-3 Location: Lewiston, New York NY.17-5 Historical Operations: Stored, shipped, and buried radioactive equipment and waste for MED and AEC containing uranium, radium, and thorium. Contains Interim Waste Containment Structure. NY.17-1 NY.17-2 NY.17-14 Eligibility Determination: Eligible NY.17-4 Radiological Survey(s): Assessment Surveys NY.17-3 NY.17-5 NY.17-6 NY.17-7 NY.17-8 NY.17-9 NY.17-10 NY.17-11 NY.17-12 NY.17-14 Site Status: Cleanup in progress by U.S. Army Corps of Engineers. NY.17-13 NY.17-14 NY.17-15 NY.17-16 USACE Website Long-term Care Requirements: To be determined upon completion.

37

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

Office of Legacy Management (LM)

e e - .' N"lr 7% PRELIMINARY SURVEY OF THE UNION CARBIDE CORPORATION METALS DIVISION PLANT, NIAGARA FALLS, NEW YORK Work performed by the Health and Safety Research Division Oak Ridge Natjonal Laboratory Oak Ridge, Tennessee 37830 December 1980 OAK RIDGE NATIONAL LABORATORY operated by UNION CARBIDE CORPORATION for the DEPARTMENT OF ENERGY as part of the Formerly Utilized Sites-- Remedial Action Program PRELIMINARY SURVEY OF THE UNION CARBIDE CORPORATION METALS DIVISION PLANT, NIAGARA FALLS, NEW YORK B. A. Berven and R. W. Doane Introduction On September 2;, 1980, two representatives from Oak Ridge National Laboratory visited Union Carbide Corporation's Metal Division Plant (UCC-MD) in Niagara Falls, New York. The purpose of the visit was to

38

PWR GASIFIER PEER REVIEW FINAL REPORT  

NLE Websites -- All DOE Office Websites (Extended Search)

PWR GASIFIER PEER REVIEW REPORT 22106 Background Pratt and Whitney Rocketdyne (PWR) signed a cooperative agreement with DOE on 93004 to develop a novel gasifier concept, which...

39

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

Energy.gov (U.S. Department of Energy (DOE))

Evaluation of Final Radiological Conditions at Areas of the Niagara Falls Storage Site Remediated under the Formerly Utilized Sites Remedial Action Program (March 2012)

40

PWR GASIFIER PEER REVIEW FINAL REPORT  

NLE Websites -- All DOE Office Websites (Extended Search)

PWR GASIFIER PEER REVIEW REPORT PWR GASIFIER PEER REVIEW REPORT 2/21/06 Background Pratt and Whitney Rocketdyne (PWR) signed a cooperative agreement with DOE on 9/30/04 to develop a novel gasifier concept, which is expected to improve the availability and efficiency of gasification-based power plants, and to reduce plant capital and operations costs. On 12/21/05, PWR submitted a proposal to continue development of their gasifier into the next phase. On January 24, 2006, a peer review was performed to review the work that PWR has done to date, their technical approach for future development, and to assess the potential benefit of the PWR gasifier and feed system technologies over state-of-the art coal gasification. The peer reviewers also evaluated a DOE analysis of the PWR refractory, and a DOE system study comparing the

Note: This page contains sample records for the topic "niagara mohawk pwr" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

PWR AXIAL BURNUP PROFILE ANALYSIS  

Science Conference Proceedings (OSTI)

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).

J.M. Acaglione

2003-09-17T23:59:59.000Z

42

Publications  

NLE Websites -- All DOE Office Websites (Extended Search)

Pricing Deliver Demand Response? A Case Study of Niagara Mohawk's Large Customer RTP Tariff." In 2004 ACEEE Summer Study on Energy Efficiency in Buildings. Pacific Grove, CA,...

43

A framework and review of customer outage costs: Integration and analysis of electric utility outage cost surveys  

E-Print Network (OSTI)

BPA, Southern Company, Duke Energy, Southern California1999) Niagara Mohawk (1985) Duke Energy Company (1992, 1997)Gas and Electric, and Duke Energy) the same customer classes

Lawton, Leora; Sullivan, Michael; Van Liere, Kent; Katz, Aaron; Eto, Joseph

2003-01-01T23:59:59.000Z

44

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

SciTech Connect

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.

Not Available

1990-05-01T23:59:59.000Z

45

An Evaluation of the Impact of the Niagara River Ice Boom on the Air Temperature Regime at Buffalo, New York  

Science Conference Proceedings (OSTI)

The objective of this study was to determine if the Niagara River ice boom has prolonged the Lake Erie ice cover at Buffalo, New York, resulting in significant changes in the spring warm-up of Lake Erie and longer, colder winters in the area. ...

Frank H. Quinn; Raymond A. Assel; Daniel W. Gaskill

1982-03-01T23:59:59.000Z

46

Barrel Bolt Cracking in a German PWR  

Science Conference Proceedings (OSTI)

Protective Insulated Coating for SCC Mitigation in BWRs · PWR Fuel Deposit Analysis at a B&W Plant with a 24-Month Fuel Cycle · PWSCC of Thermocoax ...

47

Niagara Prospects.  

E-Print Network (OSTI)

??This thesis proposes a fresh engagement with the idea of the archaic as a means to recover and replenish some of the lost vitality suffered… (more)

Wong, Johnathan

2009-01-01T23:59:59.000Z

48

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

Office of Legacy Management (LM)

COT"IPREITENS IVE RADIOLOGICAL COT"IPREITENS IVE RADIOLOGICAL SURVEY OFF-SITE PROPERTY P NIAGARA FALIS STORAGE SITE LEWISTON, NEW YORK Prepared for U.S. DePartment of EnergY as part of the Formerly Utilized Sites - Remedial ActLon Program J . D . B e r g e r P r o j e c t S t a f f J. Burden* w.L. Smlth* R.D. Condra T.J. Sowell J.S . Epler* G.M. S tePhens P.Iil. Frame L.B. Taus* W . 0 . H e l t o n C . F . W e a v e r R . C . G o s s l e e B . S . Z a c h a r e k d I I Prepared bY Radiological Slte Assessoent Progran Manpower Educailon Research, and Training Dlvision Oak Ridge Assoclated Universlties Oak Ridge, Tennessee 3783f-0117 I FINAL REPORT March 1984 Thts report ls based on work performed under contract number DE-AC05-760R00033 wiLh the DePartment of EnergY. *Evaluatlon Research Corporatlon, Oak Ridge, Tennessee TABLE OF CONTENTS L i s t o f F i g u

49

Harquahala Valley Pwr District | Open Energy Information  

Open Energy Info (EERE)

Harquahala Valley Pwr District Harquahala Valley Pwr District Jump to: navigation, search Name Harquahala Valley Pwr District Place Arizona Utility Id 8139 Utility Location Yes Ownership P NERC Location WECC NERC WECC Yes Activity Buying Transmission Yes Activity Buying Distribution Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png Gin Commercial Irrigation Pumping Commercial Non-Irrigation Agriculture Commercial Average Rates Industrial: $0.0565/kWh References ↑ "EIA Form EIA-861 Final Data File for 2010 - File1_a" Retrieved from "http://en.openei.org/w/index.php?title=Harquahala_Valley_Pwr_District&oldid=410799

50

Oxygen Control in PWR Secondary Coolant  

Science Conference Proceedings (OSTI)

Installation of two deaeration devices at a reference 1100-MWe PWR could improve deaeration of secondary coolant. Direct contact feedwater heaters (DCHs) are readily retrofittable in the condenser neck; condensers too small to accommodate DCHs could incorporate external bubbling devices.

1988-12-14T23:59:59.000Z

51

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

Office of Legacy Management (LM)

Army Corps of Engineers Army Corps of Engineers Buffalo District Office 1776 Niagara Street, Buffalo, New York, 14207 Explanation of Significant Differences for the Rattlesnake Creek Portion of the Ashland Sites Tonawanda, New York September 20, 2004 Formerly Utilized Sites Remedial Action Program Explanation of Significant Differences for the Rattlesnake Creek Portion of the Ashland Sites Table of Contents I. INTRODUCTION 1 II. SITE mSTORY, CONTAMINATION AND SELECTED REMEDy 2 A. Site History 2 B. Original Remedy 3 III. BASIS FOR TmS DOCUMENT.................................*...................*..........*................*.**** 3 A. Summary of Additional Information 3 B. References 4 IV. DESCRIPTION OF SIGNIFICANT DIFFERENCES 4 V. SUPPORT AGENCY COMMENTS 5 VI. STATUTORY DETERMINATIONS 5

52

"1. Robert Moses Niagara","Hydroelectric","New York Power Authority",2353  

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

York" York" "1. Robert Moses Niagara","Hydroelectric","New York Power Authority",2353 "2. Ravenswood","Gas","TC Ravenswood LLC",2330 "3. Nine Mile Point Nuclear Station","Nuclear","Nine Mile Point Nuclear Sta LLC",1773 "4. Oswego Harbor Power","Petroleum","NRG Oswego Harbor Power Operations Inc",1648 "5. Northport","Gas","National Grid Generation LLC",1569 "6. Astoria Generating Station","Gas","U S Power Generating Company LLC",1315 "7. Roseton Generating Station","Gas","Dynegy Northeast Gen Inc",1212 "8. Blenheim Gilboa","Pumped Storage","New York Power Authority",1160

53

Northwest Rural Pub Pwr Dist | Open Energy Information  

Open Energy Info (EERE)

Northwest Rural Pub Pwr Dist Jump to: navigation, search Name Northwest Rural Pub Pwr Dist Place Nebraska Utility Id 13805 Utility Location Yes Ownership P NERC Location WECC NERC...

54

Vermont Yankee Nucl Pwr Corp | Open Energy Information  

Open Energy Info (EERE)

Yankee Nucl Pwr Corp Jump to: navigation, search Name Vermont Yankee Nucl Pwr Corp Place Vermont Utility Id 19796 Utility Location Yes Ownership I NERC Location NPCC NERC NPCC Yes...

55

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

Science Conference Proceedings (OSTI)

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.

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

56

Storage of burned PWR and BWR fuel  

SciTech Connect

In the last few years, credit for fuel burnup has been allowed in the design and criticality safety analysis of high-density spent-fuel storage racks. Design and operating philosophies, however, differ significantly between pressurized water reactor (PWR)- and boiling water reactor (BWR)-type plants because: (1) PWR storage pools generally use soluble boron, which provides backup criticality control under accident conditions; and (2) BWR fuel generally contains gadolinium burnable poison, which results in a characteristically peaked burnup-dependent reactivity variation. In PWR systems, the reactivity decreases monotonically with burnup in a nearly linear fashion (excluding xenon effects), and a two-region concept is feasible. In BWR systems, the reactivity is initially low, increases as fuel burnup progresses, and reaches a maximum at a burnup where the gadolinium is nearly depleted. In any spent-fuel storage rack design, uncertainties due to manufacturing tolerances and in calculational methods must be included to assure that the highest reactivity (k/sub eff/) is less than the 0.95 US Nuclear Regulatory Commission limit. In the absence of definitive critical experiment data with spent fuel, the uncertainty due to depletion calculations must be assumed on the basis of judgment. High-density spent-fuel storage racks may be designed for both PWR and BWR plants with credit for burnup. However, the design must be tailored to each plant with appropriate consideration of the preferences/specifications of the utility operating staff.

Turner, S.E.

1987-01-01T23:59:59.000Z

57

Oxygen Control in PWR Makeup Water  

Science Conference Proceedings (OSTI)

Three fixed-bed processes can accelerate hydrazine-oxygen reactions in PWR makeup water and reduce oxygen levels to below 5 ppb. In this comparative-test project, activated carbon based systems offered the best combination of low cost, effectiveness, and commercial availability. A second process, employing palladium-coated anion resin, is also commercially available.

1988-02-03T23:59:59.000Z

58

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

Office of Legacy Management (LM)

7At 7At a z'/a tlYr'/ ORNL/RASA-85/ 42 RESULTS OF RADIOLOGICAL MEASUREMENTS TAKEN NEAR JUNCTION OF HIGHWAY 3I AND MILITARY ROAD IN NIAGARA FALLSI NEI{ YOR Accesr to thc inlormalion in thlt rcport is limitcd to tho!. indacatod on the di3tribution lilt and to oopartmont ot Encrgy lnd Oeplrtmcnt of Enorgy Contracton vd' This report was prepared as an account of work sponsored by an agency of the U nited States Government. N€ither the U nited States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied' or assum€s any legal liability or responsibility for the accuracy, completeness' or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not inf ringe privately owned rights. Reference herein

59

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

Office of Legacy Management (LM)

9s' 9s' RESULTS OF RADIOLOGICAL I'IEASUREMENTS HIGHT{AYS 18 AI.ID IO4 IN NIAGARA az76 rl//.ry' ORNL/RASA.85/ 40 TAKEN AT JUNCTION FALLS, NEH YORK Accesr to the information in thit rcport ir limiled to tho!' inOllateO on tho dl3tribution li3t and to OePartment ot Encrgy and Oepartmcnt ol Enotgy Gontracton This report was prepared as an account of work sponsored by an agency of the United States Government. Neitherth€ U nited StatesGovernment norany agency thereof, nor any of their employees, makes any warranty, express or implied' or assumes any legal liability or responsibility for the accuracy, completenessi or usefulness of any information, apparatus, product, or,process disclosed, or represents that its usewould not inlringe privately owned rights" Reference herein

60

Horizontal Drop of 21- PWR Waste Package  

SciTech Connect

The objective of this calculation is to determine the structural response of the waste package (WP) dropped horizontally from a specified height. The WP used for that purpose is the 21-Pressurized Water Reactor (PWR) WP. The scope of this document is limited to reporting the calculation results in-terms of stress intensities. This calculation is associated with the WP design and was performed by the Waste Package Design group in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 16). AP-3.12Q, ''Calculations'' (Ref. 1 1) is used to perform the calculation and develop the document. The sketches attached to this calculation provide the potential dimensions and materials for the 21-PWR WP design.

A.K. Scheider

2007-01-31T23:59:59.000Z

Note: This page contains sample records for the topic "niagara mohawk pwr" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

Cold leg integrity evaluation. Final report. [PWR  

SciTech Connect

The objective of this study was to evaluate the margin of safety against a large break in the cold leg piping system of a Pressurized Water Reactor (PWR) power plant. The study focused on the cold leg piping systems of Arkansas Nuclear-1, St. Lucie-1, and Farley-1 PWR power plants. All components of the cold leg piping systems were examined with the exception of the pressure vessel nozzles and the injection laterals. Both axial and circumferential cracks were postulated to exist at critical areas within the piping system. Their growth as part through and then through wall cracks was synthesized within the framework of Linear Elastic Fracture Mechanics (LEFM). The margin of safety was assessed in terms of through-wall crack development, leak rate, and the formation of a large break. The latter criterion was implemented in terms of both LEFM and plastic collapse concepts.

Mayfield, M.E.; Forte, T.P.; Rodabaugh, E.C.; Leis, B.N.; Eiber, R.J.

1980-02-01T23:59:59.000Z

62

PWR RCS Elevated Silica - Fuel Surveillance  

Science Conference Proceedings (OSTI)

Many PWR plants have recently experienced silica concentration as high as 2-5 ppm in the primary water at startup. That level exceeds the prevailing industry diagnostic limit of 1 ppm for safeguarding fuel from potential deposition of tenacious silicates. The high silica experience is primarily limited to plants using silica-containing Boroflex storage racks, which tend to decay in the intense radiation environment in the storage pool. Some plants using recycled boric acid have also experienced high star...

1999-07-28T23:59:59.000Z

63

Trojan PWR Decommissioning: Large Component Removal Project  

Science Conference Proceedings (OSTI)

While the decommissioning of large commercial nuclear plants in the United States is in its infancy, the technical challenges with associated radioactive waste management are clear. This report describes the removal and disposal of four steam generators and one pressurizer from the Trojan nuclear power plant, the first large PWR to be decommissioned in the United States. The report chronicles the problems, successes, and lessons learned in this project, which was completed on schedule and under budget in...

1997-09-29T23:59:59.000Z

64

PWR Cores with Silicon Carbide Cladding  

Science Conference Proceedings (OSTI)

The feasibility of using present-generation pressurized water reactor (PWR) fuel design, with silicon carbide rather than zirconium-based alloy cladding, to reach higher operational power levels and discharge burnups has been evaluated. A preliminary fuel design using fuel rods with the same dimensions as Westinghouse robust fuel assemblies (RFA), but with fuel pellets that have 10 volume percent central holes, has been adopted. The central holes mitigate the higher fuel temperatures that occur due to th...

2011-07-15T23:59:59.000Z

65

Secondary Side Corrosion of French PWR Steam Generator ... - TMS  

Science Conference Proceedings (OSTI)

Aug 1, 1999 ... Secondary Side Corrosion of French PWR Steam Generator Tubing: Contribution of Surface Analyses to the Understanding of the Degradation ...

66

Hydrogen Effects on PWR SCC Mechanisms in Monocrystalline and ...  

Science Conference Proceedings (OSTI)

Aug 1, 1999... 600 PWR SCC resistance has been assessed by slow strain rate tests in primary water at 360°C. Crack initiation and propagation resistance ...

67

Proposed Coordinated U.S. PWR Reactor Vessel Surveillance ...  

Science Conference Proceedings (OSTI)

Protective Insulated Coating for SCC Mitigation in BWRs · PWR Fuel Deposit Analysis at a B&W Plant with a 24-Month Fuel Cycle · PWSCC of Thermocoax ...

68

Severe accident sequences analyzed for a two-loop PWR  

Science Conference Proceedings (OSTI)

Different severe accident sequences have been analyzed for a two-loop Westinghouse pressurized water reactor (PWR) using the MELCOR code, version 1.8.4. The purpose of this study was to calculate source terms and the timing of events for severe accident sequences at this type of PWR to be used in the HAS-CAL code .The results calculated by MELCOR have been compared to results from the individual plant examination (IPE) of the Kewaunee nuclear power plant, also a two-loop Westinghouse PWR. The results of the Kewaunee IPE were obtained with the severe accident code MAAP.

Carbajo, J.J. [Oak Ridge National Lab., TN (United States)

1997-12-01T23:59:59.000Z

69

East River Elec Pwr Coop, Inc | Open Energy Information  

Open Energy Info (EERE)

Elec Pwr Coop, Inc Elec Pwr Coop, Inc Jump to: navigation, search Name East River Elec Pwr Coop, Inc Place South Dakota Utility Id 5552 Utility Location Yes Ownership C NERC Location MRO NERC MRO Yes Activity Transmission Yes Activity Buying Transmission Yes Alt Fuel Vehicle Yes Alt Fuel Vehicle2 Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png No rate schedules available. Average Rates No Rates Available References ↑ "EIA Form EIA-861 Final Data File for 2010 - File1_a" Retrieved from "http://en.openei.org/w/index.php?title=East_River_Elec_Pwr_Coop,_Inc&oldid=410618"

70

Michigan South Central Pwr Agy | Open Energy Information  

Open Energy Info (EERE)

Pwr Agy Pwr Agy Jump to: navigation, search Name Michigan South Central Pwr Agy Place Michigan Utility Id 12807 Utility Location Yes Ownership M NERC Location RFC NERC RFC Yes Operates Generating Plant Yes Activity Generation Yes Activity Transmission Yes Activity Buying Transmission Yes Activity Wholesale Marketing Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png No rate schedules available. Average Rates No Rates Available References ↑ "EIA Form EIA-861 Final Data File for 2010 - File1_a" Retrieved from "http://en.openei.org/w/index.php?title=Michigan_South_Central_Pwr_Agy&oldid=41107

71

Vermont Public Pwr Supply Auth | Open Energy Information  

Open Energy Info (EERE)

Public Pwr Supply Auth Public Pwr Supply Auth Jump to: navigation, search Name Vermont Public Pwr Supply Auth Place Vermont Utility Id 19780 Utility Location Yes Ownership P NERC Location NPCC NERC NPCC Yes ISO NE Yes Operates Generating Plant Yes Activity Generation Yes Activity Transmission Yes Activity Wholesale Marketing Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png No rate schedules available. Average Rates No Rates Available References ↑ "EIA Form EIA-861 Final Data File for 2010 - File1_a" Retrieved from "http://en.openei.org/w/index.php?title=Vermont_Public_Pwr_Supply_Auth&oldid=411933"

72

Puerto Rico Electric Pwr Authority | Open Energy Information  

Open Energy Info (EERE)

Rico Electric Pwr Authority Rico Electric Pwr Authority Jump to: navigation, search Name Puerto Rico Electric Pwr Authority Place Puerto Rico Utility Id 15497 Utility Location Yes Ownership S NERC Location PR Operates Generating Plant Yes Activity Generation Yes Activity Transmission Yes Activity Distribution Yes Alt Fuel Vehicle Yes Alt Fuel Vehicle2 Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png 312 - Industrial General at Primary Distribution Average Rates No Rates Available References ↑ "EIA Form EIA-861 Final Data File for 2010 - File1_a" Retrieved from "http://en.openei.org/w/index.php?title=Puerto_Rico_Electric_Pwr_Authority&oldid=411417"

73

BWR Fuel Assembly BWR Fuel Assembly PWR Fuel Assembly  

National Nuclear Security Administration (NNSA)

Spacer Grid Structural Guide Tube End Fitting Fuel Rod Upper Tie Plate ULTRAFLOW Spacer Water Channel Part-length Fuel Rod Lower Tie Plate PWR pressurized water reactor BWR ...

74

Fuel cycle optimization of thorium and uranium fueled PWR systems  

E-Print Network (OSTI)

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

Garel, Keith Courtnay

1977-01-01T23:59:59.000Z

75

Central Electric Pwr Coop, Inc | Open Energy Information  

Open Energy Info (EERE)

Coop, Inc Jump to: navigation, search Name Central Electric Pwr Coop, Inc Place South Carolina Utility Id 40218 Utility Location Yes Ownership C NERC Location SERC NERC SERC Yes...

76

Rushmore Electric Pwr Coop Inc | Open Energy Information  

Open Energy Info (EERE)

Rushmore Electric Pwr Coop Inc Rushmore Electric Pwr Coop Inc Jump to: navigation, search Name Rushmore Electric Pwr Coop Inc Place South Dakota Utility Id 16443 Utility Location Yes Ownership C NERC Location MRO NERC MRO Yes Activity Transmission Yes Activity Buying Transmission Yes Activity Wholesale Marketing Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png Residential Average Rates No Rates Available References ↑ "EIA Form EIA-861 Final Data File for 2010 - File1_a" Retrieved from "http://en.openei.org/w/index.php?title=Rushmore_Electric_Pwr_Coop_Inc&oldid=41147

77

Renville-Sibley Coop Pwr Assn | Open Energy Information  

Open Energy Info (EERE)

Renville-Sibley Coop Pwr Assn Renville-Sibley Coop Pwr Assn Jump to: navigation, search Name Renville-Sibley Coop Pwr Assn Place Minnesota Utility Id 15845 Utility Location Yes Ownership C NERC Location MRO NERC MRO Yes ISO MISO Yes Activity Distribution Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png No rate schedules available. Average Rates Residential: $0.0883/kWh Commercial: $0.0867/kWh Industrial: $0.0471/kWh References ↑ "EIA Form EIA-861 Final Data File for 2010 - File1_a" Retrieved from "http://en.openei.org/w/index.php?title=Renville-Sibley_Coop_Pwr_Assn&oldid=41144

78

PWR Primary Water Chemistry Guidelines: Volume 1 Revision 4  

Science Conference Proceedings (OSTI)

State-of-the art water chemistry programs help ensure the continued integrity of reactorcoolant system (RCS) materials of construction and fuel cladding, ensure satisfactorycore performance, and support the industry trend toward reduced radiation fields. These revised PWR Primary Water Chemistry Guidelines, prepared by a committee ofindustry experts, reflect the recent field and laboratory data on primary coolant systemcorrosion and performance issues. PWR operators can use these Guidelines to updatethei...

1999-03-31T23:59:59.000Z

79

Condensate Polishing Guidelines for PWR and BWR Plants -- 1997 Revision  

Science Conference Proceedings (OSTI)

Successful condensate polishing operations maintain control of ionic and particulate impurity transport to the PWR steam generator and the BWR reactor and recirculation system. This report presents revisions of EPRI's 1993 nuclear industry consensus guidelines for the design and operation of deep bed and filter demineralizer condensate polishers. This advice is consistent with the 1996 revisions of EPRI's BWR Water Chemistry Guidelines (TR-103515-R1) and PWR Secondary Water Chemistry Guidelines (TR-10213...

1997-10-31T23:59:59.000Z

80

A PWR Thorium Pin Cell Burnup Benchmark  

SciTech Connect

As part of work to evaluate the potential benefits of using thorium in LWR fuel, a thorium fueled benchmark comparison was made in this study between state-of-the-art codes, MOCUP (MCNP4B + ORIGEN2), and CASMO-4 for burnup calculations. The MOCUP runs were done individually at MIT and INEEL, using the same model but with some differences in techniques and cross section libraries. Eigenvalue and isotope concentrations were compared on a PWR pin cell model up to high burnup. The eigenvalue comparison as a function of burnup is good: the maximum difference is within 2% and the average absolute difference less than 1%. The isotope concentration comparisons are better than a set of MOX fuel benchmarks and comparable to a set of uranium fuel benchmarks reported in the literature. The actinide and fission product data sources used in the MOCUP burnup calculations for a typical thorium fuel are documented. Reasons for code vs code differences are analyzed and discussed.

Weaver, Kevan Dean; Zhao, X.; Pilat, E. E; Hejzlar, P.

2000-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "niagara mohawk pwr" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

82

Fort Drum integrated resource assessment. Volume 3, Resource assessment  

Science Conference Proceedings (OSTI)

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.

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

83

East Mississippi Elec Pwr Assn | Open Energy Information  

Open Energy Info (EERE)

Mississippi Elec Pwr Assn Mississippi Elec Pwr Assn Jump to: navigation, search Name East Mississippi Elec Pwr Assn Place Mississippi Utility Id 5578 Utility Location Yes Ownership C NERC Location SERC NERC SERC Yes Activity Distribution Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png GENERAL POWER RATE--SCHEDULE GSA--Rate 50 Commercial GENERAL POWER RATE--SCHEDULE GSA--Rate 51 Commercial GENERAL POWER RATE--SCHEDULE GSA--Rate 54 GENERAL POWER RATE--SCHEDULE GSA--Rate 80 Commercial GENERAL POWER RATE--SCHEDULE SGSB--Rate 56 General Power- Rate 40 Commercial LIGHTING--SCHEDULE LS--Rate SL

84

Grand Valley Rrl Pwr Line, Inc | Open Energy Information  

Open Energy Info (EERE)

Pwr Line, Inc Pwr Line, Inc Jump to: navigation, search Name Grand Valley Rrl Pwr Line, Inc Place Colorado Utility Id 7563 Utility Location Yes Ownership C NERC Location WECC NERC WECC Yes Activity Distribution Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png Commercial and Small Power Service, Three Phase Schedule (25)-CSP-1 Commercial Farm and Home (Residential) Service Schedule (10)-FH-1 Residential Industrial Service Schedule (50) -IND-1 Industrial Irrigation Service Schedule (40)-I-1 Commercial Large Power Service Schedule (30) -LP-1 Industrial Nonresidential - General Schedule (20)-NRG-1 Commercial

85

Cuming County Public Pwr Dist | Open Energy Information  

Open Energy Info (EERE)

Cuming County Public Pwr Dist Cuming County Public Pwr Dist (Redirected from Cuming County Public Power District) Jump to: navigation, search Name Cuming County Public Pwr Dist Place West Point, Nebraska Utility Id 4632 Utility Location Yes Ownership P NERC Location MRO NERC MRO Yes Activity Distribution Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] SGIC[2] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png Gravel Pit Service Rate Industrial High Voltage Delivery Rate Industrial Irrigation Service Anytime Control Industrial Irrigation Service Every-Other-Day Control Industrial Irrigation Service No Control Industrial Large Power Rate Industrial

86

Keosauqua Municipal Light & Pwr | Open Energy Information  

Open Energy Info (EERE)

Keosauqua Municipal Light & Pwr Keosauqua Municipal Light & Pwr Jump to: navigation, search Name Keosauqua Municipal Light & Pwr Place Iowa Utility Id 10181 Utility Location Yes Ownership M NERC Location SERC NERC SERC Yes ISO Other Yes Activity Distribution Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png Commercial Public Commercial Demand Rate Industrial Industrial Rate Industrial Residential Residential Security Light Lighting Average Rates Residential: $0.1040/kWh Commercial: $0.0858/kWh Industrial: $0.1190/kWh References ↑ "EIA Form EIA-861 Final Data File for 2010 - File1_a"

87

Seward County Rrl Pub Pwr Dist | Open Energy Information  

Open Energy Info (EERE)

Seward County Rrl Pub Pwr Dist Seward County Rrl Pub Pwr Dist Jump to: navigation, search Name Seward County Rrl Pub Pwr Dist Place Nebraska Utility Id 16954 Utility Location Yes Ownership P NERC Location SPP NERC SPP Yes RTO SPP Yes Activity Distribution Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png Cabins Commercial Commercial Electric Speace Heating L29 Commercial Commercial Electric Speace Heating L30 Commercial Grain Dryer Accounts Single Phase Commercial Grain Dryer Accounts Three Phase Commercial Industrial Service Industrial Industrial Service L75 Industrial Irrigation Service 3 Day control Industrial

88

Elkhorn Rural Public Pwr Dist | Open Energy Information  

Open Energy Info (EERE)

Rural Public Pwr Dist Rural Public Pwr Dist Jump to: navigation, search Name Elkhorn Rural Public Pwr Dist Place Nebraska Utility Id 5780 Utility Location Yes Ownership P NERC Location SPP NERC SPP Yes Activity Distribution Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png Mercury Vapor Lamp 175 Watt Lighting Mercury Vapor Lamp 250 Watt Lighting Mercury Vapor Lamp 400 Watt Lighting Metal Halide 1000 Watt Lighting Metal Halide 1500 Watt Lighting Metal Halide 400 Watt Lighting RATE 1,3- Farm Residential, Commercial, Cabins, Seasonal--Single Phase Commercial RATE 12, 69- Urban Commercial Electric Space Heating, Single Phase

89

Pearl River Valley El Pwr Assn | Open Energy Information  

Open Energy Info (EERE)

El Pwr Assn El Pwr Assn Jump to: navigation, search Name Pearl River Valley El Pwr Assn Place Mississippi Utility Id 14563 Utility Location Yes Ownership C NERC Location SERC NERC SERC Yes Activity Distribution Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png 1 GS General Service 10 LGS-6 Large General Service 2 GS-DG General Service Distributed Generation 20 LP-6 Large Power 21 LP-AE-2 Large Power All Electric 22 LP-PM-6 Large Power Primary Meter 23 LP-PM-AE-2 Large Power Primary Metering All Electric 3 GS-TWH General Service Tankless Water Heater 3 TGS-1 Temporary General Service

90

South Mississippi El Pwr Assn | Open Energy Information  

Open Energy Info (EERE)

El Pwr Assn El Pwr Assn (Redirected from South Mississippi Electric Power Association (SMEPA)) Jump to: navigation, search Name South Mississippi El Pwr Assn Place Mississippi Utility Id 17568 Utility Location Yes Ownership C NERC Location SERC NERC SERC Yes Operates Generating Plant Yes Activity Generation Yes Activity Transmission Yes Activity Buying Transmission Yes Activity Wholesale Marketing Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png No rate schedules available. Average Rates No Rates Available References ↑ "EIA Form EIA-861 Final Data File for 2010 - File1_a"

91

Polk County Rural Pub Pwr Dist | Open Energy Information  

Open Energy Info (EERE)

County Rural Pub Pwr Dist County Rural Pub Pwr Dist Jump to: navigation, search Name Polk County Rural Pub Pwr Dist Place Nebraska Utility Id 15188 Utility Location Yes Ownership P NERC Location MRO Activity Distribution Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png Cabin Residential City of Stromsburg - CYC 3/REV 8 Commercial Commercial Elec Space Heating - One Phase Commercial Commercial Elec Space Heating - Three Phase Commercial General Service CYC 2/REV1 - Three Phase Residential General Service CYC/Rev1 - Single Phase Residential Irrigation CYC 1 - Five 1/2 day - Single Phase Industrial

92

Crawfordsville Elec, Lgt & Pwr | Open Energy Information  

Open Energy Info (EERE)

Crawfordsville Elec, Lgt & Pwr Crawfordsville Elec, Lgt & Pwr Jump to: navigation, search Name Crawfordsville Elec, Lgt & Pwr Place Indiana Utility Id 4508 Utility Location Yes Ownership M NERC Location RFC NERC RFC Yes ISO MISO Yes Operates Generating Plant Yes Activity Generation Yes Activity Transmission Yes Activity Distribution Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png Commercial General Power- Single Phase Commercial Commercial General Power- Three Phase Commercial Industrial Primary Power Industrial Outdoor Lighting- 100 watt sodium vapor Lighting Outdoor Lighting- 175 watt mercury vapor Lighting

93

Singing River Elec Pwr Assn | Open Energy Information  

Open Energy Info (EERE)

Singing River Elec Pwr Assn Singing River Elec Pwr Assn Place Alabama Utility Id 17252 References Energy Information Administration.[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png No rate schedules available. Average Rates Residential: $0.1080/kWh Commercial: $0.1110/kWh The following table contains monthly sales and revenue data for Singing River Elec Pwr Assn (Alabama). Month RES REV (THOUSAND $) RES SALES (MWH) RES CONS COM REV (THOUSAND $) COM SALES (MWH) COM CONS IND_REV (THOUSAND $) IND SALES (MWH) IND CONS OTH REV (THOUSAND $) OTH SALES (MWH) OTH CONS TOT REV (THOUSAND $) TOT SALES (MWH) TOT CONS 2009-03 34.718 304.244 306 4.774 40.12 19 39.492 344.364 325

94

South Central Public Pwr Dist | Open Energy Information  

Open Energy Info (EERE)

Pwr Dist Pwr Dist Jump to: navigation, search Name South Central Public Pwr Dist Place Nebraska Utility Id 17548 Utility Location Yes Ownership P NERC Location MRO NERC MRO Yes RTO SPP Yes Activity Distribution Yes Activity Retail Marketing Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png Area&Directional Lighting Service District Owned Pole Rental Lighting Area&Directional Lighting Service District Owned Span of Conductor Rent Lighting Area&Directional Lighting Service Rate N721 Lighting Area&Directional Lighting Service Rate N721 - Fringe Lighting

95

Prentiss County Elec Pwr Assn | Open Energy Information  

Open Energy Info (EERE)

Prentiss County Elec Pwr Assn Prentiss County Elec Pwr Assn Jump to: navigation, search Name Prentiss County Elec Pwr Assn Place Mississippi Utility Id 15334 Utility Location Yes Ownership C NERC Location SERC NERC SERC Yes Activity Distribution Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png General Power GSA 1 (50 kW and Under) CT Metered Single-Phase Commercial General Power GSA 1 (50 kW and Under) Multi-Phase Commercial General Power GSA 1 (50 kW and Under) Single-Phase Commercial General Power GSA 2 (51 kW - 1000 kW) Commercial General Power GSA 3 (1001 kW - 5000 kW) Industrial

96

Fuel Reliability Guidelines: PWR Grid-to-Rod Fretting  

Science Conference Proceedings (OSTI)

Relative motion between the fuel rods and fuel assembly spacer grids can lead to excessive fuel rod wear and, in some cases, to fuel rod failure. Based on industry data, grid-to-rod-fretting (GTRF) has been the number one cause of fuel failures within the U.S. pressurized water reactor (PWR) fleet, accounting for more than 70% of all PWR leaking fuel assemblies. Most of the GTRF failures have occurred during an assembly's final cycle of operation when it is located on the periphery of the core adjacent t...

2008-07-28T23:59:59.000Z

97

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

Office of Legacy Management (LM)

Nf7 n-q Nf7 n-q gz75 tLtY r 1 irl,r:'a :.a l: i , l : i l ',:lr.:'. itl:t i .,,::l ' i , t . . ORNL/RASA-85/1 RESULTS OF THE II4OBILE GAMMA SCANNING ACTIVITIES IN NIAGARA FALLS, NEvl YORK AREA Access to the information in this report is limited to thoss indicated on the distribution list and io Department ol Energy ancl Depsrtment of Energy Contractors This report was prepared as an account ol work sponsored by an agency of the United States Government. Neither the U nited StatesGovernment nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any informalion, apparatus, product, or process disclosed, or represents thal its use would not inf ringe

98

niagaraVP  

Office of Legacy Management (LM)

at the site. Over the years, contaminated materials stored at the site were subject to wind and erosion, causing contaminants to migrate off site onto other properties. Referred...

99

Present status and development of PWR fuel in China  

Science Conference Proceedings (OSTI)

This paper describes the nuclear power plant fuel design, manufacture and R and D capability in Republic of China; the progress condition of Qinshan fuel assemblies and preparatory work for Guandong and other PWR's fuel assemblies; and the program of research and development of high performance fuel.

Shouhui, D.; Yinian, Z; Dingcang, T.

1988-01-01T23:59:59.000Z

100

Analysis of PWR RCS Injection Strategy During Severe Accident  

Science Conference Proceedings (OSTI)

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

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

2004-05-15T23:59:59.000Z

Note: This page contains sample records for the topic "niagara mohawk pwr" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


101

Studies of a small PWR for onsite industrial power  

SciTech Connect

Information on the use of a 300 to 400 MW(t) PWR type reactor for industrial applications is presented concerning the potential market, reliability considerations, reactor plant description, construction techniques, comparison between nuclear and fossil-fired process steam costs, alternative fossil-fired steam supplies, and industrial application.

Klepper, O.H.; Smith, W.R.

1977-04-19T23:59:59.000Z

102

North Central Public Pwr Dist | Open Energy Information  

Open Energy Info (EERE)

North Central Public Pwr Dist North Central Public Pwr Dist Place Nebraska Utility Id 13698 Utility Location Yes Ownership P NERC Location MRO NERC MRO Yes RTO SPP Yes Activity Distribution Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png 1- Single Phase Farm Residential Residential 10- Irrigation - Pivot Drive Commercial 11- Irrigation - No Control Commercial 12- Irrigation - 2 Day Control & Sunday Control Commercial 13- Irrigation - 3 Day Control & Sunday Control Commercial 14- Irrigation - Anytime Control & Sunday control Commercial 19- Single Phase Farm Residential (Leased areas) Commercial

103

South Mississippi El Pwr Assn | Open Energy Information  

Open Energy Info (EERE)

Assn Assn Jump to: navigation, search Name South Mississippi El Pwr Assn Place Mississippi Utility Id 17568 Utility Location Yes Ownership C NERC Location SERC NERC SERC Yes Operates Generating Plant Yes Activity Generation Yes Activity Transmission Yes Activity Buying Transmission Yes Activity Wholesale Marketing Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png No rate schedules available. Average Rates No Rates Available References ↑ "EIA Form EIA-861 Final Data File for 2010 - File1_a" Retrieved from "http://en.openei.org/w/index.php?title=South_Mississippi_El_Pwr_Assn&oldid=411549

104

Cuming County Public Pwr Dist | Open Energy Information  

Open Energy Info (EERE)

Cuming County Public Pwr Dist Cuming County Public Pwr Dist Place West Point, Nebraska Utility Id 4632 Utility Location Yes Ownership P NERC Location MRO NERC MRO Yes Activity Distribution Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] SGIC[2] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png Gravel Pit Service Rate Industrial High Voltage Delivery Rate Industrial Irrigation Service Anytime Control Industrial Irrigation Service Every-Other-Day Control Industrial Irrigation Service No Control Industrial Large Power Rate Industrial Municipal Commercial Demand Rate Commercial Municipal Commercial Demand Rate Three Phase V2 Commercial Municipal Commercial Electric Space Heating Rate Commercial

105

Fort Drum integrated resource assessment  

Science Conference Proceedings (OSTI)

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.

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

106

Modeling the Oxygen - Hydrazine Reaction in PWR Secondary Feedwater  

Science Conference Proceedings (OSTI)

The proper control of oxygen in primary water reactor (PWR) secondary feedwater, using hydrazine, has been an enduring issue. The requirements on the oxygen concentration are partly opposing. Fully deoxygenated conditions in the steam generators are essential to minimize corrosion. On the other hand, some oxygen in the feedwater counteracts corrosion of carbon steel surfaces and the transport of corrosion products to the steam generators. Optimization is, therefore, essential. This work applies the frame...

2008-06-26T23:59:59.000Z

107

PWR Axial Offset Anomaly (AOA) Guidelines, Revision 1  

Science Conference Proceedings (OSTI)

Axial offset anomaly (AOA) is defined as a significant negative axial offset deviation from the predicted nuclear design value. AOA results from the incorporation of boron within corrosion product deposits on the upper spans of high-duty pressurized water reactor (PWR) fuel assemblies. The consequences of this process are an erosion of shutdown margin and loss of operational flexibility by control room operators, particularly during power transients.

2004-06-28T23:59:59.000Z

108

Design study of long-life PWR using thorium cycle  

SciTech Connect

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.

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

2012-06-06T23:59:59.000Z

109

Materials Reliability Program: Material Selection for the PWR Supplemental Surveillance Program (PSSP) (MRP-364)  

Science Conference Proceedings (OSTI)

This report describes the results of the methodical selection of previously irradiated and tested pressurized water reactor (PWR) surveillance specimens for use in a PWR Supplemental Surveillance Program (PSSP). The PSSP will consist of two supplemental surveillance capsules that will be irradiated in host PWR plants, thereby increasing the fluence of the surveillance specimens. When tested in ~2025, the PSSP capsules will yield a significant amount of high-fluence transition temperature shift (TTS) ...

2013-06-25T23:59:59.000Z

110

Ageing and Toughness of a Mn-Ni-Mo PWR Steel  

Science Conference Proceedings (OSTI)

Abstract Scope, Mn-Ni-Mo steels are widely used in the fabrication of pressurisers, steam generators and pressure vessels of pressurised water reactors (PWR).

111

[en] THERMOHIDRAULIC MODEL FOR A TYPICAL STEAM GENERATOR OF PWR NUCLEAR POWER PLANTS.  

E-Print Network (OSTI)

??[pt] Muitas centrais nucleares do tipo PWR utilizam vapor produzido em geradores de vapor do tipo tubos em U invertido, com recirculação interna natural, nos… (more)

CARLOS VALOIS MACIEL BRAGA

2011-01-01T23:59:59.000Z

112

Materials Reliability Program: Strategies for Managing Aging Effects in PWR Vessel Internals - Interim Update (MRP-99)  

Science Conference Proceedings (OSTI)

This report updates the previous EPRI report on developing strategies for managing aging effects in pressurized water reactor (PWR) internals during the license renewal term.

2003-12-04T23:59:59.000Z

113

SCC Behavior of Alloy 52M/182 Weld Overlay in a PWR Environment  

Science Conference Proceedings (OSTI)

NRC/EPRI Welding Residual Stress Validation Program (Phase III) · On the Microstructure of Alloy 600 SCC Cracks Observed by TEM on PWR SG Pulled Tubes ...

114

PWR Fuel Deposit Analysis at a B&W Plant with a 24-Month Fuel ...  

Science Conference Proceedings (OSTI)

The patented AREVA sampling method was applied at the end of cycle ... Crack Growth Rates of Irradiated Commercial Stainless Steels in BWR and PWR ...

115

SCC Crack Growth Rate of Alloy 82 in PWR Primary Water Conditions  

Science Conference Proceedings (OSTI)

Protective Insulated Coating for SCC Mitigation in BWRs · PWR Fuel Deposit Analysis at a B&W Plant with a 24-Month Fuel Cycle · PWSCC of Thermocoax ...

116

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

Science Conference Proceedings (OSTI)

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.

Not Available

1993-05-01T23:59:59.000Z

117

Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-175)  

Science Conference Proceedings (OSTI)

This report provides screening criteria and their technical bases for age-related degradation evaluation of Pressurized Water Reactor (PWR) internals component items. It is a key element in an overall strategy that uses knowledge of internals design, materials, and material properties and applies screening methodologies for known age-related degradation mechanisms to manage the effects of aging in PWR internals.

2005-12-12T23:59:59.000Z

118

21-PWR WASTE PACKAGE WITH ABSORBER PLATES LOADING CURVE EVALUATION  

Science Conference Proceedings (OSTI)

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.

J.M. Scaglione

2004-12-17T23:59:59.000Z

119

TRANSPORT CHARACTERISTICS OF SELECTED PWR LOCA GENERATED DEBRIS.  

SciTech Connect

In the unlikely event of a Loss of Coolant Accident (LOCA) in a pressurized water reactor (PWR), break jet impingement would dislodge thermal insulation from nearby piping, as well as other materials within the containment, such as paint chips, concrete dust, and fire barrier materials. Steam/water flows induced by the break and by the containment sprays would transport debris to the containment floor. Subsequently, debris would likely transport to and accumulate on the suction sump screens of the emergency core cooling system (ECCS) pumps, thereby potentially degrading ECCS performance and possibly even failing the ECCS. In 1998, the U. S. Nuclear Regulatory Commission (NRC) initiated a generic study (Generic Safety Issue-191) to evaluate the potential for the accumulation of LOCA related debris on the PWR sump screen and the consequent loss of ECCS pump net positive suction head (NPSH). Los Alamos National Laboratory (LANL), supporting the resolution of GSI-191, was tasked with developing a method for estimating debris transport in PWR containments to estimate the quantity of debris that would accumulate on the sump screen for use in plant specific evaluations. The analytical method proposed by LANL, to predict debris transport within the water that would accumulate on the containment floor, is to use computational fluid dynamics (CFD) combined with experimental debris transport data to predict debris transport and accumulation on the screen. CFD simulations of actual plant containment designs would provide flow data for a postulated accident in that plant, e.g., three-dimensional patterns of flow velocities and flow turbulence. Small-scale experiments would determine parameters defining the debris transport characteristics for each type of debris. The containment floor transport methodology will merge debris transport characteristics with CFD results to provide a reasonable and conservative estimate of debris transport within the containment floor pool and subsequent accumulation of debris on the sump screen. The complete methodology will, of course, include a means of estimating debris generation, transport to the containment floor, transport to the sump screen, and the resulting loss of NPSH.

A. K. MAJI; B. MARSHALL; ET AL

2000-10-01T23:59:59.000Z

120

Wolverine Pwr Supply Coop, Inc | Open Energy Information  

Open Energy Info (EERE)

Supply Coop, Inc Supply Coop, Inc Jump to: navigation, search Name Wolverine Pwr Supply Coop, Inc Place Michigan Utility Id 20910 Utility Location Yes Ownership C NERC Location RFC NERC RFC Yes Operates Generating Plant Yes Activity Generation Yes Activity Transmission Yes Activity Buying Transmission Yes Activity Distribution Yes Activity Wholesale Marketing Yes Activity Bundled Services Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png LP-C ( Large Power Choice) Schedule A- Residential Residential Schedule AH- Residential Residential Schedule AS- Residential Residential

Note: This page contains sample records for the topic "niagara mohawk pwr" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

Benefits of Integrating PWR and RTI Advanced Gasification Technologies for  

NLE Websites -- All DOE Office Websites (Extended Search)

Syngas Processing Systems Syngas Processing Systems Benefits of Integrating PWR and RTI Advanced Gasification Technologies for Hydrogen-Rich Syngas Production Research Triangle Institute (RTI) Project Number: FE0012066 Project Description The project will assess the potential for integrated advanced technologies to substantially reduce capital and production costs for hydrogen-rich syngas with near-zero emissions from coal gasification for power production with carbon capture and for coal-to-liquids (specifically methanol) with carbon capture. These integrated technologies include those already tested successfully at pilot-scale with a new and innovative water-gas-shift technology, to show how multiple advanced technologies will leverage each other for significant cost and efficiency gains.

122

Stanton County Public Pwr Dist | Open Energy Information  

Open Energy Info (EERE)

(Redirected from Stanton County Public Power District) (Redirected from Stanton County Public Power District) Jump to: navigation, search Name Stanton County Public Pwr Dist Place Stanton, Nebraska Utility Id 17979 Utility Location Yes Ownership P NERC Location SPP NERC SPP Yes Activity Distribution Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] SGIC[2] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Stanton County Public Power District Smart Grid Project was awarded $397,000 Recovery Act Funding with a total project value of $794,000. Utility Rate Schedules Grid-background.png Commercial Pumping Single Phase Commercial Commercial Pumping Three Phase Commercial Commercial Single Phase Commercial Commercial Three Phase Commercial

123

Red River Valley Coop Pwr Assn | Open Energy Information  

Open Energy Info (EERE)

Assn Assn Jump to: navigation, search Name Red River Valley Coop Pwr Assn Place Minnesota Utility Id 26939 Utility Location Yes Ownership C NERC Location MRO NERC MRO Yes Activity Distribution Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png Decorative Post Light Lighting General - Three Phase Industrial Post Light Lighting Residential - Off-Peak Residential Residential - Single Phase Residential Security Parking Light Lighting Yard Light Lighting Average Rates Residential: $0.0892/kWh Commercial: $0.0883/kWh References ↑ "EIA Form EIA-861 Final Data File for 2010 - File1_a"

124

Singing River Elec Pwr Assn (Mississippi) | Open Energy Information  

Open Energy Info (EERE)

Assn (Mississippi) Assn (Mississippi) Jump to: navigation, search Name Singing River Elec Pwr Assn Place Mississippi Utility Id 17252 Utility Location Yes Ownership C NERC Location SERC NERC ERCOT Yes NERC SERC Yes ISO Other Yes Activity Distribution Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1]Energy Information Administration Form 826[2] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png Commercial Delivery Point Service (Over 2500 kW) Commercial Commercial General Service (25 kw - 249 kW) Commercial Commercial General Service (25 kw - 249 kW) Primary Voltage Commercial Commercial Large General Service (250 kW - 749 kW) Commercial

125

Nuclear Plant Design and Modification Guidelines for PWR Steam Generator Reliability  

Science Conference Proceedings (OSTI)

Operating and maintenance experience relative to PWR steam generator reliability has produced a variety of "lessons learned." This information has been incorporated in a series of guidelines to aid utilities in major plant modifications and new plant construction.

1991-09-25T23:59:59.000Z

126

Cost Impact of Using ISG-8 Rev. 3 for PWR Spent Fuel Pool Criticality Analysis  

Science Conference Proceedings (OSTI)

Nuclear Regulatory Commission (NRC) guidance for applying burnup credit in criticality analyses for spent fuel storage and transportation requirements recently changed with the release of Interim Staff Guidance (ISG) 8 Revision 3, Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks. If ISG-8 Rev. 3 were imposed upon pressurized water reactor (PWR) spent fuel pool (SFP) criticality analyses, the burnup requirements for loading would ...

2012-11-21T23:59:59.000Z

127

Evaluation and Categorization of Secondary System Layup and Cleanup Practices for PWR Plants  

Science Conference Proceedings (OSTI)

To determine ways to minimize corrosion-product transport to the secondary side of PWR steam generators, layup and post-shutdown cleanup practices now in use or proposed by utilities with operating PWR plants were examined. The results show that about 30% of the plants attempt routine layup of secondary systems during plant outages and about 60% attempt system cleanup before and during startup.

1982-12-01T23:59:59.000Z

128

Materials Reliability Program: Fracture Toughness Testing of Decommissioned PWR Core Internals Material Samples (MRP-160)  

Science Conference Proceedings (OSTI)

Pressurized water reactor (PWR) cores operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs requires detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel (internals) that are inherent to such conditions. This report contains the results of PWR environment fracture toughness testing of samples machined from decommissione...

2005-09-26T23:59:59.000Z

129

Containment integrity of SEP plants under combined loads. [PWR; BWR  

Science Conference Proceedings (OSTI)

Because the containment structure is the last barrier against the release of radioactivity, an assessment was undertaken to identify the design weaknesses and estimate the margins of safety for the SEP containments under the postulated, combined loading conditions of a safe shutdown earthquake (SSE) and a design basis accident (DBA). The design basis accident is either a loss-of-coolant accident (LOCA) or a main steam line break (MSLB). The containment designs analyzed consisted of three inverted light-bulb shaped drywells used in boiling water reactor (BWR) systems, and three steel-lined concrete containments and a spherical steel shell used in pressurized water reactor (PWR) systems. These designs cover a majority of the containment types used in domestic operating plants. The results indicate that five of the seven designs are adequate even under current design standards. For the remaining two designs, the possible design weaknesses identified were buckling of the spherical steel shell and over-stress in both the radial and tangential directions in one of the concrete containments near its base.

Lo, T.; Nelson, T.A.; Chen, P.Y.; Persinko, D.; Grimes, C.

1984-06-01T23:59:59.000Z

130

Analysis of Potential Hydrogen Risk in the PWR Containment  

SciTech Connect

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

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

2006-07-01T23:59:59.000Z

131

ERNEST ORLANDO LAWRENCE BERKELEY NATIONAL LABORATORY  

E-Print Network (OSTI)

in this report was coordinated by the Consortium for Electric Reliability Technology Solutions and funded at Niagara Mohawk Power Corporation that have faced day-ahead electricity market prices as their default and do not reschedule usage. Average price response estimates are modest: the overall substitution

132

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Assembly Shaker Test for Determining Loads on a PWR Assembly Assembly Shaker Test for Determining Loads on a PWR Assembly under Surrogate Normal Conditions of Truck Transport R0.1 Fuel Assembly Shaker Test for Determining Loads on a PWR Assembly under Surrogate Normal Conditions of Truck Transport R0.1 The United States 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 with the trend of nuclear power plants using reactor fuel for a longer time, creates questions concerning the ability of this aged, high-burnup fuel to withstand stresses and strains seen during normal conditions of transport from its current location to a future consolidated storage facility or permanent repository. UNFD R&D conducted testing employing surrogate instrumented

133

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Embrittlement and DBTT of High-Burnup PWR Fuel Cladding Alloys Embrittlement and DBTT of High-Burnup PWR Fuel Cladding Alloys Embrittlement and DBTT of High-Burnup PWR Fuel Cladding Alloys Structural analyses of high-burnup (HBU) fuel require cladding mechanical properties and failure limits to assess fuel behavior during long-term dry-cask storage and transportation. Pre-storage drying-transfer operations and early stage storage subject cladding to higher temperatures and pressure-induced tensile hoop stresses relative to in-reactor operation and pool storage. Under these conditions, radial hydrides may precipitate during slow cooling and provide an additional embrittlement mechanism as the cladding temperature decreases below the ductile-to-brittle transition temperature (DBTT). Graphic and photographic details of the testing are

134

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

SciTech Connect

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.

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

1993-06-01T23:59:59.000Z

135

Evaluation and categorization of secondary system layup and cleanup practices for PWR plants  

Science Conference Proceedings (OSTI)

The EPRI Program S113-1, Evaluation of Secondary System Layup and Cleanup Proctices was established to study ways to minimize the transport of corrosion products into the secondary side PWR steam generators that occurs during plant startups following extended outages. As part of the EPRI Program, Task 200 objective was to identify and categorize the layup and cleanup practices now in use or proposed by utilities for PWR plants. The task study consisted of gathering information by conducting site visits to fourteen representative PWR plants in the USA, Europe and Japan, by conducting a search of the open literature, reviews of related EPRI Programs, and by evaluating the practices in terms of their potential effectiveness. The results show that about 30% of the plants attempt routine layup of secondary systems during outages and about 60% perform some form of system cleanup during the return to power following extended outages.

Cleary, W.F.

1982-12-01T23:59:59.000Z

136

EPRI/C-E PWR Safety Valve Test Report, Volume 10: Piping Structural Response Results  

Science Conference Proceedings (OSTI)

EPRI at the request of the PWR Utilities developed an overall program for the testing of PWR primary system safety and relief valves. This program was in response to NUREG 0578 Item 2.1.2 and NUREG 0737 Item II.D.l.A requirements.This report documents the results of safety valve testing performed as part of the overall program at the EPRI/C-E Valve Test Facility located at Combustion Engineering's Kreisinger Development Laboratory, Windsor, Connecticut. Seven safety valves ...

1983-04-01T23:59:59.000Z

137

Endurance Tests of Valves With Cobalt-Free Hardfacing Alloys: PWR Phase Final Report  

Science Conference Proceedings (OSTI)

Gate valves hard-faced with three galling-resistant iron-base alloys and a cobalt-base standard were subjected to demanding cycling conditions in an environment simulating typical PWR primary chemistry conditions. Extensive nondestructive and destructive examinations showed that all valves with iron-base trim performed better than the cobalt-base standard after 1900 operating cycles.

1992-05-01T23:59:59.000Z

138

Improvement of the Stress Corrosion Resistance of Alloy 718 in the PWR Environment  

Science Conference Proceedings (OSTI)

The costs associated with replacement of high-strength, nickel- base in-core components has led to efforts to improve corrosion resistance by various thermal, chemical and mechanical means. This report describes efforts designed to optimize the SCC resistance of alloy 718 in the PWR environment.

1996-08-01T23:59:59.000Z

139

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

E-Print Network (OSTI)

RIS�-M-2264 CONSTRUCTION OF PWR NUCLEAR CROSS SECTIONS FOR TRANSIENT CALCULATIONS. TEST OF THE ANTI PROGRAM AGAINST TWODIM Bjørn Thorlaksen Abstract. Nuclear cross sectxons for fuel assemblies of the more for Westinghouse reference plants 38 " B. Material number densities 41 " C. Nuclear cross sections for each fuel

140

Hydrazine Usage or Corrosion Control in PWR Plants With Powdered Resin Condensate Polishers  

Science Conference Proceedings (OSTI)

This report documents testing performed at North Anna-2 to obtain data for determining the optimal amount of hydrazine to use--and the optimal injection point--for oxygen control in PWR units with condensate polishing demineralizers. Conclusions and recommendations are presented.

1983-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "niagara mohawk pwr" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

Effect of Lithium Hydroxide on Zircaloy Corrosion in the Ringhals-3 PWR Plant  

Science Conference Proceedings (OSTI)

A predictive model for Zircaloy corrosion has been compared to Zircaloy oxide thickness measurements made after operation with elevated lithium hydroxide concentrations and pH at the Ringhals-3 PWR plant. Within the accuracy of the measurements and predictions, there was no significant effect of lithium hydroxide on Zircaloy corrosion.

1992-03-01T23:59:59.000Z

142

Materials Reliability Program: Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internals (MRP-232)  

Science Conference Proceedings (OSTI)

This report summarizes the aging management strategy development for Westinghouse and Combustion Engineering (CE) reactor internals. This report provides the technical basis for the aging management requirements of Westinghouse and CE reactor internals in the Pressurized Water Reactor (PWR) internals I&E guidelines (MRP-227-Rev. 0).

2008-12-22T23:59:59.000Z

143

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

Science Conference Proceedings (OSTI)

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.

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

2006-05-31T23:59:59.000Z

144

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

E-Print Network (OSTI)

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) ...

Liu, Jack S. H.

1979-01-01T23:59:59.000Z

145

Mechanics and Mechanisms of Environmentally Assisted Cracking of Alloys 132/182 in BWR and PWR Environments  

Science Conference Proceedings (OSTI)

This report documents research on the mechanics and mechanisms of environmentally assisted cracking of Alloys 132/182 in boiling water reactor (BWR) and pressurized water reactor (PWR) environments.

2004-10-18T23:59:59.000Z

146

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

E-Print Network (OSTI)

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 ...

Tsai, Chon-Kwo

1984-01-01T23:59:59.000Z

147

Materials Reliability Program: Review of Stress Corrosion Cracking of Alloys 182 and 82 in PWR Primary Water Service (MRP-220)  

Science Conference Proceedings (OSTI)

Since 1999, there have been several incidences involving primary water stress corrosion cracking (PWSCC) of Alloy 182/82 butt welds in pressurized water reactor (PWR) plants in the United States and abroad. These events resulted in unplanned or extended outages with associated economic costs. This report summarizes the available information on PWSCC of Alloy 182 and 82 weld metals observed in PWR primary circuit components up to the end of 2006. Relevant data from laboratory stress corrosion testing are ...

2007-10-29T23:59:59.000Z

148

Materials Reliability Program: Fracture Toughness Testing of Decommissioned PWR Core Internals Material Samples (MRP-160): Non-Proprietary Version  

Science Conference Proceedings (OSTI)

Pressurized water reactor (PWR) cores operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs requires detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel (internals) that are inherent to such conditions. This report contains the results of PWR environment fracture toughness testing of samples machined from decommissione...

2005-09-26T23:59:59.000Z

149

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

SciTech Connect

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

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

1993-07-01T23:59:59.000Z

150

Nuclear plant design and modification guidelines for PWR steam generator reliability  

Science Conference Proceedings (OSTI)

Operating experience gathered from PWR plant operation during the 1960's and 1970's has been incorporated into a series of design guidelines for secondary plant systems and steam generators. Specific guidelines included in this volume are: plant design for PWR steam generator inspection and nondestructive testing, revision 1; guidelines for design of steam generator blowdown systems, revision 1; plant design guidelines for layup and cleanup of steam, feedwater, and condensate systems, revision 1; design guidelines for plant secondary systems, revision 1 and plant design for steam generator replaceability, revision 1. The guidelines are intended to address those aspects of new plant design which will minimize corrosion damage to steam generators by controlling impurity ingress, facilitate steam generator nondestructive testing and provide for eventual replacement of steam generator if necessary. The guidelines, last revised in 1986, are primarily applicable to new plant construction, however, some of the guidelines may also be applicable to major backfits to existing plants.

Not Available

1991-09-01T23:59:59.000Z

151

Evaluation of secondary-system layup and cleanup practices and processes. Final report. [PWR  

Science Conference Proceedings (OSTI)

The study of PWR secondary system layup and cleanup practices was undertaken to evaluate current and proposed methods of corrosion product control associated with extended plant outages. The overall goal was to evaluate means for significantly minimizing the steam generator sludge burden. The study included a field survey of 14 representative PWR plants, an extensive literature search and an evaluation of corrosion product transport data. Recommendations for layup and cleanup system processes were derived from these practices and related information. Estimates of the potential benefits to be expected in the control of corrosion products by controlled layup environments during extended outages and by cleanup following such outages are provided. Cleanup during all, or most, phases of operation is indicated as being most beneficial. Layup and cleanup system process design information is also provided.

Cleary, W.F.

1983-04-01T23:59:59.000Z

152

Contain analysis of hydrogen distribution and combustion in PWR dry containments  

DOE Green Energy (OSTI)

Hydrogen transport and combustion in a PWR dry containment are analyzed using the CONTAIN code for a multi-compartment model of the Zion plant. The analysis includes consideration of both degraded core and full core meltdown accidents initiated by a small break LOCA. The importance of intercell flow mixing on distributions of gas composition and temperature in various compartments are evaluated. Thermal stratification and combustion behavior are discussed. 4 refs., 8 figs., 2 tabs.

Yang, J.W.; Nimnual, S.

1991-01-01T23:59:59.000Z

153

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

SciTech Connect

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.

Cardoni, Jeffrey

2010-11-01T23:59:59.000Z

154

An Assessment of PWR Water Chemistry Control in Advanced Light Water Reactors: U.S. EPR™  

Science Conference Proceedings (OSTI)

The Electric Power Research Institute (EPRI) developed the Water Chemistry Guidelines to support improved industry water chemistry operations. This report reviews the AREVA US EPR design to assess the applicability of the EPRI Pressurized Water Reactor (PWR) Primary and Secondary Water Chemistry Guidelines to that design. It is anticipated that the timely identification of any inconsistencies will allow EPRI and its member utilities to resolve them before the first US EPR plant begins operation.

2011-12-15T23:59:59.000Z

155

EPRI PWR Fuel Cladding Corrosion (PFCC) Model: Volume 2: Corrosion Theory and Rate Equation Development  

Science Conference Proceedings (OSTI)

The EPRI PWR Fuel Cladding Corrosion (PFCC) model has been developed to help utilities manage high burnup fuel cladding corrosion and hydriding issues. The model predicts the peak oxide thickness with 92 percent confidence of being within plus or minus 10 micrometers of the measured value, with a conservative bias of 7 micrometers when the metallurgical variables are well characterized. This volume documents the evolution of the rate equation for predicting Zircaloy cladding corrosion and the database us...

1997-03-04T23:59:59.000Z

156

Boric Acid Corrosion Guidebook, Revision 1: Managing Boric Acid Corrosion Issues at PWR Power Stations  

Science Conference Proceedings (OSTI)

Boric acid corrosion (BAC) represents a significant maintenance concern at many pressurized water reactor (PWR) plants because of the large number of potential leakage sources -- flanged joints, valve packing, mechanical seals, and fittings. This report compiles information that can help utility staff reduce the potential for leakage, properly and uniformly evaluate individual incidents, mitigate potential damage, and justify continued operation with leakage when appropriate. BAC does not represent a sig...

2001-11-01T23:59:59.000Z

157

An Assessment of PWR Water Chemistry in Advanced Light Water Reactors: US-APWR  

Science Conference Proceedings (OSTI)

The Electric Power Research Institute (EPRI) developed the Water Chemistry Guidelines to support improved industry water chemistry operations. This report reviews the Mitsubishi US-APWR design to assess the applicability of the EPRI Pressurized Water Reactor (PWR) Primary and Secondary Water Chemistry Guidelines to that design. It is anticipated that the timely identification of any inconsistencies will allow EPRI and its member utilities to resolve them before the first Mitsubishi Nuclear Energy Systems...

2012-01-31T23:59:59.000Z

158

An Assessment of PWR Water Chemistry Control in Advanced Light Water Reactors: APR1400  

Science Conference Proceedings (OSTI)

The Electric Power Research Institute (EPRI) developed the Water Chemistry Guidelines to support improved industry water chemistry operations. This report reviews the Korea Hydro & Nuclear Power (KHNP) APR1400 design to assess the applicability of the EPRI Pressurized Water Reactor (PWR) Primary (Volume 1) and Secondary (Volume 2) Water Chemistry Guidelines to that design. The timely identification of any inconsistencies and technical gaps will allow EPRI and its member utilities to resolve them ...

2012-12-13T23:59:59.000Z

159

Materials Reliability Program: Lessons Learned from PWR Thermal Fatigue Management Training (MRP-83)  

Science Conference Proceedings (OSTI)

In January 2001, The EPRI Materials Reliability Program (MRP) issued an Interim Guideline (MRP-24) for the management of thermal fatigue in non-isolable piping attached to reactor coolant piping in pressurized water reactor (PWR) plants (EPRI report 1000701). To assist utility personnel in understanding the potential for thermal fatigue in this piping, the MRP also conducted plant-specific workshops at plant sites. These workshops offered training on fatigue and fatigue cracking in non-isolable piping, a...

2002-12-05T23:59:59.000Z

160

Hot Cell Examination of ZIRLO PWR Fuel: Irradiated to 70 GWd/MTU  

Science Conference Proceedings (OSTI)

A set of eight Westinghouse pressurized water reactor (PWR) fuel rods from Dominion Generation's North Anna Power Station represented the lead exposure for ZIRLO cladding in a U.S. plant, with rod average burnups of 70 GWd/MTU. These rods, part of the original ZIRLO demonstration program, were reconstituted into a once-burned assembly and operated for a fourth 18-month cycle. Results from the Robust Fuel Program (RFP) and Westinghouse-sponsored poolside examination were reported previously (EPRI 1003216)...

2003-12-09T23:59:59.000Z

Note: This page contains sample records for the topic "niagara mohawk pwr" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

Materials Reliability Program: Experimental Study of Stress Corrosion Cracking Initiation in Austenitic Stainless Steels in Off-Normal Chemistry PWR Primary Water Environments (MRP-363)  

Science Conference Proceedings (OSTI)

PWR operating experience of Type 304 and 316 austenitic stainless steels and their L grade equivalents in PWR primary circuits has been generally excellent, but a recent review of all known incidents of stress corrosion cracking (SCC) of austenitic stainless steels exposed to PWR primary water environments identified a significant number of incidents that occurred in low flow or stagnant zones in dead leg situations where the primary water chemistry was probably contaminated by impurities. The ...

2013-11-27T23:59:59.000Z

162

Materials Reliability Program: Inspection and Evaluation Guidelines for Reactor Vessel Bottom-Mounted Nozzles in U.S. PWR Plants (MR P-206)  

Science Conference Proceedings (OSTI)

This report presents inspection and evaluation guidelines for reactor vessel bottom-mounted nozzles in U.S. pressurized water reactor (PWR) plants.

2009-03-23T23:59:59.000Z

163

Importance of thermal nonequilibrium considerations for the simulation of nuclear reactor LOCA transients. [PWR  

SciTech Connect

The purpose of this paper is to show the importance of considering thermal nonequilibrium effects in computer simulations of the refill and reflood portions of pressurized water reactor (PWR) loss-of-coolnat accident (LOCA) transients. Although RELAP4 assumes thermodynamic equilibrium between phases, models that account for the nonequilibrium phenomena associated with the mixing of subcooled emergency cooling water with steam and the superheating of vapor in the presence of liquid droplets have recently been incorporated into the code. Code calculated results, both with and without these new models, have been compared with experimental test data to assess the importance of including thermal nonequilibrium phenomena in computer code simulations.

Fischer, S.R.; Nelson, R.A.; Sullivan, L.H.

1980-01-01T23:59:59.000Z

164

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

Science Conference Proceedings (OSTI)

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

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

2012-07-01T23:59:59.000Z

165

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

Science Conference Proceedings (OSTI)

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

Biswas, D.; Rathbun, R.; Lee, Si Young [Westinghouse Savannah River Co., Aiken, SC (United States); Rosenthal, P. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

1993-12-31T23:59:59.000Z

166

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

Science Conference Proceedings (OSTI)

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

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

2000-10-03T23:59:59.000Z

167

EPRI/C-E PWR Safety Valve Test Report Volume 3: Test Results for Dresser Safety Valve Model 31739A  

Science Conference Proceedings (OSTI)

EPRI at the request of the PWR utilities developed an overall program for the testing of PWR primary system safety and relief valves. This program was in response to NUREG 0578 Item 2.1.2 and NUREG 0737 Item II.D.1.A requirements. This report documents the results of safety valve testing performed as part of the overall program at the EPRI/C-E Valve Test Facility located at Combustion Engineering's Kreisinger Development Laboratory, Windsor, Connecticut. Seven safety valves representative of those utiliz...

1983-02-01T23:59:59.000Z

168

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

SciTech Connect

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.

Shaver, Mark W.; Lanning, Donald D.

2010-02-01T23:59:59.000Z

169

Study of enhanced droplet cooling across grid spacer in LOCA reflood of PWR by LDA measurement  

SciTech Connect

An experimental investigation of droplet-vapor mist flow across a test grid spacer was conducted. The study sought to simulate the grid spacer enhanced droplet cooling under reflooding conditions in the loss-of-coolant accident (LOCA) of a pressurized water reactor (PWR). The study also provides a fundamentally improved understanding of the basic mechanisms concerning droplet dynamics and droplet-vapor mist flow heat transfer. An understanding of water droplet dynamics is necessary for explaining the sharp drop in cladding temperature observed immediately downstream of the grid spacer. A test channel that simulates the PWR reactor rod bundle was built to conduct both temperature measurements of the rod cladding and vapor flow, and droplet dynamics measurements before and after the grid spacer. The large droplets ( >1mm), which thermally are relatively inactive, are intercepted by the grid spacer and broken down into smaller, thermally more active, droplets (<200 /sup +/m). This investigation discusses droplet dynamics and heat transfer mechanisms across the grid spacer. It also provides a detailed discussion of the grid spacer's quenching behavior, the cooling downstream, and the droplet enhancement cooling, combined with the results of the full-length rod-bundle test.

Sheen, H.J.

1987-01-01T23:59:59.000Z

170

Impact of PWR spent fuel variations on TRU-fueled VHTRS  

E-Print Network (OSTI)

Several alternative strategies are being considered as spent nuclear fuel (SNF) management options. Transuranic nuclides (TRU) are responsible for the SNF long-term radiotoxicity beyond the first 500 years. One of the most viable approaches suggests creating new transmutation fuels containing TRUs for use in thermal and fast nuclear reactors. Irradiation of TRUs results in their transmutation and ultimate incineration by fission. The objective of this thesis is to analyze the impact of conventional PWR spent fuel variations on TRU-fueled Very High Temperature Reactor (VHTR) systems. This effort was focused on the prismatic core configuration. The 3D core models were created for use in calculations with the SCALE 5.1 code system. As part of the research effort, basic nuclear characteristics of TRUs were taken into consideration. The potential variations of PWR spent fuel compositions were modeled with the International Atomic Energy Agency (IAEA) Nuclear Fuel Cycle Simulation System, VISTA. The VHTR configurations with varying TRU compositions were analyzed assuming a single-batch core operation. Their performance was compared to the VHTR cases with low enriched uranium (LEU). The analysis shows that TRUs can be effectively utilized in the VHTR systems. The TRU-fueled VHTRs exhibit favorable performance characteristics.

Alajo, Ayodeji Babatunde

2007-12-01T23:59:59.000Z

171

Recommendations for Addressing Axial Burnup in the PWR Burnup Credit Analyses  

SciTech Connect

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

Wagner, J.C.

2002-10-23T23:59:59.000Z

172

Parametric Study of CHF Data, Volume 2: A Generalized Subchannel CHF Correlation for PWR and BWR Fuel Assemblies  

Science Conference Proceedings (OSTI)

This volume describes the development of a generalized subchannel critical heat flux (CHF) correlation for PWR and BWR fuel assemblies. The effects of nonuniform axial heat flux, cold walls, and grid spacers are discussed, and the correlation's performance is compared with a wide range of data.

1983-01-01T23:59:59.000Z

173

End-of-life destructive examinations of Zircaloy maximum depletion blanket fuel plates from the Shippingport PWR Core 2  

DOE Green Energy (OSTI)

Destructive examinations were performed on four Shippingport PWR Core 2 maximum fluence and depletion blanket plates for surface integrity, corrosion oxide thickness, and hydrogen absorption of the Zircaloy-4 cladding. The Shippingport PWR Core 2 operated for 23,360 effective full power hours (EFPH) (62,235 hot hours) at an average coolant temperature of 536{degrees}F (280{degrees}C) and a peak neutron flux of 0.6{times}10{sup 14}n/cm{sup 2}/s. The end-of-life examination program included measurements on three PWR-2 beta-quenched blanket fuel plates and one alpha-annealed blanket end plate. The examinations consisted of optical and scanning electron microscopy (SEM) inspections, direct metallographic oxide thickness measurements, and hydrogen extraction analyses on a joined element pair from the peak fluence (132{times}10{sup 20} n/cm{sup 2}), maximum depletion (13.5{times}10{sup 20} fissions/cc)PWR-2 blanket cluster.

Clayton, J.C.; Kammenzind, B.F.; Senio, P.; Sherman, J.

1993-10-01T23:59:59.000Z

174

Materials Reliability Program: Hot Cell Testing of Baffle/Former Bolts Removed from Two Lead PWR Plants  

Science Conference Proceedings (OSTI)

Irradiation-assisted stress corrosion cracking (IASCC) has been observed in core shroud baffle former bolts in pressurized water reactor (PWR) internals. This report describes hot cell testing results for bolts removed from one Westinghouse three-loop nuclear power plant, Farley Unit 1, and one two-loop plant, Point Beach Unit 2.

2001-11-05T23:59:59.000Z

175

O:ELECTRICPP-190.PDF  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

authority to grant Presidential permits for the construction of electric transmission lines at the U.S. authority to grant Presidential permits for the construction of electric transmission lines at the U.S. 1 international border was transferred from the FPC to the Department of Energy by Executive Order 12038. PRESIDENTIAL PERMIT NIAGARA MOHAWK POWER CORPORATION PERMIT NO. PP-190 I. BACKGROUND The Office of Fossil Energy (FE) of the Department of Energy (DOE) has the responsibility for implementing Executive Order (EO) 10485, as amended by EO 12038, which requires the issuance of Presidential permits for the construction, connection, operation, and maintenance of electric transmission facilities at the United States international border. In an application dated July 21, 1998, and amended October 29, 1998, Niagara Mohawk

176

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

SciTech Connect

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.

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

1985-04-01T23:59:59.000Z

177

Environmental Distribution of Petroleum Hydrocarbons at a Utility Service Center  

Science Conference Proceedings (OSTI)

This report presents the results of a field study at a utility service center located in western New York where a petroleum product had leaked into the subsurface over a number of years. The study was a tailored collaboration effort between the Electric Power Research Institute (EPRI) and the Niagara Mohawk Power Corporation, aimed at delineating the nature and extent of migration of the dissolved hydrocarbons. The information is of interest to many utilities as they develop and implement management prac...

1999-06-30T23:59:59.000Z

178

BIOMASS REBURNING - MODELING/ENGINEERING STUDIES  

SciTech Connect

This project is designed to develop engineering and modeling tools for a family of NO{sub x}control technologies utilizing biomass as a reburning fuel. During the eighth reporting period (July 1--September 26, 1999), Antares Group Inc, under contract to Niagara Mohawk Power Corporation, evaluated the economic feasibility of biomass reburning options for Dunkirk Station. This report includes summary of the findings; complete information will be submitted in the next Quarterly Report.

Vladimir Zamansky; Chris Lindsey

1999-10-29T23:59:59.000Z

179

Coal switch helps New York plants stay competitive  

Science Conference Proceedings (OSTI)

NRG Energy bought the Dunlook and Huntley Generating Stations in 1999 from Niagara Mohawk Power Corp. and has since then invested millions of dollars in converting them from bituminous coal to low sulphur Powder River Basin coal, combustion tuning and routine maintenance to help provide reliable stable-priced electricity to New York. The plants have reduced NOx, SO{sub 2} and particulate emissions. 1 photo.

Blankinship, S.

2009-04-15T23:59:59.000Z

180

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

Office of Scientific and Technical Information (OSTI)

WAPD-SC-545 WAPD-SC-545 HYDROGEN FLAMMABILITY DATA AND APPLICATION TO PWR LOSS-OF-COOLANT ACCIDENT CONTRACT A T - I M - G E N - H BETTIS PLANT PITTSBURGH, PENNSYLVANIA Operated for the U.S. ATOMIC ENERGY COMMISSION by WESTINGHOUSE ELECTRIC CORPORATION DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency Thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product,

Note: This page contains sample records for the topic "niagara mohawk pwr" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
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181

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

Science Conference Proceedings (OSTI)

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

Gilles Youinou; Andrea Alfonsi

2012-03-01T23:59:59.000Z

182

Application of American and French rules for the next Belgian PWR  

Science Conference Proceedings (OSTI)

The licensing practice in Belgium is evolving from the precedent compliance with the USNRC rules (as applied to the 4 last Belgian PWRs) to a more sophisticated approach applied to the next Belgian PWR (N8), which incorporates a mixed compliance with the USNRC or with French rules, depending on the equipment, the structure or the system considered. In this paper is presented the approach concerning the licensing rules applicable to N8. The following aspects are covered: rules applicable to the NSSS, rules applicable to the BOP (codes of design for systems and structures), rules applicable to the equipment (code of construction for mechanical and electrical components), and impact on the lay-out of the plant. Some examples of application of the methodology are given.

Roch, M.; Cavaco, A.

1988-01-01T23:59:59.000Z

183

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

DOE Green Energy (OSTI)

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

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

1987-05-21T23:59:59.000Z

184

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

DOE Green Energy (OSTI)

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.

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

185

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

Science Conference Proceedings (OSTI)

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.

MacArthur, A.D.

1983-05-01T23:59:59.000Z

186

Assessment of Reactivity Margins and Loading Curves for PWR Burnup Credit Cask Designs  

SciTech Connect

This report presents studies to assess reactivity margins and loading curves for pressurized water reactor (PWR) burnup-credit criticality safety evaluations. The studies are based on a generic high-density 32-assembly cask and systematically vary individual calculational (depletion and criticality) assumptions to demonstrate the impact on the predicted effective neutron multiplication factor, k{sub eff}, and burnup-credit loading curves. The purpose of this report is to provide a greater understanding of the importance of input parameter variations and quantify the impact of calculational assumptions on the outcome of a burnup-credit evaluation. This study should provide guidance to regulators and industry on the technical areas where improved information will most enhance the estimation of accurate subcritical margins. Based on these studies, areas where future work may provide the most benefit are identified. The report also includes an evaluation of the degree of burnup credit needed for high-density casks to transport the current spent nuclear fuel inventory. By comparing PWR discharge data to actinide-only based loading curves and determining the number of assemblies that meet the loading criteria, this evaluation finds that additional negative reactivity (through either increased credit for fuel burnup or cask design/utilization modifications) is necessary to accommodate the majority of current spent fuel assemblies in high-capacity casks. Assemblies that are not acceptable for loading in the prototypic high-capacity cask may be stored or transported by other means (e.g., lower capacity casks that utilize flux traps and/or increased fixed poison concentrations or high-capacity casks with design/utilization modifications).

Wagner, J.C.

2002-12-17T23:59:59.000Z

187

Generic Requirements Specification for Upgrading the Safety-Related Reactor Trip and Engineered Safety Features Actuation Systems in Westinghouse PWR Nuclear Power Plants  

Science Conference Proceedings (OSTI)

To address obsolescence concerns, a generic requirements specification for digital upgrades to existing reactor trip systems and engineered safety features actuation systems in a Westinghouse pressurized water reactor (PWR) was developed. System requirements are based on a 4-loop PWR with a solid-state protection system since this typifies the most advanced capability level. However, the specification is applicable to relay-based 2- and 3-loop plants where some or all of the advances in the newest solid-...

2001-10-19T23:59:59.000Z

188

Influence of Defect Kind and Size on Margins With Respect to Fast Fracture of Irradiated PWR Vessels: Joint EPRI-CRIEPI RPV Embrittl ement Studies  

Science Conference Proceedings (OSTI)

This report examines the effects of improved analysis techniques on the calculated margins against crack initiation and fast fracture in PWR vessels. A comparison of this work with previous studies showed that use of more in-depth analytic techniques may lead to more-realistic treatment of surface flaws. Such findings are particularly valuable to utilities developing nondestructive inspection techniques and evaluation methods for demonstrating the acceptability of an irradiated PWR vessel.

1994-04-26T23:59:59.000Z

189

Materials Reliability Program: Cluster Dynamics Prediction of Void Swelling in Austenitic Stainless Steels Under PWR Conditions (MRP-321)  

Science Conference Proceedings (OSTI)

This report describes the development of an improved physics-based model for the prediction of void swelling in austenitic stainless steel materials exposed to neutron irradiation under typical PWR operating conditions. This physics-based model is based on the cluster dynamics approach that models the evolution of defect clusters as a function of dose rate, accumulated dose, and temperature, with key model parameters calibrated using laboratory data. The physics-based void swelling model developed in thi...

2011-11-14T23:59:59.000Z

190

Program on Technology Innovation: A Preliminary Hybrid Model of Nickel Alloy Stress Corrosion Crack Propagation in PWR Primary Water Environments  

Science Conference Proceedings (OSTI)

Susceptibility to stress corrosion cracking (SCC) of nickel-based alloys in pressurized water reactor (PWR) primary water environments is a well-known issue in the nuclear power industry. Empirical models have been developed for this alloy/environment system, including the Scott model; the similar Materials Reliability Program, MRP-55 model for Alloy 600; and the MRP-115 model for weld metal. A model of the effects of dissolved hydrogen concentration, temperature, and the resulting electrochemical potent...

2008-12-17T23:59:59.000Z

191

Inhibition of IGA/SCC on Alloy 600 Surfaces Exposed to PWR Secondary Water: Volume 3: Precracking Model Boiler Tests  

Science Conference Proceedings (OSTI)

Intergranular attack/stress corrosion cracking (IGA/SCC) of Alloy 600 steam generator tubing in alkaline environments continues to be a serious problem. EPRI has an extensive program devoted to qualifying corrosion inhibitors for use in PWR steam generators. Researchers have identified several potential inhibitor materials in laboratory tests. This report documents testing of these potential inhibitors in model boilers contaminated with sodium hydroxide.

1998-12-10T23:59:59.000Z

192

Materials Reliability Program: Characterization of Decommissioned PWR Vessel Internals Material Samples - Tensile and SSRT Testing ( MRP-129)  

Science Conference Proceedings (OSTI)

Pressurized water reactor (PWR) cores operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs requires detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel (internals) subjected to such conditions. This project studied the effects of reactor service on the mechanical and corrosion properties of samples of baffle plate, fo...

2004-09-28T23:59:59.000Z

193

MHK Projects/Mohawk MHK Project | Open Energy Information  

Open Energy Info (EERE)

MHK Project MHK Project < MHK Projects Jump to: navigation, search << Return to the MHK database homepage Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":5,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"500px","height":"350px","centre":false,"title":"","label":"","icon":"File:Aquamarine-marker.png","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[]}

194

ANALYSIS OF STRESSES AND DEFLECTIONS IN TOP SUPPORT GRID, PWR REACTOR. Final Report  

SciTech Connect

The top grid of the PWR reactor core assembly is treated as a simply supported circular plate. The theory of plate is applied to the grid umder mechanical loads. The thermal stress problem is analyzed by treating the plate as under combined action of a laterally distributed load and forces in the middle plane of the plate. The load distribution is calculated from the temperature variation over the grid. The thermal stress problem then is equivalent to two problems: one, of bendimg of plate; and, the other, a plane stress problem. The theoretical formulation for plates under nonuniform heating is developed by neglecting the effect of uneven expansion in the direction perpendicular to the plane of the plate. In replacing the partial dffferential equations by difference equations, the latter are modified to take into account the change in tbickmess and spacing of the grid webs near the boundary. Twentythree difference equations for the twenty-three stations in one octant of the grid are obtained for each second order partial differential equation. The difference equations are solved by assuming that the twisting moments and shearing stresses in the plane of the grid vanish at the boundary. The stresses and deflections due to mechanical loads and thermal expansion are then superposed. (auth)

Yen, T.C.; Vining, R.E. Jr.

1957-06-01T23:59:59.000Z

195

MELCOR 1.8.2 assessment: Surry PWR TMLB` (with a DCH study)  

SciTech Connect

MELCOR is a fully integrated, engineering-level computer code, being developed at Sandia National Laboratories for the USNRC. This code models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As part of an ongoing assessment program, the MELCOR computer code has been used to analyze a station blackout transient in Surry, a three-loop Westinghouse PWR. Basecase results obtained with MELCOR 1.8.2 are presented, and compared to earlier results for the same transient calculated using MELCOR 1.8.1. The effects of new models added in MELCOR 1.8.2 (in particular, hydrodynamic interfacial momentum exchange, core debris radial relocation and core material eutectics, CORSOR-Booth fission product release, high-pressure melt ejection and direct containment heating) are investigated individually in sensitivity studies. The progress in reducing numeric effects in MELCOR 1.8.2, compared to MELCOR 1.8.1, is evaluated in both machine-dependency and time-step studies; some remaining sources of numeric dependencies (valve cycling, material relocation and hydrogen burn) are identified.

Kmetyk, L.N.; Cole, R.K. Jr.; Smith, R.C.; Summers, R.M.; Thompson, S.L.

1994-02-01T23:59:59.000Z

196

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

SciTech Connect

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

Alexandreanu, B.; Chopra, O. K.; Shack, W. J. (Nuclear Engineering Division); ( EVS); ( ESE)

2009-01-01T23:59:59.000Z

197

Key Issues for the control of refueling outage duration and costs in PWR Nuclear Power Plants  

SciTech Connect

For several years, EDF, within the framework of the CIDEM1 project and in collaboration with some German Utilities, has undertaken a detailed review of the operating experience both of its own NPP and of foreign units, in order to improve the performances of future units under design, particularly the French-German European Pressurized Reactor (EPR) project. This review made it possible to identify the key issues allowing to decrease the duration of refueling and maintenance outages. These key issues can be classified in 3 categories Design, Maintenance and Logistic Support, Outage Management. Most of the key issues in the design field and some in the logistic support field have been studied and could be integrated into the design of any future PWR unit, as for the EPR project. Some of them could also be adapted to current plants, provided they are feasible and profitable. The organization must be tailored to each country, utility or period: it widely depends on the power production environment, particularly in a deregulation context. (author)

Degrave, Claude [Electricite de France, EDF-SEPTEN, 12-14 avenue Dutrievoz 69628 Villeurbanne Cedex (France)

2002-07-01T23:59:59.000Z

198

Endurance tests of valves with cobalt-free hardfacing alloys: PWR phase  

SciTech Connect

Atomic Energy of Canada Limited (AECL) is conducting endurance tests on valves hard-faced with four cobalt-free alloys. The first phase of the program, in which PWR primary heat transport conditions were simulated in AECL's valve test loop, has been completed. The candidate alloys are NOREM 01, NOREM 04, EB 5183 and EVERIT 50. One valve with Stellite 6 trim served as the standard. Prior to loop testing, a baseline inaugural inspection was performed. During testing the loop was shutdown at approximately 500 cycle intervals, and the valves were disassembled for examination. The examinations included seat leak tests, profilometry, nondestructive inspection and finally destructive examination. Corrosion coupons in the loop were used to monitor any material loss due solely to corrosion mechanisms. This report summarizes the final examination results and discusses the relative performance of the candidate alloys. The results indicate that, based upon the sliding wear damage assessment and seat leakage test results, all the candidate alloys perform better than the Stellite 6 control sample. On the same basis, NOREM 04 and EB 5183 are the best of the candidate alloys, although there are only minor differences in performance among the four alloys.

Murphy, E.V.; Inglis, I. (Atomic Energy of Canada Ltd., Mississauga, ON (Canada))

1992-05-01T23:59:59.000Z

199

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

SciTech Connect

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.

Crowell, Shannon L.; Alzheimer, James M.

2011-08-08T23:59:59.000Z

200

Compatibility of PWR gasket and packing materials and resins with organic amines  

SciTech Connect

The objectives of this testing program were two-fold: (1) to examine the compatibility of morpholine and five other amines with several synthetic polymeric materials useful for gaskets and seals in pressurized water reactor (PWR) secondary cycles and (2) to examine the potential chemical degradation of ion exchange (IX) resins by morpholine and ethanolamine. The screening of the polymeric materials in the amines was performed by heating small samples of the materials in the amines for one week to one month. Interaction of the amines with the materials was accelerated by testing at elevated temperatures and at high amine concentrations. Two materials (Kalrez and EPDM) that are potentially useful in high-temperature and high-pressure steam systems were tested in morpholine solutions in sealed bombs at 260{degrees}C (500{degrees}). After heating in the aqueous amine solutions, changes in weight were measured and the samples were visually examined for physical changes, such as swelling or cracking. Selected materials underwent testing for hardness, elongation, and tensile strength after heating in morpholine for one month. This document provides the results of this testing program.

Keneshea, F.J.; Hobart, S.A. (Adams and Hobart Consulting Engineers, Fremont, CA (United States)); Camenzind, M.J. (Balazs Analytical Lab., Mountain View, CA (United States))

1992-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "niagara mohawk pwr" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

Modeling and design of a reload PWR core for a 48-month fuel cycle  

Science Conference Proceedings (OSTI)

The objective of this research was to use state-of-the-art nuclear and fuel performance packages to evaluate the feasibility and costs of a 48 calendar month core in existing pressurized water reactor (PWR) designs, considering the full range of practical design and economic considerations. The driving force behind this research is the desire to make nuclear power more economically competitive with fossil fuel options by expanding the scope for achievement of higher capacity factors. Using CASMO/SIMULATE, a core design with fuel enriched to 7{sup w}/{sub o} U{sup 235} for a single batch loaded, 48-month fuel cycle has been developed. This core achieves an ultra-long cycle length without exceeding current fuel burnup limits. The design uses two different types of burnable poisons. Gadolinium in the form of gadolinium oxide (Gd{sub 2}O{sub 3}) mixed with the UO{sub 2} of selected pins is sued to hold down initial reactivity and to control flux peaking throughout the life of the core. A zirconium di-boride (ZrB{sub 2}) integral fuel burnable absorber (IFBA) coating on the Gd{sub 2}O{sub 3}-UO{sub 2} fuel pellets is added to reduce the critical soluble boron concentration in the reactor coolant to within acceptable limits. Fuel performance issues of concern to this design are also outlined and areas which will require further research are highlighted.

McMahon, M.V.; Driscoll, M.J.; Todreas, N.E. [Massachusetts Inst. of Tech., Cambridge, MA (United States)

1997-05-01T23:59:59.000Z

202

Integrated Radiation Transport and Thermo-Mechanics Simulation of a PWR Assembly  

SciTech Connect

The Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) is focused on predicting the temperature and strain within a nuclear fuel assembly to evaluate the performance and safety of existing and advanced nuclear fuel bundles within existing and advanced nuclear reactors. AMPFuel was extended to include an integrated nuclear fuel assembly capability for (one-way) coupled radiation transport and nuclear fuel assembly thermo-mechanics. This capability is the initial step towards incorporating an improved predictive nuclear fuel assembly modeling capability to accurately account for source terms, such as the neutron flux distribution, coolant conditions, and assembly mechanical stresses, of traditional (single-pin) nuclear fuel performance simulation. AMPFuel was used to model an entire 17 x 17 Pressurized Water Reactor (PWR) fuel assembly with many of the features resolved in three dimensions (for thermo-mechanics and/or neutronics), including the fuel, gap, and cladding of each of the 264 fuel pins, the 25 guide tubes, top and bottom structural regions, and the upper and lower (neutron) reflector regions. The final full-assembly calculation was executed on Jaguar (Cray XT5) at the Oak Ridge Leadership Computing Facility using 40,000 cores in under 10 hours to model over 162 billion degrees of freedom for 10 loading steps.

Clarno, Kevin T [ORNL; Hamilton, Steven P [ORNL; Philip, Bobby [ORNL; Sampath, Rahul S [ORNL; Allu, Srikanth [ORNL; Berrill, Mark A [ORNL; Barai, Pallab [ORNL; Banfield, James E [ORNL

2012-01-01T23:59:59.000Z

203

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

SciTech Connect

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.

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

204

PWR core and spent fuel pool analysis using scale and nestle  

SciTech Connect

The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

Murphy, J. E.; Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee, Knoxville, TN 37996-2300 (United States); St Clair, R.; Orr, D. [Duke Energy, 526 S. Church St, Charlotte, NC 28202 (United States)

2012-07-01T23:59:59.000Z

205

Griffiss Air Force Base integrated resource assessment. Volume 3, Electric resource assessment  

Science Conference Proceedings (OSTI)

The US Air Force Air Combat Command (ACC) has tasked the US Department of Energy (DOE) Federal Energy Management Program (FEMP) to identify, evaluate, and assist in acquiring all cost-effective energy projects at Griffiss Air Force Base (AFB). FEMP, with support from the Pacific Northwest Laboratory (PNL), is designing this model program for federal customers served by the Niagara Mohawk Power Company. The program with Griffiss AFB will (1) identify and evaluate all cost-effective electric energy projects; (2) develop a schedule 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 them procure the necessary contractors to perform detailed audits and install the technologies. This report provides the results of the electric energy resource opportunity (ERO) assessments performed by PNL at one of Niagara Mohawk`s primary federal facilities, the ACC Griffiss AFB facility located near Rome, New York. The results of the analyses of EROs are presented in seven common energy end-use categories. A narrative description of each ERO provides information on the initial cost, energy and dollar savings; impacts on operations and maintenance (O&M); and, when applicable, a discussion of energy supply and demand, energy security, and environmental issues. The evaluation methodology and technical and cost assumptions are also described for each ERO. Summary tables present the operational performance of energy end-use equipment before and after the implementation of each ERO and the results of the life-cycle cost analysis indicating the net present value (NPV) and savings-to-investment ratio (SIR) of each ERO.

Armstrong, P.R.; Shankle, S.A.; Elliott, D.B.; Stucky, D.J.; Keller, J.M.; Wahlstrom, R.R.; Dagle, J.E.; Gu, A.Y.

1993-09-01T23:59:59.000Z

206

Addendum 1 to CSER 78-001 PWR Core 2 Blanket Fuel Storage Cell 4 221T building  

SciTech Connect

Irradiated pressurized water reactor (PWR) Core 2 (PWR-2) blanket fuel assemblies from the Shippingport PWR have been stored in the 221-T canyon water pool for twenty years. The fuel is in the form of small wafers of UO{sub 2}, which were initially natural enriched uranium (0.72% {sup 235}U). The uranium oxide wafers have a pyrolytic carbon coating, which prevents the fuel from reacting with a zircaloy-4 grid which provides structural strength and holds the wafers in place to form fuel plates. Thirty fuel plates comprise a sub-assembly which are held together by zircaloy-4 end plates. Two identical oxide fuel plate sub-assemblies are welded together to form a square structure with two zircaloy-4 extensions welded to the ends. Seventy-two PWR-2 assemblies are stored in the 221-T canyon water pool. Eight of these assemblies were irradiated in the center of the reactor core to an average burnup of 24,538 Mwd/MTU. The remaining assemblies had a burnup of 16,200 Mwd/MTU. These assemblies were placed in the canyon in 1978 and 1979 (WHC 1996). The original Criticality Safety Analysis Report (CSAR) (WHC 1990) analyzed the criticality safety of their storage and concluded that they were safe from a criticality standpoint. It was also mentioned in this CSAR that the assemblies were scheduled to be stored for twenty years. The Criticality Prevention Specification (CPS) for this storage configuration (RHO 1978), included in (WHC 1990), specifies that the fuel ''will be stored in Cell 4 up to 20 years'', and that ''no special handling or storage requirements for criticality control during interim storage up to 20 years'' were necessary. The purpose of this addendum is to extend the period of coverage for this material. The analysis examines zircaloy-clad fuel degradation and extends the permitted storage time by ten years for Shippingport Core 2 blanket fuel assemblies in the 221-T, Cell 4 storage pool.

GOLDBERG, H.J.

1999-12-03T23:59:59.000Z

207

Welding and Repair Technology Center: Boric Acid Attack of Concrete and Reinforcing Steel in PWR Fuel Handling Buildings  

Science Conference Proceedings (OSTI)

Spent fuel pool (SFP) leakage is common throughout the U.S. PWR fleet, with some plants experiencing leakage since early in plant life. The U.S. Nuclear Regulatory Commission (NRC) issued Information Notice 2004-05 describing leakage from the SFP at Salem Generating Station that migrated outside the building. The contamination was limited to the vicinity of the fuel handling building (FHB) and was contained and remediated within the confines of the protected area. It did not reach either underground aqui...

2012-05-14T23:59:59.000Z

208

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

E-Print Network (OSTI)

The nuclear waste currently generated in the United States is stored in spent fuel pools and dry casks throughout the country awaiting a permanent disposal solution. One efficient solution would be to remove the actinides from the waste and transmute these isotopes in a fast spectrum reactor. Currently this technology is unavailable on a commercial scale and a considerable amount of research and development is still required. An alternate solution is to reprocess and recycle the used fuel in thermal reactors, creating new fuel while reducing the amount of waste and its impact to the environment. This thesis examines the possibility of multi-recycling the transuranics (Pu, Np, Am, and Cm) in a standard pressurized water reactor (PWR). Two types of recycling strategies will be examined: one where Pu, Np, and Am are recycled (TRU-Cm) and a second where the previous isotopes as well as Cm are recycled (TRU+Cm). To offset the hardened neutron spectrum that results from the inclusion of the transuranics, a smaller fuel pin is employed to provide additional moderation. Computer simulations are used to model the in-reactor physics and long-term isotopic decay. Each fuel type is assessed based on the required U-235 enrichment, void coefficient, transuranic 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 enrichment, compared to the other multi-recycled fuel and, after a significant number of recycles, can result in the required enrichment to decrease. This fuel type also maintains a negative void coefficient for each recycle. The void coefficient of the fuel type without Cm becomes positive after the third cycle. The transmutation destruction of the two multi-recycled assemblies is less than that of a MOX assembly, but the transmutation efficiency of the multi-recycled assemblies exceeds the MOX assemblies. The radiotoxicity of both multi-recycled assemblies is significantly lower than the UOX and MOX with the TRU+Cm fuel being the lowest. When Curium is recycled only 28,000 years are required for the radiotoxicity of the waste to reach that of natural Uranium and when Cm is not recycled, the amount of time increases to 57,000 years.

Chambers, Alex

2011-08-01T23:59:59.000Z

209

In-Vessel Retention of Molten Core Debris in the Westinghouse AP1000 Advanced Passive PWR  

SciTech Connect

In-vessel retention (IVR) of molten core debris via external reactor vessel cooling is the hallmark of the severe accident management strategies in the AP600 passive PWR. The vessel is submerged in water to cool its external surface via nucleate boiling heat transfer. An engineered flow path through the reactor vessel insulation provides cooling water to the vessel surface and vents steam to promote IVR. For the 600 MWe passive plant, the predicted heat load from molten debris to the lower head wall has a large margin to the critical heat flux on the external surface of the vessel, which is the upper limit of the cooling capability. Up-rating the power of the passive plant from 600 to 1000 MWe (AP1000) significantly increases the heat loading from the molten debris to the reactor vessel lower head in the postulated bounding severe accident sequence. To maintain a large margin to the coolability limit for the AP1000, design features and severe accident management (SAM) strategies to increase the critical heat flux on the external surface of the vessel wall need to be implemented. A test program at the ULPU facility at University of California Santa Barbara (UCSB) has been initiated to investigate design features and SAM strategies that can enhance the critical heat flux. Results from ULPU Configuration IV demonstrate that with small changes to the ex-vessel design and SAM strategies, the peak critical heat flux in the AP1000 can be increased at least 30% over the peak critical heat flux predicted for the AP600 configuration. The design and SAM strategy changes investigated in ULPU Configuration IV can be implemented in the AP1000 design and will allow the passive plant to maintain the margin to critical heat flux for IVR, even at the higher power level. Continued testing for IVR phenomena is being performed at UCSB to optimize the AP1000 design and to ensure that vessel failure in a severe accident is physically unreasonable. (authors)

Scobel, James H.; Conway, L.E. [Westinghouse Electric Company LLC, P.O. Box 355, Pittsburgh, PA 15230-0355 (United States); Theofanous, T.G. [Center for Risk Studies and Safety, University of California Santa Barbara (United States)

2002-07-01T23:59:59.000Z

210

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

Science Conference Proceedings (OSTI)

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.

DOE

1997-04-01T23:59:59.000Z

211

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

Science Conference Proceedings (OSTI)

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)

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

212

Simulation of German PKL refill/reflood experiment K9A using RELAP4/MOD7. [PWR  

Science Conference Proceedings (OSTI)

This paper describes a RELAP4/MOD7 simulation of West Germany's Kraftwerk Union (KWU) Primary Coolant Loop (PKL) refill/reflood experiment K9A. RELAP4/MOD7, a best-estimate computer program for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This study was the first major simulation using RELAP4/MOD7 since its release by the Idaho National Engineering Laboratory (INEL). The PKL facility is a reduced scale (1:134) representation of a typical West German four-loop 1300 MW pressurized water reactor (PWR). A prototypical scale of the total volume to power ratio was maintained. The test facility was designed specifically for an experiment simulating the refill/reflood phase of a Loss-of-Coolant Accident (LOCA).

Hsu, M.T.; Davis, C.B.; Behling, S.R.

1981-11-01T23:59:59.000Z

213

Materials Reliability Program: Determination of Crack Growth Rates for Alloy 82 at Low K Values Under PWR Primary Water Environment: 2011 Interim Report (MRP-337)  

Science Conference Proceedings (OSTI)

Crack propagation experiments, which were performed in the past on nickel-based materials under pressurized water reactor (PWR) primary water environments, have left some open questions that need to be answered. In particular, no crack growth rate (CGR) data for control rod drive mechanism (CRDM) nozzle materials are available at low stress intensity (K) values (K 15 MPam). This interim report summarizes the work done during 2011 on a cooperative project to generate CGR data at low K values for alloy 82 ...

2012-04-30T23:59:59.000Z

214

Inhibition of IGA/SCC on Alloy 600 Surfaces Exposed to PWR Secondary Water: Volume 2: Titanium and Cerium Acetate Model Boiler Testi ng  

Science Conference Proceedings (OSTI)

EPRI has devoted an extensive program to qualifying corrosion inhibitors for use in PWR steam generators. This report addresses one phase of model boiler testing using mill-annealed alloy 600 tubing with drilled-hole carbon steel tube support plate simulators in caustic environments. In two tests, investigators added inorganic inhibitors to the caustic environment. In another test, they exposed alloy 600 tubing to an acidic environment high in sulfates then to a caustic environment. Nondestructive and de...

1998-06-30T23:59:59.000Z

215

Design, Operation, and Performance Data for High Burnup PWR Fuel from the H. B. Robinson Plant for Use in the NRC Experimental Progr am at Argonne National Laboratory  

Science Conference Proceedings (OSTI)

This report presents the background information -- design, irradiation history, and performance data -- for twelve high-burnup pressurized water reactor (PWR) fuel rods that are being provided to the U.S. Nuclear Regulatory Commission (NRC) for use in experiments designed to study the response of highly irradiated fuel to transient accidents and long-term storage conditions. This information will establish the starting conditions needed to correctly interpret future experimental results.

2001-05-04T23:59:59.000Z

216

Materials Reliability Program: Characterizations of Type 316 Cold-Worked Stainless Steel Highly Irradiated Under PWR Operating Condi tions (MRP-73)  

Science Conference Proceedings (OSTI)

Irradiation-induced material degradations such as irradiation-assisted stress corrosion cracking (IASCC), irradiation-induced void swelling, and irradiation-caused embrittlement have been observed in core internals components in pressurized water reactors (PWRs). This report describes hot cell testing and characterization of a bottom-mounted instrument tube (flux thimble) that was exposed in an operating PWR for about 23 years, providing valuable data for assessing radiation effects in PWRs.

2002-08-26T23:59:59.000Z

217

Program on Technology Innovation: Hybrid Models of Stress Corrosion Crack Propagation for Nickel Alloy Welds in Low-Electrochemical Potential (ECP) Pressurized Water Reactor (PWR) Primary Water Environments  

Science Conference Proceedings (OSTI)

EPRI has developed hybrid models of pressurized water reactor (PWR) primary water stress corrosion cracking (PWSCC) in nickel alloy welds.  These models are able to account for differences in tensile properties of each heat, applied stress intensity factor, dissolved hydrogen, water temperature, and the increase in local strain rate caused by the moving crack. The new models show promise for reducing uncertainty in predicting PWSCC for nickel alloy welds by a statistically and practically ...

2012-10-30T23:59:59.000Z

218

Materials Reliability Program: Boric Acid Corrosion Guidebook, Revision 2: Managing Boric Acid Corrosion Issues at PWR Power Station s (MRP-058, Rev 2)  

Science Conference Proceedings (OSTI)

Boric acid corrosion (BAC) represents a significant maintenance concern at many pressurized water reactor (PWR) plants because of the large number of potential leakage sourcesflanged joints, valve packing, mechanical seals, and fittings. This report compiles information that can help utility staff reduce the potential for leakage, properly and uniformly evaluate individual incidents, mitigate potential damage, and justify continued operation with leakage when appropriate. BAC does not represent a signifi...

2012-07-11T23:59:59.000Z

219

Materials Reliability Program: Characterization of Type 316 Cold Worked Stainless Steel Highly Irradiated Under PWR Operating Conditions (International IASCC Advisory Committee Phase 3 Program Final Report) (MRP-214)  

Science Conference Proceedings (OSTI)

Various types of irradiation-induced material degradation such as irradiation-assisted stress corrosion cracking (IASCC), irradiation-induced void swelling, and irradiation-caused embrittlement have been observed in core internals components in pressurized water reactors (PWR). This report describes hot cell testing and characterization of bottom-mounted instrument tubes (flux thimble) that were exposed in operating PWRs for about 10 to 20 effective full power years (EFPY), providing valuable data for as...

2007-09-06T23:59:59.000Z

220

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

SciTech Connect

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.

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

2012-10-01T23:59:59.000Z

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221

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

Science Conference Proceedings (OSTI)

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.

Hermann, O.W.

2000-02-01T23:59:59.000Z

222

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

Science Conference Proceedings (OSTI)

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

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

1991-01-01T23:59:59.000Z

223

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

Science Conference Proceedings (OSTI)

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

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

1991-12-31T23:59:59.000Z

224

Record of Decision on the Disposal of the S3G and D1G Prototype Reactor Plants  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

5 5 Federal Register / Vol. 63, No. 18 / Wednesday, January 28, 1998 / Notices owned by Basin Electric Power Cooperative, Bonneville Power Administration, Citizens Utilities, Detroit Edison Company, Eastern Maine Electric Cooperative, Joint Owners of the Highgate Project, Maine Electric Power Company, Maine Public Service Company, Minnesota Power and Light Company, Minnkota Power Cooperative, New York Power Authority, Niagara Mohawk Power Corporation, Northern States Power, and Vermont Electric Transmission Company. Each of these transmission facilities, as more fully described in the application, has previously been authorized by a Presidential permit issued pursuant to Executive Order 10485, as amended. Procedural Matters Any persons desiring to become a party to this proceeding or to be heard

225

HTS Cable Projects  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Superconductivity Superconductivity Partnerships with Industry ANL Air Liquide DOE Golden LANL AEP ORNL Nexans Niagara Mohawk Super Power American Superconductor NYSERDA BOC Praxair W ? tion systems. This is the most the nation. W superconductivity? HTS Cable Projects www.oe.energy.gov Phone: 202 \ 586-1411 Office of Electricity Delivery and Energy Reliability, OE-1 U.S. Department of Energy - 1000 Independence Avenue, SW - Washington, DC 20585. Plugging America Into the Future of Power "A National Effort to Introduce New Technology into

226

Comparison of intergrated coal gasification combined cycle power plants with current and advanced gas turbines  

Science Conference Proceedings (OSTI)

Two recent conceptual design studies examined ''grass roots'' integrated gasification-combined cycle (IGCC) plants for the Albany Station site of Niagara Mohawk Power Corporation. One of these studies was based on the Texaco Gasifier and the other was developed around the British Gas Co.-Lurgi slagging gasifier. Both gasifiers were operated in the ''oxygen-blown'' mode, producing medium Btu fuel gas. The studies also evaluated plant performance with both current and advanced gas turbines. Coalto-busbar efficiencies of approximately 35 percent were calculated for Texaco IGCC plants using current technology gas turbines. Efficiencies of approximately 39 percent were obtained for the same plant when using advanced technology gas turbines.

Banda, B.M.; Evans, T.F.; McCone, A.I.; Westisik, J.H.

1984-08-01T23:59:59.000Z

227

BIOMASS REBURNING - MODELING/ENGINEERING STUDIES  

DOE Green Energy (OSTI)

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

Vladimir Zamansky; Chris Lindsey; Vitali Lissianski

2000-01-28T23:59:59.000Z

228

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

Science Conference Proceedings (OSTI)

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)

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

229

Materials Reliability Program: Determination of Crack Growth Rates for Alloy 82 at Low K Values Under PWR Primary Water Environment (MRP-256)  

Science Conference Proceedings (OSTI)

Crack propagation experiments, which were performed in the past on nickel-based materials in a PWR primary water environment, have left some open questions that need to be answered. In particular, no crack growth rate (CRD) data for control rod driving mechanism (CRDM) nozzle materials are available at low stress intensity (K) values (K 15 MPam). This interim report describes the planning and first stages of a cooperative project to generate crack growth data under low K values for alloy 82 weld metal.

2008-12-23T23:59:59.000Z

230

Graphical and tabular summaries of decay characteristics for once-through PWR, LMFBR, and FFTF fuel cycle materials. [Spent fuel, high-level waste fuel can scrap  

SciTech Connect

Based on the results of ORIGEN2 and a newly developed code called ORMANG, graphical and summary tabular characteristics of spent fuel, high-level waste, and fuel assembly structural material (cladding) waste are presented for a generic pressurized-water reactor (PWR), a liquid-metal fast breeder reactor (LMFBR), and the Fast Flux Test Facility (FFTF). The characteristics include radioactivity, thermal power, and toxicity (water dilution volume). Given are graphs and summary tables containing characteristic totals and the principal nuclide contributors as well as graphs comparing the three reactors for a single material and the three materials for a single reactor.

Croff, A.G.; Liberman, M.S.; Morrison, G.W.

1982-01-01T23:59:59.000Z

231

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

Science Conference Proceedings (OSTI)

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.

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

1980-01-01T23:59:59.000Z

232

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

Open Energy Info (EERE)

0944999°, -79.0567111° 0944999°, -79.0567111° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":43.0944999,"lon":-79.0567111,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

233

Niagara Falls, NY Natural Gas Imports by Pipeline from Canada  

Annual Energy Outlook 2012 (EIA)

GA Lake Charles, LA LNG Imports from Indonesia Lake Charles, LA LNG Imports from Malaysia Gulf Gateway, LA Lake Charles, LA LNG Imports from Nigeria Cove Point, MD Elba...

234

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

Office of Legacy Management (LM)

Site Fairfield Site Falls City Site Fernald Preserve Gasbuggy Site General Atomics Geothermal Gnome-Coach Site Grand Junction Sites Granite City Site Green River Site Gunnison...

235

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

Open Energy Info (EERE)

3119496°, -78.7476208° 3119496°, -78.7476208° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":43.3119496,"lon":-78.7476208,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

236

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

U.S. Energy Information Administration (EIA)

Decade Year-0 Year-1 Year-2 Year-3 Year-4 Year-5 Year-6 Year-7 Year-8 Year-9; 1990's: NA: NA: 2000's: NA: 2.49: 5.04: 6.77: 6.99----- 2010's--4.76: 4.08-

237

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

U.S. Energy Information Administration (EIA)

Year Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec; 2011: 734: 660: 860: 860: 194: 307: 295: 1,107: 376: 151: 415: 576: 2012: 583: 468: 175: 58: 8,823: 13,281: 2013 ...

238

MHK Projects/Niagara Community | Open Energy Information  

Open Energy Info (EERE)

Community Community < MHK Projects Jump to: navigation, search << Return to the MHK database homepage Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":5,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"500px","height":"350px","centre":false,"title":"","label":"","icon":"File:Aquamarine-marker.png","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":43.0643,"lon":-79.0278,"alt":0,"address":"","icon":"http:\/\/prod-http-80-800498448.us-east-1.elb.amazonaws.com\/w\/images\/7\/74\/Aquamarine-marker.png","group":"","inlineLabel":"","visitedicon":""}]}

239

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

Science Conference Proceedings (OSTI)

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)

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

240

Natural-gas-fired CC unit holds NO[sub x] emissions below 9. 0 ppm  

Science Conference Proceedings (OSTI)

This article describes the East Syracuse generating plant, one of first commercial stations to include LM6000 gas turbines, designed to solve noise and emissions problems. This natural-gas-fired, combined-cycle cogeneration facility provides 97 MW of power to Niagara Mohawk Power Corp and up to 80,000 lb/hr of process steam to a nearby Bristol-Myers Squibb Co plant. The plant's original design had contemplated a base-loaded facility. This stemmed from the original power sales agreement with Niagara Mohawk Power Corp. Flexibility of original design proved advantageous to the East Syracuse (NY) plant when, during the latter stages of construction, the original agreement was renegotiated into a schedulable'' contract. The new agreement now in force, providing for limited dispatch of one of the two gas turbines, is designed around other pre-existing project agreements. Design flexibility and the choice of two gas turbines made the plant capable of meeting dispatch requirements with only minor modifications of plant design and staffing.

Grunbeck, G.

1994-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "niagara mohawk pwr" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

Coal Feed Pump- PWR  

E-Print Network (OSTI)

The goal of the Gasification Program is to reduce the cost of electricity, while increasing power plant availability and efficiency, and maintaining the highest environmental standards “Federal support of scientific R&D is critical to our economic competitiveness“

Jenny B. Tennant; Dr. Steven Chu; Secretary Of Energy; Coe Reduction; Coe Decrease; Advanced Gasification; Low-rank Coal

2010-01-01T23:59:59.000Z

242

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

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.

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

2000-09-01T23:59:59.000Z

243

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

Science Conference Proceedings (OSTI)

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.

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

2011-10-01T23:59:59.000Z

244

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

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.

Ham, Y S; Sitaraman, S

2008-12-24T23:59:59.000Z

245

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

Science Conference Proceedings (OSTI)

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.

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

2008-05-05T23:59:59.000Z

246

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

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.

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

1982-05-01T23:59:59.000Z

247

Materials Reliability Program: Assessment of the Current Status and Completeness of Work on Inner and Outer Diameter Stress Corrosion Cracking of Austenitic Stainless Steels in PWR Plants (MRP-352)  

Science Conference Proceedings (OSTI)

Field experience with austenitic stainless steel in operating pressurized water reactors (PWRs) has, in general, been good, with a relatively small number of failures due to stress corrosion cracking (SCC) observed worldwide. Nevertheless, the number and nature of these failures are not insignificant and could potentially become more important as the age of the existing PWR fleet increases. In light of this, it has been identified that an ongoing focused research and plant management program is ...

2013-03-31T23:59:59.000Z

248

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

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.

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

2006-10-20T23:59:59.000Z

249

Application of MELCOR Code to a French PWR 900 MWe Severe Accident Sequence and Evaluation of Models Performance Focusing on In-Vessel Thermal Hydraulic Results  

Science Conference Proceedings (OSTI)

In the ambit of the Severe Accident Network of Excellence Project (SARNET), funded by the European Union, 6. FISA (Fission Safety) Programme, one of the main tasks is the development and validation of the European Accident Source Term Evaluation Code (ASTEC Code). One of the reference codes used to compare ASTEC results, coming from experimental and Reactor Plant applications, is MELCOR. ENEA is a SARNET member and also an ASTEC and MELCOR user. During the first 18 months of this project, we performed a series of MELCOR and ASTEC calculations referring to a French PWR 900 MWe and to the accident sequence of 'Loss of Steam Generator (SG) Feedwater' (known as H2 sequence in the French classification). H2 is an accident sequence substantially equivalent to a Station Blackout scenario, like a TMLB accident, with the only difference that in H2 sequence the scram is forced to occur with a delay of 28 seconds. The main events during the accident sequence are a loss of normal and auxiliary SG feedwater (0 s), followed by a scram when the water level in SG is equal or less than 0.7 m (after 28 seconds). There is also a main coolant pumps trip when {delta}Tsat < 10 deg. C, a total opening of the three relief valves when Tric (core maximal outlet temperature) is above 603 K (330 deg. C) and accumulators isolation when primary pressure goes below 1.5 MPa (15 bar). Among many other points, it is worth noting that this was the first time that a MELCOR 1.8.5 input deck was available for a French PWR 900. The main ENEA effort in this period was devoted to prepare the MELCOR input deck using the code version v.1.8.5 (build QZ Oct 2000 with the latest patch 185003 Oct 2001). The input deck, completely new, was prepared taking into account structure, data and same conditions as those found inside ASTEC input decks. The main goal of the work presented in this paper is to put in evidence where and when MELCOR provides good enough results and why, in some cases mainly referring to its specific models (candling, corium pool behaviour, etc.) they were less good. A future work will be the preparation of an input deck for the new MELCOR 1.8.6. and to perform a code-to-code comparison with ASTEC v1.2 rev. 1. (author)

De Rosa, Felice [ENEA, Italian National Agency for New Technologies, Energy and the Environment (Italy)

2006-07-01T23:59:59.000Z

250

The role of Hydrogen and Creep in Intergranular Stress Corrosion Cracking of Alloy 600 and Alloy 690 in PWR Primary Water Environments ? a Review  

DOE Green Energy (OSTI)

Intergranular attack (IGA) and intergranular stress corrosion cracking (IGSCC) of Alloy 600 in PWR steam generator environment has been extensively studied for over 30 years without rendering a clear understanding of the essential mechanisms. The lack of understanding of the IGSCC mechanism is due to a complex interaction of numerous variables such as microstructure, thermomechanical processing, strain rate, water chemistry and electrochemical potential. Hydrogen plays an important role in all these variables. The complexity, however, significantly hinders a clearer and more fundamental understanding of the mechanism of hydrogen in enhancing intergranular cracking via whatever mechanism. In this work, an attempt is made to review the role of hydrogen based on the current understanding of grain boundary structure and chemistry and intergranular fracture of nickel alloys, effect of hydrogen on electrochemical behavior of Alloy 600 and Alloy 690 (e.g. the passive film stability, polarization behavior and open-circuit potential) and effect of hydrogen on PWSCC behavior of Alloy 600 and Alloy 690. Mechanistic studies on the PWSCC are briefly reviewed. It is concluded that further studies on the role of hydrogen on intergranular cracking in both inert and primary side environments are needed. These studies should focus on the correlation of the results obtained at different laboratories by different methods on materials with different metallurgical and chemical parameters.

Rebak, R B; Hua, F H

2004-07-12T23:59:59.000Z

251

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

SciTech Connect

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

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

2012-11-01T23:59:59.000Z

252

MOTION TO INTERVENE OF THE NEW YORK TRANSMISSION OWNERS PP-230-4 |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

MOTION TO INTERVENE OF THE NEW YORK TRANSMISSION OWNERS PP-230-4 MOTION TO INTERVENE OF THE NEW YORK TRANSMISSION OWNERS PP-230-4 MOTION TO INTERVENE OF THE NEW YORK TRANSMISSION OWNERS PP-230-4 Pursuant to Rules 212 and 214 of the Rules of Practice and Procedure, 18 C.F.R. §§ 385.212 and 385.214 (2010), Central Hudson Gas & Electric Corporation, Consolidated Edison Company of New York, Inc., Long Island Power Authority, New York Power Authority, New York State Electric & Gas Corporation, Niagara Mohawk Power Corporation d/b/a National Grid, Orange and Rockland Utilities, Inc., and Rochester Gas and Electric Corporation (referred to herein as the "New York Transmission Owners"), individually and collectively move to intervene in the above-captioned proceeding and request an opportunity to comment on International Transmission Company's d/b/a/ ITCTransmission

253

O:ELECTRICEA-148-A.PDF  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

8-A 8-A I. BACKGROUND Exports of electricity from the United States to a foreign country are regulated and require authorization under section 202(e) of the Federal Power Act (FPA) (16 U.S.C. §824a(e)). On August 13, 1997, the Office of Fossil Energy (FE) of the Department of Energy (DOE) authorized Aquila Energy Marketing Corporation (AEM) to transmit electric energy from the United States to Canada as a power marketer using the international electric transmission facilities of Basin Electric Power Cooperative, Bonneville Power Administration, Citizens Utilities, Detroit Edison, Eastern Maine Electric Cooperative, Joint Owners of the Highgate Project, Maine Electric Power Company, Maine Public Service Company, Minnesota Power, Inc., Minnkota Power, New York Power Authority, Niagara Mohawk Power Corp., Northern States

254

Page not found | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

51 - 560 of 31,917 results. 51 - 560 of 31,917 results. Download WA_96_018_NIAGARA_MOHAWK__POWER_CORPORATION_Waiver_of_U.S._a.pdf http://energy.gov/gc/downloads/wa96018niagaramohawkpowercorporationwaiverofusapdf Download WA_00_008_PLUG_POWER_Waiver_of_Patent_Rights_in_Performance_.pdf http://energy.gov/gc/downloads/wa00008plugpowerwaiverofpatentrightsinperformancepdf Download WA_06_015_PPG_INDUSTRIES_Waiver_of_Patent_Rights_Under_a_DOE.pdf http://energy.gov/gc/downloads/wa06015ppgindustrieswaiverofpatentrightsunderadoepdf Download WA_1995_001_US_AUTO_MATERIALS_PARTNERSHIPS_Waiver_of_Patent_.pdf http://energy.gov/gc/downloads/wa1995001usautomaterialspartnershipswaiverofpatentpdf Download WA_00_005_GENERAL_ELECTRIC_Waiver_of_Government_US_and_Forei.pdf http://energy.gov/gc/downloads/wa00005generalelectricwaiverofgovernmentusandforeipdf

255

Wethersfield Wind Power Wind Farm | Open Energy Information  

Open Energy Info (EERE)

Wethersfield Wind Power Wind Farm Wethersfield Wind Power Wind Farm Facility Wethersfield Wind Power Sector Wind energy Facility Type Commercial Scale Wind Facility Status In Service Owner Enel North America Developer Western NY Wind Power Partners Energy Purchaser Niagara Mohawk Location WY County NY Coordinates 42.667741°, -78.219803° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":42.667741,"lon":-78.219803,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

256

Nuclear Regulatory Commission issuances, May 1993. Volume 37, No. 5  

SciTech Connect

This report contains the issuances received during the specified period (May 1993) from the Commission (CLI), the Atomic Safety and Licensing Boards (LBP), the Administrative Law Judges (ALJ), the Directors` Decisions (DD), and the Denials of Petitions for Rulemaking (DPRM). The summaries and headnotes preceding the opinions reported herein are not deemed a part of these opinions or have any independent legal significance. Contents of this document include an Issuance of the Nuclear Regulatory Commission with respect to the Sacramento Municipal Utility District and Issuances of Directors` Decisions concerning the Interstate Nuclear Service Corporation; Niagara Mohawk Power Corporation; and Texas Utilities Electric Company, et al. and All Nuclear Power Plants with Thermo-Lag Fire Barriers.

Not Available

1993-05-01T23:59:59.000Z

257

Automation uses common data base. [Power Resource Optimization by Electronics (PROBE)  

SciTech Connect

Supervisory-control and data-acquisition (SCADA) systems are not new to electric utilities, but the extension of this technology into substation and distribution-feeder automation has been slow. General Electric Co. and Commonwealth Edison Co. have installed a substation/distribution-automation research system at the latter's LaGrange Park substation to demonstrate feasibility and gain field experience. Niagara Mohawk Power Corp has recently joined in evaluating the functional specifications, in planning and in evaluating field tests, and is sharing in the funding. The project has been named PROBE (power resource optimization by electronics), and its key concept is use of a common data base for substation and distribution information. The project will include three phases of which the initial trial evaluation at LaGrange Park is identified as PROBE-1.

1976-09-01T23:59:59.000Z

258

BIOMASS REBURNING - MEDELING/ENGINEERING STUDIES  

SciTech Connect

This project is designed to develop engineering and modeling tools for a family of NO{sub x} control technologies utilizing biomass as a reburning fuel. During the seventh reporting period (April 1--June 30, 1999), no information was received at EER on scheduled FETC R&D group's project activities. EER activities were on hold due to the pending purchase of the Niagara Mohawk's Dunkirk Station, a target demonstration site in this program, and then by the actual purchase of the Station by NRG. This report includes information about the current project status, recently submitted to NRG for soliciting their interest to proceed with biomass reburn demonstration, and notes on alternative demonstrative partners.

Vladimir Zamansky; Michael Booth

1999-07-30T23:59:59.000Z

259

O:ELECTRICORDERSEA-168-B.PDF  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

8-B 8-B I. BACKGROUND Exports of electricity from the United States to a foreign country are regulated and require authorization under section 202(e) of the Federal Power Act (FPA) (16 U.S.C. §824a(e)). On February 25, 1998, the Office of Fossil Energy (FE) of the Department of Energy (DOE) authorized PG&E Energy Trading-Power, L.P. (PGET-Power) to transmit electric energy as a power marketer from the United States to Canada using the international transmission facilities of Detroit Edison, Minnesota Power, Niagara Mohawk and New York Power Authority. On August 25, 1998, in Order EA-168-A, DOE amended PGET-Power's electricity export authorization to add the remaining major transmission interconnections with Canada. That two- year authorization expired on August 25, 2000. On July 6, 2000, PGET-Power filed an

260

DOE/OR/20722-133 POST-REMEDIAL ACTION REPORT FOR THE NIAGARA...  

Office of Legacy Management (LM)

some of the radioactive materials stored at the NFSS were subject to water and wind erosion over the years. , As a result, radioactive materials migrated off-site, chiefly...

Note: This page contains sample records for the topic "niagara mohawk pwr" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
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261

MERCURY CONTROL USING SORBALIME AT AMERICAN REF-FUEL'S NIAGARA FALLS FACILITY  

E-Print Network (OSTI)

), and nitrogen oxides (NOx), respectively. The ESP, FGD, and SCR change the composition of the flue gas to the process following the FGD, just before the flue gases go to the stack, as shown in Figure 2. The balance + CO2 C2H4OHNH3 + + HCO3 - (1) The remaining flue gases are washed to remove any residual MEA

Columbia University

262

Case study of the propagation of a small flaw under PWR loading conditions and comparison with the ASME code design life. Comparison of ASME Code Sections III and XI  

SciTech Connect

A cooperative study was performed by EG and G Idaho, Inc., and Oak Ridge National Laboratory to investigate the degree of conservatism and consistency in the ASME Boiler and Pressure Vessel Code Section III fatigue evaluation procedure and Section XI flaw acceptance standards. A single, realistic, sample problem was analyzed to determine the significance of certain points of criticism made of an earlier parametric study by staff members of the Division of Engineering Standards of the Nuclear Regulatory Commission. The problem was based on a semielliptical flaw located on the inside surface of the hot-leg piping at the reactor vessel safe-end weld for the Zion 1 pressurized-water reactor (PWR). Two main criteria were used in selecting the problem; first, it should be a straight pipe to minimize the computational expense; second, it should exhibit as high a cumulative usage factor as possible. Although the problem selected has one of the highest cumulative usage factors of any straight pipe in the primary system of PWRs, it is still very low. The Code Section III fatigue usage factor was only 0.00046, assuming it was in the as-welded condition, and fatigue crack-growth analyses predicted negligible crack growth during the 40-year design life. When the analyses were extended past the design life, the usage factor was less than 1.0 when the flaw had propagated to failure. The current study shows that the criticism of the earlier report should not detract from the conclusion that if a component experiences a high level of cyclic stress corresponding to a fatigue usage factor near 1.0, very small cracks can propagate to unacceptable sizes.

Yahr, G.T.; Gwaltney, R.C.; Richardson, A.K.; Server, W.L.

1986-01-01T23:59:59.000Z

263

PWR cores with silicon carbide cladding  

Science Conference Proceedings (OSTI)

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)

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

264

Containment condensing heat transfer. [PWR; BWR  

SciTech Connect

This report presents a mechanistic heat-transfer model that is valid for large scale containment heat sinks. The model development is based on the determination that the condensation is controlled by mass diffusion through the vapor-air boundary layer, and the application of the classic Reynolds' analogy to formulate expressions for the transfer of heat and mass based on hydrodynamic measurements of the momentum transfer. As a result, the analysis depends on the quantification of the shear stress (momentum transfer) at the interface between the condensate film and the vapor-air boundary layer. In addition, the currently used Tagami and Uchida test observations and their range of applicability are explained.

Gido, R.G.; Koestel, A.

1983-01-01T23:59:59.000Z

265

Nuclear reactor containment spray testing system. [PWR  

SciTech Connect

Disclosed is a method for periodic testing of a spray system in a nuclear reactor containment. The method includes injecting a gas into the spray system such that a temperature differential exists between the gas and the containment atmosphere. Scanning the gas jet discharged from the spray nozzles with infrared apparatus then provides a real-time thermal image on a monitor, such as a cathode ray tube, and detects any partially or completely blocked nozzles in the spray system. The scanning may be performed from the containment operating deck. 1 claim, 4 figures.

Rubin, K.

1978-01-10T23:59:59.000Z

266

PWR Standard Radiation Monitoring Program: 2013 Summary  

Science Conference Proceedings (OSTI)

The Nuclear Energy Institute/Institute of Nuclear Power Operations/Electric Power Research Institute (EPRI) Radiation Protection “RP 2020” Initiative was developed to promote radiation dose reduction by emphasizing radiological protection fundamentals and reducing radioactive source term. EPRI was charged as the technical lead in the area of source term reduction, and EPRI’s Radiation Management Program initiated a multi-year program to develop an understanding of source term ...

2013-06-27T23:59:59.000Z

267

Two-phase jet loads. [PWR  

Science Conference Proceedings (OSTI)

Two-phase jets are currently being studied to improve engineering models for the prediction of loads on pipes and structures during LOCAs. Multi-dimensional computer codes such as BEACON/MOD2, CSQ, and TRAC-P1A are being employed to predict flow characteristics and flow-structure loading. Our ultimate goal is to develop a new approximate engineering model which is superior to the F.J. Moody design model. Computer results are compared with data obtained from foreign sources, and a technique for using the TRAC-P1A vessel component as a containment model is presented. In general, good agreement with the data is obtained for saturated stagnation conditions; however, difficulties are encountered for subcooled stagnation conditions, possibly due to nucleation delay and non-equilibrium effects.

Tomasko, D.

1980-01-01T23:59:59.000Z

268

CGS-IAH Conf. Niagara Falls 2002 (to be published -confidential) LABORATORY MEASUREMENTS AND PREDICTIVE EQUATIONS FOR  

E-Print Network (OSTI)

in a biogas digester as revealed by denaturing gradient gel electrophoresis and 16S rDNA sequencing analysis

Aubertin, Michel

269

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

SciTech Connect

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.

Hopper, Nicole; Goldman, Charles; Neenan, Bernie

2006-05-12T23:59:59.000Z

270

Impact of the Demand-Side Management (DSM) Program structure on the cost-effectiveness of energy efficiency projects  

SciTech Connect

Pacific Northwest Laboratory (PNL) analyzed the cost-effective energy efficiency potential of Fort Drum, a customer of the Niagara Mohawk Power Corporation (NMPC) in Watertown, New York. Significant cost-effective investments were identified, even without any demand-side management (DSM) incentives from NMPC. Three NMPC DSM programs were then examined to determine the impact of participation on the cost-effective efficiency potential at the Fort. The following three utility programs were analyzed: (1) utility rebates to be paid back through surcharges, (2) a demand reduction program offered in conjunction with an energy services company, and (3) utility financing. Ultimately, utility rebates and financing were found to be the best programs for the Fort. This paper examines the influence that specific characteristics of the DSM programs had on the decision-making process of one customer. Fort Drum represents a significant demand-side resource, whose decisions regarding energy efficiency investments are based on life-cycle cost analysis subject to stringent capital constraints. The structures of the DSM programs offered by NMPC affect the cost-effectiveness of potential efficiency investments and the ability of the Fort to obtain sufficient capital to implement the projects. This paper compares the magnitude of the cost-effective resource available under each program, and the resulting level of energy and demand savings. The results of this analysis can be used to examine how DSM program structures impact the decision-making process of federal and large commercial customers.

Stucky, D.J.; Shankle, S.A.; Dixon, D.R.; Elliott, D.B.

1994-12-01T23:59:59.000Z

271

Biomass Reburning: Modeling/Engineering Studies  

SciTech Connect

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.

Vladimir M. Zamansky

1998-01-20T23:59:59.000Z

272

Market potential of electrolytic hydrogen production in three northeastern utilities' service territories. Final report  

SciTech Connect

The study develops a method for exploring the market potential for electrolytic hydrogen. The service areas of three northeastern utilities - Public Service Electric and Gas, Niagara Mohawk, and Northeast Utilities - are examined, and results reported on the effort to locate specialty hydrogen users, determine patterns of hydrogen utilization, and assess the possibility of satisfying this hydrogen demand by electrolytic hydrogen from advanced electrolyzers. Hydrogen users were sought in six major product categories: chemicals, pharmaceuticals, oils, metals, electronics and float glass. Identification of users through appropriate standard industrial classification codes served as a basis for locating possible users in each of the service areas. Mailed questionnaires sought information on hydrogen demand, characteristics of hydrogen use, present hydrogen supply costs, and factors that would influence the purchase of an electrolyzer. In the three utility service areas examined, electrolytic hydrogen can be expected to have limited success competing with merchant hydrogen. Specific hydrogen users may be found whose location with respect to the source of merchant hydrogen may put electrolytic hydrogen at an economic advantage. Reduction in electrolyzer plant costs may be necessary to expand the possibilities for electrolysis. Annual power requirements for current potential demand for electrolytic hydrogen in three utilities was estimated at 140 x 10/sup 6/ kWh, which could expand to 240 x 10/sup 6/ kWh in ten years.

Fein, E.; Edwards, K.

1984-05-01T23:59:59.000Z

273

Stanton County Public Pwr Dist | Open Energy Information  

Open Energy Info (EERE)

Place Stanton, Nebraska Place Stanton, Nebraska Utility Id 17979 Utility Location Yes Ownership P NERC Location SPP NERC SPP Yes Activity Distribution Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] SGIC[2] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Stanton County Public Power District Smart Grid Project was awarded $397,000 Recovery Act Funding with a total project value of $794,000. Utility Rate Schedules Grid-background.png Commercial Pumping Single Phase Commercial Commercial Pumping Three Phase Commercial Commercial Single Phase Commercial Commercial Three Phase Commercial Discounted Small Public Authority Services Single Phase Commercial Discounted Small Public Authority Services Three Phase Commercial

274

PWR Primary-to-Secondary Leak Guidelines - Revision 3  

Science Conference Proceedings (OSTI)

Primary-to-secondary leakage of steam generator tubes in pressurized water reactors (PWRs) can result from mechanisms that propagate slowly or rapidly. This report represents Revision 3 of industry-wide guidelines first proposed in 1995 to address the issue of leakage.

2004-12-16T23:59:59.000Z

275

Estimation of structural reliability under combined loads. [PWR; BWR  

DOE Green Energy (OSTI)

For the overall safety evaluation of seismic category I structures subjected to various load combinations, a quantitative measure of the structural reliability in terms of a limit state probability can be conveniently used. For this purpose, the reliability analysis method for dynamic loads, which has recently been developed by the authors, was combined with the existing standard reliability analysis procedure for static and quasi-static loads. The significant parameters that enter into the analysis are: the rate at which each load (dead load, accidental internal pressure, earthquake, etc.) will occur, its duration and intensity. All these parameters are basically random variables for most of the loads to be considered. For dynamic loads, the overall intensity is usually characterized not only by their dynamic components but also by their static components. The structure considered in the present paper is a reinforced concrete containment structure subjected to various static and dynamic loads such as dead loads, accidental pressure, earthquake acceleration, etc. Computations are performed to evaluate the limit state probabilities under each load combination separately and also under all possible combinations of such loads.

Shinozuka, M.; Kako, T.; Hwang, H.; Brown, P.; Reich, M.

1983-01-01T23:59:59.000Z

276

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

SciTech Connect

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

Amos, C N; Schrock, V E

1983-09-01T23:59:59.000Z

277

Isotopic Generation and Confirmation of the PWR Application Model   

SciTech Connect

The objective of this calculation is to establish an isotopic database to represent commercial spent nuclear fuel (CSNF) from pressurized water reactors (PWRs) in criticality analyses performed for the proposed Monitored Geologic Repository at Yucca Mountain, Nevada. Confirmation of the conservatism with respect to criticality in the isotopic concentration values represented by this isotopic database is performed as described in Section 3.5.3.1.2 of the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000). The isotopic database consists of the set of 14 actinides and 15 fission products presented in Section 3.5.2.1.1 of YMP 2000 for use in CSNF burnup credit. This set of 29 isotopes is referred to as the principal isotopes. The oxygen isotope from the UO{sub 2} fuel is also included in the database. The isotopic database covers enrichments of {sup 235}U ranging from 1.5 to 5.5 weight percent (wt%) and burnups ranging from approximately zero to 75 GWd per metric ton of uranium (mtU). The choice of fuel assembly and operating history values used in generating the isotopic database are provided is Section 5. Tables of isotopic concentrations for the 29 principal isotopes (plus oxygen) as a function of enrichment and burnup are provided in Section 6.1. Results of the confirmation of the conservatism with respect to criticality in the isotopic concentration values are provided in Section 6.2.

L.B. Wimmer

2003-11-10T23:59:59.000Z

278

Survey of Iron and Nickel Concentrations in PWR Primary Coolant  

Science Conference Proceedings (OSTI)

The concentrations of iron and nickel corrosion products in primary coolant water were measured at eleven different pressurized water reactors. Two reactors experienced anomalies in the axial power distribution during the cycles that were sampled. The axial power distribution anomalies appeared to be associated with high-coolant nickel concentrations early in the fuel cycle.

2001-07-27T23:59:59.000Z

279

Reactor physics assessment of thick silicon carbide clad PWR fuels  

E-Print Network (OSTI)

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 ...

Bloore, David A. (David Allan)

2013-01-01T23:59:59.000Z

280

Wire wrapped fuel pin hexagonal arrays for PWR service  

E-Print Network (OSTI)

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 ...

Diller, Peter Ray

2005-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "niagara mohawk pwr" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


281

Proceedings of the June 2004 EPRI PWR Primary Shutdown Workshop  

Science Conference Proceedings (OSTI)

There have been a number of outages within the last couple of years that were significant enough to warrant review and examination for lessons learned. Industry feedback suggested a workshop be conducted to address these experiences, including issues associated with shorter outage campaigns and the effects of core design changes. Thus, a two day workshop was jointly coordinated by Chemistry and FRP and held on June 9-10, 2004 at the EPRI offices in Charlotte, NC so that lessons learned could potentially ...

2004-08-23T23:59:59.000Z

282

Optimizing Site-Specific ALARA Assessments: PWR Methodology Development  

Science Conference Proceedings (OSTI)

Over the past decade utilities have made significant progress towards reducing personnel exposure to levels that are As Low As Reasonably Achievable (ALARA). This document presents a methodology and protocol for assisting utility ALARA program and Health Physics Managers in evaluating and enhancing utility ALARA programs with a view to facilitate implementation of further reasonable exposure reduction measures.

1999-09-16T23:59:59.000Z

283

Virgin Islands Wtr&Pwr Auth | Open Energy Information  

Open Energy Info (EERE)

Utility Location Yes Ownership S NERC Location VI Operates Generating Plant Yes Activity Generation Yes Activity Distribution Yes References EIA Form EIA-861 Final Data File for...

284

Method for quantitative assessment of nuclear safety computer codes. [PWR  

SciTech Connect

A procedure has been developed for the quantitative assessment of nuclear safety computer codes and tested by comparison of RELAP4/MOD6 predictions with results from two Semiscale tests. This paper describes the developed procedure, the application of the procedure to the Semiscale tests, and the results obtained from the comparison.

Dearien, J.A.; Davis, C.B.; Matthews, L.J.

1979-01-01T23:59:59.000Z

285

LOFT contribution to PWR fuel analysis and testing  

SciTech Connect

The paper provides a brief overview of the Loss-of-Fluid Test (LOFT) Facility, a review of the key results from LOCA tests performed to date, and plans for the future of LOFT, as an introduction to the more specific information about LOFT fuel that is to follow. The purpose of the LOFT tests is to provide data for assessing the accuracy of the analytical models used in evaluating the safety of commercial nuclear power plants. While the main purpose of LOFT is to investigate phenomena important to the loss-of-coolant accident (LOCA), the 55 MWt LOFT reactor system has been scaled to commercial power reactors and heavily instrumented so that information on fuel and system behavior over a range of normal and off-normal conditions can be obtained.

Leach, L.P.

1978-01-01T23:59:59.000Z

286

Coolant monitoring apparatus for nuclear reactors. [PWR; BWR  

DOE Patents (OSTI)

A system for monitoring coolant conditions within a pressurized vessel is described. A length of tubing extends outward from the vessel from an open end containing a first line restriction at the location to be monitored. The flowing fluid is cooled and condensed before passing through a second line restriction. Measurement of pressure drop at the second line restriction gives an indication of fluid condition at the first line restriction. Multiple lengths of tubing with open ends at incremental elevations can measure coolant level within the vessel.

Tokarz, R.D.

1981-08-06T23:59:59.000Z

287

PNL technical review of pressurized thermal-shock issues. [PWR  

SciTech Connect

Pacific Northwest Laboratory (PNL) was asked to develop and recommend a regulatory position that the Nuclear Regulatory Commission (NRC) should adopt regarding the ability of reactor pressure vessels to withstand the effects of pressurized thermal shock (PTS). Licensees of eight pressurized water reactors provided NRC with estimates of remaining effective full power years before corrective actions would be required to prevent an unsafe operating condition. PNL reviewed these responses and the results of supporting research and concluded that none of the eight reactors would undergo vessel failure from a PTS event before several more years of operation. Operator actions, however, were often required to terminate a PTS event before it deteriorated to the point where failure could occur. Therefore, the near-term (less than one year) recommendation is to upgrade, on a site-specific basis, operational procedures, training, and control room instrumentation. Also, uniform criteria should be developed by NRC for use during future licensee analyses. Finally, it was recommended that NRC upgrade nondestructive inspection techniques used during vessel examinations and become more involved in the evaluation of annealing requirements.

Pedersen, L.T.; Apley, W.J.; Bian, S.H.; Defferding, L.J.; Morgenstern, M.H.; Pelto, P.J.; Simonen, E.P.; Simonen, F.A.; Stevens, D.L.; Taylor, T.T.

1982-07-01T23:59:59.000Z

288

Modeling PWR Fuel Corrosion Product Deposition and Growth Process  

Science Conference Proceedings (OSTI)

This report describes further developments to the crud chemistry model developed in 2003. This model calculates the thermal and chemical conditions within fuel crud deposits. The model was developed to understand fuel crud scrape observations and help alleviate AOA in plants suffering this problem. The model predicts the main form of boron in thick crud (30µm) is precipitated LiBO2. This precipitates because Li and boric acid concentrate in the bottom of the deposit due to evaporation. The rise in temper...

2005-09-26T23:59:59.000Z

289

Detonation of hydrogen-air mixtures. [PWR; BWR  

DOE Green Energy (OSTI)

The detonation of a hydrogen-air cloud subsequent to an accidental release of hydrogen into ambient surroundings cannot be totally ruled out in view of the relative sensitivity of the hydrogen-air system. The present paper investigates the key parameters involved in hydrogen-air detonations and attempts to establish quantitative correlations between those that have important practical implications. Thus, for example, the characteristic length scale lambda describing the cellular structure of a detonation front is measured for a broad range of hydrogen-air mixtures and is quantitatively correlated with the key dynamic detonation properties such as detonability, transmission and initiation.

Lee, J.H.S.; Knystautas, R.; Benedick, W.B.

1983-01-01T23:59:59.000Z

290

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

DOE Patents (OSTI)

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

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

1980-06-06T23:59:59.000Z

291

End effects on elbows subjected to moment loadings. [PWR; BWR  

SciTech Connect

So-called end effects for moment loadings on short-radius and long-radius butt welding elbows of various arc lengths are investigated with a view toward providing more accurate design formulas for critical piping systems. Data developed in this study, along with published information, were used to develop relatively simple design equations for elbows attached at both ends to long sections of straight pipe. These formulas are the basis for an alternate ASME Code procedure for evaluating the bending moment stresses in Class 1 nuclear piping (ASME Code Case N-319). The more complicated problems of elbows with other end conditions, e.g., flanges at one or both ends, are also considered. Comparisons of recently published experimental and theoretical studies with current industrial code design rules for these situations indicate that these rules also need to be improved.

Rodabaugh, E.C.; Moore, S.E.

1982-01-01T23:59:59.000Z

292

Developing PWR Aging-Management Strategies for Reactor Vessel ...  

Science Conference Proceedings (OSTI)

AREVA Fuel Condition Index for a Pressurized Water Reactor .... Stress Corrosion Cracking Behavior near the Fusion Boundary of Dissimilar Weld Joint with ...

293

Modeling PWR Fuel Corrosion Product Deposition and Growth Processes  

Science Conference Proceedings (OSTI)

Development of axial offset anomaly (AOA) in pressurized water reactors (PWRs) drove industry to conduct crud scrape campaigns at a number of units to characterize and better understand the material being deposited on the fuel clad surface. This report describes the first phase of a program to develop models that describe the crud deposition and growth process, including the many phenomena that influence not only the deposit mass, but the composition. The models will be benchmarked against published crud...

2004-12-09T23:59:59.000Z

294

Fuel Reliability Guidelines: PWR Fuel Cladding Corrosion and Crud  

Science Conference Proceedings (OSTI)

Developed in collaboration with utilities, industry organizations, and fuel vendors, a series of new EPRI guidelines capture state-of-the-art knowledge and describe best practices for eliminating fuel failures at nuclear power plants. The guidelines provide mandatory, needed, and best practice recommendations based on a thorough review of operating experience, fuel failure analyses, and fuel design and manufacturing procedures. More than 200 industry experts reviewed the guidelines to ensure accuracy and...

2008-04-01T23:59:59.000Z

295

PWR Field Experience with Elevated, Constant pH  

Science Conference Proceedings (OSTI)

This report analyzes chemistry data from several pressurized water reactors (PWRs) operating reactor coolant with elevated, constant pH to determine its effectiveness for managing core performance and dose rate problems stemming from corrosion product and activity transport. The operating data are supplemented with relevant shutdown and startup experience to identify the overall impact of elevated, constant pH on primary system operation.

2001-12-06T23:59:59.000Z

296

Method and apparatus for monitoring two-phase flow. [PWR  

DOE Patents (OSTI)

A method and apparatus for monitoring two-phase flow is provided that is particularly related to the monitoring of transient two-phase (liquid-vapor) flow rates such as may occur during a pressurized water reactor core blow-down. The present invention essentially comprises the use of flanged wire screens or similar devices, such as perforated plates, to produce certain desirable effects in the flow regime for monitoring purposes. One desirable effect is a measurable and reproducible pressure drop across the screen. The pressure drop can be characterized for various known flow rates and then used to monitor nonhomogeneous flow regimes. Another useful effect of the use of screens or plates in nonhomogeneous flow is that such apparatus tends to create a uniformly dispersed flow regime in the immediate downstream vicinity. This is a desirable effect because it usually increases the accuracy of flow rate measurements determined by conventional methods.

Sheppard, J.D.; Tong, L.S.

1975-12-19T23:59:59.000Z

297

Fuel performance annual report for 1981. [PWR; BWR  

SciTech Connect

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

Bailey, W.J.; Tokar, M.

1982-12-01T23:59:59.000Z

298

PWR Fuel Shipping Limits & RNP Core Design  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Fuel Fuel Transportation Experience Steven Edwards, Progress Energy September 21, 2005 2 Discussion Topics Progress Energy Transportation History Success Factors Shipment Security Dedicated Trains Emergency Response Public Communication/Participation Summary 3 Brunswick Harris Crystal River Robinson Progress Energy Nuclear Plants 4 Spent Fuel Management Strategy Maintain operating reserve at all nuclear units Spent fuel shipping program to reduce inventories at Brunswick and Robinson Maximize use of Harris spent fuel pools 5 Transportation Experience 191 shipments 1,000 MTU transported 4,541 spent fuel assemblies transported 6 Transportation Experience First Shipment - 1977 Active spent fuel transportation program since 1989 12 to 15 shipments per year

300

Notice of Intent to prepare an Environmental Impact Statement for the Construction and Operation of the Proposed Wellton-Mohawk Generating Facility, Yuma County, AZ (5/19/03)  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

056 056 Federal Register / Vol. 68, No. 96 / Monday, May 19, 2003 / Notices 3 The ALJ report will not be an initial decision, so we will not entertain the filing of briefs on or opposing exceptions. Further, we do not anticipate the need for cross-examination of witnesses. The Judge need not create an exhaustive record, but may work with the parties to create the record that provides a thorough picture of the facts, problems, and opportunities, and lessons to be learned. date of this order. 3 State Commission ALJs or expert staff that participate in the fact-finding may offer written comments or conclusions that will be appended to the Commission ALJ's report. The Commission Orders (A) The Secretary is hereby directed to publish this order in the Federal Register.

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301

Page not found | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Exclusion Determination United States Army Corps Niagara River, New York Small Hydropower CX(s) Applied: A9, A11 Date: 11302009 Location(s): Niagara River, New York...

302

niagarastoragesite  

Office of Legacy Management (LM)

Niagara Falls Storage Site, New York, is a Niagara Falls Storage Site, New York, is a 191-acre site located on Pletcher Road in the towns of Lewiston and Porter, Niagara County, in northwestern New York. It is approximately 10 miles north of the city of Niagara Falls and 19 miles northwest of Buffalo, New York. The site is a remnant of the U.S. Army's 7,500-acre Lake Ontario Ordnance Works. The property includes a 10-acre interim waste

303

CX-009805: Categorical Exclusion Determination | Department of...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Exclusion Determination CX-009805: Categorical Exclusion Determination Maintenance and Wood Pole Replacement along the Gila Wellton Mohawk 161 Kilovolt Transmission Line CX(s)...

304

Characteristics and trends in a National Study of Consumer Outage Costs  

E-Print Network (OSTI)

1999) Niagra Mohawk (1985) Duke Energy Company (1992, 1997)Gas and Electric and Duke Energy) the same customer classesindustrial concerns (i.e. , Duke Energy, Southern Company,

Lawton, Leora; Eto, Joseph H.; Katz, Aaron; Sullivan, Michael

2003-01-01T23:59:59.000Z

305

Customer Strategies for Responding to Day-Ahead Market HourlyElectricity Pricing  

Science Conference Proceedings (OSTI)

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.

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

2005-08-25T23:59:59.000Z

306

California State University, Desert Studies Consortium and LSA Associates, Inc. Old Ores: mines and mineral marketing  

E-Print Network (OSTI)

466 (gas pipeline road) and proceed to the Mohawk Mine. 43.3 (4.8) Turn left (north) toward the Mohawk a brief duel at the range cabin, one gunman was killed; the other died later of bullet wounds. Report and the Vanderbilt Mine. When the Tonopah & Tidewater Railroad siphoned off trade in 1907, the Santa Fe built

de Lijser, Peter

307

www.eia.gov  

U.S. Energy Information Administration (EIA)

Corning City of Corning 7 IC 6 IN Jasper ... Forest Creek Wind Farm LLC ... Hawaii Hawi Renewable Development LLC Hawi Wind Farm V-47 Niagara

308

ESTIMATE OF RADIUM-226 CONCENTRATIONS IN RUBBLED PCB WAREHOUSE...  

Office of Legacy Management (LM)

IN RUBBLED PCB WAREHOUSE ON VICINITY PROPERTY B ADJACENT TO THE NIAGARA FALLS STORAGE SITE MAY 1987 Prepared for UNITED STATES DEPARTMENT OF ENERGY OAK RIDGE OPERATIONS...

309

Secondary system modeling and method for a nuclear power plant training simulator. [PWR  

SciTech Connect

Disclosed is a method and system for the real-time simulation of the dynamic operation of a nuclear power plant in which a secondary system for operating the steam turbine includes reheaters for increasing the steam temperature and pressure between the low and high pressure turbine stages, an electrohydraulic controller for operating the turbine at required speed, a condenser for condensing turbine exhaust steam, a condensate and feedwater system for pumping condensed steam back to the secondary side of a steam generator, a gland steam system for preventing leakage from and to the atmosphere, and a cooling system for the main generator, utilizes apparatus that includes a digital computer for calculating output data corresponding to physical values for the operation of the plant in accordance with input data. A control console includes automatic and manually operable devices corresponding to plant control devices for varying the input data, and indicating devices responsive to the output data for monitoring the operation of the plant. 40 claims, 45 figures.

Johnson, S.J.

1977-08-16T23:59:59.000Z

310

Influence of Radiolysis and Hydrogen Embrittlement on the In-Service Cracking of PWR Internal Structures  

Science Conference Proceedings (OSTI)

Irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steel components exposed to a high neutron flux in light water reactors is becoming an increasing concern for nuclear power plant owners. This study investigates the interaction between hydrogen in the metal and the water environment that may exacerbate the cracking process in PWRs.

1999-11-10T23:59:59.000Z

311

Evaluation of Zinc Addition to Primary Coolant of Farley-2 PWR  

Science Conference Proceedings (OSTI)

A demonstration project at Southern Nuclear Operating Company's Farley Unit 2 showed that the addition of 35-45 ppb zinc to the primary coolant resulted in lower radiation dose rates and an apparent decrease in primary water stress corrosion cracking (PWSCC) of Alloy 600 steam generator tubing. The zinc had no significant effect on fuel cladding corrosion.

1996-10-31T23:59:59.000Z

312

Guidelines for Fabrication and Assembly of Alloy 690 Components in PWR Primary Systems  

Science Conference Proceedings (OSTI)

Although test data and field experience indicate that Alloy 690 is significantly more resistant to primary water stress corrosion cracking (PWSCC) degradation than Alloy 600, laboratory testing has shown that thermo-mechanical processing and surface bulk cold work can introduce vulnerabilities to PWSCC. This document provides guidelines for the fabrication, assembly, and installation of heavy section Alloy 690 parts and components in the primary system of pressurized water reactors (PWRs). The ...

2013-12-02T23:59:59.000Z

313

SUMMARY OF REACTOR DESIGN INFORMATION FROM THREE YEARS' OPERATION OF A SMALL PWR  

SciTech Connect

Reactor design information obtained from 3 years' operation of a small pressurized-water reactor, the SM-1 (formerly APPR-l), is presented and discussed. The SM-1 reactor, designed to produce 10 Mw(t) power, employs fully enriched uranium fuel in the form of UO/sub 2/ dispersed in stainless-steel fuel plates. The reactor is cooled by water at 1200 psia and mean temperature of 44) deg F. Core-physics measurements were performed of temperature coefficient, pressure coefficient, rod calibration, stuck rod position, and transient xenon as a function of core burn-out. Core burn-out characteristics were compared with few- group calculations, and reasonable agreement was obtained. Thermal-heat-balance data were obtained on the reactor core. The temperature pattern in the nominal and hot channels under operating conditions was calculated. These calculations indicated that certain of the fuel channels operated in the nucleate boiling regime. Examination of one of the fuel channels suspected of nucleate boiling indicated no adverse effects. The system response to load perturbations and during pump coast-down was measured utilizing plant instrumentation. This response was compared with analytical predictions using a lumped kinetic model, and reasonable agreement was found. Both neutron and gamma traverses were made through the primary shield during reactor operation. Gamma traverses were also made through the primary shield as a function of time after reactor shutdown. Conventional shielding calculational methods are found to give agreement with experiment sufficient for design purposes. An absolute ionization chamber was employed to measure N/sup 16/ activity in the reactor coolant. These measurements were compared with N/sup 16/ calculated from the (n,p) reaction on O/ sup 16/. (auth)

Gallagher, J.G.

1960-09-01T23:59:59.000Z

314

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

Science Conference Proceedings (OSTI)

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

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

2011-01-01T23:59:59.000Z

315

Development of a reliability-analysis method for category I structures. [PWR; BWR  

DOE Green Energy (OSTI)

The present paper develops a reliability analysis method for category I nuclear structures, particularly for reinforced concrete containment structures subjected to various load combinations. The loads considered here include dead loads, accidental internal pressure and earthquake ground acceleration. For mathematical tractability, an earthquake occurrence is assumed to be governed by the Poisson arrival law, while its acceleration history is idealized as a Gaussian vector process of finite duration. A vector process consists of three component processes, each with zero mean. The second order statistics of this process are specified by a three-by-three spectral density matrix with a multiplying factor representing the overall intensity of the ground acceleration. With respect to accidental internal pressure, the following assumptions are made: (a) it occurs in accordance with the Poisson law; (b) its intensity and duration are random; and (c) its temporal rise and fall behaviors are such that a quasi-static structural analysis applies. A dead load is considered to be a deterministic constant.

Shinozuka, M.; Kako, T.; Hwang, H.; Reich, M.

1983-01-01T23:59:59.000Z

316

Evaluation of the effects of detonation in a spherical bomb. [PWR; BWR  

DOE Green Energy (OSTI)

An analysis is presented of the time-dependent pressure forces and impulse loadings on the walls of the hemispherical dome of a nuclear reactor pressure vessel arising from a centrally ignited hydrogen-oxygen detonation. Investigated in this context are the effects of richness of the detonable gas mixture as well as those due to the inclusion of water vapor. In the analysis the gas mixture was treated as a perfect gas, and the partial differential equations governing the gasdynamic flow were integrated using the CLOUD CODE - a finite-difference technique set in Lagrangian coordinates and incorporating the smoothing action of artificial viscosity. The most interesting results pertain to the ringing of pressure pulses at the walls. Their frequency is quite uniform, and their pressure peaks, at levels significantly higher than that of combustion at constant volume, decay at a negligible rate.

Kurylo, J.; Oppenheim, A.K.

1979-11-01T23:59:59.000Z

317

Steam Generator Management Program: PWR Steam Generator Top-of-Tubesheet Denting  

Science Conference Proceedings (OSTI)

Denting of steam generator tubing is the reduction in tube diameter due to the forces exerted by corrosion products on the outer diameter surfaces. This deformation can increase the risk of stress corrosion cracking due to the high stresses, strains, and cold work developed in the tube. Historically, denting at carbon steel tube support plate locations was a significant factor necessitating the early replacement of several steam generators. Currently, denting and stress corrosion cracking are being exper...

2012-06-06T23:59:59.000Z

318

Application of the EPRI Standard Radiation Monitoring Program for PWR Radiation Field Reduction  

Science Conference Proceedings (OSTI)

The NEI/INPO/EPRI RP 2020 Initiative was developed to promote radiation dose reduction by emphasizing radiological protection fundamentals and reducing radioactive source term. EPRI was charged as the technical lead in the area of source term reduction. EPRI's Radiation Management program initiated a multi-year program to develop an understanding of source term generation and transport with the eventual goal of providing plant specific recommendations for source term reduction. Reinstatement of the Stand...

2007-11-28T23:59:59.000Z

319

Steam Generator Management Program: PWR Primary-to-Secondary Leak Guidelines Revision 4  

Science Conference Proceedings (OSTI)

Primary-to-secondary leakage of steam generator tubes in PWRs can result from mechanisms that propagate slowly or rapidly. Control room operators rely on online data for a rapid assessment of tube leakage conditions to ensure that the plant is maneuvered safely and to minimize the risk of tube rupture. Industry experts prepared and reviewed these revised guidelines to incorporate recent industry operating experience and technology improvements and to review the technical bases for action levels. This rep...

2011-09-12T23:59:59.000Z

320

Materials Reliability Program: PWR Bottom Mounted Nozzle (BMN) Issue Response Handbook (MRP-372)  

Science Conference Proceedings (OSTI)

Resolution of a case of primary water stress corrosion cracking (PWSCC) detected in a bottom mounted nozzle (BMN) could be expected to require substantial utility resources, even if only a single nozzle were to be affected, because of the challenges in accessing it in pressurized water reactors (PWRs) to perform inspections and repairs. This Handbook compiles the available knowledge and reference information to guide utility development of a site-specific response plan to an emergent BMN ...

2013-11-14T23:59:59.000Z

Note: This page contains sample records for the topic "niagara mohawk pwr" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


321

Isotopic validation for PWR actinide-only burnup credit using Yankee Rowe data  

Science Conference Proceedings (OSTI)

Safety analyses of criticality control systems for transportation packages include an assumption that the spent nuclear fuel (SNF) loaded into the package is fresh or unirradiated. In other words, the spent fuel is assumed to have its original, as-manufactured U-235 isotopic content. The ``fresh fuel`` assumption is very conservative since the potential reactivity of the nuclear fuel is substantially reduced after being irradiated in the reactor core. The concept of taking credit for this reduction in nuclear fuel reactivity due to burnup of the fuel, instead of using the fresh fuel assumption in the criticality safety analysis, is referred to as ``Burnup Credit.`` Burnup credit uses the actual physical composition of the fuel and accounts for the net reduction of fissile material and the buildup of neutron absorbers in the fuel as it is irradiated. Neutron absorbers include actinides and other isotopes generated as a result of the fission process. Using only the change in actinide isotopes in the burnup credit criticality analysis is referred to as ``Actinide-Only Burnup Credit.`` The use of burnup credit in the design of criticality control systems enables more spent fuel to be placed in a package. Increased package capacity results in a reduced number of storage, shipping and disposal containers for a given number of SNF assemblies. Fewer shipments result in a lower risk of accidents associated with the handling and transportation of spent fuel, thus reducing both radiological and nonradiological risk to the public. This paper describes the modeling and the results of comparison between measured and calculated isotopic inventories for a selected number of samples taken from a Yankee Rowe spent fuel assembly.

NONE

1997-11-01T23:59:59.000Z

322

Acoustic emission: flaw relationship for inservice monitoring of nuclear reactor pressure boundaries. [PWR; BWR  

Science Conference Proceedings (OSTI)

The objective of the acoustic emission (AE)/flaw characterization program is to provide an experimental feasibility evaluation of using the AE method on a continuous basis (during operation and during hydrotest) to detect and analyze flaw growth in reactor pressure vessels and primary piping. This effort is based on the philosophy that AE shows demonstrated capability for being a valuable addition to current nondestructive inspection (NDI) methods with unique capability for continuous monitoring, high sensitivity and remote flaw location.

Not Available

1981-10-01T23:59:59.000Z

323

Heat transfer to water from a vertical tube bundle under natural-circulation conditions. [PWR; BWR  

SciTech Connect

The natural circulation heat transfer data for longitudinal flow of water outside a vertical rod bundle are needed for developing correlations which can be used in best estimate computer codes to model thermal-hydraulic behavior of nuclear reactor cores under accident or shutdown conditions. The heat transfer coefficient between the fuel rod surface and the coolant is the key parameter required to predict the fuel temperature. Because of the absence of the required heat transfer coefficient data base under natural circulation conditions, experiments have been performed in a natural circulation loop. A seven-tube bundle having a pitch-to-diameter ratio of 1.25 was used as a test heat exchanger. A circulating flow was established in the loop, because of buoyancy differences between its two vertical legs. Steady-state and transient heat transfer measurements have been made over as wide a range of thermal conditions as possible with the system. Steady state heat transfer data were correlated in terms of relevant dimensionless parameters. Empirical correlations for the average Nusselt number, in terms of Reynolds number, Rayleigh number and the ratio of Grashof to Reynolds number are given.

Gruszczynski, M.J.; Viskanta, R.

1983-01-01T23:59:59.000Z

324

Thermal-radiation heat-transfer model for degraded cores. [PWR; BWR  

SciTech Connect

One consequence of the accident at the Three Mile Island Unit 2 (TMI-2) nuclear power plant is a realization by the nuclear power technical community that there is a need for calculational tools that can be used to analyze the TMI-2 accident and to investigate hypothetical situations involving degraded light-water reactor (LWR) cores. As a result, there are now several ongoing modeling and code development efforts in the United States among which is the development of the MIMAS (Multifield Integrated Meltdown Analysis System code) at the Los Alamos National Laboratory. This paper describes a thermal-radiation heat-transfer model for LWR degraded cores that has been developed for the MIMAS code.

Tomkins, J.L.

1983-01-01T23:59:59.000Z

325

PWR Steam Generator Secondary-Side IGA/SCC: Correlations with Deposit Lead and Phosphate History  

Science Conference Proceedings (OSTI)

This report covers an investigation into possible correlations between steam generator deposit lead concentrations and rates of intergranular attack/stress corrosion cracking (IGA/SCC). It also considers the relationship between prior phosphate injection and the rate of IGA/SCC. Prompting the study is recent identification by analytical transmission electron microscopy (ATEM) of large concentrations of lead in oxides in secondary-side cracks in tube samples from steam generators where significant involve...

2005-07-05T23:59:59.000Z

326

1996 international joint power generation conference: Proceedings. Volume 2; PWR-Volume 30  

SciTech Connect

This is volume 2 of the proceedings of the 1996 International Joint Power Generation Conference held in Houston, Texas. The topics of the paper include emerging technologies for heat exchangers, maintenance and repair of feedwater and service water heat exchangers, steam surface condensers, understanding performance test codes, reliability, availability and maintainability of units and components, economics and reliability, Kalina cycle technologies, systems development under DOE`s combustion 2000 program, improvements in turbine materials and operating environment, combined cycle steam turbine application, case histories of turbine improvements, advanced generator mechanical design improvements and upgrades, steam turbine performance improvements, improvements in turbine materials and operating environment, combustion turbines for power generation, optimization of boiler performance using CEMS, international power plant design and restructuring issues, recent improvements in utility operations, turbine generator assessment technology, environmental compliance for industrial operations, industrial energy systems and services, industrial steam generation options.

Kielasa, L. [ed.] [Detroit Edison Co., MI (United States); Weed, G.E. [ed.] [Eastman Kodak Co., Rochester, NY (United States)

1996-12-31T23:59:59.000Z

327

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

E-Print Network (OSTI)

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 ...

Beccherle, Julien

2007-01-01T23:59:59.000Z

328

Numerical Modeling Study and Assessment of PWR Fuel Rod and Assembly Distortion  

Science Conference Proceedings (OSTI)

Fuel assembly and rod distortion experienced in pressurized water reactors (PWRs) result in numerous operational challenges to plant operators such as mechanical interference between distorted assembly and control rods, difficulties in unloading and reloading cores during outages, and possibly anomalous fuel performance due to atypical water gaps. Therefore, an improved understanding of the various parameters contributing to distortion is important in order to manage or otherwise eliminate these ...

2012-11-30T23:59:59.000Z

329

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

SciTech Connect

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.

Joy L. Rempe; Darrell L. Knudson

2013-03-01T23:59:59.000Z

330

Pwr fuel assembly optimization using adaptive simulated annealing coupled with translat  

E-Print Network (OSTI)

Optimization methods have been developed and refined throughout many scientific fields of study. This work utilizes one such developed technique of optimization called simulated annealing to produce optimal operation parameters for a 15x15 fuel 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 burnable absorber pins are also inserted. Both of these optimizations can be performed by coupling Adaptive Simulated Annealing to TransLAT which successfully searches the optimization space for a fuel assembly layout that produces the minimized pin power peaking factor. Within given time constraints this package produces optimal layouts within a given set of assumptions and constraints. Each layout is forced to maintain the fuel assembly average 235U enrichment as a constraint. Reductions in peaking factors that are produced through this method are on the order of 2% to 3% when compared to the baseline results. As with any simulated annealing approach, families of optimal layouts are produced that can be used at the engineer’s discretion.

Rogers, Timothy James

2008-08-01T23:59:59.000Z

331

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

E-Print Network (OSTI)

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 ...

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

2011-01-01T23:59:59.000Z

332

High-temperature oxidation of Zircaloy in hydrogen-steam mixtures. [PWR; BWR  

DOE Green Energy (OSTI)

Oxidation rates of Zircaloy-4 cladding tubes have been measured in hydrogen-steam mixtures at 1200 to 1700/sup 0/C. For a given isothermal oxidation temperature, the oxide layer thicknesses have been measured as a function of time, steam supply rate, and hydrogen overpressure. The oxidation rates in the mixtures were compared with similar data obtained in pure steam and helium-steam environments under otherwise identical conditions. The rates in pure steam and helium-steam mixtures were equivalent and comparable to the parabolic rates obtained under steam-saturated conditions and reported in the literature. However, when the helium was replaced with hydrogen of equivalent partial pressure, a significantly smaller oxidation rate was observed. For high steam-supply rates, the oxidation kinetics in a hydrogen-steam mixture were parabolic, but the rate was smaller than for pure steam or helium-steam mixtures. Under otherwise identical conditions, the ratio of the parabolic rate for hydrogen-steam to that for pure steam decreased with increasing temperature and decreasing steam-supply rate.

Chung, H.M.; Thomas, G.R.

1982-09-01T23:59:59.000Z

333

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

E-Print Network (OSTI)

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 ...

Stempien, John D. (John Dennis)

2011-01-01T23:59:59.000Z

334

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

E-Print Network (OSTI)

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 ...

Ellis, Tyler Shawn

2006-01-01T23:59:59.000Z

335

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

E-Print Network (OSTI)

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 ...

Liu, Jack S. H.

336

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

E-Print Network (OSTI)

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

Fujita, Edward Kei

337

Steam Generator Management Program: Empirical Model for Predicting Recirculating PWR Steam Generator Broached-Hole Blockage  

Science Conference Proceedings (OSTI)

Since their initial use in commercial plants in the 1960s, the steam generators (SGs) in pressurized water reactors (PWRs) have exhibited a number of reliability problems. Even though many of these are related to the integrity of the heat-transfer tubing and other internal components or to decreases in heat-transfer efficiency, some SG designs have been subject to a different issuedeposit-induced blockage of the broached flow holes in the tube support plates (TSPs) located within the SG shell. This study...

2011-04-29T23:59:59.000Z

338

Materials Reliability Program: Functionality Analysis for Babcock & Wilcox Representative PWR Internals (MRP-229-Rev. 3)  

Science Conference Proceedings (OSTI)

This report summarizes the functionality assessment of degradation for Babcock Wilcox (BW) designed reactor internals. The components analyzed include the core barrel assembly and selected Alloy X-750 and Alloy A-286 structural bolts.

2010-12-20T23:59:59.000Z

339

Three dimensional effects in analysis of PWR steam line break accident  

E-Print Network (OSTI)

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) ...

Tsai, Chon-Kwo

340

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

SciTech Connect

The once-through fuel cycle has been analyzed to see if there are substantial prospects for improved uranium ore utilization in current light water reactors, with a specific focus on pressurized water reactors. The types of changes which have been examined are: (1) re-optimization of fuel pin diameter and lattice pitch, (2) axial power shaping by enrichment gradation in fresh fuel, (3) use of 6-batch cores with semi-annual refueling, (4) use of 6-batch cores with annual refueling, hence greater extended (approximately doubled) burnup, (5) use of radial reflector assemblies, (6) use of internally heterogeneous cores (simple seed/blanket configurations), (7) use of power/temperature coastdown at the end of life to extend burnup, (8) use of metal or diluted oxide fuel, (9) use of thorium, and (10) use of isotopically separated low sigma/sub a/ cladding material. State-of-the-art LWR computational methods, LEOPARD/PDQ-7/FLARE-G, were used to investigate these modifications.

Fujita, E.K.; Driscoll, M.J.; Lanning, D.D.

1978-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "niagara mohawk pwr" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

Laboratory program to examine effects of layup conditions on pitting of Alloy 600. Final report. [PWR  

Science Conference Proceedings (OSTI)

The effect of various layup conditions on pitting of Alloy 600 tubing for steam generators was studied in beaker tests at 40/sup 0/C (104/sup 0/F). In addition, several methods to inhibit pitting were studied. The pitting solutions studied were copper chloride or seawater plus simulated sludge containing copper, copper oxide and magnetite. Results show that the pitting of Alloy 600 initiated in less than three weeks in the copper chloride solution with about 700 ppM chloride present. The same results were obtained in the seawater solution with about 6000 ppM chloride present. Retardation of pitting was achieved in three different ways: (1) decreasing the oxygen content of the solution, (2) decreasing the copper content of the solution, and (3) increasing the pH of the solution.

Whyte, D.D.

1983-04-01T23:59:59.000Z

342

CRACK GROWTH RESPONSE OF ALLOY 690 IN SIMULATED PWR PRIMARY WATER  

SciTech Connect

The stress corrosion crack growth response of three extruded alloy 690 CRDM tube heats was investigated in several thermomechanical conditions. Extremely low propagation rates (< 1 x 10{sup -9} mm/s) were observed under constant stress intensity factor (K) loading at 325-350 C in the as-received, thermally treated (TT) materials despite using a variety of transitioning techniques. Post-test observation of the crack-growth surfaces revealed only isolated intergranular (IG) cracking. One-dimensional cold rolling to 17% reduction and testing in the S-L orientation did not promote enhanced stress corrosion rates. However, somewhat higher propagation rates were observed in a 30% cold-rolled alloy 690TT specimen tested in the T-L orientation. Cracking of the cold-rolled material was promoted on grain boundaries oriented parallel to the rolling plane with the % IG increasing with the amount of cold rolling.

Toloczko, Mychailo B.; Bruemmer, Stephen M.

2009-12-01T23:59:59.000Z

343

Characterization of Corrosion Products on the Callaway Cycle 9 PWR Core  

Science Conference Proceedings (OSTI)

Analyses of Callaway fuel deposits have demonstrated that the deposits at this high-duty plant are qualitatively, as well as quantitatively, different from deposits in lower-duty (non-boiling) pressurized water reactors (PWRs). In particular, large amounts of the nickel-iron oxyborate, bonaccordite, were found in the Callaway deposits, while nickel ferrite was at most a minor component. Further, a significant quantity of monoclinic zirconia was incorporated in the Callaway crud. Such new findings may hav...

2001-08-28T23:59:59.000Z

344

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

DOE Patents (OSTI)

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.

Tokarz, R.D.

1981-10-27T23:59:59.000Z

345

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

E-Print Network (OSTI)

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

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

2008-01-01T23:59:59.000Z

346

Review of the United Kingdom PWR Primary-Circuit-Chemistry Program, Progress Report No. 1  

Science Conference Proceedings (OSTI)

Plant, laboratory, and theoretical studies in the United Kingdom are investigating the main factors influencing radiation-field buildup in PWRs. Data from these studies indicate that cobalt input from hard-facing alloys is particularly detrimental early in plant life. However, increasing coolant pH from 6.9 to 7.4 significantly reduces transport of soluble and particulate cobalt.

1989-05-09T23:59:59.000Z

347

Lead Adsorption on Nickel Alloys and Magnetite Under Faulted PWR Secondary Side Conditions  

Science Conference Proceedings (OSTI)

Lead-assisted stress corrosion cracking (PbSCC) is a serious concern that can affect all steam generator tubing materials currently employed. A better understanding of lead (Pb) behavior is needed at steam generator and feedwater temperature before possible mitigation techniques can be successfully developed. This report documents results of an experimental program that investigated adsorption of lead onto nickel alloys and magnetite. Fundamental information regarding this adsorption phenomenon is import...

2006-07-12T23:59:59.000Z

348

Application of the TEMPEST computer code for simulating hydrogen distribution in model containment structures. [PWR; BWR  

SciTech Connect

In this study several aspects of simulating hydrogen distribution in geometric configurations relevant to reactor containment structures were investigated using the TEMPEST computer code. Of particular interest was the performance of the TEMPEST turbulence model in a density-stratified environment. Computed results illustrated that the TEMPEST numerical procedures predicted the measured phenomena with good accuracy under a variety of conditions and that the turbulence model used is a viable approach in complex turbulent flow simulation.

Trent, D.S.; Eyler, L.L.

1982-09-01T23:59:59.000Z

349

and Post-Transition Corrosion of Zr Alloys in PWR Coolant  

Science Conference Proceedings (OSTI)

Applicability of Lean Grade of Duplex Stainless Steels in Nuclear Power Plants ... Crack Growth Rates of Irradiated Commercial Stainless Steels in BWR and ...

350

Some hydrogen-control considerations for ice-condenser nuclear plants. [PWR  

DOE Green Energy (OSTI)

The proposal of the Tennessee Valley Authority (TVA) for coping with the problem of accident generated hydrogen gas in its Sequoyah ice-condenser plant was to put in place glow-plug igniters so that any hydrogen that is evolved during an accident could be burnt before accumulating into a dangerously large mass. Since it was desired to install these igniters in the Sequoyah and other plants as quickly as possible, the NRC asked the Lawrence Livermore National Laboratory (LLNL) to carry out some experiments on these igniters to delineate the region of their applicability.

Hubbard, H.W.; Hammond, R.P.; Zivi, S.M.

1981-02-01T23:59:59.000Z

351

Prediction of departure from nucleate boiling in PWR fast power transients  

E-Print Network (OSTI)

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 ...

Lenci, Giancarlo

2013-01-01T23:59:59.000Z

352

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

E-Print Network (OSTI)

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 series of tests to generate data for code assessment against AP600 relevant phenomena. The AP600 design relies only on passive safety features such as gravity driven draining pressurized tanks, and battery power logic and actuators for its safety functions. The inclusion of Core Makeup Tanks and passive heat removal systems into the design increase its dynamic complexity well beyond that of any conventional pressurized water reactor, in which the safeties can be treated as imposed boundary conditions. The bond graph methodology was used to formulate the equations and their topology, as they are used to characterize such a complex system. This characterization was applied to one of the Rig of Safety Assessment (ROSA) program transients, the one-inch cold leg break (AP-CL-03), to construct a mathematical model of the system. The model's constitutive equations were linearized for a selected period of the transient that is of particular importance to the safety analysis. These equations were used for the linear analysis of the system.

Smith, Simon Gregory

1995-01-01T23:59:59.000Z

353

Investigation of Unreinforced Branch Connections on Elbows (PWRMRP-04): PWR Materials Reliability Project (PWRMRP)  

Science Conference Proceedings (OSTI)

Branch connections are installed on elbows because of flow considerations. The qualification of these branch connection/elbow configurations is a concern in the design and qualification of certain piping systems. This report presents the results of an investigation of the stress intensification factors, indices, and flexibility factors for branch connections on elbows. The results of new tests are included.

1999-10-28T23:59:59.000Z

354

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

Science Conference Proceedings (OSTI)

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

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

2007-07-01T23:59:59.000Z

355

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

Science Conference Proceedings (OSTI)

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)

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

356

PWR Primary-Side Gas Management in Advanced Light Water Reactors  

Science Conference Proceedings (OSTI)

The designs for advanced light water reactors (ALWRs) have incorporated new water chemistry controls that have been developed over the past few decades to improve material and equipment reliability and fuel performance and to minimize radionuclide production and transport. It is important to ensure that the new designs operate within ranges that are considered safe based on current knowledge and that industry guidance for normal operation, startup, and shutdown are updated to account for expanding ...

2013-07-17T23:59:59.000Z

357

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

E-Print Network (OSTI)

& Schittkowski (1981) . . . . . . . 94 3.6.2 Example 2: Nuclear Reactor Reload Pattern Design . 95 3.6.3 Example techniques. In particular: · In Section 3.6.2, we present a small but difficult instance of a nuclear reactor Optimization in Continuous and Mixed-Integer Nonlinear Programming Theory, Algorithms, Software

358

Potential for containment leak paths through electrical penetration assemblies under severe accident conditions. [PWR; BWR  

Science Conference Proceedings (OSTI)

The leakage behavior of containments beyond design conditions and knowledge of failure modes is required for evaluation of mitigation strategies for severe accidents, risk studies, emergency preparedness planning, and siting. These studies are directed towards assessing the risk and consequences of severe accidents. An accident sequence analysis conducted on a Boiling Water Reactor (BWR), Mark I (MK I), indicated very high temperatures in the dry-well region, which is the location of the majority of electrical penetration assemblies. Because of the high temperatures, it was postulated in the ORNL study that the sealants would fail and all the electrical penetration assemblies would leak before structural failure would occur. Since other containments had similar electrical penetration assemblies, it was concluded that all containments would experience the same type of failure. The results of this study, however, show that this conclusion does not hold for PWRs because in the worst accident sequence, the long time containment gases stabilize to 350/sup 0/F. BWRs, on the other hand, do experience high dry-well temperatures and have a higher potential for leakage.

Sebrell, W.

1983-07-01T23:59:59.000Z

359

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

E-Print Network (OSTI)

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 ...

Shirvan, Koroush

2010-01-01T23:59:59.000Z

360

Proceedings of the August 2004 EPRI PWR Primary Zinc Addition Workshop  

Science Conference Proceedings (OSTI)

Approximately 60 people attended the workshop, including representatives from 21 utilities, three nuclear system and/or fuel suppliers (Areva, ENUSA, and Westinghouse), EPRI, INPO, ORNL, KAPL, the University of New Brunswick, and several consulting organizations. Personnel from Canada, France, Germany, Spain, Sweden, and the United States were present. Presentations and discussions covered the following areas: • Background Science • Status of Research on Application • Field Experience &bul...

2005-01-17T23:59:59.000Z

Note: This page contains sample records for the topic "niagara mohawk pwr" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

MODELING THE PERFORMANCE OF HIGH BURNUP THORIA AND URANIA PWR FUEL  

E-Print Network (OSTI)

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

Long, Y.

362

Hydrodynamics of adiabatic inverted annular flow: an experimental study. [PWR; BWR  

SciTech Connect

For low-quality film boiling in tubes or rod bundles, the flow pattern may consist of a liquid jet-like core surrounded by a vapor annulus, i.e., inverted annular flow. The stability, shape, and break-up mechanisms of this liquid core must be understood in order to model correctly this regime and to develop appropriate interfacial transfer correlations. This paper reports on a study in which inverted annular flow was simulated in an adiabatic system. Turbulent water jets, issuing downward from long-aspect nozzles were enclosed within cocurrent gas annuli. Jet-core diameter and velocity, and gas-annulus diameter, velocity, and species were varied, yielding liquid Reynolds numbers up to 33,000, void fractions from 0.29 to 0.95, and relative velocities from near zero to over 80 m/s. Jet-core break-up lengths and secondarily, core break-up mechanisms, were observed visually, using strobe lighting.

De Jarlais, G.; Ishii, M.

1983-01-01T23:59:59.000Z

363

Combustion of hydrogen:air mixtures in the VGES cylindrical tank. [PWR; BWR  

DOE Green Energy (OSTI)

Sandia National Laboratories is currently involved in a number of experimental projects to provide data that will help quantify the threat of hydrogen combustion during nuclear plant accidents. Several experimental facilities are part of the Variable Geometry Experimental System (VGES). The purpose of this report is to document the experimental results from the first round of combustion tests performed at one of these facilities: a 5-m/sup 3/ cylindrical tank. The data provided by tests at this facility can be used to guide further testing and for the development and assessment of analytical models to predict hydrogen combustion behavior.

Benedick, W.B.; Cummings, J.C.; Prassinos, P.G.

1984-05-01T23:59:59.000Z

364

Steam Generator Management Program: Empirical Model for Predicting Recirculating PWR Steam Generator Broached-Hole Blockage  

Science Conference Proceedings (OSTI)

Since their initial use in commercial plants in the 1960s, the steam generators (SGs) in pressurized water reactors (PWRs) have exhibited a number of reliability problems. Even though many of these are related to the integrity of the heat-transfer tubing and other internal components or to decreases in heat-transfer efficiency, some SG designs have been subject to a different issue—deposit-induced blockage of the broached flow holes in the tube support plates (TSPs) located within the SG ...

2012-12-12T23:59:59.000Z

365

Process development and fabrication for sphere-pac fuel rods. [PWR; BWR  

Science Conference Proceedings (OSTI)

Uranium fuel rods containing sphere-pac fuel have been fabricated for in-reactor tests and demonstrations. A process for the development, qualification, and fabrication of acceptable sphere-pac fuel rods is described. Special equipment to control fuel contamination with moisture or air and the equipment layout needed for rod fabrication is described and tests for assuring the uniformity of the fuel column are discussed. Fuel retainers required for sphere-pac fuel column stability and instrumentation to measure fuel column smear density are described. Results of sphere-pac fuel rod fabrication campaigns are reviewed and recommended improvements for high throughput production are noted.

Welty, R.K.; Campbell, M.H.

1981-06-01T23:59:59.000Z

366

Modeling and Analysis of Pressurized Water Reactor (PWR) Primary Coolant Zinc Transients  

Science Conference Proceedings (OSTI)

Analysis of plant responses to transients in power production and zinc injection rates has the potential to reveal additional information about how, where, and at what rate zinc is deposited and incorporated into the films on primary system surfaces. Although the process of zinc transport and incorporation is complicated by the numerous mechanisms and surfaces available for incorporation, a control theory type analysis (linear systems analysis) could be useful for the analysis of transients, including in...

2009-09-24T23:59:59.000Z

367

PWR Activity Transport and Source Term Assessment: Surface Activity Concentrations by Gamma Scanning  

Science Conference Proceedings (OSTI)

Measurement of surface isotopic concentrations, known as gamma scanning in the United States, is an invaluable tool in assessing variations in shutdown dose rates and the impacts of chemistry and operational changes.

2011-07-25T23:59:59.000Z

368

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

SciTech Connect

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.

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

369

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

Science Conference Proceedings (OSTI)

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

Matzie, R A [ed.

1980-04-01T23:59:59.000Z

370

Comparison of homogenized and enhanced diffusion solutions of model PWR problems  

SciTech Connect

Model problem comparisons in slab geometry are made between two forms of homogenized diffusion theory and enhanced diffusion theory. The pin-cell discontinuity factors for homogenized diffusion calculations are derived from homogenized variational nodal P1 response matrices and from standard finite differencing. Enhanced diffusion theory consists of applying quasi-reflected interface conditions to reduce variational nodal Pn response matrices to one degree of freedom per interface, without homogenization within the cell. As expected both homogenized diffusion methods preserve reaction rates exactly if the discontinuity factors are derived from the P 11 reference solutions. If no reference lattice solution is available, discontinuity factors may be approximated from single cells with reflected boundary conditions; the computational effort is then comparable to calculating the enhanced diffusion response matrices. In this situation enhanced diffusion theory gives the most accurate results and finite difference discontinuity factors the least accurate. (authors)

Lewis, E. E. [Dept. of Mechanical Engineering, Northwestern Univ., 2145 Sheridan Rd., Evanston, IL 60208 (United States); Smith, M. A. [Nuclear Engineering Div., Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States)

2012-07-01T23:59:59.000Z

371

Design of a low-cost underwater acoustic modem for short- range sensor networks  

E-Print Network (OSTI)

Table 7.3: FPGA Power Consumption Device Q Pwr D Pwr T Pwr (in device families known for their low power consumption (

Benson, Bridget

2010-01-01T23:59:59.000Z

372

The Influence of Terrain on the Severe Weather Distribution across Interior Eastern New York and Western New England  

Science Conference Proceedings (OSTI)

Forecasters have surmised that prominent mountain ranges and river valleys in eastern New York and western New England (e.g., Hudson and Mohawk River valleys; Adirondack, Catskill, Green, and Berkshire Mountains) affect convective initiation and ...

Alicia C. Wasula; Lance F. Bosart; Kenneth D. LaPenta

2002-12-01T23:59:59.000Z

373

American Studies Department Student Papers Collection Summary  

E-Print Network (OSTI)

by drug traffickers. The United States Drug Enforcement Administration notes the Blackfeet Indian on the Blackfeet Indian Reservation; more than 30 convictions for cocaine trafficking resulted.31 In 1999, Mohawk

374

Page not found | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

81 - 21290 of 31,917 results. 81 - 21290 of 31,917 results. Download Evaluation of Final Radiological Conditions at Areas of the Niagara Falls Storage Site Remediated under the Formerly Utilized Sites Remedial Action Program Evaluation of Final Radiological Conditions at Areas of the Niagara Falls Storage Site Remediated under the Formerly Utilized Sites Remedial Action Program (March 2012) http://energy.gov/lm/downloads/evaluation-final-radiological-conditions-areas-niagara-falls-storage Download Evaluation of Final Radiological Conditions at Areas of the Niagara Falls Storage Site Remediated under the Formerly Utilized Sites Remedial Action Program Paper presented at the Waste Management 2012 Conference.February 26 through March 1, 2012, Phoenix, Arizona. http://energy.gov/lm/downloads/evaluation-final-radiological-conditions-areas-niagara-falls-storage-site-remediated

375

Drag-disc turbine transducer data evaluation methods for dynamic steam-water mass flow measurements. [PWR  

SciTech Connect

The mechanical design of a two-phase mass flow rate transducer for a highly corrosive, high temperature (651 K) hot water environment is presented. Performance data for transient steam-water flows are presented. Details of the applications of the device during loss-of-coolant experiments in a pressurized water reactor environment are discussed.

Winsel, C.E.; Fincke, J.R.; Deason, V.A.

1979-01-01T23:59:59.000Z

376

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

SciTech Connect

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.

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

2006-02-28T23:59:59.000Z

377

Some mechanistic observations on the crack growth characteristics of pressure vessel and piping steels in PWR environment  

SciTech Connect

The fatigue crack growth behavior of A533B and A508 pressure vessel steel and AISI Types 304 and 316 steels used in reactor coolant piping have been studied in a pressurized water reactor environment at 288/sup 0/C (550/sup 0/F). The influence of stress ratio (P/sub min//P/sub max/), frequency, ramp times, specimen orientation and material microstructures were included in the study. While none of the materials showed evidence of static crack growth in the environment, the ferritic steels did show an enhanced fatigue crack growth rate at test frequencies of five cycles per minute and lower. Based on fractographic examinations the enhanced growth rate is not the result of environmentally induced intergranular or cleavage modes of crack propagation. Instead, striation spacing measurements were found to agree with the macroscopic crack growth rate, demonstrating a time dependent environmental interaction which introduces a frequency dependent enhancement of the mechanically developed striations. Crack growth experiments using hold times have confirmed the absence of any superimposed contribution of static crack growth components. Fatigue crack growth tests were conducted in an environment of Hydrogen Sulfide gas to establish the contribution of hydrogen embrittlement and will also be described.

Bamford, W.H.; Moon, D.M.

1979-01-01T23:59:59.000Z

378

Direct Experimental Evaluation of the Grain Boundaries Gas Content in PWR fuels: New Insight and Perspective of the ADAGIO Technique  

SciTech Connect

Over the last decades, many analytical experiments (in-pile and out-of-pile) have underlined the active role of the inter-granular gases on the global fuel transient behavior under accidental conditions such as RIA and/or LOCA. In parallel, the improvement of fission gas release modeling in nuclear fuel performance codes needs direct experimental determination/validation regarding the local gas distribution inside the fuel sample. In this context, an experimental program, called 'ADAGIO' (French acronym for Discriminating Analysis of Accumulation of Inter-granular and Occluded Gas), has been initiated through a joint action of CEA, EDF and AREVA NP in order to develop a new device/technique for quantitative and direct measurement of local fission gas distribution within an irradiated fuel pellet. ADAGIO technique is based on the fact that fission gas inventory (intra and inter-granular parts) can be distinguished by controlled fuel oxidation, since grain boundaries oxidize faster than the bulk. The purpose of the current paper is to present both the methodology and the associated results of the ADAGIO program performed at CEA. It has been divided into two main parts: (i) feasibility (UO{sub 2} and MOX fuels), (ii) application on high burn up UO{sub 2} fuel. (authors)

Pontillon, Y.; Noirot, J.; Caillot, L. [Commissariat a l'Energie Atomique, DEN/DEC/SA3C, Centre d'Etudes de Cadarache, BP1, 13108 Saint Paul Les Durance (France); Muller, E. [Commissariat a l'Energie Atomique, DEN/DEC/SESC, Centre d'Etudes de Cadarache, BP1, 13108 Saint Paul Les Durance (France)

2007-07-01T23:59:59.000Z

379

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

SciTech Connect

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

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

1983-01-01T23:59:59.000Z

380

Materials Reliability Program: Pressurized Water Reactor Issue Management Table, PWR-IMT Consequence of Failure (MRP-156)  

Science Conference Proceedings (OSTI)

The Industry Initiative on the Management of Materials Issues provides a proactive, safety-focused approach to the management of materials degradation. In support of this initiative, EPRI formed the Materials Degradation Assessment/Issue Management Table Ad-Hoc Committee and developed an Issue Management Table (IMT) for reactor coolant system components. This report provides initial input to the IMT to address the consequences of failure for the identified components in the reactor coolant system for ope...

2005-12-16T23:59:59.000Z

Note: This page contains sample records for the topic "niagara mohawk pwr" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

Program on Technology Innovation: Mitigation of Flow-Accelerated Corrosion by Titanium Injection in PWR Secondary Systems  

Science Conference Proceedings (OSTI)

This report summarizes the results of a study to determine the potential of titanium-containing compounds for mitigating the flow-accelerated corrosion (FAC) of carbon steel piping in the feedwater train of PWRs. Currently, plants of all types are investigating methods to reduce the severity of FAC in vulnerable systems. Titanium dioxide (TiO2) has been a promising additive under Canadian deuterium/uranium (CANDU) primary conditions, and Ti has been used routinely to mitigate localized corrosion in steam...

2010-03-30T23:59:59.000Z

382

Short-Crack Response of Alloy 182 Weld Metal Undergoing Stress Corrosion Cracking in High-Temperature PWR Primary Water  

Science Conference Proceedings (OSTI)

Mechanistic investigations of environmentally assisted cracking (EAC), to date, have focused more on propagation relative to initiation. At the same time, components spend most of their life in the initiation and "short-crack" growth regimes. Prior exploratory work conducted at General Electric Global Research Center (GE GRC) showed that stainless steels, Alloy 600, and Alloy 182 weld metal exhibit lower average growth rates when the cracks are very short or small (1050 m). In those tests, the transition...

2008-10-31T23:59:59.000Z

383

Proceedings of the 2000 International Conference on Fatigue of Reactor Components (MRP-46): PWR Materials Reliability Program (PWRMR P)  

Science Conference Proceedings (OSTI)

This report contains information presented at the First International Conference on Fatigue of Reactor Components held July 31 - August 2, 2000, in Napa, California. The conference -- sponsored by EPRI, the Organisation for Economic Co-operation and Development Nuclear Energy Agency/Committee on the Safety of Nuclear Installations (OECD NEA/CSNI), and the U.S. Nuclear Regulatory Commission (U.S. NRC) -- provided a forum for the technical discussion of fatigue issues that affect the integrity and operatio...

2001-06-25T23:59:59.000Z

384

Assessment of Experimental Data to Support Computational Fluid Dynamics Analysis of PWR Rod Bundle Heat Transfer Studies  

Science Conference Proceedings (OSTI)

Crud-induced cladding corrosion (CILC) is a localized phenomenon, which is directly related to subcooled nucleate boiling (SNB) in rod bundles of pressurized water reactors (PWRs). Local boiling on fuel rod surfaces leads to preferential deposition of corrosion products circulating in the reactor coolant. Typical thermal hydraulic codes/methods used in core design do not have sufficient resolution to predict susceptible "hot spots" on fuel rod surfaces when SNB is elevated. Hence, detailed local computat...

2010-08-30T23:59:59.000Z

385

Experimental investigation of H/sub 2/ combustion in the Sandia VGES intermediate-scale burn tank. [PWR; BWR  

DOE Green Energy (OSTI)

Sandia National Laboratories is presently involved in several NRC-sponsored experimental projects to provide data that will help quantify the threat of hydrogen combustion during LWR accidents. One project, which employs several experimental facilities: is the Variable Geometry Experimental System (VGES). The purpose of this paper is to present the experimental results from one of these facilities; the intermediate-scale burn tank (approx.5m/sup 3/). The data provided by this facility can be used in the development and assessment of analytical models used to predict hydrogen combustion behavior.

Benedick, W.B.; Cummings, J.C.; Berman, M.; Prassinos, P.G.

1983-01-01T23:59:59.000Z

386

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

SciTech Connect

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.

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

2004-05-18T23:59:59.000Z

387

The Czech National R&D Program of Nuclear Incineration of PWR Spent Fuel in a Transmuter with Liquid Fuel  

E-Print Network (OSTI)

The principle drawbacks of any kind of solid nuclear fuel are listed and briefly analysed in the first part of the paper. On the basis of this analysis, the liquid fuel concept and its benefits are introduced and briefly described in the following parts of the paper allowing to develop new reactor systems for nuclear incineration of spent fuel from conventional reactors and a new clean source of energy. As one of the first realistic attempts to utilise the advantages of liquid fuel, the reactor/blanket system with molten fluoride salts in the role of fuel and coolant simultaneously, as incorporated in the accelerator-driven transmutation technology (ADTT) being proposed in [1], has been proposed for a deeper, both theoretical and experimental studies in [2]. There will be a preliminary design concept of an experimental assembly LA-0 briefly introduced in the paper which is under preparation in the Czech Republic for such a project [3]. 1

M. Hron

1998-01-01T23:59:59.000Z

388

Analysis of Advanced Liquid Waste Minimization Techniques at a PWR: Advanced Media, Pleated Filters, and Ecomomic Evaluation Tools  

Science Conference Proceedings (OSTI)

Utilities may employ a number of options for processing radioactive liquids or improving processing system O&M. This report summarizes low level waste minimization studies for the Diablo Canyon Power Plant. These studies involved the performance of selective ion media, optimization of the chemical volume control system (CVCS) demineralizers, performance assessment of the application of advanced minimum precoat elements for processing condensate demineralizer system rinse water, and evaluation of the econ...

1998-06-30T23:59:59.000Z

389

The Influence of the In-Situ Clad Staining on the Corrosion of Zircaloy in PWR Water Environment  

DOE Green Energy (OSTI)

Zircaloy cladding tubes strain in-situ during service life in the corrosive environment of a Pressurized Water Reactor for a variety of reasons. First, the tube undergoes stress free growth due to the preferential alignment of irradiation induced vacancy loops on basal planes. Positive strains develop in the textured tubes along prism orientations while negative strains develop along basal orientations (Reference (a)). Second, early in life, free standing tubes will often shrink by creep in the diametrical direction under the external pressure of the water environment, but potentially grow later in life in the diametrical direction once the expanding fuel pellet contacts the cladding inner wall (Reference (b)). Finally, the Zircaloy cladding absorbs hydrogen as a by product of the corrosion reaction (Reference (c)). Once above the solubility limit in Zircaloy, the hydride precipitates as zirconium hydride (References (c) through (j)). Both hydrogen in solid solution and precipitated as Zirconium hydride cause a volume expansion of the Zircaloy metal (Reference (k)). Few studies are reported on that have investigated the influence that in-situ clad straining has on corrosion of Zircaloy. If Zircaloy corrosion rates are governed by diffusion of anions through a thin passivating boundary layer at the oxide-to-metal interface (References (l) through (n)), in-situ straining of the cladding could accelerate the corrosion process by prematurely breaking that passivating oxide boundary layer. References (o) through (q) investigated the influence that an applied tensile stress has on the corrosion resistance of Zircaloy. Knights and Perkins, Reference (o), reported that the applied tensile stress increased corrosion rates above a critical stress level in 400 C and 475 C steam, but not at lower temperatures nor in dry oxygen environments. This latter observation suggested that hydrogen either in the oxide or at the oxide-to-metal interface is involved in the observed stress effect. Kim et al. (Reference (p)) and Kim and Kim (Reference (q)) more recently investigated the influence that an applied hoop stress has on the corrosion resistance of Zircaloy tubes in a 400 C steam and in a 350 C concentrated lithia water environment. Both of these studies found the applied tensile hoop stress to have no effect on cladding corrosion rates in the 400 C steam environment but to have accelerated corrosion in the lithiated water environment. In both cases, the corrosion acceleration in the lithiated water environment was attributed to the accumulation of the increased hydrogen picked up in the lithiated environment into the tensile regions of the test specimen. Dense hydride rims have been shown, independent of clad strain, to accelerate the corrosion of Zirconium alloys (References (r) and (s)), suggesting that the primary effect of applied stresses on the corrosion of Zircaloy in the above studies is through the accumulation of hydrogen at the oxide-to-metal interface and not through a direct mechanical breakdown of the passivating boundary layer. To further investigate the potential role of in-situ clad straining (or stress) on Zircaloy corrosion rates, two experimental studies were performed. First, several samples that were irradiated with and without an applied stress were destructively examined for the extent of corrosion occurring in strained and nonstrained regions of the test samples. The extent of corrosion was determined, posttest, by metallographic examination. Second, the corrosion process was monitored in-situ using electrochemical impedance spectroscopy on samples exposed out-of-reactor with and without an applied stress. Post test, these autoclave samples were also metallographically examined.

Kammenzind, B.F., Eklund, K.L. and Bajaj, R.

2001-06-21T23:59:59.000Z

390

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

SciTech Connect

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.

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

391

Preliminary results of thermal igniter experiments in H/sub 2/-air-steam environments. [PWR; BWR  

DOE Green Energy (OSTI)

Thermal igniters (glow plugs), proposed by the Tennessee Valley Authority for intentional ignition of hydrogen in nuclear reactor containment, have been tested for functionability in mixtures of air, hydrogen, and steam. Test environments included 6% to 16% hydrogen concentrations in air, and 8%, 10%, and 12% hydrogen in mixtures with 30% and 40% steam fractions. All were conducted in a 10.6 ft/sup 3/ insulated pressure vessel. For all of these tests the glow plug successfully initiated combustion. Dry air/hydrogen tests exhibited a distinct tendency for complete combustion at hydrogen concentrations between 8% and 9%. Steam suppressed both peak pressures and completeness of combustion. No combustion could be initiated at or above a 50% steam fraction. Circulation of the mixture with a fan increased the completeness of combustion. The glow plug showed no evidence of performance degradation throughout the program.

Lowry, W.

1981-01-01T23:59:59.000Z

392

Materials Reliability Program: Validation of Welding Residual Stress Models for PWR Piping Dissimilar Metal Welds (MRP-271)  

Science Conference Proceedings (OSTI)

The residual stresses imparted by the welding process are a principal factor in primary water stress corrosion cracking (PWSCC) of Dissimilar Metal (DM) piping butt welds in PWRs. Analytical models are frequently used to simulate the welding process in order to predict the residual stress distribution in the weld and base material as an input to crack growth calculations. The crack growth calculations have demonstrated a high sensitivity to the welding residual stress distribution inputs. As part of the ...

2009-12-22T23:59:59.000Z

393

Analysis of the PWR Control Rod Ejection Event with High-Fidelity 3D-DeCART Code  

Science Conference Proceedings (OSTI)

DeCART is a high fidelity three-dimensional (3D) whole core transport calculation methodology for light water reactor (LWR) core simulations. During the past few years, the principal focus of DeCART applications has been on steady-state LWR problems in support of the Electric Power Research Institute’s (EPRI’s) Boiling Water Reactor (BWR) Crud Deposition Analysis program, with the most recent emphasis on depletion analysis of BWR fuel assemblies. Recently, the DeCART methodology has been extended to tran...

2010-04-23T23:59:59.000Z

394

Critical heat-flux experiments under low-flow conditions in a vertical annulus. [PWR; BWR; LMFBR  

Science Conference Proceedings (OSTI)

An experimental study was performed on critical heat flux (CHF) at low flow conditions for low pressure steam-water upward flow in an annulus. The test section was transparent, therefore, visual observations of dryout as well as various instrumentations were made. The data indicated that a premature CHF occurred due to flow regime transition from churn-turbulent to annular flow. It is shown that the critical heat flux observed in the experiment is essentially similar to a flooding-limited burnout and the critical heat flux can be well reproduced by a nondimensional correlation derived from the previously obtained criterion for flow regime transition. The observed CHF values are much smaller than the standard high quality CHF criteria at low flow, corresponding to the annular flow film dryout. This result is very significant, because the coolability of a heater surface at low flow rates can be drastically reduced by the occurrence of this mode of CHF.

Mishima, K.; Ishii, M.

1982-03-01T23:59:59.000Z

395

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

Science Conference Proceedings (OSTI)

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

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

2012-06-06T23:59:59.000Z

396

WASTE DISPOSAL TREATMENT OF PWR HOT LAUNDRY AND DECONTAMINATION ROOM WASTES. Appendix I: SURVEY OF APPLICATION OF STANDARD WATER CLARIFICATION PROCEDURES TO PWR LAUNDRY AND DECONTAMINATION ROOM WASTES. Appendix II: CONFERENCE BETWEEN R. LLOYD AND J.R. POINTE TO ESTABLISH TENTATIVE PROCEDURES AND DETERMINE EQUIPMENT FOR APPLYING ADSORPTION-FLOCCULATION TREATMENT TO PWR LAUNDRY W  

SciTech Connect

This report and three appendixes were issued separately, but are cataloged as a unit. The necessity for treatment of hot laundry and decontamination room wastes prior to disposal at the out, and means for accomplishing this are discussed. A feasible procedure suggested consists of an adsorptionflocculation treatment with supernate disposal by dilution, pass through an evaporator, transfer to surge and decay tanks, with final sludge concentration in drums for retention and burial at sea. (T.R.H.)

Cohen, P.; Lloyd, R.; LaPointe, J.R.; Abrams, C.S.

1956-03-23T23:59:59.000Z

397

Surplus Facilities Management Program (SFMP) Contract No. DE-AC05-810R20722  

Office of Legacy Management (LM)

'^ l '"17 '^ l '"17 ^' ~/t~ >7~ 6~'1 ~DOE/OR/20722-18 Surplus Facilities Management Program (SFMP) Contract No. DE-AC05-810R20722 NIAGARA FALLS STORAGE SITE ENVIRONMENTAL MONITORING REPORT Calendar Year 1983 July 1984 Bechtel National, Inc. Advanced Technology Division DOE/OR/20722-18 NIAGARA FALLS STORAGE SITE ENVIRONMENTAL MONITORING REPORT CALENDAR YEAR 1983 July 1984 Prepared for U.S. DEPARTMENT OF ENERGY OAK RIDGE OPERATIONS OFFICE Under Contract No. DE-AC05-810R20722 By Bechtel National, Inc. Advanced Technology Division Oak Ridge, Tennessee 37830 Bechtel Job No. 14501 *4:F~~~~ ^ABSTRACT 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

398

The Value of Distributed Generation under Different Tariff Structures  

E-Print Network (OSTI)

Laboratory. [ConEd] Consolidated Edison Company of New York,$/a) FLT TOU Consolidated Edison RTP FLT RTP FLT TOU NiagarakW) FLT TOU Consolidated Edison RTP FLT RTP FLT TOU Niagara

Firestone, Ryan; Magnus Maribu, Karl; Marnay, Chris

2006-01-01T23:59:59.000Z

399

U.S. LNG Imports from Indonesia  

Annual Energy Outlook 2012 (EIA)

NY Niagara Falls, NY Waddington, NY Sumas, WA Highgate Springs, VT North Troy, VT LNG Imports into Cameron, LA LNG Imports into Cove Point, MD LNG Imports into Elba Island,...

400

U.S. LNG Imports from Australia  

Annual Energy Outlook 2012 (EIA)

NY Niagara Falls, NY Waddington, NY Sumas, WA Highgate Springs, VT North Troy, VT LNG Imports into Cameron, LA LNG Imports into Cove Point, MD LNG Imports into Elba Island,...

Note: This page contains sample records for the topic "niagara mohawk pwr" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


401

U.S. LNG Imports from Equatorial Guinea  

Gasoline and Diesel Fuel Update (EIA)

NY Niagara Falls, NY Waddington, NY Sumas, WA Highgate Springs, VT North Troy, VT LNG Imports into Cameron, LA LNG Imports into Cove Point, MD LNG Imports into Elba Island,...

402

U.S. LNG Imports from Other Countries  

Annual Energy Outlook 2012 (EIA)

NY Niagara Falls, NY Waddington, NY Sumas, WA Highgate Springs, VT North Troy, VT LNG Imports into Cameron, LA LNG Imports into Cove Point, MD LNG Imports into Elba Island,...

403

U.S. LNG Imports from Trinidad/Tobago  

Gasoline and Diesel Fuel Update (EIA)

NY Niagara Falls, NY Waddington, NY Sumas, WA Highgate Springs, VT North Troy, VT LNG Imports into Cameron, LA LNG Imports into Cove Point, MD LNG Imports into Elba Island,...

404

U.S. LNG Imports from Yemen  

Gasoline and Diesel Fuel Update (EIA)

NY Niagara Falls, NY Waddington, NY Sumas, WA Highgate Springs, VT North Troy, VT LNG Imports into Cameron, LA LNG Imports into Cove Point, MD LNG Imports into Elba Island,...

405

U.S. LNG Imports from Peru  

Gasoline and Diesel Fuel Update (EIA)

NY Niagara Falls, NY Waddington, NY Sumas, WA Highgate Springs, VT North Troy, VT LNG Imports into Cameron, LA LNG Imports into Cove Point, MD LNG Imports into Elba Island,...

406

U.S. Natural Gas Exports to Mexico  

Gasoline and Diesel Fuel Update (EIA)

NY Niagara Falls, NY Waddington, NY Sumas, WA Highgate Springs, VT North Troy, VT LNG Imports into Cameron, LA LNG Imports into Cove Point, MD LNG Imports into Elba Island,...

407

U.S. Total Exports  

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

NY Niagara Falls, NY Waddington, NY Sumas, WA Highgate Springs, VT North Troy, VT LNG Imports into Cameron, LA LNG Imports into Cove Point, MD LNG Imports into Elba Island,...

408

U.S. LNG Imports from Nigeria  

Gasoline and Diesel Fuel Update (EIA)

NY Niagara Falls, NY Waddington, NY Sumas, WA Highgate Springs, VT North Troy, VT LNG Imports into Cameron, LA LNG Imports into Cove Point, MD LNG Imports into Elba Island,...

409

U.S. LNG Imports from Malaysia  

Gasoline and Diesel Fuel Update (EIA)

NY Niagara Falls, NY Waddington, NY Sumas, WA Highgate Springs, VT North Troy, VT LNG Imports into Cameron, LA LNG Imports into Cove Point, MD LNG Imports into Elba Island,...

410

U.S. LNG Imports from Oman  

Annual Energy Outlook 2012 (EIA)

NY Niagara Falls, NY Waddington, NY Sumas, WA Highgate Springs, VT North Troy, VT LNG Imports into Cameron, LA LNG Imports into Cove Point, MD LNG Imports into Elba Island,...

411

U.S. LNG Imports from Egypt  

Annual Energy Outlook 2012 (EIA)

NY Niagara Falls, NY Waddington, NY Sumas, WA Highgate Springs, VT North Troy, VT LNG Imports into Cameron, LA LNG Imports into Cove Point, MD LNG Imports into Elba Island,...

412

U.S. LNG Imports from Norway  

Annual Energy Outlook 2012 (EIA)

NY Niagara Falls, NY Waddington, NY Sumas, WA Highgate Springs, VT North Troy, VT LNG Imports into Cameron, LA LNG Imports into Cove Point, MD LNG Imports into Elba Island,...

413

U.S. LNG Imports from Algeria  

Gasoline and Diesel Fuel Update (EIA)

NY Niagara Falls, NY Waddington, NY Sumas, WA Highgate Springs, VT North Troy, VT LNG Imports into Cameron, LA LNG Imports into Cove Point, MD LNG Imports into Elba Island,...

414

U.S. Natural Gas Exports to Canada  

Annual Energy Outlook 2012 (EIA)

NY Niagara Falls, NY Waddington, NY Sumas, WA Highgate Springs, VT North Troy, VT LNG Imports into Cameron, LA LNG Imports into Cove Point, MD LNG Imports into Elba Island,...

415

U.S. LNG Imports from Brunei  

Annual Energy Outlook 2012 (EIA)

NY Niagara Falls, NY Waddington, NY Sumas, WA Highgate Springs, VT North Troy, VT LNG Imports into Cameron, LA LNG Imports into Cove Point, MD LNG Imports into Elba Island,...

416

U.S. Liquefied Natural Gas Exports To Brazil  

Annual Energy Outlook 2012 (EIA)

Babb, MT Havre, MT Port of Morgan, MT Pittsburg, NH Grand Island, NY Massena, NY Niagara Falls, NY Waddington, NY Sumas, WA Sweetgrass, MT Total to Chile Sabine Pass, LA Total to...

417

The Better Buildings Neighborhood View - April 2013  

NLE Websites -- All DOE Office Websites (Extended Search)

for State and Local Communities May 30-31, 2013 Washington, DC ACEEE Summer Study on Energy Efficiency in Industry July 23-26, 2013 Niagara Falls, NY Innovation Nation Beyond...

418

Materials for Nuclear Power: Digital Resource Center - BOOK: 11th ...  

Science Conference Proceedings (OSTI)

Feb 12, 2007... includes papers included cover degradation phenomena particular to the various reactor systems, BWRs, PWR primary and PWR secondary, ...

419

 

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

energy energy The proposed activity includes implementing the following list of energy efficiency improvements in existing Tribal buildings: * Foam Insulation for the Administration Building * Install new Energy Star oil furnace at the Maintenance Garage * Install new Energy Star oil furnace at the Construction Office * Window replacement at the Family Support Building and Department of Social Services using Energy Star qualified Double Hung Vinyl Windows * Install new Energy Star oil furnace at Terrance House Energy Efficiency and Conservation Block Grants Saint Regis Mohawk Tribe Energy Efficiency Retrofits Saint Regis Mohawk Tribe New York Nov 2, 2009 Jane Summerson Print Form for Records Submit via E-mail Billie Newland Digitally signed by Billie Newland

420

NOAA Technical Memorandum ERL GLERL-22 CHARACTERISTICS OF THE OSWECO RIVER PLUME AND ITS INFLUENCE  

E-Print Network (OSTI)

-Mohawk Steam Station flow and sampling periods during 1972. Configuration of plume and specific conductance of plume and specific conductance at surface, 22 June. Surface temperature distribution, 22 June to the thermocline, 22 June. Mean area1 distribution of plume. Results based on specific conductance and expressed

Note: This page contains sample records for the topic "niagara mohawk pwr" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


421

Building a Culturally Sensitive portal for Indigenous Nations The eHumanity Project, developed by Indiana University in collaboration with the  

E-Print Network (OSTI)

quotation actually came from a 21st-century Blackfeet woman working in a high administrative position, Cheyenne. Still others will mention the Blackfeet, Huron, Cherokee, Mohawk, and even Inuit. These names content in terms of their own personal tribal heritage. A member of the Blackfeet tribe pointed out

422

Guidelines for PWR Steam Generator Tubing Specifications and Repair: Volume 2, Revision 1: Guidelines for Procurement of Alloy 690 S team Generator Tubing  

Science Conference Proceedings (OSTI)

This revised document provides guidelines for procuring Alloy 690 steam generator tubing and sleeve material.

1999-04-14T23:59:59.000Z

423

Materials Reliability Program, Environmental Fatigue Testing of Type 304L Stainless Steel U-Bends in Simulated PWR Primary Water (MR P-188)  

Science Conference Proceedings (OSTI)

Laboratory data generated in the past decade indicate a significant reduction in component fatigue life when reactor water environmental effects are experimentally simulated. However, these laboratory data have not been supported by nuclear power plant component operating experience. In an earlier comprehensive review of laboratory component and structural test data performed through the EPRI Materials Reliability Program (MRP), flow rate was identified as a critical variable that was generally not consi...

2006-02-28T23:59:59.000Z

424

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

E-Print Network (OSTI)

, Nucl. Technology 56:484. Tylee, J.L., 1983, On­line failure detection in nuclear power plant in nuclear power plants, Proceedings of the Topical Meeting on Advances in Human Factors Research on Man networks to the operations of nuclear power plants, Nuclear Safety, 32:68 Upadhyaya, B.R., and Kitamura, M

Pázsit, Imre

425

Seismic Safety Margins Research Programs. Assessment of potential increases in risk due to degradation of steam generator and reactor coolant pump supports. [PWR  

Science Conference Proceedings (OSTI)

During the NRC licensing review for the North Anna Units 1 and 2 pressurized-water reactors (PWRs), questions were raised regarding the potential for low-fracture toughness of steam-generator and reactor-coolant-pump supports. Because other PWRs may face similar problems, this issue was incorporated into the NRC Program for Resolution of Generic Issues. The work described in this report was performed to provide the NRC with a quantitative evaluation of the value/impact implications of the various options of resolving the fracture-toughness question. This report presents an assessment of the probabilistic risk associated with nil-ductility failures of steam-generator and reactor-coolant-pump structural-support systems during seismic events, performed using the Seismic Safety Margins Research Program codes and data bases.

Bohn, M. P.; Wells, J. E.; Shieh, L. C.; Cover, L. E.; Streit, R. L.

1983-08-01T23:59:59.000Z

426

Materials Reliability Program: Environmental Fatigue Testing of Type 304L Stainless Steel U-Bends in Simulated PWR Primary Water (MRP-137)  

Science Conference Proceedings (OSTI)

Laboratory data generated in the past decade indicate a significant reduction in component fatigue life when reactor water environmental effects are experimentally simulated. However, these laboratory data have not been supported by nuclear power plant component operating experience. In a recent comprehensive review of laboratory, component, and structural test data performed through the EPRI Materials Reliability Program, flow rate was identified as a critical variable that was generally not considered ...

2004-12-22T23:59:59.000Z

427

A THEORETICAL INVESTIGATION OF AEROSOL RETENTION WITHIN THE SECONDARY SIDE OF A STEAM GENERATOR UNDER A SGTR SEVERE ACCIDENT SEQUENCE IN A PWR NUCLEAR POWER PLANT.  

E-Print Network (OSTI)

??Las secuencias de accidente con rotura de tubos en el generador de vapor (secuencias SGTR) están consideradas como contribuyentes del riesgo en reactores de agua… (more)

LÓPEZ DEL PRÁ, CLAUDIA

2012-01-01T23:59:59.000Z

428

Comparisons of ASME-code fatigue-evaluation methods for nuclear Class 1 piping with Class 2 or 3 piping. [PWR; BWR  

SciTech Connect

The fatigue evaluation procedure used in the ASME Boiler and Pressure Vessel Code, Sect. III, Nuclear Power Plant Components, for Class 1 piping is different from the procedure used for Class 2 or 3 piping. The basis for each procedure is described, and correlations between the two procedures are presented. Conditions under which either procedure or both may be unconservative are noted. Potential changes in the Class 2 or 3 piping procedure to explicitly cover all loadings are discussed. However, the report is intended to be informative, and while the contents of the report may guide future Code changes, specific recommendations are not given herein.

Rodabaugh, E.C.

1983-06-01T23:59:59.000Z

429

Materials Reliability Program: PWR Internals Age-Related Material Properties, Degradation Mechanisms, Models, and Basis Data - State of Knowledge (MRP-211)  

Science Conference Proceedings (OSTI)

The purpose of this report is to summarize the current state-of-knowledge of neutron irradiation-induced property changes in austenitic stainless steels, principally solution-annealed Type 304 and 304L materials, cold-worked and solution-annealed Type 316 and 316L materials, and Type 308 weld metal. The age-related degradation models were evaluated by an expert panel assembled by EPRI and the Reactor Internals Focus Group (RI-FG). This panel endorsed models to be used in functionality evaluations and sug...

2007-12-19T23:59:59.000Z

430

Materials Reliability Program: Safety Evaluation for Boric Acid Wastage of PWR Reactor Vessel Bottom Heads Due to Bottom-Mounted Noz zle Leakage (MRP-167)  

Science Conference Proceedings (OSTI)

This safety assessment addresses one of the potential safety issues associated with aging degradation of reactor vessel bottom head penetrations: bottom mounted nozzles (BMNs). Specifically, this report evaluates the concern that BMN leakage due to primary water stress corrosion cracking (PWSCC) of the Alloy 600 nozzle and/or Alloy 82/182 J-groove attachment weld could lead to significant wastage of the low-alloy steel head shell material due to concentration of the boric acid present in the leaking prim...

2008-07-02T23:59:59.000Z

431

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

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.

Felde, D. K.

1982-04-01T23:59:59.000Z

432

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

SciTech Connect

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.

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

1982-03-01T23:59:59.000Z

433

Steam Generator Management Program: PWR Steam Generator Tube Wear - Alloy 690/Foreign Objects, Alloy 600/Carbon Steel, Alloy 690/Car bon Steel Support  

Science Conference Proceedings (OSTI)

Wear at tube support plates and wear resulting from foreign objects (FOs) can damage tubes in replacement steam generators. To date, however, limited data have been available on wear rates for Alloy 690 tubing. Under the Steam Generator Management Program, the Electric Power Research Institute (EPRI) has sponsored a series of experiments to determine the wear coefficients between combinations of Alloy 690 steam generator tube material and relevant support and FO materials. This report describes the test ...

2008-12-22T23:59:59.000Z

434

CX-000219: Categorical Exclusion Determination | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

9: Categorical Exclusion Determination 9: Categorical Exclusion Determination CX-000219: Categorical Exclusion Determination United States Army Corps Niagara River, New York Small Hydropower CX(s) Applied: A9, A11 Date: 11/30/2009 Location(s): Niagara River, New York Office(s): Energy Efficiency and Renewable Energy, Golden Field Office The United States Army Corps of Engineering (USACE) would demonstrate the feasibility of low-head hydropower generation along the Niagara River. The work would take place at the USACE offices in Buffalo, New York and Portland, Oregon. The project consists of two steps, Reconnaissance Analysis and Feasibility Study. DOCUMENT(S) AVAILABLE FOR DOWNLOAD CX-000219.pdf More Documents & Publications Dams and Energy Sectors Interdependency Study, September 2011 EIS-0145: Final Environmental Impact Statement

435

PDF Document (3851k)  

Office of Legacy Management (LM)

REFERENCES REFERENCES B 1234 1. Bechtel National, Inc. Engineering Evaluation of Alternatives for the Disposition of Niagara Falls Storage Site, Its Residues and Wastes, DOE/OR/20722-1, prepared for U.S. Department of Energy, Oak Ridge Operations, Oak Ridge, TN January 1984. 2. Aerospace Corporation. Background and Resurvey Recommendations for the Atomic Energy Commission Portion of the Lake Ontario Ordinance Works, ATR-82 (7963-04)-1, prepared for U.S. Department of Energy, washington, DC, 1982. 3. Battelle Columbus Laboratories. A Comprehensive Characterization and Hazard Assessment of the DOE Niagara Falls Storage Site, BMI-2074, Columbus, OH, 1981. 4. Battelle Columbus Laboratories. Preliminary Smelting of Afrimet Residues, Columbus, OH, 1983. 5. Bechtel National, Inc. Geologic Report, Niagara Falls Storage Site, DOE/OR/20722-8, prepared for the U.S. Department of

436

/ J8Y.17 I E(DE86008418)  

Office of Legacy Management (LM)

/,, DOE/EIS-0109F /,, DOE/EIS-0109F / J8Y.17 I E(DE86008418) O&4 .48 FINAL Environmental Impact Statement Long-Term Management of the Existing Radioactive Wastes and Residues at the Niagara Falls Storage Site ( Se(42Ho- /,02.F7fro^' t^ K- A~-) April 1986 U. S. Department of Energy Washington, D.C. 20585 DOE/EIS-0109F (DE86008418) Distribution Category UC-70A FINAL ENVIRONMENTAL IMPACT STATEMENT LONG-TERM MANAGEMENT OF THE EXISTING RADIOACTIVE WASTES AND RESIDUES AT THE NIAGARA FALLS STORAGE SITE April 1986 U.S. Department of Energy Washington, D.C. COVER SHEET FINAL ENVIRONMENTAL IMPACT STATEMENT LONG-TERM MANAGEMENT OF EXISTING RADIOACTIVE WASTES AND RESIDUES AT THE NIAGARA FALLS STORAGE SITE a) Lead Agency: U.S. Department of Energy (DOE) b) Proposed Action: Long-term stabilization and control of existing radioactive wastes and residues

437

# Energy Measuremenfs Group  

Office of Legacy Management (LM)

ri EECE ri EECE # Energy Measuremenfs Group SUMMARY REPORT . AiRIAL R4DIOLOGICAL SURVEY - NIAGARA FALLS AREA NIAGARA FALLS, NEh' YORK DATE OF SURVEY: SEPTEMBER 1979 APPROVED FOR DISTRIBUTION: P Stuart, EC&G, Inc. . . Herbirt F. Hahn, Department of Energy PERFDRflED BY EGtf, INC. UNDER CONTRACT NO. DE-AHO&76NV01163 WITH THE UNITED STATES DEPARTMENT OF ENERGY II'AFID 010 November 30, 1979 - The Aerial Measurements System (A%), operated by EC&t, Inc< for the Un i ted States Department of Energy, was used during November 1976 to conduct an exploratory aerial radiological survey in-the greater Niagara Fails area. The purpose of that survey was to identify locations having concentrations of terrestrial radioactivity not typical of the radiation

438

Microsoft Word - S06246_VP_Report  

Office of Legacy Management (LM)

Formerly Utilized Sites Formerly Utilized Sites Remedial Action Program Niagara Falls Storage Site Vicinity Properties, New York: Review of Radiological Conditions at Six Vicinity Properties and Two Drainage Ditches October 2010 LMS/NFS/S06246 This page intentionally left blank LMS/NFS/S06246 Formerly Utilized Sites Remedial Action Program Niagara Falls Storage Site Vicinity Properties, New York: Review of Radiological Conditions at Six Vicinity Properties and Two Drainage Ditches October 2010 This page intentionally left blank U.S. Department of Energy NFSS Vicinity Property Report October 2010 Doc. No. S06246 Page i Contents Abbreviations .................................................................................................................................. v

439

FORMERLY UTILIZED SITES REMEDIAL ACTION PROGRAM ELIMINFA;RIOflH;EPORT  

Office of Legacy Management (LM)

ELIMINFA;RIOflH;EPORT ELIMINFA;RIOflH;EPORT FORMER ELECTRO METALLURGICAL COMPANY NIAGARA FALLS, NEW YORK U.S. DEPARTMENT OF ENERGY OFFICE OF NUCLEAR ENERGY OFFICE OF REMEDIAL ACTION AND WASTE TECHNOLOGY DIVISION OF FACILITY AND SITE DECOMMISSIONING PROJECTS CONTENTS INTRODUCTION BACKGROUND Site Function Site Description Radiological History and Status ELIMINATION ANALYSIS SUMMARY OF FINDINGS REFERENCES Page 1 : 3 5 7 8 ELIMINATION REPORT FORMER ELECTRO METALLURGICAL COMPANY NIAGARA FALLS, NEW YORK INTRODUCTION From 1942 through 1953, the Electra Metallurgical' Company ("Electromet"), a subsidiary of Union Carbide and Carbon Corporation (now Umetco Minerals Corporation, a subsidiary of Union Carbide Corporation) performed work with radioactive materials under contract to the Manhattan

440

China energy, environment, and climate study: Background issues paper  

E-Print Network (OSTI)

Plant Name Capacity Technology Qinshan I 300 MW Chinese PWRDaya Bay 2x900 MW French PWR Qinshan II 2x600MW Chinese PWR Qinshan III* 2x740 MW Canadian HWR Ling'ao

Sinton, Jonathan E.; Fridley, David G.; Logan, Jeffrey; Guo, Yuan; Wang, Bangcheng; Xu, Qing

2000-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "niagara mohawk pwr" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


441

Pennsylvania Nuclear Profile - Three Mile Island  

U.S. Energy Information Administration (EIA)

snpt3pa8011 805 6,634 94.1 PWR Three Mile Island Unit Type Data for 2010 PWR = Pressurized Light Water Reactor. Note: Totals may not equal sum of ...

442

Missouri Nuclear Profile - Callaway  

U.S. Energy Information Administration (EIA)

snpt3mo6153 1,190 8,996 86.3 PWR Callaway Unit Type Data for 2010 PWR = Pressurized Light Water Reactor. Note: Totals may not equal sum of components ...

443

Kansas Nuclear Profile - Wolf Creek Generating Station  

U.S. Energy Information Administration (EIA)

snpt3ks210 1,160 9,556 94.0 PWR Wolf Creek Generating Station Unit Type Data for 2010 PWR = Pressurized Light Water Reactor. Note: Totals may not ...

444

Optimization of sparse matrix-vector multiplication on emerging multicore platforms  

Science Conference Proceedings (OSTI)

We are witnessing a dramatic change in computer architecture due to the multicore paradigm shift, as every electronic device from cell phones to supercomputers confronts parallelism of unprecedented scale. To fully unleash the potential of these systems, ... Keywords: Autotuning, Cell, HPC, Multicore, Niagara, Performance, Sparse

Samuel Williams; Leonid Oliker; Richard Vuduc; John Shalf; Katherine Yelick; James Demmel

2009-03-01T23:59:59.000Z

445

PLEAEERUSH ANALYTICAL DA-~-A SHEET  

Office of Legacy Management (LM)

for F Alpham Remarks NIAGARA pALI+S* N.Y. U Beta Bldg. 103 - furnace room - -NO, Ra Oil PH Be Th Sample No. Hour Sample Description I I I--- R ) T 1 Q I I I 7392 1100 GA...

446

Evaluation of the SUN UltraSparc T2+ Processor for Computational Science  

Science Conference Proceedings (OSTI)

The Sun UltraSparc T2+ processor was designed for throughput computing and thread level parallelism. In this paper we evaluate its suitability for computational science. A set of benchmarks representing typical building blocks of scientific applications ... Keywords: Computational Science, Evaluation, Niagara2, Sun UltraSparc T2+

Martin Sandrieser; Sabri Pllana; Siegfried Benkner

2009-05-01T23:59:59.000Z

447

Capturing episodes: may the frame be with you  

Science Conference Proceedings (OSTI)

We are interested in detecting episodes in a data stream that are characterized by a period of time over which a condition holds, usually with a minimum duration. For example, we might want to know whenever any router has a packet-drop rate above 0.3% ... Keywords: DSMS, NiagaraST, data streams, episodes, frames, windows

David Maier; Michael Grossniklaus; Sharmadha Moorthy; Kristin Tufte

2012-07-01T23:59:59.000Z

448

The frequency that wouldn't die hydroelectric generators  

Science Conference Proceedings (OSTI)

North America's Niagara River is the site of operating 25 hertz hydroelectric generators that date to the dawn of the electrical age. The reasons why 25 Hz was chosen for such a large block of power and why that obsolete frequency has lived on for the ...

R. D. Barnett

1990-10-01T23:59:59.000Z

449

Categorical Exclusion Determinations: Office of Energy Efficiency and  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

2, 2009 2, 2009 CX-000172: Categorical Exclusion Determination Minnesota City St. Paul CX(s) Applied: A9, A11, B5.1 Date: 11/02/2009 Location(s): St. Paul, Minnesota Office(s): Energy Efficiency and Renewable Energy, Golden Field Office November 2, 2009 CX-000035: Categorical Exclusion Determination Washington's Energy Efficiency Retrofits CX(s) Applied: B5.1 Date: 11/02/2009 Location(s): Washington, Utah Office(s): Energy Efficiency and Renewable Energy November 2, 2009 CX-000034: Categorical Exclusion Determination Saint Regis Mohawk Tribe Energy Efficiency Retrofits CX(s) Applied: B5.1, B2.5 Date: 11/02/2009 Location(s): New York Office(s): Energy Efficiency and Renewable Energy November 2, 2009 CX-000033: Categorical Exclusion Determination Saint Regis Mohawk Tribe Energy Efficiency and Conservation Programs for

450

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

E-Print Network (OSTI)

signals measured at a Finnish nuclear Pressurized Water Reactor (PWR) located in Loviisa. References signals monitored at a Finnish Pressurized Water Reactor (PWR) nuclear power plant. 1 Introduction the reconstruction of a data set of 215 signals measured at a Finnish nuclear Pressurized Water Reactor (PWR) located

451

Two novel procedures for aggregating randomized model ensemble outcomes for robust signal reconstruction in nuclear power plants monitoring systems  

E-Print Network (OSTI)

monitored at a Finnish nuclear Pressurized Water Reactor (PWR) and 920 simulated signals of the Swedish the reconstruction of a data set of 215 signals measured at a Finnish nuclear Pressurized Water Reactor (PWR) located at a nuclear Pressurized Water Reactor (PWR) located in Loviisa, Finland; the second addresses

452

Carnegie Mellon University CARNEGIE INSTITUTE OF TECHNOLOGY  

E-Print Network (OSTI)

DUPIC (PWR to CANDU) PUREX (PWR to PWR) CO2 (ton/GWh) 5.7 12 8.5 3.1E-05 9.4E-05 4.5E-06 33 17 8.8 SO2

453

CMAD IV 11/14/96 Information Security  

E-Print Network (OSTI)

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 Commission (FERC) 889 ­ information on transmission availability and prices. ­ equal access for wholesale

California at Davis, University of

454

Prepared by: Assessment Unit Staff  

E-Print Network (OSTI)

GAMBLE CLALLAM 1 1 ASSINIBOINE 1 1 POTAWATOMIE 1 1 BLACKFEET 1 1 PUEBLO 1 1 BLACKFOOT SIOUX 1 1 PUYALLUP AMERICAN INDIAN 7 3 10 MOHAWK 1 1 BAY MILLS CHIPPEWA 1 1 NAVAJO 3 1 4 BLACKFEET 1 1 NORTHERN CHEYENNE 1 1 - NO DOC 4 2 6 AMERICAN INDIAN 1 1 COLVILLE 1 1 ARAPAHO 1 1 COWLITZ 1 1 BLACKFEET 1 1 HAIDA 1 1 BLACKFOOT

Kaminsky, Werner

455

Revised Petition To Establish A New Generic Subclass  

E-Print Network (OSTI)

Petition to the Federal Trade Commission (the “Commission”) for the establishment of a new generic subclass within the existing polyester category for fibers made from poly(trimethylene terephthalate) (“PTT”) to restate and supplement the contents of the Petition dated February, 21, 2006 in order to address certain questions raised by the Commission. Petitioners note the Commission’s April 18, 2006 action taken in response to the February 21, 2006 Petition assigning the designation “PTT001 ” for PTT fiber for temporary use until a final determination is made by the Commission as to the merits of this Petition. Petitioners propose in order of preference the following names for a new generic subclass of polyester that may be used with respect to PTT fibers: 1. triexta; 2. resisoft; and 3. durares. 1 Mohawk was founded more than 120 years ago and today is the leading producer and distributor of flooring worldwide. Mohawk products serve all major flooring categories: carpet, rugs, hardwood, laminate, ceramic tile, and vinyl flooring. Mohawk has launched a line of carpets manufactured from PTT and sells such carpets under the trademark SmartStrand. 2 Founded in 1802, DuPont is a science company operating in more than 70 countries. DuPont offers a wide range of innovative products and services for markets including agriculture, nutrition, electronics, communications, safety and protection, home and construction, transportation and apparel. DuPont markets PTT under the trademark Sorona®. 3

unknown authors

2006-01-01T23:59:59.000Z

456

operation_tbl1_October_2011M.xlsx  

Gasoline and Diesel Fuel Update (EIA)

2009 Summer 2009 Summer Capacity 2010 Annual Generation Capacity Factor Net MW(e) 1 Net MWh 2 Percent 3 Arkansas Nuclear One 1 AR PWR 842 6,607,090 90 Arkansas Nuclear One 2 AR PWR 993 8,415,588 97 Beaver Valley 1 PA PWR 892 7,119,413 91 Beaver Valley 2 PA PWR 885 7,874,151 102 Braidwood Generation Station 1 IL PWR 1,178 9,196,689 89 Braidwood Generation Station 2 IL PWR 1,152 10,003,246 99 Browns Ferry 1 AL BWR 1,066 8,072,298 86 Browns Ferry 2 AL BWR 1,104 8,842,513 91 Browns Ferry 3 AL BWR 1,105 7,856,326 81 Brunswick 1 NC BWR 938 6,808,445 83 Brunswick 2 NC BWR 920 8,000,043 99 Byron Generating Station 1 IL PWR 1,164 10,337,288 101 Byron Generating Station 2 IL PWR 1,136 9,518,424 96 Callaway 1 MO PWR 1,190 8,996,033 86 Calvert Cliffs Nuclear Power Plant 1 MD PWR 855 6,755,043 90 Calvert Cliffs Nuclear Power Plant 2 MD PWR 850 7,238,905 97 Catawba 1 SC

457

Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants  

Science Conference Proceedings (OSTI)

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.

Woo, H.H.; Lu, S.C.

1981-09-15T23:59:59.000Z

458

DOE - Office of Legacy Management -- Love Canal - NY 24  

Office of Legacy Management (LM)

Love Canal - NY 24 Love Canal - NY 24 FUSRAP Considered Sites Site: LOVE CANAL (NY.24 ) Eliminated from consideration under FUSRAP Designated Name: Not Designated Alternate Name: None NY.24-1 Location: Region running from Old Military Road from the 16-acre rectangular piece of land in the southeast corner of Niagara Falls into the Township of Lewiston , Niagara Falls , New York NY.24-3 Evaluation Year: 1987 NY.24-1 Site Operations: Chemical storage and disposal. NY.24-1 NY.24-3 Site Disposition: Eliminated - No residual radioactive material found NY.24-1 Radioactive Materials Handled: None Indicated Primary Radioactive Materials Handled: None Indicated Radiological Survey(s): Yes NY.24-5 Site Status: Eliminated from consideration under FUSRAP Also see Documents Related to LOVE CANAL

459

Prepared  

Office of Legacy Management (LM)

Prepared Prepared by Oak Ridge Associated Universities Prepared for Division of Remedial Action Projects U.S. Department of Energy COMPREHENSIVE RADIOLOGICAL SURVEY OFF-SITE PROPERTYM NIAGARA FALLS STORAGE SITE LEWISTON, NEW YORK B.P. ROCCO Radiological Site Assessment Program Manpower Education, Research, and Training Division FINAL REPORT May 1983 B.P. Rocco FINAL REPORT Prepared for A.M. pitt T.J. Sowell C.F. Weaver T.S. Yoo Project Staff Prepared by J.D. Berger R.D. Condra R.C. Gosslee J.A. Mattina OFF-SITE PROPERTY M NIAGARA FALLS STORAGE SITE LEWISTON, NEW YORK CO~WREHENSIVE RADIOLOGICAL SURVEY Radiological Site Assessment Program Manpower Education, Research, and Training Division Oak Ridge Associated Universities Oak Ridge, Tennessee 37830 U.S. Department of Energy as part of the Formerly Utilized Sites -- Remedial Action Program May 1983 -- til - This

460

ORO-845 REV.  

Office of Legacy Management (LM)

ORO-845 ORO-845 REV. 1 SUPPLEMENT Surplus Facilities Management Program (SFMP) Contract No. DE-AC05-810R20722 NIAGARA FALLS STORAGE SITE PROJECT MANAGEMENT INFORMATION SUPPLEMENT Prepared by U.S. DEPARTMENT OF ENERGY OAK RIDGE OPERATIONS OFFICE .....- - - - - - - - - - LEGAL NonCE - - - - - - - - - - - - . This report was prepared as an account of work sponsored by the United States Government. Neither the United States nor the United States Department of Energy, nor any of their employees, nor any of their contractors, subcontractors, or their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness or usefulness of any information, apparatus, product or process disclosed, or represents that its use would not infringe privately owned rights. I I I I I " +..'."~ I I * * * * * * * * * * * * ORO-845 Rev. 1 NIAGARA FALLS

Note: This page contains sample records for the topic "niagara mohawk pwr" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
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461

IE  

Office of Legacy Management (LM)

/ / / / / ORAU 89/J-178 IE Prepared by Oak Ridge Associated VERIFICATION Universities OF Prepared for the Division of 1983 AND 1984 REMEDIAL ACTIONS Facility and Site Decommissioning NIAGARA FALLS STORAGE SITE I P rojets VICINITY PROPERTIES U.S. Department of Energy LEWISTON, NEW YORK I~~~| ~S. A. WICAL, M. R. LANDIS, and A. J. BOERNER I I I I I Environmental Survey and Site Assessment Program l3*~~~~~~~ ~~~Energy/Environment Systems Division FINAL REPORT DECEMBER 1989 I VERIFICATION OF 1983 AND 1984 REMEDIAL ACTIONS NIAGARA FALLS STORAGE SITE VICINITY PROPERTIES LEWISTON, NEW YORK 3~~ ~~~~~~1 ~Prepared by S.A. Wical, M.R. Landis, and A.J. Boerner Environmental Survey and Site Assessment Program Energy/Environment Systems Division Oak Ridge Associated Universities Oak Ridge, TN 37831-0117

462

OAK RIDGE NATIONAL LABORATORY RESULTS OF RADIOLOGICAL  

Office of Legacy Management (LM)

2 7% 2 7% d &y / 7 ORNL/TM- 10076 OAK RIDGE NATIONAL LABORATORY RESULTS OF RADIOLOGICAL ~-T-m -~=- -~ w-~- -"" * ,<.~- ~w&$UREMENTs: TAKEN IN THE NIAGARA FALLS, NEW YORK, AREA (NF002) J. K. Williams B. A. Berven ~.~~;:;-~~~ ~. -,' - ~~ 7, OPERATED BY MARTIN MARIDTA ENERGY SYSTEMS, INC, FOR THE UNITED STATES DEPARTMENT OF ENERGY --... ORNL/TM-10076 HEALTH AND SAFETY RESEARCH DIVISION Nuclear and Chemical Waste Programs (Activity No. AH 10 05 00 0; ONLWCOI) RESULTS OF RADIOLOGICAL MEASUREMENTS TAKEN IN THE NIAGARA FALLS, NEW YORK, AREA (NFOO2) J. K. Williams* and B. A. Berven *Biology Division Date Published November 1986 Investigation Team B. A. Berven - RASA Program Manager W. D. Cottrell - FUSRAP Project Director W. H. Shinpaugh - Field Survey Supervisor

463

I I~ I~~" Oak Ridge As.slJj:iatea I Universities Prepared  

Office of Legacy Management (LM)

I~ I~ I~~" Oak Ridge As.slJj:iatea I Universities Prepared for Division of ;' ' 'I Remedial A'ci';on Projects " , : I u.S. Department of Ener'1N I '. I I 'I I I I I [(-24194 COMPREHENSIVE RAOIOLOGYe.AL~\iAVEY , . ~,- ... . -" . OFF-SITE PROPE~tV,i~~ , . NIAGARA FALLS STORA~ESJt~; LEWISTON, NEW YORK A. J. BOERNER Radiological Site Assessment Program Manpower Educatiou t Research, and Training Division FINAL REPORT March 1984 . .. '~-".".'''~'''.;:-- .. ~ -, , I A. J. Boerner FINAL REPORT March 1984 W.O. Helton T.J. Sowell C.F. Weaver B.S. Zacharek Prepared by Project Staff Prepared for OFF-SITE PROPERTY K NIAGARA FALLS STORAGE SITE LEWISTON, NEW YORK J.D. Berger M.J. Brennan R.D. Condra P.W. Frame R.C. Goss1ee COMPREHENSIVE RADIOLOGICAL SURVEY Radiological Site Assessment Program

464

FORMERLY UTILIZED SITES REMEDIAL ACTION PROGRAM ELIMINATION REPORT  

Office of Legacy Management (LM)

FORMERLY UTILIZED SITES REMEDIAL ACTION PROGRAM FORMERLY UTILIZED SITES REMEDIAL ACTION PROGRAM ELIMINATION REPORT FOR OCCIDENTAL CHEMICAL CORPORATION ( FORMER HOOKER ELECTROCHEMICAL COMPANY ) NIAGARA FALLS, NEW YORK SEP 30 1985 Department of Energy Office of Nuclear Energy Office of Remedial Action and Waste Technology Division of Facility and Site Decommissioning Projects ELIMINATION REPORT FOR OCCIDENTAL CHEMICAL CORPORATION (FORMER HOOKER ELECTROCHEMICAL COMPANY) L NIAGARA FALLS, NEW YORK- INTRODUCTION The Department ' of Energy (DDE), Office of Nuclear Energy, Office of Remedial Action and Waste Technology, Division of Facility and Site Decommissioning Projects (and/or the predecessor agencies, offices, and divisions), has reviewed the past activities of the Manhattan Engineer District (MED) and the Atomic Energy Commission (MED/AEC) at

465

Steuben Rural Elec Coop, Inc | Open Energy Information  

Open Energy Info (EERE)

Rural Elec Coop, Inc Rural Elec Coop, Inc Jump to: navigation, search Name Steuben Rural Elec Coop, Inc Place New York Utility Id 18102 Utility Location Yes Ownership C NERC Location NPCC NERC NPCC Yes ISO NY Yes Operates Generating Plant Yes Activity Generation Yes Activity Buying Transmission Yes Activity Distribution Yes References EIA Form EIA-861 Final Data File for 2010 - File1_a[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Utility Rate Schedules Grid-background.png Electric Thermal Storage-TOU Commercial Electric Thermal Storage-TOU(Niagara Hydro Plus Program Participant) Commercial Industrial Industrial Industrial(Niagara Hydro Plus Program Participant) Industrial Residential Seasonal Residential

466

Oak Ridge Associated  

Office of Legacy Management (LM)

])\D ])\D Oak Ridge ,.(\\~ Associated ru~ Universities Post Office Box 11 7 Oak Ridge, Tennessee 37831-0117 October 21, 1986 Manpower Education, Research, and Training Division Mr. Edward G. Delaney, Director Division of Facility and Site Decommissioning Projects Office of Nuclear Energy U.S. Department of Energy Washington, DC 20545 Subject: VERIFICATION OF NIAGARA FALLS STORAGE SITE VICINITY PROPERTIES - 1983/1984 REMEDIAL ACTIONS Dear Mr. Delaney: Oak Ridge Associated Universities Niagara Falls Storage Site, which were remediated during the 1983 and 1984 construction seasons. Based on the results of document reviews, confirmatory sample analyses, and independent site surveys it is ORAU's opinion that the remedial actions have been effective in meeting the DOE radiological

467

OE/EV-0005/2 Formerly Utilized MED/AEC Sites Remedial Action Program  

Office of Legacy Management (LM)

OE/EV-0005/2 OE/EV-0005/2 Formerly Utilized MED/AEC Sites Remedial Action Program Radiological Survey of the Hooker Chemical Company Niagara Falls, New York January 1977 Final Report Prepared for U.S. Department of Energy Division of Environmental Control Technology Washington, D.C. 20545 DOE/EV-0005/2 UC-70 Formerly Utilized MED/AEC Sites Remedial Action Program Radiological Survey of the Hooker Chemical Company Niagara Falls, New York January 1977 Final Report Prepared for U.S. Department of Energy Division of Environmental Control Technology Washing-ton, D.C. 20545 Under Contract No. W-7405-ENE-26 Oak Ridge National Laboratory Oak Ridge, Tennessee 3783C NOTICE This report was prepared as an account of work sponsored by the United States Government. Neither the United States nor the United

468

Microsoft Word - NY.17-16.doc  

Office of Legacy Management (LM)

Printed with soy ink on recycled paper Printed with soy ink on recycled paper Department of Energy Washington, DC 20585 Ms. Judith Leithner Project Manager, Buffalo District U.S. Army Corps of Engineers Department of the Army 1776 Niagara Street Buffalo, New York 14207-3199 Dear Ms. Leithner: This is in reference to the Niagara Falls Storage Site (NFSS) Vicinity Properties E', E, and G located in Lewiston, New York. In accordance with the terms of the March 1999 Memorandum of Understanding (MOU) between the Department of Energy (DOE) and the U.S. Army Corps of Engineers (U.S. ACE), DOE is in the process of completing closure documentation for several sites remediated by DOE prior to assignment of the Formerly Utilized Sites Remedial Action Program (FUSRAP) to the U.S. ACE. Under contract to the U.S. ACE (U.S. ACE Contract

469

Canrom Photovoltaics Inc | Open Energy Information  

Open Energy Info (EERE)

Canrom Photovoltaics Inc Canrom Photovoltaics Inc Jump to: navigation, search Name Canrom Photovoltaics Inc Place Niagara Falls, New York Zip 14305 Sector Solar Product Developer of a thin-film CdTe based solar electric module. References Canrom Photovoltaics Inc[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. Canrom Photovoltaics Inc is a company located in Niagara Falls, New York . References ↑ "Canrom Photovoltaics Inc" Retrieved from "http://en.openei.org/w/index.php?title=Canrom_Photovoltaics_Inc&oldid=343203" Categories: Clean Energy Organizations Companies Organizations Stubs What links here Related changes Special pages Printable version Permanent link Browse properties

470

Natural Gas - U.S. Energy Information Administration (EIA) - U.S. Energy  

Gasoline and Diesel Fuel Update (EIA)

9, 2013 | Release Date: October 10, 9, 2013 | Release Date: October 10, 2013 | Next Release: October 17, 2013 Previous Issues Week: 01/19/2014 (View Archive) JUMP TO: In The News | Overview | Prices/Demand/Supply | Storage In the News: Northeast net imports from Canada plummet, driven by export growth at Niagara Falls Northeast U.S. net natural gas imports from Canada declined by more than half for the first nine months of 2013 compared to the same period in 2012, from 939 million cubic feet per day (MMcf/d) to 438 MMcf/d, according to U.S. Energy Information Administration (EIA) calculations using data from Bentek Energy LLC. Most of this 502 MMcf/d decrease resulted from a 433 MMcf/d decrease in net imports at Kinder Morgan's Tennessee Gas Pipeline (TGP) and TransCanada Pipeline's border crossing between Niagara Falls,

471

DOE - Office of Legacy Management -- Titanium Alloys Manufacturing Co Div  

NLE Websites -- All DOE Office Websites (Extended Search)

Titanium Alloys Manufacturing Co Titanium Alloys Manufacturing Co Div of National Lead of Ohio - NY 41 FUSRAP Considered Sites Site: TITANIUM ALLOYS MANUFACTURING CO., DIV. OF NATIONAL LEAD OF OHIO (NY.41) Eliminated from consideration under FUSRAP Designated Name: Not Designated Alternate Name: Titanium Alloy Metals Titanium Alloy Manufacturing Division Titanium Alloy Manufacturing (TAM) Division of National Lead Company The Titanium Pigment Co. NL Industries ICD/Niagara NY.41-1 NY.41-2 NY.41-3 Location: Niagara Falls , New York NY.41-1 Evaluation Year: 1993 NY.41-4 Site Operations: Produced commercial grade zirconium tetrachloride; conducted research and development relating to solid metallic hydride moderators; and experimental work relative to the conversion of thorium scrap to anhydrous tetrachloride. NY.41-5

472

Page not found | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

81 - 19090 of 26,764 results. 81 - 19090 of 26,764 results. Download CX-000219: Categorical Exclusion Determination United States Army Corps Niagara River, New York Small Hydropower CX(s) Applied: A9, A11 Date: 11/30/2009 Location(s): Niagara River, New York Office(s): Energy Efficiency and Renewable Energy, Golden Field Office http://energy.gov/nepa/downloads/cx-000219-categorical-exclusion-determination Download CX-000098: Categorical Exclusion Determination Churchill County's Heating, Ventilating, and Air Conditioning Retrofit CX(s) Applied: B5.1 Date: 11/24/2009 Location(s): Churchill County, Nevada Office(s): Energy Efficiency and Renewable Energy http://energy.gov/nepa/downloads/cx-000098-categorical-exclusion-determination Download CX-000091: Categorical Exclusion Determination Cheektowaga's Installation of Photovoltaic Solar Panels

473

Page not found | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

01 - 610 of 29,416 results. 01 - 610 of 29,416 results. Download Evaluation of Final Radiological Conditions at Areas of the Niagara Falls Storage Site Remediated under the Formerly Utilized Sites Remedial Action Program Paper presented at the Waste Management 2012 Conference.February 26 through March 1, 2012, Phoenix, Arizona. http://energy.gov/lm/downloads/evaluation-final-radiological-conditions-areas-niagara-falls-storage-site-remediated Download EIS-0463: Amended Notice of Intent To Modify the Scope of the EIS and Conduct Additional Public Scoping Meetings, Notice of Floodplains and Wetlands Involvement Department of Energy - Presidential Permit Application for Northern Pass Transmission, New Hampshire http://energy.gov/nepa/downloads/eis-0463-amended-notice-intent-modify-scope-eis-and-conduct-additional-public-scoping

474

U.S. Spent Nuclear Fuel Data as of December 31,2002 -Table 2  

Gasoline and Diesel Fuel Update (EIA)

6 6 Table 1 | Table 3 Table 2. Nuclear Power Plant Data as of December 31, 2002 Reactor Name State Reactor Type Reactor Vendor a Core Size (number of assemblies) Startup Date (year)b License Expiration (year) Actual Retirement (year) Arkansas Nuclear 1 AR PWR B&W 177 1974 2034 Arkansas Nuclear 2 AR PWR CE 177 1978 2018 Beaver Valley 1 PA PWR WE 157 1976 2016 Beaver Valley 2 PA PWR WE 157 1987 2027 Big Rock Point MI BWR GE 84 1962 2000 1998 Braidwood 1 IL PWR WE 193 1987 2026 Braidwood 2 IL PWR WE 193 1988 2027 Browns Ferry 1 AL BWR GE 764 1973 2013 Browns Ferry 2 AL BWR GE 764 1974 2014 Browns Ferry 3 AL BWR GE 764 1976 2016 Brunswick 1 NC BWR GE 560 1976 2016

475

BWRVIP-196: BWR Vessel and Internals Project, Assessment of Mixing Tee Thermal Fatigue Susceptibility in BWR Plants  

Science Conference Proceedings (OSTI)

In 1998, a French pressurized water reactor (PWR) plant experienced leakage due to thermal fatigue from piping immediately downstream of a residual heat removal (RHR) heat exchanger. EPRI report 1013305, Materials Reliability Program: Assessment of RHR Mixing Tee Thermal Fatigue in PWR Plants (MRP-192), December 2006, was prepared so that owners of PWR plants could determine if their RHR piping systems might be susceptible to similar thermal fatigue cracking and if additional inspection should be recomme...

2008-09-23T23:59:59.000Z

476

Steam Generator Reference Book, Revision 1: Volume 1  

Science Conference Proceedings (OSTI)

The Steam Generator Reference Book documents the state of the art in PWR steam generator technology, providing a comprehensive source for operators, owners, and designers of PWR nuclear power plants. The book summarizes pertinent steam generator operating issues and provides recommendations to improve operational efficiency. Information in the book represents 15 years of research and development activity over the course of several hundred research projects involving PWR steam generator issues.

1994-12-31T23:59:59.000Z

477

Quantification of Ultrasonic Fuel Cleaning Performance  

Science Conference Proceedings (OSTI)

Ultrasonic Fuel Cleaning (UFC) is a process developed by EPRI to remove deposits on PWR fuel assemblies. The process was first used at Callaway in 2001 and up to December 2009 a total of 96 UFC campaigns have been performed at twenty eight PWR plants worldwide.ObjectiveThe project objectives are:Compile and interpret the results of recent PWR fuel cleaning campaigns supplied to EPRI by utilities using a standard template ...

2012-10-30T23:59:59.000Z

478

METHODOLOGIES FOR REVIEW OF THE HEALTH AND SAFETY ASPECTS OF PROPOSED NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL SITES AND FACILITIES. VOLUME 9 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network (OSTI)

Fossil-Fuel-Fired Steam Generators," U.S. Environmentalbasin Boiler or PWR Steam Generator Blowdown Transmissionreactor coolant pumps, steam generators, piping, main stream

Nero, A.V.

2010-01-01T23:59:59.000Z

479

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

E-Print Network (OSTI)

?? 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)

Bertran Morancho, Joan

2011-01-01T23:59:59.000Z

480

Formation and Quantification of Corrosion Deposits in the Power Industry.  

E-Print Network (OSTI)

??The presence of deposits on the secondary side of pressurized water reactor (PWR) steam generator systems is one of the main contributors to the high… (more)

Namduri, Haritha

2007-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "niagara mohawk pwr" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


481

The use of dispersants in pressurised water reactor steam generators.  

E-Print Network (OSTI)

??Environmental degradation promoted by the presence of sludge piles in the steam generators of Pressurised Water Reactors (PWR) can pose a threat to their safe… (more)

Tulloch, Sam

2011-01-01T23:59:59.000Z

482

Uncertainties in the Anti-neutrino Production at Nuclear Reactors  

E-Print Network (OSTI)

directly in the steam generator. Smaller corrections arefeed-water to the steam generator (PWR) or reactor vessel (secondary side of the steam generator or reactor vessel. The

Djurcic, Zelimir

2009-01-01T23:59:59.000Z

483

Strain Rate Sensitivity of Alloy 718 Stress Corrosion Cracking  

Science Conference Proceedings (OSTI)

Stress corrosion cracking (SCC) tests were conducted in 36O'C pressurized- water-reactor. (PWR) primary water using. Alloy 718 heat-treated to produce.

484

Glossary - U.S. Energy Information Administration (EIA)  

U.S. Energy Information Administration (EIA)

W#wind_pwr. There are no record(s) that match your search criteria. Please try again! Thank You. We welcome your comments or suggestions (optional).

485

Oak Ridge National Laboratory - Nuclear Science and Engineering  

NLE Websites -- All DOE Office Websites (Extended Search)

supercomputer research such as this simulation of a Westinghouse PWR900 pressurized water reactor core. Visualization by Tom Evans, ORNL Nuclear Energy Innovation Hub ORNL will...

486

A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network (OSTI)

LOGA LOFT LWR MWe MWt NRC PWR RSR (systems) - EnvironmentalABC been difficult (now NRC) licensing procedure havethe "Acceptance The criteria go (NRC) licensing procedure.

Nero, A.V.

2010-01-01T23:59:59.000Z

487

Louisiana Nuclear Profile - Waterford 3  

U.S. Energy Information Administration (EIA)

1,168 8,949 87.5 PWR Waterford 3 Unit Summer Capacity (MW) Net Generation (Thousand MWh) Summer Capacity Factor (Percent) Type Commercial Operation ...

488

Materials for Nuclear Power: Digital Resource Center - ARTICLE ...  

Science Conference Proceedings (OSTI)

Nov 25, 2007 ... The paper starts with a review of our present capability to predict the materials degradation modes encountered in the current BWR and PWR ...

489

Effects of Thermo-Mechanical Treatments on Deformation Behavior ...  

Science Conference Proceedings (OSTI)

NRC/EPRI Welding Residual Stress Validation Program (Phase III) · On the Microstructure of Alloy 600 SCC Cracks Observed by TEM on PWR SG Pulled Tubes ...

490

Research and Evaluation of Low Temperature Crack Propagation of ...  

Science Conference Proceedings (OSTI)

NRC/EPRI Welding Residual Stress Validation Program (Phase III) · On the Microstructure of Alloy 600 SCC Cracks Observed by TEM on PWR SG Pulled Tubes ...

491

Neutron Dose Rate Effect on Radiation Hardening of Type 316L ...  

Science Conference Proceedings (OSTI)

NRC/EPRI Welding Residual Stress Validation Program (Phase III) · On the Microstructure of Alloy 600 SCC Cracks Observed by TEM on PWR SG Pulled Tubes ...

492

Comparison of the Oxidation Behavior of the 14CrODS Alloy in ...  

Science Conference Proceedings (OSTI)

NRC/EPRI Welding Residual Stress Validation Program (Phase III) · On the Microstructure of Alloy 600 SCC Cracks Observed by TEM on PWR SG Pulled Tubes ...

493

Development of the Extremely Low Probability of Rupture (xLPR)  

Science Conference Proceedings (OSTI)

NRC/EPRI Welding Residual Stress Validation Program (Phase III) · On the Microstructure of Alloy 600 SCC Cracks Observed by TEM on PWR SG Pulled Tubes ...

494

Awardee AwardeeHeadquarters RecoveryFunding TotalValue Denmark  

Open Energy Info (EERE)

Netherlands Germany Greece Malaga Spain Snohomish County PUD No Everett Washington South Kentucky Rural Electric Coop Corp Somerset Kentucky South Mississippi El Pwr Assn...

495

Irradiation-Assisted Stress Corrosion Cracking of Austenitic ...  

Science Conference Proceedings (OSTI)

Protective Insulated Coating for SCC Mitigation in BWRs · PWR Fuel Deposit Analysis at a B&W Plant with a 24-Month Fuel Cycle · PWSCC of Thermocoax ...

496